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Nuclear Development
ACTINIDE AND FISSION PRODUCT
PARTITIONING AND TRANSMUTATION
6 th Information Exchange Meeting
Madrid, Spain,
11-13 December 2000
In co-operation with the
European Commission
EUR 19783 EN
Hosted by CIEMAT and ENRESA
Centro de Investigaciones Energéticas,
Medioambientales y Tecnológicas
Empresa Nacional de
Residuos Radioactivos, SA
NUCLEAR ENERGY AGENCY
ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT
Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on
30th September 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies
designed:
− to achieve the highest sustainable economic growth and employment and a rising standard of living in
Member countries, while maintaining financial stability, and thus to contribute to the development of the
world economy;
− to contribute to sound economic expansion in Member as well as non-member countries in the process of
economic development; and
− to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance with
international obligations.
The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece,
Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United
Kingdom and the United States. The following countries became Members subsequently through accession at the dates
indicated hereafter: Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th
May 1973), Mexico (18th May 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd
November 1996), Korea (12th December 1996) and the Slovak Republic (14 December 2000). The Commission of the
European Communities takes part in the work of the OECD (Article 13 of the OECD Convention).
NUCLEAR ENERGY AGENCY
The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC
European Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its first
non-European full Member. NEA membership today consists of 27 OECD Member countries: Australia, Austria, Belgium,
Canada, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg,
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and the United States. The Commission of the European Communities also takes part in the work of the Agency.
The mission of the NEA is:
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scientific, technological and legal bases required for a safe, environmentally friendly and economical use of
nuclear energy for peaceful purposes, as well as
− to provide authoritative assessments and to forge common understandings on key issues, as input to
government decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy
and sustainable development.
Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive waste
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law
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participating countries.
In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency in
Vienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.
©OECD 2001
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FOREWORD
The objective of the OECD/NEA Information Exchange Programme on Actinide and Fission
Product Partitioning and Transmutation, established in 1989, is to enhance the value of basic research
in this area by facilitating the exchange of information and discussions of programmes, experimental
procedures and results. This Programme was established under the auspices of the NEA Committee
for Technical and Economic Studies on Nuclear Energy Development and the Fuel Cycle and is
jointly co-ordinated by the NEA Nuclear Development Division and the NEA Nuclear Science
Division.
The scope of the Programme includes information on all current and past research related to the
following areas:
1. Physical and chemical properties of elements generated in the nuclear fuel cycle:
a) Chemical properties and behaviour of the actinide species in aqueous and organic solution.
b) Analytical techniques and methods.
c) Physical and chemical properties of various actinide compounds.
d) Collection and evaluation of nuclear and thermodynamic data of relevant elements.
2. Partitioning technology:
a) Partitioning of high-level liquid waste with wet and dry processes.
b) Platinum-group metals-recovery technology.
c) Fabrication technology of the fuel and target materials.
d) Partitioning in the reprocessing process.
3. Transmutation:
a) Transmutation with fast reactors.
b) Transmutation with TRU burner reactors.
c) Transmutation with proton accelerators.
d) Transmutation with electron accelerators.
4. Applications
Other activities related to nuclear data, benchmark exercises and more basic science studies in
relation to this Programme are conducted by the NEA Nuclear Science Division and the NEA Data
Bank.
The Programme is open to all interested NEA Member countries contributing to the information
exchange activities and the Commission of the European Communities. All participants designated a
liaison officer who is a member of the Liaison Group (see Annex 1, CD-ROM).
The Information Exchange Meetings form an integral part of the Programme and are intended to
provide a biennial review of the state of the art of partitioning and transmutation. They are co-organised by
the NEA Secretariat and major laboratories in Member countries.
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An overview of NEA activities on partitioning and transmutation and relevant publications are
available at http://www.nea.fr.html/pt/welcome.html
These proceedings include the papers presented at the 6 th Information Exchange Meeting in Madrid
(Spain) on 11-13 December 2000, held in co-operation with the European Commission. The opinions
expressed are those of the authors only, and do not necessarily reflect the views of any OECD/NEA
Member country or international organisation. These proceedings were co-edited by OECD/NEA and the
European Commission. They are published on the responsibility of the Secretary-General of the OECD.
Acknowledgements
The OECD/NEA gratefully acknowledges CIEMAT and ENRESA for hosting the 6 th Information
Exchange Meeting on Actinide and Fission Product Separation and Transmutation. We also
gratefully acknowledge the European Commission for their support. A special thanks goes to
Ms. Frédérique Joyeux who edited these proceedings within OECD/NEA.
4

TABLE OF CONTENTS
Foreword ........................................................................................................................................ 3
Executive Summary ...................................................................................................................... 7
Scientific Programme.................................................................................................................... 11
Welcome Addresses
CIEMAT................................................................................................................ 17
L. Izquierdo
European Commission .......................................................................................... 19
M. Hugon
OECD Nuclear Energy Agency ............................................................................ 21
Ph. Savelli
Session I: Overview of National and International Programmes .................................... 25
Chairs: J.L. Diaz-Diaz (CIEMAT) – Ph. Savelli (OECD/NEA)
Session II: The Nuclear Fuel Cycle and P&T ...................................................................... 29
Chairs: J. Bresee (DOE) – J.P. Schapira (CNRS)
Closing the Nuclear Fuel Cycle: Issues and Perspectives ..................................... 31
P. Wydler (PSI) and L. Baetslé (SCK•CEN)
Session III: Partitioning........................................................................................................... 51
Chairs: J.P. Glatz (ITU) – J. Laidler (ANL)
Overview of the Hydrometallurgical and Pyrometallurgical Processes
Studied Worldwide for the Partitioning of High Active Nuclear Wastes.............. 53
Ch. Madic (CEA)
Session IV: Basic Physics, Materials and Fuels .................................................................... 65
Chairs: S. Pilate (BN) – H. Takano (JAERI)
Transmutation: A Decade of Revival
Issues, Relevant Experiments and Perspectives..................................................... 67
M. Salvatores (CEA)
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Session V: Transmutation Systems and Safety ................................................................... 93
Chairs: Y. Arai (JAERI) – W. Gudowski (KTH)
Safety Considerations in Design of Fast Spectrum ADS
for Transuranic or Minor Actinide Burning:
A Status Report on Activities of the OECD/NEA Expert Group .......................... 95
D. Wade (ANL)
Poster session: Partitioning ......................................................................................................... 119
Chair: M.J. Hudson (University of Reading)
Poster session: Basic Physics: Nuclear Data and Experiments
and
Materials, Fuels and Targets ............................................................................. 121
Chair: P. D’Hondt (SCK•CEN)
Poster session: Transmutation Systems ...................................................................................... 125
Chair: T.Y. Song (KAERI)
CD-ROM: Contents and Instructions for Use ............................................................................ 127
6

EXECUTIVE SUMMARY
More than 160 participants from 15 countries and three international organisations gathered for
the sixth time since 1989 to exchange information on the various aspects of partitioning and
transmutation (P&T). This 6 th OECD/NEA Information Exchange Meeting was generously hosted by
CIEMAT and ENRESA and was held in co-operation with the European Commission.
Since 1989, the OECD/NEA has conducted an international Information Exchange Programme on
Actinide and Fission Product Partitioning and Transmutation within which the most visible activity has
been the biennial Information Exchange Meeting. Previous meetings were held in Mito City (1990),
Argonne National Laboratory (1992), Cadarache (1994), Mito City (1996) and Mol (1998). This
6 th meeting closed the first ten years of information exchange and opened the discussion on what the next
years should bring. Indeed, while the objectives of these meetings have remained unchanged, the focus
has changed according to the developments and expectations in the subject field.
This 6 th meeting highlighted the developments in P&T according to the themes which were the
subject of 5 sessions, i.e.:
• International collaboration and institutional aspects were addressed in the first session
“Overview of National and International Programmes on P&T”.
• The role of P&T in advanced nuclear fuel cycles and especially the link with waste
management was considered in Session II “The Nuclear Fuel Cycle and P&T”.
• Partitioning was addressed in Session III.
• Session IV addressed basic physics aspects (i.e. nuclear data and experiments), material and fuel
developments, as well as an insight into specific reactor physics aspects of transmutation systems.
• Finally, Session V addressed several concepts of transmutation systems and highlighted
especially the safety considerations.
An overview of the presentations and discussions during these sessions has been given by the
session chairs. This executive summary brings an overview of the discussions held during the
technical sessions and reports on the panel discussion during the closing session.
The discussions during the closing session highlighted the current issues and situation of P&T
research and development in the individual countries and on an international level. A lively session
addressed several aspects:
• The increasing need for interaction between the P&T and the radioactive waste management
community.
• The organisation, planning and role of international collaboration in the development of P&T
and especially of ADS.
• Multi-purpose or single-purpose development of ADS.
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• The need for consensus on figures of merit for P&T, etc.
P&T has made various advances in the past ten years. It has been shown that separation of the
minor actinides is a feasible process, exhibiting high separation factors at the laboratory level; figures of
merit are needed in the future. Indications are that all separation processes, hydro-reprocessing as well as
pyro-reprocessing, are becoming so complex that simplicity and thus cost-reductions should become
prime criteria for future development. A large variety of transmutation concepts have been proposed in
response to which partitioning has been continuously adapted (increasing recovery yields). As further
development in partitioning and in transmutation becomes more expensive, choices on performance and
specific objectives (i.e. criteria and indicators) for P&T will be needed. Future reprocessing processes are
closely linked to fuel-fabrication aspects (fissile content of fuel, type of fuel). Pyro-processes will most
probably be on a batch basis with low throughput because of the limiting transfer capacity between
process steps. Continuous pyro-processes may be envisaged in a few decades time. The specific question
of whether to include curium in a transmutation scheme influences the partitioning processes to be
developed to a pre-industrial stage. The discussions indicated that no common view exists in that respect.
Whereas fuel fabricators indicated a receptiveness to separate the curium and to store it in order to decay
to plutonium for recycling after about 100 years; others indicated that this (concentrated) storage would
involve difficult problems such as criticality, heat removal, doses.
The discussion also covered the question of why P&T might be needed and the objectives
involved. P&T may be justified in order to increase the utilisation of the natural uranium resources
while minimising the possible impact of the long-lived radionuclides on the biosphere. Different
strategies could be envisaged. Multiple recycling has to be considered both in “double strata”
strategies and in standard critical reactors to consistently reduce the final potential radiotoxicity of
waste (by factors up to 200-300). “Once-through” strategies would only allow lesser reductions
(factors 30-40). It was pointed out that, to obtain significant P&T effectiveness in reducing the
potential radiotoxicity of waste in a deep geological repository, process losses in the fuel cycle as low
as 0.1% for all transuranics are required.
Another option which might improve the disposal strategy would be to adopt partitioning and
conditioning (P&C) in which specific conditioning of the minor actinides (MAs) and some of the
long-lived fission products (LLFPs) is applied.
Even before deploying P&T in a fuel cycle, questions arise about the role of plutonium in a truly
sustainable nuclear option. Sustainable development may depend on a plutonium economy in order to
extend the time span over which the natural resources are available to generate electricity. In this
context, one should mention that the thorium option also involves specific problems that will require
substantial developments in the area of fuel cycle.
One may not forget that the time scales for implementation of P&T and of geological disposal are
very different. While P&T could indeed reduce the mass and radioactivity of long-lived waste, for
example by a factor of 100, achieving this reduction would take several decades to equilibrium and a
few centuries to achieve the potential impact on the final waste disposal. This means an institutional
decision to act on P&T early and for several decades in order to reduce potential very long-term
impacts. Currently, the time period from decision to implementation of geological disposal is only
about 40-50 years.
There has been growing interest and activity in basic science supporting P&T. This is particularly
the case in the high power proton accelerator field where a synergy can be envisaged between
transmutation and other applications of intense neutron sources using these accelerators. In the nuclear
physics field, nuclear data measurements have experienced a revival as deficiencies in data are
8

identified. In the fuel area, laboratories for MA-handling are being built in which, over the next
decade, options should be experimentally proven. A difficulty however has been identified, i.e. the
reduced number of fuel irradiation facilities. Other crucial elements in the development of fuel and
materials relate to corrosion and irradiation resistance; these are supported in well-structured R&D
programmes. New technologies emerging from other fields of science and technology could also play
an interesting role in the R&D for P&T (hollow fibres, nano-materials, other uses of Pb-Bi).
The panel discussion indicated that streamlining and prioritisation of R&D is needed in the future
as P&T-related R&D starts to demand more resources and, as mentioned above, would benefit of a
convergence of ideas and consensus on the selection of desirable fuel cycles schemes. Choices will
need to be made in the coming years. Criteria and indicators will therefore need to be identified for
future guidance of work. Basic science developments (materials, nuclear data and simulation,
chemistry, etc.) will be very important and this will demand an international collaborative effort. The
link with waste management was considered as a priority for future work by international
programmes. Finally, simplicity in P&T should be sought.
The closing session ended with the statement that “the P&T community will have to make up its
mind!” in the near future in order to keep the R&D well supported and well focused on the ultimate
objective. This issue will be the focus of the 7 th Information Exchange Meeting which is tentatively
planned to be held at the end of the year 2002 and will be hosted by the Republic of Korea.
9

SCIENTIFIC PROGRAMME 1
Monday 11 December 2000
8:30-9:15 Registration
9:15-9:45 Welcome addresses
Mrs. Lucila Izquierdo, General Secretary of External and Institutional Relations, CIEMAT
Mr. M. Hugon, Co-ordinator, P&T and Future Systems, EC-DGXII
Mr. Ph. Savelli, Deputy Director, Science Computing and Development, (OECD/NEA)
9:45-11:00 Session I: Overview of National and International Programmes
J.L. Diaz-Diaz (CIEMAT) – Ph. Savelli (OECD/NEA)
• Research and Development of Technologies for Partitioning and Transmutation of
Long-lived Nuclides in Japan –Status and Evaluation–, S. Aoki (STA).
• French Research Programme to Reduce the Mass and Toxicity of Long-lived Highly
Radioactive Nuclear Waste, P. Bernard et al. (CEA).
• The Status of the US Accelerator Transmutation of Waste Programme, J. Bresee et al. (DOE, ANL).
11:00-11:20 Coffee break
11:20-12:50 • IAEA Activities in the Area of Emerging Nuclear Energy Systems, A. Stanculescu (IAEA).
• Accelerator Driven Sub-critical Systems for Waste Transmutation: Co-operation and Coordination
in Europe and the Role of the Technical Working Group, M. Salvatores et al. (TWG)
• Partitioning and Transmutation in the EURATOM Fifth Framework Programme,
M. Hugon et al. (EC).
• Activities of OECD/NEA in the Frame of P&T, L. Van den Durpel et al. (OECD/NEA).
12:50-13:00 Introduction of poster sessions
13:00-14:30 Lunch
14:30-17:00 Session II: The Nuclear Fuel Cycle and P&T
J J. Bresee (DOE) - P. Schapira (CNRS)
Overview paper Closing the Nuclear Fuel Cycle: Issues and Perspectives,
P. Wydler (PSI) and L. Baetslé (SCK•CEN)
• Recent Topics in R&D for the OMEGA Project in JAERI, T. Osugi et al. (JAERI).
• Transuranics Transmutation on Fertile and Inert Matrix Lead-bismuth Cooled ADS,
E. González et al. (CIEMAT).
• Actinide and Fission Product Burning in Fast Reactors with a Moderator, I. Krivitski et al. (IPPE).
• Assessment of Nuclear Power Scenarios Allowing for Matrix Behaviour in Radiological
Impact Modelling of Disposal Scenarios, H. Boussier et al. (CEA).
• Disposal of Partitioning-transmutation Wastes with Separate Management of High-heat
Radionuclides, Ch.W. Forsberg (ORNL).
• The AMSTER Concept, D. Lecarpentier et al. (EdF, Ministère de l’Éducation, CEA, CNRS).
17:00-17:20 Coffee break
1 The four scientific Sessions (II-V) included an invited overview paper, which is reproduced in this booklet.
The contributed papers are available on the CD-ROM.
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17:20-19:05 Session III: Partitioning
J.P. Glatz (ITU) - J. Laidler (ANL)
Overview paper Overview of the Hydro-metallurgical and Pyro-metallurgical Processes
Studied Worldwide for the Partitioning of High Active Nuclear Wastes, Ch. Madic (CEA)
Sub-session III-A: Aqueous Reprocessing
• Partitioning-separation of Metal Ions Using Heterocyclic Ligands, M.J. Hudson et al.
(University of Reading, CEA).
• Separation of Minor Actinides from a Genuine MA/LN Fraction, J.P. Glatz et al.
(EC/JRC/ITU).
• Partitioning Anionic Agents Based on 7,8-Dicarba-Nido-Undecaborate for the Remediation of
Nuclear Wastes, F. Teixidor et al. (Institut de Ciència de Materials de Barcelona).
19:05-19:15 End of the session and summary of the first day
Tuesday 12 December 2000
9:00-10:35 Session III: Partitioning (Cont’d)
Sub-session III-B: Dry Reprocessing
• Pyrochemical Processing of Irradiated Transmuter Fuel, J. Laidler et al. (ANL, DOE).
• R&D of Pyrochemical Partitioning in the Czech Republic, J. Uhlir (R ]).
• Demonstration of Pyrometallurgical Processing for Metal Fuel and HLW,
J.P. Glatz et al. (CRIEPI, JRC/ITU).
• Development of Plutonium Recovery Process by Molten Salt Electrorefining with Liquid
Cadmium Cathode, M. Iizuka et al. (CRIEPI, JAERI).
10:35-10:55 Coffee break
10:55-13:05 Session IV: Basic Physics, Materials and Fuels
S. Pilate (BN) – H. Takano (JAERI)
Overview paper Transmutation: A Decade of Revival Issues, Relevant Experiments and Perspectives,
M. Salvatores (CEA)
Sub-session IV-A: Basic Physics
• Nuclear Data Measurements for P&T and Future Plans in JNC, K. Furutaka et al. (JNC).
• New Data and Monte Carlo Simulations on Spallation Reactions Relevant for the Design
of ADS, J. Benlliure (Universidad de Santiago de Compostela).
• The Muse Experiments for Sub-critical Neutronics Validation and Proposal for a
Computer Benchmark on Simulation of Masurca Critical and Sub-critical Experiments,
R. Soule et al. (MUSE collaboration).
• OECD/NEA Benchmark Calculations for Accelerator Driven Systems, M. Cometto et al.
(CEA/PSI, OECD/NEA).
13:05-14:30 Lunch
14:30-15:40 Sub-session IV-B: Materials
• Stainless Steel Corrosion in Lead-bismuth under Temperature Gradient,
D. Gómez Briceño et al. (CIEMAT).
• Accumulation of Activation Products in Pb-Bi, Tantalum, and Tungsten Targets of ADS,
G.V. Kiselev et al. (RF SSC ITEP).
• Thermal and Stress Analysis of HYPER Target System, T.Y. Song et al. (KAERI,
Seoul National University, Gyeongsang National University).
15:40-16:40 Poster session
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16:40-18:20 Sub-session IV-C: Fuels & Targets
• Fuel/Target Concepts for Transmutation of Actinides, D. Haas et al. (EC/JRC/ITU).
• Americium Targets in Fast Reactors, S. Pilate et al. (BN, EdF).
• Research on Nitride Fuel and Pyrochemical Process for MA Transmutation,
Y. Arai et al. (JAERI).
• Transmutation Studies in France: R&D Programme on Fuels and Targets,
M. Boidron et al. (CEA, EdF).
• Fission Product Target Design for HYPER System, W.S. Park et al. (KAERI,
Seoul National University).
18:20-18:30 End of the session and summary of the second day.
20:30 Conference Dinner kindly offered by CIEMAT and ENRESA.
Wednesday 13 December 2000
9:00-11:15 Session V: Transmutation Systems and Safety
Y. Arai (JAERI) – W. Gudowski (KTH)
Overview paper Safety Considerations in Design of Fast Spectrum ADS for Transuranic or
Minor Actinide Burning: A Status Report on Activities of the OECD/NEA Expert Group,
D. Wade (ANL)
• Safety Analysis of Nitride Fuels in Cores Dedicated to Waste Transmutation,
J. Wallenius et al. (KTH).
• Aspects of Severe Accidents in Transmutation Systems, H. Wider et al. (EC/JRC/Ispra).
• A Simple Model to Evaluate the Natural Convection Impact on the Core Transients in
Liquid Metal Cooled ADS, A. D’Angelo et al. (ENEA, Politecnico di Torino).
• Comparative Study for Minor Actinide Transmutation in Various Fast Reactor Core
Concepts, S. Ohki (JNC).
• Study on a Lead-bismuth Cooled Accelerator Driven Transmutation System,
H. Takano et al. (JAERI).
11:15-11:40 Coffee break
11:40-13:30 • Transuranics Elimination in an Optimised Pebble-bed Sub-critical Reactor,
P. León et al. (ETSI, Soreq NRC).
• Transmutation of Nuclear Wastes with Gas-cooled Pebble-bed ADS, A. Abánades et al.
(LAESA, ORNL, Universidad Politécnica de Valencia and Madrid).
• Myrrha, a Multi-purpose ADS for R&D as First Step Towards Waste Transmutation –
Current Status of the Project, H. Aït Abderrahim et al. (SCK•CEN, IBA).
• ADS: Status of the Studies Performed by the European Industry, B. Carluec et al.
(Framatome, Ansaldo).
• Helium-cooled Reactor Technologies for Accelerator-transmutation of Nuclear Waste,
A. Baxter et al. (General Atomics).
13:30-14:45 Lunch
14:45-17:00 Closing session: P&T in the Future?
V. Gonzalez (ENRESA) – M. Salvatores (CEA)
• Session chair summaries.
• Chairman IEM Summary introducing key-questions and discussion.
• Closing remarks by Mr. M. Hugon, EC.
• Closing remarks by Mr. L. Van den Durpel, OECD/NEA.
• Closing address by Dr. A. Colino, President ENRESA.
17:00 End of the 6 th IEM.
13

Poster sessions
Poster session: Partitioning
M.J. Hudson (University of Reading)
• Studies on Behaviour of Selenium and Zirconium in Purex Process, A.G. Espartero et al.
(CIEMAT).
• Solubilization Studies of Rare Earth Oxydes and Oxohalides. Application of
Electrochemical Techniques in Pyrochemical Processes, C. Caravaca et al. (CIEMAT,
Universidad de Valladolid).
• Calix[6]arenes Functionalised with Malondiamides in the Upper Rim as Possible
Extractants for Lanthanide and Actinide Cations, S. Esperanza et al. (Universidad
Autónoma de Madrid).
• Actinide(III)/Lanthanide(III) Partitioning Using N-PR-BTP as Extractant: Extraction
Kinetics and Extraction Test in a Hollow Fiber Module, A. Geist et al. (FZK).
• The Potential of Nano- and Microparticles for the Selective Complexation and Separation
of Metal Ions/Radionuclides, G. Grüttner et al. (Micromod, Universität Potsdam).
• New Extractants for Partitioning of Fission Products, B. Grüner et al. (Institute of
Inorganic Chemistry, Nuclear Research Institute, Katchem).
• Influence of Intermediate Chemical Reprocessing on Fuel Lifetime and Burn-up,
A.S. Gerasimov et al. (RF SSC ITEP).
• Recent Progresses on Partitioning Study in Tsinghua University, C. Song et al.,
(Tsinghua University).
Poster session: Basic Physics: Nuclear Data and Experiments
and
Materials, Fuels and Targets
P. D’Hondt (SCK•CEN)
• Design and Characteristics of the n_TOF Experiment at CERN, D. Cano-Ott (CIEMAT).
• Recent Capture Cross-sections Validation on 232 Th from 0.1 eV to 40 keV and
Self-shielding Effect Evaluation, A. Billebaud et al. (CNRS/IN2P3/UJF).
• Double Differential Cross-section for Protons Emitted in Reactions of 96.5 MeV Neutrons on
Enriched 208 Pb Targets, F.R. Lecolley et al. (CNRS/IN2P3, Subatech, Uppsala University,
ULB, IreS).
• Measurements of Particule Emission Spectra in Proton Induced Reactions of Interest for
the Development of Accelerator Driven Systems, N. Marie et al. (CNRS/IN2P3, Subatech,
IPN, ULB, CEA, IreS).
• Intermediate Energy Neutron-induced Fission Cross-sections for Prospective Neutron
Production Target in ADS, A.N. Smirnov et al. (Khlopin, Uppsala University).
• Nucleon-induced Fission Cross-sections Calculations and Development of Transmutationactivation
Data Library for Transitive Energy Region 20-200 MeV, S. Yavshits et al.
(Khlopin, Institute of Nuclear Power Engineering).
• Neutron Radiative Capture Cross-section of 232 Th in the Energy Range from 0.06 to
2 MeV, D. Karamanis et al. (CEN, ISN, CERN).
• Determination of the Neutron Fission Cross-section for 233 Pa from 0.5 to 10 MeV Using
the Transfer Reaction 232 Th( 3 He, pf) 234 Pa, M. Petit et al. (CEN, ISN).
• Measurement of Double Differential Cross-sections for Light Charged Particles
Production in Neutron Induced Reactions at 62.7 MeV on Lead Target, M. Kerveno et al.
(Subatech, IPN, Institute of Atomic Physics Bucharest, LPCC Caen).
• High and Intermediate Energy Nuclear Data for Accelerator Driven Systems – The HINDAS
Project, J.P. Meulders et al. (UCL, KVI, University of Santiago de Compostela, CEA,
Université de Liège, Subatech, FZ Juelich, NRCG, LPCC/CNRS/ISMRa/Université de
Caen, ZSR-Hannover, Uppsala University, Darmstadt, Braunschweig, ETHZ-Zurich, PSI).
• A Study on Burnable Absorber for a Fast Sub-critical Reactor HYPER, Y.H. Kim et al.
(KAERI, Seoul National University).
14

Poster session: Transmutation Systems
W.S. Park (KAERI)
• MA and LLFP Transmutation in MTRs and ADSs: The Typical SCK•CEN Case of
Transmutations in BR2 and Myrrha. Position with Respect to the Global Needs,
Ch. De Raedt et al. (SCK•CEN).
• Enhancement of Actinide Incineration and Transmutation Rates in ADS EAP-80 Reactor
Core with MOX PuO 2 &UO 2 Fuel, S. Kaltcheva-Kouzminova et al. (Petersburg Nuclear
Physics Institute, ENEA).
• Remarks on Kinetics Parameters of a Sub-critical Reactor for Nuclear Waste
Incineration, J. Blázquez (CIEMAT).
• Noise Method for Monitoring the Sub-criticality in Accelerator Driven Systems,
J.L. Muñoz-Cobo et al. (Universidad Politécnica de Valencia, ORNL, LAESA).
• Molten Salts as Possible Fuel Fluids for TRU Fuelled Systems: ISTC #1606 Approach,
V. Ignatiev et al. (RRC-Kurchatov, VNIITF).
• Comparative Assessment of the Transmutation Efficiency of Plutonium and Minor
Actinides in Fusion/Fission Hybrids and ADS, M. Dahlfors et al. (Uppsala University,
CERN).
• Deep Underground Transmutor (Passive Heat Removal of LWR with Hard Neutron
Energy Spectrum), H. Takahashi (BNL).
• Radiation Characteristics of PWR MOX Spent Fuel After Long-term Storage Before
Transmutation in Accelerator Driven Systems, B.R. Bergelson et al. (RF SSC ITEP).
• Radiation Characteristics of Uranium-Thorium Spent Fuel in Long-term Storage for
Following Transmutation in Accelerator Driven Systems, B.R. Bergelson et al.
(RF SSC ITEP).
• International Co-operation on Creation of a Demonstration Transmutation Accelerator
Driven System, A.S. Gerasimov et al. (RF SSC ITEP).
• On Necessity of Creation of Accelerator Driven System with High Density of Thermal
Neutron Flux for Effective Transmutation of Minor Actinides, A.S. Gerasimov et al.
(RF SSC ITEP).
• New Original Ideas on Accelerator Driven Systems in Russia as Base for Effective
Incineration of Fission Products and Minor Actinides, G.V. Kiselev (RF SSC ITEP).
• Conditions of Plutonium, Americium and Curium Transmutation in Nuclear Facilities,
A.S. Gerasimov et al. (RF SSC ITEP).
• Demonstration Accelerator Driven Complex for Effective Incineration of 99 Tc and 129 I,
A.S. Gerasimov et al. (RF SSC ITEP).
• Critical and Sub-critical GT-MHRs for Waste Disposal and Proliferation-Resistant Fuel
Cycles, A. Ridikas et al. (CEA).
• The Use of Pb-Bi Eutectic as the Coolant of an Accelerator Driven System,
A. Peña et al. (ETSII, JRC/Ispra).
• One Way to Create Proliferation-protection of MOX Fuel, V.B. Glebov et al. (MEPhI,
SEC NRC).
• Transmutation of Long-lived Nuclides in the Fuel Cycle of BREST-Type Reactors,
A.V. Lopatkin et al. (RDIPE).
15

WELCOME ADDRESS
Lucila Izquierdo
General Secretary of External and Institutional Relations
CIEMAT
Avda. Complutense 22, 28040 Madrid, Spain
Ladies and Gentlemen,
Dear Michel Hugon, Phillipe Savelli, dear participants,
It is my pleasure, as Secretary of External and Institutional Relations of CIEMAT and in the name
of its Director General, to welcome all of you to CIEMAT for the 6 th OECD/NEA Information
Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation.
We are glad that Madrid joins the list of cities that have hosted this series of meetings that started
in Mito and have got consecutive success at Argonne, Cadarache and Mol.
We would like to thank the organisers both NEA/OCDE and the European Commission for their
invitation for the joint hosting of the meeting by CIEMAT and ENRESA (the Spanish body for
radioactive waste management).
CIEMAT is the Spanish Public Organism for Research and Technological Development
supported by the Ministry of Science and Technology responsible of finding solutions to improve the
use of resources and energy generation systems, to develop alternative energy sources and to solve the
problems of the Spanish companies regarding energy and its effects on the environment.
CIEMAT is largely involved in the research on future and present nuclear energy sources through
the programmes of Nuclear Fusion and Nuclear Fission, whose activities include many projects related
to the back-end of the nuclear fuel cycle.
At CIEMAT we find in Partitioning and Transmutation (P&T) one very interesting element for
the Nuclear Waste management. Ideally, it will allow to achieve large reduction on the inventories of
long-lived radioactive wastes contained in the nuclear waste, in particular the actinides, reducing the
concerns about our use of nuclear energy for future generations. Not to forget the positive effect on the
public acceptance of nuclear waste management programmes and the potential capability to produce
huge amounts of electricity.
After the participation of CIEMAT in the FEAT and TARC experiments at CERN related to the
Energy Amplifier project, already in 1994, the Nuclear Fission Department has initiated a wide P&T
research programme in 1997.
17

This programme includes six main lines: the advanced hydro-metallurgic reprocessing techniques
to separate all the Transuranium elements and some long-lived fission fragments from the spent fuel of
the present nuclear power plants; the pyro-metallurgic technologies for new ADS fuel recycling; the
corrosion of materials in molten lead alloys; the behaviour of materials in extreme irradiation and
temperature conditions as expected for the spallation target windows in ADS systems; the computer
simulation of transmutation devices and strategies; and the participation on basic experimental
research for transmutation and ADS.
All this work is strongly integrated in the international research on P&T, including the
participation on 6 contracts of the present 5 th Framework Programme of the European Union, and on
the activities on OCDE/NEA, IAEA and the European ADS Technical Working Group. In addition,
we have established bilateral collaboration agreements directly with CEA and with the ITU through
ENRESA. Further contracts with CRIEPI and contacts with several USA laboratories open our
activities in the field outside Europe.
Inside Spain, all this research is performed in close collaboration with ENRESA and several
Universities distributed over the Spanish geography.
I wish that the efforts of all of you present here, the laboratories and institutions from were you
are coming, and the international organisations represented in the room, will make soon the P&T
dream promises a reality. In this sense the opportunities of information exchanges provided by
OECD/NEA and other forums, and the continuation and increase of the support from the national and
international funding agencies, as well as the progressive involvement of industry, are key elements
for the success. I am sure that, with your collaboration, this meeting will represent an important step
forward in this direction.
I hope that beside the intense work schedule of the meeting you can find some time to enjoy
Madrid and the surrounding cities.
I want finally thank all the speakers and poster authors for sharing their work and results with all
of us, and for their collaboration in the organisation of the meeting, and to all of you for coming, and I
hope, for your active participation on the meeting.
Thank you and welcome to CIEMAT.
18

WELCOME ADDRESS
Michel Hugon
Co-ordinator, P&T and Future Systems
European Commission – Research Directorate-General
200 rue de la Loi, 1049 Brussels, Belgium
Ladies and Gentlemen,
It is a great pleasure for me to welcome you today to this 6 th Information Exchange Meeting on
Actinide and Fission Product Partitioning and Transmutation.
I would also address more especially my warm welcome to the many young and enthusiastic
people, who are starting to work in this very exciting field.
It is the second time that the European Commission co-organises this Information Exchange
Meeting with the Nuclear Energy Agency of the OECD. I would like to express here my deep
satisfaction of the excellent relationships that we have established with OECD/NEA in the field of
Partitioning and Transmutation (P&T), where we share a common understanding. This synergy
enables both international organisations to maximise the co-operation in this area between their
different member countries and also to invite representatives of China and Russia to participate to this
meeting.
I would also like to thank both CIEMAT and ENRESA for hosting this meeting and for the hard
preparatory work they have done to make this meeting on its way to a great success.
I am happy that this meeting is taking place in Spain, a country which is very active in the field of
nuclear waste management and disposal. Earlier this year, in March, an International Conference on
the Safety of Radioactive Waste Management was held in Cordoba and organised by IAEA in cooperation
with the EC, the OECD/NEA and the World Health Organisation. From what I heard, it was
a great success.
As you all know, P&T aims at reducing the inventories of long-lived radionuclides in radioactive
waste by separating them from the waste and then transmuting them into radionuclides with a shorter
lifetime. However, there will be always a need for appropriate geological disposal for the existing high
level waste and the waste containing the long-lived radionuclides, which cannot be transmuted.
Nevertheless, the techniques used to implement P&T could alleviate the problems linked to waste
disposal. P&T is still at the research and development stage, which will require long lead-times.
There has been a renewal of interest in P&T worldwide at the end of the eighties (OMEGA
programme in Japan, SPIN programme in France). Meanwhile, sufficient progress has been made in
accelerator technology to consider as feasible the use of accelerator driven systems (ADS) for waste
incineration. Proposals to develop ADS have been made during the nineties by the Los Alamos National
19

Laboratory in the USA with the ATW (Accelerator driven Transmutation of Waste) programme, by
CERN in Europe with the Energy Amplifier (EA) and by JAERI in Japan. In addition, there is a number
of research activities on ADS going on in several EU countries (Belgium, France, Germany, Italy, Spain,
Sweden), Czech Republic, Switzerland, Korea and Russia.
The interest for P&T in the EU is reflected in the increase of funding in this area over the last three
EURATOM Framework Programmes, 4.8, 5.8 and about 26 million for the 3 rd , 4 th and 5 th Framework
Programmes respectively. Research work is also carried out at the Joint Research Centre of the EC,
mainly at the Institute for Transuranium Elements in Karlsruhe.
Ladies and Gentlemen, I would also like to take this opportunity to inform you about some recent
thoughts about the future energy supply in Europe that has been developed by the European
Commission. At the Industry-Energy Council on 5 December, the Vice-President of the Commission
in charge of Energy and Transport, Ms. Loyola de Palacio, presented a Green Paper entitled “Towards
a European Strategy for the Security of Energy Supply” in order to launch a debate.
The starting points are:
• If no measures are taken, in the next 20 to 30 years, about 70% of the Union’s energy
requirements will have to be covered by imported products (today 50%). The energy dependence
of the Union will be increasingly alarming. This will affect all sectors of the economy.
• The fight against the climate change is difficult: inversion of the trends is more difficult than
it appeared to be three years ago. Thus, while the Union stabilised its emissions of
greenhouse gases in 2000, the forecasts of the European Environment Agency consider that
they will increase by 5.2% between now and 2010.
The Green Paper offers for discussion a plan for a long-term energy strategy, in 5 main fields:
• A genuine change in consumer behaviour and energy consumption.
• A truly alternative transport policy.
• Doubling the share of renewable energies from 6 to 12% in the energy balance between now
and 2010 (financial measures).
• Solutions at the Community level (e.g. reinforced strategic oil and gas stocks, a fiscal policy
for energy to steer towards more environmentally friendly sources).
• To analyse the medium-term contribution of nuclear power taking into account the phasing
out decisions of the majority of the Member States and issues related to waste management,
global warming, security of supply and sustainable development.
It is proposed that the European Union must retain its leading position in the field of civil nuclear
technology, in order to retain the necessary expertise and develop more efficient fission reactors and
enable fusion to become a reality.
Research on the technologies of waste management and their practical implementation under
optimum safety conditions has actively to be continued. This applies to geological disposal as well as
to partitioning and transmutation.
Ladies and Gentlemen, our work these coming days are thus of great interest. I wish you all a
very fruitful and successful meeting as well as a nice stay in Madrid.
Thank you for your attention.
20

WELCOME ADDRESS
Philippe Savelli
Deputy Director for Science, Computing and Development
OECD Nuclear Energy Agency
Le Seine St-Germain, 12, Boulevard des Iles, 92130 Issy-les-Moulineaux, France
Ladies and Gentlemen,
It is a real pleasure for me to welcome you to this 6 th Information Exchange Meeting organised by
the OECD Nuclear Energy Agency (NEA).
Back in 1988, the Japanese government asked the NEA to launch an international information
exchange programme on partitioning and transmutation (P&T). At that time, only a few countries were
really active in this field. In the 1960s and 1970s, preliminary studies and experiments had been
conducted in the USA, Japan and within several European countries, as well as the European
Commission. The conclusions of some of the assessments published in the late 1970s and early 1980s
clearly stated that the transmutation of minor actinides was considered theoretically possible, but that
it was not obvious whether the potential long-term risk reduction for the waste disposal site was
overall beneficial, because of the increase in short-term risks for the workers. Those studies also
concluded that there were no obvious direct cost or safety incentives for P&T of actinides for waste
management purposes. However, it was recognised that further investigation of advanced reprocessing
techniques for conditioning of plutonium and minor actinides would be valuable.
A second phase of interest in P&T emerged at the end of the 1980s, partly based on the growing
awareness of the difficulties in licensing large nuclear waste repositories and certain delays in the
related R&D projects. There was a need to re-examine the validity of the P&T option in the light of
the more recent results. This led Japan, France, USA and other countries to start new studies,
complemented by an experimental R&D programme.
In the early 1990s, new assessment reports were published by France and the USA, as well as
studies conducted under the auspices of IAEA or EC. The Nuclear Energy Agency undertook a
systems study in early 1996 and published the Status and Assessment Report on Minor Actinide and
Fission Product Partitioning and Transmutation in April 1999. It is worth emphasising four of the
main conclusions of this report:
• P&T will not replace the need for appropriate geological disposal of high level waste.
• The recycling of plutonium and minor actinides could stabilise the transuranium nuclide
inventory. However, multiple recycling of transuranium nuclides is a long-term venture for
which it may take decades to reach equilibrium.
21

• Partitioning methods for long-lived radiotoxic elements have been developed on a laboratory
scale and could be very useful to condition separated long-lived nuclides in appropriate
matrices or in irradiation targets. These matrices could be selected to be less soluble than
glass in geological media.
• Last but not least, fundamental R&D for the implementation of P&T needs long lead-times
and would require large investments in dedicated fast neutron spectrum devices, extension of
reprocessing plants and the construction of remotely manipulated fuel and target fabrication
plants.
During the 1990s, we also noticed a renewed interest in accelerator driven systems (ADS). Today,
as the participation in this meeting shows, several countries active in P&T emphasise the ADS-line. We
have seen increasing international activity, especially in Europe; a growing number of bilateral and
multilateral co-operations have been established. Examples of these are the collaboration of Japanese
institutes with European Joint Research Centres, the 5 th Framework R&D projects sponsored by the
European Commission, the Technical Working Group in Europe under the chairmanship of Carlo Rubia,
the ISTC activities with our Russian colleagues and the foreseen increased collaboration between USA
and France.
It was in response to this emerging interest that the NEA launched new studies under the auspices
of its Nuclear Development and Nuclear Science Committees. Both committees have, together with
the NEA Data Bank, developed several well co-ordinated activities, covering a diverse set of issues
related to P&T, such as nuclear data and benchmarks, partitioning techniques and also more strategic
systems studies. Today, more exchange with the NEA Radioactive Waste Management Committee is
sought and we view Session II (The Nuclear Fuel Cycle and P&T) of this meeting as a welcome step
in this direction. A new Working Party on Scientific Issues in P&T has been launched and, in fact,
held its first meeting yesterday here in Madrid. Other Working Parties and Expert Groups, as well as
specific Workshops and Information Exchange Meetings will remain part of our work programme and
they will be tailored in response to your demands. In addition, our P&T activities are now organised as
a horizontal project and in that respect, a single NEA web page on P&T will announce all our projects
and programmes in the future. A separate presentation in Session I this morning will cover our
activities more in detail.
Ladies and Gentlemen, in the light of these past and ongoing developments, I consider it
appropriate to raise two items that I regard as important for future activities on P&T.
The first is the increasing importance that nuclear power could play in response to the need for a more
sustainable energy development. During the debate on climate change in den Hague two weeks ago, some
delegates indicated that nuclear should be recognised as part of a future energy mix. It is for example
encouraging to note that the European energy and transport commissioner, Ms. Loyola de Palacio,
recognises this role of nuclear, despite some EC countries having embarked on a nuclear phase-out
strategy.
If the concern for our future would be translated into a continued demand for nuclear energy,
several of the developments discussed in this meeting could help reply to some of the questions
regarding nuclear energy. The public would only accept an increased use of nuclear, if today’s
concerns about safety, waste and proliferation could be satisfied. P&T is one approach that could
contribute to the sustainability of nuclear energy.
A second item relates to the assessment of P&T and especially the question of objectives and
indicators to be applied. Society demands more clear objectives and indicators before embarking on
22

developments. This will surely become the case if society accepts increased reliance on nuclear
energy. Society claims an economically viable energy resource, showing an excellent safety level,
dealing in an efficient way with waste and other residuals and finally respecting the environment in the
short and long term. Therefore, nuclear, and clearly also P&T, will have to face this kind of
evaluation. We consider that an honest reflection on applicable objectives and criteria would be a
worthwhile undertaking in the future.
In this context, we are all aware of the declining trend in nuclear education and availability of
infrastructure. In today’s context of deregulation and increasing competitive pressure on the utilities
and on research institutes, P&T will have to face this additional challenge of limited resources and
infrastructure. However, a positive factor is that the different P&T projects have presented new and
challenging scientific problems that are attracting young scientists to enter the nuclear field.
This Information Exchange Meeting is again in co-operation with the European Commission and
I wish to thank them for their valuable support. I believe that the co-operation we have established is a
good example of how scarce resources can be shared, based on mutual understanding.
In ending my talk, I would first of all like to thank CIEMAT and its Director-General
Dr. Felix Yndurain Muñoz, as well as ENRESA and its President Dr. Antonio Colino, who are jointly
hosting this meeting and have ensured the success of, what I am convinced, will be a very enjoyable
stay here in Madrid.
Ladies and Gentlemen, may I wish you a fruitful meeting. The numerous participation gives me
confidence that the scientific programme has captured your interest and that these Information
Exchange Meetings respond to your wishes. I am glad that we can also welcome participation from
non-OECD countries. I invite all of you to help us shape our activities in the future and your advice or
comments will certainly be taken into account and be reflected in our future programme of work.
Thank you for your attention.
23

SESSION I
Overview of National and International Programmes
Chairs: J.L. Diaz-Diaz (CIEMAT) – Ph. Savelli (OECD/NEA)
_____________________
SUMMARY
This session gave an overview of the major activities in the field of partitioning and transmutation
(P&T) in some of the Member countries, i.e. Japan, France and the USA as well as by international
groups (TWG) and organisations (EC, IAEA, NEA).
Significant progress has been made in many countries since the previous Information Exchange
Meeting in Mol, November 1998. The OMEGA programme in Japan underwent a review by the
AEC’s Advisory Committee on Nuclear Fuel Cycle Back-End Policy that released a report entitled
Research and Development of Technologies for Partitioning and Transmutation of Long-lived
Nuclide: Status and Evaluation Report (March 2000). This report gave an overview of ongoing and
planned R&D by the different governmental and private organisations in Japan and identified future
activities. The report concluded that the R&D at the three research institutes (JAERI, JNC and
CRIEPI) had resulted in the establishment of processes for P&T technology with the expected
performance. The aims of the Phase I R&D have thus been achieved according the expectations
included in the OMEGA-programme. R&D in Phase II has experienced some delays, the primary
reason being that Japan is redefining its entire FBR programme, and facilities to handle MAs and other
materials have yet to be constructed. The report also mentioned that in carrying out further R&D, it
would be important to promote co-operation with domestic and foreign organisations in order that
experimental facilities – including those for engineering experiments – can be used efficiently.
The three Japanese research organisations indicated in the report that a common issues is the
implementation of experiments to demonstrate processes using actual HLLW. In addition, the
preparation of a database on fuel irradiation behaviour for performance analysis and the development
of fuel fabrication technology were considered an issue as well.
While several R&D activities are planned in the future, e.g. economic aspects, P&T technology as
part of the fuel cycle, system design and others, it was considered appropriate to conduct R&D in
these areas on a time schedule compatible with nuclear fuel cycle R&D. At present, feasibility studies
on commercialised FBRs and related fuel cycles system is being carried out under the collaborative
efforts of JNC, electric utilities, CRIEPI and JAERI. In this study, R&D scenarios toward
commercialisation of fast reactor system will be reviewed by about the year 2005. Thus, around the
year 2005 is deemed to be an appropriate time to reconsider all R&D scenarios of P&T including the
25

use of FBRs for transmutation together with power generation, and the double-strata fuel cycle.
Thereafter, progress, results and R&D policy will be checked and reviewed every five years or so.
Evaluations of P&T technology system concepts, and reviews of introduction scenarios, should also be
conducted.
In France, research is conducted with the goal of establishing by 2006 the scientific feasibility of
transmutation in various types of nuclear reactors (PWR, innovative reactors) and the technical
feasibility of intensive separation downstream from reprocessing at La Hague, as well as of the
specific conditioning of separated long-lived radionuclides. The research is conducted in co-operation
with partners in the nuclear industry, EdF, COGEMA and FRAMATOME, as well as with CNRS and
universities. It benefits form significant co-operation at the European and international level. It is
constantly evaluated by the National Evaluation Commission, which draws up and publishes an
evaluation report annually. The procedure involves identifying a set of complementary scientific and
technical solutions, which serve to define open and flexible strategies for the back-end of the cycle and
lay the groundwork for a decision in 2006.
Concerning partitioning research, a reference programme has been defined for an advanced
separation process for the main long-lived radionuclides in waste. The families of extractors were
defined, the principal reference molecule synthesised, and their performances verified experimentally
on real radioactive solutions in the ATALANTE facility in order to reach the stage of scientific
feasibility in 2001. The next stage will be that of technical feasibility, moving from the molecule to the
overall chemical process, which will be defined and validated in 2005. Experimental studies on fuel
for the transmutation in fast neutron reactors have been launched, in particular in the PHENIX reactor,
whose irradiation programme has focused on this research since 1998 and which therefore has been
the object of inspection, renovation and maintenance, in view of a power increase in 2001.
In addition, teams at CEA and CNRS, in co-operation with industrial partners, have provided the
technical data for a request for an experimental demonstration model of a hybrid reactor for
transmutation, in a European and international framework.
Since the last meeting on P&T in 1998 in Mol, the US accelerator transmutation of waste (ATW)
programme has changed significantly. Two years ago, the only effort was the preparation of a research
plan for developing ATW technology. Today, a significant research effort is underway, and the US is
seeking opportunities to collaborate with other national programmes. Transmutation R&D in the US
initially has been focused on ADS and has involved a series of trade-off studies. In all cases, it has
been assumed that uranium remaining in civilian spent fuel elements would be recovered, probably by
a modified PUREX process called UREX. Initial studies of the UREX process have shown that the
uranium product will meet US Class C requirements and could be disposed of as low level waste or be
stored for possible future use in a nuclear fuel cycle. Various combinations of proton accelerator
designs, spallation neutron sources, and transmutation target have been evaluated for technology
readiness, and assumed irradiated targets have been studied for the effectiveness of chemical
processing to recycle untransmuted long-lived isotopes. These evaluations have resulted in a base-line
design which includes a linear proton accelerator, a lead-bismuth spallation target, and sodium-cooled
metallic or ceramic dispersion transmutation target/blanket non-fertile fuel elements. Another
interesting transmutation system design currently being evaluated consists of a “dual strata” approach
which would involve a thermal critical reactor within which plutonium and minor actinides would
fission and 99 Tc/ 129 I would be subjected to a thermal neutron flux.
The ATW programme during the Fiscal Year 2001 involved approximately a doubling of the
Fiscal Year 2000 funding. This will allow an expansion of experimental programmes, and DOE’s
Office of Nuclear Energy, Science and Technology (NE) is actively seeking opportunities for
26

collaborative research with foreign ADS programmes. Meanwhile, the programme is being
reorganised to combine the objectives of the DOE Defence Programme’s Accelerator Production of
Tritium (APT) programme with those of NE’s ATW efforts. The combined programme is known as
Advanced Accelerator Application (AAA), and it will be administered by NE. Congress has requested
a report by March 1, 2001 on how the new activity will be carried out.
In Europe, a Technical Working Group (TWG) was established with the task of identifying the
critical technical issues in which R&D is needed to develop an European demonstration programme of
ADS over a 10-year time scale. The TWG, currently extended to ETWG, started in 2000 an intensive
work aimed at defining a European Roadmap towards an experimental ADS, called XADS. The
roadmap document is expected to be issued in the first half of 2001. The first goal of the roadmap is to
propose a technological route to reduce the risks associated with nuclear waste, based on the
transmutation of this waste using an ADS. The second and main goal of the roadmap is to prepare a
detailed technical programme, with cost estimates, which could lead to the demonstration of an
experimental ADS in 10 years. The programme as described in the roadmap will lead to a
rationalisation of human resources and experimental facilities, a training ground for young researchers,
the development of innovative fuels and reprocessing technology, spin-offs in the fields of
accelerators, spallation sources, liquid metal technology, radioisotope production and actinide physics
and chemistry. Hence, a final goal of the roadmap is to identify possible synergies and rationalisation
that this programme could have within the nuclear community, indicate potential spin-offs, show how
competence can be maintained in a currently stagnating field.
The European Commission has included in its previous Framework Programmes and in the
current ongoing Fifth one, FP5 (1998-2002), several activities related to P&T. These activities address
the chemical separation of long-lived radionuclides and the acquisitions of technological and basic
data, necessary for the development of an ADS. Collaboration is also being implemented in this field
between scientists of the European Union (EU) and the Commonwealth of Independent States (CIS).
The interest for P&T in the EU is reflected in the increase of funding over the EURATOM
Framework Programme, i.e. 4.8, 5.8 and about 26 million for the Third, Fourth and Fifth Framework
Programmes respectively.
The P&T projects in FP5 have been grouped in three clusters. The experimental investigation of
efficient hydro-metallurgical and pyrochemical processes for the chemical separation of long-lived
radionuclides from HLW is carried out in the cluster on partitioning. The work on transmutation is
mainly related to the acquisition of data, both technological and basic, necessary for the development of
an ADS. The cluster on transmutation-technological support deals with the investigation of radiation
damage induced by spallation reactions in materials, of the corrosion of structural materials by lead alloy
and of fuels and targets for actinide incineration. In the cluster transmutation-basic studies, basic nuclear
data for transmutation and ADS engineering design are collected and sub-critical neutronics are
investigated. Additional projects will be funded in 2001, such as preliminary engineering design studies
for an ADS demonstrator, complementary projects for technological support and networking.
The Commissioner responsible for research in the EC launched the idea of an “European
Research Area” in January 2000. The intention is to contribute to the creation of better overall working
conditions for research in Europe. In view of the future research programme, the EURATOM
Scientific and Technical Committee has prepared a report on the strategic issues to be considered in
the development of the appropriate nuclear energy research strategies in a 20-50 year perspective. In
the area of nuclear fission, continued support should be given to maintain and develop the competence
needed to ensure the safety of existing and future reactors. In addition, support should be given to
explore the potential for improving present fission technology form a sustainable development point of
27

view, i.e. better use of uranium and other nuclear fuels, whilst reducing the amount of long-lived
radioactive waste produced.
The session ended with two presentations by the international nuclear agencies, IAEA and NEA,
showing their programmes of work in the area of P&T. A more detailed overview of their activities is
given in the respective papers. The need for further international co-operation was once again repeated
and strengthening of existing co-operations between countries as well as in the framework of
international organisations has been requested in order to secure the effective use of scarce resources.
This need for international co-operation will also be needed as prioritisation of R&D will be needed in
the nearby future while decisions for new infrastructure will be requested.
The discussion during this session already highlighted the need for convergence of R&D in the
future. A selection of fuel cycle schemes and an associated precise work-programme or strategy was
requested where new international studies should focus on fuel cycle impacts and R&D needs.
One may remark that quite significant resources are spent in R&D in the various areas of P&T.
While some countries perform in essence theoretical studies on possible P&T scenarios, others are
embarked in experimental programmes and commit resources for construction of specific facilities. In
the meantime, these countries also recognise that P&T is a long-term endeavour and that no immediate
decisions are needed before about 2005. In France and Japan, a review of the P&T programme in the
light of a long-term back-end policy is foreseen by the middle of this decade, while real
implementation of a P&T scheme would still need an additional 20 years.
28

SESSION II
The Nuclear Fuel Cycle and P&T
Chairs: J. Bresee (DOE) – J.P. Schapira (CNRS)
_____________________
SUMMARY
The introduction by L.H. Baetslé and P. Wydler to this session gave a general overview of new
partitioning and transmutation (P&T) fuel cycles based on present technologies (LWR, FR, uraniumplutonium
fuels) altogether with their main issues. In order to minimise back end risks, plutonium
management has to be addressed first. On the other hand, any P&T fuel cycle assessment has to take
into account all the wastes and nuclear materials involved as well as their global impact (radiological,
heat, secondary solid wastes, liquid and gaseous effluents). P&T strategy based on multirecycle in FRs
leads to a waste radiotoxicity reduction factor of about 3 to 10 (depending on the date) if only
plutonium is recycled and of about 100 if all the TRU are recycled within different scenarios (double
strata, double components). These figures do not take into account the fuel cycle inventory
radiotoxicity. They are strong constraints related to safety coefficients degradation (this can be
alleviated by using sub-criticality), to neutron economy and to performances (inventory, burn-up
achievable, losses). This paper shows that a factor 100 of mass reduction for TRU can be achieved if
losses are less than 0.18% and burn-up greater than 15%, which are real challenges. In this respect, the
normal PUREX process might be inadequate. Finally, new options such as pyrochemistry and the use
of sub-criticality with accelerator driven systems, which allow the use of fertile free fuels, will
probably be needed to achieve such performances. A comprehensive view of the principal actinide
transmutation strategies is given using evolutionary and innovative approaches and according to the
principal driving force: resource, waste, and proliferation. Concerning long-lived fission products,
such as 129 I, 99 Tc, 135 Cs, 93 Zr and 126 Sn, this paper gives a good review of the various difficulties to
include them in a P&T strategy (if chemically separated, most of them might be embedded in a more
stable matrix than glass). Finally, the impacts of various P&T options on geological disposal are
described.
The other papers describe some work related to P&T carried out in various laboratories as well as a
new proposal. In the T. Osugi et al. paper, some recent results related to the JAERI double strata strategy
are given: nitrides fuels, pyrochemical processes, 800 MWth ADS Pb-Bi target and coolant. Technical
issues related to ADS will be studied using the experimental facility at KEK, with the prospect for
commercial ADS around 2035. A 800 MWth ADS based on the lead coolant, Rubbia’s concept (but with
forced convection) is studied at CIEMAT in Spain, with the emphasis on new types of fuel (thorium and
nitride inert matrix) applied to TRU burning in LWR (normal MOX fuel) then in ADS.
29

W. Forsberg proposes to reduce the high-level waste heat loading by chemically separating five
heat generating nuclei: Cs, Sr (both to be put in an interim storage for heat decay) Pu, Am and Cm (to
be transmuted by P&T). The remnant wastes containing the low-heat radionuclides will be
geologically disposed at low cost (no interim storage and reduced disposal surface needed).
H. Boussier et al. calculates the potential risk of geological disposal within various P&T
scenarios described in the overview paper. The originality of this paper is to consider not the global
inventory radiotoxicity but that of the quantities which escapes from the waste deduced from the
matrix fraction subjected to alteration over time. However, the difficulty is to get accurate values for
such fractions.
The session ended up with the presentation by J. Vergnes et al. from EdF of a new concept of a
molten salt fuel critical and thermal reactor which is able to produce energy with a very low amount of
long lived wastes. In such a reactor called AMSTER (Actinides Molten Salt Transmuter) actinides are
recycled, fission products and eventually 236 U continuously extracted. At equilibrium and in a fissile
isogenerating mode, only fertile material (natural uranium or thorium) is fed into the reactor.
AMSTER can incinerate TRU produced in PWR or recycle its own actinides only. First theoretical
studies show that a reduction by several decades in the TRU quantities is expected, leading to a “clean
energy” reactor, especially if the thorium fuel cycle is used.
30

CLOSING THE NUCLEAR FUEL CYCLE: ISSUES AND PERSPECTIVES
Peter Wydler 2
5452 Oberrohrdorf, Switzerland
Leo H. Baetslé
SCK•CEN
Boeretang 200, 2400 Mol, Belgium
Abstract
Partitioning and transmutation (P&T) aims at making nuclear energy more sustainable from the
viewpoint of the back-end of the fuel cycle by minimising the high-level waste with respect to its mass,
radiotoxicity and (possibly) repository risk. P&T mainly deals with the management – i.e. transmutation
and/or special conditioning and confinement – of minor actinides and fission products, but involves the
closure of the fuel cycle for plutonium as a necessary first or parallel step. The conditions for a
completely closed fuel cycle, the goals for transmutation, and the implications for the reactor and fuel
cycle technology are overviewed and discussed, and the currently favoured transmutation strategies are
compared with respect to achievable waste radiotoxicity reduction and impact on the releases of
potentially troublesome actinides from a repository for vitrified high-level waste.
2 Under contract with Federal Office of Energy, CH-3003 Bern, Switzerland.
31

1. From the open fuel cycle to the semi-closed fuel cycles for plutonium management
The LWR once-through fuel cycle at a mean burn-up of 50 GWd/t HM produces (a) spent fuel
consisting of fission products (1.1 t/GWe-a), irradiated uranium, plutonium, and minor actinides
(about 20 t/GWe-a, mostly uranium), and (b) depleted uranium from the enrichment process (about
170-190 t/GWe-a, depending on the 235 U concentration of the tails).
If these materials are not further utilised, they have to be considered as nuclear waste. The
preferred option is currently to store the irradiated fuel elements after appropriate cooling in suitable
geological formations. The LWR once-through fuel cycle has the advantage that it avoids the
difficulties of reprocessing; however, it can only extract about half a percent of the energy content of
the mined uranium.
Depleted uranium stored as UF6 is a chemical hazard and becomes radiotoxic in the long-term.
Therefore, it has to be transformed into a more stable form (e.g. UO2 or U3O8) and appropriately stored
for use in future breeder reactors, or adequately disposed. In the nuclear waste discussion, not much
attention has yet been given to the management of the increasing stocks of depleted uranium.
1.1 Incentive for reprocessing
The incentive to reprocess irradiated LWR fuel arises primarily from the desire to improve the
uranium utilisation and implies the recycling of the bred plutonium (about 250 kg/GWe-a) in MOX-
LWRs or future high-conversion and fast-spectrum reactors. Multi-recycling of uranium and
plutonium in LWRs in the so-called self-sufficient mode would allow the uranium utilisation to
improve by about a factor of two, but involves a higher 235 U and plutonium enrichment due to the
“degradation” of the uranium and plutonium isotopic composition. A much better (close to 100%)
uranium utilisation could be achieved with fast reactors (FR).
Industrial reprocessing is currently based on the PUREX process which allows “clean” uranium
and plutonium to be separated from the fuel and was initially developed for military applications. The
remaining high-level waste (HLW), consisting mostly of fission products and minor actinides, is
converted into a stable form for ultimate disposal, with the normal method being the storage of
vitrified HLW in geologic repositories.
The separation of uranium and plutonium from the spent fuel has the advantages of reducing the
actinide mass and the plutonium content of the HLW. In combination with vitrification, it minimises
the risk of a clandestine recovery of fissile material from a waste repository; however, it does not
significantly reduce the long-term toxicity of the HLW. Drawbacks are the extra investment in
complex technology for reprocessing and α-active fuel fabrication, and the potential proliferation risk
associated with the handling of pure fissile materials. Since the balance of advantages and drawbacks
depends on regional boundary conditions and political factors, there are currently contradicting
policies regarding the recycling of plutonium.
1.2 Plutonium stock management
In some countries, an early commitment to industrial reprocessing, combined with the delayed
introduction of fast reactors, has led to large stocks of separated plutonium. The available fabrication
capacity for LWR-MOX fuel of about 300 tHM/a, however, will now allow some 25 tPu/a to be
recycled, which balances the current output of the reprocessing facilities [1].
32

Relative to the self-sufficient recycling mode, which has a zero plutonium balance, the plutonium
consumption of an LWR can be enhanced by increasing the number of MOX fuel elements in the core,
up to a full MOX core. An even higher plutonium consumption – up to the theoretical limit indicated
in Figure 1 for uranium-free concepts – could be achieved by reducing the uranium content of the fuel.
However, it should be noticed that LWRs alone cannot completely burn plutonium because the
buildup of the even isotopes in a thermal neutron spectrum constrains the number of recycles to two or
three at most. The remaining degraded plutonium has to be disposed, or stored until it can be utilised
in fast reactors, which offer similar net consumption rates as the LWRs (cf. Figure 1).
It is obvious that plutonium stocks can be managed effectively with LWRs and fast reactors; new
types of burner reactors or reprocessing methods are not needed, but plutonium could, of course, also
be managed with other types of reactors. The respective issues, including fuel developments, have
been discussed in the framework of working parties and workshops of OECD/NEA [2,3] and in many
international conferences. It should be emphasised that the closure of the fuel cycle for plutonium is a
prerequisite for, but not a direct issue of P&T. Therefore, plutonium management as such is not in the
focus of this paper.
Figure 1. Net plutonium production of different reactor types (kg/GWe-a)
-1050
FR, burner, U-free
-600
FR, burner, MOX
FR, reduced blanket
FR, breeder
250
-1260
LWR, U-free
-500
LWR, 100 % MOX
LWR, self sufficient
LWR, once-through
250
-1400 -1200 -1000 -800 -600 -400 -200 0 200 400
2. Fully closed fuel cycles with P&T
P&T aims at making nuclear energy more sustainable from the viewpoint of the back-end of the
fuel cycle and implies the separation and further utilisation of valuable materials as well as the
minimisation of the remaining HLW with respect to its mass, radiotoxicity and (perhaps) repository
risk. It thus responds to current concerns of the public and politicians who are not satisfied with a
radiological hazard which extends over millions of years, although the associated long-term risk in
terms of annual individual dose is very small.
33

Figures 2 and 3 show the radiotoxicity of the HLW produced for a 120 GWe-a scenario after
reprocessing and the resulting annual individual dose, estimated for vitrified waste emplaced in
cristalline host rock [4]. Evaluations for other scenarios and repository concepts give comparable
results [5,6] and confirm the following general observations:
• From the viewpoint of the radiotoxicity, which plays a role in accidental intrusion scenarios,
P&T must first be concerned with the actinides, particularly the minor actinides americium
and neptunium, the toxicity of the fission products (shaded in Figure 2) laying at least two
orders of magnitude below that of the actinides after 10 3 years.
• From the viewpoint of the long-term risk of a geologic repository, the relatively mobile
fission products are more important than the actinides, the fission products 135 Cs, 79 Se, 99 Tc
and 126 Sn being dominant dose contributors in vitrified HLW scenarios 3 .
• The fission product risk peaks in the time span 10 4 to 10 6 years after the closure of a
repository, whereas an actinide risk arises “only” after one million years.
2.1 Goal for minor actinide mass reduction
Figure 2 shows that the radiotoxicity of the actinides requires more than ten-thousand years to
decay to the toxicity level, Unat(LWR), given by the consumed natural uranium. With a hundred-fold
reduction in the actinide content, this goal could be reached already after a few hundred years. The fact
that the “natural toxicity” level for a pure fast reactor strategy, Unat(FR), is about hundred times
smaller than that for an LWR strategy speaks also for the goal of a reduction in the actinide content of
the HLW by a factor of 100.
It is obvious that a hundred-fold reduction of the actinide mass cannot be achieved in a single
pass through a reactor. Hence, multi-recycling will be essential. In fact, the ideal P&T system has a
fuel cycle which is fully closed for the actinides, meaning that only fission products are separated from
the spent fuel and all actinides are returned to the reactor, together with a “top-up” (make-up) of new
fuel replacing the fuel which was fissioned. It is also clear that such a system must be operated for
many decades before the core – and hence the composition of the discharged fuel, which determines
the specific waste radiotoxicity – reach an equilibrium.
2.2 Goal for actinide recovery
In practice, the actinides cannot be recovered completely from the spent fuel, and the remainder
will go to waste. For a system with a fully closed fuel cycle, the mass of actinides going to waste is:
M W = δ L M F
where M F is the total mass of actinides fissioned, L is the actinide loss fraction during reprocessing
and fuel fabrication, and the burn-up factor, δ, can be evaluated from the fraction of heavy metal
fissioned, B, as (1 - B) / B. Under equilibrium conditions, M F equals the top-up fuel mass, M T , which,
in general, can be divided into the mass, M B , of transuranic or minor actinides to be burnt
(i.e. transmuted and ultimately fissioned), and a diluent, usually consisting of fertile uranium (normal
critical burner cores are not suited for fertile-free top-up fuel).
3 The long-lived fission product 129 I, which is known to dominate the repository risk in direct storage scenarios,
is not present in vitrified HLW because it is discharged to the sea during reprocessing. Since sea disposal will
not be a desirable feature of a “clean” nuclear energy, 129 I is also a candidate for P&T.
34

Figure 2. Radiotoxicity of LWR spent fuel after uranium and plutonium separation
Figure 3. Annual individual dose for vitrified HLW emplaced in cristalline host rock
10 2
annual individual dose
from all pathways [mSv y -1 ]
10 1
10 0
10 -1
10 -2
10 -3
10 -4
10 -5
10 -6
10 -7
10 -8
Natural radiation exposures in Switzerland
Regulatory Guideline: 0.1 mSv y -1
Sum over nuclides.
4N + 3 chain
4N + 2 chain
4N + 1 chain
4N chain
126 Sn
135 Cs
59 Ni
79 Se
99 Tc
93 Zr
10 -9
10 2 10 3 10 4 10 5 10 6 10 7
time after repository closure [years]
35

Denoting the transuranic (TRU) or minor actinide (MA) fraction of the top-up fuel, M B /M T , by τ
and the “waste mass reduction factor”, M B /M W , by R M , one obtains the simple expression
L = τ / (δ R M ),
which gives the allowable losses as a function of the waste mass reduction factor. For the desired
reduction factor of 100, an achievable average fuel burn-up of 15%, and a top-up fuel without a fertile
component (τ = 1), the expression yields L = 0.18%. Since average burn-ups beyond 15% have not yet
been proven with known fuel technologies, a target value of 99.9% for the actinide recovery yield
must consequently be set for an effective P&T system.
3. Neutronic requirements for fully closed fuel cycles and role of the ADS
For neutronic reasons, not all reactors can operate with a fully closed fuel cycle. To assess the
suitability of a reactor in terms of neutron multiplication, the production-to-absorption ratio of the
actinides in the equilibrium core, ηec, is a useful parameter. Alternatively, the overall neutron balance
for the complete fissioning of actinides can be measured in terms of the “fuel neutron production
parameter” – D [7].
An ηec value smaller than 1 means that the fuel of the equilibrium core cannot maintain a chain
reaction; a negative -D value indicates that an actinide or an actinide mixture cannot be completely
fissioned. It can be shown that the parameters are mainly influenced by the neutron spectrum and flux
of the system and that the two approaches lead to the same conclusions.
The ηec and -D values in Table 1 refer to realistic reactor concepts including an ATW-type subcritical
thermal transmuter, a CAPRA-type fast plutonium and MA burner, a critical (sodium-cooled)
fast TRU burner, a sub-critical (LBE-cooled) fast TRU burner, and a dedicated sub-critical MA
burner. For the actinide feed, plutonium and TRU mixtures separated from PWR spent fuel and the
MA mixture from the first stratum of a typical “double strata” fuel cycle strategy [8] are assumed. The
values demonstrate that minor actinides cannot be completely burnt in thermal systems and that fast
plutonium and TRU burners offer more surplus neutrons than the respective thermal systems. The
surplus neutrons could be used for burning fission products.
Table 1. Neutronic performance of plutonium, minor actinide and transuranics burners
Thermal (ADS) Fast (critical) Fast ADS
Actinide feed a η ec -D η ec -D η ec -D
Plutonium
Minor actinides
Transuranics
1.15 b
0.89
1.11
0.40 b
-0.37
0.30
1.64
1.28
2.00
1.18
0.71
1.52
1.80
1.33
1.75
1.34
0.79
1.29
a Plutonium and TRU from PWR spent fuel with a burn-up of 50 GWd/t HM , minor actinides from a park
with PWR-UOX reactors (70%), PWR-MOX reactors (10%), and CAPRA reactors (20%).
b A MOX-PWR with self-sufficient plutonium recycling has a similar neutron economy.
36

3.1 Core design constraints
In practice, the design of a TRU or MA burner core, like that of any reactor core, is not only
constrained by the above-mentioned basic neutronic criterion, but also by performance and safety
parameters, such as the reactivity swing during burn-up, coolant void reactivity effect, Doppler
coefficient, effective delayed-neutron fraction, etc. In particular, for a sodium-cooled fast reactor core,
the substitution of normal MOX fuel by TRU- or MA-dominated fuel has an unfavourable influence
on several of these parameters, and this is one of the reasons for the recent revival of various fast and
thermal reactor concepts which were studied in the past, but have not been introduced as commercial
systems. For example, the (positive) coolant void effect in sodium-cooled fast reactors, which
deteriorates in a minor actinide burning regime, can be mitigated by substituting the sodium by lead,
or even eliminated by substituting the liquid metal by a gas coolant.
To ensure that a critical burner core performs satisfactorily and has acceptable safety parameters,
it is usually necessary to blend the TRU or minor actinides with the fertile materials uranium or
thorium. However, blending reduces the transmutation effectiveness of the system. In this context,
accelerator-driven systems offer interesting additional parameters of freedom by removing the
criticality constraint and increasing the safety margin to prompt criticality. The latter feature is
particulary important for MA burners, which are difficult to control as critical systems because the
effective delayed-neutron fraction is only about half of that of a normal fast reactor.
In response to the new core design issues raised by P&T and the increased interest in advanced
reactor technology in general, government and industry funded design teams in Europe, the Far East
and the USA are currently spending a considerable effort on the optimisation of a broad range of
advanced reactor designs featuring both critical and accelerator-driven cores.
3.2 Transmutation effectiveness
Various, sometimes conflicting definitions for transmutation effectiveness, usually based on the
minor actinide balance of a particular core, can be found in the literature (see, for instance, [6]). For a
system with fully closed fuel cycle, the difficulty of defining a core-specific transmutation effectiveness
is circumvented by defining an “overall transmutation effectiveness” as the relative content of the top-up
fuel in transuranic and minor actinides, M B /M T , i.e. the already discussed quantity τ.
It is interesting to notice that the overall transmutation effectiveness does not depend directly on
the choice of the neutron spectrum, the fuel type and the coolant, but is governed by the abovementioned
performance and safety constraints. For a critical TRU burner based on liquid metal
technology, τ is smaller than about 0.5, and in the case of homogeneous MA recycling in a EFR-type
fast reactor τ is less than 0.1. The possibility to operate sub-critical MA and TRU burner cores with a
fertile-free top-up fuel and hence 100% overall transmutation effectiveness, i.e. τ = 1, is probably the
most important advantage of accelerator-driven systems.
3.3 Neutronic transmutation effect
Using the same notation as before, the radiotoxicity reduction relative to the top-up fuel, R T (t), is:
R T (t) = R N (t) M T / M W
37

or, in terms of the fuel burn-up and the fuel loss,
R T (t) = R N (t) / (δ L)
where R N (t) is a time-dependent “neutronic transmutation factor” 4 . It should be noticed that the
toxicity reduction relative to the actinides to be burnt, which is of direct relevance for the assessment
of transmutation schemes, equals R T for all practical purposes because the toxicity of the fertile
component of the top-up fuel is negligible compared with that of the actinides to be burnt.
Figure 4 compares the neutronic transmutation factors for three systems: (a) an ALMR-type
critical TRU burner, (b) an ATW-type sub-critical TRU burner with a fast-neutron spectrum, and (c) a
minor actinide burner operating in the P&T cycle of a double strata fuel cycle. It can be seen that the
critical burner has a small “neutronic” advantage over the sub-critical burners, but the difference is not
significant compared with the goal for the total toxicity reduction by a factor of 100.
The general conclusions to be drawn from this discussion are that:
• Regarding the neutronic transmutation factor, no single transmutation system has a significant
advantage over other systems.
• Radiotoxicity reduction has to be achieved primarily by an actinide mass reduction which
implies the maximisation of fuel burn-up and the minimisation of reprocessing and fuel
fabrication losses.
The importance of advanced reprocessing and fuel technologies for P&T is thus confirmed.
Figure 4. Neutronic transmutation factors for transuranics and minor actinide burners
10
Neutronic transmutation factor
1
Critical TRU Burner
Subcritical TRU Burner
Subcritical MA Burner
0.1
1.E+2 1.E+3 1.E+4 1.E+5 1.E+6 1.E+7
Time after fuel reprocessing (a)
4 R N (t) is a characteristic of the core and is sometimes called “neutronic toxicity reduction”.
38

4. Fuel cycle problems and challenges
Current proposals for P&T technology rely on the aqueous (PUREX-type) reprocessing of spent
fuel as a preliminary step preceding minor actinide partitioning. But, even in case of pyrometallurgical
processing of TRUs, a mechanical head-end and an aqueous processing step (called UREX) for the
prior removal of uranium as the main fertile element in any case precedes the sequence of
pyrometallurgical separation steps.
The drastic increase of the plutonium content from 15% up to 43% and the fuel burn-up from
80 GWd/tHM up to 210 GWd/tHM for fast reactor concepts such as CAPRA make aqueous reprocessing
more difficult because of the low solubility of plutonium and the radiation damage to the organic
extractant (tributyl phosphate). Industrial “pilot” scale work at Dounreay and Marcoule has shown
that, in specific technological conditions (pin-chopper, fast contactors), aqueous reprocessing can be
considered as valid for fast reactor and future ADS fuel, if the decay heat can be mitigated by cooling
or by dilution with LWR fuel. Special chopping or shearing systems have been developed for fast
reactor (FR) fuel, e.g. for FFTF, in order to combine the single or multiple pin shearing with a more
economical bundle-shear approach.
Spent fuel arising from a reactor park composed of 70% LWR-UOX reactors, 10% LWR-MOX
reactors and 20% FR-MOX reactors has to be reprocessed in order to guarantee the plutonium recycle
requirements in a stable scenario. The corresponding spent fuel discharges per 100 Gwe-a delivered by
the reactor park are summarised in Table 2. Table 3 gives the residual decay heat for these fuel types,
assuming the reprocessing operations to be carried out 5 years after the discharge of the fuel from the
reactor.
Table 2. Spent fuel arisings from a composite nuclear reactor park per 100 GWe-a
Fuel type
Burn-up
(GWd/t HM )
Delivered energy
(GWe)
Spent fuel discharge
(t HM )
LWR-UOX
LWR-MOX
FR-MOX
50
50
150-200
69.7
10.3
20.0
1 520
224
124-153
Table 3. Decay heat of spent fuel 5 years after discharge from reactor
Fuel type
Burn-up
(GWd/t HM )
Total decay heat
(kW/t HM )
Fission products
(kW/t HM )
Actinides
(kW/t HM )
LWR-UOX
LWR-MOX
FR-MOX
FR-MOX
50
50
150
210
2.76
6.56
30.3
33.7
2.28
1.97
6.60
8.14
0.477
4.60
23.7
25.6
4.1 Issues of aqueous reprocessing
The present state-of-the-art in aqueous reprocessing based on the PUREX process is almost
satisfactory with regard to uranium and plutonium separation. Recovery yields of nearly 99.7% have
been achieved and 99.9% is potentially achievable in a near future. However, reprocessing of industrial
39

quantities of LWR-MOX and FR-MOX within short cooling times will require gradual adaptations of
the PUREX flow-sheet and involve the construction of additional extraction rigs.
In order to improve the plutonium dissolution yield and to avoid solvent radiation peaks during the
extraction, a separate dissolver dedicated to FR-MOX treatment could be installed and connected to the
main dissolver by a metering system. By connecting the dedicated FR-MOX dissolver to the main LWR
dissolver, a constant radiation level can be kept throughout a given processing campaign. The second
dissolver could also serve as “residue dissolver” by making use of highly oxidising compounds (e.g.
electrochemically generated Ag(II)) to dissolve the insoluble fraction of the initial plutonium inventory
and to reduce the transfer of insoluble plutonium residue to the waste stream.
The decay heat of the separated plutonium fraction enters the organic phase and remains in contact
with the actinide fraction (U, Pu, Np) until separation and purification of the major actinides occurs. The
238 Pu concentration in the plutonium stream is the main source of radiation damage to the tributyl
phosphate extractant, a strong source of neutrons in the separation plant, and an additional radiolysis
agent during the oxalate precipitation and conversion to PuO2. On the other hand, the high 238 Pu
concentration in the separated PuO2 for the same reasons (heat, spontaneous neutrons) improves the nonproliferation
resistance of the product during storage. The MOX fuel fabrication will have to adapt its
handling technology to reduce the radiation dose to the workers in the plant and during the transport of
MOX fuel assemblies.
The separation of the minor actinides is currently under intense investigation throughout the
OECD/NEA countries. The separation of the bulk of 237 Np from the aqueous product solution during the
reprocessing operations is technically feasible but requires an adaptation of the first cycle extraction
flow-sheet. The residual 237 Np which goes directly to the HLW solution could, in principle, be recovered
by recycling the HLW solution through a secondary recovery cycle after a valence adjustment. With
liquid extraction processes, the separation yield can be increased up to the desired value of 99.9% by
increasing the number of extraction stages.
The separation of americium and curium is much more difficult, but considerable progress has been
made during the past decade. Bulk separation of minor actinides together with lanthanides has been
demonstrated under hot-laboratory conditions with several new processes such as TRUEX, DIDPA,
DIAMEX and TRPO [6,9]. The separation of minor actinides from lanthanides, however, remains an
obstacle for the industrial up-scaling of Am-Cm separation from HLW. Several promising laboratory
methods – e.g. the ALINA and CYANEX301 processes using DTPA [10] and BTP [11] as specific
extractants for the separation of minor actinides from lanthanides – have been tested at the ITU of JRC
Karlsruhe. Yields of 99% and 97.6% have been achieved for americium and curium, respectively.
Further improvements are necessary in order to include these methods in a cycle of multiple
reprocessing.
By progressively incorporating an additional extraction rig for minor actinides from HLW coupled
to a conventional vitrification process, a much less toxic HLW could be obtained, keeping the actinides
in the fuel cycle for ultimate fissioning, and keeping the fission products in the glass.
4.2 Pyrochemical reprocessing
The proliferation risk potentially associated with the clean plutonium produced by the aqueous
reprocessing has drawn the attention on the pyrochemical reprocessing which makes it difficult to
separate individual TRU elements. Whereas the pioneering work was performed by ANL in the USA
[12], most of the recent advances have been made in Russia [13] and Japan [14]. In the Russian
40

Institute of Atomic Research (RIAR), a pyrochemical process has been demonstrated with highly
irradiated spent oxide fuel with burn-ups of 210-240 GWd/tHM. Good results were obtained during the
demonstration. The recovered PuO2 will be mixed with UO2 and processed by vibropacking into fresh
FR-MOX fuel. Pyrochemical reprocessing of metallic fuel was developed in the USA in the
framework of the integral fast reactor (IFR) programme with capability for actinide burning. Recovery
yields of 95% for uranium and 99% for mixtures of uranium and minor actinides have been obtained
on laboratory and pilot-plant scales at the Idaho Argonne West Laboratories.
The advantages and disadvantages of the pyrochemical reprocessing can be summarised as
follows:
• Highly concentrated TRU product streams can be handled without major radiation
degradation of the reagents, allowing the facilities to be more compact than aqueous
reprocessing facilities for the same TRU throughput. Because of the absence of water, much
smaller criticality risks during purification and metallic fuel re-fabrication arise when
processing industrial quantities of TRUs.
• Since all TRU elements remain together throughout the process, the proliferation risk is much
reduced. The separation of the TRUs from the lanthanides, however, is difficult, requiring the
development of multistage countercurrent extraction using highly corrosive reagents at high
temperature. The most challenging issue is the selection and industrial manufacture of
corrosion resistant equipment which must be designed for remote operation and maintenance.
• For spent LWR-UOX fuel, a genuine pyrochemical process has to cope with the problem of
the elimination of the uranium, which is the major constituent of the LWR fuel. Therefore, a
mixed approach using a (P)UREX-type process for uranium-neptunium elimination has been
selected as the first step of the “ATW road map” project. Fluoride volatility has also been
suggested as an alternative, but the mixed volatility of U-Pu and zirconium leads to complex
waste streams which can be difficult to control.
• Starting from metal or nitride TRU fuel, a complete pyrometallurgical process involving a
series of electro-refining steps in a wide range of molten salt baths has recently attracted
much interest throughout the nuclear research communities in the USA, Russia, Japan and
France. Most of the respective flow-sheets remain in the pre-conceptual phase and will
require several decades of R&D to become a mature technology comparable to the present
aqueous reprocessing.
• Whereas the aqueous reprocessing consists of an independent industrial process which supports
a large reactor park and can operate independently on continental or even world scales, the
pyrochemical process will predominantly be applied in small facilities installed in the
immediate vicinity of the reactors.
• However, in the long-term, pyrochemical reprocessing will become indispensable, if very hot
fuel has to be multi-recycled in fast reactor or ADS facilities on an industrial scale with a
limited out-of-pile inventory.
5. Principal actinide transmutation strategies
Depending on regional boundary conditions and political factors, countries with P&T
programmes have developed different transmutation strategies. As to the transmutation of actinides,
Table 4 provides an overview of the principal approaches and indicates respective driving forces. In
41

view of the historic development, the table distinguishes between evolutionary (right) and innovative
(left) approaches.
The evolutionary approach, adopted mainly in Europe and Japan, aims at closing the fuel cycle in
successive steps, starting with the recycling of plutonium in LWRs and later in fast reactors using
conventional reprocessing and MOX fuel technology, and finally complementing the system with a
dedicated P&T cycle which features MA burners with fast neutron spectra. This approach has the
advantage that it can respond flexibly to changes in the priorities for plutonium and minor actinide
management, and that new technologies have to be developed only for a comparatively small number
of MA burners which support a large system of conventional LWRs and fast reactors.
Table 4. Principal actinide transmutation strategies
TRU burning
Separation of uranium and TRU from spent
LWR fuel, TRU remain together.
TRU recycled in thermal or fast critical or subcritical
reactors with fully closed fuel cycles.
Dry reprocessing particularly suited for closed
fuel cycles and highly active fuels.
Principal driving force:
non-proliferation
Pu recycling
Separation of uranium and Pu from spent LWR fuel.
Pu recycled in thermal and later in fast reactors
(limited number of “thermal” recycles).
PUREX-type wet reprocessing methods are appropriate.
Flexible thermal: fast reactor ratio from about 4 to zero.
Possibility to move to a pure fast reactor strategy.
Principal driving force:
resource management
With thermal
neutrons
With fast neutrons Without transmutation With MA transmutation
TRU burning
in thermal reactor.
TRU burning in FR
Pu burning
Double strata fuel cycle
TRU burning
in thermal ADS
Pure burner strategy.
(thermal ATW system)
Flexible thermal: fast
reactor ratio from
about 2 to zero.
Possibility to move to
a pure FR strategy
(IFR system).
TRU burning in fast
ADS
Thermal:fast reactor
ratio of about 3.
Possibility to move to
a pure ADS strategy
by adding fertile fuel
(Energy Amplifier).
Fully closed fuel cycle for
Pu.
MA and FP conditioning
by vitrification and/or
dedicated insolubilisation
with ceramics technology.
Principal driving force:
waste management.
MAs (and FPs) burnt in a
fully closed P&T cycle.
Fast spectrum required for
MA transmutation,
ADS has safety
advantages.
Dry reprocessing
particularly suited for
fully closed cycle.
One MA burner supports
some 15 “normal”
reactors.
Principal driving force:
waste management.
The innovative approach, first suggested in the USA, aims at co-processing plutonium and minor
actinides to avoid the use of technologies with a potentially high proliferation risk. After the initial
42

separation of the uranium from the LWR spent fuel, the unseparated transuranic actinides are recycled
in a transuranic burner with a closed fuel cycle based on pyrochemical reprocessing. Compared with
the double strata strategy, the number of burners in the equivalent system is four to six times larger,
but the burners are not subjected to a (fast) neutron-spectrum condition. Nevertheless, most of the
currently evaluated critical and sub-critical transuranic burners feature a fast neutron spectrum.
5.1 Strategies studied by OECD/NEA
An expert group of the OECD/NEA Nuclear Development Committee is currently comparing the
principal actinide burning and transmutation strategies in more detail. The investigated strategies
comprise:
a) Plutonium burning in LWRs and CAPRA-type fast reactors.
b) The double strata strategy with LWRs and CAPRA reactors in the first stratum and
accelerator-driven MA burners in the second stratum.
c) TRU burning in critical fast reactors (IFR concept).
d) TRU burning in sub-critical fast reactors (ADS).
e) A heterogeneous strategy in which americium and curium are recycled in targets.
Table 5 summarises the most important assumptions for strategies a to d, and Figure 5 gives
preliminary results for the achievable actinide waste radiotoxicity reduction relative to the open fuel
cycle. The figure shows that transmutation strategies b, c and d all meet the goal of a hundred-fold
reduction, and confirms that plutonium recycling alone is not effective in reducing the actinide waste
radiotoxicity.
Figure 5. Actinide waste radiotoxicity reduction relative to open fuel cycle
(Preliminary results of OECD/NEA study)
1000
Actinide waste toxicity reduction
100
10
1
Plutonium burning
TRU burning in FR
TRU burning in ADS
Double strata
0.1
1.E+0 1.E+1 1.E+2 1.E+3 1.E+4 1.E+5 1.E+6 1.E+7
Time after fuel reprocessing (a)
43

Table 5. Assumptions for transmutation strategies compared by OECD/NEA
Strategies Reactor/ADS Fuel
Av. burn-up
(GWd/t HM )
Reprocessing
method
Recovery
yield (%)
All
a, b
b
c
d
LWR (N4)
FR (CAPRA)
MA Burner, ADS
TRU Burner, IFR
TRU Burner, ADS
UOX/MOX
MOX
AcN-ZrN
Ac-Zr
Ac-Zr
50
185
140
140
140
wet
wet
dry
dry
dry
99.9
99.9
99.9
99.9
99.9
6. Fission product transmutation
The neutron capture process is currently the only promising nuclear reaction for transmuting
fission products. Other processes which have been proposed in the past rely on technologies which are
still at a very early stage of development (e.g. fusion neutron sources) and generally suffer from a poor
energy balance. The capture process consumes neutrons but, theoretically, fast reactors could deliver
enough surplus neutrons to allow the potentially troublesome long-lived fission products to be
completely transmuted to shorter-lived or stable species (cf. Table 1).
The transmutation of a fission product makes sense only if the reaction rate (microscopic crosssection
times neutron flux) is higher than the natural decay rate of the nuclide. With the practically
achievable neutron fluxes, this condition cannot be met for the most abundant fission products 137 Cs
and 90 Sr with half-lives of only about 30 years, i.e. these fission products are “non-transmutable”.
However, since their radioactive life is limited to less than 300 years, they can be safely enclosed
using engineered barriers only. On the other hand, the fission products determine the size of the
vitrified waste repository which, consequently, cannot be much reduced by P&T operations.
The fission products which influence the long-term risk of a repository are, in order of radiologic
importance, 129 I, 99 Tc, 135 Cs, 93 Zr, and 126 Sn. Activation products ( 14 C and 36 Cl) can also contribute to
the dose. Obviously, the relative contribution of these nuclides to the integral risk varies according to
the type of repository host rock. In the following, the problems associated with the transmutation of
these five fission products are discussed separately. It should be noted that the determination of the
separation yield and the decontamination factor (DF) for HLW depends to a great extent on policy
decisions and that, depending on the nuclide, special conditioning and confinement is an alternative to
transmutation.
6.1 Iodine-129 (T 1/2 = 1.6·10 7 a)
In most of the direct storage concepts for spent fuel, 129 I is the first nuclide to enter into the
biosphere due to its very high mobility. During aqueous reprocessing, iodine is removed from the
dissolver solution with a yield of 95-98%. To improve the separation yield, more complex chemical
treatments are necessary. Better separation yields may also be achieved with high-temperature
pyrochemical processes. Special methods for conditioning in the form of AgI, PbIO4, etc. have been
developed but, because of the extremely long half-life, the eventual migration to the environment
cannot be excluded.
44

Since the radiotoxicity of 129 I is the highest of the fission products and equivalent to that of
actinides, it would be advisable to increase the separation yield to achieve a DF of about 1 000. The
necessity of an isotopic separation and the limited stability of the target material, however, make the
transmutation of 129 I difficult, and conditioning and confinement seems to be the best method to reduce
its radiological impact. Nevertheless, the present method of diluting iodine in the sea may still be
defendable because of the enormous dilution of 129 I in the (natural) 127 I present in seawater. The
storage in a salt dome, consisting of evaporated seawater, is an another alternative which still has its
merits.
6.2 Technetium-99 (T1/2 = 2.1·10 5 a)
The radiologic significance of 99 Tc is important if the repository surroundings are slightly oxidic.
In reducing conditions 99 Tc is remarkably stable and insoluble as technetium metal or TcO2 suboxide.
Partitioning of 99 Tc is not an easy task because it occurs as insoluble metal and as soluble technetate
ion in solution. Separation from aqueous effluents is possible in an advanced PUREX scheme, but
recovery from insoluble residues is difficult, with the present recovery yield at best reaching 80%.
Improving this yield significantly implies the development of new separation technologies such as the
not yet proven conversion of the technetium into a single chemical species. Alternatively, a group
separation together with the platinum metals may be carried out using pyrometallurgical processes. If
separated in metallic form, transmutation appears to be feasible because of its stability and relatively
large neutron capture cross-section.
6.3 Caesium-135 (T1/2 = 2.3·10 6 a), Zirconium-93 (T1/2 = 1.5·10 6 a) and Tin-126 (T1/2 = 1.0·10 5 a)
Caesium occurs in the form of the isotopes 133 (stable), 135 and 137. In terms of radiologic
significance, 137 Cs is the major constituent of HLW (see above). The activity of the long-lived 135 Cs in
HLW is a million times lower. However, once released from a matrix as glass, caesium is very mobile.
Transformation of 135 Cs to stable 136 Ba is possible from a neutronics point of view, but probably
impracticable because a close to 100% isotopic separation efficiency would be required (traces of
133 Cs in the target would generate new 135 Cs during the irradiation).
Zirconium-93 is somewhat similar to 135 Cs, it has also a very long half-life and a small isotopic
abundance (about 14% of the total Zr). An isotopic separation would be necessary, and its
transformation to stable 94 Zr would be very slow because the thermal capture cross-section is about
five times smaller than that of 135 Cs.
Tin-126 is partly soluble in HLW from aqueous reprocessing but occurs also as an insoluble
residue, similar to technetium. Isolation involves a special treatment of the HLW, and the use of
isotopic separation techniques. Transmutation of 126 Sn is questionable due to the very low neutron
capture cross-section.
7. Consequences for geologic disposal
As mentioned before, the primary concern of geologic repositories are possible releases of the
relatively mobile fission products. Since the fission product yields are not very sensitive to the fuel
composition and the neutron spectrum of the reactor, the fission product risk depends primarily on the
number of fissions, i.e. the energy, produced in the fuel. This means that the fission product risk
cannot be much influenced by the actinide transmutation strategy and can only be mitigated by
45

separating troublesome fission products from the waste. The choice of the reactor can, however,
influence the in-situ transmutation of a fission product. For example, capture of thermal neutrons in
the precursor 135 Xe reduces the 135 Cs production in an LWR by as much as 70% [15].
As to the influence of different transmutation strategies on the release of actinides from a
repository, the usual approach is to perform complete nuclide transport calculations. Such an integral
approach was chosen, for instance, by the authors of a recent study of the European Commission who
compared nuclide fluxes at the clay-aquifer interface [16]. However, a difficulty of this approach is the
strong dependence of the results on the host rock characteristics, which can vary considerably
depending on the structure of the host rock and the storage concept (storage in salt, clay or granite).
For generic studies, it may therefore be preferable to perform site-independent comparisons on the
basis of releases of potentially troublesome actinides from the well-defined near-field of the
repository 5 , allowing a subsequent folding with site-specific geosphere and biosphere responses to be
performed independently and according to the needs of specific repository projects [15].
Adopting the latter approach, near-field release rates have been evaluated for the strategies with
vitrification investigated in the framework of the afore-mentioned OECD/NEA study. The results in
Figure 6 indicate a strongly non-linear relationship between release rates and actinide concentrations
in the glass. Important conclusions are that the plutonium burning strategy generally increases the
maximum release rates, and the addition of the P&T cycle does not reduce maximum release rates for
all potentially troublesome actinides.
Figure 6. Maximum release rates from near-field relative
to LWR once-through/vitrification case
(Strategies investigated in the framework of the OECD/NEA study)
Max. release rate relative to LWR-OT/vitrification
1.E+02
1.E+01
1.E+00
1.E-01
1.E-02
1.E-03
Np-237 Th-229 Pu-242 Ra-226 Pu-239 Pa-231
Plutonium burning
Double strata
5 Actinides to be considered in this context are 231 Pa, 237 Np and their respective daughter products 227 Ac and
229 Th, 226 Ra, a decay product of 234 U, and the long-lived plutonium isotopes 239 and 242.
46

8. Summary
The principal points and conclusions arising from this overview discussion of the nuclear fuel
cycle and P&T are summarised as follows:
The closure of the nuclear fuel cycle with P&T will be a long-term endeavour and becomes a
central issue in the development of a future sustainable nuclear energy system; the P&T strategy will
directly influence the choices of new reactor and reprocessing technologies.
Plutonium recycling is a first step in this direction. Plutonium can be managed effectively with
existing LWRs which, later on, should be complemented with fast reactors to burn the plutonium
completely. The necessary extension of the proven, PUREX-type reprocessing technology to cope
with the highly active fuels arising from plutonium-burning strategies appears to be feasible. The
motivation for the utilisation of plutonium is its energy content, but not a mitigation of the long-term
radiological hazard associated with the back-end of the fuel cycle.
Partitioning and transmutation aims at making the fuel cycle more sustainable from the viewpoint
of the back-end by reducing the HLW radiotoxicity and (possibly) the migration of radiotoxic nuclides
from HLW repositories to the biosphere. To this end, it introduces the separation and transmutation
(or, alternatively, improved immobilisation) of minor actinides and fission products, assuming that the
fuel cycle is already (or simultaneously) closed for plutonium.
Transmutation implies the development of advanced and innovative reactor and fuel cycle
technologies, including ADS reactor technology, fuels with very high burn-up capability, and pyrochemical
reprocessing methods. It can be shown that the goals of transmutation cannot be achieved
without the implementation of fast-neutron-spectrum systems in some form, and that the overriding
requirement is that for high fuel burn-ups and actinide recovery yields. Regarding neutronics, it
appears that no single transmutation system has a significant advantage over other systems.
As to the separation of minor actinides from HLW using aqueous reprocessing, promising new
processes have recently been developed. The results achieved at laboratory and pilot-plant scales give
confidence that the required high recovery yields can ultimately be achieved on an industrial scale.
Nevertheless, the long-term future appears to belong to the pyrochemical reprocessing method, which
is not yet mature, but is intrinsically better suited for a fully closed fuel cycle with highly active fuels
and may be more proliferation resistant because all actinides remain together, and the relatively
compact plants can be collocated with the reactors.
A comparison of the double strata and the TRU burning strategy indicates that, under comparable
assumptions, the two strategies are equivalent and that both have the potential of achieving a hundredfold
radiotoxicity reduction of the actinides in the HLW. These (or any other) actinide transmutation
strategies, however, are not effective in reducing the total mass of the HLW which is dominated by the
fission products and, hence, mainly depends on the total energy produced.
The primary risk of geologic repositories is related to the release of long-lived fission products.
With the exception of 99 Tc, however, the transmutation of long-lived fission products appears to be
difficult because of low neutron reaction cross-sections and the necessity of isotopic separations. This
means that, for most fission products, special conditioning and confinement is the only practical
method to reduce the radiological impact. Finally, it is shown that, due to the non-linear relationship
between release rates from a repository and actinide concentrations in the glass, P&T does not
necessarily result in reduced release rates for all potentially troublesome actinides as one could expect.
47

REFERENCES
[1] L.H. Baetslé, Ch. De Raedt and G. Volckaert, Impact of Advanced Fuel Cycles and Irradiation
Scenarios on Final Disposal Issues, Proc. Int. Conf. on Future Nuclear Systems, Global’99,
29 August-3 September 1999, Jackson Hole, Wyoming, Paper 017 (CD ROM).
[2] OECD Nuclear Energy Agency, Physics of Plutonium Recycling: Issues and Perspectives,
Vol. 1, Paris, France, (1995).
[3] OECD Nuclear Energy Agency, Advanced Reactors with Innovative Fuels, Workshop
Proceedings, Villigen, Switzerland, 21-23 October 1998, Paris, France, (1999).
[4] Nagra, Kristallin-I Safety Assessment Report, Nagra Technical Report NTB 93-22, Nagra,
Wettingen, Switzerland, (1994).
[5] G. Volckaert et al., Long-term Environmental Impact of Underground Disposal of P&T Waste,
Proc. 5 th Int. Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, 25-27 November 1998, Mol, Belgium, p. 463, EUR 18898 EN, OECD Nuclear
Energy Agency, Paris, France, (1999).
[6] OECD Nuclear Energy Agency, Actinide and Fission Product Partitioning and Transmutation;
Status and Assessment Report, Paris, France, (1999).
[7] M. Salvatores, I. Slessarev and M. Uematsu, A Global Physics Approach to Transmutation of
Radioactive Nuclei, Nucl. Sci. Eng., 116, 1-18 (1994).
[8] T. Mukaiyama, Importance of Double Strata Fuel Cycle for Minor Actinide Transmutation,
Proc. 3 rd Int. Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, 12-14 December 1994, Cadarache, France, p. 30, OECD Nuclear Energy
Agency, Paris, France, (1995).
[9] OECD Nuclear Energy Agency, Actinide and Fission Product Partitioning and Transmutation,
Proc. 5 th Int. Information Exchange Meeting, 25-27 November 1998, Mol, Belgium, p. 111-192,
EUR 18898 EN, Paris, France, (1999).
[10] G. Modolo and R. Odoj, Actinides(III)-lanthanides Group Separation from Nitric Acid Using
New Aromatic Diorganyldithiophosphinic Acids, ibid., p. 141.
[11] P.Y. Cordier, C. Hill, C. Madic and Z. Kolarik, New Molecules for An(III)/Ln(III) Separation by
Liquid Liquid Extraction, ibid., p. 479.
[12] Y.I. Chang, The Integral Fast Reactor, Nucl. Technol., 88, 129 (1989).
48

[13] A.V. Bychkov et al., Pyroelectrochemical Reprocessing of Irradiated FBR-MOX Fuel. III.
Experiment on High Burn-up Fuel of the BOR-60 Reactor, Proc. Int. Conf. on Future Nuclear
Systems, Global’97, 5-10 October 1997, Yokohama, Japan, p. 912.
[14] T. Inoue and H. Tanaka, Recycling of Actinides Produced in LWR and FBR Fuel Cycles by
Applying Pyrometallurgical Process, ibid., p. 646.
[15] P. Wydler and E. Curti, Closing the Fuel Cycle: Consequences for the Long-term Risk, ibid., p. 201.
[16] Evaluation of Possible Partitioning and Transmutation Strategies and of Means for
Implementing Them, European Commission, Project report, EUR 19128 EN (2000).
49

SESSION III
Partitioning
Chairs: J.P. Glatz (ITU) – J. Laidler (ANL)
_____________________
SUMMARY
Professor Charles Madic of DEA-DCC (Saclay) presented an overview paper on chemical partitioning
in which he described in considerable detail the French processes for the extraction of minor actinides and
their separation from lanthanide elements. He noted that the number of aqueous processes is burgeoning
and that the processes are becoming too complex. He registered a plea for process simplification and for
reduction in the size of process equipment, achievable perhaps by pre-concentration of the solutions to be
processed.
Professor Michael Hudson of University of Reading gave a richly detailed description of the
structure and characteristics of heterocyclic ligands that can be exceedingly useful in the extraction of
lanthanides and minor actinides. He presented a comprehensive model for designing ligands to
function as specialized extractants for specific lanthanide elements.
Dr. Jean-Paul Glatz of the European Commission Joint Research Centre/Institute for Transuranium
Elements (EC-JRC/ITU) reported on the successful demonstration of minor actinide/lanthanide
separations using a nitric acid PUREX raffinate solution containing minor actinides and lanthanides. Bistriazinepyridine
(BTP) was used as the extractant in a 16-stage centrifugal contactor train operating in
the counter-current mode. Reasonable decontamination factors and recovery efficiencies were achieved.
The work is particularly noteworthy because it is the first demonstration of minor actinide/lanthanide
separation using an actual waste stream.
Dr. James Laidler of Argonne National Laboratory presented a paper describing a pyrochemical
process being developed for use with a non-fertile metallic transmuter blanket fuel. This chloride
volatility process involves digestion of the inert zirconium matrix by formation of volatile ZrCl 4 .
Transuranic elements are subsequently recovered from the residual salt by electrowinning. The unit
operations comprising this process have all been successfully demonstrated with simulated fuel.
Dr. Jan Uhlir of the Nuclear Research Institute, Rez, Czech Republic, proposed the use of a fluoride
volatility method as a continuous or semi-continuous process for partitioning molten salt fuels in a
molten salt transmutation reactor scheme. He maintained that a practical near-term application of
fluoride volatility processing may be as a means for processing oxide fuels to remove uranium and
51

ecover transuranics for fissioning in a transmuter system. He cited experience in processing BOR-60
oxide fuel at Dimitrovgrad in the 1980s.
The CRIEPI/Transuranium Institute collaborative effort was described by Dr. Jean-Paul Glatz.
This study involves the processing of metal alloy fuels (U-Pu-Zr and U-Pu-MA-Ln-Zr) using molten
salt electrorefining and reductive extraction. A capability for small-scale hot processing has been
established at EC-JRC/ITU, and preliminary experiments have been carried out with the electrorefiner,
including both solid and liquid cathode deposition. A molten salt/metal reductive extraction process
has been used to demonstrate the cleanup of electrolyte salt as well as for the treatment of high-level
liquid waste.
Dr. M. Iizuka of CRIEPI described their work in development of the liquid cadmium cathode.
The work was performed with a small (9 mm dia.) cathode crucible, without stirring or agitation. The
result was that complete recovery of Pu could be obtained at low current densities and low
concentrations, but that Pu loss occurred at higher current densities by rapid growth of dendrites. The
work led to a projection of a Pu collection rate of nearly 300 grams per hour in a crucible of practical
dimensions, with Pu loadings in the cathode approaching 5 wt.%.
52

OVERVIEW OF THE HYDROMETALLURGICAL AND
PYRO-METALLURGICAL PROCESSES STUDIED WORLDWIDE
FOR THE PARTITIONING OF HIGH ACTIVE NUCLEAR WASTES
Charles Madic
DEN, CEA/Saclay
91191 Gif-sur-Yvette, France
E-mail: madic@amandin.cea.fr
Abstract
For more than 10 years, a revival of the interest for the partitioning of high active nuclear wastes
(HAWs) exists worldwide in connection with possible improvements of the management of the HAWs
actually produced and with futuristic nuclear fuel cycles. The main aim of the partitioning processes is
to separate, from the complex mixtures of HAWs, the long-lived radionuclides (LLRNs), belonging
either to the minor actinides (MAs) or to the fission products (FPs) families of elements, in order to
prepare fuels and/or targets suitable for their subsequent transmutation (P&T strategy). An other
possible strategy consists in the special conditioning of the separated LLRNs into stable dedicated
matrices (P&C strategy).
The LLRNs considered for partitioning are the MAs, neptunium, americium and curium, but also the
FPs, caesium, technetium and iodine.
Most of the partitioning processes studied so far belongs to the domain of hydrometallurgy, but,
recently, a new impetus was observed in the field of pyrometallurgical processes.
The main aim of this talk is to present a brief status of the development of both “hydro” and “pyro”
processes for the partitioning of LLRNs developed worldwide, with a special emphasis on the
benefits/drawbacks of each process.
53

1. Introduction
Since the end of the 80s, a renewal of interest is observed worldwide for LLRNs partitioning
techniques from nuclear wastes (HAWs). This interest is connected with two main fields:
• The conventional LWR closed fuel cycle using the PUREX process. New management
methods for nuclear wastes are considered, the so-called partitioning-transmutation (P&T)
and partitioning-conditioning (P&C) scenarios. For this domain, hydrometallurgy is the main
route for LLRNs partitioning process development, while pyrometallurgy is also subject of
some research.
• “New” fuel cycles associated with the development of fast reactors (FRs), accelerator driven
systems (ADSs) and fused salt reactors (FSRs). For this field, pyrometallurgy is the main
route considered for spent fuel reprocessing and wastes partitioning, while a small interest
still remains for the hydrometallurgical route.
The aim of the present article is to give a brief overview of the progress realised worldwide in the
recent years in the field of partitioning of LLRNs by hydrometallurgical and pyrometallurgical
processes.
2. Partitioning processes: an overview
2.1 General considerations [1]
2.1.1 Target elements for the separations
Actinides: for P&T and P&C scenarios, the elements considered for partitioning are the so-called
minor actinides (MAs): neptunium (Np), americium (Am), and curium (Cm), while for “new” fuel cycles
scenarios, uranium (U), plutonium (Pu) and the MAs are all concerned with partitioning/reprocessing
process development.
Fission products: for P&T and P&C scenarios, iodine (I), technetium (Tc) and caesium (Cs) are
the three main elements considered for partitioning. Some P&C scenarios also consider the
partitioning of caesium and strontium.
In Japan, the separation of the platinum group metals (PGMs) from nuclear wastes is also studied
for industrial uses of the separated PGMs.
2.1.2 Goals for the separations
The most important goals for the separations are:
• Minimisation of the long-term radiotoxic inventory of the wastes conditioned in
“conventional matrices”, e.g. in glasses (removal of LLRNs of MAs and FPs families).
• Minimisation of the heat load of the conditioned wastes (removal of 137 Cs and 90 Sr).
• More often, high purities of the separated LLRNs are required, for target or fuel fabrication
for subsequent transmutations of these LLRNs.
54

2.1.3 Consequences
Owing to the fact that efficient separation methods are needed with low losses of LLRNs and
high purities of the separated LLRNs, multi-stage processes are most often necessary.
2.2 Hydro processes for actinides and FPs partitioning
2.2.1 Examples of separation strategies
Examples of separation strategies are given for some countries and some research organisations.
2.2.1.1 Japan
• JNC
For FRs closed fuel cycle, JNC develops an integrated approach based on hydrometallurgical
steps including the: (i) dissolution of MOX FR spent fuels with an aqueous nitric acid solution,
(ii) iodine volatilisation, (iii) electrolytic extraction of technetium (Tc), palladium (Pd) and
selenium (Se), (iv) crystallisation of most of the uranium contained within the spent fuel
dissolution liquor in the form of hydrated uranyl nitrate, (v) single PUREX extraction step for
recovery of remaining U+Pu+Np, (vi) SETFICS process for Am and Cm partitioning.
• JAERI
JAERI proposed to separate Np and Tc during the implementation of the PUREX process. This
organisation develops also, since many years, the so-called four-group partitioning process for
the treatment of the wastes issuing the reprocessing by the PUREX process of UOX or MOX
LWR spent fuels. This partitioning process includes the following steps: (i) MAs partitioning
(Np, Am and Cm), (ii) Cs+Sr extraction, (iii) PGMs extraction. The remaining mixture of
wastes constitutes the 4 th category of elements of the initial mixture treated.
2.2.1.2 USA
The situation in the USA is peculiar because partitioning processes developed concern the treatment
of defence wastes in particular those accumulated at DOE’s Hanford site during the cold war. Several
processes were developed for the partitioning of radionuclides from the wastes: (i) TRUEX process for
transuranic extraction, (ii) SREX process for Sr removal and (iii) CSEX process for Cs extraction. It
should be also noted that in 1999, a Report named A Roadmap for Developing Accelerator
Transmutation of Wastes (ATW) Technology was published by the DOE [2], which considers the
possible treatment of the LWR spent fuels accumulated in the USA in order to separate: (i) U for final
disposition as low level waste and (ii) TRUs for burning in ATW systems. The processes considered for
these separations are: (i) the UREX process, which consists in a modified PUREX process aiming to
only extract U, (ii) pyrometallurgical partitioning process for TRU separation from the UREX wastes
and for the ATW fuel cycle.
2.2.1.3 France (CEA)
The strategy developed by the CEA for partitioning the nuclear wastes of LWR closed cycle
concerns 6 LLRNs to separate from the wastes: 3 MAs (Np, Am and Cm) and 3 FPs (I, Tc and Cs).
This strategy is based on the development of successive liquid-liquid extraction processes: (i) the
improved PUREX process for U, Pu, Np, I and Tc separations, (ii) the DIAMEX process for trivalent
Am+Cm extraction (FP lanthanides (III), Ln, are co-extracted), (iii) the SANEX process for
55

Am+Cm/Ln(III) separation, (iv) the SESAME process for Am/Cm separation, (v) the CALIX-
CROWN process for Cs separation.
In some organisations (e.g. in USA, Japan, Czech Republic or Russian Federation), instead of
developing a succession of separation processes for peculiar LLRNs, the integration of processes for
MAs and FPs extraction are studied. For example, the use of a solvent containing a mixture of cobalt
dicarbollide+dioxide of diphosphine allows the combined extraction of Cs+Sr+(Am+Cm)+Ln.
2.2.2 Minor actinides partitioning
One cycle processes
• DIDPA process (JAERI, Japan)
This process for MAs partitioning is based on the use of di-isodecylphosphoric acid (DIDPA).
The extraction mechanism is the following:
n+
M + n (HA) ⇔ M(HA ) + n H
(1)
2 2 n
+
The separation of the TRU elements is done by successive stripping from the loaded solvent,
including the use of diethylenetriaminopentaacetic acid (DTPA) complexing agent for
actinides(III)/Ln(III) separation (TALSPEAK like process, vide infra).
The DIDPA process was recently tested successfully in the BECKY hot-cell at NUCEF
(JAERI, Tokai-Mura).
Among the possible drawbacks of this process one can mention the: (i) required feed acid
adjustment, (ii) solvent degradation and its delicate clean-up, (iii) limited solvent loading with
metal ions.
• SETFICS (JNC, Japan)
The SETFICS process constitutes a modification of the TRUEX process (vide infra) based on
the use of the extractant di-isobutyl-phenyl-octylcarbamoylmethylphosphine oxide
(ΦC8H17P(O)CH2C(O)(i-C4H9)2 = CMPO)
The extraction mechanism is as follows:
n+
-
M + n NO + m CMPO ⇔ M(NO ) (CMPO)
(2)
3
3 n m
The separation of TRUs is done by successive stripping from the loaded solvent, including also
the use of DTPA for An(III)/Ln(III) separation. This process has not been tested yet with
genuine HAWs. The possible drawbacks of this process are: (i) the limited stripping efficiency,
(ii) the management of salts and DTPA containing effluents.
• PALADIN (CEA, France)
This process is based on the use of a mixture of extractants: a malonamide (DIAMEX process
extractant)+di-ethylhexylphosphoric acid (HDEHP), the extractant of the TALSPEAK process.
56

The extraction and separation mechanisms are the following:
Extraction:
3+ −
M + 3NO + 2 DIAM ⇔M(NO ) (DIAM)
(3)
3 3 3 2
An(III)/Ln(III) separation: done after contacting the loaded solvent with a pH adjusted aqueous
solution containing DTPA selective trivalent actinide complexing agent. For pH range HDEHP
is the extractant, while at the metal nitrate extraction step, carried out with acidic feeds 3 to
5 mol/L in nitric acid, trivalent An and Ln are extracted with the malonamide.
This process was recently successfully tested in the ATALANTE facility (CEA/Marcoule, France).
The possible drawbacks of this process are the: (i) necessity to use 2 extractants, (ii) need of pH
adjustment, (iii) co-extraction of numerous ions, (iv) solvent clean-up not yet defined.
Multi-cycle processes
• 1 st step: An+Ln co-extraction
TRUEX (USA, Japan, Russian Federation, Italy and India)
This process is based on the use of the CMPO extractant. This process was developed by
Horwitz (ANL) and Schulz (Hanford) in the USA in the 80s. The advantages of the TRUEX
process are the following: (i) it can extract An (and Ln) salts from acidic feeds, (ii) its efficiency
has been demonstrated with genuine HAWs, (iii) a large experience has been gained worldwide.
The main drawbacks of the TRUEX process are the: (i) necessity to use large concentration of
tri-n-butylphosphate (TBP) as solvent modifier added to the solvent to prevent third phase
formation, (ii) stripping of metal ions which are not so efficient, (iii) delicate solvent clean-up.
DIAMEX (France, Italy, Germany, Europe, Japan, USA and India)
This process is based on the use of a malonamide extractant. The main interests of the process
are: (i) An (and Ln) salts can be extracted from acidic feeds, (ii) its efficiency has been
demonstrated on genuine HAWs, (France, Europe), (iii) no secondary solid wastes generated
owing to the CHON character of the malonamide extractant. Its main drawback relies in the
partial co-extraction of palladium (Pd) and ruthenium (Ru) with the MAs.
A process based on a new type of diamide, a diglycolamide (DGA), which is a terdendate
ligand having better affinity for An(III) than the malonamide, is under development at JAERI
(Tokai, Japan).
TRPO (INET, Tsinghua University, China)
The TRPO process is based on the use of a mixture of tri-alkyl phosphine oxides (R3P(O), with
R = alkyl groups) as extractant. This process has been tested successfully in China with genuine
HAW. Its main drawbacks concern the necessity: (i) to adjust the feed acidity, (ii) to use a
concentrated nitric acid solution for An(III)+Ln(III) stripping, which complicates the
subsequent An(III)/Ln(III) partitioning step.
57

• 2 nd step: An(III)/Ln(III) separation
TALSPEAK and CTH processes
The TALSPEAK process, developed at ORNL (USA) in the sixties and then adapted (CTH
process) at Chalmers University, Göteborg, Sweden, can be considered as the reference process
for An(III)/Ln(III) group separation. It is based on the use of HDEHP as extractant and DTPA
as the selective An(III) complexing agent. The An(III)/Ln(III) separation is performed by the
selective stripping of An(III) from the HDEHP solvent loaded with the mixture of
An(III)+Ln(III) under the action of an aqueous solution containing DTPA and an
hydroxocarboxylic acid, like lactic, glycolic or citric acids. The advantages of this process are:
(i) the large experience gained worldwide, (ii) its good efficiency. Among the main drawbacks
one can cite: (i) the necessity to adjust the pH of the feed, (ii) the limited solvent loading of
metal ions, (iii) the difficult solvent clean-up.
SANEX concept (acidic S-bearing extractants)
– CYANEX 301 process (China, USA, Germany)
The CYANEX 301 extractant consists in a dialkyldithiophosphinic acid (R2PSSH, with
R = an alkyl group). Its use for An(III)/Ln(III) was first proposed by Zhu at Beijing (China)
in 1995. The main interest of the process relies in: (i) the large efficiency for An(III)/Ln(III)
separation, (ii) the fact that the process has been tested with genuine An(III)+Ln(III)
mixtures. Nevertheless, for an efficient use of this process the feed solution should be
adjusted to pH 3 to 5, which is not so easy to carried out industrially. Moreover, the solvent
clean-up is also a weak point.
– ALINA process (Germany)
To cope with the main drawbacks of the CYANEX 301 process mentioned above, Odoj and
Modolo at Jülich (Germany) proposed the use of a syngergistic mixture made of
bis(chlorophenyl)dithio-phosphinic acid ((ClΦ)2PSSH)+tri-n-octylphosphine oxide
(TOPO) to perform the An(III)/Ln(III) group separation. If the separation factors between
An(III) and Ln(III) are less than those observed with CYANEX 301, the concentration of
nitric acid in the feed can be as high as 1.5 mol/L, which makes the ALINA process more
attractive than the CYANEX 301 one. The ALINA process was successfully tested with
genuine wastes. The possible drawbacks of this process are: (i) the solvent clean-up process
not yet defined, (ii) the generation of P- and S-bearing wastes (from the degraded
extractants) which should be managed.
SANEX concept (neutral N-bearing extractants)
– BTPs (Germany, France, Europe)
After the discovery by Kolarik at FZ Karlsruhe (Germany) of the astonishing properties of
the bis-triazinyl-1,2,4-pyridines (BTPs) for An(III)/Ln(III) separation, a process was readily
developed and tested in the frame of the European so-called NEWPART project [3].
Successful hot tests were achieved both at the CEA/Marcoule and at the ITU in Karlsruhe
using the n-propyl-BTP. Large efficiency of the BTP process was obtained. One should
mention also that the feed of the n-propyl-BTP process can be acidic ([HNO3] = 1 mol/L).
Nevertheless, even if this system seems very promising, an instability of the n-propyl-BTP
58

extractant was observed. As a consequence, efforts are underway at the CEA to modify the
solvent formulation to suppress this major drawback.
– TMAHDPTZ+octanoïc acid (CEA, France)
A synergistic mixture made of the terdendate N-ligand, 2-(3,5,5-trimethylhexanoylamino)-
4,6-di-(pyridin-2-yl)-1,3,5-triazine (TMAHDPTZ), and octanoïc acid was developed at
CEA/Marcoule. A process flowsheet was defined and successfully tested with genuine
effluent with good efficiency. The main drawbacks of this process are: (i) the required pH
adjustment of the feed, (ii) the management of secondary wastes not yet defined.
• 3 rd step: Am/Cm separation
For this step, processes based on the selective oxidation of Am at the +VI or +V oxidation states
are developed, the curium remaining unchanged as Cm(III), allowing simple Am/Cm separation
processes to be defined.
SESAME process (CEA, France, Hitachi, Japan)
In strong oxidising conditions, Am can be oxidised from Am(III) to Am(VI). This can be done,
for example, by electrolysis in the presence of heteropolyanions (HPA) acting as catalyst. The
so-generated Am(VI) can be separated from Cm(III) by extraction, for example by TBP. This is
the principle of the so-called SESAME process developed at CEA/Marcoule. At Hitachi
(Hitachi city, Japan), oxidation of Am to Am(VI) is obtained by the use of ammonium
persulphate. Then, Am(VI) is extracted by TBP. The SESAME process exhibits a great
efficiency for Am/Cm separation. A large experience was obtained at the CEA at pilot scale
during the last twenty years with a SESAME like process (kg amounts of 241 Am were purified).
Nevertheless, the industrialisation of the process is faced with difficulties such as: (i) the
instability of Am(VI), (ii) the non-easiness to develop a multi-stage process, (iii) the generation
of secondary solid waste (made of HPA constituents).
Am(V) precipitation (JNC, Japan)
The selective precipitation of double carbonate of Am(V) and potassium (K) is one of the oldest
method for Am/Cm or Am/Ln separations, developed at the end of the 60’s in the USA. This
method requires the use a 2 mol/L K 2 CO 3 solution in which the mixture of Am(III) and Cm(III)
is dissolved. After chemical or electrochemical oxidation of Am(III) to Am(V), Am(V)
precipitates from the solution as the solid crystalline K5AmO2(CO3)3 nH2O, while Cm(III)
remains in solution. After filtration, Am is separated from Cm. This process: (i) is simple, (ii) is
selective for Am, (iii) has been largely used worldwide. The process main drawbacks are: (i) the
Am losses with Cm, which are not so low, (ii) the fact that it exists only one stage for the
process, (iii) the large amounts of secondary wastes generated.
2.2.3 Fission products partitioning
Iodine ( 129 I)
The separation of iodine is done just after the spent fuel dissolution step within the PUREX process.
Oxidation of iodide ion, I - , into elemental iodine (I2) induces its transfer to the dissolver off-gases
(DOGs) where iodine is recovered through DOGs basic washing. To recover most of the iodine spent
fuel inventory at that step, slight improvement of the efficiency of the transfer of iodine from the
dissolution liquor to the DOGs seems required.
59

Technetium ( 99 Tc)
The soluble fraction of Tc contained in the spent fuels exists in the dissolution liquor as Tc(VII)
(TcO4 - ). Its co-extraction with Zr(IV), then U(VI) and Pu(IV), by TBP is well known. So, for example,
the separation of the Tc soluble fraction is achieved through a solvent special scrubbing step in the
course of the implementation of the PUREX process at COGEMA La Hague reprocessing plants. If high
Tc partitioning yield is required, the main problem concerns the recovery of the Tc fraction that is
contained within the solid insoluble residues remaining after spent fuel dissolution. A special process is
required for this Tc recovery, which actually does not exist.
Caesium and strontium or caesium alone
Many processes were developed worldwide in this field, including the use of:
– Inorganic sorbents, like for the JAERI’s 4 group partitioning process.
– Crown-ether extractants, like for the SREX and CSEX processes developed in the USA (ANL).
– Cobalt dicarbollide extractants, as developed in Czech Republic, Russia and Western Europe.
– Calix-crown extractants, as developed in France, Western Europe and in the USA.
Most of these processes were successfully tested with radioactive effluents.
2.3. Pyro processes for actinide partitioning [4]
2.3.1 Selected media and possible separation techniques
Selected media
Most of the “pyro” processes developed so-far are based on the use of one or two of the following
high temperature liquid phases:
• Fused salts. The most popular fused salts studied are:
– Molten chloride eutectic, such as LiCl+KCl.
– Molten fluoride eutectic, such as LiF+CaF2.
• Fused metal, such as Cd, Bi, Al, etc.
Separation techniques
To partition the actinides contained within the fused salt baths, three main techniques are studied
and developed:
• Actinide electrodeposition on solid (pyrographite or metal) or liquid metal cathodes.
• Liquid-liquid extraction of actinide between fused salt bath and a metal bath containing a
reductive metal solute (Li for example).
• Actinide oxide precipitation from the fused salt under the proper control of the oxygen
thermodynamic activity within the salt bath.
60

2.3.2 Examples of strategies and “pyro” processes
2.3.2.1 USA (ANL, Chicago)
A “pyro” process was developed at ANL in relation with the treatment of FR metallic fuels (EBR II’s
type) for stabilisation of these Na bonded fuels. The aims of the process is limited. It consists in the
separation of the spent fuel into three major fluxes: (i) most of the uranium as a low level waste,
(ii) cladding+noble metals+Zr as metallic waste, (iii) TRUs+FPs+Na+salt incorporation into a zeolithe
matrix in order to obtain a ceramic waste after hot pressing. The key step consists, after the oxidative
dissolution of the spent metallic fuel in LiCl+KCl eutectic bath at 500°C, into the separation of most of the
uranium by electrorefining on a solid cathode. A demonstration campaign involving the treatment of
100 core assemblies (0.4 ton of spent HEU) and 25 blanket assemblies (1.2 tons of spent depleted U) was
successfully carried out at Argonne West in the recent years. License for pyroprocessing the whole EBR II
spent fuel inventory was obtained recently.
2.3.2.2 Russian Federation (RIAR, Dimitrovgrad)
The pyro-process developed at RIAR concerns the treatment of spent oxide fuels (UOX and MOX)
in order to recover U and Pu for MOX fuel re-fabrication by the vibro-compaction process.
The spent oxide fuel is dissolved by chloration in a Li, Na, K, Cs chloride fused salt bath at 650-
700°C. Separation of U, Pu or mixture of U+Pu from the salt bath can be obtained by electrodeposition
or precipitation. For example:
• U can be separated as UO2 (which is a good electric conductor) by electrodeposition on a
cathode made of pyrographite, while chlorine gas is generated at the anode.
• As PuO 2 is a bad electric conductor, it cannot be electrodeposited on solid cathode. But PuO 2
can be selectively separated by precipitation after bubbling a mixture of Cl2+O2 gases into the
fused salt bath.
• Under the addition of a mixture of Cl2+O2 gases into the fused salt bath, which stabilises Pu
as oxychlorides, electrolysis generates a deposit of (U,Pu)O2 onto the pyrographite cathode
while chlorine gas evolves at the anode.
An important experience with spent fuel pyroprocessing has been obtained at RIAR with the
treatment of:
• 3.3 kg of UO2 spent fuel (1% burn-up) from the VK-50 reactor, done in 1968.
• 2.5 kg of UO2 spent fuel (7.7% burn-up) from the BOR-50 reactor, done in 1972-73.
• 4.1 kg of (U,Pu)O2 spent fuel (4.7% burn-up) from the BN-350 reactor, done in 1991.
• 3.5 kg of (U,Pu)O2 spent fuel (21-24% burn-up) from the BOR-60 reactor, done in 1995.
2.3.2.3 Japan
• CRIEPI
The “pyro” process developed at CRIEPI concerns both the treatment of spent fuels from
LWRs (oxide fuels) or FRs (metallic fuels) and the partitioning of TRU elements from the
wastes issuing the reprocessing of spent LWR fuels by the PUREX process. The fused salt
selected is the LiCl+KCl eutectic in which the spent fuels or the oxides of high active wastes
61

are dissolved by a carbo-chloration technique. After dissolution, U can be electrodeposited as a
metal on a solid cathode, then the TRUs can be recovered by electrolysis using a liquid
cadmium or bismuth cathode. CRIEPI is also studying the partitioning of MAs by liquid-liquid
extraction using a LiCl+KCl salt bath and Cd or Bi metallic solvents containing Li as a
reducing agent. In this case, liquid-liquid extraction corresponds to the reductive transfer of a
metal from the salt bath, where it exists as M n+ cation, to the metal solvent, where M exists as a
M 0 solute.
The equation of the extraction reaction can be written as follows:
n+
0
0
+
M + n Li ⇔ M + n Li
(4)
(salt)
(Cd or Bi metal phase)
(Cd or Bi metal phase)
Large expertise has been gained by CRIEPI in this field but only with surrogates of actinides. A
joint CRIEPI-ITU programme is under way to test the process with actinides.
• JAERI
JAERI is studying pyroprocessing for the possible treatment of nitride, oxide or metallic spent
fuels in order to prepare nitride fuels enriched with 15 N for FRs. After dissolution of the spent
fuels into a LiCl+KCl eutectic salt bath, the actinides will be electrodeposited on solid or liquid
(Cd) cathodes. The recovered actinide metals will then be converted into actinides nitrides after
their dissolution in liquid cadmium. The nitruration agent will be N2 or Li3N.
• JNC
JNC is also engaged in the development of pyro processes aiming to reprocess FR spent fuels.
The method selected are similar to those studied by CRIEPI: (i) choice of LiCl+KCl eutectic bath,
(ii) electrodeposition method, (iii) liquid-liquid extraction between salt bath and liquid Cd metal.
2.3.2.4 France
Two years ago, the CEA has launched a programme dedicated to the partitioning of MAs by “pyro”
processing. A research team was created at CEA/Marcoule and special hot facilities have been created. The
programme selected is rather wide. It will consider both chloride and fluoride melts and the most important
separation techniques known to be effective in “pyroprocessing”, i.e. (i) electrodeposition, (ii) oxide
precipitation, (iii) liquid-liquid extraction between fused salt bath and a metallic solvent. The results
obtained to-date concern mainly the basic understanding of the chemistry of actinides (U, Pu and Am) in
solution in the fused melts. Process developments are also underway and active tests on irradiated objects
are foreseen to be done before 2005.
2.3.2.5 Czech Republic
At Rez Institute, Czech scientists are developing a process based on the dual use of actinide
hexafluoride volatilisation and pyroprocessing of the wastes from a fluoride melt. This research
programme is connected with the interest of Czech Republic for the development of the molten salt
reactor (MSR) technology. Facilities are under construction at Rez Institute for testing the pyroprocesses.
(salt)
62

3. Conclusions and perspectives
3.1 Conclusions
Numerous concepts have been consolidated or newly developed during the last few years, both in
“Hydro” and “Pyro” processing of HAWs or spent fuels and targets for “new” nuclear systems.
Tests on “real objects” were carried out successfully in several countries, including the EBR II
demonstration test at Argonne-West (UA) on pyroprocessing of FRs spent fuels.
In the domain of “Hydro”, blooming of concepts is observed. Multi-step processes look
promising but most of the systems developed so far appear complex. Efforts to simplify the processes
seem required.
In the domain of “Pyro”, strong consolidations of “old concepts” were obtained, including
fluoride volatilisation.
3.2 Perspectives
• Hydro
It seems important to work in order to increase the “simplicity” and “compacity” of the MAs
and LLFPs separation processes. Some routes for improvement can be proposed:
– One cycle process.
– Consideration of High Active Concentrates instead of High Active Raffinates for process
development (large volume reduction factor).
– Integration of MAs and LLFPs separation processes.
– Consideration of possibly new LLFPs for partitioning.
– Maintaining alive the “CHON principle” (minimisation of secondary solid wastes).
• Pyro
Directions for improving the processes appear to be:
– Minimisation of TRU losses in wastes and increase of the purities of the separated.
– Actinides which can be obtained through the combined use of several separation
techniques and multi-stage techniques.
– The waste problem, which is mostly corrosion related owing to the aggressive.
– Character of the media and the high process temperatures, needs to be precisely estimated.
– Consideration of the possible separation of LLFPs.
• Collaborations
It seems a pressing necessity to maintain, or best to increase, the collaborations in this complex
field at:
– National levels: maintain or create network(s) between academic and applied research
bodies. As example in France it exist two networks working under the auspices of the
63

December 1991 Nuclear Waste Act: the so-called PRACTIS and NOMADE Groupes de
Recherches.
– Bi-national levels: numerous collaborations exist, e.g. CRIEPI-ANL, CEA-JNC, CEA-
JAERI, etc.
– Regional level. As example at the European level it exists common works partly financed
by the EU, e.g. the PARTNEW, CALIXPART and PYROREP programmes within the
5 th FWP of EU (2000-2003). The role of ITU at Karlsruhe is also very important for
European and wider collaborations,
– At the International level, the roles of OECD/NEA for Workshops and Working Parties
managements and also of IAEA appear essential.
So, within a few years, one predicts that a large array of robust “hydro” and “pyro” processes will
be available for the definition of new scenarios for the management of nuclear wastes generated
through LWRs and FRs closed fuel cycles, but also for the fuel cycles of futuristic nuclear systems,
such as the MSRs or the ADSs.
REFERENCES
[1] Proceedings of EURADWASTES Conference, Luxembourg, November 1999.
[2] Anonymous, DOE/RW-0519, (1999), A Roadmap for Developing Accelerator Transmutation of
Waste (ATW) Technology. A Report to Congress.
[3] C. Madic; M.J. Hudson; J.O. Liljenzin; J.P. Glatz; R. Nannicini, A. Facchini; Z. Kolarik and
R. Odoj, (2000), New Separation Techniques for Minor Actinides, EUR 19149 EN.
[4] Pyrochemical Separations, Workshop Proceedings, Avignon, France, 14-16 March 2000,
OECD/NEA.
64

SESSION IV
Basic Physics, Materials and Fuels
Chairs: S. Pilate (BN) – H. Takano (JAERI)
_____________________
SUMMARY
Session IV was subdivided into three parts devoted respectively to basic physics, materials, fuels
and targets.
1. Basic physics
Nuclear data measurements performed or supported by JNC in Japan concern fission and capture
cross-sections for Am, Np and long-lived fission product nuclides. In the case of 99 Tc, activation
measurements revealed significantly different from previous results. It was noted that a co-operation
with ORNL is planned.
Residue production in spallation reactions is being studied by researchers from Spain, Germany
and France. New experiments are made at Saturne and NESSI (GSI) to identify the numerous isotopes
created; an accuracy of 10% on their production is aimed at.
The experimental programme MUSE in MASURCA at CEA-Cadarache, launched in 1995, now
continues with 16 different partners and with EC funding. In the present MUSE-4 configuration, a
strong pulsed neutron source is fed from the coupled accelerator GENEPI. The k-value will
progressively be decreased to 0.95. After the Na-cooled core, a gas-cooled core will be mocked-up.
In parallel, a complementary programme SAD will be run at Dubna (RF), using real spallation
sources produced by a synchrotron.
A simplified version of MUSE-4 in RZ geometry is also proposed as a benchmark; the
calculations already available indicate differences of e.g. 0.8% on the k-level.
Results of ADS benchmark calculations have been gathered by OECD/NEA and PSI/CEA. NEA
had already organised a first, preliminary ADS exercise in 1994. This follow-up exercise started in
1999. Pb-Bi is retained as ADS target and reactor coolant. Two cores are considered: at start-up and at
equilibrium. The external source is pre-defined. Significant discrepancies can be observed among the
65

7 solutions, obtained with the 3 basic data files ENDF/B-VI, JEF 2.2 and JENDL 3.2. The k-values
differ by as much as 3% dk, and this is not only an effect of basic data libraries.
A next exercise is planned on a transient ADS benchmark (beam trip). It was pointed out that the
partners should give more details on their data processing.
2. Materials
The corrosion of stainless steels in a Pb-Bi circuit has been studied by CIEMAT for temperatures
ranging from 400°C (cold leg) to 550°C (hot leg). The loop was made of austenitic steel while test
samples were made of 2 martensitic steels. The oxide layer formation was recorded after different
operation times up to 3 000 hours. A gas with 10 ppm O2 was bubbling in the hot leg.
The oxide protection layer grew with time. The coolant dissolved some elements of the steel,
mainly nickel.
Such experiments, crucial for the use of Pb-Bi coolants, should be made again, provided the exact
O2 activity be well monitored.
Two other papers were devoted:
• A Russian one, to the production of residues from spallation (as above).
• A Korean one, to thermal and stress analysis for the HYPER target; HYPER is the accelerator
driven system developed by KAERI, based on the use of Pb-Bi coolant (another KAERI paper
also considered the problem of transmuting 99 Tc and 129 I in HYPER, what is very difficult).
3. Fuels and targets
While a paper by industrial companies stressed the interest to develop practical concepts of Am
targets, to be placed in special, moderated positions in fast reactor cores, two very interesting papers
by ITU and CEA described the experimental programmes devoted to new fuels and targets in Europe.
Both are complementary, and the EC sponsorship stimulates a vast European collaboration.
Different promising concepts will be examined and tested, as for example IMMOX (Inert Matrix
MOX), THOMOX (where ThO2 is a “quasi-inert matrix”), MATINA (with macromasses instead of
micro-dispersion), ECRIX/CAMIX/COCHIX to be loaded in Phénix.
Two laboratories for minor actinides have recently been built in Europe, the one at Marcoule
(ATALANTE) and the other one at ITU Karlsruhe, allowing handling Am in the kg range.
In Japan, JAERI also builds a new facility TRU-HITEC, for high-temperature chemistry of Am
and Cm. Available at Tokai in 2002, it will allow to handle dozens of grams of Am and Np, in
addition to Pu. The research is centered there on nitride fuel with inert matrix, and on pyro-processing
(as described in a JAERI paper on ADS transmutation in Session V).
The set-up of these three laboratories for minor actinides handling is a concrete result of the
“decade of revival” for transmutation research, which had been illustrated in the bright overview paper
introducing Session IV, given by Mr. M. Salvatores.
The coming decade should now help identify the most efficient transmutation options thanks to
irradiation experiments.
66

TRANSMUTATION: A DECADE OF REVIVAL
ISSUES, RELEVANT EXPERIMENTS AND PERSPECTIVES
OVERVIEW PAPER
M. Salvatores
CEA, Nuclear Reactor Directorate, Cadarache, France
Abstract
For more than a decade, transmutation studies have been again a topic of wide interest and have
triggered numerous international activities, like bilateral/multilateral collaborations, information
exchanges, state-of-the-art reports, conferences, but also some co-ordinated programmes and
experiments.
It is legitimate to ask at this point, whether transmutation studies are still “fashionable” and why; what
is known, what has been done and what should be done.
Since the motivations of national programmes are often different, due to a different context, we will
take for granted that transmutation is generally seen as an option for the back-end of the fuel cycle in
order to reduce the burden of potential geological storages of radioactive wastes (whatever their
nature).
Finally, we also acknowledge the fact that some highly respected scientists have at several occasions
during this decade expressed their doubts about the value of the transmutation option. A typical
example is the position expressed by Pigford and Rasmussen, reporting the results of a study for the
US National Research Council (IAEA-TECDOC-990, 2/12/1997).
67

1. Introduction
To give a state-of-the-art of the transmutation studies, one could make use of international
publications or proceedings of the specialised conferences that have been mushrooming in this field.
Of course a significant example is the OECD/NEA state-of-the-art report published in 1999: “Actinide
and Fission Product Partitioning and Transmutation. Status and Assessment Report”.
We will limit our analysis to a few points that we consider of special relevance. Successively, we
will review some ongoing research and experimental validation studies, in order to provide a list of
relevant expected results, which should have impact to shape (or re-shape) future programmes.
In this perspective, we will also indicate which are in our opinion the “missing” experiments or
studies, the absence of which could jeopardise the process of decision making.
The nuclear energy “environment” is a changing one, and it is of interest to review some
activities/studies/concept proposals, which are not strictly speaking in the “transmutation” domain, but
which could have an impact on the conclusions which could be drawn on the potential role of
transmutation.
Finally we will attempt to summarise a list of open questions and an analysis of possible
(re)orientation of priorities. Of course, this paper does not deal with chemistry issues, but rather with
reactor and fuel cycle technology, since the physics of transmutation is today well understood (see for
example [1]).
2. Where are we?
Transmutation is of course an R&D endeavour. The potential “customers” of such R&D can be
found more in the society at large and its political representative bodies than in industry. By the way, it
is not evident that even a fundamental feature of transmutation (i.e. the need to reprocess the fuel) is
clearly understood in that context.
In so far as customers, one should not forget that utilities look probably with some apprehension
and scepticism to studies which could offer options for the back-end of the fuel cycle but which could
potentially have a non negligible impact on the cost of the electricity generation, without a clear
definition of criteria to evaluate costs versus benefits, and in a frame of a highly competitive
environment. In fact, utilities are ready to contribute to the R&D studies but to establish sound figures
for induced or direct costs on the kWh!
The nuclear reactor and fuel cycle industry has an obvious interest to be well informed of the
transmutation R&D issues and results, but does not finance these activities other than marginally. This
oversimplified analysis is only meant to make clear that there is an inherent difficulty to evaluate the
real status of the research in the transmutation field with respect to the potential utilisation of the
technology in a reasonable time horizon within stated performance and cost criteria.
In view of this difficulty, we have preferred to single out some specific scientific results, which
could characterise our present understanding of transmutation and its use as an option for the fuel cycle.
68

2.1 The IFR concept and the homogeneous recycling
The IFR concept [2] is still the most outstanding example of an “inherently” transmutation
concept in the so-called “homogeneous” recycling mode (see Figure 1 and [3]). The IFR concept can
be seen as an energy producing system capable to recycle Pu and minor actinides (MA), to reach
equilibrium, both stabilising the Pu and MA mass flows, and sending to the wastes only a very small
fraction of the radiotoxic isotopes. This fraction is of the order of 0.1% or less, according to the
announced performances of the pyrochemical process involved, which has still to be demonstrated at
large scale in the frame of the transmutation application.
Figure 1. Pu and MA management in P/T strategies 6
P&T strategies
Pu and MA handled
together
Pu and MA handled
separately
Open cycle
Irradiated fuel
to storage
Dedicated cores
Pu + MA fuel w/o
fertile (critical or
preferably, ADS)
Homogeneous recycling
− in LWRs
− in FRs (e.g. IFR
concept),
− molten salts reactors
Heterogeneous recycling
(once-through or multirecycling)
Pu and MA in the same
core :
− Pu as standard fuel
for the core
− MA targets in S/A at
the core periphery
Dedicated cores
Two separate fuel cycle
strata :
− Pu in main stratum
− MA (with some Pu) as
core fuel in burner
reactors of a separate
fuel cycle stratum
(critical or preferably,
ADS)
The appealing aspects of the IFR concept in the frame of transmutation are:
• The concept is mainly designed to produce energy, making an optimised use of resources and
using a robust reactor and fuel cycle layout.
• The fuel cycle does not imply the separation of Pu and MA.
• The concept can accommodate in principle several options in terms of reactor size, reactor
coolant, waste-forms, etc.
In general, the homogeneous recycling has equivalent performances for whatever the type of fuel
in the fast reactor. In fact, if the losses at reprocessing are assumed to be of the order of 0.1%, the
homogeneous recycling allows to reach a reduction of the potential radiotoxicity with respect to the
open cycle scenario of a factor of 200 and more, and this over all the time scale (10 2 → 10 6 years) [4].
However, the consequences on the fuel cycle have to be taken into account (see Table 1) and their
impact evaluated.
6 If LLFP management is required, they can in principle be handled in the different scenarios as targets to be
irradiated at the periphery of the different core types.
69

Table 1. Consequences on the fuel cycle of MA recycling in FR a) .
Variation expressed as ratio with respect to the corresponding values
for the reference case: PWR-MOX Pu content: 12%, taken as 1.
Fabrication
Reprocessing b)
Activity heat due
to:
α
β
γ
Neutron source
Activity heat due
to:
α
β
γ
Neutron source
PWR – MOX
12%
a) Oxide fuel, EFR type.
b) 5 years cooling time.
c) Effect due mainly to 252 Cf.
d) Heterogeneous recycling total fission rate: 90% (see text).
1
1
1
1
1
1
1
1
1
1
FR: Multirecycling
of Pu, Am, Cm
0.1
0.5
0.2
1.5
30
0.1
0.2
0.5
0.2
0.4
FR: Multi-recycling
of Pu, once-through
irradiation of
Am + Cm targets d)
0.1
1.7
0.4
8.7
104
0.10
0.11
0.09
0.06
245 c)
2.2 Heterogeneous recycling and its potential limitations
An option has been explored, mainly in Europe and in particular at CEA in France [5] and at JNC
in Japan [6], to perform the transmutation of MA in the form of targets to be loaded in critical cores of
a “standard” type. The mode of recycling has been called “heterogeneous” (see Figure 1), the potential
advantage being to concentrate in a specific fuel cycle the handling of a reduced inventory of MA
(separated from Plutonium). The major obstacles to that approach are:
• The very high irradiation doses needed to fission a significant amount of MA (which implies
very high damage rates).
• The need to separate Am and Cm from Pu and to keep them (Am and Cm) together, in order
to reach high values (~30) for the radio-toxicity reduction [3].
• The need to load the MA targets in a very large fraction (∼30 ÷ 50%) of the reactor park,
possibly made of fast reactors, due to their favourable characteristics for this mode of
recycling (high fluxes, which can be easily tailored in energy to increase fission rates).
• Consequences on the power distributions and their evolution with time.
In any case, the consequences on the fuel cycle are relevant, if one wants to reach a factor of
radiotoxicity reduction of ~30 ÷ 40 (see Table 1 and [7]).
70

2.3 Dedicated systems
Making again reference to the scheme of Figure 1, a possible approach to keep the MA fuel cycle
and the transmutation technology separated from the electricity production, is the one which calls for
the use of “dedicated” cores, where the fuel is heavily (>30%) loaded with MA, the rest being,
e.g. plutonium (the ratio Pu/(Pu + MA) being

exception of activities started at CRIEPI (Japan) and now extended to TUI-Karlsruhe. These new
activities concern also the fuel reprocessing (see [12] at this conference). In particular, an experiment
(METAPHIX) is planned, in order to irradiate metal fuel pins, loaded with MA and rare earths (RE)
([13] and Figure 2).
Figure 2. Arrangement of fuel pins in a rig
for the METAPHIX experiment (CRIEPI-TUI) (from [13])
Nine metallic fuel pins are prepared for the METAPHIX irradiation study: three pins of UPuZr,
three pins of UPuZr-MA2%-RE2%, three pins of UPuZr-MA5%, and UPuZr-MA5%-RE5%. They are
planned to be inserted in the positions 1, 2 and 3, respectively, in the rig. Three rigs consisting of three
sample metallic fuel pins and sixteen driver oxide pins will be prepared. Three rigs correspond to three
different values of the burn-up: 1.5, 5 and >10%, respectively.
As far as homogeneous recycling in standard oxide fuels, some experimental knowledge has been
obtained with the SUPERFACT experiment [14]. More will come from experimental programmes
conceived at JNC and which should take place in JOYO beyond 2003. On the contrary, no experience
exists on MA-loaded oxide fuels in standard light water reactors.
Finally, it has to be noted that the present revival of interest for pyroprocessing techniques, has
been largely motivated by the relevance of these techniques to handle “hot” fuels, like those which are
foreseen for transmutation. It has to be noticed that a modest but significant programme, PYROREP,
has been launched as an EU contract for the 5 th FWP.
3.2 Heterogeneous recycling
Apart from conceptual studies at JNC and CEA, experimental activities have been launched in
Europe (e.g. the EFTTRA collaboration) and some useful indications have been gathered [15]. The
EFTTRA-T4 and T4-bis experiments concern 241 Am, at a 12% volume fraction, inside a matrix of
MgAl2O4, for a maximum fission rate of 28%. The swelling due to the decay of the 242 Cm produced by
neutron capture, has been relevant, and triggered further research on the form of inert matrix/actinide
fabrication (e.g. micro-dispersion versus macro-dispersion). It is worth to notice that experiments
performed up to now, did not cover the presence of 243 Am and the presence of Cm.
Further experiments are planned in France, and in particular the ECRIX experiment, which
should take place in PHENIX and the CAMIX and COCHIX experiments (J.C. Garnier, CEA, Private
communication), also planned in PHENIX. The CAMIX experiment will provide information on
“micro-dispersion” of a (Am, Zr, Y)O2-x compound in MgO and COCHIX information on the same
72

compound “macro-dispersed” in MgO or (Zr0.6 Y0.4)O1.8. All 3 experiments are planned to reach a
fission rate equivalent to 30 at%.
A significant global experiment is presently worked-out in the frame of the collaboration between
MINATOM (Russia) and CEA (France) with the participation of FZK and TUI-Karlsruhe, as partners
of CEA. In this experiment (AMBOINE), Am targets AmO2+UO2 and AmO2+MgO should be
fabricated at RIAR according to the VIPAC process. These targets should be irradiated in BOR-60 and
reprocessed by pyroprocess after irradiation again at RIAR, providing in that way a full validation of
the whole fabrication – irradiation – reprocessing cycle for CERCER targets (S. Pillon, CEA – Private
communication).
3.3 Dedicated systems
3.3.1 Fuels for dedicated systems
For both critical and sub-critical dedicated cores, the major issue in the path towards feasibility
demonstration, is the fuel development. Many candidates have been considered (see for example
Table 2), but limited experimental work has been done, in order to characterise the basic properties of
these potential fuels, their fabrication processes and their behaviour under irradiation.
Table 2. Dedicated Pu + MA fuels (adapted from [16])
Metal fuels
Oxide fuels
Nitride fuels
Composite fuels:
the role of Zr
Coated particle fuels
− Need to improve thermal properties ⇒ add non-fissile metal with high
melting point (e.g. Zr) ⇒ Pu-MA-Zr alloy.
− However: mutual solubility of Np and Zr?
− Mixed transmutation oxides as a logical extension of MOX.
− However: smaller margin to melting (low thermal conductivity).
− Good thermal behaviour.
− However: need to enrich in 15 N.
− Lower stability against decomposition at high temperatures.
Ad-hoc “tailoring”:
− MgO + (Zr, An)O2-X (CERAMIC-CERAMIC).
− Zr + (Zr, An)O2-X (CERAMIC-METALLIC).
− Zr + (An, Zr) alloy (METAL-METAL).
However, fabrication can be difficult (also: size and distribution of the
disperse actinide phase).
Special form of composite fuels. However in the case of fast spectra, little
is known on potential candidates (TiN?).
⇒ A generic problem: the high He production under irradiation.
A common feature for these fuels is to be fertile-free, or, at least, “U-free”, since Th is sometimes
considered as an acceptable support, in particular for strategies that promote the replacement of the
U-cycle with the Th-cycle.
Well-structured programmes for fuel development are missing in practically all the major
transmutation programmes. A significant exception is the JAERI programme, focused on nitride fuels.
73

The CONFIRM project, sponsored by the EU 5 th FWP is also devoted to nitride fuels ((Pu, Zr) N
and (Am, Zr) N). The project aims to the fabrication, characterisation and irradiation of these fuels,
and addresses also the issue of 15 N enrichment.
Finally, we recall the initiative of the European Technical Working Group (TWG) on ADS,
chaired by Professor Rubbia, which has set up an ad-hoc task force an Fuel Fabrication and
Processing, in order to produce a state of the art report and an agreed work plan in the frame of a roadmapping
towards ADS deployment [16].
As far as reprocessing, the situation is, obviously, not fully satisfactory either. The use of
dedicated fuels imposes their reprocessing in all considering schemes. Again, the programme proposed
by JAERI, includes laboratory scale experiments of reprocessing, together with a flow diagram of a
process to enrich in 15 N the fuel [17].
The present European efforts are reviewed in [12,16]. A programme for pyrochemistry
development is also being set up in France at CEA. The rationale for it can be found in the report:
“Assessment of Pyrochemical Processes for Separation/Transmutation Strategies: Proposed Areas of
Research – CEA/PG – DRRV/Dir/00-92, March 2000”.
3.3.2 ADS systems
A special case is the research activity in the ADS domain.
In the last two years, relevant initiatives have taken place. The ATW Roadmapping in the US
[18], should give rise to a focused programme in very near future. The joint KEK-JAERI project, has
given a place to ADS development in Japan, in the frame of a multipurpose facility [19].
In France, the GEDEON programme [20] has gathered a large community of physicists around
the basic physics items of research for ADS (nuclear data, spallation physics, sub-critical core
neutronics, materials, but also pyrochemistry, molten salts, thorium and system studies).
In Europe the Technical Working group mentioned above has been established. In that frame two
concepts for ADS are being studied (see Figure 3) and a rationale is emerging for a “step-by-step”
validation and demonstration of the ADS concept (see Figure 4) and its waste transmutation potential,
in the frame of a specific road-mapping, which is being finalised at present. Few comments will be
made in what follows, on some ongoing experimental steps, like the MEGAPIE project and the MUSE
programme. Finally, the European Union is supporting a number of projects, in the frame of the 5 th
R&D Framework Programme.
The issues and programmes related to high power proton accelerators (HPPA), although essential,
will not be dealt with here. We only remind the R&D work that has been initiated on the topic of
“accelerator reliability” and which has been the subject of two NEA Nuclear Science Committee
workshops (Mito, 1998, Aix-en-Provence, 1999).
74

Figure 3. Sketch of ADS, liquid metal cooled (right) and gas cooled (left) (not to scale),
representative of the European EADS proposals (ANSALDO and FRAMATOME).
Potentially the same fuel assembly (e.g. SNR-300 S/A with MOX fuel).
75

Figure 4. A step-by-step approach to the validation and demonstration of the ADS concept
HIGH INTENSITY
ACCELERATOR
A
Systems to be validated:
A B C
⇒
Experimental ADS
(with irradiation
capability
A B D
⇒
Experimental ATW
(transmutation
demonstration
A
: i p ≥ 5 mA, Ep ~ 600 - 1 000
B
C
D
: P ≥ 1 MWt
: k eff
~ 0.90 ÷; P : 40 ÷ 100 MW th
: MA-dominated fuels in the same core as
C
TARGET
(Ex. Pb/Bi)
B
External source
MULTIPLYING MEDIUM
WITH STANDARD FUEL
C
WITH DEDICATED FUEL
D
A path towards validation:
A
: the IPH project (High Intensity Proton Injector) or TRASCO
program and follow-up programs (e.g. superconducting cavities)
A
+
B
: the MEGAPIE project (with “known” A )
B
D
+ C : the MUSE programme (with “known” B )
Next steps:
A
+
B
A B C
+ +
: spallation source (1 ÷ 5 MW th )
: experimental ADS (40 ÷ 100 MW th ; k eff ~ 0.90 ÷ 0.98) with standard fuel
(e.g. SNR-300 MOX fuel) and high flux (≈ 10 15 n/cm 2 .s) (time horizon ≈ 2015)
A
+ B +
D
: experimental ATW (time horizon ≈ 2025)
76

3.3.2.1 The MEGAPIE project [40]
MEGAPIE is an international (CEA, PSI, FZK, CNRS France, ENEA, SCK•CEN. JAERI will
join soon) experiment to be carried out in the SINQ target location at the Paul Scherrer Institute in
Switzerland and aims at demonstrating the safe operation of a liquid metal target at a beam power in
the region of 1 MW. The minimum design service life will be 1 year (6 000 mAh).
The target material will be the PbBi eutectic mixture. Existing facilities and equipment at PSI will
be used to the largest possible extent. In fact, the MEGAPIE target will be used in the existing target
block of SINQ.
A vertical cut through this target block and parts of the proton beam line is shown in Figure 5.
Figure 5. Vertical cut through the target block
and part of the proton beam transport line of SINQ
The target’s outer dimensions must be such that it fits into the target position of the SINQ facility,
the existing target exchange flask including its contamination protection devices and the existing
target storage positions.
The target will be designed for 1 MW of beam power at a proton energy of 575 MeV, i.e. a total
beam current of ip = 1.74 mA.
It is also important to realise that the stability of beam delivery cannot be guaranteed at all times.
The MEGAPIE heat removal system must be able to cope with frequent short beam trips and
occasional unstable operation i.e. up to days long shutdown periods.
A sketch of the MEGAPIE spallation target is given in Figure 6.
77

The major objectives of the MEGAPIE initiative are:
• Full feasibility demonstration of a spallation target system.
• Evaluation of radiation and damage effects of structures and beam window in a realistic
spallation spectrum.
• Effectiveness of the window cooling under realistic conditions.
• Liquid metal/metal interactions under radiation and stress.
• Post irradiation examinations (PIE).
• Demonstration of decommissioning.
Figure 6. Sketch of the 1 MW exploratory liquid lead-bismuth spallation target MEGAPIE
It has to be reminded that two EU contracts, established in the frame of the 5 th FWP [21], SPIRE
(material irradiation) and TECLA (physico-chemical properties of lead alloys: corrosion...), provide a
relevant R&D back-up to the MEGAPIE project. Moreover, experimental laboratories have been
launched in support of these activities (like the KALLA laboratory in FZK-Karlsruhe) or are reoriented
(like the ENEA laboratory in Brasimone: the CIRCE loop).
78

3.3.2.2 The MUSE experiments
The MUSE experiments, launched in 1995 [22], provide a simulation of the neutronics of a
source-driven sub-critical system, using the physics characteristics of the separation of the effects due
to the presence of an external neutron source from the effects of the neutron multiplication. In fact for
a wide range of sub-criticality values (e.g. keff: 0.9 ÷ 0.99) the space dependence of the energy
distribution of the source neutrons is quickly (in approximately one mean free path) replaced by the
fission-dominated neutron energy distribution.
In practice, external known neutron sources have been introduced at the centre of a sub-critical
configuration in the MASURCA reactor. The more recent of these experiments is made of a deuton
accelerator and a target (deuterium or tritium) at the centre of a configuration, where actual target
materials (like lead) are loaded, to provide the neutron diffusion representative of an actual spallation
target (see Figure 7 and [23]). The neutrons issued from (d,d) and (d,t) reactions provide a reasonable
simulation of the spallation neutrons, in terms of energy distribution (see Figure 8).
Static (e.g. flux distributions, spectrum indexes, importance of source neutrons) and kinetic
parameters (e.g. time dependence of neutron population, effective delayed neutron fraction, with
appropriate weighting, etc.) have been or will be measured (see [23]). Sub-criticality itself, is
measured by static and dynamic techniques.
Finally, the proposed experiment MUSE-4 start-up procedure i.e. 1) critical configuration with
accelerator hole but no beam, 2) sub-critical configuration with accelerator hole but no beam, 3) same,
but with beam on, allows to establish a precise reactivity scale in step 1, which can be used both to
calibrate eventual control rods and to measure in a standard way (e.g. with the modified source
multiplication, MSM, method) the level of sub-criticality of steps 2 and 3.
The MUSE-4 experiments, described in a separate paper at this conference [23], are also partly
supported by an EU contract for the 5 th FWP.
3.3.2.3 Streamlining basic physics experiments
The ADS research development has also motivated a significant number of experimental
activities in the field of spallation physics and nuclear data measurement and evaluation (mainly
actinides and LLFP). Examples will be found in papers at this conference.
A major experiment takes place at GSI, defined in order to gather much needed information on
spallation product yields and distributions in (A, Z) [24].
Also in this area, the EU supports projects in the frame of contracts for the 5 th FWP [21].
If present uncertainties in nuclear data allow making reasonable pre-conceptual design
assessments, future detailed studies will require more accurate data, with drastically reduced
uncertainties. The relevant sensitivity studies have started (see [25]), but they have not yet tackled in
satisfactory way the problem of the accuracy needs in the intermediate (i.e. 20 MeV ≤ E ≤ 200 MeV)
energy range.
79

Figure 7. The MASURCA installation for the MUSE programme
The GENEPI
deuton accelerator
Vertical loading of
MASURCA tube
MUSE configuration
Deuton beam (from GENEPI)
MASURCA tube (10 × 10 cm)
Steel shielding
Réflecteur
NaSS reflector
NaSS
Core (Na-UO 2
- PUO 2
)
Lead (simulating spallation target)
Tritium or deuterium target
Beam pulse width ∼ 1 µsec
Source intensity ≈ 2 × 10 10 n/s
Sub-critical core: k eff = 0.95 ÷ 0.98
A sub-critical MUSE-4 configuration
80

Figure 8. Comparison of the neutron spectra obtained with (D,D) and (D,T) neutron sources
(as in MUSE experiments), with the reference spallation source, in the same configuration
81

3.3.3 Scenarios studies
Scenario studies have allowed during this decade to get a global picture of the transmutation
potential, mass flows at equilibrium, and consequences on the power park structure. In the illustration
of Figure 9 [26], the same type of ADS is used in order to transmute MA (double strata approach), or
Pu + MA (double component type of power park [27]).
The fractions of ADS in the park at equilibrium are shown (respectively 3.4% and 16%), and also
the MA and Pu yearly mass flows, including total losses towards a deep geological storage.
3.4 LLFP
In this area, after the performance of irradiation experiments on 99 Tc and Iodine [28], not much is
being done apart from conceptual studies, that underline the need to use high fluxes and thermalised
spectra, like in the so-called “Leakage-with-Slowing Down” (LSD) approach [11].
Projects related to the transmutation of 90 Sr and 137 Cs have finally been abandoned everywhere.
Cs transmutation is more and more considered as non-realistic (even with isotopic separation).
Other activation products have been mentioned as candidates for transmutation, but the inherent
difficulty of high neutron-consuming processes has discouraged further experimental programmes.
82

Figure 9. Scenarios at equilibrium for a 60 GWe Park – Mass flows/year (t) (only TRU)
Double strata: to transmute MA (separated from Pu).
Double component: to transmute Pu + MA (non-separated).
Same type of sub-critical ADS (gas-cooled, particle-fuel).
Power: 1 500 MWth.
Initial sub-criticality: keff = 0.98 (ip ≈ 17 mA, Ep = 1 GeV).
U
43%
UOX
PWR
235 U
EFR
Am: 0.27
Cm: 0.05
Pu: 4.4
Np: 0.3
U
Am: 1.07
Cm: 0.13
54%
Pu: 30
Np: 0.5
U: 98
U
To transmute Pu + MA:
84% U 16%
UOX
PWR
Pu: 8.6
Np: 0.6
Am: 0.5
Cm: 0.1
235 U
ADS
Pu: 10
Np: 0.1
Am: 1.1
Cm: 1.6
ADS
3.4%
Pu: 2
Am: 0.87
Cm: 0.91
Total losses:
Pu: 0.036
Am: 0.002
Cm: 0.001
Np: 0.001
Total losses:
Pu: 0.019
Np: 0.001
Am: 0.0016
Cm: 0.0016
% : fraction in the power park.
UOX-PWR: 4.9% 235 U enrichment. BU: 60 GWd/t.
Cooling time: 5 years.
Ageing before irradiation: 2 years.
Losses to the wastes: U, Pu, MA= 0.1%.
83

4. Major missing points
The analysis of the ongoing activities and of their relevance allows attempting an indication of
areas where further efforts are needed, in order to consolidate the present knowledge and to provide
elements to judge feasibility.
• It seems that what is probably still needed with a high priority is a (re-)assessment of the criteria
to judge the performance of transmutation systems. Transmutation as a waste management
option, is indissolubly related to a constant or expanded use of nuclear energy and its impact
should be evaluated on the full fuel cycle (cost and licensing of new installations, doses to the
workers, secondary wastes, acceptability, ...).
• Experimentation about fuels is a priority. No concept can be considered seriously, if the
appropriate fuels are not defined, which means characterised, fabricated, irradiated and
reprocessed. Now, very limited facilities are available to deal with MA fuel fabrication and
reprocessing (wet or dry routes) and their workload is already very demanding. Moreover, the
problem of Cm has been up to now somewhat “forgotten” and, on the contrary, it can be crucial
to define an optimum transmutation strategy or even to identify potential “show stoppers”.
To start with compounds and fuel basic properties assessment should be a priority. An
international co-ordination and share of work should be envisaged.
Also, irradiation tools with fast neutrons will be dramatically reduced in the coming years with
the remarkable exception of the JOYO reactor (at least up to ~2015). Again, an international
initiative could be envisaged to harmonise programmes and to allow the best use of existing
resources.
Reprocessing of irradiated fuels should be foreseen as an essential step of any programme on
fuels, for homogeneous recycling of fuels, heterogeneous recycling of targets, or dedicated
fuels. The priority, in the opinion of the author is with fuels for homogeneous recycling and
with dedicated fuels, if the double strata, (or the “double component”) approach is accepted.
• In the case of heterogeneous recycling, the feasibility of a fission rate >90% should be
experimentally verified, and that demonstration should be made in the case of a target
containing both Am and Cm. In fact, multi-recycling of targets should be avoided.
• It is more and more evident that transmutation studies could not necessarily require the
separation of individual MA and MA from Pu. Quite the opposite option seems to be more
attractive [3]. In that respect, partitioning/conditioning strategy [29] could represent an option
to be investigated, and which could justify partitioning by itself.
• In the case of ADS, an Experimental ADS (EADS) realisation at the 2015 horizon, is a need,
in order to prove the technology at a significant scale. The priority is the engineering concept
(i.e. component coupling, control, reliability and licensing) validation. The double role of an
EADS as a facility to validate the concept but also able to provide the appropriate fast neutron
field for advanced fuel irradiation at high damage rates, is a strong point to be made.
The present status and the necessary research in the accelerator field are the subject of another
paper at this meeting. It is however necessary to remind here that this is an essential issue,
since the expected performances of the “dedicated” high power accelerators are very
demanding in terms of reliability and availability.
• LLFP transmutation research, if at all needed, should concentrate on a realistic approach for 129 I
handling, if any. Reliable target materials and high transmutation rates should be the priority
84

goals in this field. Since once-through transmutation is hard to be envisaged, the recovery of
iodine in the irradiated target and its reprocessing, should also be the object of research.
5. A changing environment
The research in transmutation experienced a revival in the mid-eighties, essentially in the context
of waste management within programmes which gave a definite value to Plutonium and which implied
the reprocessing of irradiated fuel.
The transmutation approach was successively identified in some countries with the approach to Pu
elimination (both weapon and civil Pu). Reactor physics problems were indeed very similar. In that way,
the interests of the two previously separated communities (i.e. Pu = resource versus Pu = liability), were
somewhat federated, in particular in terms of fuels development and their reprocessing by pyroprocesses.
The third step in the evolution of the transmutation approach is underway at present, since the
objectives of a “Generation IV” or, in general, the objectives of a future nuclear power development
(beyond the horizon 2030-2050) are being globally re-discussed.
Transmutation and waste minimisation are then part of the potential criteria to define future
energy systems (reactor plus fuel cycle).
In this changing environment, it can be useful to single out some concepts or research areas,
which can have impact on the future of transmutation studies.
5.1 Evolutionary reactor concepts
A few well worked-out reactor concepts have emerged in the last few years, which, besides
attractive safety and economics characteristics, have a potential to be “inherent” MA transmuters in
the homogeneous recycling mode.
Besides the IFR concept, often quoted as a paradigm in the present paper and the Energy
Amplifier proposed by C. Rubbia, we can remind the BREST lead-cooled fast reactor concept
developed in Russia [30], the CAPRA reactor in France [31], the SCR (Super-critical Water Cooled)
concept, developed in Japan [32], but also the APA concept [33], despite the fact that it concerns
mostly an innovative assembly design for PWRs.
In particular, the nitride fuel foreseen for the BREST reactor, favours the MA transmutation by
neutron spectrum hardening. Using a pyrochemical process, it is possible to envisage for this fuel, by
multi-recycling, the transmutation of the actinide produced during irradiation. This mode, close to the
one indicated for the IFR concept, has the same advantages and of course similar drawbacks, in
particular due to the build-up of spontaneous fission neutron emitters (Cm isotopes, cf. isotopes, see
§2.1 and Table 1).
Finally, interest in gas-cooled fast reactors has been renewed, in particular to keep open the fast
reactor (FR) option, due to FR flexibility with respect to resources utilisation and their potential for
waste minimisation. In view of the “political” opposition to Na as coolant in some countries, the gas
cooling is being revisited.
85

The potential of any fast reactor, whatever the fuel type (oxide, nitride, metal) and whatever the
coolant [34], indicates that the priority for GCFRs is to design a viable reactor (in terms of safety) with
a realistic fuel form (e.g. particle, avoiding as far as possible graphite) for which no firm candidate has
been proposed up to now.
5.2 Molten salts
Molten salt reactors, besides their specific interest as energy producing systems, have also a
number of perceived advantages for transmutation, in particular:
• High burn-up potential (up to 600 MWd/t) limited only by absorption due to fission products,
minimising the quantity of fuel to be reprocessed (a few litres of salt per day).
• Actinide losses minimised in the ultimate waste form (0.01-1% of the actinide inventory).
Molten salts characteristics result also in increased flexibility:
• Continuous input of purified salt fuel and output of irradiated salt: by adjusting the TRU
concentrations, reactivity can be controlled without the use of poison or fertile material.
• Long-lived fission products (Zr, I, Tc, Cs, ...) can be added directly to the salt with no
detrimental effect on its physicochemical properties.
• In the case of thorium cycle, fertile thorium can be added to the salt to fabricate 233 U rather
than TRUs; the 233 U can be “quickly” extracted without 232 U.
Recent studies on molten salts concepts [35,36] point out the application to MA and Pu
elimination, but also indicate the way towards improved fuel cycle and waste management scenarios
(e.g. the TASSE system [37]).
5.3 The thorium cycle
A recent study of the European Union (in the frame of a contract for the completed 4 th FWP, to be
continued in the 5 th FWP) has addressed the issue of “Thorium as a Waste Management Option” [38].
The objective of the work was a re-assessment of Thorium cycles in the context of limitation of
nuclear waste production and prospects for waste burning. The aim was to obtain a review of the
major steps of the fuel cycle, focusing to the waste aspect. A restriction was made regarding reactor
types: PWR, FR and ADS.
The final report of that study shows that there are important advantages of thorium cycles with
respect to the waste issue that we will quote in detail from [38]:
• Long-lived radio-toxicity of mining waste is expected to be relatively small, which leads to
more manageable waste as compared to the uranium case.
• Fabrication of Th/Pu-MOX fuels is comparable with U/Pu-MOX fabrication methods as long
as fresh Th, fresh U and recycled Pu are used. Recycling of U bred from Th, however, needs
remote handling and reprocessing techniques specific to Thorium.
86

• The use of Thorium in PWRs always requires make-up fuel and therefore a self-sustaining
mode is impossible in such a reactor.
• To reduce the radio-toxicity of PWR waste in an once-through mode, one has to avoid 238 U
and therefore use thorium together with make-up fuel like 233 U or highly-enriched 235 U.
Advantages in terms of waste radio-toxicity are seen during the first 10 000 years of disposal.
Recycling gives a further reduction of radio-toxicity up to 10 000 to 50 000 years of disposal.
• The long-term residual risk of directly disposed fuel in a thorium matrix is still not known
very well, but there are indications on improved performance. Further experimental work is
needed to clarify this point.
• Th-assisted Pu burning, using a Th/Pu-MOX type of fuel in a PWR, is an attractive option
with respect to mass reduction of Pu.
• Fast neutron reactors and accelerator-driven systems offer both (with similar characteristics)
the possibility of a closed Th cycle without make-up fuel, except to start the cycle, reducing
mining needs and radiological risks. Full recycling of actinides gives impressively low radiotoxicity
results for the wastes over a long period of disposal, starting after the bulk of fission
products has decayed.
Non-proliferation concerns are also treated in the report.
The interest of the Th cycle should justify a number of experimental developments, in particular:
• Reprocessing and fabrication techniques of Th fuels could be extended from laboratory scale
to industrial scale and further optimised.
• Co-extraction of actinides by pyrochemistry in molten salts, aiming at losses of the order of
0.1% should be demonstrated.
• Nuclear data and physico-chemical data should be established or improved.
• Some simple irradiation experiments should be foreseen, (as it is the case in the new thorium
project for the EU 5 th FWP).
Finally, the considerations of [37], further enhance the potential of thorium, if powerful
accelerators are used.
5.4 Multipurpose neutron source installations
Recently, the “transmutation” community has become involved in the discussions around
“multipurpose” facilities, based on a high power proton accelerator, which provides neutrons, by
spallation on one (or several) target(s) for different applications.
The most known example is of course the joint KEK-JAERI project [19]. A new initiative is
under study in Europe. The ADS experimental installation could be one of the “potential” customers
of the neutrons (as it is in Japan), and, consequently, the transmutation community could be interested
both to the possibility to demonstrate the concept, and to irradiate the dedicated fuels and targets
needed to assess feasibility (see §4).
87

However, the need to single out a “leading” customer, can somewhat jeopardise the performance
allowable for the “lesser” customers. A typical example, is the debate on the pulsed or continuous
mode of operation of the high intensity proton accelerator.
6. Healthy criticism
In the reference quoted at the beginning of this report, Rasmussen and Pigford express their
doubts about the value of P&T with arguments that should be carefully taken into account still today.
Three of their arguments seem of particular relevance, besides the economical and institutional issues,
which, although of fundamental importance, have to be adapted to each specific situation:
1. The total inventory of untransmuted radioactivity in the reactor and fuel cycle must also
(besides what is sent to the repository as losses at reprocessing) be considered as a potential
waste and it takes centuries to reduce it.
2. In the search for an adequate measure of performance, the repository “intrusion” scenario, is
claimed to be the most affected by P&T. However, if one considers “intrusion” in a
repository, why not to consider “intrusion” in the installations of the fuel cycle, where most
of the inventory is kept!
3. P&T will increase to significant amounts new secondary wastes.
As far as arguments 1 and 2, it is clear (and should be always made clear in front of any type of
audience), that P&T strategies are definitely associated to an (expanded) use of nuclear energy, with
fuel (re)processing and relevant investments in new facilities. However, this (expanded) use of nuclear
energy can be made acceptable to the public by the very fact that the burden to repositories is reduced
by P&T, and that potentially physical means to eliminate all nuclear materials are provided by the
same P&T technologies, even if they should be operated for long periods of time.
Finally, the problem of secondary wastes, and the more general problem of the impact of P&T on
the fuel cycle installations, often mentioned in the present report has to be carefully quantified, in
particular in terms of social acceptability.
In conclusions, arguments against P&T can be seen simply as arguments in favour of a simplified
fuel cycle, and not necessarily in favour of the once-through cycle based on Uranium utilisation.
7. Conclusions and perspectives
Transmutation of wastes has been revisited in the last decade and, although no spectacular
breakthrough has been made, a number of significant results have been obtained.
Besides the relevant results in the aqueous chemical separation process domain (which have not
been reviewed here), one can quote:
• Understanding of the physics of transmutation and of the “neutron availability” concept.
• Understanding of the role of innovative fuels (including molten salts and particle fuels) to
improve the characteristics of the fuel cycle and to minimise wastes.
• Understanding, in that context, of the potential of pyrochemical processes both for fuel
fabrication and for irradiated fuel reprocessing.
88

• Understanding of the role of ADS to handle Pu and MA, but also to provide an option for an
extended use of the thorium cycle.
• Understanding of the role of fast neutron spectra and their flexibility. In this frame, the
discussion around the coolants for FR would benefit from a better international agreement on
pro and cons of the different options.
Since fuels play a central role in all scenarios of waste minimisation and nuclear power
development, an international share of efforts around nitrides, oxides and metals should be organised
in order to insure an optimum use of resources in the few existing laboratories to handle very active
fuels. In that frame, the availability of irradiation facilities, in particular able to provide fast spectra
(and high damage rates) is a key point and a major concern.
No convincing case can be made in favour of transmutation, without the full experimental
demonstration of its feasibility. Experiments are then needed and the relevant installations should be
kept available, with enough experienced teams. Besides the case of the installations for fuel
characterisation, fabrication and irradiation, often mentioned in this report, installations related to
basic physics (nuclear data and neutronics) will remain vital for all scenarios of development.
In the field of ADS, the development of high power proton accelerators and the construction of a
60 ÷ 100 MWth Experimental facility, at a realistic but not too far away, time horizon, seem to be
necessary in order not to loose credibility.
International initiatives should be upgraded and, besides the very valuable information exchange
goal should address the practical share of work in key fields and should help to focus on some most
promising concepts, promoting joint experiments and avoiding dispersion of efforts. In this respect, a
co-ordinated activity on pyrochemical processing is strongly suggested.
Finally, the results of the new study of OECD/NEA presently underway will certainly help to
better understand and to agree on the relative merits of two of the major options (i.e. critical fast
reactors and ADS) for waste transmutation [39].
Acknowledgements
The author acknowledges the relevance of many results provided by M. Delpech sometimes before
publication and the fruitful discussions with him, as well as I. Slessarev, J.P. Schapira and A. Zaetta.
89

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92

SESSION V
Transmutation Systems and Safety
Chairs: Y. Arai (JAERI) – W. Gudowski (KTH)
_____________________
SUMMARY
Session V started with a very interesting and important overview paper by Dr. Dave Wade (ANL)
summarising safety and operational concerns for ADS. Some of these concerns were also addressed in
some later presentations but this paper gave an insight in a logic deduction of the safety and
operational aspects that is defining an ADS.
Different presentations were given on concepts or precise projects for ADS facilities. One should remark
that some of these proposals involve multi-purpose machines based on Pb-Bi technology (i.e. KEK/JAERI,
Myrrha) where some proposals involve gas-cooled systems (i.e. GA, LAESA pebble-bed). One may remark
that our community could benefit of checking non-nuclear experiences with Pb-Bi coolants in order to shorten
our learning curve.
It should also be remarked that pyroprocessing technology can be applied to pebble-bed’s
TRISO-type of fuel which indicates the potential for closed fuel cycles by recycling this fuel.
The session also showed good arguments in favour of nitride fuels.
93

SAFETY CONSIDERATIONS IN DESIGN OF FAST SPECTRUM ADS FOR
TRANSURANIC OR MINOR ACTINIDE BURNING:
A STATUS REPORT ON ACTIVITIES OF THE OECD/NEA EXPERT GROUP
OVERVIEW PAPER
D.C. Wade
Argonne National Laboratory
9700 S. Cass Avenue, Building 208, Argonne, IL 60439, USA
Abstract
The Nuclear Development Committee of the OECD/NEA convened an expert group for a
“Comparative Study of Accelerator Driven Systems (ADS) and Fast Reactors (FR) in Advanced
Nuclear Fuel Cycles”. The expert group has studied complexes (i.e. energy parks) of fission-based
energy production and associated waste management facilities comprised of thermal and fast reactors,
and ADS. With a goal to minimise transuranic (TRU) flows to the repository per unit of useful energy
provided by the complex, the expert group has studied homogenous and heterogeneous recycle of
TRU and minor actinides (MA) in the facilities of the complex using aqueous or dry recycle in single
and double strata architectures. In the complexes considered by the expert group the ADS is always
assigned a TRU or MA (and sometimes a LLFP) incineration mission – with useful energy production
only as a secondary ADS goal to partially offset the cost of its construction and operation.
Ancillary issues have also been considered – including ADS safety challenges and strategies for
resolving them. This paper reports on the status of the expert group’s considerations of ADS safety
strategy.
95

1. Introduction
The term ADS comprehensively includes all non-self sustaining fissioning neutron multiplying
assemblies which are driven by an external neutron source provided by a charged particle accelerator
and a neutron producing target. ADS systems under current study worldwide include both thermal and
fast neutron multiplying media comprised of either liquid or solid (lattice) fuel and driven by either
cyclotron or linear proton accelerators and spallation targets (liquid and solid) of various heavy metals.
The underlying missions targeted for ADS systems span the range from nuclear waste incineration
with ancillary power production through power production with integral waste self-incineration to
finally, excess neutron production for the purpose of isotope production via neutron capture reactions
on targets.
The OECD/NEA expert group on “Comparative Study of ADS and FR’s in Advanced Fuel Cycles”
[1] confined its scope of inquiry to a subset of ADS configurations – those targeted for nuclear waste
incineration with ancillary power production, and specifically those which operate on a fast neutron
spectrum with a solid fuel pin lattice. Moreover, the expert group set a requirement of maximum
“support ratio” (i.e. maximum energy from the reactors in the complex compared to energy from the
ADS in the complex) which leads to inert matrix fuel (i.e. 238 U and 232 Th – free fuel) for the ADS. The
scope of this discussion of safety strategy is similarly confined in scope. Even within the limited scope, a
range of possibilities exists. The ADS might be a minor actinide (MA) burner or a TRU burner; the
physics and safety characteristics of these cases differ because of differences in their values of βeff
(which helps to set the degree of sub-criticality of the ADS) and in their reactivity burn-up swing (which
helps to set the control strategy). The choice of coolant (liquid metal or gas) and fuel type (oxide, nitride,
metal) also distinguishes members of the ADS class considered by the expert group. The choice of
recycle (partitioning) technology (aqueous, dry) directly affects the architecture of the energy complex
and indirectly affects the ADS itself. Figure 1 illustrates the several energy producing complexes which
were considered by the expert group and identifies the waste management function of the ADS studied in
the single strata architecture (3B) and (the double strata architecture (4)).
The fast spectrum, solid fuel, waste incinerating class of ADS considered by the expert group
shares with all ADS a distinction from critical reactors in relying on an external neutron source rather
than a self generated delayed neutron source for maintaining the neutron population in balance – with
attendant changes in dynamic response and in control strategy. However, the class of ADS considered
here offers design and safety challenges which are unique vis-à-vis other ADS classes in the areas of
burn-up control compensation and reactivity feedback characteristics; these unique challenges are
traceable to a small number of salient design features which derive directly from the requirements of
the TRU or MA incineration mission – with ancillary power production. The salient design features of
the ADS whose safety features were considered by the OECD/NEA expert group include the
following:
• Fertile-free transuranic or minor actinide fuel.
• Multiple recycle of fuel to (near) complete fission incineration of transuranics.
• Fast neutron spectrum.
• Choice of coolant (Na, Pb-Bi or He).
• Sub-delayed critical operating state.
• External neutron source created via spallation reactions of high-energy protons on a heavy
atom spallation target.
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The approach taken by the expert group to identify ADS safety issues and discuss strategies for
addressing them was as follows. First, the top level safety functions to be satisfied for any fission
chain reacting system (reactor or ADS) were enumerated. Then, the distinguishing features of the class
of ADS considered here were traced back to the mission assigned to it in the complex; as a way to
indicate which features (and safety issues) would be changed by a change in mission requirements.
Then, an impact matrix was constructed (with “safety function” columns and “distinguishing feature”
rows) to identify where the ADS distinguishing features have raised safety-relevant challenges which
are different from the more familiar situation for a fast reactor. Finally, for each identified challenge, a
set of alternative safety strategies for addressing it were discussed with the views that:
Figure 1. Nuclear fuel cycle schemes
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97

• Safety should be “designed in” from the outset.
• The vast experience base from fast reactor development should be exploited where possible, e.g.:
− Defence in depth principles.
− Single failure criterion.
− Exploit passive safety principles.
And that in devising ADS strategies, one must:
• Bear in mind that safety implications on recycle and re-fabrication operations accrue to
choices made for the ADS per se and must be factored in.
The work of the expert group is ongoing with a target for completion by late spring 2001. This
paper provides a status report.
2. Safety functions and strategies for fissioning systems
2.1 Basic safety functions for fissioning systems
At a basic level, there are six safety functions to be fulfilled when operating fissioning systems.
1) The nuclear fuel must remain contained within a controlled space because of its radiotoxicity;
this is traditionally accomplished by use of multiple containment barriers.
2) Shielding must be kept in place between fissioning and fissioned fuel and humans to avoid
suffering radiation damage.
3) A heat-transport path must be in place to carry energy away from the fissioning medium to a
heat sink; usually an energy conversion plant.
4) The rate of release of fission energy in the chain reacting medium must be regulated to
remain in balance with the rate of energy delivery to the heat sink, so as not to overheat the
containment barriers around the fuel and challenge their integrity; a capacity to store heat in
the reacting medium and heat transport channel will buffer mismatches of short duration or
small amplitude.
5) Since some 5% of the energy liberated from each fission event is initially retained in nuclear
bonds of unstable fission products, and since these fission products subsequently decay
according to their natural radioactive-decay time constants, a means must be provided for
transporting heat from the fission products and transuranics in the fuel for all times
subsequent to the fission event. Failure to satisfy the latter two safety functions could lead to
overheating of the fuel with the potential to defeat the integrity of the containment and
shielding.
6) Operation of the fissioning device in a quasi steady state mode requires a balance of neutron
production and destruction rates from one fission chain generation to the next – even as the
composition of the chain reacting medium changes due to transmutation and as the
absorption, leakage, and neutron production properties of the fissioning medium change with
changes in composition and temperatures.
98

2.2 Safety strategies
Strategies to fulfil the six basic safety functions have been developed and refined over many years for
conventional (critical) reactors. The strategy employs defence in depth such that any single failure will not
defeat the strategies for meeting safety functions and thereby result in unacceptable release of radiotoxicity;
multiple barriers (fuel cladding, primary coolant boundary, and reactor containment building) are used to
prevent release of radiation even under accident conditions. Highly reliable (diverse and redundant)
systems for controlling and terminating the chain reaction are used to match heat production to removal
rate. Highly reliable, redundant/diverse systems for decay heat removal are provided. High quality
construction and verification norms minimise manufacturing flaws, and rigorous maintenance, formal
procedures, and exhaustive training and certification of operators and maintenance workers are used to
minimise the occurrence of human error which could subvert the achievement of the safety functions. Once
safety is “designed into” the system, its efficacy is judged by an independent safety regulative authority on
a plant-by-plant basis prior to deployment and during its operation.
In recent years, the fast reactor safety design strategy has gone beyond those traditional measures, and
the system architecture consisting of the reactor heat source coupled to the balance-of-plant heat engine is
configured to achieve the safety functions by exploiting the natural laws of physics to the maximum degree
achievable. This passive safety approach partially supplants the traditional engineered devices by
exploiting passive systems or inherent characteristics that play the role of “functional redundancies” (i.e.
they can, in case of failure of the upstream line of defence achieve the same safety related mission); the
approach is so implemented to assure safe response 7 – even if the engineered systems which require
assured sources of power and highly reliable “active” sensing and switching equipment were to fail, or if
multiple, compounding failures and human errors were to occur simultaneously. The passive safety
approach can be applied for all the defence in depth levels, i.e. accident prevention, accident management
and consequence mitigation. The passive concepts can employ inherent reactivity feedbacks to keep heat
production and removal in balance. Designs having minimal reactivity loss upon burn-up and minimal
reactivity vested in control rods can preclude reactivity addition accidents. Designs having large margins to
damage temperatures and large thermal mass provide reactivity feedbacks with room to operate without
reaching damage temperatures or conditions. Designs using buoyancy-driven flows and always-operating
heat transport paths to ambient remove decay heat without systematic reliance on switching of valve
alignments or active monitoring. These passive safety approaches for fast reactors have been demonstrated
[2] in full-scale tests at EBR-II, RAPSODIE, FFTF, BOR-60, etc.
Given that the safety approaches for FRs are well known, the plan for this paper is to first describe the
chain of logic which gives rise to the salient differences between FRs and that class of ADS studied by the
OECD/NEA expert group. Then a broad survey is made of each salient difference of the ADS design as
compared to a FR to identify which of the six basic safety functions might be affected by this particular
salient difference. Following that, for each case having an identified difference; potential strategies for
fulfilling the function for an ADS are discussed.
7 A passive system should be theoretically more reliable than an active one. The reasons are that it does not need
any external input or energy to operate and it relies only upon natural physical laws (e.g. gravity, natural
convection, conduction, etc.) and/or on inherent characteristics (properties of materials, internally stored energy,
etc.) and/or “intelligent” use of the energy that is inherently available in the system (e.g. decay heat, chemical
reactions, etc.). Nevertheless passive devices can be subject to specific kinds of failure like, e.g. structural failure,
physical degradation, blocking, etc. Generally speaking, the reliability of passive systems depends upon:
• The environment that can interfere with the expected performance.
• The physical phenomena that can deviate from the expectation.
• The single components reliability.
99

3. Minor actinide and/or transuranic burning ADS design rationale and distinguishing design
features
Overall purpose; support ratio, and fertile-free fuel – First, the overall purpose of the class of ADS
considered here is to function as one element of an integrated nuclear power enterprise comprised of
conventional and advanced power reactors for energy production and ADS for reducing the radiotoxicity of
the nuclear waste produced by these power reactors – prior to its entombment in a geologic repository. The
radiotoxic materials targeted for incineration may be minor actinides (MA) or may be transuranics (TRU),
depending on the configuration of the overall enterprise. The ADS also may incinerate selected fission
products (see Figure 1, which illustrates the several power production complexes considered by the expert
group). The ATW [3] (US design) is an example of a TRU incinerator such as case 3B; the JAERI double
strata ADS [4] is an example of a MA incinerator such as case 4.
The transuranics are fissioned in the ADS to transmute them to fission products with the concomitant
release of heat amounting to about 1 gm TRU incinerated per MWth day energy release. For a fissioning
device, the incineration rate of TRU depends on the power rating of the heat removal equipment – be it an
ADS or a reactor and while the ADS plant will likely use the liberated heat for power production to offset the
cost of its operation, its primary function is to reduce the transuranic and long-lived fission product inventories
emanating from the power reactors deployed in the nuclear energy complex. The “support ratio” of the
integrated power producing complex is the ratio of power of the reactors in the enterprise to the power of the
ADSs in the enterprise. A large support ratio is targeted for the class of ADS designed for waste incineration
with ancillary power production considered here so as to relax the demands on ADS cost and energy
conversion ratio – inasmuch as the ADS will then comprise a smaller segment of the overall integrated energy
supply complex. The primary purpose of the ADS is to maximise incineration rate per unit of heat that has to
be removed. Fertile-free fuel is the first salient design feature shared by proposed ADS systems of the class
considered here – to avoid in situ production and incineration of new transuranics. A 3 000 MWth ADS plant
operating for 300 days per year transmutes about 900 kg of TRU into nearly 900 kg of fission products and
releases 900 Gigawatt days of thermal energy.
Multiple recycle – The ADS will operate on a closed fuel cycle with a feedstream of transuranics or
minor actinides arriving from the power producing reactors and with fission-product-containing (largely
actinide-free) waste forms leaving destined for a geologic repository. Internal multiple recycle of the ADS
fuel will be required to reconstitute the fuel into fresh cladding because the fluence required for total fission
consumption exceeds the neutron damage endurance of any known cladding. Recycle is required also to inject
new feedstock into the ADS lattice to sustain the neutron multiplication within its design range as well as to
extract the fission products destined for geologic disposal. Although not unique to ADS, a need for multiple
recycle constitutes a second salient feature of the ADS considered here.
Except for the “once-through cycle” (Case 1 of Figure 1), the recycle step in the overall complex is
where the waste stream to the geologic repository is generated. It is comprised of fission products and of TRU
or MA trace losses which escape the recycle/refab. processes back to the ADS or FR. These trace losses of
TRU or MA to waste are to be minimised if the ADS is to achieve its assigned mission; it is clear that both the
trace loss per recycle pass and the number of recycle passes fully control the ADS contribution to the
complex’s total loss – and that therefore a high average discharge burn-up from the ADS is desirable.
Moreover, since the radiotoxicity per gram and also the half life of the various TRU or MA isotopes vary, it is
desirable that the transuranic isotopic spectrum achieved upon multiple recycle should be favourable in terms
of long-term toxicity (including accounting all post emplacement decay daughters 8 ) – the ADS neutron
spectrum is controlling in that regard.
Fast neutron spectrum – Upon multiple recycle to achieve total fission incineration, the TRU or the MA
isotopic composition of the LWR or FR spent fuel feedstock evolves to an asymptotic ADS recycle feed
8 For example for the US (oxidising environment) geologic repository, 237 Np (a post emplacement daughter in the
241 Pu→ 241 Am decay chain) dominates the long term toxicity.
100

composition; this composition depends on the neutron spectrum to which it is subjected. (see Table 1) The ADS
considered here is designed to operate on fission chains in the fast neutron range so that all transuranic elements
stand a good chance to fission upon a single neutron absorption and thereby to minimize the development of an
isotopic spectrum which is skewed toward heavier mass transuranic isotopes. Table 2 shows [5] that a fast
neutron spectrum (having high fission probability for all TRU) is entirely essential to achieve total consumption
in an MA burning ADS and it is preferable to a thermal spectrum for TRU burning. Table 3 indicates [6] that the
asymptopic isotopic spectrum from multi-recycle in a fast spectrum will lead to a more favourable long term
radiotoxicity burden that does that arising from thermal spectrum burning. A fast neutron spectrum is the third
salient design feature of the class of ADS systems considered here.
Table 1. Equilibrium distribution of transuranic isotopic masses
for high fluence exposure to thermal and fast neutron spectra
Isotope
Thermal
neutron spectrum
Fast
neutron spectrum
237 Np 5.51 0.75
238 Pu 4.17 0.89
239 Pu 23.03 66.75
240 Pu 10.49 24.48
241 Pu 9.48 2.98
242 Pu 3.89 1.86
241 Am 0.54 0.97
242m Am 0.02 0.07
243 Am 8.11 0.44
242 Cm 0.18 0.40
243 Cm 0.02 0.03
244 Cm 17.85 0.28
245 Cm 1.27 0.07
246 Cm 11.71 0.03
247 Cm 0.75 2.E-3
248 Cm 2.77 6.E-4
249 Bk 0.05 1.E-5
249 Cf 0.03 4.E-5
250 Cf 0.03 7.E-6
251 Cf 0.02 9.E-7
252 Cf 0.08 4.E-8
Total 100.00 100.0
Note: all values are atom % of transuranic inventory built up as a result of extended exposure to a
neutron flux. (Calculated as the steady-state solution of the depletion-chain equations independent of
criticality considerations.)
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Table 2. Values of Dj (neutron consumption per fission) for isotopes j or for a fuel type (Dj

Table 3. Radiotoxicity data (CD = Cancer Dose Hazard)
Isotope
Toxicity factor
CD/Ci
Half-life
Years
Toxicity factor
CD/g
Actinides and their daughters
210 Pb 455.0 22.3 3.48E4
223 Ra 15.6 0.03 7.99E5
226 Ra 36.3 1.60E3 3.59E1
227 Ac 1185.0 21.8 8.58E4
229 Th 127.3 7.3E3 2.72E1
230 Th 19.1 7.54E4 3.94E-1
231 Pa 372.0 3.28E4 1.76E-1
234 U 7.59 2.46E5 4.71E-2
235 U 7.23 7.04E8 1.56E-5
236 U 7.50 2.34E7 4.85E-4
238 U 6.97 4.47E9 2.34E-6
237 Np 197.2 2.14E6 1.39E-1
238 Pu 246.1 87.7 4.22E3
239 Pu 267.5 2.41E4 1.66E1
240 Pu 267.5 6.56E3 6.08E1
242 Pu 267.5 3.75E5 1.65E0
241 Am 272.9 433 9.36E2
242m Am 267.5 141 2.80E4
243 Am 272.9 7.37E3 5.45E1
242 Cm 6.90 0.45 2.29E4
243 Cm 196.9 29.1 9.96E3
244 Cm 163.0 18.1 1.32E4
245 Cm 284.0 8.5E3 4.88E1
246 Cm 284.0 4.8E3 8.67E1
Short-lived fission products
90 Sr 16.7 29.1 2.28E3
0 Y 0.60 7.3E-3 3.26E5
137 Cs 5.77 30.2 4.99E2
Long-lived fission products
99 Tc 0.17 2.13E5 2.28E-3
129 I 64.8 1.57E7 1.15E-2
93 Zr 0.095 1.5E6 2.44E-4
135 Cs 0.84 2.3E6 9.68E-4
14 C 0.20 5.73E3 8.92E-1
59 Ni 0.08 7.6E4 6.38E-3
63 Ni 0.03 100 1.70E0
126 Sn 1.70 1.0E5 4.83E-2
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Table 4. Delayed neutron fraction
Isotope γ d /γ total ⇒ 10% Fertile fission raises β in fertile containing fast reactor fuel
238 U 0.0151
232 Th 0.0209
235 U 0.00673
239 Pu 0.00187
241 Pu 0.00462
β( 238 U)
0.10 × 0.0151
= 0.00151
+
+
β( 239 Pu)
0.90 × 0.00187
0.00168
242 Pu 0.00573
237 Np 0.00334
241 Am 0.00114
243 Am 0.00198
242 Cm 0.00033
= 0.00319
(Nearly doubles β vis-à-vis fertile-free fuel)
Sub-delayed critical operating state – Transuranic fuel containing no fertile ( 238 U and 232 Th)
atoms exhibits a delayed neutron fraction for fast fission which is in the range of 0.0015 to 0.0020 i.e.,
about half the value for a conventional fast reactor and about a sixth the value for a conventional
LWR. Table 4 displays β for fast fission of various actinide isotopes and shows that even at only 10%
contribution to fissions, as is typical for a fast reactor, fertile 238 U or 232 Th contribute very significantly
to delayed neutron fraction – doubling its value from what applies for fertile-free fuel. For fertile-free
TRU or MA fuel compositions the delayed neutron fraction is remarkably small and therefore the
margin to prompt critical is correspondingly small. This feature, when combined with reactivity
feedback considerations discussed next, leads to further salient design features of ADS specifically as
a safety strategy approach.
The neutron leakage in a fast neutron lattice is sensitive to the assembly geometry because of the
long neutron mean free path. Subtle thermo-structural-induced geometry changes which are dependent
on power to flow ratio (P/F) – such as fuel bowing, grid plate expansion, etc. – change the neutron
leakage fraction in response to power and flow changes. For example Figure 2 illustrates for the FFTF
sodium cooled fast reactor the normalised power to flow ratio dependence of the radial expansion plus
bowing component of thermostructural reactivity feedback. Several features are notable [10]; first the
amplitude is nontrivial with respect to β over the range 0 < P/F < 1; clamping and duct wall tolerances
and stiffness are designed [7] so as to assure negative bowing reactivity feedback at P/F in the vicinity
of the operating point, P/F ~ 1; and reactivity increases with decreasing P/F may become
indeterminant [7] at low values of P/F owing to the “unlocking” of above core load pad structural
contact. Numerous other leakage dependant thermostructural reactivity feedbacks (grid plate
expansion, fuel axial expansion, etc.) are also individually nontrivial in amplitude compared to β, as
illustrated in Figure 3 for a power change transient in the modular PRISM reactor [8].
ADS designs using fertile-free fuel have high values of k ∞ and correspondingly high neutron
leakage fractions [9]. With a reduced delayed neutron fraction of 0.002 or less and even assuming an
unrealistically small neutron leakage fraction of only 5%, a change in leakage fraction of only a few
per cent of its value – induced by thermostructural feedbacks – would exceed the .002 ∆k/k offset
from prompt critical. Not only is it impossible to design for and to control thermostructural response to
that degree of precision [10], but variability as well as controllability is the issue here. In an ADS
functioning as a waste burner, the fuel composition itself and its η value and β value can be expected
104

to vary from loading to loading as LWR spent fuel and/or FR spent fuel of differing burn-up, differing
cooling times and differing origins supply the ADS fuel feedstock. These feedstock variabilities
change not only k ∞ and the concomitant leakage fraction and resulting amplitudes of thermostructural
feedbacks, but they also change the delayed neutron fraction and offset from prompt criticality itself.
Consider the effect on core k ∞ of even small variability in 239 Pu/ 241 Pu/ 241 Am ratios in ADS feedstock 9
as indicated by their vastly differing η values illustrated in Figure 4. Or, referring to Table 4, consider
the effect on β of the transformation of 241 Pu (14.35 year half life) to 241 Am over different cooling
periods prior to introduction into the ADS – a factor of four change in β contribution.
Figure 2. Reactivity from radial core expansion as a function of normalised power-to-flow ratio,
comparing the FFTF correlation and the SASSYS/SAS4A calculation
(β = 0.003)
Figure 3. Unprotected transient overpower for ALMR (β = 0.003)
9 Local power peaking in reload assemblies also are affected by these variabilities.
105

Figure 4. η (Neutrons released per neutron absorbed) for several isotopes
Taken all together, the variability and non controllability of the reactivity state of a fertile-free
MA or TRU lattice relative to the reduced offset from delayed critical to prompt critical requires a
safety strategy to assure that no potential can exist for power/flow induced reactivity feedback to carry
the lattice into the super prompt critical range – even accounting for the variability in β, k∞, and
leakage fraction which result from feedstock of varying composition. The ADS strategy is one
approach to this need – an increased offset of operating state from prompt critical is achieved by
operating sub delayed critical. So as to avoid any potential for power/flow induced reactivity
feedbacks to inadvertently carry the system into the super prompt critical regime, the geometry and
composition of the ADS lattice is configured so that the operating margin to prompt critical will
always substantially exceed the maximum power/flow reactivity feedback – while accounting for
expected variability in the values of η and β owing to differing feedstock compositions. But the
resulting offset then exceeds the value of the delayed neutron fraction itself – so it makes the operating
point of the ADS lattice sub-delayed critical. An external source is required, therefore, to drive a
continuing fission reaction. The fissioning system multiplies the externally supplied neutron source. A
sub-delayed critical operating state driven by an external neutron source is the fourth and defining
design feature of all ADS.
Spallation neutron source – At 1 gm of TRU or MA incinerated per MWthermal day, ADS facility
heat ratings must lie in the range of 1 000 MWth or more to support any reasonably – sized energy
complex. The size of the neutron source required to drive a sub-delayed-critical ADS depends on both
106

the desired heat rating and on the degree of neutron self multiplication of the lattice, which depends on
the degree of sub-criticality (see Equation 1, below). With the required offset from prompt critical no
less than 2 or 3% ∆k/k (i.e. a source neutron fission chain multiplication no greater than 30 to 50), it is
clear that no passive neutron-emitting source is strong enough to meet the requirement for
~1 000 MWth power rating. However, plausible extensions in proton beam current capability which are
now achieved in linear accelerators (i.e. beams of multi megawatt levels), could achieve required
neutron source strength by driving a heavy metal spallation target. This leads to the fifth salient design
feature of an ADS; namely the external source must derive from a spallation neutron target driven by
a high power proton accelerator.
3.1 Summary of distinguishing features for TRU or MA burning ADS
The distinguishing features of the type of ADS considered by the expert group derive directly
from the mission assigned to it in the energy complex: namely – TRU or MA (and LLFP) incineration
for waste management in the integrated energy complex with power generation only secondary to
partially offset cost, – combined with the a-priori assumptions on scope of cases considered by the
expert group: namely fast spectrum, solid fuel, and maximised support ratio.
The resulting distinguishing features are:
Those shared with FR:
• Fast neutron spectrum.
• Solid fuel lattice.
• Multiple recycle.
• Choice of coolant: Na, Pb-Bi, gas.
And those unique to ADS:
• Fertile free fuel.
• Sub-critical operating state.
• Spallation neutron source driven by a high power proton beam.
4. Safety-related challenges arising specifically from ADS design features
The salient design features of ADS give rise, in some cases, to different safety-related challenges
and different approaches to fulfilling the six cardinal safety functions for fissioning systems as
compared with the more familiar issues and safety strategy which apply for a fast reactor. Table 5,
which tabulates salient design feature (matrix rows) versus required safety function (matrix columns),
identifies where these differences exist. In Table 5 the effect of salient design features on strategy for
meeting safety functions is indicated for both normal operational safety and for off-normal safety
situations. The Table 5 impact matrix is overviewed here and the safety strategy options to
accommodate the new issues are briefly discussed.
107

Table 5. ADS Distinguishing feature vs. basic safety function impact matrix
Neutron balance Heat removal Regulation of
power/flow
Containment Shielding Decay heat
removal
Fast
spectrum
Choice of
coolant
Fertile-free
fuel
Sub-critical
state
Spallation
source
Multiple
recycle
Normal Offnormal
Normal Offnormal
Normal Offnormal
Normal Offnormal
Normal Offnormal
Normal Offnormal
Legend: new issues arise vis-à-vis a fast reactor.
Distinguishing features
108

Proton beam transport tube and spallation neutron source effects – The most readily obvious
geometrical difference occurs via the introduction of the proton beam tube. First is its topological
effect on the multiple containment barrier defence in depth containment safety strategy. In fast
reactors, the fuel is contained first by its cladding (or by multiple layer ceramic barriers in particle
fuel), then by the primary cooling circuit boundary and lastly by the containment building. For the
ADS based on linacs, the proton beam tube penetrates the last of these and employs a metallic proton
beam window as a topological continuation of the primary coolant boundary. The safety issue pertains
to the preservation of defence in depth for the containment and shielding functions. In a fast reactor
similar topologies result from steam lines which penetrate the containment, and from IHX tubes which
comprise a topological extension of the primary coolant boundary. In BWRs, the steam lines penetrate
both the containment and the reactor vessel. Fast acting values at the containment boundary of steam
pipes and robust heat exchanger tube walls are the safety strategies used for these reactors. Fast acting
valve safety strategies employed for reactors will be considered for the ADS. For the ADS, the beam
window operates in an especially hostile environment in light of its temperature and the proton and
neutron bombardment that it experiences; the hazard deriving from the multi-megawatt proton beam
potentially impinging on any of these barrier boundaries or valves is unique in an ADS.
The beam tube introduces new issues as well in the area of shielding safety function – comprising
a streaming path from the fissioning lattice to the exterior of the vessel as well as an unshieldable
pathway for radiation activation of the bending magnets or accelerating structures of the accelerator.
Finally, the several tens of centimetre diameter evacuated beam tube presents a new issue in the area
of a potential positive reactivity effect should the beam tube flood and decrease the neutron leakage;
the degree of reactivity offset from prompt critical must be sufficient to safely accommodate such
potential flooding.
The presence of an strong spallation neutron source has an effect on power density peaking factor
[11] and on the change in power peaking as k ∞ of the lattice changes with burn-up and as the ratio of
source to fission multiplied neutrons is altered by changes in source strength or reactivity. Also,
depending on beam tube entry geometry, the fuel-loading pattern may be non-azimuthally symmetric –
again affecting power density profile. Tailored spatial k ∞ zoning can be applied and, a design strategy,
which relies on, increased margins so as to accommodate local shifts in power/flow ratio – while
undesirable for a dedicated power producer, is quite consistent with the ADS mission where power
production is an ancillary function only.
The ADS power output is proportional to spallation source strength and sub-critical reactivity
offset via the relationship:
S S
P α =
1 − ρ
− 1
k
k −1
where ρ =
k
(1)
Asymptotic adjustment of power (or power density) scale to first order with source changes or
reactivity changes as:
δP
δS
δP
= −
P S ρ
0 0 0
(2)
109

While a favourable ADS safety feature derives from its asymptotic rather than rising period response
to a positive reactivity insertion [12], a safety challenge still remains in assuring that positive source
strength or source importance changes cannot take the ADS to damaging overpower conditions.
Equation 2 indicates that e.g. at an beginning of cycle offset of -ρo equals 3% ∆k/k and a burn-up
reactivity loss of 6% ∆k/k, the source for maintaining end of cycle power level would have to exceed
beginning of cycle requirement by 100%, leading to a factor of two overpower potential should the full
source strength be introduced prematurely. Options to minimise burn-up reactivity loss include multibatch
fuel loading [9] and optimal mixes of plutonium and minor actinides [13] to flatten the reactivity
change with burn-up. However, given fertile-free fuel, it has proven impossible for ADS designers to
achieve small burn-up reactivity loss; so that compensation by either external reactivity changes
(control rods) or by source strength or importance changes is unavoidable. In every case then, an
overpower potential exists.
At the other extreme, if ADS heat removal equipment were to fail (loss of flow or loss of heat
sink), then the beam would be required to trip off within seconds to avoid overheating and melting of
the fuel [14].
Such considerations of controlling ADS on the basis that the beam current will assume the
functions assigned to control rod in fast reactors might lead to a “nuclear safety grade” designation for
the accelerator equipment and its maintenance, or at least for its controller – having significant
unfavourable cost implications. Alternately, the proton beam might be operated at 100% strength at all
times with a safety grade scram circuit, while burn-up reactively loss could be compensated by (safety
grade) control rod actuators. Using the same example as above, a control rod bank worth of 6% ∆k/k
would accomplish the same burn-up reactivity compensation as a factor of two larger proton
accelerator – with a significant favourable cost advantage likely. Or, mechanical adjustments of
neutron source importance via changes in source location or spectral importance may be options. Even
adjustable volume fraction mixes of various spallation target materials having differing neutron yield
per proton might be considered. In all cases a safety hazard exists in potential for premature actuation
of the excess source or reactivity prior to burnout of the lattice; it simply cannot be avoided, short of
letting the power rating decrease with burn-up.
Coolant choice effects – The distinguishing characteristic of the coolant choices relate to system
pressure, lattice power density, effect on neutron spectrum, and chemical activities – as tabulated in Table 6.
Table 6. Coolant characteristic features
Na Pb-Bi He
System pressure Low Low High
Lattice power density High Low Lower
Neutron spectrum Hard Harder Harder
Chemical activity High Low Null
These distinguishing features permeate the entire design approach for ADS and fast reactor alike
and influence the resulting safety strategies. High pressure gas cooling introduces a loss of coolant
vulnerability but eliminates chemical compatibility issues. Gas cooling shares with Pb-Bi cooling the
need for a low power density, open fuel pin lattice – which leads to potential for significant reactivity
additions upon hypothetical pin disruption or compaction but reduces potential for blockage from
foreign objects.
110

Freezing temperatures and coolant/structural/fuel chemical interactions and potential for “local
fault propagation” into flow blockages are important safety relevant issues for liquid metal coolant
choice, and given that every fertile-free fuel under consideration for ADS use lacks a data base of
inservice experience, this issue will require a substantial R&D effort in every case.
The coolant voiding reactivity coefficient is of reduced safety relevance because the ADS
provides an added degree of freedom in the sub-critical offset from prompt critical sufficient to cover
voiding worths [15].
The high density of Pb alloy coolant introduces several new issues for ADS and fast reactor alike;
first is the structural support and the seismic structural response of large reactor vessels when filled
with dense lead alloy. Second is the design of refuelling equipment and fuel assembly hold-down
devices for the case where the fuel and the structures are less dense than the coolant and tend to float
in it. Its high boiling point, on the other hand, provides more than sufficient margin to boiling.
A significant safety-relevant issue for fast reactors and ADS also is the consequence of failing to
maintain leak tightness of the primary coolant system. Rank ordering of coolant favours liquids over
gas for this issue because only gas operates at above-ambient pressure. However, each coolant
displays a vulnerability which is unique to itself. Since gas-cooled systems operate at high pressure, a
loss of integrity anywhere in the gas heat transport circuit could lead to a loss of coolant accident. Loss
of coolant accidents are of extremely low probability for liquid metal cooled systems using a pool
layout but each liquid metal displays a safety vulnerability upon leakage of primary coolant. Sodium
burns in air, creating an aerosol containing (24-hr γ-emitting) radioactive 24 Na. Pb-Bi alloy does not
burn but none-the-less releases 138-day (α-emitting) 210 Po. Safety approaches have been developed in
the fast reactor communities to mitigate and recover from Na and Pb-Bi leakage events and, as
compared with a gas leakage loss of coolant vulnerability in a gas cooled fast reactor, the liquid
coolant mitigation technologies are at a more mature state of development. However, in-service
inspection and repair are a serious vulnerability for opaque liquid metal cooling as compared with gas
cooling.
For fast spectrum ADS applications, safety-related issues upon loss of primary boundary integrity
should be evaluated first at the particular point of vulnerability innate to ADS: the single thin-wall
boundary between the transmuter coolant and the vacuum extension of the proton beam tube leading
into the spallation target located at the centre of the core. The window operates in a hostile
environment of proton and neutron damage and it alone lies between the centre of the fissioning lattice
and the proton accelerating structures external to the containment building. Beam tube melt-through
upon a beam misalignment similarly presents a loss of containment boundary vulnerability.
Fertile-free fuel effects – Fertile-free fuel has a high value of η (see Figure 4) and requires a
design strategy for safely disposing of excess neutrons. The options are leakage or internal parasitic
capture – either discrete absorbers or absorbers homogeneously mixed with the fuel. Thermostructural
reactivity feedback variations can be reduced the smaller is the leakage fraction and this is desirable
for reasons discussed above. Recycle/re-fabrication batch sizes may benefit from the homogeneous
absorber option. On the other hand, radial k ∞ zoning using only a single fuel pin fabrication
specification may be achievable using discrete absorber pins.
As already discussed above, the absence of internal conversion of fertile to fissile species with
burn-up will place demands for burn-up reactivity compensation on other design approaches – such as
source strength or source importance changes, batch refuelling, or absorber control rod changes. For
minor actinide burners, in situ isotopic transmutations mitigate but do not eliminate this issue.
111

Optimised mixes of MA and Pu can be tailored [13] to limit burn-up swing; but in every case source or
reactivity compensation strategies are needed.
An off-normal safety related challenge derives from fertile-free fuel – which excludes the
traditional Doppler contributor to prompt negative reactivity power feedback in a FR. Small (but not
zero) Doppler feedback has been accommodated (and beneficially exploited for a passive safety
mechanism) in metal-fuelled fast reactors. However, an HCDA termination mechanism will have to be
devised for an ADS having fertile free fuel [16], high melting point oxide-fuelled FRs traditionally
rely heavily on prompt Doppler feedback to limit the pre-disassembly energy generation which
controls severity of HCDAs. Inertial resistance to disassembly in an HCDA sequence is an additional
issue with Pb-Bi cooling.
Pure TRU or MA fuel presents issues in recycle batch sizes and processing geometries because of
a small critical mass. Experience exists with metal-fuel/pyro recycle where discrete rather than
continuous processing is employed and batch size is limited by relatively larger fast criticality
constraints; this issue would require special care in the case of continuous aqueous reprocessing
having very small critical masses.
Sub-critical operating state and dynamics effects – A fundamental distinction between ADS and
critical reactor safety-relevant control arises because of the dramatic differences in dynamic response
of critical versus sub-critical source-driven lattices. In a source-driven system, a change in source
strength or in source importance or a change in reactivity will cause the neutron population and power
level to promptly 10 adjust to a new asymptotic level in accordance with Equation 2; whereas in a
critical reactor a change in reactivity leads (absent reactivity feedbacks) to an asymptotic period (or
exponential time change) of neutron population, the promptness of which is controllable by the
reactivity insertion magnitude. In a critical reactor, the period of power adjustment is chosen to match
the thermal and structural time constants – which are in the range of 0.1 to 100 seconds (see Figure 3).
The dynamics and control challenges can be illustrated under the excellent assumption that the
neutron population, n(t) is in prompt quasi-static equilibrium with the source. For a reactor it is the
delayed neutron source; for the ADS it is the external spallation source plus the delayed neutron
source:
dn
dt
ρ − β
= 0 = n+ λC + S
∧
∧ ⎡neutrons
⎤
n() t = [ λC() t + S(); t ] [ units] = * Vol of Core
3
β − ρ()
t
⎢
⎣ cm
⎥
⎦
(3)
where:
Λ = prompt neutron generation time ~ 10 -7 [sec]
1/λ = delayed neutron precursor lifetime ~ 10 [sec]
β = delayed neutron fraction ~ .002
β-ρ(t) = β - ρo - ∆ρ(t)
β-ρ o =
⎡∆k
⎤
reactivity offset from prompt critical
⎢
⎣ k ⎥
⎦
∆ρ(t) = feedback + external control reactivity
10 The adjustment will occur within 30 to 50 prompt neutron generation times for sub-criticality of 2 to 3% ∆k/k.
Given a generation time of ~10 -7 sec, prompt means several microseconds adjustment times for an ADS.
112

The prompt neutron population establishes equilibrium immediately (

Moreover, the fuel is where neutron and heat removal time constants clash continually; giving
rise to new requirements on the fuel also, specifically it must be structurally tough to thermal shocks,
and must have heat storage capacity to slow down heat release transients.
Controller options include traditional control rod actuators as well as actuators controlling source
strength or source importance (either spatial or spectral dependencies). As in the case of burn-up
reactivity swing compensation, the control actuator will likely be required to assume a “nuclear safety
grade” level of reliability.
In summary, the dependence on spallation source neutrons rather than on delayed neutrons to
maintain the fission chain reaction in balance from one fission chain generation to the next leads to
extremely abrupt response to control actions, reduced influence of power/flow dependent reactivity
feedbacks, and puts added importance on the fuel and on the control actuator itself to moderate the
vastly different time constants of nuclear and heat removal processes.
Beam reliability effects – Current multi megawatt proton beam accelerators have not been
designed for second-to-second reliability; they trip off many times a week due to accelerator cavity
sparking; they restart after a spectrum of time delays ranging from a fraction of a second to tens of
minutes [17]. Since ADS power scales linearly with the source strength (Equation 2) such source trips
lead to ADS power trips which in turn induce abrupt fuel and coolant temperature transients. Such
temperature transients induce thermal stresses in the core support and heat transport heavy-walled
structures; repeated trips give rise to life-shortening low cycle fatigue damage of these structures [18].
Moreover, if the restart delay exceeds several minutes, the balance of plant must undergo a major
restart procedure to connect to the grid [4].
Accelerator designers are devising means to reduce the frequency of trips – but do not foresee
means to reduce frequency to only a few per year (similar to frequency of unplanned reactor trips).
Thus, design options to mitigate their effects on the transmuter core and the heat transport circuits
have been studied. Options include multiple accelerators to avoid total loss of power given any single
accelerator trip. Other options include power density de-rating to lessen the amplitude of temperatures
swings. Thermal storage – in the fuel [18], in the coolant, and in the steam generator [4] – are also
considered so as to mitigate the abruptness of downstream temperature swings.
Application of passive safety principles – For liquid metal cooled ADS, the passive decay heat
removal strategies used for fast reactors apply without modification.
Passive power self-regulation based on thermostructural reactivity feedbacks which has been
exploited for fast reactor passive safety [19] is precluded by the fundamental feature of ADS subcritical
source-driven systems. For an ADS, the operating point is offset from prompt criticality by
(β - ρ o ) where -ρ o is the sub-criticality operating point. This is compared to an offset of only β for a
critical reactor. As is evident from observation of the denominator in Equation 3, the effect is to
decrease sensitivity of ADS power level to reactivity feedbacks as compared to a reactor. Moreover, as
is also evident from the inhomogeneous source term in Equations 3 and 4, the power can never be
driven to zero by reactivity changes as long as the spallation source is nonzero. These differences give
rise to a need in the ADS for different strategies for employing passive concepts to keep heat
production and removal in balance. Specifically, some means for source strength or source importance
to be adjusted passively in response to power changes is needed. Options include accelerator powered
by ADS-generated electricity 11 [20]; or source – transmuter coupling which is dependent on coolant
temperature or density. Absorber curtains or moderator curtains in the buffer surrounding the source or
target spatial relocations (all activated by coolant temperature or density changes) affect coupling and
might offer opportunities to apply passive source feedbacks analogous to the passive reactivity
11 The exceedingly long time constant for feedback presents a major challenge with this option.
114

feedbacks successfully exploited and demonstrated in fast reactors as the means to passively self
regulate the heat production rate to match the heat removal rate.
Activation product effects – Given the ADS function to reduce waste losses to the repository from
the fission energy complex per unit of useful energy from the complex, care must be taken in ADS
design to minimise production of incremental waste. This is a safety-related issue not only for the long
term, but for operational safety as well. Issues arise in choice of coolant [21], choice of spallation
target material [22] and level of sub-criticality operating state and in beam misalignment and halo
effects on activation of accelerator structures [23].
Recycle facility safety – The recycle and re-fabrication processes for TRU and MA fuel introduce
safety issues of criticality, pyrophoricity and atmosphere control; these do not differ in character from
the recycle safety issues for reactors intended for TRU or MA use. In either case, however, small
critical mass of fertile-free fuel and the demands on shielding and atmosphere control when working
with high concentrations of minor actinides (displaying characteristics of spontaneous fission, neutron
emission, and low temperature volatility) raise new challenges compared to current practice with UOX
or MOX fuel.
Accelerator safety – The accelerator brings with it the usual accelerator safety issues (highvoltage
safety, control of worker dose owing to components activated by beam divergence, etc.).
These issues are not peculiar to ADS applications except for shielding issues at the ADS/accelerator
beam tube interface, already discussed.
5. Summary
The work of the expert group in studying the safety issues of a specific class of ADS – that
employing fast neutron spectrum, solid fertile-free fuel, and multiple recycle – with a primary mission
of TRU or MA incineration has comprised an ancillary element of the OECD/NEA “Comparative
Study of Fast Reactors and ADS in Advanced Fuel Cycles”. Safety-related challenges which derive
from the distinguishing design features of the ADS for this specific mission have been identified. The
expert group has discussed safety strategy options available for addressing each safety-relevant issue
based on a presumption that safety should be designed in from the start; that relevant fast reactor
safety principles and practices should be applied where applicable and that safety of the entire cycle
(including recycle, refab, and waste disposal operations) should be kept in mind during each ADS
safety-related design decision.
While the work is still ongoing, multiple options for addressing nearly all issues have been
developed, drawing on experience from reactor safety approaches. An impact matrix (of ADS
distinguishing design feature vs. required safety function) has been developed, and a tracking of
distinguishing design feature back to specified ADS mission element has been produced. These
materials will be useful to designers and safety analysts in optimising the design of their ADS
concepts within their specific set of constraints and mission requirements.
Several issues merit special note. First, the issue of ADS dynamic response to reactivity or source
changes and the achievement of buffering between nuclear and thermo/structural time constants
without benefit of delayed neutron buffering is the area of greatest difference between fast reactor and
ADS safety-related characteristics and an area where no precedents exist in the fast reactor experience
base. Second, a prompt negative feedback mechanism for quenching HCDA sequences will have to be
developed for the ADS which will rely on phenomenology other than the traditional Doppler
coefficient of reactivity of fertile atoms in the fuel – or else the maximum support ratio requirement
which dictated fertile-free fuel could be relaxed. Finally, opportunities for application of passive safety
principles can be foreseen and should be exploited; straightforward adoptions are available for passive
decay heat removal; innovation will be required to achieve passive power self-regulation.
115

REFERENCES
[1] L. Van den Durpel et al., Activities of OECD/NEA in the Frame of P&T, Proceedings of the 6 th
Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN, OECD Nuclear Energy
Agency, Paris, France, (2001).
[2] Fistedis (ed), The Experimental Breeder Reactor II Inherent Safety Demonstration, Elsevier
Science Publishers, B.V. Holland, (1987).
[3] US-DOE, A Roadmap for Developing Accelerator Transmutation of Waste (ATW) Technology:
A Report to Congress, (Oct. 1999).
[4] H. Takano et al., Study on a Lead-bismuth Cooled Accelerator Driven Transmutation System,
Proceedings of the 6 th Information Exchange Meeting on Actinide and Fission Product Partitioning
and Transmutation, Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN, OECD Nuclear Energy
Agency, Paris, France, (2001).
[5] M. Salvatores, I. Slessarev, and K. Uematsu, A Global Physics Approach to Transmutation of
Radioactive Nuclei, Nucl. Sci. & Eng’g. 116, p. 1-18, (1994).
[6] a) R.N. Hill et al., Physics Studies of Weapons Plutonium Disposition in the Integral Fast
Reactor Closed Fuel Cycle, p. 17, Nucl. Sci. Eng’g., Vol. 121, #1 (Sept 1995).
b) R.N. Hill, D.C. Wade, E.K. Fujita, and H. Khalil, Physics Studies of Higher Actinide
Consumption in an LMR, Proc. Int. conf. Physics of Reactors, Marseille, France, April 23-27,
1990, p. I-83, American Nuclear Society French Section.
c) V.N. Mikhailov, V.M. Mourogov, V.S. Kagramanian, N.S. Rabotnov, V.Ya. Rudneva and
M.F. Troyanov, Plutonium in Nuclear Power Industry, IAEA Consultancy on Long-term
Options for Plutonium Disposition, Vienna, Austria, (1994).
[7] T.J. Moran, Core Restraint Design for Inherent Safety, Proceedings of the Joint American
Nuclear Power Conference, Myrtle Beach, South Carolina, (April 1988).
[8] E. Gluekler, US Advanced Liquid Metal Reactor (ALMR), Progress in Nuclear Energy, Vol. 31,
No. ½, pp. 43-61, (1997).
[9] a) W. Yang and H. Khalil, Neutronics Design Studies of Pb-Bi Cooled ATW Blanket,
Proceedings IAEA Technical Committee Meeting on Core Physics and Engineering Aspects of
Emerging Nuclear Energy Systems For Energy Generation and Transmutation, Argonne, IL,
(Nov. 28-Dec.1, 2000), (to be published).
b) R. Hill and H. Kahlil, Physics Studies for a Na-cooled ATW Design, Proceedings IAEA
Technical Committee Meeting on Core Physics and Engineering Aspects of Emerging Nuclear
Energy Systems For Energy Generation and Transmutation, Argonne, IL, (Nov. 28-Dec.1, 2000),
(to be published).
116

[10] a) R.A. Wigeland, Comparison of the SASSYS/SAS4A Radial Core Expansion Reactivity Feedback
Model and the Empirical Correlation for FFTF, Trans. Am. Nucl. Soc., 55, p. 423 (1987).
b) R.A. Wigeland and T.J. Moran, Radial Core Expansion Reactivity Feedback in Advanced
LMRs: Uncertainties and their Effects on Inherent Safety, ANS Topical Meeting on the Safety
of Next Generation Power Reactors, May 1-5, 1988.
[11] C.H.M. Broeders, A Comparison of Some Neutronics Characteristics of Critical Reactors and
Accelerator Driven Sub-critical Systems, Proceedings of the 5 th International Information Exchange
Meeting on Actinide and Fission Product Partitioning and Transmutation, EUR 18898 EN, Mol,
Belgium, 25-27 Nov., 1998, OECD Nuclear Energy Agency, Paris, France, (1999).
[12] G. Ritter et al., Comparison Study of Hybrid vs. Critical Systems in Point Kinetics, Proceedings
of the 5 th International Information Exchange Meeting on Actinide and Fission Product
Partitioning and Transmutation, EUR 18898 EN, Mol, Belgium, 25-27 Nov., 1998, OECD
Nuclear Energy Agency, Paris, France (1999).
[13] E. Gonzalez et al., Transuranics Transmutation on Fertile and Inert Matrix Lead-bismuth
Cooled ADS, Proceedings of the 6 th Information Exchange Meeting on Actinide and Fission
Product Partitioning and Transmutation, Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN,
OECD Nuclear Energy Agency, Paris, France, (2001).
[14] a) F. Lypsch and R. Hill, Development and Analysis of a Metal – Fuelled Accelerator-driven
Burner, Proceedings of the American Institute of Physics International Conference on Acceleratordriven
Transmutation Technologies and Applications, Las Vegas, Nevada, (July 1994).
b) H. Wider et al., Aspects of Severe Accidents in Transmutation Systems, Proceedings of the
6 th Information Exchange Meeting on Actinide and fission Product Partitioning and
Transmutation, Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN, OECD Nuclear Energy
Agency, Paris, France, (2001).
[15] J. Wallenius et al., Analysis of Nitride Fuels in Cores Dedicated to Waste Transmutation,
Proceedings of the 6 th Information Exchange Meeting on Actinide and Fission Product
Partitioning and Transmutation, Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN, OECD
Nuclear Energy Agency, Paris, France, (2001).
[16] W. Maschek et al., Safety Analysis for ADS Cores With Dedicated Fuel, and Proposals for
Safety Improvements, Proceedings IAEA Technical Committee Meeting on Core Physics and
Engineering Aspects of Emerging Nuclear Energy Systems for Energy Generation and
Transmutation, Argonne, IL (Nov. 28-Dec. 1, 2000), (to be published).
[17] OECD Nuclear Energy Agency, Workshop on Utilisation and Reliability of High Power
Accelerators, Proceedings, Aix-en-Provence, France, (Nov. 22-24, 1999), in preparation.
[18] a) F. Dunn, Design Criteria and Mitigation Options for Thermal Fatigue Effects in ATW
Blankets, Proceedings IAEA Technical Committee Meeting on Core Physics and Engineering
Aspects of Emerging Nuclear Energy Systems for Energy Generation and Transmutation,
Argonne, IL (Nov. 28-Dec. 1, 2000), (to be published).
b) T. Takizuka et al., Development of Accelerator-driven Transmutation System Concept and
Related R&D Activities at JAERI, Proceedings IAEA Technical Committee Meeting on Core
Physics and Engineering Aspects of Emerging Nuclear Energy Systems For Energy Generation
and Transmutation, Argonne, IL (Nov. 28-Dec. 1, 2000), (to be published).
117

[19] D.C. Wade and E. Fujita, Trends Versus Reactor Size of Passive Reactivity Shutdown and
Control Performance, Nucl. Sci. Eng’g. 103, p.182 (1988).
[20] A. Gandini, I. Slessarev et al., ADS Performance in the Safety and Reliability Perspectives,
Proceedings, OECD/NEA Workshop on Utilisation and Reliability of High Power Accelerators,
Aix-en-Provence, France, (Nov. 22-24, 1999), in preparation.
[21] V. Oussanov et al., Long-lived Residual Activity Characteristics of Some Liquid Metal Coolants
for Advanced Nuclear Energy Systems, Proceedings of Global’99, International Conference,
Jackson Hole, Wyoming (Sept 1999).
[22] a) M. Saito et al., Long-lived Spallation Products in Accelerator-driven Systems, Proceedings
IAEA Technical Committee Meeting on Core Physics and Engineering Aspects of Emerging
Nuclear Energy Systems For Energy Generation and Transmutation, Argonne, IL (Nov. 28-
Dec.1, 2000), (to be published).
b) Benlliure et al., New Data and Monte Carlo Simulations on Residue Production in Spallation
Reactions Relevant for Design of ADS, Proceedings of the 6 th Information Exchange Meeting on
Actinide and Fission Product Partitioning and Transmutation, Madrid, Spain, 11-13 Dec. 2000,
EUR 19783 EN, OECD Nuclear Energy Agency, Paris, France, (2001).
c) Gerasinov et al., Accumulation of Activation Products in Pb-Bi, Tantlalum, and Tungsten
Targets of ADS, Proceedings of the 6 th Information Exchange Meeting on Actinide and Fission
Product Partitioning and Transmutation, Madrid, Spain, EUR 19783 EN, 11-13 Dec. 2000,
OECD Nuclear Energy Agency, Paris, France, (2001).
[23] J. Klein, Structural Activation, Energy Deposition and Shielding Calculations Due to Proton
Beam Loss in A High Proton Power Linear Accelerator, Proceedings, OECD/NEA Workshop on
Utilisation and Reliability of High Power Accelerators, Aix-en-Provence, France, (Nov. 22-24, 1999)
(to be published).
118

POSTER SESSION
Partitioning
Chair: M.J. Hudson (University of Reading)
_____________________
SUMMARY
Amongst the developments within the poster presentations there was a high degree of novelty and
innovation. In all, there were twelve interesting poster presentations – the list for which is given
elsewhere. All of the presentations were of a high standard and the authors are to be congratulated for
the hard work that they put into the presentations. The paper by Suarez et al., for example, indicated
that selenium and zirconium isotopes remained in the raffinate within the PUREX process.
Caravaca et al. showed that within electrochemical processes in pyrochemical systems, when
LiCl/KCl is used as the electrolyte, the nucleation and crystal growth of the rare earth metal seems to
be the controlling step for deposition. Following the success of recent European Projects, which
studied amides, Almaraz et al. have managed to bind malondiamides onto calix[6]arenes which may
have potential as solvent extraction reagents. Using hollow fibre techniques, Geist et al. showed that
over ninety per cent of americium might be extracted from the feed phase when nPr-BTP is used as an
extractant. The kinetics of the extraction seemed to be rather slow. Dicarbolide studies are also
continuing and Plesek et al. showed that COSAN might be used for the separation of isotopes of
strontium and actinides. The influence of intermediate chemical processing of nuclear fuel has been
studied by Gerasimov et al. The extent of burn up increased with the amount of enrichment.
Song et al. discussed the developments that have been taking place at Tsinghua University. They
produced a flow sheet for the total partition process for commercial HLLW, which was the focus of
much attention.
The paper by Gruettner et al. indicated that nanotechnology should be considered more seriously
for the selective complexation of radionuclides. Especial importance must be directed in the future the
great potential that nanoparticles and nanostructured materials may have in the future. Thus the
particles themselves, or the functional groups on the surface may be used to interact with
radionuclides.
119

POSTER SESSION
Basic Physics: Nuclear Data and Experiments
and
Materials, Fuels and Targets
Chair: P. D’Hondt (SCK•CEN)
_____________________
SUMMARY
As mentioned during the poster session introduction, research on ADS has incited a revival of
interest in nuclear cross-sections of many nuclides in a large energy range with a special interest for
the higher energies.
In this poster session, 12 papers were foreseen from which 11 were presented, a success in itself. I
will go through each of them following the order of the programme to give you a flavour of what was
presented.
The first paper was on the n-TOF experiment at CERN and was concerned with the actual design
of the installation, with special emphasis on those aspects particular to the n-TOF: namely the
excellent energy resolution and the high-energy spectrum of the neutrons. Worthwhile to mention is
that the neutron energy spectrum induced by 20 GeV protons on a lead target, after a 200 m flight path
only contain 7% of the neutrons with energies higher than 20 MeV.
The second poster reported on recent capture cross-sections validation on 232 Th from 0.1 eV to
40 keV. It was shown that it is possible to determine cross-sections with a precision of 5% by making
use of a slowing-down spectrometer associated to a pulsed neutron source. These measurements
performed at ISN (Grenoble) with GENEPI show good agreement with ENDF/B-VI and JEF-2.2 in
the energy range 10 eV to 40 keV. Discrepancies were observed with JENDL 3.2 in the energy range
from 300 eV to 3 keV.
The third poster reported on double differential cross-sections for protons and light charged
particles emitted in reactions of 100 MeV neutrons on enriched 208 Pb targets. The measurements were
performed at Upsalla, Sweden. Preliminary results were presented for different angles relative to the
beam.
121

The fourth paper reported on measured double differential cross-sections of neutrons produced in
reactions induced by a proton beam of 62.5 MeV on a lead target. This experiment was performed on the
S-line of the CYCLONE facility of Louvain-la-Neuve in Belgium. Results were shown for 5 different angles.
The fifth paper reported on a joint work between Russian and Swedish colleagues. This work
focused on neutron-induced fission cross-sections of tantalum, tungsten, lead, mercury, gold and
bismuth. Results were presented for 10 different neutron energies ranging from the fission threshold
up to 175 MeV.
Paper 6 was entitled: “Neutron Radiative Capture Cross-sections of 232 Th in the Energy Range
from 60 keV to 2 MeV”. This paper reported on the work performed at the 4 MW Van der Graaf of the
CEN-Bordeaux. The activation technique was used and the cross-section was measured relative to the
197
Au(n,γ) standard cross-section up to 1 MeV. The results indicate that the cross-sections are close to
the JENDL database for values up to 800 keV and over 1.4 MeV. For energies in the intermediate
range, values are slightly lower to the ones from the libraries.
The seventh paper was related to the determination of the neutron fission cross-section for 233 Pa
from 0.5 to 10 MeV using the transfer reaction technique. This common work of CEN-Bordeaux,
CEN-Saclay and ISN-Grenoble is a first attempt to determine the neutron induced fission cross-section
of 233 Pa in the fast neutron energy range as a product of the fission probability of 234 Pa and the same
compound nucleus formation cross-section. Although the results are preliminary, they tend to agree
with the JENDL evaluation at least for energies greater than 4 MeV.
The experiment described in the eighth paper has been performed at the CYCLONE installation
of Louvain-la-Neuve. Double differential cross-sections for light charged particles production in
neutron induced reactions at 62.7 MeV on lead target were presented. Special attention was devoted to
the correction procedures coming from the use of a thick target and collimators. Measurements were
done with good statistics and are in good agreement with other experimental data. The comparison
with some well-known theoretical total production cross-sections data still shows large discrepancies.
Need for improvements of the theoretical models are still necessary.
The HINDAS project, which was presented as paper 9, will provide similar data for Fe and U. The
general objective of HINDAS, accepted within the 5 th framework programme, is to obtain a complete
understanding and modelling of nuclear reactions in the 20-200 MeV region, in order to build reliable
and validated computational tools for the detailed design of the spallation module of an ADS.
The tenth paper was also concerned with an accepted project within the 5 th
framework
programme, namely: the CONFIRM programme. This project aims at investigating the feasibility of a
high burn-up, high linear rating uranium free fuel, by means of modelling, fabricating and irradiating
transuranium nitride fuels. Some preliminary results from the safety analysis, pellet/pin design and
data requirements for irradiation modelling were presented.
Finally, the eleventh paper was concerned with one of the major problems in ADS namely the
large burn-up reactivity swing and the consequent unfavourable slanting of the radial power
distribution over the depletion period. In this work two concepts of the burnable absorber application
were considered, homogeneous and heterogeneous loading of B 4 C. The homogeneous application of
the B 4 C burnable absorber can be effectively used in reducing the burn-up reactivity swing but is not
favourable in terms of source neutron multiplication. Loading of burnable absorbers in the outer zones
can minimise the parasitic spallation neutron absorption as well as mitigate the slanting phenomenon
of the radial power distribution. Outer zone loading leads to longer cycle lengths compared to unpoisoned
reference cores.
122

To conclude, this poster session included new information on cross-section measurement. There
is still need for improvement of the theoretical models. New initiatives were presented in terms of
programmes and installations and I am looking forward to seeing new results at the next Information
Exchange Meeting in Korea.
123

POSTER SESSION
Transmutation Systems
Chair: Dr. T. Y. Song (KAERI)
_____________________
SUMMARY
In the poster session of transmutation system, all 18 papers were submitted. Many technical
aspects related to transmutation system were described.
ITEP submitted 7 papers which are mainly related to accelerator driven system. They cover
radiotoxicity and decay heat calculations of PWR spent fuels using Th, U and Pu. They also showed the
calculation results of Pu, Am, Cu, Tc and I transmutation in ADS, and suggested some ideas for ADS.
B.R. Bergelson from ITEP presented changes of radiotoxicity and decay heat power of actinides
from spent uranium and uranium-plutonium nuclear fuel of PWR-type reactors at long-term storage. In
another paper, he presented the same changes from spent thorium-uranium nuclear fuel.
A.S. Gerasimov from ITEP suggested the principal opportunity for development of a project for a
Demonstration Transmutation ADS (DTADS) as an international collaboration in Russia. He also
discussed the opportunity to use high thermal neutron flux for effective incineration of fission products
and minor actinides. He analysed weapon-grade plutonium burning and transmutation of the
americium and curium isotopes from spent fuel in reactor or accelerator-driven installations with
various neutron fluxes and spectra. Finally, he discussed the design of ADS complex for transmutation
of 99 Tc and 129 I. G.V. Kiselev from ITEP introduced ideas and suggestions related to ADS technology.
They include broad range of topics such as accelerator, target, sub-critical blanket, sectioned blanket,
necessity of high neutron flux, two blankets installation with fluid fuel etc.
Two other papers dealed with gas-cooled reactors and one paper dealed with the topic of molten
salt. A. Baxter from General Atomics reported work on the development of two concepts using
helium-cooled reactor technologies for transmutation. Both concepts make use of thermal and fast
energy spectra. D. Ridikas from CEA discussed gas-cooled target and assemblies, and considered both
fast and thermal sub-critical assemblies. It was suggested that the best features of both critical and subcritical
systems can be merged by combining the GT-MHR with an accelerator driven sub-critical
assembly. V. Ignatiev from Kurchatov Institute reported molten salt reactor developed in the
framework of the ISTC#1606. ISTC#1606 includes experimental study of the key properties of the
selected molten salt fuel composition, experimental verification of structural materials and physics &
fuel cycle considerations on molten salt reactor.
125

Five other papers related to ADS covered kinetics, sub-criticality measurement, fuels for Energy
Amplifier, Pb-Bi coolant and comparison of neutron sources. Only one paper covered transmutation in a
fast reactor. J. Blazquez from CIEMAT remarked the subtleties behind the questions related to ADS
such as sub-criticality, spallation source, kinetic parameters etc. Y. Rugama from Universidad
Politecnica de Valencia presented an absolute measurement technique for the sub-criticality
determination based on the Stochastic Neutron and Photon Transport Theory. S. Kaltcheva-Kouzminova
from Petersburg Nuclear Physics Institute presented neutronics calculations of the accelerator driven
reactor core EAP-80 with UO 2 & PuO 2 MOX fuel elements and Pb-Bi coolant. A. Pena from ETSII e IT
showed calculation results obtained by using two different CFD codes for Pb-Bi coolant in ADS. It
shows that Pb-Bi coolant circulation by buoyancy forces is an important result. M. Dahlfors from
Uppsala University showed a preliminary comparative assessment relevant to the transmutation
efficiency of plutonium and minor actinides. It has been performed in the case of ANSALDO’s Energy
Amplifier Demonstration Facility with two different neutron sources. A.V. Lopatkin from Research and
Development Institute of Power Engineering introduced RDIPE’s work on a concept of a fast leadcooled
reactor with UN-PuN fuel (BREST series).
The other two papers are related to comparison of BR2 and MYRRHA in transmutation
characteristics, and usage of ADS for proliferation creation of MOX fuel. Ch. De Raedt from SCK•CEN
presented the performances of the high flux materials testing reactor BR2 and compared them with those
of the ADS prototype MYRRHA in its present development stage. Finally, V.B. Glebov from MEPhI
evaluated the potential of the accelerator driven systems for enhancing the proliferation resistance of
LWR MOX fuel. 232 U is added to the MOX fuel and irradiated in the ADS blanket to create the inherent
reaction barrier.
126

CD-ROM CONTENTS AND INSTRUCTIONS FOR USE
This CD-ROM contains all the papers presented during the 6 th OECD/NEA Information Exchange
Meeting on Actinide and Fission Product Partitioning and Transmutation. The content of the booklet
accompanying this CD-ROM is also available in electronic form on this CD-ROM.
A composite file enables the reader to do full text searches in all the papers and summaries of this
meeting. Individual papers may be printed by selecting those in the programme overview.
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127

SESSION I
OVERVIEW OF NATIONAL AND
INTERNATIONAL PROGRAMMES
J.L. Diaz-Diaz (CIEMAT) – Ph. Savelli (OECD/NEA)
129

RESEARCH AND DEVELOPMENT OF TECHNOLOGIES FOR PARTITIONING
AND TRANSMUTATION OF LONG-LIVED NUCLIDES IN JAPAN
– STATUS AND EVALUATION –
Sanae Aoki
Director for Planning,
Radioactive Waste Policy Division, Science and Technology Agency (STA)
2-2-1, Kasumigaseki, Chiyoda-ku ,Tokyo, 100-8966, Japan
E-mail: s3aoki@sta.go.jp
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1. Current activities for radioactive waste management
Measures to treat and dispose of radioactive waste are one of the most important issues in the
development and application of nuclear energy. The Atomic Energy Commission (AEC) has carefully
considered how to classify radioactive waste properly and how to dispose of it according to these
classifications.
Japanese basic policy regarding disposal of high-level radioactive waste (HLW) is to solidify it
into stabilized form, to store it for 30-50 years to be cooled, and to dispose of it deep to the
underground (geological disposal).
In April 1997, the Advisory Committee on Nuclear Fuel Cycle Back-end Policy, AEC, laid down
the guidelines on future research and development of the disposal of HLW. In accordance with the
report, the Japan Nuclear Cycle Development Institute (JNC released the report on the outcome of
R&D activities to elucidate technological reliability of the geological disposal, and to provide
technical ground for selecting repository sites and for establishing safety requirement in the form of
the second progress report in December 1999, with the co-operation of related institutions, such as the
Japan Atomic Energy Research Institute (JAERI), the Central Research Institute (CRIEPI), the
Geological Survey of Japan (GSJ), the National Research Institute for Earth Science and Disaster
Prevention (NIED), and university researchers.
In parallel with the R&D programme, there has also been an effort to make a system for
implementing HLW disposal. In the Special Committee on High-Level Radioactive Waste Disposal,
AEC, and in the nuclear sub-committee of the Advisory Committee for Energy, Ministry of
International Trade and Industry, various aspects of HLW disposal were considered, including social
and economic aspects. A law 1 for implementing geological disposal was passed in the Diet in
May 2000. Based on the law, an implementing entity 2 for HLW disposal was established in
October 2000. The programme then moves from the generic into the site-specific phase. Thereafter we
are looking to start operation of the repository between 2030 and the mid-2040s at the latest.
2. Status of partitioning and transmutation study
At the same time, recognizing the nature of the radioactive nuclides contained in HLW, the aim
since the early days of nuclear energy has been to develop technology either to separate useful
elements and nuclides in order to re-use them effectively, or to transmute long-lived nuclides into
short-lived or stable – i.e. non-radioactive – forms by irradiation.
In Japan, reference to P&T technology for long-lived and other nuclides first appeared in the
Long-term Programme for Nuclear Research, Development and Utilisation (or “long-term nuclear
programme”) back in 1972. That programme noted the need for research and development in order to
ensure effective processing of radioactive waste.
In a 1976 interim report by the AEC’s Technical Advisory Committee on Radioactive Waste, the
relationship between P&T technology and the disposal of radioactive waste was specifically pointed
out. Specifically, if radioactive nuclides in waste could be appropriately separated into groups, waste
management could become more flexible. This is because the amount of radioactive nuclides requiring
strict control would be reduced, and treatment and disposal appropriate for the half-life of each group
would become possible. In addition, if long-lived nuclides could be transmuted into short-lived ones
by nuclear reaction, the burden of long-term waste management could also be reduced.
Specified Radioactive Waste Final Disposal Act.
Nuclear Waste Management Organisation of Japan (NUMO).
1
2
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The long-term nuclear programme issued in 1987 stated that P&T technology was very important
from the viewpoint of recycling HLW and enhancing disposal efficiency. It also stated that systematic
R&D would be carried out jointly by JAERI, the then Power Reactor and Nuclear Fuel Development
Corp. (PNC, now JNC) and others.
Under that programme, in 1988, the AEC’s Advisory Committee on Radioactive Waste Measures
issued a report entitled, Long-term Programme for Research and Development on Nuclide Partitioning
and Transmutation Technology 3 This can be considered to have been the first systematic R&D
programme on P&T technology in Japan. It presented a plan for R&D that ran from 1988 to 2000 and
was divided into two phases: Phase I, covering the first four to nine years, which included evaluation
of various concepts and R&D on key technologies; and Phase II, covering the next four to nine years,
which included engineering experiments on key technologies and demonstrations.
The long-term nuclear programme issued in 1994 stated that each research institute would carry
out basic studies on P&T technologies and evaluate each technology at some time in the mid-1990s to
determine how to proceed thereafter.
Meanwhile, in 1998, the AEC’s Special Committee on the Disposal of High-Level Radioactive
Waste released Basic Concepts in the Disposal of High-level Radioactive Waste. This stated that, in
order to gain public understanding of disposal technology, “research on waste reduction with the aim
of achieving safer and more efficient geological disposal, as well as more efficient use of waste, would
be carried out on a regular basis”. It also said that it was “important to have mechanisms to respond
flexibly to any dramatic progress in P&T technology”.
Under these circumstances, the AEC’s Advisory Committee on Nuclear Fuel Cycle Back-end
Policy investigated and considered matters concerning P&T technology for long-lived and other
nuclides, based on the evaluation schedule stated in the long-term nuclear programme issued in 1994.
The Committee issued a report entitled, Research and Development of Technologies for Partitioning
and Transmutation of Long-lived Nuclide Status and Evaluation Report in March 2000. The brief
summary of this report is as follow.
3. Results to date and analyses of current status
3.1 Elements subject to P&T
Long-lived (i.e. long half-life). In particular:
• High radiotoxity due to the emission of rays.
• Fast migration through geological formations via underground water, when disposed of
geologically.
Relatively short-lived and heat-generating, producing most of the heat in HLW.
Rare and useful elements.
3
Called the “OMEGA Programme” – an acronym for Options Making Extra Gains from Actinides and Fission Products.
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3.2 Results to date and analyses of current status
3.2.1 The partitioning process
3.2.1.1 JAERI
JAERI is developing a four-group partitioning process, in which elements in concentrated highlevel
liquid waste HLLW are separated into four groups: MAs, Tc-platinum group metals, Sr-Cs, and
others. Basic experiments using simulated HLLW helped to establish the concept. Test runs with both
simulated and actual HLLW on a scale 1/1000 that of an actual plant confirmed the expected
capabilities of group partitioning. A recovery rate for MAs of 99.95% or better was achieved.
3.2.1.2 JNC
JNC is developing an improved PUREX process – an advanced version of the conventional
reprocessing process – to recover Np. It is also developing a CMPO-TRUEX process to separate MAs
from highly concentrated HLLW and an electrolytic extraction method to separate Tc-platinum group
and other elements from aqueous reprocessing solutions. In the CMPO-TRUEX process research, it
was demonstrated that Am and all nuclides can be recovered to a level of 99.9% or more under
standard extraction conditions.
3.2.1.3 CRIEPI
CRIEPI is developing a reductive-extraction process using molten chlorides and liquid metal
solvents. Basic data were obtained for the behavior of elements in this type of molten-salt-and-liquidmetal
system. Experiments on the separation of recovery of TRUs were carried out using some
10 milligrams of TRUs and some 100 grams of chlorides, which confirmed recovery of more than
99% of the TRUs and adequate separation of TRUs from REs.
3.2.2 The transmutation cycle
JNC and CRIEPI are studying transmutation technology using fast breeder reactors (FBRs). This
concept is centered on the use of fast reactors for electricity generation. In this scenario , FBR will
take over the role of light-water reactor (LWR) in future, with power generation and the transmutation
of MAs and other elements carried out simultaneously by the FBR. In contrast, JAERI’s concept is
the “double strata fuel cycle”, where a dedicated system for transmuting MAs, such as an actinide
burner fast reactor (ABR) or an accelerator-driven subcritical system (ADS) is at the centre of the
transmutation cycle, allowing a commercial power generation cycle and a transmutation cycle to be
developed and optimized independently for their individual purposes.
3.2.3 The fuel production process
3.2.3.1 JAERI
JAERI is developing MA-nitride fuel. The basic data on thermal properties of MA-nitrides
necessary to design the fuel, as well as thermal properties of Tc alloys, were obtained. It was
confirmed that the MA nitrides could be prepared by means of carbothermic synthesis and uranium
nitrides microspheres can be produced via the sol-gel process. In addition, through irradiation tests of
U-Pu mixed nitride fuel produced on a trial basis, it was ascertained that the fuel element was intact
after a burn-up rate of 5.5% achieved.
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3.2.3.2 JNC
JNC is developing fuel in which Np and/or Am is added to the MOX fuel that JNC has developed
for FBRs. For the addition of Np to MOX fuel, a vibro-packing process is being developed in a joint
international research effort. For the addition of Am to MOX fuel, in addition to irradiation
experiments being carried out at the experimental fast reactor “Joyo”, remote fuel fabrication facilities,
including those to produce pellets and inspect fuel pins, were established and performance tests were
carried out.
3.2.3.3 CRIEPI
CRIEPI is developing metallic fuel – an MA-content U-Pu-Zr ternary alloy under the
international collaborations. Fuel pins have been made on a trial basis and physical characteristics and
other basic data have been obtained. It has been determined that an MA content of about 5% does not
affect the characteristics of the fuel, and it is expected that MAs can be mixed homogeneously during
fabrication of the fuel.
3.2.4 The transmutation process
3.2.4.1 JAERI
JAERI is developing concepts for ADSs and ABRs. Nuclear data on MA nuclides were obtained
through international co-operation and were verified while developing a database, carrying out integral
experiments and analysing irradiated MA samples. In the development of a proton accelerator for the
ADS, major key technologies were developed, and the highest level performance has been
demonstrated.
3.2.4.2 JNC
JNC is developing a MOX-fuelled FBR. The prototype reactor “Monju” already exists – i.e.
construction of a “prototype plant” has been achieved – but the development of other key technologies
will be required to add MAs to MOX fuel. Nuclear data on MA nuclides were obtained and evaluated
via nuclear reactors and accelerators. Design studies were carried out on acceptable amounts of MAs
and rare-earth elements, and on actual fuel loading.
3.2.4.3 CRIEPI
CRIEPI is developing the concept of metallic-fuelled FBRs. Regarding nuclear data on MA
nuclides, analysis programmes for MA transmutation were developed and analyses were carried out.
3.2.5 Fuel processing
3.2.5.1 JAERI
JAERI is developing a pyrochemical process similar to dry reprocessing. Molten-salt electrolysis
of U-nitride, Np-nitride and Pu-nitride on a gram scale were carried out to confirm that transuranic
metals can be recovered. For the recovery and recycling of N-15, nitrogen (N 2 ) release in molten salts
was studied, and it was confirmed that almost 100% of the nitrides can be released in the form of N 2 .
135

3.2.5.2 JNC
JNC is considering the same method as for the partitioning process.
3.2.5.3 CRIEPI
CRIEPI is developing molten-salt electrorefining and reductive extraction, which is similar to the
partitioning process. Electrorefining forms the main part of the pyro-reprocessing method. Feasibility
was confirmed through an in-house study and international joint research, and the process is now at
the stage of engineering experiments. Feasibility of the conversion of oxides of spent fuel to metal,
and of the spent-salt treatment process, are still to be confirmed.
3.2.6 Conclusion
R&D at the three research institutes has resulted in establishment of processes for P&T
technology with the expected performance. The aims of Phase I R&D have thus been achieved. R&D
in Phase II has experienced some delays, the primary reasons being that Japan is redefining its entire
FBR programme, and facilities to handle MAs and other materials have yet to be constructed. In
carrying out further R&D, it is important to promote cooperation with domestic and foreign
organizations in order that experimental facilities – including those for engineering experiments – can
be used efficiently.
3.3 Technical issues
The implementation of experiments to demonstrate processes using actual HLLW is an issue
common to the three organizations. Also common to JAERI and JNC, which are developing aqueous
partitioning processes, are development of a method to more efficiently separate MAs and rare-earth
elements, and technologies to reduce the volume of secondary waste. At CRIEPI, where dry
partitioning is being developed, the main technical issues in the handling of molten chlorides are
material development and molten-salt transport technology. Common issues to the three organisations
are preparation of a database on fuel irradiation behavior for performance analysis, and development
of fuel fabrication technology.
4. Effects and significance of partitioning and transmutation technology
4.1 Radioactive inventory in waste
It is a social imperative to minimise, as far as possible, the generation of hazardous waste
produced by industrial activities. Reduction of long-term radioactive inventory through the removal of
long-lived nuclides from HLW by P&T technology helps to meet this requirement. If, for example,
99% of MAs contained in spent fuel can be removed, the toxicity of the spent fuel after several
hundred years following reprocessing will be equal to the toxicity of the same amount of natural
uranium as used in the production of the original fuel.
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4.2 Effects on geological disposal
4.2.1 Long-term safety
• Effects on the underground water migration scenario:
Maximum dose can be reduced by about two orders of magnitude by separating and removing
99% of the 135 Cs, and 99% or more of the Np and Am, which are parent nuclides of 229 Th.
• Effects on the human intrusion scenario:
Risks can be reduced by two orders of magnitude by separating and removing 99-99.99% of
the actinide elements.
4.2.2 Impact on geological disposal site design
Approximately two-thirds of the heat from HLW is generated by Cs and Sr, and separation and
removal of these elements would shorten the required storage period and reduce the size of the site.
The HLW storage period could be reduced by separating and removing exothermic nuclides. In
addition, disposal site design could be rationalized by, for example, disposing of the HLW in one large
cavity rather than in several smaller ones.
4.3 Effective use of resources
Among the materials in HLW, some can be used effectively as resources. For example platinum
group elements are widely used as catalysts to reduce nitrogen oxides in vehicle exhaust gases, and so
on. However, prior to actual application, clearance level issues are to be solved.
4.4 Reduction of MA and LLFP inventory, and the times required
Even if P&T technology is employed, some MAs and LLFPs will remain in the waste, and final
disposal of such waste will eventually be necessary. An oxide-fuelled or metallic-fuelled fast reactor
can transmute MAs from more than five or six LWRs every year. While an ADS with one quarter of
thermal output of a LWR can transmute MAs from more than ten LWRs (about 250 kg per year).
4.5 Generation of secondary waste
The aqueous partitioning process, like PUREX reprocessing, generates secondary waste.
Compared with reprocessing, however, the volume of secondary waste is lower because P&T
technology deals with the very limited quantities of HLLW generated by reprocessing.
In the dry-type partitioning process, molten salts and liquid metals are used as solvents, and
metallic Li is used as a reducer. These solvents and reducers are less susceptible to degradation by
radiation, and may be recycled in the system. But there is almost no experience of using this process
on an industrial scale.
To evaluate the secondary waste volume, further investigations and experience using actual
materials are needed.
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4.6 Short-term increase in radiation dose
When P&T technology is employed in the nuclear fuel cycle, exposure dose could increase in the
short term as new processes and facilities are introduced and as the volume of MAs and LLFPs to be
dealt with in such processes increases. It will be possible, however, to keep exposure dose for workers
and the public below the statutory standard, and as low as reasonably possible, by measures such as
enhanced shielding.
4.7 Economy
R&D to date has focused on basic studies to obtain fundamental data for designing processes and
systems, and no one is in a position to make a reliable cost estimate for the P&T technology.
Nevertheless, very preliminary cost estimates of three organisations indicate a few percent of LWR
electricity generation cost for PT implementation.
5. Future research and development
Modern society demands maximum control of hazardous waste produced by industry, as well as
recycling to preserve resources and protect the environment. It goes without saying that the nuclear
industry, too, must take all effective measures. P&T technology applied to long-lived nuclides can be
useful in, for example, reducing the long-term radioactive inventory in nuclear waste, and it is
appropriate that R&D should be carried out on an ongoing basis.
5.1 P&T technology and the nuclear fuel cycle
P&T technology is a part of the nuclear fuel cycle. The question of how and in what part of the nuclear
fuel cycle P&T technology should be incorporated in order to optimize the cycle, should be considered.
The purpose of R&D is to suggest scenarios for introducing P&T systems into the nuclear fuel cycle, and to
develop designs and establish key technologies for such systems. Keeping the totality of the fuel cycle in
mind, it is necessary to correctly evaluate issues of economy, energy security, and reduction of the
radioactive inventory in waste, and to analyse the trade-off among those factors.
5.2 Research on system design and development of key technologies
R&D for P&T technology consists of research on system design with the aim of introducing this
technology into the nuclear fuel cycle, and development of the key technologies necessary to realise
the system. Development of key technologies takes time and should be carried out in a progressive
fashion, while at the same time being properly integrated with research on system design.
5.3 How to proceed R&D in future
P&T technology based on the use of power-generating fast reactors, and P&T technology based
on the double-strata fuel-cycle concept, on which R&D is being carried out in Japan, have their own
distinctive features. These two concepts also provide new options for the fuel cycle and it is therefore
appropriate at this stage to continue the development of both. The objectives of further R&D are to
study scenarios, including a possible blending of these two concepts, in order to introduce a feasible
P&T technology system into the nuclear fuel cycle, and to develop the necessary technologies. System
138

design and P&T introduction scenarios will continue to be studied. According to the scenarios thus
defined, and based on the results of R&D to date, small-scale experiments to demonstrate the
feasibility of a series of processes will then be conducted. Following this, systems whose feasibility
has been successfully demonstrated will be subjected to engineering tests in order to obtain data on
their safety. It will be important at this stage to manage the R&D under a system of checks and
reviews, and to update the scenarios on a regular basis.
5.4 Co-operation for R&D
Although there are differences in the concepts, reactor types and systems, many common issues
exist in R&D on P&T technology. JAERI, JNC and CRIEPI should work together in an effort to resolve
those common issues, strengthening their co-operation through the sharing of their R&D results. At the
same time, it is important to carry out the R&D effectively by, for example, cooperating with other
domestic organizations and using their existing facilities. In Western nations, the trend is toward
international co-operation in various areas of P&T technology R&D, and the three organisations should
actively join such cooperation frameworks and make use of research facilities available overseas. They
are also encouraged to exchange information through the existing OECD/NEA framework.
5.5 R&D schedule and evaluation
P&T technology is inseparable from the nuclear fuel cycle, and it is therefore appropriate to
conduct R&D in this area on a time schedule compatible with nuclear fuel cycle R&D. At present,
feasibility study on commercialised FBRs and related fuel cycle system is being carried out under the
collaborative efforts of JNC, electric utilities, CRIEPI and JAERI. In this study, R&D scenarios
toward commercialization of fast reactor system will be reviewed by about 2005. Thus, around the
year 2005 is deemed to be an appropriate time to reconsider all R&D scenarios of PT including the use
of FBRs for transmutation together with power generation, and the double-strata fuel cycle. Thereafter,
progress, results and R&D policy will be checked and reviewed every five years or so. Evaluations of
P&T technology system concepts, and reviews of introduction scenarios, should also be conducted.
6. Concluding remarks
P&T technology belongs to a world quite different from that of ordinary chemical reactions in the
sense that, in P&T, materials are transformed at the atomic level. Its potential is not limited simply to
transmutation. In addition, its development raises issues that will be difficult to resolve with existing
technology alone, overcoming these difficulties could lead to other exciting technological breakthroughs.
This, indeed, should be one reason for young engineers scientists to want to become involved in the
field. Research on this kind of advanced technology can be expected to make a great contribution to
revitalizing nuclear research generally. P&T technology research should be actively promoted in order to
nurture the development of human resources in the nuclear field.
In doing this, however, it is important to create an environment in which innovative ideas can be
adopted without “interference” from existing systems. P&T technology requires open-minded R&D.
Inspired by Japan’s OMEGA programme, many similar programmes have been established around
the world – in France, other European countries, and the United States. It is expected for Japan to play an
important on-going role in this area – and to do so as part of her international contribution, while
conducting timely evaluation of R&D progress.
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Annex 1
R&D scheme for partitioning and transmutation of long-lived radioactive waste in Japan
JNC,CRIEPI
✟ Advanced fuel cycle
✟ Commercial FR for transmutation
✟ Oxide fuel, Metal fuel
✟ 2~5% MA content (homogeneous)
Advanced fuel cycle
JAERI
■ Double-strata fuel cycle concept
❐ Combination of a power reactor fuel cycle
& an independent P-T cycle
✟ Dedicated systems for P-T
✟ Accelerator-Driven System (ADS)
Double strata fuel cycle
1st Startum of Fuel Cycle
LWR
Common technologies
LWR
1000MWe
LWR 10- 14units
Fuel Fabrication
Plant
MOX- LWR
FR
U,TRU(P u,MA),LLFP
Advanced
Reprocessing
HLW
Final Disposal
MA ; Minor Actinide
LLFP; Long- lived FP
SLFP; Short- lived FP
R&D items to be developed
in collaboration
✟ Separation chemistry
✟ Reactor physics
✟ Fuel basic property
✟ Irradiation test
✟ Nuclear data
Fuel Fabrication
Plant
2nd Startum of Fuel Cycle
(P-T Cycle)
Spent
Fuel
Storage
Dry Reprocessing
Plant
Final Disposal
MOX- LWR
FR
99.5%U,P u
U,P u
MA
LLFP
Transmutation
System Fuel Fabri-
MA Nit ride
800- 1000MWt cation Plant
MA,LLFP
SLFP
Reprocessing
Plant
HLW(TRU,FP )
P a r t it ioning
P la nt
HLW
Storage
140

Annex 2
JAERI JNC CRIEPI
141
Elements Partitioning MAs (Np, Am, Cm), Pu, Tc
MAs (Np, Am, Cm), Pu, Tc MAs (Np, Am, Cm), Pu
subject to
Platinum group (Ru, Rh, Pd), Sr, Cs Platinum group (Ru, Rh, Pd)
P&T Transmutation MAs (Np, Am, Cm), Tc, I MAs (Np, Am, Cm), Pu, Tc, I MAs (Np, Am, Cm), Pu
Partitioning process
4-group partitioning process
(wet process)
Advanced reprocessing nuclide
partitioning system (wet process)
MA separation DIDPA extraction CMPOTRUEX process
Improved PUREX process
Dry process
Reductive extraction
(molten salt/liquid metal)
Tc-Platinum group Precipitation by de-nitration Electrolytic extraction –
Sr-Cs
Column absorption with
inorganic ion exchangers
– –
Transmutation cycle Double strata fuel cycle MOX fuelled FBR Metallic-fuelled FBR
Fuel type MA-nitride fuel MA-MOX fuel U-Pu-Zr ternary alloy
Transmutation
process
Accelerator driven sub-critical system (ADS)
Actinide burner fast reactor(ABR)
FBR
Fuel processing Molten-salt electrolysis (pyroprocess) Wet process Molten-salt electro-refining and
reductive extraction
FBR

Annex 3
Members of the Advisory Committee on Nuclear Fuel Cycle Back-end Policy
Atomic Energy Commission of Japan
Yumi Akimoto
Kenkichi Ishigure
Nobuo Ishizuka
Michiko Ichimasa
Yoichiro Ohmomo
Yoshiaki Oka
Takeki Kawahito
Keiji Kanda
Tomoko Kusama
Nobuaki Kumagai
Keiji Kojima
Osamu Konishi
Kisaburo Kodama
Shinzo Saito
Shiro Sasaki
Atsuyuki Suzuki
Hiroshi Sekimoto
Satoru Tanaka
Yasumasa Tanaka
Akira Tokuyama
Hiroyuki Torii
Yasuo Nakagami
Tadashi Nagakura
Kunio Higashi
Junsuke Fujioka
Hajimu Maeda
Miyako Matsuda
Hirotake Moriyama
Yoshiaki Yamanouchi
President, Mitsubishi Materials Corporation
Professor, Saitama Institute of Technology
Member of Board and Secretary General,
Japan Atomic Industrial Forum, Inc.
Professor, Ibaraki University
Senior Executive Director, Institute for Environmental Sciences
Professor, University of Tokyo
Chairman, Radioactive Waste Management Center
Professor, Kyoto University
President, Oita University of Nursing and Health Sciences
Professor Emeritus, Osaka University
Representative, Geospace Laboratory
Editor, Nippon Hoso Kyokai (Japan Broadcasting Corporation)
Director General, Geological Survey of Japan
Vice President, Japan Atomic Energy Research Institute
Technical consultant, Japan Nuclear Fuel Limited
Professor, University of Tokyo
Professor, Tokyo Institute of Technology
Professor, University of Tokyo
Professor, Gakushuin University
President, Fuji Tokoha University
Editorial Writer, Nihon Keizai Shimbun, Inc.
Executive Vice President,
Japan Nuclear Cycle Development Institute
Senior Advisor Emeritus, Central Research Institute of the
Electrical Power Industry
Professor, Kyoto University
Managing Director, Japan Radioisotope Association
Chairman, Nuclear Task Force,
Federation of Electric Power Companies
Commentator on consumer and environmental affairs
(Issues related to waste and recycling)
Professor, Kyoto University
Attorney at Law
142

FRENCH RESEARCH PROGRAMME TO REDUCE THE MASS AND
TOXICITY OF LONG-LIVED HIGHLY RADIOACTIVE NUCLEAR WASTE
Jacques Bouchard, Director of Nuclear Energy
Patrice Bernard, Director of Nuclear Development and Innovation
CEA, Nuclear Energy Division
31, rue de la Fédération, 75752 Paris Cedex 15
France has launched a process of optimising waste management by separating and recycling
recoverable energy materials, reducing, conditioning and storing final waste. In addition, it has initiated
operations to clean up and dismantle older facilities (first-generation reactors, cycle plants, etc.) through
the development of related technologies.
French research and industry have developed processes and technologies which ensure
downstream management of the fuel cycle and waste (reprocessing of spent fuel, recycling of
plutonium, processing and conditioning of waste from nuclear plants and cycle facilities,
development of containers and interim storage facilities – clean up and dismantling).
Once operations in the downstream part of the cycle have been completed, long-lived highly
radioactive waste (vitrified class C waste) represents the major portion of the waste’s total
radioactivity, conditioned in a small volume (

plutonium can be recycled in pressurized water reactors on a recurring basis under
economically acceptable conditions.
• Minimising long-lived highly radioactive final waste – studies here concern separation
(intensive chemical separation during reprocessing for which new very selective molecules
have been developed), transmutation (transformation in industrial or specialized nuclear
reactors into non-radioactive elements or with a much shorter life), specific conditioning
(incorporation of separated elements, which cannot be transmuted, within the crystalline
network of almost unalterable materials on a time scale characteristic of disappearance through
radioactive decay) of the main long-lived radionuclides (minor actinides and certain very longlived
fission products abundant in spent fuel and potentially more mobile in the environment 3 )
present in highly radioactive waste;
Research is conducted with the goal of establishing by 2006 the scientific feasibility of
transmutation in various types of nuclear reactors (PWR, innovative reactors) and the technical
feasibility of intensive separation downstream from reprocessing at La Hague, as well as of the
specific conditioning of separated long-lived radionuclides.
Studies are conducted in the framework of the French law of December 30, 1991, on the
management of long-lived highly radioactive waste, which established a structured research programme
with three focuses: separation-transmutation, storage in deep geological formations, conditioning and
long-term interim storage.
• Focus 1 studies the various solutions envisaged to ensure a substantial reduction of the mass
and toxicity of waste, for the same amount of energy produced, by separating and transmuting
the waste.
• Focus 2 studies storage in deep geological formations, a situation which could become
definitive in the absence of human intervention, since the geological medium makes it possible
to ensure containment on a time scale characteristic of long-lived radionuclides, while
maintaining for a certain period an option of reversibility so that our descendants might be able
to recover the packages.
• Focus 3 concerns management procedures under the responsibility of the society, in long-term
interim storage facilities above or under ground, which make it possible to protect the packages,
by having previously conditioned them in a form which ensures long-lasting containment and
the possibility to recover the packages under conditions that are safe and defined by technical
specifications.
The law defined a calendar which stipulates that a comprehensive report evaluating research will
be submitted to the French Parliament in 2006. Public authorities have designated a pilot for each
focus: the French Atomic Energy Commission (CEA) for Focuses 1 and 3 and Andra for Focus 2.
This research is conducted in co-operation with partners in the nuclear industry, EdF, COGEMA
and FRAMATOME, as well as with CNRS and universities. It benefits from significant co-operation
at the European and international level. It is constantly evaluated by the National Evaluation
Commission, which draws up and publishes an evaluation report annually.
3
Mainly iodine 129, caesium 125, technetium 99.
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The procedure involves identifying a set of complementary scientific and technical solutions,
which serve to define open and flexible strategies for the downstream part of the cycle and lay the
groundwork for a decision in 2006.
Concerning partitioning research, if studies on uranium and plutonium separation from the other
fission products depend on a mature chemical process, intensive separation of minor actinides and
these three fission products was not possible using industrial processes. Through studies conducted
since 1991, a reference programme was defined for an advanced separation process for the main longlived
radionuclides present in waste:
• Neptunium, iodine and technetium could be separated by adapting the PUREX process used
industrially in the reprocessing facilities at La Hague.
• To separate americium, curium and cesium, it was necessary to develop new chemical
separation processes by devising very selective molecules capable of separating these elements.
The families of extractors were defined, the principal reference molecules synthesised, and their
performances verified experimentally on real radioactive solutions in the ATALANTE facility at
CEA-Marcoule, in order to reach the stage of scientific feasibility (2001). The next stage will be that
of technical feasibility, moving from the molecule to the overall chemical process, which will be
defined and validated in 2005. The existence of the reprocessing industry makes the implementation
of these processes a real possibility.
In addition, research conducted in recent years has pointed up the performances of the control of
plutonium and of transmutation in different types of electronuclear power plants, showing:
• That it is possible to stabilize over time the quantity of plutonium by consuming it completely
with advanced plutonium fuel; an overall balance can thus be reached between the formation
and the consumption of plutonium; the amount of final waste is consequently divided by three
compared with the open cycle.
• That by separating and multi-recycling the minor actinides in a reactor (in order to transmute
them), the mass and toxicity of the waste is divided by 100 on a similar basis, and studies show
that the innovative reactors (electricity-producing reactors or those dedicated to transmutation)
which present the characteristics adapted to these performances (great capacity to consume
plutonium as well as long-lived radionuclides and ability to use to the best advantage the
energy contained in the fuel: rapid spectrum, fuel with a very high combustion rate, reactor/
integrated cycle, etc.).
Experimental studies on fuel for the transmutation in rapid neutron reactors have been launched,
in particular in the PHENIX reactor, whose irradiation programme has focused on this research since
1998 and which has therefore been the object of inspection, renovation and maintenance, in view of a
power increase in 2001.
Teams at CEA and CNRS, in cooperation with industrial partners, have provided the technical
data for a request for an experimental demonstration model of a hybrid reactor for transmutation, in a
European and international framework.
This research is conducted to provide, in particular with reference to the report which must be
submitted to Parliament in 2006, the scientific and technical elements which may contribute to the
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choices and the implementation of management options for radioactive waste based on three guiding
principles: the minimization of waste, containment and reversibility.
The strategy targeting the control of plutonium and the minimisation of final waste aims to
achieve a significant long-term reduction in the mass and toxicity of waste to be stored and promotes
progressive implementation.
• The implementation of the separation-conditioning strategy may be programmed as an
extension of existing industrial capability: development, in a reprocessing facility, of
complementary processes which would make it possible to carry out the intensive separation of
long-lived radioactive elements, then their conditioning in specific matrixes designed to last a
very long time.
• The transmutation potential of light water reactors (current and EPR to satisfy the need for new
reactors), based on the development of new plutonium fuel, could be used to consume all the
plutonium, with the other long-lived elements benefiting from the separation-conditioning
strategy; the amount of final waste is thus divided by three compared with the open cycle.
• Future nuclear power production systems, studied in light of objectives of economic
competitiveness, optimum utilisation of natural resources in fuel, significant capacity for the
consumption of plutonium and long-lived radionuclides, would make it possible to reduce even
further the mass and toxicity of waste to be stored.
In addition, research on conditioning, interim storage and long-term storage is conducted, in
cooperation with the producers of waste and Andra, in order to develop:
• Processes for processing-conditioning for all types of nuclear waste for which an industrial
conditioning process does not exist.
• Containers which are certified for long-term interim storage, compatible with recovery or final
repository.
• Knowledge of the characteristics and performances of the long-term behaviour of all the
packages.
• Detailed preliminary projects for interim storage facilities, above or under ground, certified for
the long-term, and ready to be built if so decided, making it possible to keep spent fuel over
time, in particular MOX fuel, in order to preserve the energy content and be able to carry out
reprocessing-separation-transmutation operations at a later date.
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THE STATUS OF THE US ACCELERATOR TRANSMUTATION OF WASTE PROGRAMME
James C. Bresee 1 , James J. Laidler 2
1
United States Department of Energy
1000 Independence Avenue, SW, Washington, DC 20585, USA
2
Argonne National Laboratory (East)
9700 South Cass Avenue, Argonne, Illinois 60439, USA
Abstract
Since the last biannual meeting on partitioning and transmutation, the US accelerator transmutation
of waste (ATW) programme has changed significantly. Two years ago, the only effort was the
preparation of a research plan for developing ATW technology. Today, a significant research effort
in underway, and the US is seeking opportunities to collaborate with other national programmes.
Although the US fuel cycle is still based on a “once-through” process, with civilian spent fuel being
stored for direct disposal in a geologic repository, the technical feasibility for transmutation is being
investigated as a possible future option. Technetium-99, iodine-129 and neptunium-237 may be
released from a repository over geologic time periods and are the principle radioisotopes for
transmutation studies. Substantial reduction in total fissile materials and generation of useful energy
are also possible benefits of ATW. New test facilities are being considered which may be useful for
future multinational studies.
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1. Introduction
Since the 5th Information Exchange Meeting in Mol, Belgium in November 1998, the US
accelerator transmutation of waste (ATW) programme and indeed the US programme for chemical
processing of spent fuel has undergone a substantial change. At the time of the Mol meeting, the US
Congress had just authorised the preparation of a “roadmap” or programme plan for the development
of ATW technology. The background provided by foreign transmutation programmes and commercial
spent fuel reprocessing was an important part of the resulting roadmap which was published in late
1999 [1]. Based on that report, the US Congress provided $9 million in Fiscal Year 2000 (October 1,
1999-September 30, 2000) to establish an ATW research and development programme in the Office of
Nuclear Energy, Science and Technology. Department of Energy appropriations for the current fiscal
year (October 1, 2000 – September 30, 2001) include a substantial increase in ATW funding. The
purpose of this paper is to describe the content of that programme and the status of the R&D effort.
2. The US civilian spent fuel management programme
The nuclear fuel cycle in the US is currently a once-through process. Spent fuel from the
approximately 100 civilian nuclear power plants is being stored at the reactor sites with the
intention of transporting it in the future to a central geologic repository for “permanent” disposal.
However, any such repository, under current Nuclear Regulatory Commission requirements, must
be designed to allow spent fuel retrieval for at least 50 years. The actual design of a proposed
repository at Yucca Mountain, Nevada involves a retrieval capability for at least 100 years. Such
retrieval is mainly for safety purposes, in the unlikely event that during performance monitoring,
the repository or its contents develop significant problems.
Long term access to the contents of the repository also increases the probability of licensing,
since some of the uncertainties about repository safety will be reduced during monitoring. Finally,
retrievability offers the possibility that future generations may decide to recover the energy in the
spent fuel (principally plutonium-239) and reduce the long half-life radioactivity in the waste
through transmutation. Thus, the existence of a US programme for the development of a “oncethrough”
geologic repository while at the same time studying the possibility of nuclear waste
transmutation represents a consistent approach and provides technical flexibility. After all of the
changes during the twentieth century, an allowance for future technologic advances in nuclear waste
management is sound public policy.
The US Yucca Mountain project has reached a critical juncture. In December 2000, the
Department of Energy is releasing to the public a Site Recommendation Consideration Report
which provides interested parties with essentially all of the information which will provided in
June 2001 to the President in a Site Recommendation Report. He will use the final version of the
report as a basis for his decision on whether to recommend the Yucca Mountain site to Congress as
one he feels meets the strict environmental and safety requirements for a permanent repository. If
he does so decide, the Nuclear Waste Policy Act of 1992, as amended, provides an opportunity for
the affected state (Nevada, in the case of Yucca Mountain) to object to the President’s decision.
Such an objection stands unless overridden by a majority of both the US Senate and the US House
of Representatives. The technical content and persuasive arguments of the Site Recommendation
Report will strongly influence such a Congressional override.
The Yucca Mountain site is arid. Its annual rainfall is only about 12 centimetres of rain, 95%
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of which runs off or is evaporated rather than penetrating the mountain. That which does penetrate
moves in unsaturated flow through cemented and uncemented volcanic rock about 300 meters
before reaching the waste site and then another 300 meters before reaching saturation. The saturated
zone under the proposed repository site then flows slowly toward Death Valley as part of a closed
hydrology region. The water eventually evaporates in Death Valley rather than being connected
with any regional river system such as the Colorado River. Approximately thirty miles from the
Yucca Mountain site, there is some farming in Amargosa Valley which uses irrigation water from
the same aquifer flowing toward death Valley. It is the safety of individuals in Amargosa Valley
which will determine the acceptability and, if the site is found to be acceptable, the ability to license
a possible Yucca Mountain repository.
The most important issues in the decision process will be the containment at the repository site
of certain long-lived fission products and heavy elements within the high level nuclear waste. All
past performance assessment studies of the Yucca Mountain site as a possible repository, including
that reported in the Site Recommendation Consideration Report, have indicated that the dominant
mobile radionuclides during the first 10 000 years of the repository life are technetium-99
(213 000 year half-life) and iodine-129 (15.7 million year half-life), both of which may be
transported by underground water to points of possible human exposure. After approximately
100 000 years, the dominant isotope is neptunium- 237, with the possibility that plutonium isotopes
may also be important if carried by colloids or if plutonium were present in the more soluble VI
oxidation state. 99 Tc, 129 I and 237 Np (and other minor actinides) have been the focus of attention in US
studies of transmutation. No transmutation evaluations have indicated that a repository programme
will not be needed in the future; all such studies have shown that transmutation, if successful, could
reduce the hazards of such repositories.
3. The current US transmutation programme
Transmutation R&D in the US initially has been focused on accelerator-driven systems and has
involved a series of trade-off studies. In all cases, it has been assumed that uranium remaining in
civilian spent fuel elements would be recovered, probably by a modified Purex process called
UREX. Initial studies of the UREX process have shown that the uranium product will meet US
Class C requirements and could be disposed of as low level waste or be stored for possible future
use in a nuclear fuel cycle. The remaining process streams would be chemically separated into
transmutation fuel material, long-lived fission product transmutation targets, and a waste stream
that can be converted into durable waste forms capable of disposal in a high-level nuclear waste
repository.
Various combinations of proton accelerator designs, spallation neutron sources, and
transmutation target have been evaluated for technology readiness, and assumed irradiated targets
have been studied for the effectiveness of chemical processing to recycle untransmuted long-lived
isotopes. These evaluation have resulted in a base-line design which includes a linear proton
accelerator (or Linac), a lead-bismuth spallation target, and sodium- cooled metallic or ceramic
dispersion transmutation target/blanket non-fertile fuel elements. The initial formation of such nonfertile
transmutation targets and their subsequent reprocessing is the subject of a paper to be
presented later in this conference [2]. Other alternative designs have included cyclotrons, tungsten
spallation targets cooled by sodium, pressurised helium, or water, and nitride transmutation targets.
Another interesting transmutation system design currently being evaluated consists of a “dual
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strata” approach which would involve a thermal critical reactor within which plutonium and minor
actinides would fission and 99 Tc/ 129 I would be subjected to a thermal neutron flux. Technetium
would probably be in metallic form and iodine as an iodide of sodium, silver or other stable cations.
The thermal spectrum reactor would be an effective plutonium-239 burner along with other
actinides with high thermal fission cross-sections. Higher actinide isotopes would be produced by
non-fission neutron capture, and after post-irradiation chemical processing, they would be the
primary targets of an accelerator-driven transmutation system. Chemical processing of such targets
after irradiation would result in actinide recycle to the ATW unit and 99 Tc/ 129 I recycle to the thermal
reactor. High-level waste streams for repository disposal would be produced by the initial
processing of civilian spent fuel, the recycle processing of spent fuel from the thermal reactor, and
the ATW recycle process.
Since transmutation produces a net energy gain, it has been of interest to design systems
capable of producing electric power to off-set transmutation expenses. One concern has been the
current high “trip” rate of present generation accelerators, which may experience several unplanned
cut-offs each day. Quite apart from thermal shock safety considerations in the transmutation
system, such interrupted power would have much lower value than conventional base-load systems.
Early analysis indicates that more than ninety percent of the energy release in the “dual strata”
would occur in the thermal reactor, so it may be possible to design the ATW system as a lowtemperature
actinide burner with much less stringent requirements for accelerator power and
stability. Materials and corrosion problems in the ATW system would also be minimised. Studies of
the concept are continuing.
4. Advanced accelerator applications
The ATW programme during the current fiscal year involves approximately a doubling of the
Fiscal Year-2000 funding. This will allow an expansion of experimental programmes, and DOE’s
Office Of Nuclear Energy, Science and Technology (NE) is actively seeking opportunities for
collaborative research with foreign ADS programmes. Meanwhile, the programme is being
reorganised to combine the objectives of the DOE Defense Programme’s Accelerator Production of
Tritium programme with those of NE’s ATW efforts. The combined programme is known as
Advanced Accelerator Application, and it will be administered by NE. Congress has requested a
report by March 1, 2001 on how the new activity will be carried out. It will be a public document,
and it may be of interest to many attending this conference. It will be available on the World Wide
Web as well as in hard copy.
One objective of the new programme will be to help strengthen the nuclear science
infrastructure in America. To accomplish this, graduate thesis projects related to the programme
objectives will be sponsored at many universities. Another objective will be to strengthen nuclear
test facilities, and an accelerator driven test facility is under active consideration. The need to make
better use of limited test facilities throughout the world is also one of the reasons why DOE will be
seeking to increase international ADS/ATW collaboration. The coming years may see a
considerable expansion of the international quest for effective transmutation systems.
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REFERENCES
[1] A Roadmap for Developing Accelerator Transmutation of Waste (ATW) Technology – A
Report to Congress, DOE/RW-0519, October 1999.
[2] James J. Laidler and James C. Bresee, Pyrochemical Processing of Irradiated Transmutation Fuel,
6th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, Madrid, Spain, Dec. 11-13, 2000, EUR 19783 EN, OECD Nuclear Energy Agency,
Paris (France), 2001.
151

IAEA ACTIVITIES IN THE AREA OF EMERGING NUCLEAR ENERGY SYSTEMS
A. Stanculescu
International Atomic Energy Agency
Division of Nuclear Power, Nuclear Power Development Section,
PO Box 100, 1400 Vienna, Austria
Abstract
Nuclear energy is a proven technology that already makes a large contribution to energy supply
worldwide. At the end of 1999, there were 433 nuclear power plants operating in the world with a total
capacity of some 349 GW(e). The average annual growth rate of electricity production from nuclear
power is estimated to be about 0.6% per year for the period from now to 2015. One of the greatest
challenges facing nuclear energy is the highly radioactive waste, which is generated during power
production. While not involving the large quantities of gaseous products and toxic solid wastes
associated with fossil fuels, radioactive waste disposal is today’s dominant public acceptance issue. In
fact, small waste quantities permit a rigorous confinement strategy, and mined geological disposal is
the strategy followed by some countries. Nevertheless, political opposition arguing that this does not
yet constitute a safe disposal technology has largely stalled these efforts. One of the primary reasons
that are cited is the long life of many of the radioisotopes generated from fission. This concern has led
to increased R&D efforts to develop a technology aimed at reducing the amount of long-lived
radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent
years, in various countries and at an international level, more and more studies have been carried out
on advanced waste management strategies (i.e. actinide separation and elimination). In the frame of
the project on Nuclear Systems for Utilisation and Transmutation of Actinides and Long-lived Fission
Products the IAEA initiated a number of activities on utilisation of plutonium and transmutation of
waste, accelerator driven systems, thorium fuel option, innovative nuclear reactors and fuel cycles,
non-conventional nuclear energy systems, and fission/fusion hybrids.
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1. Introduction
In the second half of the 20 th century, nuclear power has evolved from an R&D environment to an
industry that supplies one sixth of the world’s electricity, one fifth of the USA’s and almost one third
of Western Europe’s. At the end of 1999, there were 433 nuclear power plants in operation and 39
under construction. Over nine thousand reactor-years of operating experience had been accumulated.
The turn of the century is a potential turning point also for nuclear power for several reasons:
• Fundamentally solid future prospects due to increasing world energy consumption, nuclear
power’s contribution to reducing greenhouse gas emissions, nuclear fuel resource
sustainability, and improvements in operation of current nuclear power plants.
• Advanced reactor designs that will improve economics and availability, and further enhance
safety.
• Continued attention to the key issues of nuclear safety, nuclear waste disposal and nonproliferation
of nuclear weapons.
From this perspective, the future prospects for nuclear power and IAEA’s role can be summarised
as follows:
Nuclear power operates in a growing market segment. Global energy demand is growing due to
industrialisation, economic development and increases in world population. It is projected to almost
triple by the middle of the 21 st century. In developing countries in the next thirty years, energy demand
is projected to increase two to three-fold, depending on the economic growth scenario. It is anticipated
that most of the world’s increase in nuclear capacity will be in Asia. The substantial increase in global
energy consumption in the coming decades will be driven principally by the economic growth and
industrialisation of developing countries, whose three quarters of the world’s inhabitants consume
only one quarter of the global energy. North America has a per capita consumption more than twice
that of Europe and almost eight times greater than that of South East Asia and the Far East. Strong
economic growth in many developing countries is already leading to sharp increases in per capita
energy consumption. Consumption will continue to rise, driven also by the projected two-fold
expansion in world population during the 21 st century that will occur overwhelmingly in the
developing regions. Globally, fossil fuels provide 87% of commercial primary energy. Nuclear power
and hydroelectric each contributes 6%. The non-hydroelectric renewables, solar, wind, geothermal and
biomass, constitute less than 1% of the energy supply. One third of commercial primary energy is
consumed in electricity generation.
Nuclear power reduces greenhouse gas emissions. Currently, nuclear power avoids annually
about 8% of global CO 2 emissions from energy production, or more than 600 million tonnes of carbon
(or 2 300 million tonnes of CO 2 ). As more and more people become convinced of the potential
consequences of global warming, and realise that the solutions are not going to be easy, the potential
for nuclear power to play an important role in the future energy mix in various regions must inevitably
become more widely recognised. The years since the Rio conference have solidified the international
consensus that increasing greenhouse gas emissions will have serious global consequences.
Nuclear power can compete with other energy sources. Despite the prevailing relatively low
fossil-fuel prices, the generating cost of nuclear electricity continues to be competitive with fossil fuel
for base-load electricity generation in many countries. Although the large capital investment required
for nuclear power plants is a disadvantage, especially in developing countries, the nuclear fuel cost is
relatively low. Moreover, the prices of fossil fuels are likely to increase over the long term (and have
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actually started to do so) because the resource is limited and also if pressures are applied – policy or
financial instruments, to discourage use. On the other hand, there is still scope in the nuclear industry
for rationalisation, standardisation, modular construction, shorter construction periods, higher burn-up
and simplification, resulting in better performance and lower electricity generation costs. Nuclear
power can thus be expected to be more competitive with fossil-fired plants in many areas of the world
in the long run.
In the early years of the next century, however, nuclear utilities will experience an operating
environment in which nuclear power plants will face increased competition, in an open energy market,
with other suppliers of electricity. In the face of this competitive pressure, nuclear power plants
worldwide are already showing a steady increase in the energy availability. The IAEA emphasises
improving the performance and reliability of nuclear power plants through the sharing of information
and experience world-wide, provides the PRIS database, as an authoritative source of information for
statistical analysis of nuclear power plant performance indicators, and conducts projects in nuclear
power plant performance assessment.
Moreover, nuclear power programs in Member States are making significant investments in
technology development and designs for the next century, focusing on substantial evolutionary
improvements of reactor systems to further enhance their economics, reliability and safety. To support
these programmes, the IAEA promotes technical information exchange and co-operation between
Member States, provides a source of balanced, objective information on developments in advanced
reactor technology, and publishes reports available to all Member States interested in the current status
of reactor development. These activities are conducted within the frames of technical working groups
for the major reactor lines: light water reactors, heavy water reactors, fast reactors, and gas cooled
reactors.
The global nuclear safety culture is extensively addressed by regulators in the Member States,
and by operators who have the prime responsibility for nuclear safety. The IAEA contributes to the
global nuclear safety culture through the introduction of binding conventions and recommended
standards, the provision of advisory services and the exchange of experience and information.
Nuclear waste disposal is often seen as the Achilles heel of the nuclear industry. Extensive
research and development in many countries has led to the general conclusion that final disposal is
technically feasible, but it still needs to be demonstrated convincingly to the public. That this has not
been done is largely attributable to public scepticism or opposition and lack of the necessary political
support. Presently, high level wastes are being stored above or below ground, awaiting policy
decisions on their long-term disposal.
The IAEA plays a major role in facilitating safe management of radioactive wastes. Support is
given to the collection, assessment and exchange of information on waste management strategies and
technologies for nuclear power plants, fuel cycle facilities, radioisotope applications, research
activities, and waste site restoration. The IAEA provides general technical guidance, assistance in
technology transfer and promotes international collaboration in optimising the development and
establishment of technical waste management infrastructures and programmes in Member States.
The IAEA plays a vital role in operating the international safeguards system that serves the
overall objective of non-proliferation of nuclear weapons. It also provides services designed to
strengthen the physical protection of nuclear materials and to combat the threat of illicit trafficking in
such materials. The safeguards system of the IAEA has been strengthened, via the so-called 93 + 2
programme, with requirements for more information and for allowing safeguards inspectors greater
access to installations, even to undeclared nuclear facilities. Through a co-operative activity between
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the IAEA Departments of Nuclear Energy and of Safeguards, guidelines for design measures to
facilitate the implementation of safeguards for future water cooled nuclear power plants have been
prepared.
To facilitate energy policy decision-making by Member States, the IAEA’s comparative
assessment programme is aimed at defining optimal strategies for the development of the energy
sector, consistent with the aims of sustainable development. This program focuses on developing and
disseminating databases and methodologies for comparative assessment of nuclear power and other
energy sources in terms of their economic, health and environment impacts; ensuring that the results of
IAEA-supported assessments are made available to relevant national and international forums (such as
the Intergovernmental Panel on Climate Change and the United Nations Framework Convention on
Climate Change); and on enhancing the capability of Member States to incorporate health and
environmental considerations in the decision making process for the energy sector.
In summary, as an international forum for exchange of scientific and technical information, the
IAEA plays a role in bringing together experts for a worldwide exchange of information about
national programmes, trends in safety and user requirements, the impact of safety objectives on plant
design, and the co-ordination of research programmes in advanced reactor technology. To support its
information exchange function, and to provide balanced and objective information to all
Member States on advances in reactor technology, the IAEA periodically prepares status reports on
advances in technology for each major reactor line.
3. IAEA activities in the area of Emerging Nuclear Energy Systems
The IAEA convened two meetings (The senior expert group and the external experts review
group) to review the Agency’s major programs. Both groups endorsed the Agency’s activities geared
towards the development and introduction of proliferation-resistant, long-lived radionuclides
incinerating/transmuting fuel cycles and innovative reactor designs in the small and medium sized
category.
During the 42 nd General Conference in September 1998, representatives from the United States
(Mr. Bill Richardson, Secretary of the US DOE), and the Russian Federation (Mr. E.O. Adamov of
Minatom, RF) promoted international R&D co-operation regarding research into the potential of
nuclear reactors and fuel cycles based on innovative technologies.
3.1 Innovative reactors and fuel cycles
Several experts’ groups at its meetings held in Vienna in 1998-2000 stated that “In response to
the new realities, it is important that nuclear technology be economically competitive, manage its
waste in a publicly acceptable manner, and contribute to sustainable utilisation of the earth’s
resources. As with other globally important technologies the nuclear community should mobilise the
intellectual and technical potential to produce innovative reactors and fuel cycles”.
In this context an IAEA initiative on innovative reactors and fuel cycles is timely and in
accordance with the interests of Member States, who are currently reviewing and discussing options
for the future direction of nuclear energy. Accordingly, the IAEA in co-operation with other relevant
international organisations, including OECD/NEA and IEA, is recommended “to provide an
international forum for discussion of innovative reactors and fuel cycles and to consider a possible
international R&D project”. In compliance with these findings, IAEA activities aiming at the
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establishment of an international R&D project on innovative reactors and fuel cycles are now in
progress.
3.2 Non-conventional Nuclear Energy Systems
As mentioned, the civilian nuclear energy enterprise is presently at a crossroads, the installed
nuclear energy market penetration having levelled off at only about 6% of the global total primary
energy consumption. This share will possibly even decrease in the near term since new nuclear power
plant construction is not likely to compensate for the decommissioning of existing ones. Underlying
this malaise are deeply held public apprehensions on issues such as reactor operational safety, longterm
spent fuel safety, and fissile material proliferation safety.
While researchers continue to investigate narrow aspects of nuclear reactor processes and
components, there exists little understanding of the integrated effect, which such investigations may
have on actual public acceptance of an entire operating nuclear energy system. It follows, therefore,
that there exists a need for the development of a complementary, client oriented “system performance
framework” to help focus nuclear energy research and development activities.
For the near term, small, factory-fabricated, modular plants which are delivered turnkey, are
supported by front end (fuel production) and back end (reprocessing and/or waste disposal) services by
the supplier, and which are offered for delivery under favourable financing might be favourable
received. They could accommodate the client’s situation of scarce financing, spare infrastructure, and
need for only small to medium sized increments in capacity. If the product mix produced by the plant
were diverse as well (producing potable water, electricity, and process heat), this would further
enhance their attractiveness to a developing country having natural resources but unskilled labour and
a need to manufacture value-added products based on their indigenous natural resources.
The process heat applications and enabling technologies to support them could lead in a natural
way to the longer term (second half of the 21 st century) nuclear power architecture which must be
based on fast spectrum systems which couple to modern energy converters such as gas turbines and
fuel cells. Technologies which support high core outlet temperature (850°C) and hydrogen production
via water cracking will be needed. The near-term developments to support process heat applications in
developing countries (such as coal reforming, heavy oil hydrogenization) will offer opportunities for
symbiosis with nuclear’s competitors for primary energy fuel supply – by helping to make fossil more
environmentally attractive – and thereby would provide incentive to large industrial groups (coal, oil)
to support innovations in nuclear technology.
3.3 Fission/fusion co-operation on technology aspects
Along with the ongoing efforts to establish fusion as an energy source, there is renewed interest in
fusion neutron source applications. In addition to fundamental neutron research, fusion R&D activities
are becoming of interest to nuclear fission power development. Indeed, for nuclear power
development to become sustainable as a long-term energy option, innovative fuel cycle and reactor
technologies will have to be developed to solve the problems of resource utilisation and long-lived
radioactive waste management. Both the fusion and fission communities are currently investigating
the potential of innovative reactor and fuel cycle strategies that include a fusion/fission hybrid. The
attention is mainly focused on substantiating the potential advantages of such hybrid systems:
utilisation and transmutation of actinides and long-lived fission products, intrinsic safety features,
enhanced proliferation resistance, and fuel breeding capabilities. An important aspect of the ongoing
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activities is comparison with the accelerator driven sub-critical system (spallation neutron source),
which is the other main option for producing excess neutrons. A consultancy held in Moscow in July
of this year initiated the preparation of a background report on the use of fusion/fission hybrids for
utilisation and transmutation of actinides and long-lived fission products, identifying the needs of the
R&D groups involved, and thus providing justifications and incentives for the Agency’s future
initiatives in this area.
3.4 The Three Agency study
“R&D on Innovative Nuclear Reactors – Status and Prospects” has been launched in 1999. The
three Agencies co-operating on the study are OECD/NEA, IAE and the IAEA. The objectives of the
study are: to trace possible paths to the future availability of nuclear technologies that are sufficiently
improved in safety, environmental performance and economics that they could conceivably be
commercially ordered in competitive energy markets within about 20 years, and to identify where
current research and development are feeding into such paths. Future nuclear technology R&D needs
will be identified and discussed in the context of decreasing public and private energy technology
R&D budgets. Particular attention will be paid to the potential role of international co-operation in
facilitating R&D and enhancing its cost effectiveness.
3.5 ADS and transmutation
In recent years, an old idea has re-surfaced and is gaining attention in nuclear technology: the
sub-critical, spallation neutron source driven nuclear system, or hybrid system. In this concept, a
powerful proton accelerator produces a spallation neutron source that drives a sub-critical core to a
relatively high fission power. Accelerator driven spallation targets and their proposed applications are
hybrid technologies, coupling the fields of accelerator design, particle beam physics, spallation target
design, and nuclear as well as reactor physics and the related engineering disciplines.
Particle accelerator and nuclear reactor technologies have developed for several decades along
parallel paths, with an important similarity being the capacity to produce large numbers of neutrons,
via fission (reactors) or spallation. Because of the improvements in particle accelerator technology and
economics, several large-scale applications of accelerator driven systems have been proposed. One of
the earliest proposed hybrid applications involved using spallation neutrons to supplement the fission
process in accelerator driven breeder reactors. More recently, a wider range of applications for hybrid
technologies was proposed to incinerate/transmute materials produced in nuclear reactors. These
applications rely on the larger availability of neutrons from hybrid systems and on their operation
flexibility as compared to critical nuclear reactors. Specifically, these applications include spallation
neutron sources, accelerator driven transmutation of waste, and accelerator driven power production.
Nuclear waste contains large quantities of plutonium, other fissionable actinides, and long-lived
fission products that pose challenges for long-term storage of waste and that are potential proliferation
concerns. If one assumes the same level of global nuclear power generation as exists today, then in the
year 2015 there will be more than 2 000 tons of plutonium in the spent fuel worldwide.
Different strategies for dealing with nuclear waste are being followed by various countries
because of their geologic situations and their views on nuclear energy, reprocessing and nonproliferation.
The current United States policy is to store unprocessed spent reactor fuel in a geologic
repository. Other countries are opting for treatment of nuclear waste, including partial utilisation of the
fissile material contained in the spent fuel, prior to geologic storage.
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The accelerator driven transmutation of waste (ATW) concept offers potential alternative paths
that would essentially eliminate plutonium, higher actinides, and environmentally hazardous fission
products from the waste stream destined for permanent storage. ATW does not threaten but instead
enhances the viability of permanent waste repositories. As such, ATW has increasingly become of
worldwide interest and could be an important component of strategies to deal with international
nuclear materials management requirements.
Accelerator driven systems: energy generation and transmutation of nuclear waste (Status
report). Participants of the special scientific programme on “Use of High Energy Accelerators for
Transmutation of Actinides and Power Production” held in Vienna, in 1994 in conjunction with the
38 th IAEA general conference recommended the IAEA to prepare a status report on accelerator driven
systems (ADS). The general purpose of the status report was to provide an overview of ongoing
development activities, different concepts being developed and their status, as well as typical
development trends in this area and to evaluate the potential of this system for power production, Pu
burning and transmutation of minor actinides and fission products. This document includes the
individual contributions by the experts from six countries and two international organisations, as well
as executive summaries in many different areas of the ADS technology. The document was published
and more than 500 copies were distributed by the IAEA in 1997 (IAEA-TECDOC-985).
Co-ordinated research project (CRP) on the Use of Thorium-based Fuel Cycles in Accelerator
Driven Systems (ADS) to Incinerate Plutonium and to Reduce Long-term Waste Toxicities is now in
progress. The participating countries and international organisations in the CRP are Belarus,
Czech Republic, France, Germany, Italy, the Netherlands, Russian Federation, Sweden, CERN and
Spain (as an observer). The purpose of the CRP is to assess the uncertainties of the calculated
neutronic parameters of a simple model of thorium or uranium fuelled ADS, in order to get a
consensus on the calculational methods and associated nuclear data. Participants identified a number
of issues, which should be considered to get a better understanding of the ADS and agreed that some
points, such as comparison of the different approaches and tools used by the different groups, should
be reviewed.
Three research co-ordination meetings (RCM) were held: in 1997 in Bologna, Italy, in 1998 in
Petten, Netherlands, and in 1999 in Vienna. Detailed papers on the results of these RCMs were
reported to several international meetings, e.g. the technical committee meeting on Feasibility and
Motivation for Hybrid Concepts for Nuclear Energy Generation and Transmutation (Madrid, Spain,
September 1997), and the Third International Conference on Accelerator Driven Transmutation
Technologies and Applications (Prague, Czech Republic, June 1999). As agreed at the consultancy in
Minsk, Belarus, in July of this year, the present stage of the CRP will be based on the YALINA set-up,
a well-defined and refined experiment considered by the experts as having the potential to resolve
some of the existing discrepancies in simulation of sub-critical systems and to give an indication on
the quality of widely used evaluated nuclear data libraries. Moreover, the present stage of the CRP
gives an opportunity to widen international participation in benchmarking and validation activities and
lays the ground for future activities in this area. The participants had committed themselves to perform
in advance “blind” test simulations of the first experiments. Already the preliminary results and
comparisons presented during the consultancy are of the great interest for the participating parties.
Technical committee meeting on Feasibility and Motivation for Hybrid Concepts for Nuclear
Energy Generation and Transmutation. The purpose of this TCM was to assess the advantages and
disadvantages of hybrid concepts for nuclear energy generation and transmutation of minor actinides
and their potential role relative to the current nuclear power programmes and potential future direction
to promote these concepts worldwide. The TCM was hosted by CIEMAT (Centro de Investigaciones
Energeticas Medicamentales y Tecnologicas) and held at its headquarters in Madrid, Spain, on
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17-19 September 1997. Several major programmes/concepts on ADS development were presented, i.e.
the CERN ADS concept, the OMEGA Program and Neutron Science Project for Developing
Accelerator Hybrid Systems at JAERI, the Los Alamos ATW Program, and the Hybrid Systems For
Nuclear Waste Transmutation Project in France.
The most salient observations resulting from the TCM were:
• Several accelerator systems and source concepts can be developed for ADS.
• Importance to have a very reliable neutron source coupled with the reactor system.
• The associated sub-critical reactor will likely be liquid lead (or lead-bismuth) cooled, with
efforts to use natural convection for coolant circulation.
• Effort to develop neutronic benchmarks and codes for ADS should be pursued at the
international level under the aegis of the Agency.
• Even if ADS is tentatively presented by some as a way to solve all nuclear waste issues, ADS
is not an alternative to geological disposal. However, ADS has the potential to drastically
reduce the waste toxicity, thanks to their capacity to burn minor actinides and fission
products. As a reprocessing stage will be required, non-proliferation concerns should be
addressed.
• Further development of ADS requires the building of a demonstration device with a thermal
power, in the 100-300 MW range. Efforts should be co-ordinated at international level on this
matter.
• This pre-industrial test should provide input on the feasibility of the industrial deployment of
ADS, including fuel cycle requirements, and a better understanding of the safety issues to be
addressed. The proceedings of this TCM were published by CIEMAT and distributed recently
by IAEA.
Database of experimental facilities and computer codes for ADS related R&D. The needs for
strengthening international co-operation in the field of the R&D for accelerator driven systems was
emphasised at several international forums, e.g.:
• Scientific program on “Use of High Energy Accelerators for Transmutation of Actinides and
Power Production”, Vienna, 21 September 1994 (in conjunction with the 38 th IAEA General
Conference).
• The Second International Conference on Accelerator Driven Transmutation Technologies and
Applications, Kalmar, Sweden, 3-7 June 1996.
• The 8 th International Conference on Emerging Nuclear Energy Systems (ICENES’96),
Obninsk, Russian Federation, 24-28 June 1996.
The consultancy on Hybrid Concepts for Nuclear Energy Generation and Transmutation held in
Vienna, in December 1996 noted that an increasing number of groups are entering this field of
research, many of these groups are not embedded in wider national activities, for these groups there is
a need for co-ordinating their efforts and jointly funding projects as also for getting access to
information from nationally or internationally co-ordinated activities.
Discussing organisational aspects of a possible IAEA involvement, the consultants came to the
conclusion that an effective co-ordination would necessitate the creation of an information document
on existing and planned experimental facilities which can be used for ADS related R&D. To
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substantiate this recommendation, several consultancies were organised in 1997-2000 to work out and
finalise the format of the document.
In June 1998, a draft of the database was distributed by the Agency to all contributors in the form
of working material. Presently an “electronic” version of the database is available on CD-ROM and
will be publicly accessible on the Internet very soon.
Advisory Group Meeting (AGM) on Review of National Accelerator Driven System (ADS)
Programs. This AGM was hosted by KAERI in Taejon, Republic of Korea, from 1-4 November 1999.
Its purpose was to review the current R&D programs in the Member States, and to assess the progress
in the development of hybrid concepts, as well as their potential role relative to both the current status
and the future direction of nuclear power worldwide. Further, the AGM participants provided advice
and guidance for the IAEA activities in the ADS area.
Technical Committee Meeting (TCM) on Core Physics and Engineering Aspects of Emerging
Nuclear Energy Systems for Energy Generation and Transmutation. This TCM was hosted by the
Argonne National Laboratory in Argonne, Illinois, USA, from 28 November-1 December 2000. Its
objective was to review the status of R&D activities in the area of hybrid systems for energy
generation and transmutation, to discuss in depth specific scientific and technical issues covering the
different R&D topics of these systems, and to recommend to the IAEA activities that would be
specifically targeted to the needs of the Member States performing R&D in this field.
Co-ordinated Research Project (CRP) on Safety, Environmental and Non-proliferation Aspects of
Partitioning and Transmutation (P&T) of Actinides and Fission Products. The overall objectives of
the CRP were to study the possibility of reduction of the long-term hazard arising from the disposal of
high level waste. More specifically, the CRP aimed to identify the critical nuclides to be considered in
a P&T strategy, to quantify their radiological importance in a global nuclear fuel cycle analysis and to
establish a priority list of radionuclides according the hazard definition. In the framework of the CRP
the radionuclides hazard was studied in order to identify the critical nuclides to be considered in a
P&T strategy and to quantify their radiological importance in a global nuclear fuel cycle analysis.
4. Thorium fuel option
Since the start of nuclear power development, thorium was considered to be the nuclear fuel to
follow uranium. The technology to utilise thorium in nuclear reactors was sought to be similar to that
of uranium, thorium resources to be larger than those of uranium, and the neutron yield of 233 U in
thermal and epithermal regions is higher than that for 239 Pu in the U/Pu fuel cycle. In more detail the
major reasons for the introduction of the thorium-based nuclear fuel cycles are: enlargement of fissile
resources by breeding 233 U; large thorium deposits in some countries, coupled with a lack of uranium
deposits in those countries; potential reduction in fuel cycle cost; reduction in 235 U enrichment
requirements; safer reactor operation because of lower core excess reactivity requirements; safe and
more reliable operation of thorium oxide fuel at high burn-up as compared to uranium oxide, due to
the higher irradiation and corrosion resistance of the former.
However, thorium has some disadvantages when compared with uranium, and this was also
recognised right from the beginning: thorium is more radioactive than uranium, making its handling in
the fabrication stage more challenging; the nuclear reactions induced by neutron absorption in thorium
and the decay schemes of the resulting nuclides are complicated, and the time for spent fuel storage in
water is longer due to the higher residual heat; potential difficulties in the back-end of the fuel cycle.
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In spite of the above-mentioned disadvantages, R&D efforts on the thorium/uranium fuel cycle
and thorium-fuelled reactor programmes started in the early 50s in several countries.
A series of three meetings was organised by IAEA in the period 1997-1999 on the thorium fuel options:
(1) Advisory group meeting on Thorium Fuel Cycle Perspectives, Vienna, Austria, 16-18 April 1997,
(2) Advisory group meeting on Thorium Fuel Utilisation: Options and Trends, Vienna, Austria, 28-30
September 1998, and (3) Technical committee meeting on Utilisation of Thorium Fuel: Options in
Emerging Nuclear Energy Systems, Vienna, Austria, 15-17 November 1999. The meetings were
organised jointly by the Nuclear Power Technology Development Section of the Division of Nuclear
Power and by the Nuclear Fuel Cycle & Materials Section of the Division of Nuclear Fuel Cycle and
Waste Technology. The purpose of the meetings was to assess the advantages, shortcomings, and options
of the thorium fuel under current conditions, with the aim of identifying new research areas and fields of
possible co-operation within the framework of the IAEA “Programme on Emerging Nuclear Energy
Systems”. Apart from current commercial reactors, the scope of the meetings covered all types of
evolutionary and innovative nuclear reactors, including molten salt reactors and hybrid systems.
Preparations for publication of the proceedings (IAEA-TECDOC) of the above mentioned
meetings is under way. Contributions to these meetings in the form of working material were
distributed to the participants.
Status report on Thorium-based Fuel Options. Within the framework of IAEA activities, the
Agency has maintained an interest in the thorium fuel cycle and its utilisation worldwide. Its periodic
reviews have assessed the current status of this fuel cycle, its applications worldwide, its economic
benefits, and its perceived advantages vis-à-vis other nuclear fuel cycles. Since 1994 the IAEA has
convened a number of technical meetings on the thorium fuel cycle and related issues. Between
1995-1997 individual contributions also were solicited from experts of France, Germany, India, Japan,
Russia and the United States of America, in many different areas of the thorium fuel cycle. They
included evaluations of the current status of the thorium fuel cycle worldwide, evaluation of new
incentives for using thorium as a result of the large stockpiles of plutonium produced in nuclear
reactors, new reactor concepts that can utilise thorium, strategies for thorium use, and an evaluation of
the toxicity of thorium fuel cycle waste as compared to other fuel cycles. The results of this updated
evaluation are summarised in the present publication “Thorium based fuel options for the generation
of electricity: developments in the 1990s”, IAEA-TECDOC-1155. Additionally, this document is a
contribution to the important task of preserving a large amount of past experience.
Co-ordinated research programme (CRP) on the Potential of Thorium-based Fuel Cycles to
Constrain Plutonium and to Reduce the Long-term Waste Toxicity. At the consultancy on “Important
Consideration on the Status of Thorium” held in Vienna from 29 November to 1 December 1994,
participants recommended the IAEA to organise a CRP on thorium-based fuel cycle issue. In 1995, the
Agency approved the topic for the CRP: “Potential of Thorium-based Fuel Cycles to Constrain
Plutonium and to Reduce Long-term Waste Toxicity”. The scope of this CRP was discussed and
agreed upon by the participants of the consultancy on “Thorium-based Fuel Cycles”, held from 6 to 9
June 1995 at the Agency’s Headquarters in Vienna. The participating countries in the CRP are: China,
Germany, India, Israel, Japan, Republic of Korea, the Netherlands, Russian Federation and the
United States of America.
This CRP examines the different fuel cycle options in which plutonium can be recycled with
thorium to incinerate plutonium. The potential of the thorium-matrix has been examined through
computer simulations. Each participant has chosen his own cycle, and the different cycles are
compared on the basis of certain predefined parameters (e.g. annual reduction of plutonium
stockpiles). The toxicity accumulation and the transmutation potential of thorium-based cycles for
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current, advanced and innovative nuclear power reactors are investigated. The research program has
been divided into three stages: (1) benchmark calculations, (2) optimisation of the incineration of
plutonium in various reactor types, and (3) assessment of the resulting impact on the waste toxicity.
The results of stage 1 were presented at ICENES 98. As agreed at the last RCM in Taejon
(Republic of Korea), in October 1999, the paper reporting the results of stage 2 was submitted to this
conference.
5. Efforts pursued jointly with other international organisations
Apart from the already mentioned “Three Agencies Study,” a collaborative effort with OECD/NEA
and IAE, two more salient recent examples of collaboration between the Agency and other international
organisations are worthwhile mentioning. The first one is the joint benchmark program set up by the
IAEA and the European Commission (EC) to assess the potential of reducing the sodium void
reactivity effect in innovative fast reactor designs and to perform comparative assessments of the
consequences of severe accident scenarios on such advanced fast reactor designs with near-zero
sodium void reactivity effect (this joint benchmark program has resulted in IAEA-TECDOC-731, and
IAEA-TECDOC-1139). The second example is the Agency’s participation in OECD/NEA’s Expert
Group on “Comparative Study of ADS and FR in Advanced Nuclear Fuel Cycles” whose objective is
to assess whether ADS deliver distinctive benefits in an advanced fuel cycle that includes P&T, as
compared to fast reactors.
6. Conclusions
The expansion of nuclear energy has been dampened in the past decade by a number of factors.
However, the prospects for the long term are positive. The global energy market is expanding and
nuclear energy has the potential to increase market share by diversification into non-electric use of
energy. Nuclear energy has two fundamental competitive advantages: long-term security of supply and
the potential for reduction of the emission of greenhouse gases. Significant investments are being
made in advanced nuclear reactor designs and technologies with the goal of having the technology
ready for the 21 st century. Continued safe operation of current reactors, implementation of nuclear
waste disposal technology, and an improved safeguards regime should reduce concerns about nuclear
power. Nuclear power can be expected to make an important contribution to global energy needs and
to the abatement of greenhouse gases in the next century and beyond.
For the last four decades, the IAEA has fulfilled the objective expressed in Article II of the
Agency’s Statute: “The Agency shall seek to accelerate and enlarge the contribution of atomic energy
to peace, health and prosperity throughout the world. It shall ensure, insofar as it is able, that
assistance provided by it or at its request or under its supervision or control is not used in such a way
as to further any military purpose”. Accordingly, the IAEA’s role is to provide all Member States with
an international source of balanced and objective information on advances in nuclear technology, and
to provide an international forum for information exchange and co-operative research. We see our role
in continuing to support joint efforts in emerging nuclear energy systems development. The IAEA will
continue to play a major role as the nuclear industry faces the challenges and opportunities of the
21 st century.
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LIST OF IAEA PUBLICATIONS
1. International Atomic Energy Agency, Evaluation of Actinide Partitioning and Transmutation,
Technical Report Series No. 214, Vienna, 1982.
2. International Atomic Energy Agency, Feasibility of Separation and Utilisation of Ruthenium,
Rhodium and Palladium from High Level Wastes, Technical Report Series No. 308, Vienna, 1989.
3. International Atomic Energy Agency, Feasibility of Separation and Utilisation of Caesium and
Strontium from High Level Liquid Wastes, Technical Report Series No. 356, Vienna, 1993.
4. International Atomic Energy Agency, Use of Fast Reactors for Actinide Transmutation, IAEA-
TECDOC-693, Vienna, 1993.
5. International Atomic Energy Agency, Safety and Environmental Aspects of Partitioning and
Transmutation of Actinides and Fission Products, IAEA-TECDOC-783, Vienna, 1995.
6. International Atomic Energy Agency, Advanced Fuels with Reduced Actinide Generation,
IAEA-TECDOC-916, Vienna, 1996.
7. International Atomic Energy Agency, Status Report on Actinide and Fission Product
Transmutation Studies, IAEA-TECDOC-948, Vienna, 1997.
8. International Atomic Energy Agency, Accelerator Driven Systems: Energy Generation and
Transmutation of Nuclear Waste, IAEA-TECDOC-985, Vienna, 1997.
9. International Atomic Energy Agency, Thorium Based Fuel Options for the Generation of
Electricity: Developments in the 1990s, IAEA-TECDOC-1155, Vienna, 2000.
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ACCELERATOR DRIVEN SUB-CRITICAL SYSTEMS FOR
WASTE TRANSMUTATION: CO-OPERATION AND CO-ORDINATION
IN EUROPE AND THE ROLE OF THE TECHNICAL WORKING GROUP
M. Salvatores, S. Monti
On Behalf of the European Technical Working Group on ADS
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1. Background
In 1998 the Research Ministers of France, Italy and Spain, recognising the potential of accelerator
driven systems (ADS) for the transmutation of long lived nuclear waste, decided to set up a Group of
Advisors (Ministers’ Advisors Group – MAG) to define a common R&D European platform on ADS.
In its meeting on May 1998, the MAG recommended an European demonstration programme over a
10-year time scale.
A Technical Working Group (TWG) under the chairmanship of Carlo Rubbia was also
established with the task of identifying the critical technical issues in which R&D, in such a
demonstration programme, is needed. In October 1998 the TWG issued an Interim Report which, in
particular, highlighted a) the need for a demonstration programme, b) the basic components and the
different options for the proposed demonstration facility, and c) the R&D directly relevant to the
realisation of such a facility.
This report was endorsed by the MAG at its meeting of March 1, 1999. In the same meeting, it
was proposed to extend participation beyond the three countries France, Spain, Italy; to consider the
role of ADS R&D within the 5th European Framework Programme (FWP); and to recognise an
eXperimental ADS (XADS) as an European goal.
As a consequence, a MAG “ad hoc” meeting open to all interested EU member states was held in
Rome on April 21, 1999. Representatives of eleven countries (Austria, Belgium, Denmark, Finland,
France, Germany, Italy, Portugal, UK, Spain and Sweden) participated in that meeting which
concluded:
It was agreed that neutron induced transmutation represents an attractive approach to radioactive
waste management, being complementary to geological disposal.
All participants appreciated the proposal to extend the participation in the initiative to other
European countries besides France, Italy and Spain, particularly considering that similar approaches
were being undertaken in the USA and Japan.
The interim report of the TWG issued in 1998 was accepted as a good basis for future work to be
carried out by an Enlarged Technical Working Group (ETWG), under the chairmanship of
Carlo Rubbia.
In September 1999, the ETWG – composed of representatives of Austria, Belgium, Finland,
France, Germany, Italy and Spain – issued a second technical report aimed at providing an overview
of the different ongoing activities on ADS in various European countries, along with an examination
of the proposals to be submitted to the 5th FWP. The report, presented to and endorsed by MAG on its
meeting of September 17, 1999, also identified a number of open points and gave recommendations
for the future development of the activities. In particular, the ETWG strongly recommended an
increased support – in particular by European Commission – and co-ordination of ADS-related
activities at multinational level.
At the beginning of 2000 the ETWG (further enlarged to representatives of the JRC, Portugal and
Sweden), recognising that the R&D programme on ADS has reached a turning point with regard to
programme co-ordination and resource deployment in Europe and taking also into account the
substantial recent progress on the subject in the United States and in Japan, issued a so-called “fourpage
document” on a strategy for the implementation of an ADS programme in Europe. In particular,
the document called for the urgent definition of a consensual European “Roadmap” towards
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demonstration of feasibility of a European waste transmutation facility and recognised its potentiallyrelevant
implications on the 6th European Framework Programme.
The “four-page document” was endorsed by the MAG on its meeting of February 25, 2000 and,
consequently, the ETWG started an intensive work aimed at defining the above mentioned European
Roadmap. In particular, in order to specifically address some relevant key issues such as accelerator,
fuel and fuel processing development, two dedicated sub-groups have been created inside the ETWG
and co-ordination with the European ADS system design group has been established.
The roadmap document is expected to be issued at the very beginning of 2001.
In the report, the ETWG will identify the steps necessary to start the construction of an XADS
towards the end of the decade. The construction and operation of an XADS at that point in time is
considered as an essential prerequisite to assess the safe and efficient behaviour of ADS for a possible
large scale deployment of ADS for transmutation purposes in the first half of the century.
The first goal of the roadmap is to propose a technological route to reduce the risks associated
with nuclear waste, based on the transmutation of nuclear waste in ADS and to assess the impact of
this approach in the reduction of the radiotoxicity of nuclear waste. The report will review historical
developments and will identify and review the status of current activities and facilities related to ADS
research in the EU and worldwide. A decision to go ahead with the project will require a detailed
planning of the technical aspects, a substantially increased budget, together with close synchronisation
with the 6th and 7th Framework Programmes. The second and main goal of the roadmap is, therefore,
to prepare a detailed technical programme, with cost estimates, which will lead to the demonstration of
an experimental ADS (XADS) in 10 years, within the 6th and 7th Framework Programmes. The
programme as described in the roadmap will lead to the development of innovative fuels and
reprocessing technology, a rationalisation of human resources and experimental facilities, a training
ground for young researchers, spin-offs in the fields of accelerators, spallation sources, liquid metal
technology, radioisotope production and actinide physics and chemistry. Hence, a final goal of the
roadmap is to identify possible synergies and rationalisations that this programme could have within
the nuclear community, indicate potential spin-offs, show how competence can be maintained in a
currently stagnating field.
The roadmap will be a result of a mandate given to the ETWG on ADS by the MAG. In the first
instance, therefore, the report will be directed at the MAG. The document will be of interest, however,
to policy makers throughout Europe, in particular to research ministries in the Member States of the
European Union, to members of the European Parliament, and to the relevant Directorates General of
the European Union. In addition, the report will be of interest to parties involved with ADS research
and development within the EU and worldwide.
2. Motivations for ADS
In contrast to standard critical nuclear reactors in which there are enough neutrons to sustain a
chain reaction, sub-critical systems used in ADS need an external source of neutrons to sustain the
chain reaction. These “extra” neutrons are provided by the accelerator. More exactly the accelerator
produces high-energy protons, which then interact with a spallation source to produce neutrons.
But why go to all this trouble to build an ADS when critical reactors already work? The answer to
this lies in the fact that one has more control and flexibility in the design of the sub-critical reactor.
This is required when the reactor is being used to transmute large amounts of nuclear waste in the
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form of minor actinides (MAs). Today it appears that ADS has great potential for waste transmutation
and that such systems may go a long way in reducing the amounts of waste and thereby reducing the
burden to underground repositories. A schematic description of how an ADS can be used in
conjunction with conventional reactors in a “Double Strata” approach is shown in Figure 1.
Figure 1. Schematic description of the transmutation of nuclear waste by ADS
within a Double Strata Fuel Cycle (LLFP: long-lived fission products, SLFP:
short-lived fission products, LLW: low level waste, HLW: high level waste)
The first stratum is based on a conventional fuel cycle and consists of standard light water
reactors (LWR) and fast neutron reactors (FNR), fuel fabrication and reprocessing plants. The
recovered plutonium is recycled as mixed oxide fuel in the thermal and fast reactors. The remaining
plutonium, MAs and long-lived fission products are partitioned from the waste and enter the second
stratum where they are transmuted in a dedicated ADS. In the second stratum, devoted primarily to
waste reduction, the Pu, MAs, and long-lived fission products are fabricated into fuels and targets for
transmutation in dedicated ADS. The use of dry reprocessing in this stratum allows for multiple
reprocessing of the fuel. A key advantage of this is that higher levels of radiation can be tolerated in
the molten salts, used in this process, and therefore allows reprocessing of spent fuel which has been
cooled for periods as short as one month.
Accelerator driven systems therefore open the possibility of “burning” waste material from LWRs
in dedicated actinide burners. These actinide burners can burn large quantities of minor actinides per
unit (in contrast to critical reactors) safely, and generate heat and electricity in doing so. In addition,
schemes have been proposed, in which the long-lived fission products are also destroyed. An
advantage of ADS is that, since there is no criticality condition to fulfil, almost any fuel composition
can be used in the system.
3. Overall roadmap towards ADT industrial implementation through an ADS demonstration
From now on, taking for granted the potential role of Accelerator Driven Technology (ADT) for
radio-toxicity reduction of the ultimate nuclear waste, the main issue is to define a path towards
industrial implementation of this very innovative system.
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An extended “skeleton” for ADS technology development is given in Figure 2. It consists of three
main phases:
• Phase 1: Development/realisation/operation of an XADS/XADT (eXperimental Accelerator
Driven System/Transmuter).
• Phase 2: Development/realisation/operation of a PROTO-ADT.
• Phase 3: Industrial application.
In particular, the preparation of the construction of the XADS must be performed in parallel to the
development and qualifications of the main components and the basic R&D programme. It starts with
a system analysis of different concepts to be performed in 2001 to 2003. Further milestones are:
• A decision on the basic features of the XADS in 2005.
• Start of detailed design work in 2005.
• Final decision and start of construction in 2009.
• Operation of XADS in 2013-2014.
In parallel to the construction of the XADS, which can be considered as the “spallation-fission
facility” oriented demonstration, pilot plants have to be built and operated in order to:
• Feed the XADS/XADT – prototype with fuel and targets.
• Get an “As Soon As Possible (ASAP)” operation feedback for further development of
industrial scale facilities.
The ADS demonstration phase is characterised, in an ASAP perspective, by the use of available
fuel technology, at least in the first sub-phase of the XADS. The two sub-phases currently considered
are:
• Sub-Phase 1: which uses available fuel technology and is devoted to the demonstration of the
ADS concept and possibly to irradiation purposes (XADS).
• Sub-Phase 2: which is devoted to the transmutation demonstration with a large number of
MA-based fuel assemblies. During this phase the XADS will be used more and more as
demonstration of a transmuter. In about 2017-2018 a decision has to be made, whether the
facility can be modified to a full transmuter (XADT) or whether a new facility would be
needed, which possibly should be available in 2025. From 2025 on in any case a full
demonstration of accelerator driven transmutation should occur. This requires successful
development of the fuel and the fuel cycle facilities till about 2020!
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Figure 2. Time schedule and milestones for the development
of an experimental accelerator driven system (ADS) and
accelerator driven transmutation (ADT) technology in Europe
4. TWG sub-group on fuels and fuel processing
As above mentioned, a specific subgroup on “Fuels and Fuel Processing”, under the chairmanship
of Rudy Konings (ITU-Karlsruhe), has been set up within the TWG. Two status reports have been
issued by this sub-group:
• Part I: The Fuel of the ASAP DEMO.
• Part II: Advanced Fuel Cycles.
As far as fuel for XADS, Mixed OXide (MOX) is considered the only “ASAP” fuel option in
Europe; indeed, for this fuel:
• Extensive knowledge exists.
• Existing fuel elements or fuel pins (SNR-300 or SPx) are available.
• The required infrastructure for fabrication of new fuel elements is existing.
• Core design has to be adapted (partially) to elements.
• Austenitic cladding of the SNR and SPx elements are compatible with He cooling and can
also be used for Pb/Bi coolant, but the oxygen content of the liquid metal should be low.
As far as advanced fuels for transmutation, following the mandate of the TWG, the following
boundary conditions were assumed:
• Solid fuel forms.
• High TRU content (up to 50% with MA/(Pu+MA) ratio between 1/5 and 5).
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• Fertile support (uranium) is not considered.
• Gas or liquid metal coolant.
• Fast neutron system.
Pros and cons of metal, oxide, nitride, carbide fuel form have been examined. Composite fuels
(including coated particle fuels), such as CERCER and CERMET, have been also analysed, as well as
fission product targets.
The subgroup also addressed the key issue of fuel reprocessing, concluding that some drawbacks
are expected for advanced fuels treated by hydro-chemical reprocessing. On the contrary, pyrochemical
reprocessing is considered a promising technology, given the fact that good fuel solubility
can be achieved even for refractory materials. Furthermore, compact facilities and small cooling times
are required. Drawbacks of this technology are: aggressive process media, secondary wastes and
sophisticated technology which is under development.
The main conclusions of the “fuels and fuel processing subgroup” were the following:
• The European fuel research should be focussed on innovative oxide-based fuels, either as a
solid solution or as a composite (CERMET or CERCER), with emphasis on the helium issue.
• Thorium-based fuel should be considered as back-up solution, e.g. (Th, TRU)O 2 .
• The pyro-chemical reprocessing of these fuel types should be studied.
• In the short term (4-5 years) an out-of-pile testing programme should be performed to
establish the properties of MA-based oxide fuels (solid solution, composite).
• In the longer term (8-10 years) a dedicated irradiation programme for these fuels should be
performed.
• In parallel a collaboration with Japan and the Russian Federation on nitride fuel must be
pursued (U-based nitrides, with limited MA content).
5. TWG Subgroup on accelerators for ADS
A second subgroup devoted to HPPA development for ADS, under the chairmanship of
Alex Mueller (CNRS-IN2P3), has been set up within the ETWG. This subgroup has issued a status
report on “Accelerators for ADS”. The main outcomes of this status report can be summarised as
follows.
The orders of magnitude of the characteristics of the accelerator for XADS were defined
assuming a 100 MW thermal power for the sub-critical core.
A proton beam of 1 GeV, 10 mA may be obtained both by linac- or cyclotron- type accelerators,
although at the very limits of potential performance gains for the latter case.
Concerning the beam structure, the preliminary analysis has not identified principle technical
difficulties comparing pulsed or continuous beam operation. However, it seems that the latter option is
overall preferred because of a closer similarity to the classical critical reactor.
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The accelerator of the XADS has not to be defined only by the needs of the XADS itself, but it
should also anticipate the requirements and the technical issues of a future industrial ADS burner. This
concerns in particular the reliability, the economical, and the nuclear safety aspects.
A superconducting linac is adapted to a high power industrial ADS burner. Its different
components are currently being investigated in several European laboratories. The studies show that a
100 mA proton beam can now be handled.
For comparative cost estimates an energy of 600 MeV has been used: it is the minimum energy
(and therefore cost) for a complete ADS demonstration; for this energy band convincing accelerator
studies exist for both types of accelerators. The extrapolation to 1 GeV, the energy of best neutron
economy, is straightforward for the linac, however delicate for the cyclotron.
The investment cost of a 20 mA, 600 MeV linac, adapted to the XADS needs, including all
industrial aspects, is preliminary evaluated to around 200 M . Its development (R&D and
construction) requires around eight years. An extension to 1 GeV would require an additional
investment cost of about 100 M .
A single cyclotron (600 MeV, 5 mA) is adapted to low power applications which is enough for
the XADS if coupling of accelerator and reactor are the basic objective. A rough cost evaluation, on a
similar base than for the linac, is around 70 M . Approximately six years are requested for
development and construction. If the industrial burner demonstration is a must, the coupling of at
least 3 such machines must be demonstrated, of which the global cost is estimated to 240 M .
6. Harmonisation of industry and R&D proposals for a preliminary design study of an XADS
Nine countries are participating to a 3-years proposal to be submitted to the second call of P&T
within the 5th European Framework Programme.
The purpose is to develop conceptual plant configurations to enable definition of the supporting
R and D, to perform objective comparisons and to propose reference solutions to be engineered in
detail.
• Preliminary design concentrate mainly on three concepts:
− A small XADS (20-40 MWt) cooled by Pb/Bi (MYRRHA-Belgium).
− A large Pb/Bi cooled XADS (~80 MWt), developed in Italy.
− A similar size gas-cooled ADS, developed in France.
Some work will be devoted to a pebble-bed concept developed in Spain.
The work will be carried out in three different phases:
• Phase 1: definition of the main technical specifications of XADS (6 months).
• Phase 2: preliminary engineering design studies (including safety approach, accelerator, core
and target, system integration), to consolidate feasibility and cost estimation and to provide
essential elements of comparison (2 years).
• Phase 3: comparisons and recommendations to implement the road-mapping towards XADS
(6 months).
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7. Conclusions
The TWG under the chairmanship of Carlo Rubbia, enlarged from the three initial partners
(France, Italy, Spain) to ten partners (Austria, Belgium, Finland, France, Germany, Italy, JRC,
Portugal, Spain, Sweden), has worked to identify critical technical issues, R&D needs and a strategy in
view of an ADS demonstration programme.
A roadmap document is being prepared, and state of the art reports have been issued in
Accelerator and Fuel and Fuel Processing Technologies.
The TWG allows also a monitoring and harmonisation of other European projects, like:
• The MEGAPIE project for the construction and operation of a 1 MWt Pb/Bi target in the
SINQ installation in PSI-Switzerland.
• The proposal for a preliminary design study of an XADS.
• Which will be both submitted to the second call of P&T within the 5th European Framework
Programme.
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PARTITIONING AND TRANSMUTATION
IN THE EURATOM FIFTH FRAMEWORK PROGRAMME
Michel Hugon, Ved P. Bhatnagar
European Commission
rue de la Loi, 200, 1049 Brussels, Belgium
E-mail: Michel.Hugon@cec.eu.int
Abstract
Partitioning and Transmutation (P&T) of long-lived radionuclides in nuclear waste is one of the
research areas of the EURATOM Fifth Framework Programme (FP5) (1998-2002). The objective of
the research work carried out under FP5 is to provide a basis for evaluating the practicability, on an
industrial scale, of P&T for reducing the amount of long lived radionuclides to be disposed of. The
content and the implementation of the EURATOM FP5 are briefly presented. The research projects on
P&T selected for funding after the first call for proposals are then briefly described. They address the
chemical separation of long-lived radionuclides and the acquisition of technological and basic data,
necessary for the development of an accelerator driven system. Other projects are expected in response
to the next call with a deadline in January 2001. International co-operation in P&T should be fostered.
Collaboration is being implemented in this field between scientists of the European Union (EU) and
the Commonwealth of Independent States (CIS). Finally, a brief outline of the discussions for the
preparation of the Sixth Framework Programme (2002-2006) is given.
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1. Introduction
The priorities for the European Union’s research and development activities for the period
1998-2002 are set out in the Fifth Framework Programme (FP5). These priorities have been identified
on the basis of a set of common criteria reflecting the major concerns of increasing industrial
competitiveness and the quality of life for European citizens. FP5 has been conceived to help solve
problems and to respond to major socio-economic challenges facing Europe. To maximise its impact,
it focuses on a limited number of research areas combining technological, industrial, economic, social
and cultural aspects.
The Fifth Framework Programme has two distinct parts: the European Community Framework
Programme covering research, technological development and demonstration activities; and the
EURATOM Framework Programme covering research and training activities (RT) in the nuclear field.
The content and the implementation of the latter are briefly presented in this paper.
Partitioning and Transmutation (P&T) of long-lived radionuclides in nuclear waste is one of the
research areas of the EURATOM FP5. This paper briefly recalls the goals of P&T, its position in the
framework of nuclear waste management and disposal and its renewed interest worldwide. The
research projects on P&T so far selected for funding in FP5 are then briefly described. Co-operation in
this field with some countries of the Commonwealth of Independent States (CIS) through the
International Science and Technology Centre in Moscow is also outlined.
Finally, some indications are given concerning the European Union’s research beyond the Fifth
Framework Programme and the “European Research Area”.
2. The EURATOM Fifth Framework Programme (FP5) (1998-2002)
The Fifth Framework Programme of the European Atomic Energy Community (EURATOM) has two
specific programmes on nuclear energy, one for indirect research and training actions managed by the
Research Directorate General (DG RTD) and the other for direct actions under the responsibility of the
Joint Research Centre of the European Commission (EC). The strategic goal of the first one,
“Research and training programme in the field of nuclear energy”, is to help exploit the full potential
of nuclear energy in a sustainable manner, by making current technologies even safer and more
economical and by exploring promising new concepts [1]. This programme includes a key action on
controlled thermonuclear fusion, a key action on nuclear fission, research and technological
development (RTD) activities of a generic nature on radiological sciences, support for research
infrastructure, training and accompanying measures. The key action on nuclear fission and the RTD
activities of a generic nature are being implemented through indirect actions, i.e. research cosponsored
(up to 50% of total costs) and co-ordinated by DG RTD, but carried out by external public
and private organisations as multi-partner projects. The total budget available for these indirect actions
during FP5 is 191 millions .
The key action on nuclear fission comprises four areas: (i) operational safety of existing
installations; (ii) safety of the fuel cycle; (iii) safety and efficiency of future systems and (iv) radiation
protection. The operational safety of existing installations deals with plant life extension and
management, severe accident management and evolutionary concepts. In the safety of the fuel cycle,
waste and spent fuel management and disposal, and partitioning and transmutation are the two larger
activities, as compared to the decommissioning of nuclear installations. The objective of safety and
efficiency of future systems is to investigate and assess new or revisited concepts for nuclear energy,
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that would be more economical, safer and more sustainable in terms of waste management, utilisation
of fissile material and safeguards. Radiation protection has four sub-areas: (i) risk assessment and
management, (ii) monitoring and assessment of occupational exposure, (iii) off-site emergency
management and (iv) restoration and long-term management of contaminated environments.
The implementation of the key action on nuclear fission is made through targeted calls for
proposals with fixed deadlines. The generic research on radiological sciences is the subject of a
continuously open call, but proposals are evaluated in batches. Following the calls for proposals made
in 1999, about 140 proposals covering all areas of the key action and of the generic research have been
accepted for a total funding of around 100 million . Most of the projects have started now. A new call
for proposals has been made on 16 October 2000 with a deadline on 22 January 2001 to select
proposals for another 50 million . The 2000 version of the Work Programme is available on the
CORDIS website (www.cordis.lu/fp5-euratom). A final call will be made in October 2001.
3. Partitioning and Transmutation (P&T)
Spent fuel and high level waste contain a large number of radionuclides from short-lived to longlived
ones, thus requiring very long time periods to be considered for their geological disposal. The
long-lived radionuclides are mainly the actinides and some fission products. Partitioning and
Transmutation aims at reducing the inventories of long-lived radionuclides in radioactive waste by
transmuting them into radionuclides with a shorter lifetime [2].
Partitioning is the set of chemical and/or metallurgical processes necessary to separate from the
high-level waste the long-lived radionuclides to be transmuted. This separation must be very efficient
to obtain a high decontamination of the nuclear waste. It should also be very selective to achieve an
efficient transmutation of the long-lived radiotoxic elements.
Long-lived radionuclides could be transmuted into stable or short-lived nuclides in dedicated
burners. These burners could be critical nuclear reactors and sub-critical reactors coupled to
accelerators, the so-called accelerator-driven systems (ADS).
If successfully achieved, P&T will produce waste with a shorter lifetime. However, as the
efficiency of P&T is not 100%, some long-lived radionuclides will remain in the waste, which will
have to be disposed of in a deep geological repository. P&T is still at the research and development
(R&D) stage. Nevertheless, it is generally accepted that the techniques used to implement P&T could
alleviate the problems linked to waste disposal.
There has been a renewal of interest in P&T worldwide at the end of the eighties (OMEGA
programme in Japan, SPIN programme in France). Meanwhile, sufficient progress has been made in
accelerator technology to consider as feasible the use of ADS for waste incineration. Proposals to
develop ADS have been made during the nineties by the Los Alamos National Laboratory in the USA
with the ATW (Accelerator driven Transmutation of Waste) programme, by CERN in Europe with the
Energy Amplifier (EA) [3] and by JAERI in Japan. In addition, there is a number of research activities
on ADS going on in several EU countries (Belgium, France, Germany, Italy, Spain, Sweden),
Czech Republic, Switzerland, Republic of Korea and Russian Federation.
The interest for P&T in the EU is reflected in the increase of funding in this area over the
EURATOM Framework Programmes, 4.8, 5.8 and about 26 million for the Third, Fourth and Fifth
Framework Programmes respectively.
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In the EURATOM Fourth Framework Programme (1994-1998), there were research activities on
P&T covering three aspects: (i) strategy studies, (ii) partitioning techniques, (iii) transmutation
techniques. The progress achieved in the field of partitioning of minor actinides by aqueous processes
suggested that, with some additional effort, a one-cycle process for the direct extraction of minor
actinides from liquid high-level waste could be demonstrated at the pilot plant scale. The conclusions
of the P&T strategy studies concerning critical and sub-critical reactors and the results obtained in the
transmutation experiments clearly indicated that the feasibility of ADS should be more thoroughly
investigated for transmutation of nuclear waste [4].
4. The research activities on P&T in the EURATOM Fifth Framework Programme
The objective of the research work carried out under FP5 is to provide a basis for evaluating the
practicability, on an industrial scale, of partitioning and transmutation for reducing the amount of long
lived radionuclides to be disposed of.
After the first call for proposals in 1999, 19 proposals were received in the area of partitioning
and transmutation, requesting about 3.8 times more than the available budget. By taking due account
of the advice of the evaluators, the Commission services selected 9 proposals for funding at a level
lower than requested due to budget limitations. All the projects have already started between August
and November 2000.
The selected projects address different scientific and technical aspects of P&T and have therefore
been grouped in three clusters. The experimental investigation of efficient hydro-metallurgical and
pyrochemical processes for the chemical separation of long-lived radionuclides from liquid high-level
waste is carried out in the cluster on partitioning. The work on transmutation is mainly related to the
acquisition of data, both technological and basic, necessary for the development of ADS. The cluster
on transmutation – technological support – deals with the investigation of radiation damage induced
by spallation reactions in materials, of the corrosion of structural materials by lead alloys and of fuels
and targets for actinide incineration. In the cluster on transmutation – basic studies, basic nuclear data
for transmutation and ADS engineering design are collected and sub-critical neutronics is investigated.
The cluster on partitioning includes three projects, the main characteristics of which are given in
Table 1. The first one, PYROREP, aims at assessing flow sheets for pyrometallurgical processing of
spent fuels and targets. Two methods, salt/metal extraction and electrorefining, will investigate the
possibility of separating actinides from lanthanides. Materials compatible with corrosive media at high
temperature will be selected and tested. It is worth noting that one of the partners of this project is
CRIEPI, the research organisation of the Japanese utilities.
The two other projects are dealing with the development of solvent extraction processes of minor
actinides (americium and curium) from the acidic high level liquid waste (HLLW) issuing the
reprocessing of spent nuclear fuel. In PARTNEW, the minor actinides are extracted in two steps.
They are first co-extracted with the lanthanides from HLLW (DIAMEX processes), then separated
from the lanthanides (SANEX processes). Basic studies will be performed for both steps, in particular
synthesis of new ligands and experimental investigation and modelling of their extraction properties.
The radiolytic and hydrolytic degradation of the solvents will be also studied and the processes will be
tested with genuine HLLW.
The CALIXPART project is dealing with the synthesis of more innovative extractants.
Functionalized organic compounds, such as calixarenes, will be synthesised with the aim of achieving
the direct extraction of minor actinides from HLLW. The extracting capabilities of the new
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compounds will be studied together with their stability under irradiation. The structures of the
extracted species will be investigated by nuclear magnetic resonance (NMR) spectroscopy and X-ray
diffraction to provide an input to the molecular modelling studies carried out to explain the
complexation data.
Table 1. Cluster on partitioning
Acronym Subject of research Co-ordinator
(country)
PYROREP
PARTNEW
CALIXPART
Pyrometallurgical
processing research
Solvent extraction processes
for minor actinides (MA)
Selective extraction of MA by
organised matrices
Number of
partners
Duration
(months)
EC funding
(Million )
CEA (F) 7 36 1.5
CEA (F) 10 36 2.2
CEA (F) 9 36 1.3
The cluster on transmutation-technological support has four projects (see Table 2). The SPIRE
project addresses the irradiation effects on an ADS spallation target. The effects of spallation products
on the mechanical properties and microstructure of selected structural steels (e.g. martensitic steels)
will be investigated by ion beam irradiation and neutron irradiation in reactors (HFR in Petten, BR2 in
Mol and BOR 60 in Dimitrovgrad). Finally, data representative of mixed proton/neutron irradiation
will be obtained from the analysis of the SINQ spallation target at the Paul Scherrer Institute in
Villigen.
The objective of TECLA is to assess the use of lead alloys both as a spallation target and as a
coolant for an ADS. Three main topics are addressed: corrosion of structural materials by lead alloys,
protection of structural materials and physico-chemistry and technology of liquid lead alloys. A
preliminary assessment of the combined effects of proton/neutron irradiation and liquid metal
corrosion will be done. Thermal-hydraulic experiments will be carried out together with numerical
computational tool development.
Fuel issues for ADS are addressed in the CONFIRM project. Computer simulation of uranium
free nitride fuel irradiation up to about 20% burn-up will be made to optimise pin and pellet designs.
Other computations will be performed especially concerning the safety evaluation of nitride fuel.
Plutonium zirconium nitride [(Pu, Zr)N] and americium zirconium nitride pellets will be fabricated
and their thermal conductivity and stability at high temperature will be measured. (Pu, Zr)N pins of
optimised design will be fabricated and irradiated in the Studsvik reactor at high linear power
(≈70 kW/m) with a target burn-up of about 10%.
The objective of the project THORIUM CYCLE is to investigate the irradiation behaviour of
thorium/plutonium (Th/Pu) fuel at high burn-up and to perform full core calculations for thoriumbased
fuel with a view to supplying key data related to plutonium and minor actinide burning. Two
irradiation experiments will be carried out: (i) four targets of oxide fuel (Th/Pu, uranium/plutonium,
uranium and thorium) will be fabricated, irradiated in HFR in Petten and characterised after
irradiation; (ii) one Th/Pu oxide target will be also irradiated in KWO Obrigheim. Though this project
was accepted for funding in the area of “safety and efficiency of future systems”, it has been grouped
with the three previous projects in the cluster on transmutation-technological support for convenience,
because it is related to fuel issues.
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Table 2. Cluster on transmutation-technological support
Acronym Subject of research Co-ordinator
(country)
SPIRE
TECLA
CONFIRM
THORIUM
CYCLE
Effects of neutron and
proton irradiation in steels
Materials and thermohydraulics
for lead alloys
Uranium free nitride fuel
irradiation and modelling
Development of thorium
cycle for PWR and ADS
Number of
partners
Duration
(months)
EC funding
(Million )
CEA (F) 10 48 2.3
ENEA (I) 16 36 2.5
KTH (S) 7 48 1.0
NRG (NL) 7 48 1.2
Finally, three projects are grouped in the cluster on transmutation-basic studies (see Table 3). The
MUSE project aims to provide validated analytical tools for sub-critical neutronics including
recommended methods, data and a reference calculation tool for ADS study. The experiments will be
carried out by coupling a pulsed neutron generator to the MASURCA facility loaded with different
fast neutron multiplying sub-critical configurations. The configurations will have MOX fuel with
various coolants (sodium, lead and gas). Cross-comparison of codes and data is foreseen.
Experimental reactivity control techniques, related to sub-critical operation, will be developed.
The last two projects are dealing with nuclear data, one at medium and high energy required for
the ADS engineering design including the spallation target (HINDAS), and the other encompassing
the lower energy in resonance regions required for transmutation (n-TOF-ND-ADS).
The objective of the HINDAS project is to collect most of the nuclear data necessary for ADS
application. This will be achieved by basic cross section measurements at different European facilities,
nuclear model simulations and data evaluations in the 20-200 MeV energy region and beyond. Iron,
lead and uranium have been chosen to have a representative coverage of the periodic table, of the
different reaction mechanisms and, in the case of iron and lead, of the various materials used for ADS.
The n-TOF-ND-ADS project aims at the production, evaluation and dissemination of neutron
cross sections for most of the radioisotopes (actinides and long-lived fission products) considered for
transmutation in the energy range from 1 eV up to 250 MeV. The project is starting with the design
and development of high performance detectors and fast data acquisition systems. Measurements will
be carried out at the TOF facility at CERN, at the GELINA facility in Geel and using other neutron
sources located at different EU laboratories. Finally, an integrated software environment will be
developed at CERN for the storage, retrieval and processing of nuclear data in their various formats.
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Table 3. Cluster on transmutation-basic studies
Acronym Subject of research Co-ordinator
(country)
Number of
Partners
Duration
(months)
EC funding
(Million )
MUSE
Experiments for subcritical
CEA (F) 13 36 2.0
neutronics
validation
HINDAS High and intermediate UCL (B) 16 36 2.1
energy nuclear data
for ADS
n-TOF-ND-ADS ADS nuclear data CERN 18 36 2.4
The second call for proposals has been published in October 2000 with a deadline in January 2001.
This call has been targeted on the areas, which were not sufficiently well covered by the projects selected
after the first call, such as preliminary engineering design studies for an ADS demonstrator and
technological support. A new item, networking, has been included in this call. But the areas of chemical
separation and basic studies are not included, as they were well covered in the first call.
5. ADS related research activities in the framework of the International Science and Technology
Centre (ISTC)
The International Science and Technology Centre (ISTC) was established by an international
agreement in November 1992 as a non-proliferation programme through science co-operation. It is an
intergovernmental organisation grouping the European Union, Japan, the USA, Norway, the Republic
of Korea, which are the funding parties, and some countries of the Commonwealth of Independent
States (CIS): the Russian Federation, Armenia, Belarus, Georgia, Kazakhstan and Kyrgyzstan. The
ISTC finances and monitors science and technology projects to ensure that the CIS scientists,
especially those with expertise in developing weapons of mass destruction, are offered the opportunity
to use their skills in the civilian fields.
A Contact Expert Group (CEG) on ADS related ISTC projects has been created in January 1998.
Its main objectives are to review proposals in this field and to give recommendations for their funding
to the ISTC Governing Board, to monitor the funded projects and to promote the possibilities of future
or joint research projects through the ISTC. Five topics have been identified for the ADS related
projects: (i) accelerator technology, (ii) basic nuclear and material data and neutronics of ADS, (iii)
targets and materials, (iv) fuels related to ADS and (v) aqueous separation chemistry. Because the
funding parties primarily respond to local scientific/political interests and pressure, it was decided in
January 2000 to reorganise the CEG into “local” CEGs (EU, Japan, Republic of Korea and USA) with
some inter-co-ordination between them. This inter-co-ordination should foster exchange of
information between ISTC projects in the same field, even if they are supported by different funding
parties.
The EU CEG should develop co-operation between ISTC and FP5 EU funded projects. In fact,
collaboration has already started between EU scientists, not necessarily belonging to the CEG, and
CIS research teams both in the preparation of ISTC proposals and in the follow-up of projects in some
specific areas. Links with FP5 projects will be established, once the projects have actually started. An
area where the co-operation between ISTC and FP5 EU funded projects could be improved is that of
basic nuclear data for ADS.
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6. Community research for the period 2002-2006
The Commissioner responsible for research in the EC launched the idea of a “European Research
Area” in a communication [5] in January 2000. The intention is to contribute to the creation of better
overall working conditions for research in Europe. The Communication is applicable to all areas of
research. The starting point was that the situation concerning research in Europe is worrying, given the
importance of research and development for future prosperity and competitiveness.
In October 2000, the Commission adopted a communication for the future of research in Europe,
which sets out guidelines for implementing the “European Research Area” initiative, and more
particularly the Research Framework Programme [6]. It is proposed to change the approach for the
next Framework Programme, based on the following principles:
• Focusing on areas where Community action can provide the greatest possible “European
added value” compared with national action.
• Closer partnership with the Member States, research institutes and companies in Europe by
networking the main stakeholders.
• Greater efficiency by channelling resources to bigger projects of longer duration.
The Commission's proposals take account of the results of the evaluation of the previous
Framework Programmes carried out by an Independent Expert Panel.
In practical terms, the following arrangements are proposed:
• Networking of national research programmes through support for the mutual opening-up of
programmes and EU participation in programmes carried out in a co-ordinated fashion.
• Creation of European networks of excellence by networking existing capacities in the
Member States around “joint programmes of activities”.
• Implementation of large targeted research programmes by consortia of companies,
universities and research centres on the basis of overall financing plans.
• Greater backing for regional and national efforts in support of innovation and research
conducted by small and medium enterprises (SME).
• More diversified action in support of research infrastructures of European interest.
• Increase in and diversification of mobility grants not only for EU researchers but also for
researchers from third countries. Measures in respect of human resources in research are
proposed, including the “Women and Science” Action Plan.
• Action to strengthen the social dimension of science, in particular in matters concerning
ethics, public awareness of science and giving young people a taste for science.
At its meeting in November 2000, the Research Council supported the general approach of the
Commission as set out in its communication aiming at the continuation of the implementation of the
“European Research Area”. It further noted the importance of the Framework Programmes as strategic
tools to achieve the creation of the “European Research Area” and to increase the efficiency of research
activities in Europe. Finally, the Council invited the Commission to transmit to it formal proposals
concerning the Sixth Framework Programme (FP6) (2002-2006) during the first quarter of 2001.
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In view of the future research programme, the EURATOM Scientific and Technical Committee
has prepared a report on the strategic issues to be considered in the development of the appropriate
nuclear energy research strategies in a 20-50 year perspective [7]. The main message is that “a key
R&D objective should be to ensure that future generations have a real selection of available
technologies to choose from when they have to decide on the energy supply system that would best
suit their needs and acceptance criteria. Therefore, R&D on technical options with a capacity to
contribute significantly to base-load electricity supply must be carried out, including the fission and
fusion options.” They also stress that, given the increased competition in the deregulated electricity
market, public financing will be increasingly needed to ensure that society maintains and develops the
scientific and technical infrastructure needed as a basis for long-term industrial development and
competitiveness. In the area of nuclear fission, continued support should be given to maintain and
develop the competence needed to ensure the safety of existing and future reactors. In addition,
support should be given to explore the potential for improving present fission technology from a
sustainable development point of view (better use of uranium and other nuclear fuels, whilst reducing
the amount of long-lived radioactive waste produced).
The detailed discussions about the content of FP6 have now started.
7. Conclusion
The research activities in the field of partitioning and transmutation under the EURATOM Fifth
Framework Programme have now begun. At present, the research projects are addressing the chemical
separation of long-lived radionuclides and the acquisition of technological and basic data, necessary
for the development of an accelerator-driven system. Other projects are expected for the next call for
proposals with a deadline in January 2001. This call is targeted on the areas, which were not
sufficiently well covered by the projects selected after the first call, such as preliminary engineering
design studies of an ADS demonstrator, technological support and networking. Both, the present
projects and those, which will be selected next year, should contribute significantly to providing a
basis for evaluating the practicability, on an industrial scale, of partitioning and transmutation for
reducing the amount of long lived radionuclides to be disposed of.
Concerning international co-operation, a Japanese research organisation is already participating in
one of the FP5 projects without EU funding. The EU Contact Expert Group on ADS is fostering
collaboration between EU and CIS research teams by linking FP5 and ISTC EU funded projects,
which are related to ADS. It is hoped that co-operation in the field of partitioning and transmutation
will be extended to other countries in the near future.
The discussions about the scientific content of the Sixth Framework Programme (FP6)
(2002-2006) have just started. FP6 is a strategic tool to achieve the creation of the “European Research
Area”, an idea which was launched by the Commissioner responsible for research in January 2000.
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REFERENCES
[1] “Council Decision of 25 January 1999 adopting a research and training programme (Euratom)
in the field of nuclear energy (1998 to 2002)”, Official Journal of the European Communities,
L 64, March 12th, 1999, p.142, Office for Official Publications of the European Communities,
L-2985 Luxembourg.
[2] OECD/NEA, Actinide and Fission Product Partitioning and Transmutation – Status and
Assessment Report, 1999, OECD Nuclear Energy Agency, Paris, France.
[3] C. Rubbia et al., Conceptual Design of a Fast Neutron Operated High Power Energy Amplifier,
CERN/AT/95-44 (ET), (1995).
[4] M. Hugon, Overview of the EU Research Projects on Partitioning and Transmutation of Longlived
Radionuclide, (2000), Report EUR 19614 EN, Office for Official Publications of the
European Communities, L-2985 Luxembourg.
[5] Towards a European Research Area, Communication from the Commission, COM (2000) 6,
18 January 2000, http://europa.eu.int/comm/research/area.html.
[6] Making a Reality of the European Research Area: Guidelines for EU Research Activities
(2002-2006), Communication from the Commission, COM (2000) 612, 4 October 2000,
http://europa.eu.int/comm/research/area.html.
[7] Scientific and Technical Committee Euratom, Strategic Issues Related to a 6th Euratom Framework
Programme (2002-2006), (2000), Report EUR 19150 EN, Office for Official Publications of the
European Communities, L-2985 Luxembourg, www.cordis.lu/fp5-euratom.
184

ACTIVITIES OF OECD/NEA IN THE FRAME OF P&T
Luc Van den Durpel, Byung-Chan Na, Claes Nordborg
OECD/Nuclear Energy Agency
Le Seine Saint-Germain, 12, Boulevard des Iles, 92130 Issy-les-Moulineaux, France
E-mail: vddurpel@nea.fr
Abstract
Back in 1989, the OECD/NEA started a comprehensive programme of work in the field of partitioning
and transmutation (P&T). This programme was initiated by a request from the Japanese government
who was launching a programme on P&T (OMEGA project) and invited the OECD/NEA to coordinate
an international information exchange programme on P&T.
This OECD/NEA Information Exchange Programme has since then resulted in several activities,
among them the Information Exchange Meetings and two state-of-the-art systems studies next to
scientific aspects being handled by the Nuclear Science Committee.
The Nuclear Science Committee covers a wide range of scientific aspects of P&T. Aspects ranging
from accelerators, nuclear data and integral experiments, chemical partitioning, issues on fuels and
materials, and physics and safety of transmutation systems. This paper will overview the activities of
the past ten years and will give insight in the ongoing projects, the results, and the perceived future
activities.
185

1. Introduction
Back in 1989, the OECD/NEA started a comprehensive programme of work in the field of
partitioning and transmutation (P&T). This programme was initiated by a request from the Japanese
government who was launching a programme on P&T (OMEGA project) and invited the OECD/NEA
to co-ordinate an international information exchange programme on P&T.
This OECD/NEA Information Exchange Programme has since then emerged in several activities,
among them the Information Exchange Meetings and two state-of-the-art systems studies.
Since the NEA was invited to take up this topic in 1988, the interest in it has grown in several of
our Member countries. The task is one of long-term scientific research, but it is recognised that certain
short- or medium-term benefits could also be derived. There is quite a rich network of bilateral
agreements on P&T between OECD Member countries. However, judging from the number of
participants who have come a long way to the Information Exchange Meetings and the different
workshops, there is a clear view that substantial benefits can be achieved from wider international
activities and co-operation.
NEA has recently reorganised the P&T activities as a horizontal project between the Nuclear
Development and Nuclear Science Committees and a restructuring of the science programme under
the umbrella of a new working party on scientific issues in P&T has recently been started.
Figure 1 shows, for general information, the past increasing interest and participation in the
Information Exchange Meetings. The shown trends let us also reflect on the future character of these
meetings.
Figure 1. Historic overview of participation to the OECD/NEA Information Exchange Meetings
200
180
160
140
120
Australia
Belgium
China
Czech Republic
Finland
France
Germany
Italy
Japan
Korea
Netherlands
Russian Federation
Spain
Sweden
Switzerland
United Kingdom
United States
EC
IAEA
OECD/NEA
# participants minus host country # countries
100
80
60
40
20
0
Mito City, 1990 Argonne, 1992 Cadarache, 1994 Mito City, 1996 Mol, 1998 Madrid, 2000
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2. The previous 10 years
The activities of OECD/NEA in the field of P&T were initiated in 1989 by the proposal from the
Japanese Government to conduct an international Information Exchange Programme under the
umbrella of the Nuclear Development Committee. The objective of this OECD/NEA Information
Exchange Programme on Actinide and Fission Product Partitioning and Transmutation were defined
as to enhance the value of basic research in the subject by facilitating the exchange of information on
and discussion of programmes, experimental procedures and results. The Information Exchange
Meetings are integral part of this Programme intending to bring a biannual review of the state-of-theart
of P&T and is co-organised by the Secretariat and major laboratories in Member countries. Next to
this Information Exchange Programme, other activities under the umbrella of the Nuclear Science
Committee are undertaken and will be detailed in following section.
The first Information Exchange Meeting was held at Mito City (Japan) in November 1990 [1].
Various scientific and policy aspects of P&T were addressed and highlighted several disparate
approaches which had been taken, covering a variety of aqueous and non-aqueous chemical
procedures and a number of different reactor and accelerator based transmutation schemes. As this
meeting stressed the need to have some small specialists meetings on suitable topics, OECD/NEA
organised two of those specialists meetings. The first one handled on partitioning technologies (Mito
City, Japan, November 1991) [2] where the second one focused on the topic of accelerator-based
transmutation (PSI, Switzerland, March 1992) [3].
In November 1992, the Argonne National Laboratory hosted the second Information Exchange
Meeting [4]. The papers presented indicated that one common thread was the need for some means of
taking an integrated view of the expected benefits and possible disadvantages of including P&T in the
nuclear fuel cycle. Among other results of such an approach would be guidance on research needs. A
number of emerging important issues were identified during the meeting, including the legal
background, the incentives and the implications for the whole fuel cycle in different countries. One of
the main conclusions was that a comparison of system studies in the field of P&T, some of them
already in progress, should form the central part of the P&T activities under the umbrella of the
Nuclear Development Committee of the OECD/NEA.
These views were carried forward at the third meeting, hosted by CEA at its Cadarache site in
December 1994 [5]. Several participants from 11 countries, together with Russia, the IAEA and the
European Commission attended the meeting that primarily focused on P&T strategic systems studies.
The meeting provided a solid basis for approaching a more co-ordinated NEA project, which was
started in early 1996, on the benefits and penalties of adding P&T to the nuclear fuel cycle. This
meeting also concluded that there was a clear need to define objectives against to which to measure
the potential benefits of P&T. However, the discussion at the Cadarache meeting indicated that a final
set could not yet be established.
The fourth meeting was hosted by STA, JAERI, JNC and CRIEPI and was held again in Mito
City, Japan, in September 1996 [6]. The goals for P&T were set clearer during this meeting, i.e. P&T
would not replace geological disposal, the potential hazard reduction was mainly associated with TRU
elements, and reduction of the dose impact to man would come from mobile fission product
radionuclides such as 129 I and 135 Cs. The main motivation for P&T was considered being based on
ethical reasons for the future generations and public claims concerning geological waste disposal sites.
The meeting also indicated that there was a need to better define the performance evaluation by the
use of criteria. Those criteria would relate to the feasibility and credibility of the achievable reductions
in mass and toxicity reduction, the corresponding cut-off period, the best way for industrial
implementation and to a reasonable level of extra costs. Achievement of this kind of evaluations
187

would need the continuation of technical studies and of systems and strategic studies including the
necessary economical evaluations.
The Fifth Information Exchange Meeting [7], hosted by the Belgian Nuclear Research Centre
SCK•CEN and co-organised by the European Commission, was held in November 1998 at Mol
(Belgium) with more than 130 participants from 15 countries and 3 international organisations. This
meeting could be characterised by two main directions; first of all, a more integrative view on
partitioning and transmutation was observed where consensus on the way to perform P&T was
achieved but where questions were raised on the added-value of P&T in the nuclear fuel cycle (and the
most appropriate way to achieve it). Secondly, the breakthrough in partitioning of minor actinides was
achieved on laboratory scale at pre-set performances.
3. Systems studies
In response to the discussions and conclusions of these meetings, two systems studies were
conducted by the NDC. Both aiming at a comprehensive authoritative state-of-the-art and assessment
report on the role, feasibility and developments of P&T.
The first report, entitled Status and Assessment Report of Actinide and Fission Product
Partitioning and Transmutation, was published in 1999 [8] and was the result of a two-year’s work by
an expert group. The report investigates different options to decrease the final radiotoxicity and
provides a limited systems analysis of the main options as a step towards clarifying choices among this
complex set of possible alternatives. The preliminary systems analysis starts from the present technical
state of the art in the fuel cycle and points to some possible developments in P&T technologies which
would result in an advanced fuel cycle with an overall reduction of the radiotoxic inventory and a
reduced impact on the biosphere. The main general conclusions of this report state:
• Fundamental R&D for the implementation of P&T needs long lead-times and requires large
investments in dedicated fast neutron spectrum devices, extension of reprocessing plants, and
construction of remotely manipulated fuel and target fabrication plants.
• Partitioning methods for long-lived radiotoxic elements have been developed on a laboratory
scale.
• Recycling of plutonium and minor actinides could stabilise the transuranium nuclides
inventory of a nuclear power park. Multiple recycling of transuranium nuclides is a long-term
venture that may take decades to reach equilibrium of inventories.
• Conditioning of separated long-lived nuclides in appropriate matrices which are much less
soluble than glass in geological media, or which could serve as irradiation matrix in a delayed
transmutation option, is a possible outcome for future decades.
• P&T will not replace the need for appropriate geological disposal of high level waste, irradiated
transuranium concentrates, and residual spent fuel loads from a composite reactor park.
The ongoing systems study on “Comparative Study of ADS and FR in Advanced Nuclear Fuel
Cycles” complements this first study by looking to the role of P&T in different nuclear fuel cycle
schemes and the specific role of accelerator-driven systems (ADS) versus fast reactors (FR). In
addition, the technological status and issues in developing these ADS or FR including the needed
developments in the fuel cycle, and finally a cost/benefit analysis and the R&D-issues related to P&T
are part of the study’s scope. An international expert group has been set-up and will finalise its
analysis by mid-2001.
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4. Scientific issues of P&T
In parallel with the activities on P&T of the Nuclear Development Division the Nuclear Science
Committee (NSC) covers various scientific issues in response to the perceived need expressed during
the Information Exchange Meetings.
Where, in principle, the NSC focuses a rather broad field of topics, its initial activities related to
P&T were in essence oriented towards two substantial aspects, i.e. the creation of data bases and the
organisation of inter laboratory comparison exercises. Later, these activities were expanded to include
also other aspects of P&T, e.g. radiochemistry, accelerators, and others.
Today, the programme of work of the NSC in the frame of P&T comprises scientific aspects
studied by its working parties and expert groups (or task forces), the NEA Data Bank activities on
nuclear data and computer programmes, and finally the organisation of workshops (see Figure 2).
Figure 2. Overview of today’s OECD/NEA’s activities within the frame of P&T
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A Task Force to investigate the physics aspects of different transmutation concepts was set-up in
the early 1990s. The resulting overview report, published in 1994 [9], described the basic features of a
number of different transmutation concepts.
In addition, the Working Party on Evaluation Co-operation (WPEC) worked previously on the
High Priority List of Measurements which covers a list of actinides needing further measurement
activity, especially in the above 20-MeV part of the neutron energy-range [10]. Co-ordination of
differential nuclear data measurements was undertaken as well as an effort was started on intermediate
energy nuclear data measurements, evaluation and nuclear model codes development and
benchmarking.
In 1996, the NSC Expert Group on Physics Aspects of Different Transmutation Concepts
organised a benchmark exercise to investigate the physics of complex fuel cycles involving
reprocessing of spent PWR reactor fuel and its subsequent reuse in different reactor types: PWRs, fast
reactors and an accelerator-driven system. The results of the comparison of the calculated activities for
189

individual isotopes as a function of time for different plutonium and minor actinide transmutation
scenarios in different reactor systems can be found in reference [11].
New activities were launched since 1998 and included workshops on high power accelerators
reliability [12]. The first of this kind was held in October 1998 aiming the exchange of information
between the accelerator and nuclear community in the light of accelerator driven systems for P&T
purposes. The NSC conducted the second workshop on high-reliability accelerators in France at 23-27
November 1999 [13].
A speciation technology workshop was held in 1999 [14] and a workshop on pyrochemistry has
recently been organised in October 2000. The NSC also reviewed in 1998 current and developing
actinide separation chemistry via a workshop and a report by an expert group [15].
The NSC and Data Bank continue to fulfil their role to provide physics tools and data that enable
improved transmutation calculations to be performed. Ongoing activities relate to:
• Nuclear data and benchmark calculations.
− Activities were conducted in the field of intermediate energy nuclear data. Three studies
were completed in the early ‘90s:
v One on the availability of experimental data and nuclear model codes.
v A second on the requirements for an evaluated nuclear data file.
v The third study was an international comparison of the performance of computer
codes used in intermediate energy calculations. The results of this exercise were
published in 1994 [16]. It was followed by a specialist’s meeting in June 1994 [17],
which recommended the systematic compilation of experimental intermediate energy
data to assist the theoreticians and evaluators in their work. The Data Bank started to
compile these data into the internationally maintained EXFOR database [18].
−
−
−
A co-operative project on an evaluated intermediate energy nuclear data file, through its
Working Party on International Evaluation Co-operation.
Related activities involved a specialist’s meeting in 1997 on the optical model dealing
with higher energies required in accelerator-driven technologies [19].
Benchmark exercises were undertaken and reported in 1997 on the predictive power of
nuclear reaction models and codes for calculation of activation yields in the intermediate
energy range (up to 5 GeV) [20].
• Shielding aspects of accelerators, targets and irradiation facilities [21].
• Criticality (including sub-criticality) benchmarks.
• A neutronic benchmark of an accelerator-driven minor actinide burner has performed. A
summary of the results is presented in this meeting [22] and the final report of the benchmark
will be issued in mid-2001
Recently, the NSC meeting of June 2000 endorsed the creation of the Working Party on Scientific
Issues in Partitioning and Transmutation (WPPT) in order to guide the P&T-related activities of NSC
in the future.
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This Working Party will envelop the scientific aspects of P&T and comprises four sub-groups:
• Group on Accelerator Utilisation and Reliability:
− This group emerges from the previous workshops on Accelerator Utilisation and
Reliability, will synthesise the improvements made and draw conclusions from each
workshop held and continue to organise such workshops. The group will also deal with
target and window performances, for instance, issues on spallation products and thermal
stress and radiation damage, respectively.
• Group on Chemical Partitioning:
− The existing expert group on Pyrochemistry moves under this WPPT where this subgroup
will first focus on the drafting of a state-of-the-art report on Pyrochemistry. Despite its
name, the group will also look into aqueous processing issues.
• Group on Fuels and Materials, as the new proposed transmutation systems will demand
specific materials to be validated or developed for use in more challenging irradiation
conditions.
• Group on Physics and Safety of Transmutations Systems:
−
This group will organise theoretical and experiment-based benchmarks to validate
nuclear data as well as calculation tools needed for simulating advanced transmutation
systems, and investigate safety aspects of transmutation systems such as the beam trip
problem of ADS.
5. Other committees
Contacts with the Radioactive Waste Management Committee (RWMC) guarantee that an
exchange of resulting information is shared among the related committees. The RWMC issued a
statement on P&T in 1992 [23] primarily to emphasise the point that actinide P&T cannot be
considered as an alternative to geologic disposal. The RWMC continues general studies related to the
development of geologic repositories but there are no activities specifically directed at P&T.
In early 2000, the NEA Steering Committee and Committee of Standing Technical Chairs
decided to extend the Information Exchange Programme as a horizontal activity within OECD/NEA,
emphasising the multi-disciplinary character of P&T and also responding to a more transparent
structure of NEA’s activities in this field. Therefore, both the NDC and NSC will co-organise this
Programme and input from other committees, especially the Radioactive Waste Management
Committee, would be searched for. A new web page has also been launched as a joint activity [24].
6. Possible future activities
Beyond the known NEA activities related to P&T, it is noted that confident decisions or actions
regarding a P&T fuel cycle requires significant extensions of existing technology throughout the backend
of the fuel cycle. However, examination of the historical literature reveals a number of important
areas of technology that have received relatively little or dispersed attention.
191

Therefore, while there is continuing need for improved data and analyses on a broad front to
provide the basis for governmental decisions, the following areas are perceived as deserving particular
attention in the future and appropriate OECD/NEA activities are considered in the future:
• Repository analysis: one of the primary benefits of P&T is believed to be a reduction in risk
from the geologic repository. Despite this, there have been very few studies to quantify the
extent and uncertainty of the risk reduction. Most of the historical studies have used hazard or
toxicity indexes that do not properly account for the migration of various radionuclides. The
recently published first phase report [9] covered into some extent the impacts of P&T on the
long-term repository risk. However, there remains a pressing need to analyse these impacts of
P&T on long-term repository risk in all future work and this will again be addressed within
the ongoing second phase P&T systems study. Such analysis will request the collaborative
effort from RWMC.
• Fuel cycle impacts: recycling MAs results in significant changes in the composition of the
fuel or targets that contain them. However, the impacts of these changes on the out-of-reactor
fuel cycle (e.g. design of facilities and transportation casks, safety impacts) are not yet fully
documented and were only partially covered in previous assessment studies. Further detailed
studies in this area are needed and are partially included in the second phase study on P&T.
• Systems analyses: despite the many studies that have been conducted over decades, the
number of systems studies that comprehensively compare the advantages and disadvantages
of standard and P&T fuel cycles is small. Such an analysis should be a key component of
NEA recommendations to member governments and, as such, constitutes a major need. The
first and second phase of NEA’s P&T systems studies and the specific NSC-activities have
been targeted to cope with this specific demand.
• Safety of P&T related installations: the current renewed interest in ADSs and also in FRs for
P&T purposes has been evolving such that currently small-scale or demonstration facilities
are proposed. Despite the still long-term venture for P&T activities and construction of
specific installations, one should already consider basic studies on safety related issues and
especially on the ADS related aspects. Some of these aspects have been covered by existing
or ongoing systems studies but input of specific expertise could be welcome.
• Nuclear data: whilst overall assessment studies are important, the need for basic nuclear data
to perform the underlying neutronic calculations are of very high importance. Especially the
move to harder neutron spectra in fast reactor systems and accelerator-driven systems, in
addition to the specific aspects of spallation neutron spectra and the increasing contents of
minor actinides in the fuel, support the need for a continuous and increased need for validated
nuclear data files. Specific effort is needed in the intermediate energy domain and NSC-Data
Bank has conducted work on this since the early 1990s.
• Materials science: the new proposed transmutation systems will demand specific materials to
be validated or developed for use in more challenging irradiation conditions. This materials
science domain including not only structural materials but also target and fuel materials
becomes more dominant in such P&T schemes. Increased effort in experience exchange is
needed and has already partially been covered by the NSC International Fuel Performance
Experiments (IFPE) database. Activities on a Material Damage Database and a report on the
correlation between dpa-calculations and real damage are planned within NSC.
• Separation chemistry, being a very important part of any P&T scheme, needs continuous
development especially in the light of new developments in aqueous as well as in dry
separation methods. NEA’s activities cover the follow-up of recent developments and needs.
192

• Demonstration Experimental Programme: while different Member countries study P&T and
plan to conduct or are conducting an experimental programme aiming the scientific and
technological demonstration of P&T and especially ADS, there is scope for an international
co-ordinated joint project in this domain. This joint undertaking would supplement the
Information Exchange Programme and aim rationalisation of the different international
initiatives in order to support collaborative development in the domain.
7. Conclusions
OECD/NEA has organised a structured programme of work during the past ten years. The two
main committees involved in this programme are currently pursuing system studies, nuclear data
evaluations and benchmarks, and organise specific workshops on scientific aspects of P&T.
The increasing interest in P&T has brought the OECD/NEA to regroup some of its scientific
activities in a new working party WPPT in order to respond better to the perceived needs of the P&Tcommunity
and governments.
REFERENCES
[1] OECD/NEA, Proceedings of the Information Exchange Meeting on Actinide and Fission
Product Separation and Transmutation, Mito City, Japan, 6-8 November 1990, (1991).
[2] OECD/NEA, Proceedings of the Workshop on Partitioning of Actinides and Fission Products,
Mito City (Japan), 19-21 November 1991, CEA-CONF-11066 (1991).
[3] OECD/NEA. Proceedings of the Specialists’ Meeting on Accelerator-based Transmutation,
PSI-Villigen, Switzerland, 24-26 March 1992, (1992).
[4] OECD/NEA, Proceedings of the Information Exchange Meeting on Actinide and Fission
Product Separation and Transmutation, Argonne National Laboratory, United-States,
11-13 November 1992, (1993).
[5] OECD/NEA, Proceedings of the Information Exchange Meeting on Actinide and Fission
Product Separation and Transmutation, Cadarache, France, 12-14 December 1994, (1995).
[6] OECD/NEA, Proceedings of the 4th Information Exchange Meeting on Actinide and Fission
Product Separation and Transmutation, Mito City, Japan, 11-13 September 1996, (1997).
[7] OECD/NEA, Proceedings of the 5th Information Exchange Meeting on Actinide and Fission
Product Separation and Transmutation, Mol, Belgium, 25-27 November 1998, EUR 18898 EN,
(1999).
[8] OECD/NEA, Status and Assessment Report on Actinide and Fission Product Partitioning and
Transmutation, Paris, France (1999).
193

[9] OECD/NEA, Overview of Physics Aspects of Different Transmutation Concepts,
NEA/NSC/DOC(94)11, (1994).
[10] http://www.nea.fr/html/science/docs/pubs/hprl.pdf.
[11] OECD/NEA, Calculations of Different Transmutation Concepts – An International Benchmark
Exercise, ISBN 92-64-17368-1, 2000.
[12] OECD/NEA, Workshop on Utilisation and Reliability of High Power Proton Accelerators,
Mito, Japan, 13-15 October 1998, (1999).
[13] http://www.nea.fr/html/science/hpa2/.
[14] OECD/NEA, Speciation, Techniques and Facilities for Radioactive Materials at Synchrotron
Light Sources Workshop Proceedings, Grenoble, France, 4-6 October 1998, (1999).
[15] OECD/NEA, Actinide Separation Chemistry in Nuclear Waste Streams and Materials, Paris,
France, (1997).
[16] OECD/NEA, International Code Comparison for Intermediate Energy Nuclear Data, Paris,
France, (1993).
[17] OECD/NEA, Intermediate Energy Nuclear Data: Models and Codes, Proceedings of a
Specialists’ Meeting, Paris, France, 30 May-1 June 1994, Paris, France (1994).
[18] http://www.nea.fr/html/dbdata/x4/welcome.html.
[19] OECD/NEA, Nucleon-Nucleus Optical Model up to 200 MeV, Proceedings of a Specialists’
meeting, Bruyères-le-Chatel, France, 13-15 November 1996, (1997).
[20] OECD/NEA, International Codes and Model Intercomparison for Intermediate Energy
Activation Yields, NSC/DOC(97)-1, (1997).
[21] OECD/NEA, SATIF-4 – Shielding Aspects of Accelerators, Targets and Irradiation Facilities.
Proceedings of 4th Specialists Meeting, Knoxville, Tennessee, USA, 17-18 September 1998,
(1999).
[22] M. Cometto, B.C. Na and P. Wydler, OECD/NEA Benchmark Calculations for Accelerator-
Driven Systems, 6th Information Exchange Meeting, Madrid, Spain, 2000, EUR 19783 EN,
OECD/NEA (Nuclear Energy Agency), Paris, France, 2001.
[23] OECD/NEA, Statement by the Bureau of the RWMC on the Partitioning and Transmutation of
Actinides, NEA/SEN/RWM(92)3, (April 1990).
[24] OECD/NEA, P&T web page can be consulted at http://www.nea.fr/html/pt/.
194

SESSION II
THE NUCLEAR FUEL CYCLE AND P&T
J. Bresee (DOE) – J.P. Shapira (CNRS)
195

RECENT TOPICS IN R&D FOR THE OMEGA PROGRAMME IN JAERI
Toshitaka Osugi, Hideki Takano, Takakazu Takizuka, Yasuo Arai,
Toru Ogawa, Shoichi Tachimori, Yasuji Morita
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Zip: 319-1195, Japan
Abstract
The R&D for partitioning and transmutation (P&T) technology has been carried out in Japan since
1988 under the OMEGA (Options for Making Extra Gains from Actinides and fission products)
programme. In this programme JAERI has proposed the double-strata fuel cycle concept as a
partitioning and transmutation system for long lived radioactive nuclides. The system consists of three
technical areas or processes of the partitioning, nuclear transmutation and fuel processes. This paper
summarises the JAERI’s activities on these topics, focusing on the recent technical achievements in
each process.
197

1. Introduction
The double-strata fuel cycle concept has been proposed by JAERI as a partitioning and
transmutation system for long lived radioactive nuclides. Mukaiyama et al., reviewed the activities in
JAERI for research and development of this concept [1]. The system consists of the following three
technical areas or processes, the partitioning, nuclear transmutation and fuel processes.
• Partitioning process: The four-group partitioning process (4-GPP) has been developed to
separate the elements in high-level liquid wastes (HLLW) into transuranium elements (TRU),
Tc and Platinum group metals (PGM), Sr-Cs group, and others. TRU are separated by
extraction with diisodecylphosphoric acid (DIDPA), Tc and PGM by precipitation through
denitration, Sr and Cs by adsorption with inorganic ion exchangers (titanic acid and zeolite).
A hot verification test was performed using concentrated real HLLW. As a modification
effort of the present 4-GPP, a more powerful ligand, tridentate diglycolamide (DGA), has
been studied to extract actinides directly from HLLW. From fundamental studies, tetraoctyl
3-oxapentandiamide (TODGA) was selected as the most proper DGA extractant.
• Transmutation process: A lead-bismuth cooled accelerator driven system (ADS) with nitride
fuel has been proposed as a dedicated transmutation system to be deployed into the second
stratum. The 800 MWt plant has a pool-type configuration and a power conversion system
operated on a saturated steam cycle.
• Fuel process: Nitride is suitable for the fuel material for MA transmutation from the
viewpoint of supporting hard neutron spectrum and heat conduction ability. In addition,
actinide mononitride with NaCl-type structure will have a mutual solubility leading to the
flexibility of fuel composition. Pyrochemical processing has several advantages over wet
process in case of treating MAs concentratedly with large decay heat and fast neutron
emission. One of the drawbacks of nitride fuel is that nitride with 15 N enriched nitrogen must
be used to minimise the 14 C production. But the pyrochemical process has the practical
feasibility of recovering expensive 15 N.
In this paper, the recent JAERI’s activities in the P&T technology development are reviewed by
focusing on the major technical achievements in each process shown in Figure 1.
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Figure 1. Partitioning and transmutation system for long-lived radioactive nuclides at JAERI
2. Partitioning process
A hot verification test of the 4-GPP with concentrated real HLLW was carried out in the
Partitioning Test Facility in the hot cell [2]. For the preparation of the concentrated HLLW, about 14L
(11 TBq) of the raffinate from the co-decontamination cycle of Purex Process were first denitrated and
then concentrated to about 2.5L. The raffinate was obtained by two reprocessing tests with about 1 kg
of spent fuel burned up to 8 000 MWd/t and with about 1.5 kg of spent fuel burned up to
31 300 MWd/t. The flow sheet is shown in Figure 2.
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Figure 2. Flow-sheet of 4-group partitioning process
Results of the present test well agreed with the either result of previous tests using the
unconcentrated real HLLW and the simulated HLLW added with a small amount of real HLLW. Table 1
summarises the fractional distribution of each element at the 1st mixer-settler. More than 99.998% of
Am were extracted from the HLLW with the organic solvent containing 0.5M DIDPA – 0.1M TBP, and
99.986% of Am were back-extracted with 4M nitric acid. Cm showed the same behaviour as Am. Np
and Pu were extracted simultaneously in a high yield, and more than 99.9% of them were back-extracted
with oxalic acid. In the denitration step for the separation of Tc and PGM, pH of the solution was
increased to 2.8 after the denitration, and then more than 90% of Rh and more than 97% of Pd were
precipitated. About half of Ru were remained in the denitrated solution, but the remaining Ru were
quantitatively precipitated after neutralization of the denitrated solution to pH 6.4, which was performed
for the preparation of the feed solution to the adsorption step with the inorganic ion exchangers. In the
adsorption step, both Sr and Cs were separated effectively. Decontamination factor for Cs was more than
10 6 in all the effluent samples.
Table 1. Fractional distribution (%) of each element at the 1st mixer-settler (from [2])
Element Raffinate Stripped with
4M HNO 3
Solvent Mass balance
Am

The present 4-GPP necessitates a pre-treatment step, i.e. denitration-filtration, to reduce the
acidity of an aqueous feed in harmony with the capability of diisodecylphosphoric acid (DIDPA).
Thus as a modification effort of the 4-GPP, to eliminate the above step, more powerful ligand to
extract actinides directly from high-level liquid waste (HLLW) has been developed in line with the
CHON principle [3]. We found that a tridentate diglycolamide (DGA) with an ether oxygen at a centre
of diamide molecule forms more stable complex with trivalent actinides and lanthanides than that of
bidentate malonamides. Aiming at a high solubility of the extracted metal-DGA complex in n-
dodecane solvent, to avoid the third-phase formation, and a high extractability toward all actinides, we
have examined the effect of chain length of alkyl groups attached to the two amidic nitrogen atoms,
and selected a ligand, tetraoctyl 3-oxapentandiamide (called TODGA hereafter) as the most proper
DGA-extractant. To facilitate the development of TODGA-partitioning process, fundamental studies
on i) the extraction behaviour of TODGA for various valency states of actinides, trivalent lanthanides,
and some fission products ii) radiolytic degradation of TODGA by 60 Co gamma rays have been carried
out. The results of above investigation revealed that the TODGA is a satisfactory extractant to be
applied to the process of separation of actinides and lanthanides (III) in the 4-GPP.
3. Transmutation process
Preliminary design of an 800-MWt lead-bismuth cooled accelerator-driven system has been
developed as a dedicated transmutation system to be deployed into the second stratum in the double
strata fuel cycle [4]. The system employs lead-bismuth for the target and primary coolant material.
The plant has a pool type configuration and a power conversion system operated on a saturated steam
cycle. An analysis of beam trip transient was made for this type of accelerator-driven transmutation
plant [5]. Transients of the primary coolant temperature, the water/steam temperature, the water/steam
pressure, the turbine flow rate, and the electric output were calculated using a simple network model
based on a simplified flow diagram of ADS plant shown in Figure 3. The plant and turbine trips were
required at 380 s after beam trip to prevent from overcooling. The maximum temperature swing was
185°C in lead-bismuth, and 82°C for in water/steam for the case when beam recovered at 370 s.
Figure 3. Simplified flow model of ADS plant
Mass balance in the proposed double strata transmutation system was analysed for cases where
the type of power reactors in the first stratum is UO 2 -LWR, MOX-LWR and FBR. The analysis shows
that the transmutation rate for (Pu, MA) composition from a MOX-LWR becomes one half than that
201

from an UO 2 -LWR. The number of transmutation systems and the amount of transmuted minor
actinide are estimated for several possible scenarios of the future nuclear power development,
assuming the deployment of transmutation systems starts in 2030. It was concluded that the
introduction of ADS could play a significant role as “transmuter” in the back-end of fuel cycle [6].
A code system “ATRAS” was developed for the neutronics design of ADS [7]. The code system
consists of the nucleon-meson transport code, Sn code and burn-up analysis code. In order to obtain
the nuclear data required for the development of ADS, the Actinide File and the High Energy File
were developed along with the JENDL General Purpose File.
4. Fuel process
Fabrication of MA nitride, irradiation tests of nitride fuel and the development of pyrochemical
process for nitride fuel have been carried out [8].
4.1 Fuel fabrication process
Fabrication of Pu and MA-bearing nitrides and preparation of the thermodynamic database have
been carried out besides the irradiation tests of (U,Pu)N fuel up to 4.6 at%. High-purity AmN and
(Pu,Cm)N were fabricated by carbothermic reduction of the dioxides by use of 243 Am and 244 Cm
nuclides. X-ray diffraction patterns showed almost the single phase of NaCl-type structure. On the
other hand, PuN pellets containing inert matrix nitrides such as ZrN and TiN were fabricated and
characterised. Vapour pressure of Np(g) over NpN, (U,Np)N and (Np,Pu)N was measured by hightemperature
mass spectrometry to clarify thermodynamic properties of the solid nitride phase.
Thermodynamic property of Np(C,N), which is an intermediate product of carbothermic reduction for
fabricating NpN from NpO 2 , was also evaluated by both experiments and calculation. The results
suggested that Np (C,N) could be treated as ideal solid solution as is the case of Pu (C,N).
Measurements of heat capacity and thermal expansion of NpN and PuN are underway by use of the
sintered sample for preventing oxidation. The irradiation of two (U 0.8 Pu 0.2 )N fuel pins at fast test
reactor JOYO was completed in 1999 under the joint research with JNC. The non-destructive post
irradiation examinations are underway and any failure of fuel pins was not observed. The destructive
examinations will start in the latter half of this year.
4.2 Fuel reprocessing
As for pyrochemical process, the electrochemical dissolution behaviour of NpN and PuN were
measured by cyclic voltammetry and the equilibrium potentials of the nitrides in LiCl-KCl eutectic melt
were determined. On the other hand, the electrochemical deposition behaviour of Pu at liquid Cd cathode
was investigated. In this case the potential of deposition and dissolution shifted positively compared with
the case of solid cathode in correspondence with a thermodynamic stabilisation by formation of
intermetallic compound. Indeed, the formation of PuCd 6 phase was observed at the cathode by
microprobe analysis. By adjusting electrochemical parameters such as current density during electrolysis,
ten-gram scale of Pu was recovered at liquid Cd cathode with high Pu concentration. In addition, the
electrochemical deposition behaviour of Np at liquid Cd cathode and the phase relationship of Am-Cd
binary system were experimentally studied. Nitrogen releasing behaviour from NpN and PuN at an
anode, and the results of distillation and nitrization of the Cd cathode after the electrolysis were
examined. It was proved that the pyrochemical process is fundamentally suitable for recovery of
expensive 15 N-enriched nitrogen gas compared with the wet process.
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5. Concluding remarks
In 1999, the Atomic Energy Commission of Japan (AEC) established the Advisory Committee on
Nuclear Fuel Cycle Back-end Policy to conduct the Check-and-Review of the outcome of the
OMEGA programme. The Committee concluded, in the report issued in March 2000, that the present
status of the programme would be the level of basic studies and tests, and that various concepts of the
P&T system were evaluated and required technologies were developed. They also concluded that the
future R&D should be proceeded in order to convert the high level waste into useful resources and to
reduce the environmental impact associated with its disposal. Their recommended processes are as
follows; to study the P&T implementation scenario taking account the situation of nuclear fuel cycle
in Japan, to carry out basic experiments to demonstrate the feasibility of the process, and to conduct
engineering scale experiments to obtain safety data of these systems.
After getting the results of the above mentioned C&R by AEC, JAERI will proceed the R&D in
each process of the P&T system from the basic experimental step to the engineering mock-up step
including the following areas Figure 4:
• Pb/Bi material test: One of the technical issues for ADS developments is the
corrosion/erosion of material in liquid-bismuth coolant. The beam window represents a major
technical challenge in ensuring the structural integrity as it suffers a high differential pressure
load as well as thermal stress and radiation damage.
• Am and Cm characteristics: A “high temperature chemical cell” is to be constructed in
Nuclear Fuel Cycle Safety Engineering Facility of JAERI for gram-scale experiments of Am
and Cm.
• ADS experimental facility: Another issues to be tested or to be demonstrated are sub-critical
reactor physics, and system operation and control. The experimental programmes to solve
these technical issues for ADS developments have been planed within the framework of the
JAERI-KEK Joint Project for High-Intensity Proton Accelerators.
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Figure 4. Scenario for development of ADS
204

REFERENCES
[1] T. Mukaiyama, T. Takizuka, M. Mizumoto, Y. Ikeda, T. Ogawa, A. Hasegawa, H. Takada and
H. Takano, Review of Research and Development of Accelerator Driven System in Japan for
Transmutation of Long-lived Nuclides, submitted to the special edition of “Progress in Nuclear
Energy” devoted to ADS R&D, Sept 2000.
[2] Y. Morita, I. Yamaguchi, T. Fujiwara, H. Koizumi and S. Tachimori, A Demonstration of
4-Group Partitioning Process with Real High-level Liquid Waste, Proceedings of the
International Conference ATALANTE2000 – Scientific Research on the Back-end of the Fuel
Cycle for the 21st Century, October 24-26, 2000, Avignon, France, Paper No. P3-37.
[3] Y. Sasaki, Y. Sugo and S. Tachimori, Actinide Separation with a Novel Tridentate Ligand,
Diglycolic Amide for Application to Partitioning Process, ATALANTE2000 – Scientific
Research on the Back-end of the Fuel Cycle for the 21st Century, October 24-26, 2000,
Avignon, France, Paper No. O2-07.
[4] K. Tsujimoto, T. Sasa, K. Nishihara, T. Takizuka, H. Takano, K. Hirata and Y. Kamishima,
Study of Accelerator-driven System for Transmutation of High-level Waste from LWR, Proc. In
7th International Conference on Nuclear Engineering, Tokyo, Japan, 19-23 April, 1999,
ICONE-7092.
[5] T. Takizuka, H. Oigawa, T. Sasa, K. Tsujimoto, K. Nishihara, H. Takano, M. Hishida and
M. Umeno, Responses of ADS Plant to Accelerator Beam Transients, OECD/NEA 2nd
Workshop on Utilisation and Reliability of HPPA, Aix-en-Provence, France, 1999
[6] H. Takano, K. Nishihara, K. Tsujimoto, T. Sasa, H. Oigawa and T. Takizuka, Transmutation of
Long-lived Radioactive Waste Based on Double-strata Concept, GENES-3, Tokyo, Japan,
14-17 Dec. 1999.
[7] T. Sasa, K. Tsujimoto, T. Takizuka and H. Takano, Accelerator-driven Transmutation Reactor
analysis Code System – ATRAS –, JAERI-Data/Code 99-007(1999).
[8] Y. Arai and T. Ogawa, Research on Nitride Fuel and Pyrochemical Process for MA
Transmutation, 6th Information Exchange Meeting on Actinide and Fission Product P&T,
Madrid, Spain, Dec. 11-13, 2000, EUR 19783 EN, OECD/NEA (Nuclear Energy Agency),
Paris, France, 2001.
205

TRANSURANICS TRANSMUTATION ON FERTILE AND
INERT MATRIX LEAD-BISMUTH COOLED ADS
E. González, M. Embid-Segura, A. Pérez-Parra
CIEMAT
Av. Complutense, 22, 28040 Madrid (Spain)
E-mail: enrique.gonzalez@ciemat.es
Abstract
Different strategies for the back-end of the nuclear waste are explored, including different strategies
of ADS application to nuclear waste transmutation. In this paper the results of the detailed simulation
studies of ADS systems, both with fertile (Th) and inert (Zr compounds) matrix fuels, but always with
lead-bismuth coolant will be presented. In addition, several options are considered for the plutonium
isotopes: direct burning in ADS together with the minor actinides, a separate partial burning in MOX
LWR before its load to the ADS and intermediate solutions. Depending on the case, the studies are
performed from two perspectives: the situation of the equilibrium of the fuel cycle and the approach
to the equilibrium from the actual LWR discharge composition.
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1. Introduction
CIEMAT is actively working on the evaluation of the possible roles of ADS systems on the
nuclear waste management within a collaboration agreement with ENRESA, the Spanish enterprise
responsible for the radioactive waste management. Different strategies for the back-end of the nuclear
waste are explored, from direct disposal to different strategies of ADS application to nuclear waste
transmutation.
In this paper the results of the detailed simulation studies of ADS systems, both with fertile (Th)
and Inert (Zr compounds) matrix fuels, but always with lead-bismuth coolant will be presented. In
addition, several options are considered for the plutonium isotopes: direct burning in ADS together
with the minor actinides, a separate partial burning in MOX LWR before its load to the ADS and
intermediate solutions. Depending on the case, the studies are performed from two perspectives: the
situation of the equilibrium of the fuel cycle and the approach to the equilibrium from the actual LWR
discharge composition.
The studies are grouped in two wide groups. The first one is based on an ADS with a MOX fuel
based on a ThO 2
matrix and the second one for the inert matrix cases is based on ZrN plus AcN. The
ADS systems are similar for the two cases but not exactly the same, on the other hand the
methodology of the detailed simulations is in both cases the same, and always based on the
EVOLCODE system [1].
2. ADS systems main characteristics and simulation methodology
The ADS concept used in all the studies includes a fast core with an hexagonal arrangement of
fuel elements cooled by lead (fertile matrix) or lead bismuth eutectic (inert matrix) in forced
convection, and operates at constant thermal power close to 800 MW th . The external neutrons are
produced in a windowed spallation target, of the same material that the main coolant, by the action of
a 1 GeV proton beam. The mass and composition of the fuel depends on the case.
Figure 1. Side and top view of the ADS core concept used for the inert matrix simulations
The ADS used for the fertile fuel case had already been presented in several papers and
conferences [2,3]. In the inert matrix cases, the core geometry has been slightly modified, including a
total of 132 fuel assemblies, to introduce 12 special rod positions (see Figure 1). These positions are
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eserved for control bars, shutdown bars, sample irradiation channels, special instrumentation and
others, however in the present studies they have been considered as filled with coolant. Table 1 gives
additional details on the inert matrix ADS concept.
Table 1. Inert matrix ADS parameters
Hexagonal fuel subassemblies
Flat to flat
Total height
Active length
Subassembly wall thickness
Power + Primary circuit
210.96 mm
150 cm
120 cm
5 mm
Nominal power
800 MW th
Coolant/Convection type Pb/Bi E./Forced
Inlet temperature 300°C
Outlet temperature 450°C
Core
Fuel
(Zr,TRU)N
TRU elements
Pu, Np, Am, Cm
Coolant and moderator
Pb/Bi
Cladding material
Steel HT9
Configuration
Hexagonal
Number of fuel assemblies 132
Number of special rod positions 12
Proton beam and spallation target
Kinetic energy
1 000 MeV
Beam pipe material
HT9
Beam window
Steel
Vacuum beam pipe thickness
3 mm
Vacuum beam pipe external diameter 200 mm
Fuel pins
Number of pins per subassembly Var. 169 – 331
Pitch (mm) Var. 15 – 10.7
External radius of fuel pins
4.1 mm
Cladding thickness
0.35 mm
Void thickness
0.1 mm
External radius of fuel pellets
3.65 mm
Internal radius of fuel pellets
0.55 mm
The simulation of the ADS systems, their k eff values, power distributions and isotopic
composition evolution during burn-up has been performed using the EVOLCODE system. The
system is based on the combination of: LAHET [4] for the simulation of the neutron spallation in lead
produced by the proton beam, and the transport of these neutrons down to 20 MeV; MCNP4B [5] for
the complete neutron transport by Monte Carlo for energies below 20 MeV, and to calculate the
neutron multiplication, the neutron flux energy spectra at different positions inside the core, the
neutron flux intensity magnitude and distribution, the specific power distributions and the energy
release by fission; and ORIGEN2.1 [6] with ad-hoc libraries for the burnup calculations. Further
details on EVOLCODE can be found in [1]. For the purpose of the simulations of material evolution
with burn-up, each fuel assembly is logically subdivided in 10 longitudinal zones.
3. Transmutation based on fertile or inert matrix ADS
Two approaches are considered in the CIEMAT transmutation studies. The first one uses a
Th matrix (ThO 2 ) for the fuel. The matrix provides chemical, mechanical and thermal characteristics
very similar to the well known MOX fuels, and in addition, the breeding required to achieve very
long burn-ups of the fuel (1 500 days). On the other hand, at the end of the transmutation process a
substantial amount of 233 U has been bred from the Th matrix. This fuel cycle concept will make sense
if the 233 U is used in the LWR substituting the 235 U or if the U-Pu cycle was to be replaced by the Th-U
cycle. This second option will provide a much smaller production of transuranics and finally the
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adiotoxicity to be managed could also be reduced. In this last approach the use of TRU from the
LWR in the ADS will be a transitory operation where most probably the equilibrium of the cycle will
not be reached before the TRU are exhausted (depending on the different countries strategies). The
study will concentrate on the first cycles of the TRU burning on ADS.
The most recent studies are devoted to evaluate the inert matrix option both in the mixed oxides
or mixed nitrides versions. This option has the advantage of not introducing new isotopes in the fuel
cycles, although the enrichment on some of the higher actinides becomes in certain phases unusually
high. In the hypothesis of stability of the fraction of energy generation from the fission process, either
the present LWR or new reactor types should provide most of energy and the ADS will only
contribute with a small fraction of the total produced energy. In this circumstances the transmutation
ADS has to handle a continuous amount of TRU regularly being produced at the same rate that they
are eliminated. As in most strategies of transmutation on ADS, fuel recycling inside the ADS is
required to obtain high elimination levels. It can be easily demonstrated that these two conditions are
sufficient to progressively approach in the ADS to an equilibrium fuel composition. The behaviour of
the system after reaching equilibrium decides the final TRU elimination efficiency of the system. For
these reasons the studies of inert matrix concentrate on the fuel cycle after the equilibrium has been
reached.
4. Fertile matrix (Th based) fuel option for TRU elimination
The concept explored in this study has been to close the LWR fuel cycle operation by
introducing all the transuranic isotopes, TRU, contained in the LWR nuclear wastes, after 40 years
cooling time, homogeneously in a fuel based on a thorium matrix. This fuel in the MOX chemical
form is used in a fast ADS using lead as coolant. The ADS is then operated for a total of 1 500 days at
a mean power of 800 MW th
reaching a burn-up of 146 GWd/THM (the burn-up for TRU reaches
238 GWd/T). The right choice of TRU/Th allows to obtain these very long burn-ups without
interruption of the ADS operation, by the precise compensation of fissile isotopes consumption and
breeding (mainly from the Th matrix). The ADS fuel is reprocessed after discharge, assuming an
uniform 99.9% efficiency for all actinides. The fission and activation products and the reprocessing
losses are stored in an appropriated repository. The uranium recovered, mainly 233 U, is available to be
used in the operation of other reactors or ADS systems devoted to energy production. The recovered
TRUs are mixed with fresh thorium and new TRUs from the LWR nuclear wastes to produce the new
cycle fuel. Figure 2 and [2,3] provide more details of the global fuel cycle.
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Figure 2. Fuel cycle assumed in the fertile fuel ADS TRU transmutation studies
The first eleven cycles of operation of such an ADS system had been carefully studied in detail.
This number of cycles might be sufficient to exhaust most of the TRU contained in the nuclear wastes
produced by one reactor generation (from beginning of nuclear reactors till the end of life of the
presently installed LWR) of a country with moderate nuclear energy production (10 GWe) in a small
number of ADS systems. The fuel composition of each reload is carefully tuned in order to maintain a
sufficiently stable neutron multiplication (between 0.96 and 0.98) and to optimise the transmutation
efficiency. The respect of the thermomechanical limits of the fuel during the burn-up is also verified.
Figure 3 shows the masses of the different components of the fuel for each reload: the TRU
recovered from the previous cycle, the new TRUs from the LWR wastes, the total Th and the total
actinide mass. In addition, the figure also shows the accumulated TRU from LWR, accumulated Th
entered in the system, the eliminated TRU and the recovered U after each cycle. It can be observed
that these last four quantities increase linearly with the cycle number after the 4th cycle. The amount
of TRU remaining in the ADS after each cycle decreases progressively approaching a constant value,
as a consequence that as more and more cycles are performed and equilibrium is approached, the
TRU transmutation efficiency increases reaching at the latest cycles very high values. The total TRU
mass loaded in the eleven cycles was 11.9 tons while at the discharge of the 11th cycle the total
remaining TRU is close to 2.0 tons. This means a global cumulative elimination ratio close to 83.1%.
When reprocessing losses are taking into account the global cumulative TRU elimination efficiency is
82.8% in 11 cycles.
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Figure 3. Mass composition of the different ADS reloads and evolution of cycle parameters
In addition to the mass reduction, the isotopic composition changes with the transmutation
cycles. This is a consequence of the difference of the cumulative elimination efficiencies for the
different isotopes. The evolution of this parameter for the most abundant TRU isotopes is presented in
Figure 4. For all the main components of the LWR TRUs the cumulative elimination efficiency
increases with the cycle number, reaching at the 11 th cycle values as high as 94% for 239 Pu, 93% for
237
Np, 91% for 241 Am, 70% for 240 Pu, 49% for 241 Pu, 56% for 242 Pu and 27% for 243 Am. 238 Pu is not
produced nor eliminated and the curium isotopes are continuously produced in the system, but in any
cases the final masses of these isotopes represent only 7% of the TRUs in the ADS discharge after the
11 th cycle.
Figure 4. Evolution of the cumulative elimination efficiencies for the most abundant TRU
The theoretical limits of this system with very large number of cycles had been computed
assuming that the 11 th cycle is a good representative for the behaviour of the system at equilibrium.
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Reprocessing losses both from the LWR and the ADS reprocessing had been taken into account
assuming extraction efficiencies of 99.9% for all TRUs. The asymptotic limits of these curves had
been computed giving as result that the 0.34% of the LWR TRU will end on the final storage, from
reprocessing losses.
5. Inert matrix fuel options for TRU elimination
A parametric study of the characteristics of different inert matrix fuels with a ZrN matrix and
different Pu-MA fractions [7] (from a cycle with UO 2
LWR and a single pass for all the Pu in LWR
MOX fuel and in an ADS with 20 tons of nitride fuel), showed that the evolution of the ADS
neutronic multiplication varies from a rapidly falling neutron multiplication to a continuous breeding
to configurations close to critical as the MA fraction increases (see Figure 5). Of particular relevance
is the existence of Pu-MA mixtures that allow achieving very long burn-ups with minimum variation
of the neutron multiplication during the ADS operation. It is also important to note that the fraction
Pu/TRU in these stable mixtures is approximately 40%, half the fractions produced in OU 2 LWR or in
similar cycles.
The detailed studies on inert matrix fast ADS applications to nuclear wastes elimination had been
performed on the scope of the cycle described in Figure 6. The UO 2
fuel is consumed in LWR. The
resulting spent fuel is reprocessed separating four streams: the recovered depleted U, the Pu, the
minor actinides (MA), and the fission fragments, activation products and reprocessing losses. The Pu
is used to produce MOX and then this fuel is used once in LWR. The Pu and MA in the spent MOX
and the MA from the LWR are send to the ADS described in Figure 1 and Table 1. The spent fuel of
the ADS is then continuously recycled after reprocessing and addition of more Pu and MA from the
spent UO 2
and MOX from the LWR.
Figure 5. Evolution of the k eff for different inert matrix
Pu-MA fuel mixtures in a fast Pb-Bi cooled ADS
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Figure 6. Fuel cycle assumed in the inert matrix ADS TRU transmutation studies
Because the transformation of the actinide isotopic composition vector along the ADS cycle
(ADS burn-up, cooling, reprocessing and storage) is contractive, the fuel composition at the
beginning of irradiation will progressively approach to an equilibrium value from cycle to cycle as far
as the feed from the LWR reprocessing is kept constant. For the long-term consideration and in the
hypothesis of maintenance of the present level of energy production from fission, the relevant
information is the performance of the ADS cycle after this equilibrium has been reached. The isotopic
composition of this equilibrium ADS fuel depend on the isotope vector coming from the ADS, the
ADS characteristics and the burn-up per ADS cycle. This last dependence is however small and
variations in the burn-up from 600 to 1 500 days introduce corrections smaller than 15% in the main
isotopes. On the other hand the ratio between the LWR (UO 2
and MOX) TRUs and the total fuel mass
in the ADS fuel depends strongly on this burn-up. The ADS equilibrium fuel composition was
computed for the cycle of Figure 6. This fuel includes 2.5% U, 3.8% Np, 72.8% Pu, 13.1% Am and
7.8% Cm. The operativity of the ADS loaded with this fuel in nitride form distributed in a ZrN matrix
was studied. Figure 7 shows the evolution of the neutron multiplication with the burn-up, for two
different fuel masses of the ADS, 10 and 20 tons. The figure shows that it will be very difficult and
expensive (from the accelerator point of view) to maintain the operation more than 150 days.
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Figure 7: Evolution of k eff of an inert matrix fast ADS
loaded with the equilibrium fuel of the cycle of Figure 6
This peculiarity of the ADS loaded with this fuel will be a serious difficulty. On one hand, it will
require frequent interruptions of the ADS that will reduce its energy production cost competitivity,
and on the other hand, it will mean many reprocessing passes for the TRU before it is significantly
reduced. Many possibilities can be envisaged to mitigate this difficulty in the ADS application to the
transmutation of TRUs in equilibrium with simple LWR energy producing cycles. One type of
possibilities already proposed by other authors are: the continuous or quasi-continuous fresh fuel
supply by means of liquid fuels (e.g. molten salts), particle fuels (e.g. pebble-bed fuels), sliding fuel
assemblies or designs of cores that allows to move fresh and spent fuel assemblies between the ADS
core and a region neutronically decoupled inside the main vessel. A different possibility is the use of
burnable absorbers or control rods in order to maintain a stable sub-criticality level. A third option,
implicitly included in the double strata concept, consist in changing the isotopic composition of the
equilibrium fuel. What is need is to severely reduce the Pu content on the equilibrium fuel. This can
be achieved by reducing the Pu from the LWR reprocessing. This plutonium can not be simply stored,
the natural option should be to use its potential as energy producing fuel by continuously reprocessing
it on (critical or sub-critical/LWR or fast) reactors devoted to energy production. In this paper two
additional options are discussed: the use of the equilibrium fuel in several batches and the use of a
partially fertile matrix for the ADS.
5.1 Inert matrix fuel ADS for TRU elimination: batches with equilibrium fuel
One possible method to extend the burn-up of the fuel in the case of equilibrium fuel in an inert
matrix ADS configuration is to irradiate the fuel in batches. Figure 8, shows a sketch of the possible
batch refueling scheme studied in this paper. The approach is OUT-IN, with the fresh fuel coming to
the ADS periphery there the fuel is irradiated for a period of time (166 days). When the neutron
multiplication has fall below the accelerator possibilities, the ADS stops and the fuel elements move
inward, extracting the inner most batch and introducing again fresh fuel in the periphery. Figure 9
shows the variation of neutron multiplication constant k eff
during one refueling bath. An equilibrium
load has been computed that allows to charge exactly the same amount of fuel per 166 days batch in
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the 4 batches scheme, allowing to achieve a fuel burn-up at discharge of 140 GWd/THM, with a
fluctuation on the k eff
from 0.956 to 0.936 during each batch. The fresh fuel composition introduced at
the periphery has 75.3% Zr, and 24.7% of TRUs, with their equilibrium composition. This OUT-IN
scheme allows also reducing the picking ratio of the power distribution inside the ADS.
Figure 8. OUT-IN refueling scheme studied for the inert matrix ADS with equilibrium fuel
Figure 9. k eff
evolution of an inert matrix ADS
with equilibrium fuel during one of the four refueling batches
5.2 Partially fertile matrix fuel ADS for TRU elimination with equilibrium fuel
Another option to extend the burn-up of the fuel per ADS cycle is the use of partially fertile
matrix adding either 238 U or 232 Th. The peculiarities of fertile fuel with Th had been described in the
previous section. The natural choice for the breeding material should be 238 U. This isotope will
introduce no new isotope in the fuel cycle and the main effect would be the reduction of
transmutation efficiency per cycle. Figure 10 shows that a fuel with 65% Zr, 12.7% U and
22.3% TRU allows to extend the operativity of the ADS loaded with 20 tons of nitride fuel for more
than 500 days. Configurations with 55% Zr, 22.5% U and 22.5% TRU and 21.7 tons of nitride fuel
216

allow to operate for more than 1 200 days. Mixtures of Th with 70% Zr, 7% U and 23% TRU and
18 tons of fuel also allow to obtain operation for more than 500 days. These configurations improve
the achievable burn-up per cycle of the ADS but reduce the transmutation efficiency of TRU. Figure
11 shows the change in TRU elimination from pure inert matrix to the 65% Zr, 13% U and 22% TRU
fuel after 500 days of irradiation. The eliminated TRU mass is 72% of what could be transmuted in a
pure inert matrix ADS if it could be operated for 500 days. The main effect of this reduction is to
increase in the complementary proportion the number of reprocessing passes and the corresponding
reprocessing losses, as well as increasing the time required for TRU elimination. Both inconveniences
are easily acceptable and well compensated for the extension of the single pass burn-up.
5.3 Reprocessing losses estimation
To estimate the TRU fraction finally going to the nuclear waste storage, from the reprocessing
losses, in the inert matrix scenario, the 4 batches refueling concept with 660 total irradiation time and
an average burn-up of 140 GWd/THM, will be used, as the simpler solution for a realistic operation
of an inert matrix TRU transmuter ADS. For these parameters and assuming that the reprocessing
efficiencies are 99.9% for all the TRUs in the reprocessing of the LWR, the MOX and the ADS spent
fuels, simple arithmetic allows to estimate the fraction of TRUs going to the repository between 0.7
and 0.8% of the originally produced. The value obtained from the detailed simulation is 0.707%.
Figure 10. k eff
evolution of an partially fertile (Zr- 238 U) matrix ADS with equilibrium fuel
6. Conclusions
The previous exercises have shown that both the inert matrix and fertile matrix allow to reduce the
amount of TRUs to be stored in the final nuclear waste repository by a factor larger than 100 if the cycle
is maintained sufficiently long. The inert matrix choice is the solution of minimum perturbation of the
present fuel cycle but it has the difficulty of short burn-up per cycle. Several solutions are possible for
this problem, again the minimum deviation from the present cycle would be the use of refueling batches
or partially fertile (Zr- 238 U) matrixes. A more advance solution is the introduction of Pu recycling in the
energy production strata, although this probably will require the use of new types of reactors. Finally the
use of Th based matrix ADS will be more justified as a transition from the U-Pu fuel cycle to the Th-U
fuel cycle for energy production, although intermediate solutions are also possible.
217

Figure 11. Transmutation efficiency per 500 days batch in inert and partially fertile matrix ADS
REFERENCES
[1] E. Gonzalez et al., EVOLCODE: ADS Combined Neutronics and Isotopic Evolution Simulation
System, Presented in the Mathematics and Computation, Reactor Physics and Environmental
Analysis in Nuclear Applications, MC’99 Conference, Madrid, September 1999, 963-974.
[2] J. García-Sanz et al., Neutronic and Isotopic Simulation of a Thorium-TRU’s Fuel Closed
Cycle in a Lead Cooled ADS, CIEMAT Report 920, 2000.
[3] J. García-Sanz et al., Isotopic Composition Simulation of the Sequence of Discharges from a
Thorium TRU’s, Lead Cooled, ADS, Presented in 3rd International Conference on Accelerator
Driven Transmutation Technologies and Applications, ADTTA’99. Prague, 1999.
[4] R.E. Prael and H. Lichtenstein, User Guide of LCS: The LAHET Code System, Group X-6.
MS-B226. LANL, 1989.
[5] J.F. Briesmeister, Editor, MCNP – A General Monte Carlo N-particle Transport Code. Version
4B, LA-12625 M. 1997.
[6] M.J. Bell, ORIGEN – The ORNL Isotope Generation and Depletion Code, V ORNL-4628. 1973.
[7] a) M. Embid et al., Performance of Different Solid Nuclear Fuels Options for TRU Transmutation
in Accelerator Driven Systems, CIEMAT DFN/TR-02/II-00. b) M. Embid et al., Neutronic
Analysis of an Accelerator-based TRU-Inert Matrix Fuel Transmutation System. Meeting of the
Sociedad Nuclear Española. León, October 2000.
218

ACTINIDE AND FISSION PRODUCT BURNING
IN FAST REACTORS WITH A MODERATOR
Igor Yu. Krivitski, Andrei L. Kochetkov
Federal Research Centre – Institute for Physics and Power Engineering
Bondarenko sq.1, 249020, Obninsk, Kaluga region, Russian Federation
E-mail: stogov@ippe.rssi.ru
Abstract
Calculations have been carried out with respect to the transmutation of long-lived wastes (minor
actinides and fission products) in fast reactors using special devices based on large amounts of
moderator material. Such devices can replace both radial and axial blanket sub-assemblies. It has been
shown that the implementation of these devices will allow the achievement of long-lived waste burnup
up to a level of 90-95%, decreasing essentially the radiotoxicity of wastes to be buried.
219

1. Introduction
Spent fuels of modern nuclear reactors contain a rather large quantity of long-lived high-active
wastes (plutonium, minor actinides (MA) and fission products (FP)). This amount of waste increases
with the electricity production. If plutonium can be used as a fuel for reactors (both thermal and fast),
ways should be found for decrease the amount of minor actinides and fission products. Therefore, the
major challenge for the future nuclear energy system is the decrease of possible contamination of the
environment by such wastes.
Nevertheless, the existing reactors are able to solve these type of problems before development
and operation of new promising nuclear systems. However, the implementation of thermal reactors for
solving these problems is not efficient due to essential limitations of core physics. The use of
traditional fast reactors allows solving the task partially. A homogeneous waste addition to the fuel
makes it possible to utilise annually only 15% of all produced MAs in this period and less than 10% of
long-lived FP. However, a noticeable degradation of neutronic parameters (increase in sodium void
reactivity effect (SVRE) and decrease in Doppler-effect) requires a search for other decisions, more
acceptable from a safety perspective.
One of the possibilities involves the implementation of special irradiation devices (ID), which can
be located either in radial or in axial blankets. In this case, the effect on core physics will be much less
as compared with homogeneous recycling. But the efficiency of transmutation in these devices will be
essentially lower.
The efficiency of these devices can be improved by introducing a rather large quantity of
moderator. This is related to the fact that practically all FPs have maximum cross-sections in the
thermal region and MA cross-sections increase as well when shifting the neutron spectrum to the
thermal region. Possibilities for MA and FP efficient transmutation in ID containing moderator are
considered in this report for a fast power reactor of BN-800 type as an example.
2. MA burning in special devices located in the radial blanket
Americium oxide located in a magnesium oxide inert matrix is considered as a fuel composition
for MA burning. A ratio between volume fraction of americium oxide and magnesium oxide can be
varied in such a way that it keeps the total americium loading in the ID.
We noted that fast reactors have an important advantage over thermal reactors for burning minor
actinides because their neutron flux is two orders as higher. However, in a fast spectrum, actinides
have lower cross-sections compared to a thermal spectrum.
On the basis of these two factors an idea appears to use moderated sub-assemblies (SAs) in fast
reactors. In this case we conserve a rather high neutron flux and essentially increase the actinide
cross-sections [1].
As an ID, we can consider a core SA in which part of the fuel pins are replaced by pins containing
americium oxide in an inert matrix and others are replaced by pins containing a moderator. Varying
the number of fuel pins with americium and moderator, one can change a moderator volume fraction
in the SA. For a more essential moderator fraction increase, on can use a promising fuel pin design in
which the central target material rod (of small diameter, with cladding or without it) is surrounded by a
rather thick moderator layer. In this design, it is very easy to vary the ratio of fuel and moderator
220

volume fractions in a wide range conserving the fuel pin external diameter. This type SA design was
used in further studies.
One of the tasks of optimisation studies was a search for moderator material allowing a more
efficient actinides transmutation, which all other things being the same. Figure 1 presents the
dependency of the capture and fission cross-sections of some actinides on material type and Figure 2 –
the dependency of the same cross-sections on moderator volume fraction in the fuel pin for zirconium
hydride as the most efficient moderator.
Figure 1. Dependency of actinide cross-sections on a moderator type
80
70
60
50
40
30
20
10
0
barn
ZrH 2 MgH 2
CeH 3
σ cPu 240 σ fPu 241 σ cAm 241 σ fAm 242m σ cAm 243 σ fCm 243
Figure 2. Dependency of actinide cross-sections on a moderator volume fraction
120
barn
100
0 25 50 75 95
80
60
40
20
0
σ Pu 240 σ Pu 241 σ Am 241 σ Am 242m σ Am 243 σ Cm 243
We noted that the transmutation of americium in such ID would be worthwhile only if the
radiotoxicity of wastes remaining after irradiation is much less than the radiotoxicity of non-irradiated
americium. Figure 3 presents the change in waste radiotixicity for different ID burn-ups in reference to
storage of non-irradiated americium
221

Figure 3. A change in radiotoxicity of ID wastes for different
burn-up relating to storage of non-irradiated americium
1
0
9% 31% 77% 98%
log10(Sv/S0)
-1
-2
-3
-4
-5
0 1 2 3 4 5 6 7
log10(year)
The dependency presented shows that for a decrease in the waste radiotoxocity of at least two
orders, it is necessary to reach 93-95% h.a. americium burn-up. Thus, when designing an ID for
americium burning, we should start from the necessity to reach just this high burn-up with maximum
possible americium loading, and at the same time not to fall outside the existing limitations for basic
structure materials of fast reactor cores.
Figure 4 presents the dependency of americium burning value on irradiation time interval for
different moderator volume fraction (zirconium hydride).
Figure 4. Average americium burn-up as a function of irradiation time
90
Average burnup, %h.a.
60
30
75 25
50 95
0
420 1500 3000 4500
Irradiation time, fpd
It is easy to see that handling the problem of reaching ~90% h.a. burning requires a long
irradiation time (~10-15 years) and a rather high volume moderator fraction in fuel pins (>90%). We
noted that long irradiation of SAs with americium will require the development of special reloading
regimes. For example, in order to eliminate a high burn-up irregularity over a SA, it is necessary to
turn it 180° during irradiation cycle. In this case the maximum damage dose will be ≈200 dpa, which
will require a high performance for the structure materials used in these ID. Large changes in ID
power with americium burn-up will require the development of a special regime for their cooling.
Besides, the introduction of the ID with moderator in the first row of radial blanket will lead to
increase in power of adjacent core SAs. The changing in power of these SAs can reach 20-30%.
However, in our opinion, this power increase is not critical and will not require special measures.
Thus, in order that americium transmutation in ID is appropriate from the standpoint of essential
decrease in actinide radiotoxicity, average americium burn-up should be not less than 90% h.a. This
222

equires irradiation times beyond 10 years, which, in turn, will require the structure material
performing at high damage doses.
The second aspect of long-time irradiation is the dependence of actinide cross-sections and burning
efficiency on cycle number. Taking into account that the irradiation time is more than 10 years and core
lifetime is approximately one and half a year it is necessary to have ~10 cycles of moderated
subassembly irradiations. Dependency of actinide cross-sections on cycle number for two type of subassemblies
are shown on Figures 5a and 5b.
Figure 5a. Dependency of cross-sections on cycle number
(without moderator)
σEDUQ
σ I
$P P
σ F
$P
&\FOHÃQXPEHU
Figure 5b. Dependency of cross-sections on cycle number
(with moderator)
σÃEDUQ
σ F $P
σ I $P P
&\FOHÃQXPEHU
This dependency shows the cross-sections in moderated sub-assembly reach the maximum value
that after approximately 8 cycles. For the non-moderated sub-assembly the cross-sections do not
practically depend on cycle number.
3. Influence of moderated sub-assembly on core parameters
To check the influence of moderated sub-assembly with americium on core parameters we
investigate the dependency of multiplication factor, target burn-up and sub-assembly power on cycle
number.
223

These dependencies are shown on Figures 6 to 8.
Figure 6. Dependence of multiplication factor on cycle number
. HII
ZLWKRXWÃPRGHUDWRU
ε P
=0.31
ε P
=0.69
ε P
=0.59
&\FOHÃQXPEHU
Figure 7. Dependence of target burn-up on cycle number
%ÃÈKD
ZLWKRXWÃPRGHUDWRU
ε
P
=0.31
ε
P
=0.59
ε
P
=0.69
&\FOHÃQXPEHU
Figure 8. Dependence of sub-assembly power on cycle number
:ÃN:O
ZLWKRXWÃPRGHUDWRU
ε
P
ε
P
ε
P
&\FOHÃQXPEHU
224

The dependencies presented allow making the following conclusions:
• The moderated sub-assemblies have a great influence on critical state of reactor and it is
necessary to take special measures to avoid under and sub criticality.
• The power of a moderated sub-assembly is drastically increased during the first cycles of
irradiation due to the formation of isotopes with large fission cross-sections (For example,
Am242m) and than drops to the nominal level.
• The burn-up of 90% h.a. is reached after 5-6 cycle and the following 5% (up to 95%) requires
approximately the same time.
So, the use of moderated sub-assembly for the americium transmutation faces large difficulties,
the main of them being the drastically increase of sub-assembly power during irradiation. This can
result in decreasing the temperature of target, moderator, steel cladding and possible melting of these
components.
4. Fission products transmutation in irradiation devises
Among different FPs, greatly contributing to a long-lived activity, there is the practice of
separating a group of several isotopes: 99 Tc, 107 Pd, 93 Zr, 135 Cs, 129 I, 126 Sn and 79 Se. The problems of
transmuting these isotopes both in thermal and fast reactors have been considered in detail in many
publications [2-4]. There is no longer any doubt that a more efficient transmutation of these nuclides is
reached in a thermal or close to thermal neutron spectra, since the basic resonance of these FPs are just
in this energy range.
Two aspects are considered in this report, connected with FP transmutation in fast reactors with
special ID implementation, containing a large moderator quantity:
• The effect of different moderators on the FP transmutation efficiency.
• The effect of irradiation devices on major neutronic core characteristics.
We considered two possible ways to locate the IDs:
• In the first row of the radial blanket.
• Without the lower blanket.
4.1 Technetium-99 transmutation
Transmutation efficiency. 99 Tc half-life period is 2.13 × 10 5 years, and its production in spent fuel
of modern power reactors comprises 3.0 kg/TWh for fast reactors and 3.2 kg/TWh for thermal
reactors.
Table 1 presents a comparison of 99 Tc transmutation when using different moderators.
It should be noted that the introduction of hydrated moderators makes it possible to obtain the
most efficient transmutation. The effect of volume moderator fraction on the transmutation rate and
absolute value of 99 Tc transmutation is presented in Table 2.
225

When increasing the moderator fraction up to 95%, it is possible to transmute up to 80% of initial
loading of the ID during a once-through irradiation, thus providing ~8% per year. Higher
transmutation rates (up to 25%/year) can be provided by an ID located in the axial blanket.
Table 1. Comparison of 99 Tc transmutation efficiency when using different moderators
Moderator Radial blanket Axial blanket
%/cycle kg/TWh %/cycle kg/TWh
CaH 2 77.3 2.29 35.7 5.33
MgH 2 77.3 2.29 35.7 5.33
TiH 2 71.5 2.12 31.2 4.66
CeH 3 74.6 2.21 33.5 5.01
ZrH 2 72.1 2.14 31.7 4.73
Be 64.0 1.90 26.2 3.92
C 60.1 1.78 24.0 3.58
Be 2 C 65.3 1.94 27.1 4.04
NbBe 17 63.5 1.88 25.9 3.87
Table 2. Effect of moderator volume fraction on
99 Tc transmutation efficiency (radial blanket/axial blanket)
Moderator volume fraction (CaH 2 ),
%
0 22 80 95
%/cycle 11.2/3.4 15.4/4.8 44.0/15.6 77.3/35.7
kg/TW*h 6.0/9.6 7.3/11.1 5.9/10.2 2.3/5.3
However, a small irradiation time (~1.5 year) allows to burn in one cycle only 35% of the loaded
technetium.
Thus, the introduction of a moderator in ID allows an important increase in the transmutation rate.
However, a decrease in this case of the total FP results in a decrease of absolute value of transmuted
technetium.
It should be noted that even an essential decrease of the quantity of long-lived FPs as a result of
the irradiation does still not solve the problem of the activity decrease. It is due to the build-up of other
nuclides from which the radioactivity could exceed the one of the target nuclide. In the case implying
99 Tc, this problem does not exist, since during the irradiation a short-lived 100 Tc isotope and two stable
ruthenium isotopes are produced. The calculation results are presented in Table 3.
226

Table 3. Different isotope contribution into activity after irradiation of 99 Tc, Cu/kg
Isotope Loading Discharge
After cooling, year
Tc 99
Tc 100 17.05
–
3.5
5.26 + 5
3 100 1000
3.5
–
3.5
–
3.5
–
Total 17.05 5.26 + 5 3.5 3.5 3.5
Thus, the ID activity will be defined by 99 Tc both before and after irradiation.
The presented results indicate a rather high efficiency of technetium transmutation in ID with a
moderator. However the required transmutation efficiency should be determined from an economical
point of view. Small technetium quantity transmuted in an irradiation cycle, even for a rather high
transmutation rate, will require much more cycles of irradiated target processing, which increase
irretrievable losses. Larger transmutation volumes, with low transmutation rate, require either a larger
number of IDs in the reactor or an increase in reactor number, which should be loaded by IDs.
4.2 The effect moderator in the blanket on core neutronic parameters
We will consider the effect of a moderator in the axial blanket on the core parameters for 99 Tc
transmutation.
First of all, we noted that the moderator location in the immediate vicinity of the core leads to the
production of thermal neutrons in the moderator which re-enter the core, increasing sharply the fission
rate and, therefore, the power in the nearest core layers. It is obvious in the Figure 9, which shows an
axial power field in most fuel pins of the core.
Figure 9. Axial power distribution for various moderators in axial blanket
50
kW/m
45
40
35
30
25
without
CaH 2
Be
20
15
Border between core and absorber layer
It is necessary in this case to pay attention to the fact that the use of hydrated moderators leads to
a very sharp power increase in the region adjacent to the blanket. The power at the boundary between
the core and the blanket increases to 80%.
227

Even greater power increases are observed when using a beryllium moderator, although in this
case the power increase takes place smoothly along the height of the lower core half. The question if
one of these cases is more dangerous from the standpoint of reactor safety requires the performance of
detailed thermal-hydraulic calculations.
However, the effect of moderator on the power distribution can be essentially decreased by the
introduction of an absorber layer between the core and the blanket with moderator. The Figures 9 and
10 present the axial power distributions when introducing a layer of different type absorber.
Figure 10. Axial power distribution for various absorber
in layer between core and axial blanket (moderator CaH 2 )
N:P
&G2
*G 2
+I
% &
Border between core and absorber layer
These distributions show that best results are achieved when using an absorbing layer with
cadmium oxide. In this case, not only the power at the boundary decreases but also the maximum
power value decreases. We noted that when beryllium is used as a moderator, the effect of absorbers
on the power fields is only observed in the immediate vicinity of the blanket.
The effect of the absorber introduction on transmutation efficiency and SVRE value are
considered further on. The calculation results are presented in Table 4.
Table 4. The effect of absorber on transmutation efficiency and SVRE value
Moderator CaH 2
Absorber – CdO Gd 2 O 3 Hf B 4 C
kg/TWh 5.29 4.59 3.36 3.79 1.90
SVRE, %∆k/k -0.169 +0.145 -0.100 -0.102 -0.311
Moderator Be
Absorber – CdO Gd 2 O 3 Hf B 4 C
kg/TWh 3.76 3.56 2.13 2.38 1.79
SVRE, %∆k/k +0.339 +0.650 +0.171 +0.188 +0.109
The results presented show that the introduction of CaH 2 as a moderator decreases SVRE value to
∼0.5%∆k/k comparing to the use of beryllium. The implementation of different absorbers changes the
SVRE value in the limits of 0.5 %∆k/k.
228

Thus, the introduction of an absorber solves the power distribution problem and at the same time
leads to a decrease in the FP transmutation efficiency. But if CdO is used as an absorber, the decrease
of the transmutation efficiency does not exceed 15%.
4.3 Iodine-129 transmutation
129 I has the largest among all long-lived wastes half-life period, equal to 1.57 × 10 7 years, and its
production amounts are about 0.7 kg/TWh for fast reactors and 0.66 kg/TWh for thermal reactors.
To consider 129 I transmutation, it is necessary to choose the best chemical form for a material
containing large quantities of 129 I. Among different chemical compositions, three compositions are
considered as target materials: BeI 2 , NaI, CeI 3 , the latter having the largest number of 129 I nuclei [5].
Table 5 presents a comparison of 129 I transmutation efficiency for different moderators and three
chemical composition stabilising iodine
Table 5. A comparison of 129 I transmutation for different chemical compositions
BeI 2 NaI CeI 3
kg/TWh %/cycle kg/TWh %/cycle kg/TWh %/cycle
MgH 2 0.71 29.3 1.17 28.6 1.25 28.8
ZrH 2 0.71 29.2 1.17 28.6 1.25 28.8
It is necessary to pay attention to the fact that the introduction of BeI 2 allows the assurance of a
maximum transmutation rate, whereas introducing CeI 3 provides the largest absolute transmutation
efficiency. The results regarding the BeI 2 composition are presented.
The comparison of 129 I transmutation efficiency for implementation of different moderators is
given in Table 6.
Table 6. I 129 transmutation efficiency when using different type moderators
Moderator Radial blanket Axial blanket
%/cycle kg/TWh %/cycle kg/TWh
CaH 2 65.2 0.32 27.0 0.66
MgH 2 68.7 0.33 29.3 0.71
TiH 2 52.2 0.25 19.8 0.48
CeH 3 67.8 0.33 28.7 0.70
ZrH 2 68.6 0.33 29.2 0.71
Be 50.3 0.24 18.7 0.46
C 41.6 0.20 14.8 0.36
Be 2 C 51.3 0.25 19.3 0.47
NbBe 17 47.7 0.23 17.6 0.43
Because the 129 I capture cross-section has a clearly defined maximum in the thermal range, the
best results are obtained when using hydrated moderators (MgH 2 , CeH 3 , ZrH 2 ).
229

The effect of a moderator volume fraction on the transmutation rate and absolute quantity of
iodine transmuted is presented in Table 7.
Table 7. The effect of moderator volume fraction
on 129 I transmutation efficiency (radial blanket/axial blanket)
Moderator volume fraction (ZrH 2 )
%
0 25 70 95
%/cycle 18.2/5.7 35.2/12.0 62.1/25.1 68.6/29.2
kg/TWh 1.8/2.3 2.7/4.6 1.3/2.6 0.33/0.71
An analysis of activity changing for 129 I and the products of its irradiation is presented in Table 8.
Table 8. Different isotope contribution to activity after 129 I irradiation, Cu/kg
Isotope Loading Discharge
129 I
130m I
130 I
131 I
131m Xe
133m Xe
133 Xe
134m Cs
134 Cs
0.18
–
–
–
–
–
–
–
–
0.05
62 272.73
1.15+5
68.18
595.45
1.68
23,55
0.8
1.75
After cooling, year
3 100 1000
0.05 0.05 0.05
– – –
– – –
– – –
– – –
– – –
– – –
0.64 – –
– – –
Total 0.18 1.77+5 0.69 0.05 0.05
Contrary to 99 Tc, an irradiation of 129 I leads to the creation of e.g. 134m Cs while the activity of these
will define the activity of irradiated targets activities during some time (~30 years). Nevertheless, the
irradiation of iodine is also a rather efficient method to decrease the FP radioactivity. However, it is
necessary to pay attention to the fact that the irradiation products are also gaseous xenon isotopes
( 130 Xe, 131 Xe and 132 Xe) creating a rather high pressure in the used targets.
4.4 Palladium-107 transmutation
107 Pd has a half-life period of 6.5 × 10 6 years, and its production in the power reactor fuel
amounts: 1.54 kg/TWh for fast reactors and 0.78 kg/TWh for thermal reactors.
The effect of different moderators on 107 Pd transmutation efficiency is presented in Table 9.
230

Table 9. 107 Pd transmutation efficiency when using different moderators
Moderator Radial blanket Axial blanket
%/cycle kg/TWh %/cycle kg/TWh
CaH 2 63.4 1.89 25.9 3.90
MgH 2 64.0 1.91 26.2 3.95
TiH 2 52.9 1.58 20.1 3.02
CeH 3 58.1 1.73 22.8 3.43
ZrH 2 57.1 1.71 22.3 3.36
Be 77.0 2.30 35.4 5.33
C 67.8 2.03 28.7 4.31
Be 2 C 77.9 2.32 36.2 5.45
NbBe 17 74.6 2.23 33.5 5.05
Contrary to the technetium and iodine, the highest efficiency of palladium transmutation is
provided by using beryllium containing moderators.
The effect of moderator volume fraction on the transmutation rate and absolute quantity of
transmuted palladium is presented in Table 10.
Table 10. The effect of moderator volume fraction on 107 Pd transmutation efficiency
Moderator volume fraction(Be 2 C)
%
0 25 70 95
%/cycle 9.6/2.9 22.7/7.2 49.7/18.2 77.9/36.2
kg/TWh 5.8/8.7 10.6/16.9 6.5/12.0 2.32/5.45
The analysis results for palladium irradiation product activity are presented in Table 11.
231

Table 11. Different isotopes contribution to activity after 107 Pd irradiation, Cu/kg
Isotope
107 Pd
108 Ag
108m Pd
108 Pd
109m Ag
110m Ag
110 Ag
111m Pd
111 Pd
111 Ag
111m Cd
113m Cd
115 Cd
Loading Discharge
0.51
–
–
–
–
–
–
–
–
–
–
–
–
0.11
0.33
5641.60
77589.41
3.53+5
1.31+5
3.30+5
5.0
31.76
32.4
102.8
0.73
2.87
After cooling, year
3 100 1 000
0.11
–
–
–
–
632.00
8.85
–
–
–
–
0.63
–
0.11
–
–
–
–
–
–
–
–
–
–
–
0.01
0.11
–
–
–
–
–
–
–
–
–
–
–
–
Total 0.51 1.06+6 642.1 0.12 0.11
The results presented show that the irradiated palladium activity increases several orders due to
the creation of short-lived nuclides ( 110m Ag, 110 Ag), which then falls quickly, and after approximately
30 years the targets activity will be again defined by palladium only.
4.5 Caesium-135 transmutation
135 Cs has a half-life time of 2.3 × 10 6 years, and its production in the power reactor fuel amounts:
3.7 kg/TWh for fast reactors and 1.4 kg/TWh for thermal reactors.
When analysing the calculation results for possibility to transmute 135 Cs in the targets with a
moderator, as presented in Table 12, a conclusion can be made that in case of 135 Cs irradiation the
choice of moderator does not play an important role, since all moderators give approximately the same
results.
It should be noted that the caesium transmutation rate is somewhat lower compared to the
nuclides considered above, which is explained by a lower capture cross-section.
232

Table 12. 135 Cs transmutation efficiency when using different moderators
Moderator
Radial blanket Axial blanket
%/cycle kg/TWh %/cycle kg/TWh
CaH 2 44.7 0.40 16.2 0.72
MgH 2 44.8 0.40 16.2 0.72
TiH 2 33.6 0.30 11.5 0.51
CeH 3 39.0 0.35 13.7 0.61
ZrH 2 40.2 0.36 14.2 0.64
Be 40.3 0.36 14.2 0.64
C 41.3 0.37 14.7 0.65
Be 2 C 47.3 0.42 17.4 0.78
NbBe 17 42.7 0.39 15.3 0.68
Besides, other caesium isotopes ( 133 Cs, 137 Cs etc.) are accumulated in power reactor spent fuel and
the Cs 135 fraction is ~10% only. When transmuting caesium without chemical isotope separation, the
transmutation efficiency decreases one order due to creation secondary 135 Cs.
Thus, the results presented show that the issue on advising to transmutate 135 Cs in reactor
conditions remains open.
The transmutation of 93 Zr (T 1/2 = 1.53 × 10 6 years, production 1.74 kg/TWh for fast reactors
2.8 kg/TWh for thermal reactors) was not considered in detail in this report due to large uncertainties
in its nuclear data.
The transmutation of such elements 79 Se (T 1/2 = 65 000 years) and 126 Sn (T 1/2 = 10 5 years) is not
considered because of their low transmutation rate.
The analysis results allow forming the final Table 13.
Table 13. A comparison of transmutation rate
for different FP and their production in power reactors.
Isotope T 1/2 Production in reactors,
kg/TWh
Transmutation efficiency
kg/TWh
without moderator with moderator
99 Tc 2.13 × 10 5 3.0/3.2 10.6 5.3
107 Pd 6.5 × 10 6 1.54/0.78 5.5 3.0
135 Cs 2.3 × 10 6 3.70/1.4 5.1 0.8
129 I 1.57 × 10 7 0.70/0.66 5.3 0.7
79 Se 6.5 × 10 4 0.03/0.02 0.09 0.01
126 Sn 10 5 0.15/0.08 0.2 0.04
93 Zr 1.53 × 10 6 1.74/2.8 2.3 1.4
233

Thus the transmutation of such FPs 99 Tc, 107 Pd and 129 I with the use of moderators in IDs is rather
proved, since their transmutation efficiency exceeds their production. The transmutation of 135 Cs, 93 Zr
in such blanket can turn out to be not useful, and a further optimisation of the moderator quantity is
necessary. Moreover, 135 Cs will require its separation from other Cs isotopes produced in spent fuel.
5. Conclusion
The considered method for radioactive wastes (actinides and fission products) transmutation in
special irradiation devices containing large moderator quantity has essential advantages over the
homogeneous method for these wastes to be recycled.
The results presented have shown that to decrease the americium radiotoxicity significantly, a
very high actinide burn-up (up to 95% h.a.) should be achieved in these irradiation devices. In this
case up to 60 kg of americium per year will be destroyed. Such quantity is accumulated presently in all
VVER-100 reactors in Russia.
But this method faces large difficulties in realising such core with moderated sub-assemblies. It is
necessary to take special measures to avoid the large power increase in moderated sub-assembly
during irradiation.
The major long-lived fission products can be also transmuted in such irradiation devices.
However, isotope separation will be needed to increase the transmutation efficiency of 135 Cs and 93 Zr
isotopes, for example. The major advantage of the use of this fission product utilisation concept
consists in the decrease of total losses in each step of waste reprocessing. Nevertheless, a serious
contradiction should be pointed out between the transmutation rate and absolute quantity of utilised
fission products.
The influence of irradiation devices with a moderator on some core neutronic parameters has
been considered. And it has been shown that the implementation of absorbing blankets, made from
cadmium oxide, will allow an essential decrease in the effect of such devices on the core parameters,
decreasing slightly the transmutation efficiency.
234

REFERENCES
[1] M. Salvatores, I. Slessarev and M. Uematsu, Physics Characteristics of Nuclear Power System
with Reduced Term Radioactivity Risk, Nuclear Science and Engineering, 120, 18 (1995).
[2] P. Brusselaers et al., Possible Transmutation of Long-lived Fission Products in Usual Reactors,
Proc. Int. Conf. on the Physics of Reactors PHYSOR 96, Mito, Japan, September 16-20, 1996,
Vol. 3, p. M-101.
[3] A.P. Ivanov, E.M. Efimenko, A.G. Tsykunov, Fast Reactor Application for the Fission
Products Burning, Proc. Int. Conf. on the Physics of Reactors PHYSOR 96, Mito, Japan,
September 16-20, 1996, Vol. 3, p. M-111.
[4] K. Kobayashi et al., Conceptual Core Design to Transmute Long-lived Radioactive Nuclides in
a Large Fast Breeder Reactor, Proc. Int. Conf. on the Physics of Reactors PHYSOR 96, Mito,
Japan, September 16-20, 1996, Vol. 3, p. D-47.
[5] H.R. Brager, L.D. Blackburn, D.W. Wootan, Development of Irradiation Targets to Transmute 129 I,
Trans. of ANS, Vol. 62, p. 103, American Nuclear Society, La Grande Park, Illinois, USA (1990).
235

ASSESSMENT OF NUCLEAR POWER SCENARIOS ALLOWING FOR MATRIX
BEHAVIOUR IN RADIOLOGICAL IMPACT MODELLING OF DISPOSAL SCENARIOS
Eric Tronche 1 , Hubert Boussier 1 , Jean Pavageau 2
1
Commissariat à l’Énergie Atomique (CEA-Valrhô), DCC/DRRV/SSP
BP 17171, 30207 Bagnols-sur-Cèze Cedex, France
2
Commissariat à l’Énergie Atomique (CEA-Cadarache), DRN/DER/SPRC
Abstract
The innovative scientific contribution of this study is to consider a third type of radiotoxic inventory:
the potential radiotoxic inventory after conditioning, i.e. taking into account the containment capacity
of the radionuclide conditioning matrices. The matrix fraction subjected to alteration over time
determines the potential for radionuclide release, hence the notion of the potential radiotoxic
inventory after conditioning. An initial comparison of possible scenarios is proposed by considering
orders of magnitude for the radionuclide containment capacity of the disposal matrices and for their
mobilisation potential. All the scenarios investigated are normalised to the same annual electric power
production so that a legitimate comparison can be established for the ultimate waste forms produced
per year of operation.
This approach reveals significant differences among the scenarios considered that do not appear when
only the raw potential radiotoxic inventory is taken into account. The matrix containment
performance has a decisive effect on the final impact of a given scenario or type of scenario. Pu
recycling scenarios thus reduce the potential radiotoxicity by roughly a factor of 50 compared with an
open cycle; the gain rises to a factor of about 300 for scenarios in which Pu and the minor actinides
are recycled. Interestingly, the results obtained by the use of a dedicated containment matrix for the
minor actinides in a scenario limited to Pu recycling were comparable to those provided by
transmutation of the minor actinides.
237

1. Introduction
Under the provisions of the “separation-conditioning” option of the strategy and program defined
under research topic 1 of the 1991 French radioactive waste management law, various fuel cycle
scenarios will be assessed and compared [1] in terms of feasibility, flexibility, cost, and ultimate
waste radiotoxic inventory. The latter criterion may be further broken down into “potential radiotoxic
inventory” (the radiotoxic inventory of all the radionuclides produced) and “residual radiotoxic
inventory” (the radionuclide fraction reaching the biosphere after migration from the repository).
The innovative scientific contribution of this study is to consider a third type of radiotoxic
inventory: the potential radiotoxic inventory after conditioning, i.e. taking into account the containment
capacity of the radionuclide conditioning matrices. The source term therefore includes only the effects
of the radionuclides released from the altered matrix. The matrix fraction subjected to alteration over
time determines the potential for radionuclide release, hence the notion of their potential radiotoxic
inventory after conditioning. The impact of the radionuclides on the human population is considered
through ingestion alone, and not by inhalation.
2. Allowance for matrix containment capacity
The ultimate containment matrices in the various scenarios considered included uranium oxide in
spent fuel and uranium ore, glass in waste packages containing fission products (FP) and minor actinides
(MA), new containment matrices (NCM) or new glass compositions, and fuel assembly structural
materials or compacted hulls and end-fittings containing structural activation products (SAP).
The potential release is taken into account by means of three parameters, as shown in Figure 1.
The first parameter is the integrity time limit (ITL) after which matrix alteration begins; the ITL could
correspond to the lifetime of the container (e.g. zircaloy cladding or steel package). The second is the
matrix alteration rate, described by the annual altered matrix fraction (AAMF), which is constant
over time; the reciprocal of the AAMF thus corresponds to the matrix lifetime. The third parameter
involves the inherent behaviour of each radionuclide, as expressed by the probability that a particular
radionuclide will be released from the altered portion of the matrix: PR (RN)
.
Figure 1. Schematic representation of radionuclide release over time
unaltered matrix altered matrix Radionuclide fraction totally released
Q 0
Q 0(t)
Qr (t)
Qr (t) = Q 0(t) * (t-ITL) * AAMF * PR (RN)
Time
I I I I
0 ITL ITL+1/(AAMF) ITL+1/(AAMF* PR (RN) )
The equation postulates that the radiotoxic inventory of the radionuclides released from the
matrix Qr (t)
is a fraction of the inventory over time in a closed system Q 0(t)
: the fraction is the integral
over time since the ITL of the product of the annual altered matrix fractions by the radionuclide
238

elease probabilities. The time distribution of the radionuclide inventory in a matrix is thus a uniform
and constant function of ITL up to ITL + 1/(AAMF × PR (RN)
). The parameter values are indicated in
Table 1.
The integrity time limit is first assumed constant for all the packages considered. The AAMF
values for spent fuel and for glass reflect the minimum performance corresponding to their maximum
leach rates; the figure for NCM represents a target value corresponding to 100 times better
containment than currently estimated for glass. Uranium mine tailings were also added as a reference
to assess the possible evolution of the equivalent of uranium ore under the same disposal conditions
as the other matrices. The quantity considered was the natural uranium requirement necessary for one
year of operation of a reactor population at equilibrium; the uranium was assumed to be at
equilibrium with its decay products, and was taken into account here in oxide form.
With regard to their radionuclide release probability factors, the actinides characterised by their
low mobility were assigned a value of 10 -2 over duration equal to 1/AAMF. The release of the highly
mobile fission products (iodine, caesium and technetium) was considered congruent with the matrix
alteration, hence their PR (RN)
value of 1. The other less characteristic chemical elements were assigned
intermediate AAMF values.
Table 1. Parameter values for the radionuclide release equation:
(boldface figures are the product of AAMF × PR (RN)
)
PR (RN)
Matrix
ITL
(years)
AAMF
(year -1 )
Pu
(10 -2 )
Am
(10 -2 )
U
(10 -2 )
I
(1)
Tc
(1)
Cs
(1)
Other FP
(10 -1 )
SAP
(10 -1 )
Uranium oxide 300 10 -4 10 -6 10 -6 10 -6 10 -4 10 -4 10 -4 10 -5
Glass 300 10 -5 10 -7 10 -7 10 -7 10 -5 10 -5 10 -5 10 -6
New containment
matrices
300 10 -7 10 -9 10 -9 10 -9 10 -7 10 -7 10 -7 10 -8
Structural materials 300 10 -4 10 -6 10 -6 10 -6 10 -4 10 -4 10 -4 10 -5 10 -5
3. Application to fuel cycle scenarios
3.1 Scenarios
All the scenarios considered were normalised with respect to an electric power production of
400 TWh/year. The scenarios also assumed quasi steady-state operation to allow valid comparisons of
the ultimate waste production over one year of operation in each case. The twelve scenarios taken into
account are briefly described below, and can be considered as belonging to four major types:
• Scenarios resulting in large quantities of plutonium and minor actinides in the waste materials.
Open-cycle scenario, and once-through-Pu scenario in which plutonium is recycled once as
MOX fuel without further reprocessing.
• Scenarios eliminating the plutonium from the waste materials.
FNR-Pu, PWR/FNR-Pu, and PWR-Pu scenarios, in which plutonium is recycled repeatedly
either as MOX fuel in pressurised water reactors (PWR) or fast neutron reactors (FNR), or as
239

PWR MIX (MOX with enriched uranium) fuel; a variant with isotopic separation of 242 Pu
was also considered: IS-Pu242.
• Scenarios eliminating the plutonium from the waste materials, with separation and specific
conditioning of the minor actinides.
These are variants of the preceding scenarios with implementation of enhanced separation
and conditioning (SC) techniques. The fuel is reprocessed in an enhanced reprocessing plant
using the new DIAMEX and SANEX processes to separate americium (Am) and/or curium
(Cm) for incorporation in a new containment matrix (NCM) with very high radionuclide
retention performance (Table 1). The vitrified waste therefore contains only fission products
(except for process losses). These scenarios are designated: PWR-Pu/MA-SC,
FNR-Pu/MA-SC, and PWR/FNR-Pu/MA-SC.
• Scenarios eliminating the plutonium and some or all of the minor actinides from the waste
materials for transmutation in PWRs or FNRs.
These are the PWR-Pu/MA, PWR-Pu/NpAm (Cm is sent to vitrification), FNR-Pu/MA, and
PWR/FNR-Pu/MA scenarios in which the minor actinides are transmuted in homogeneous
mode, and the PWR/FNR-AmCm-target and PWR/FNR-Am-target scenarios in which the
actinides are transmuted in heterogeneous mode as once-through targets; after irradiation,
90% of the minor actinides are transmuted into fission products.
3.2 Fuels
The burn-up is assumed equal to 60 GWd·t -1
for all the PWR fuels, and approximately
140 GWd·t -1 for the fast neutron reactor fuels. The U and Pu reprocessing losses are assumed equal to
0.1%. All the fuel compositions and their annual flows are determined by neutronic feasibility
analysis, calculated by the Reactor and Fuel Cycle Physics Department of the CEA’s Nuclear Reactor
Division (DRN/SPRC) [2]. The comparisons were performed under steady-state conditions based on
the total annual production.
The potential radiotoxic inventory of the ultimate wasteforms over time were calculated by
multiplying the activities of each radionuclide (determined by decay using the JEF2.2 data [3]) by the
dose-per-unit-intake factors (Sv·Bq -1 ) from ICRP72 [4].
4. Results
4.1 Mass balance and “raw” potential radiotoxic inventory
The mass balance was established for the fission products, minor actinides and plutonium
released from the waste (Table 2). The structural activation products have a lower radioactive and
radiotoxic impact. The reprocessed uranium is not considered as an ultimate waste form, unlike the
uranium reprocessing losses (estimated at 0.1%).
240

Table 2. Annual heavy nuclide contribution (kg/year)
to ultimate waste form for each scenario
Recycling policy
No
recycling
Partial Pu
Pu recycling
Scenario
Open cycle Once-through
Pu
Isotopic
separation 242 Pu
MIX
Pu
FNR
Pu
PWR/FNR
Pu
U 754 736 79 105 743 747 238 440
Pu 10 332 3 175 2 021 17 56 35
Np 746 683 632 674 172 499
Am 645 1 229 1 852 1 853 1 425 1 406
Cm 113 286 360 951 112 196
Total (excl. U) 11 835 8 373 4 865 3 495 1 776 2 136
Recycling policy
Pu + MA recycling
Scenario
MIX
Pu + MA
MIX
Pu + AmNp
PWR/FNR
Am/Cm targets
PWR/FNR
Am targets
PWR/FNR
Pu + MA
FNR
Pu + MA
U 735 720 437 438 378 236
Pu 25 22 96 95 39 58
Np 1 1 1 1 1 1
Am 3 3 16 15 2 2
Cm 3 1 668 94 252 1 1
Total (excl. U) 32 1 694 207 363 43 62
When the raw radiotoxic inventory results are plotted relative to the “open-cycle” reference
scenario (Figure 2), three main categories of scenarios can be distinguished.
• The first includes the scenarios with significant quantities of residual plutonium (open and
once-through cycles).
• The second, with a potential radiotoxic inventory 4 to 10 times lower, comprises the multiple
plutonium recycling scenarios (MIX-Pu, FNR-Pu, and PWR/FNR-Pu).
• The third includes the scenarios in which both plutonium and the minor actinides are
recycled (MIX-Pu/MA, FNR-Pu/MA, and PWR/FNR-Pu/MA), resulting in a potential
radiotoxic inventory some 100 times lower than in the reference scenario.
Note: From the standpoint of the potential radiotoxic inventory, there are no differences between
the basic scenarios and the variants involving a separation and conditioning strategy.
241

Figure 2. Reduction of potential raw radiotoxic inventory
for each scenario compared with open cycle
4.2 Potential radiotoxic inventory after conditioning
Allowing for the containment capacity of the conditioning matrix, the logical outcome of the
processes implemented in spent fuel reprocessing, provides a new basis for comparing the possible
scenarios.
The matrix containment performance has a decisive influence on the final impact of a given
scenario or group of scenarios. The Pu recycling scenarios provide a gain by a factor of about 50 over
the open cycle in terms of the potential radiotoxic inventory after conditioning; the scenarios in which
both plutonium and minor actinides are recycled result in a gain by a factor of about 300.
It is interesting to note that specific conditioning of the minor actinides in high-performance
containment matrices in a scenario in which Pu alone is recycled would be as effective as transmuting
the minor actinides. Moreover, the effectiveness of the scenarios in which the minor actinides are
recycled as once-through targets would be no better, under the hypothetical conditions of this study,
than recycling plutonium alone, as the glass matrix provides better containment than the
unreprocessed targets.
242

Figure 3. Reduction of potential radiotoxic inventory after
conditioning for of each scenario compared with open cycle
5. Conclusion
The notion of the radiotoxic inventory after conditioning, by taking into account the respective
containment properties of each ultimate wasteform, provides a means for distinguishing three
categories of fuel cycle management routes according to the potential release of the radiotoxic
inventory.
Fuel cycle management strategies in which plutonium is recycled partially or not at all yield the
poorest performance; multiple plutonium recycling strategies are about 50 times more effective in this
respect, and multiple recycling of plutonium and the minor actinides is even more effective (some
300 times more than the open cycle).
Enhanced reprocessing together with the use of dedicated matrices having a containment
capacity 100 times better than glass taken into account in this study would result in performance
factors equivalent to those of an enhanced reprocessing/transmutation cycle without requiring the use
of burner reactors.
243

REFERENCES
[1] Stratégie et programme des recherches au titre de la loi du 30 décembre 1991 relative à la
gestion des déchets radioactifs à haute activité et à vie longue: 1999-2006, (April 1999).
[2] J.P. Grouiller, J.L. Guillet, H. Boussier, J.L. Girotto, Nuclear Materials Recycling in
Conventional or Advanced Reactors: a Scenario Study, International Conference on Future
Nuclear Systems; Global’99, Jackson Hole, Wyoming, (Aug 29–Sept 3, 1999).
[3] R. Mills, A. Tobias, J. Blachot, JEF-2.2 Radioactive Decay Data, JEFF Report 13. OECD
Nuclear Energy Agency, Paris, France (August 1994).
[4] Age-dependent Doses to Members of the Public from Intake of Radionuclides. Part 5:
Compilation of Ingestion and Inhalation Dose Coefficients, Annals of the ICRP: ICRP
Publication 72, Vol. 26, No. 1, ISSN 0146-6453, Pergamon Press, Elsevier Science Ltd. (1996).
244

DISPOSAL OF PARTITIONING-TRANSMUTATION WASTES
WITH SEPARATE MANAGEMENT OF HIGH-HEAT RADIONUCLIDES
Charles W. Forsberg
Oak Ridge National Laboratory
P.O. Box 2008, Oak Ridge, Tennessee 37831-6180, USA
E-mail: forsbergcw@ornl.gov
Abstract
An alternative approach is proposed for disposing of partitioning-transmutation (P&T) wastes to (1)
reduce repository costs and (2) improve repository performance. Radioactive decay heat controls the
size and cost of the repository. It is proposed that P&T wastes be separated into a high-heat
radionuclide (HHR) fraction and a very-low-heat-radionuclide (VLHR) fraction to bypass this
repository design constraint. There are five repository HHRs in spent nuclear fuel: caesium, strontium,
plutonium, americium, and curium. P&T, by destroying the long-lived HHRs (plutonium, americium,
and curium), is an enabling technology for separate low-cost disposal of the remaining HHRs ( 137 Cs
and 90 Sr), which have limited half-lives (T 1/2 = ~30 year), small volumes, and high heat-generation
rates. These characteristics allow the use of lower-cost disposal methods for these HHR wastes.
Eight HHR disposal options are identified and described. With the removal of the HHRs, there are
lower-cost, higher performance methods for disposal of the remaining VLHRs.
245

1. Introduction
Repository design and performance are primarily controlled by radioactive decay heat. Consider
the proposed Yucca Mountain (YM) repository [1] in the United States. It is designed for ~70,000 t of
spent nuclear fuel (SNF) and high-level waste (HLW). If there were no radioactive decay heat, the
entire volume could be placed in a cube, which would be ~30 m on each side. The cost of such a
repository would be very low. However, radioactive waste generates heat. To ensure repository
performance, the repository temperatures are limited. The temperature is limited by packaging the
wastes in ~11,000 long-lived, expensive waste packages (WPs) and dispersing the WPs over 100 km
of tunnels. The repository program will cost several tens of billions of dollars.
From a distance, a schematic of the proposed YM repository (Figure 1) appears as a large planar
structure—like a horizontal underground car radiator. This typical characteristic of geological
repositories is a consequence of the need to limit repository temperatures and dissipate decay heat. If
waste partitioning and transmutation (P&T) is to have a major impact on the repository cost, it must be
by changing how decay heat is managed in a repository.
Figure 1. Schematic of the proposed YM repository
ORNL DWG 2000-272
Mountain
Ridge
N o r t h R a m p
S o u t h R a m p
Emplacement
Block
2. Radioactive decay heat: sources and impacts
There are several temperature limits [2] on the repository: (1) waste-form limit, (2) package limit,
(3) near-field rock limit, and (4) various far-field limits. Each limit is imposed to prevent damage to
one or more barriers to radionuclide migration from the waste form to the accessible environment.
Almost all repository decay heat from SNF (Figure 2) is produced from five elements: caesium ( 137 Cs),
strontium ( 90 Sr), plutonium (Pu: multiple isotopes), americium (Am: multiple isotopes), and curium
(Cm: multiple isotopes). While there are other heat-generating radionuclides, these decay away
quickly. These high-heat radionuclides (HHRs) can be divided into two categories: shorter-lived
HHRs ( 137 Cs and 90 Sr) and long-lived HHRs (Pu, Am, and Cm).
246

Figure 2. Decay heat from SNF
2000
ORNL DWG 92C-242R
1000
1.21X
1.31X
THERMAL POWER (W /MTIHM)
100
Sr + Cs
SPENT FUEL
LESS Sr
AND Cs
ACTINIDES IN
SPENT FUEL
1.42X
ACTINIDE-FREE
HIGH-LEVEL
WASTE
TOTAL SPENT FUEL
3X
129X
BASIS:
PWR FUEL
235
3.2 wt % U
33 GWd/MTIHM
1.0 MTIHM
10
10 100 1000
DECAY TIME AFTER DISCHARGE (years)
The temperature limits in and near the WP are controlled by decay heat from the shorter-lived
HHRs – 90 Sr and 137 Cs. The long-term temperature limits far from the WP are usually controlled by the
longer-lived actinides. It takes a significant amount of decay heat over a long time to heat large
quantities of rock to unacceptable temperatures. The removal of either the shorter-lived or longer-lived
HHRs radionuclides from the waste provides some benefits to the repository, but the benefits are
limited because both sets of radionuclides impose temperature limits on the repository-one set in the
near term and the second set in the longer term.
If the HHRs are removed from the waste, alternative repository design options [3] exist that
significantly reduce the size and cost of a repository. A large repository is replaced with a
mini-repository, and the size of the mini-repository is controlled by the fraction of the HHRs that are
not removed. For this to occur, alternative methods for management of the HHRs are required.
• Long-lived HHRs. This conference is examining P&T of actinides, including Pu, Am, and
Cm. If the P&T technology is successful and economically viable, this approach can be used
to destroy these troublesome HHRs.
• Shorter-lived HHRs. The 90 Sr and 137 Cs must be separately managed. Because the
characteristics of shorter-lived HHR wastes are different than those of HLW and SNF,
low-cost disposal methods may be available. These HHRs differ from SNF and HLW in four
ways (characteristics):
−
Half-life. The short half-life (T 1/2 ≈ 30 years) allows the use of options that are safe for
disposal of these materials but that would be difficult to demonstrate as safe over
geological times if disposing of SNF or HLW with their large inventories of long-lived
radionuclides.
247

−
−
−
Waste volume. The quantities of HHRs (caesium and strontium) are small. One tonne of
40 000-MWd light-water reactor (LWR) SNF contains 4.1 kg of caesium and strontium.
Heat-generation rate. High heat-generation rates create options that require decay heat to
function.
Fissile content. These wastes include no fissile materials, and thus there are no safeguards
or criticality concerns.
There is experience [4] in separating and packaging shorter-lived HHRs from HLW. Over 10 8 Ci
of these HHRs were separated from defence HLW at Hanford, Washington, to minimise the cost of
storing HLW in tanks. The decay heat, not the tank volume, limited tank capacity. The HHRs were
packaged in 6.67-cm-diam capsules.
3. Management of shorter-lived HHRs
There are many methods to manage 90 Sr and 137 Cs. The method selected by any nation will
depend upon institutional factors and the geology available to each nation to manage such wastes.
Near term and more speculative options are described herein to emphasise that when the
characteristics of the waste change, the disposal options change. This is an area of waste management
where very few investigations have been conducted; thus, many of these options are not well
understood. In parenthesis are the characteristics of the short-lived HHRs that are important for the
disposal option.
3.1 Long-term storage (half-life)
The HHR wastes can be placed in long-term, dry-storage facilities, which are similar to those
used for HLW and SNF. After the decay of most of the HHRs, the wastes can be disposed of in the
repository.
3.2 Extended dry repository (waste volume, heat-generation rate, half-life)
The HHR capsules could be disposed of in a separate section of a dry repository above the water
table [5]. The proposed YM repository in the United States is of this type, and thus this is a potential
option for the United States. Long boreholes would be drilled into the rock from a central tunnel and
then filled with small-diameter HHR capsules. The heat load would be controlled by placing
low-volume HHR capsules in small-diameter horizontal boreholes (

Figure 3. Shorter-lived HHR repository with boreholes (rather than tunnels)
used to distribute decay heat load radionuclide
ORNL DWG 99C-391R
High-
Temperature
Rock
Access
Drift
Hundreds of
meters
HHR
Capsules
Horizontal
Borehole
The HHR section of the repository would be designed as an “extended-dry” repository in
unsaturated rock. By placing the boreholes closely together to obtain higher local heat loads and
higher local temperatures (but sufficiently apart to avoid HHR capsule damage), the local rock
temperature would be above the boiling point of water for thousands of years. If the rock temperature
is above the boiling point of water, there can be no groundwater flow near the capsules and no
migration of radionuclides in groundwater. The shorter-lived HHRs decay before the high-heat section
of the repository cools below the boiling point of water. The large heat capacity of the rock maintains
higher temperatures for extended periods of time. The need for high temperatures requires closer
spacing of the HHRs than is used for SNF and HLW; and thus a correspondingly smaller repository
section for these wastes.
The YM repository project investigated SNF extended-dry repository concepts [6] because of
economic advantages. Such concepts have not been adopted for SNF or HLW because of the
uncertainties in predicting long-term, extended-dry repository behaviour after the repository cools.
These uncertainties do not exist for HHRs that decay before the high-heat section of the repository
cools down.
3.3 Conventional repository (half-life, waste volume)
The HHR wastes could be disposed of in a conventional repository. However, the repository size
and cost may be significantly reduced. Boreholes, not tunnels, are needed for HHR placement.
Elimination of long-term heat-decay loads and most long-term performance requirements
(>1 000 years) allows the use of simpler WPs and other simplifications.
3.4 Saltdiver (heat-generation rate, half-life)
Natural salt domes contain relatively pure salt in the shape of a mushroom with diameters
measured in kilometers. The vertical dimension may be as large as 10 000 m. The saltdiver repository [3]
uses the high-heat generation rates of HHR capsules to allow disposal at depths up to 10 000 m
249

underground in salt domes. The HHRs are packaged into moderately large containers (saltdivers) that
are placed in a salt dome. The high-density, high-temperature WPs sink by heating the salt under the
WP until the salt becomes plastic or melts (at 800°C).
3.5 Rock melt (heat-generation rate, half-life)
In the melt-rock repository [7], a large, spherical, underground cavity would be constructed
several hundred to several thousand meters underground. Large quantities of HHRs would be placed
in the cavity. During loading operations, active cooling systems control temperatures. After the cavity
is loaded, the cavity would be sealed, and the cooling systems would be shut off. The HHRs would
melt and then melt the surrounding rock. The radionuclides would then be incorporated into the
molten rock. It is large-scale vitrification of waste. Ultimately, as the decay-heat levels decrease, the
molten rock would solidify into solid rock.
During periods of high-temperature operations, the high temperatures cause plastic deformation
of the rock beyond the melt zone, which seals all cracks. Several uncertainties [8] have been identified
with this disposal option. However, the identified uncertainties apply only to HLW and SNF that
contain long-lived radionuclides, not disposal of shorter-lived HHRs. Further analysis would be
required to determine if there are unidentified failure modes when disposing of HHRs.
3.6 Borehole (waste volume)
The use of deep vertical boreholes (>5 km) has been considered for the disposal of various
radioactive wastes. However, a major drawback is that a borehole has very limited volume. Drilling
deep, wide boreholes is expensive. For short-lived HHRs, the volumes are very small; thus, this may
be a viable low- cost option for these specific wastes.
3.7 Seabed (waste volume, half-life)
International programs [9] have investigated seabed disposal of SNF and HLW. Seabed disposal
involves placing WPs into the clay layer, which covers most of the ocean’s seabed. The clay layer has
potentially excellent waste-isolation properties, and the ocean provides an independent backup
mechanism (ocean dilution) if there were failures. There are major institutional problems and some
technical problems associated with this option.
Demonstration of disposal viability of short-lived HHRs would be simpler than for other wastes
because the shorter-lived HHRs remain hazardous for a much shorter period of time. Furthermore,
there are no fissile materials associated with HHRs. Recent analysis has raised questions about the
viability of disposal of wastes with fissile materials using this technology. New off-shore oil recovery
technologies are making it increasingly easier to recover objects from the ocean seabed; thus, there is a
concern about the recovery of any fissile materials by unknown parties if the disposal site is the ocean
seabed.
3.8 Shallow-land disposal, half-life
The limited half-life may allow shallow-land disposal of the shorter-lived HHRs under some
circumstances [3].
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4. Management of low-heat, long-lived radionuclides
With most of the HHRs removed, the repository for the remaining wastes becomes a small
facility [3,5]. The required repository would contain two sections: a section for wastes with significant
decay heat and a section for the very-low-heat radionuclides (VLHR). Existing vitrified HLW and
some P&T target wastes (deep-burn, once-through targets; certain target-processing wastes) would be
disposed of in a repository section similar to existing repository designs. Because of the small
quantities of these wastes, this repository section would be relatively small.
The wastes from processing light-water reactor SNF, after removal of the shorter-lived ( 90 Sr and
137 Cs) and long-lived (Pu, Am, and Cm) HHRs would be a VLHR waste. These wastes may be
disposed of in a few lower-cost, high-performance silos without exceeding temperature limits.
Depending upon the geology and efficiency of removal of HHRs, such a silo might accept the wastes
from up to 10 000 tonnes of SNF.
There is experience with waste silos [10]. Sweden (Figure 4) and Finland have constructed and
are operating underground silos for the disposal of intermediate-activity wastes. The heat-generating
characteristics of these wastes are somewhat similar to VLHR wastes. The Swedish waste silos are
about 50 m high and 25 m in diameter. The costs per unit volume are a fraction of the cost of
traditional WPs.
Figure 4. Swedish SFR silo for intermediate wastes
ORNL DWG 89C-1100
Silo Crane
Greater Than 50 M
To Surface
Transporter
Waste
Shipping
Container
Clay
Barrier
Transported
Waste Package
Rock
Cavern
Silo
(Final Waste
Package)
VLHR silos would be located in the middle of the repository at full repository depth to take
advantage of the waste-isolation capabilities of the repository. The repository provides a major barrier
against human intrusion, and the geology provides several barriers against radionuclide releases to the
accessible environment. Silos are an alternative WP, not a replacement for the repository.
The replacement of WPs with large silos may result in significant improvements in the
performance of the engineered barriers to radionuclide releases. The release of radionuclides from a
failed WP is proportional to (1) the groundwater flow through the WP and the (2) solubility limits of
the radionuclides in groundwater. By concentrating the VLHR wastes from up to 10 000 t of SNF in
1 silo rather than spreading it over ~1 000 WPs, the groundwater flow through the wastes per unit
volume is reduced by a factor of 100 to 1 000. With the reduction of groundwater flow per unit
251

quantity of waste, radionuclide releases are proportionally reduced. The large waste silo has a smaller
surface-to-volume ratio than does each WPs.
5. Other considerations
5.1 Scaling factors
No detailed economic analysis of these repository benefits has been conducted. However, some
comparisons [3] between conventional repositories and these alternative designs can be made.
Consider the case where (1) P&T destroys the long-lived HHRs, (2) the shorter-lived HHRs are
disposed of in an extended-dry repository such as YM, and (3) the long-lived, low-heat wastes are
disposed of in silos.
In the conventional YM repository design, the SNF is disposed of in large WPs in 5.5-m-diam
tunnels. For every 100 m of tunnel required for disposal of SNF, about 71 m of boreholes would be
required to dispose of the shorter-lived HHR wastes from the SNF. For every 100 SNF WPs, 71 HHR
capsules of similar length and a small fraction of a silo would be required for disposal of the
HHR-VLHR wastes from that SNF. In effect, there are three major changes: (1) substitution of 5.5-m
disposal tunnels with 15-cm boreholes for the HHRs, (2) substitution of thousands of high-performance,
expensive WPs with a few silos, and (3) reduced heat load from destruction of the longer-lived HHRs.
The impact of these changes would be to significantly reduce the operational costs for the
repository. Operational costs include the mining of disposal drifts for the WPs. It may not impact
siting or licensing costs – an important fraction of the total costs. The economic incentives are
dependent upon the size of the repository. As the repository capacity increases and the cost per unit of
waste decreases, operational costs become a larger fraction of disposal costs. Siting and licensing costs
are essentially fixed costs.
The economic cost for the repository gains in an actinide P&T fuel cycle is the necessity to
separate the caesium and strontium from the other waste streams. This cost is dependent upon the
specific separation processes.
5.2 Caesium-135
The short-lived HHRs contain one long-lived radionuclide, 135 Cs. It has a half-life of
3 × 106 years. Performance assessments of proposed repositories [11] indicate that this long-lived
radionuclide is not usually a significant risk to man nor a significant factor in terms of repository
performance. There are several reasons for this:
• Geochemistry. Radionuclides, such as 129 I, 237 Np, and 99 Tc, which dominate the long-term
risks from a repository are those most easily transported by groundwater with little retention
by the geology. There is significant retention of caesium in most types of rock and
ion-exchange of radioactive caesium isotopes with non-radioactive caesium in the rock.
• Biological effects. Differences in the accumulation rate of different radionuclides in specific
human organs determines their relative hazards. The hazard from 135 Cs is low compared to
many other radionuclides because of its low rate of bioaccumulation.
For any HHR disposal option, a performance assessment of the risks from this radionuclide will
be required. There are major engineering questions about the feasibility of isotopically separating this
252

isotope from other caesium isotopes; thus, disposal with the other caesium isotopes is likely to be the
most practical route. Such an assessment would be significantly simpler to make than for HLW or
SNF because there is only a single radionuclide.
6. Conclusions
Repository designs and costs are controlled by radioactive decay heat. Any P&T option that
destroys long-lived HHRs (Pu, Am, and Cm) is an enabling technology that may allow for lower-cost,
higher performance repositories by separate management of (1) the shorter-lived HHRs and (2) the
VLHR wastes. These repository benefits may exceed the other waste management benefits of actinide
P&T fuel cycles such as reductions in radiotoxicity. The cost for these benefits is the requirement to
separate caesium and strontium from the other P&T wastes.
An understanding of the costs and benefits of separate management of shorter-lived HHRs should
be a high priority within any investigation of actinide P&T fuel cycles. This is an appropriate area for
international co-operation. Most of the issues (selection of radionuclides to be destroyed by P&T,
transmutation efficiencies, solidification of short-lived HHRs, repository design for low-heat,
long-lived radionuclides, etc.) are common issues for all. It is an area of waste management where
only very limited studies have been undertaken.
REFERENCES
[1] US Department of Energy, Draft Environmental Impact Statement for a Geological Repository
for the Disposal of Spent Nuclear Fuel and High-level Radioactive Waste at Yucca Mountain,
Nye County, Nevada, DOE/EIS-0250D, July 1999, US Department of Energy, Washington, D.C.
[2] A.G. Croff, A Concept for Increasing the Effective Capacity of a Unit Area of a Geologic
Repository, Radioactive Waste Management and Environmental Protection, 1994, 18, 155–180.
[3] C.W. Forsberg, Rethinking High-level Waste Disposal: Separate Disposal of High-heat
Radionuclides ( 90 Sr and 137 Cs), Nuclear Technology, August 2000, 131, 252–268.
[4] R.R. Jackson, Hanford Waste Encapsulation: Strontium and Caesium, Nuclear Technology,
January 1977, 32(1), 10.
[5] C.W. Forsberg, Disposal of Partitioning-transmutation Wastes in a Yucca-Mountain-Type
Repository With Separate Management of High-heat Radionuclides ( 90 Sr and 137 Cs), 4th Topical
Mtg on Nuclear Applications of Accelerator Technology, Embedded Topical Mtg American
Nuclear Society, 2000 Winter Meeting, November 2000 (American Nuclear Society).
[6] T.A. Buscheck and J.J. Nitao, Repository-heat Driven Hydrothermal Flow at Yucca Mountain,
Part I: Modelling and Analysis, Nuclear Technology, December 1993, 104(3), 418.
253

[7] J.J. Cohen, A.E. Lewis and R.L. Braun, In-situ Incorporation of Nuclear Waste in Deep Molten
Silicate Rock, Nuclear Technology, April 1972, 13, 76.
[8] A.S. Kubo, Technical Assessment of High-level Nuclear Waste Management, Ph.D. degree thesis,
May 1973, Department of Nuclear Engineering, Massachusetts Institute of Technology, Cambridge.
[9] NEA (1988), Feasibility of Disposal of High-level Radioactive Wastes Into the Seabed, OECD
Nuclear Energy Agency, Paris, France.
[10] J. Carlsson, Nuclear Waste Management in Sweden, Radwaste Magazine, 1998, 5(6), 25.
[11] TRW Environmental Safety Systems, Inc., Total System Performance Assessment–viability
Assessment (TSPA-VA) Analysis Technical Basis Document, B00000000-01717-4301-00001
REV 01, November 13, 1998, Las Vegas, Nevada, USA.
254

THE AMSTER CONCEPT
J. Vergnes 1 , D. Lecarpentier 2 , C. Garzenne 1
1 EdF-R&D, 1 Avenue du Général de Gaulle, 92140 Clamart, France
2 Ministère de l’Éducation nationale, de la recherche et de la technologie,
Conservatoire National des Arts et Métiers, Laboratoire de physique,
292 rue Saint-Martin, 75141 Paris Cedex, France
H. Mouney
EdF, Pôle Industrie, Division Service et Ingéniérie
CAP AMPERE, 1 Place Pleyel, 93282 Saint Denis Cedex, France
G. Ritter, M. Valade,
CEA/Cadarache, Bât. 320, 13108 St Paul Lez Durance, France
A. Nuttin, D. Heuer, O. Meplan, J.M. Loiseaux
Institut des Sciences Nucléaires, (UJF-IN2P3-CNRS)
53 rue des Martyrs, 38026 Grenoble, France
Abstract
AMSTER is a concept for a graphite-moderated molten salt reactor, in which the salt treatment
installation has been redesigned in order to reduce waste production. Using this concept, one can
define a large number of configurations according to the products loaded and recycled. This document
presents a configuration which self-consumes transuranium elements and generates fissile material
with a mixed thorium and uranium support. This gives a reactor, which is highly economical in
uranium and thorium consumption, leaving only a few grams of transuranium elements per billion
KWhe in the ultimate wastes to be disposed of.
255

1. Introduction
On 30 December 1991 the French parliament adopted a law concerning research into the
management of radioactive waste [1]. This law stipulated that three types of work should be
conducted simultaneously:
• A search for solutions allowing the separation and transmutation of long-life radioactive
elements present in this waste.
• A study into the possibility of reversible or irreversible storage in deep geological
formations, in particular by creating underground laboratories.
• A study into long-term packaging and storage conditions for this waste on the surface.
The studies presented in this article aim to provide a foundation for work in the first area of
research defined by this law.
Analysis of the problem of reducing long-life radioactive products leads us to propose a new
concept: AMSTER (Actinides Molten Salt TransmutER). We will present this concept and the results
of a preliminary study into the reactor physics associated with this concept.
2. Description of the AMSTER concept
AMSTER is a continuously reloaded, graphite-moderated molten salt critical reactor, using a
Uranium 238 or Thorium 232 support, slightly enriched with 235 U if necessary.
2.1 General presentation
Critical molten salt reactors were extensively studied in the 60s and 70s. Research was carried
out in the Oak Ridge National Laboratory, where an 8 Mwth prototype was operated, the Molten Salt
Reactor Experiment (MSRE). This experiment was followed by a 1 Gwe project, the Molten Salt
Breeder Reactor (MSBR) [2], on which certain aspects of AMSTER are based. It should also be noted
that the MSBR project was extensively examined in France (at the CEA and at EdF) in the 1970s.
The aim at the time was breeder reactors. Today, this type of reactor is again of interest to the
specialists, owing to its incinerating capacity.
Figure 1 gives the basic layout of this type of reactor.
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Figure 1. Layout diagram of the molten salt reactor
Molten salt
Pump
800°C
He
Combined
cycle turbine
Graphite
550°c
On-line
reprocessing
unit
SALT/SALT exchanger
SALT/He exchanger
2.2 Reactor core
The core of a molten salt reactor consists of an array of graphite hexagons identical to those of
the Saint-Laurent B1 UNGG reactor. Each hexagon contains a hole through which the salt circulates.
We used the results of a study conducted by the EdF reactor physics department in 1976, which
defines a salt hole diameter of 8 cm for a hexagon 13 cm on a side. The diameter of this hole was
optimised to favour reactor conversion (transmuting as much 238 U as possible into 239 Pu).
Figure 2. AMSTER reactor cell
Graphite density
of 2.25
SALT
8 cm
.
13 cm
We used salt of the same type as that of the MSBR project [2]. The composition adopted,
61LiF – 21BeF 2 – 18NLF 4 , enables a moderate quantity of uranium and transuranium nuclei to be
introduced into the core (NL here stand for heavy nuclei).
2.3 Operating principle
When the salt enters the array, it becomes critical and heats up. It enters at a temperature of about
550°-600°C and leaves at a temperature of 800°C.
The core inlet temperature is determined by the salt melting temperature, which itself depends on
the composition of the salt (500 to 600°C).
The outlet temperature is determined by the strength of the materials other than graphite
(hastelloy).
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Once heated, the salt is entrained by pumps and passes through salt/salt exchangers which enable
the thermal energy produced to be recovered. On leaving the core, an on-line reprocessing unit
which we have entirely redesigned, takes a small fraction of the fuel for reprocessing, in other
words it extracts the fission products from it. This reprocessing is accompanied by injection into
the salt of new nuclei, 235 U, 232 Th, 238 U or transuranium elements, to replace the heavy nuclei
already fissioned.
The secondary salt heats up either steam or helium, which feeds either a combined cycle turbine
plus alternator, or thermal applications such as the production of hydrogen if the salt temperature is
high enough.
2.4 Configurations examined
A large number of configurations can be envisaged with this type of reactor, depending on:
• The products placed in the reactor: isotopes of uranium, transuranium elements, isotopes of
Thorium, long-life fission products (LLFP).
• Substances multi-recycled in the reactor.
• The substances leaving the reactor:
−
−
Vitrified losses from reprocessing of transuranium elements, uranium, thorium, LLFP
and all the short-life fission products (SLFP) considered being waste.
Depleted uranium when the 235 U currently stored is used for enrichment.
The various configurations are characterised by:
• The type of the support: uranium, thorium, mixture of uranium and thorium. The thorium
support has the advantage of producing significantly fewer transuranium elements than the
uranium support and of having a thermal spectrum regenerating more fissile material than the
uranium support.
• The presence of fertile blankets allowing regeneration of the fissile nuclei.
• The possible input of transuranium elements coming from other reactors, determining the
incinerating capacity of the reactor. If no transuranium elements are input from the outside,
the reactor consumes its own transuranium elements. The reactor loaded with outside
transuranic become an incinerator.
We thus examined 7 configurations. Table 1 describes the various configurations examined.
There are 4 configurations without fertile blanket and 3 configurations with fertile blanket.
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Table 1. Definition of configurations examined
Without fertile blanket
Support Uranium Thorium Uranium + thorium
Self consuming X X
Incinerating X X
With fertile blanket
Support Uranium Thorium Uranium + thorium
Self consuming X X X
Incinerating
Below, we will present the transuranium elements consumption configuration, with fertile
blanket and a mixed uranium-thorium support.
Processing unit: for each configuration examined, more or less complex salt reprocessing is
required.
2.5 General recycling principle
Below, we present the recycling principle for the uranium support reactor. For the thorium
support, the principle is the same, with the thorium separated after the transuranium elements and
before the FP.
Figure 3. Layout diagram of the AMSTER salt processing unit
Vitrification
Used salt
Partitioning
U
UF6
Natural uranium F6
Partitioning
TRU/PF
Fission products
Centrifugation
Depleted U + 236 U
TRU
Enriched U
Fresh salt
More specifically, the transuranium elements are confined in the core – reprocessing unit
assembly by separate extraction of the transuranium elements and rare earths by a Bismuth counterflow.
259

The salt processing unit includes the cycle front-end (salt enrichment) and back-end
(FP extraction).
• Cycle front-end
Uranium is first of all extracted from the salt (Figure 3). It is in UF 6 form, which is mixed
with an adjusted mass of natural uranium (also in UF 6 form). This mixture is enriched with
235 U if necessary, to the required new salt 235 U enrichment value, for example using an
ultra-centrifuge.
The residual depleted uranium is evacuated, taking with it a large proportion of the 236 U in
the spent salt (about 35 %) [3]. This prevents the core being poisoned with this isotope.
This solution would require a small number of centrifuges, owing to the small quantity of
235 U to be added.
• Cycle back-end
The transuranium elements are separated in a salt – liquid metal exchanger. Given the good
separation factor in this operation (about 10), and by using 6 consecutive stages, the salt
would only contain a residue of about 10 -5 times the initial mass of transuranium elements.
Then the thorium and the fission products, except for the LLFP to be incinerated, are
extracted from the salt with no need for a high separation capacity. The residual
transuranium elements in the salt are extracted with the FP, which can be vitrified and stored
in the same way as fission products today.
3. Numerical simulation principle
Numerical simulation of this type of reactor requires an iterative working method, the principle
of which is described below.
We begin with an APOLLO 1 type calculation of a cell evolving in an infinite medium. We
chose APOLLO 1, as this code requires little calculation time and has already been used to simulate
and manage Gas Graphite Natural Uranium Reactors (UNGG). We checked the accuracy of the
calculations against a reference calculation using the Monte Carlo TRIPOLI 4 code (discrepancy of
400 pcm for a k ∞ of 1.2).
At the end of this calculation step, we extract the FP from a fraction of the core. They are
replaced by a mixture of TRU, 235 U and 238 U so as to guarantee the reactivity at the end of the step.
To limit the calculation time, the time between two reprocessing operations must be sufficiently
long. We achieved initial equilibrium with a pitch of 10 days. Then, in the light of the first results
(slight evolution of k ∞ ) we raised this pitch to 100 days.
We adopted a calculation pitch of 100 days and a cell k ∞ at the end of evolution of 1.05 (to take
account of leaks), an electrical power produced by AMSTER of 1 GWe or 2 250 MWth, a salt volume
of 48 m 3 (30 in the active core and 18 in the auxiliaries) for a reactor without blanket.
We defined a reference case in which we extract one third of the core every 100 days (burn-up of
300 efpd). This initial simulation showed that it was necessary to purge all or part of the 236 U formed
by 235 U capture.
260

We therefore adopted a partial purge (30%) of the 236 U for the rest of the study.
The transuranium elements, enriched uranium, thorium and possibly other transuranium
elements, would be re-injected into the salt and then into the reactor.
4. Feasibility of an AMSTER self-generating uranium concept
4.1 The concept
In the case of the thorium support, the consumption of fissile uranium is low enough for it to be
produced in the form of 233 U, in an extra core zone. This fertile zone located on the core periphery,
would be under-moderated by increasing the diameter of the salt hole.
In this concept, the size of the fertile zone would be adapted to make the reactor only just a 233 U
self-generator (the production of the fertile zone would exactly compensate the consumption of the
fissile zone).
Salt processing would simply be by removal of the FP and replacement by the same mass of 232 Th
or (and) 238 U support.
4.2 Calculation method
Only an equilibrium situation using a complete core calculation would be able to validate this
concept completely. However, its feasibility can be evaluated by cells calculations.
In the fertile zone, the radius of the cylinder in which the salt circulates is set at 8 cm. This value
will then be optimised when the core calculation verification is made.
We also supposed that the power density in the fertile zone was half that in the fissile zone. An
exact determination of this ratio is only possible with a core calculation.
We calculate the volume of fissile and fertile salt to obtain the power level sought.
In these conditions, the mean power density of the fertile salt is equal to one-eighth the mean
power density of the fissile salt. So that heating of the fertile salt is the same as that of the fissile salt,
we slow down the salt flowrate in the blanket by a factor of 8 and thus multiply the time the salt
spends in the blanket by 8. The production of heavy nuclei in the blanket is thus equivalent to the
production from a volume of salt during the time the salt passes through the fissile core.
Given these hypotheses, the reactor was thus balanced for a given ratio of fertile and fissile salt
volumes. At each time step ∆t:
• The fuel was placed in each zone for ∆t, and the 2 zones were mixed pro rata the core
volumes.
• Part of the fission products is removed and is replaced by the same mass of a mixture of 233 U
and 232 Th.
• The enrichment of this mixture is calculated to keep the fissile core k ∞ at 1.05.
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The volume of salt in the fertile area is that for which the mass of 233 U to be added at equilibrium
is zero.
Equilibrium was achieved for several ratios of salt volume in fertile zone to salt volume in fissile
zone. For reasons of precision, we took a time pitch of 10 efpd
4.3 The thorium support
In the fertile zone the 233 U balance is positive: More 233 U comes out of the blanket than goes in. If
we increase the blanket volume, we increase the quantity of 233 U produced and at any given moment,
the 233 U contribution of the blankets will compensate the 233 U consumption by the fissile core. The
core will generate its own uranium. To determine the volume of the blankets needed, we varied the
ratio between the fertile salt volume and the fissile salt volume between 1.5 and 2.5.
For each of these ratios, we give (Table 2) the consumption of 233 U, the masses in the reactor at
equilibrium, the increase in core size (volume and radius) necessary, and the proportion of power
given off in the fissile zone.
These calculations show the feasibility of the concept. This configuration is particularly
interesting because:
• It does away with the uranium enrichment phase.
• It requires no extraction of 236 U.
• It offers a very small transuranium elements inventory.
• It reduces consumption of heavy nuclei to 100 kg of 232 Th per TWhe, thus reducing mining
waste accordingly.
• It considerably reduces the masses of depleted and reprocessed uranium.
4.4 The thorium-uranium support
It is also possible to add uranium (i.e. depleted U in stock) to the thorium support. We thus
increase the quantity of transuranium elements at equilibrium, while remaining within a reasonable
range. But we then burn 238 U and make 232 Th savings. We significantly reduce the fissile nuclei
enrichment of the uranium in the core, thus making the reactor non-proliferating.
As a counterpart, we will have to increase the production of 233 U and thus the volume of the
blankets.
We thus varied the proportion of 238 U added to the load from 0 to 100% for 2 volumes of fertile
salts: 2 times the fissile volume and 2.5 times the fissile volume.
Figure 4 shows the mass of 233 U consumed (or produced) versus the percentage of uranium 238
in the fertile material added;
The more uranium is placed in the support, the more 233 U must be added and the more the volume
of the blanket must be increased. Between 0 and 50 % 238 U the 233 U input is small enough to be
262

conceivable. Thus by increasing the relative volume of fertile salt to 2.5, self-generation is achieved
for 50% 238 U.
Table 2. Characteristics of the two-zone AMSTER, depending on the ratio of salt volume in
fertile core to the salt volume in fissile core with a 100% thorium support
Fertile salt volume to fissile salt volume ratio 1.5 2 2.5
Core inventories (kg/GWe)
Th 138 800 153 580 167 360
Uranium 3 354 3 650 3 900
Transuranium elements 53 54 54
233 U consumption (kg/Twhe) 2.11 -0.12 +1.91
Core radius (m) 4.95 5.04 5.11
Total volume of salt in the reactor (m 3 ) 82 91 99
Power produced in the fissile area (MWe) 1 895 1 800 1 714
The presence of 238 U increases the mass of transuranium elements at equilibrium. This rises
almost linearly with the percentage of 238 U and is equal to 1 500 kg for a percentage of 50%.
The presence of 238 U also modifies the isotopic composition of the transuranium elements.
Figure 5 shows the different masses in the reactor for 0%, 50% and 100 % 238 U in the fertile material
inputs.
Figure 4. Mass of 233 U consumed (or produced) versus
the percentage of 238 U in the fertile material input
18
16
14
233 U top-up (Kg/TWhe)
12
10
8
6
4
Vfertile/Vfissile=2
2
Vfertile/Vfissile=2.5
0
0
-2
20 40 60 80 100
-4
% of uranium in the inputs
263

Figure 5. Core mass balance for 0.50% and 100% 238 U
in the fertile material input for Vfertile/Vfissile = 2
1 000 000
Loaded mass in reactor (kg)
100 000
10 000
1 000
100
10
1
TH
32
PA
33
U2
33
U2
34
U2
35
U2
36
U2
38
PU
38
PU
39
PU
40
PU
41
PU
42
AM
41
AM
43
100%Th 50% Th 50% U 100% U
NP
37
CM
42
CM
43
CM
44
CM
45
5. A major safety asset: core drainage
The salt is at its maximum reactivity in the graphite. A safe fallback position can thus be
obtained by draining the core. As the fuel is liquid, it can be extracted from the core at any moment.
For this, we adopted a concept proposed by EdF and the CEA, which consists in placing a drain tank
under the core, which is permanently connected to it. Salt is confined within the core by a helium
back-pressure (Figure 6). One therefore need simply interrupt the electric power supply to the He
compressor for gravity drainage of the core. This feature, allied with the considerable thermal inertia
of the reactor and the difficulty of rapidly inserting reactivity, should make the reactor particularly
safe.
Figure 6. Core drainage principle
Thermal detector
Reactor
Salt
Compressor
He
264

6. R&D needed to validate the AMSTER concept
To make the transition from concept to technology, much research and experimentation is
required.
Among the subjects covered will be:
• Salt chemistry, structural corrosion by salt.
• Processing chemistry, containment of transuranium elements in the salt.
• Reactor dynamics.
• Safety.
It should be recalled that major R&D work was already carried out in the 60s/70s and a prototype
functioned remarkably well (MSRE). A detailed preliminary project for a breeder reactor (MSBR)
was conducted. The AMSTER concept is similar to these two reactors and the experience acquired
would be directly applicable to it.
7. Fertile and fissile material utilisation
The AMSTER concept should allow the production of energy with very reduced quantities of
transuranium element waste (a few g per TWhe), and with no transportation of highly radiotoxic
substances.
AMSTER with a mixed thorium-uranium support with a peripheral fertile zone, breeding its own
uranium, should further improve this performance. This reactor should in fact offer incinerating
performance identical to that of the thorium AMSTER supplied with 235 U, while eliminating the need
for uranium enrichment and making the reactor non-proliferating. Furthermore, this self-breeding
reactor would consume only 50 kg of thorium and 50 kg of 238 U per TWhe, which in the light of
estimated resources (1 to 4 million tonnes of thorium and 1 to 3 million tonnes of uranium), would
allow the production of 20 to 70 million TWhe (the world’s annual electricity consumption from all
sources together is about 15 000 TWhe); The depleted uranium stored in France (200 000 t), together
with 200 000 t of thorium, could produce 4 million TWhe (annual production is 400 TWhe).
Figure 7 compares the mass balance entering and leaving the reactor for a standard open-cycle
PWR and a self-generating AMSTER with a support of 50% U and 50% Th.
The inputs are natural or depleted uranium and thorium.
The outputs comprise:
• Transuranium element losses, U and Th vitrified with the FP.
• Depleted uranium produced at enrichment of the support.
This theoretical study shows that:
• AMSTER would be far “cleaner” than a PWR (4 decades reduction of transuranium element
waste).
265

• As we saw earlier, it offers “virtually inexhaustible” resources.
• It should be safer, owing to the fact that the fuel can be rapidly extracted from the core if
necessary.
Figure 7. Mass balance entering and leaving the reactor for a standard open-cycle PWR
and for a self-generating AMSTER with a support of 50% U and 50% thorium
Depleted U
OUT
Th
FP
TRU
Vitrified
IN
Thorium
Natural U
1
g
10
100
1 000
kg
10 000
100 000
1 000 000
t
10 000 000
100 000 000
REP One Through
AMSTER Self-generating ThU
8. Initial conclusions
Although new, molten salt reactor technology was already experimented at Oak Ridge in the
1960s. The prototype built at the time operated remarkably well. Furthermore, EdF and the CEA
studied molten salt reactors until 1983. This preliminary study should be followed by a minimal R&D
and engineering program in order to evaluate both the pyrochemical reprocessing process (loss rates
to be confirmed) and the technological feasibility and safety of AMSTER.
The reactor specifications would then have to be optimised: unit power, size of graphite array,
salt burn-up, quantity of transuranium elements in the reactor, etc., and its economics evaluated.
REFERENCES
[1] Journal officiel: “Loi n° 91-1381, du 30 décembre 1991, relative aux recherches sur la gestion
des déchets radioactifs”.
[2] Conceptual Design Study of a Single Fluid Molten Salt Breeder Reactor, Oak Ridge National
Laboratory Report, ORNL- 4541. June 1971.
[3] J. Vergnes et al., The AMSTER Concept (Actinide Molten Salt TransmutER), Physor 2000
Conference, Pittsburgh, USA, May 2000.
266

SESSION III
PARTITIONING
J.P. Glatz (ITU) – J. Laidler (ANL)
267

SESSION III
PARTITIONING
SUB-SESSION III-A:
AQUEOUS REPROCESSING
269

PARTITIONING-SEPARATION OF METAL IONS USING HETEROCYCLIC LIGANDS
Michael J. Hudson 1 , Michael G.B. Drew 1 , Peter B. Iveson 1 , Charles Madic 2 , Mark L. Russell 1
1 Department of Chemistry, University of Reading
Box 224, Whiteknights, Reading, RG6 6AD, United Kingdom
2 Commissariat à l’Énergie Atomique, B 171, Bagnols-sur-Cèze, 30207 France
Abstract
Some guidelines are proposed for the effective design of heterocylic ligands for partitioning because
there is no doubt that the correct design of a molecular extractant is required for the effective
separation of metal ions such as actinides(III) from lanthanides(III). Heterocyclic ligands with
aromatic ring systems have a rich chemistry, which is only now becoming sufficiently well understood
in relation to the partitioning process. The synthesis, characterisation and structures of some chosen
molecules will be introduced in order to illustrate some important features. For example, the molecule
N-carboxybutyl-2-amino-4,6-di (2-pyridyl)-1,3,5-triazine (BADPTZ), which is an effective solvent
extraction reagent for actinides and lanthanides, has been synthesised, characterised and its interaction
with lanthanide ions studied. The interesting and important features of this molecule will be compared
with those of other heterocyclic molecules such as 2,6-bis(5-butyl-1,2,4-triazol-3-yl) pyridine
(DBTZP), which is a candidate molecule for the commercial separation of actinides and lanthanide
elements.
Primary co-ordination sphere
One of the most critical features concerning whether a molecule is a suitable extraction reagent is the
nature of the binding and co-ordination in the primary coordination sphere of the metal. The resultant
effects for partitioning will be considered briefly for selected heterocylic molecules. It will be shown
how the structural types change as the complete lanthanide series is traversed from lanthanum to
lutetium. For effective solvent extraction, the ligand(s) should be able completely to occupy the
primary co-ordination sphere of the metal ion to be extracted. Interactions in the secondary coordination
sphere are of less importance.
Inter-complex (Hydrogen Bonding) interactions
Another feature that will be briefly considered is the intermolecular binding between ligands when
bound to the metal ion. Thus the intermolecular structures between complex molecules will be
considered where these have relevance to the extraction process. For effective separations, the
intermolecular interactions should be minimised such that there are only weak van der Waals
interactions arising from the hydrophobic exteriors of the complexes.
Implications for partitioning
The effectiveness of the above heterocyclic reagents will be considered in relation to the interactions
in the primary co-ordination sphere of the metal and the intermolecular interactions.
271

1. Introduction
There is no doubt that the correct design of a molecule is required for the effective separation of
metal ions such as actinides(III) from lanthanides using solvent extraction reagents. Recent attention
has been directed towards malonamides for coextraction but heterocyclic ligands with aromatic ring
systems have a rich chemistry, which is only now becoming sufficiently well understood in relation to
the partitioning process for the selective extraction of the actinides [1].
A future goal in nuclear fuel reprocessing may be the conversion or transmutation of the
long-lived radioisotopes of minor actinides, such as americium, into short-lived isotopes by irradiation
with neutrons [1]. In order to achieve this transmutation, it is necessary to separate the trivalent minor
actinides from the trivalent lanthanides by solvent extraction, otherwise the lanthanides absorb
neutrons too effectively and hence limit neutron capture by the transmutable actinides. Solvent
extraction using ligands containing only carbon, hydrogen, nitrogen and oxygen atoms is desirable
because they are completely incinerable and thus the final volume of waste is minimised [2]. Nitric
acid is used in the extraction experiments because it is envisaged that the An(III)/Ln(III) separation
process could take place after the existing PUREX process. For the ensuing discussion the
heterocyclic ligands are considered as free bases but in practice one or more of the nitrogen atoms will
be protonated – this should not influence the overall discussion since co-ordination to metal ion
dominates over protonation particularly in the chelating molecules under consideration.
There is clearly a need to study future processes such as SANEX in which the minor actinides are
selectively separated from the lanthanides. For the reasons outlined below, the solvent extraction
reagents depicted in Figure 1 have been evaluated.
Figure 1. The ligands L 1 , L 2 , L 3 , L 4 and L 5
272

The terpyridyl reagent (ligand L 1 ) has been extensively studied previously [3]. With common
ligands such as L 1 and 2,4,6-tri(2-pyridyl)-1,3,5-triazine (L 2 ), Am(III)/Eu(III) separation factors
between 7 and 12 have been obtained when 2-bromohexanoic acid is used as a synergistic reagent [4].
The alkyl derivatives of tripyridyl–s-triazines (L 2 ) may well be good candidates for future studies but
difficulties of synthesis has limited the availability of these reagents. The design of molecules such as
the 2,6-bis-(5,6-dialkyl-1,2,4-triazin-3-yl)- pyridines (BTPs) (L 5 ) has lead to improved separation
factors [5] without the requirement for reagents such as 2-bromohexanoic acid, which is required with
most other heterocyclic reagents. The evaluation of the structures of complexes formed by L 4 with
some trivalent metal ions has enabled the co-ordination of these ligands with the entire range of
lanthanides to be evaluated. This has meant that the limitations on co-ordination caused by the
lanthanide contraction are now better understood.
We have studied the molecule N-carboxybutyl-2-amino-4,6-di(2-pyridyl)-1,3,5-triazine
(BADPTZ), which is an effective solvent extraction reagent for actinides and lanthanides. This
molecule co-ordinates to the metal ions to L 4 but has improved Am(III)/Eu(III) separation factors of
over 10. Thus it will shortly be possible to correlated precise solvent extraction data with known
metal-ligand interactions.
2. Guidelines
One of the most important features concerning whether or not a molecule is a suitable extraction
reagent is the nature of the binding and co-ordination in the primary and secondary co-ordination
spheres of the metal ion. Thus it is possible to formulate some general guiding suggestions (not rules)
for the purposeful design of ligands:
• The extractant molecules (ligands) should be designed so as to exploit the differences in the
co-ordination chemistries (including ion-pair formation) of the ions to be separated.
• For effective solvent extraction, the ligand molecule(s) should be able completely to occupy
the primary co-ordination sphere of the metal ion to be extracted.
• Interactions in the secondary co-ordination sphere are of less importance even though the
counter-ions are in the secondary co-ordination sphere.
• There should be limited intermolecular interactions between the extracted species – preferably
only van der Waals interactions such that even hydrogen bonding is minimised.
• There should only be a single species that is extracted.
• The chemical bonds within the primary co-ordination sphere should be strong but not so
strong as to prevent subsequent bond breaking for the subsequent stripping process.
• There should be acceptable resistance to radiolysis.
• The likely decomposition products should have minimal interference with the solvent
extraction process.
2.1 Primary co-ordination sphere
The primary co-ordination sphere of the metal ion is defined by the covalent binding between the
metal atom and the immediate ligand atoms and the stereochemical arrangement of the ligands (or
solvent extraction reagents). For lanthanides and minor actinides, metal-ligand binding is considered
273

to be much more ionic than is the case with the trivalent transition metals for which Crystal Field
Effects play an important thermodynamic part. Thus the ligands are much more loosely bound in many
cases – such that the metal-ligand bonds are broken and reformed much more readily than is the case
with the transition metals or with the major actinides prior to Am. The design of polydentate nitrogen
ligands, which complex minor actinides such as Am(III) preferentially to lanthanides, has proved to be
challenging because the chemistries are so similar. Such differences that can be identified include the
marginally enhanced selectivity of “soft” nitrogens in a heteraromatic ring for Am(III) than for
Eu(III). Moreover, there is a body of evidence to suggest that the polydentate ligands are more firmly
bound to the metal than is the case with monodentate ligands. Accordingly, we have been varying the
nature of the donor nitrogen ligands in order to enhance these small differences so that they are
manifested in increased separation factors (Am(III)/Ln(III) [2,3].
In order to do this, however, it is essential to understand the fundamental chemistries involved and
to evaluate the structural types that are formed with different stereochemistries and bound ligand atoms.
For example, the molecule N-carboxybutyl-2-amino-4,6-di-(2-pyridyl)-1,3,5-triazine (BADPTZ), which
is an effective solvent extraction reagent for actinides and lanthanides, has been synthesised,
characterised and its interaction with metal ions studied – see Figure 2. A low pressure cyclisation
route has proved to be particularly useful in the synthesis of these reagents and has enabled a wide
range to be prepared. In particular, the ligand ADPTZ (L 4 ) in which there are no alkyl groups on the
amide group, has enabled the study of the range of structures that are manifest across the whole
lanthanide series [6]. There are five structure types, which can be broadly divided into two equal
groups. The larger lanthanides from La to Sm form structure types in which the co-ordination numbers
are 11 and 10. La is the only lanthanide to form an 11-coordinate complex La(ADPTZ)(NO 3 )(H 2 O) 2 in
which the ligand is tridentate and the nitrates are all bidentate. The La cation is too large to fit into the
tridentate cavity of the ligand and sits outside. One general feature of the lanthanide complexes and
the minor actinides is that there is rather rich co-ordination chemistry and several complexes have
been identified for the same metal and ligands. For example, Nd and Sm both form two different
10-coordinate complexes. The first, M(ADPTZ)(NO 3 ) 3 (H 2 O) is neutral with three bidentate nitrates
and one water molecule. The second, [M(ADPTZ)(NO 3 ) 2 (H 2 O) 3 ] + is a cation with two bidentate
nitrates and three water molecules. For the smaller lanthanides, there are two structure types, each of
which is nine co-ordinate. One of these structures, [LnL 4 (NO 3 ) 2 (H 2 O) 2 ](NO 3 ), is rather unusual in that
a dication is formed together with two nitrate anions. The complex [Yb(ADPTZ)(NO 3 ) 3 (H 2 O)] is
unusual in that it contains a monodentate nitrate anion. Without exception, the structures show
intermolecular hydrogen bonding between the amine group of the metal complexes or solvent
molecules. Such an observation is contrary to Guideline 4 but the inclusion of an alkyl or other
hydrophobic group on the amine group greatly restricts the intermolecular hydrogen bonding so that
BADTPTZ remains a candidate molecule for partitioning [7].
274

Figure 2. The synthesis of ADTPTZ and then BADTPTZ
Thus, it can be seen even from this limited account of structures formed in the solid state that
there is a wide range of structural types to be found with the lanthanides (and minor actinides).
Probably the same also holds true for the actinides in that there is a wide range of structures for a
given ligand plus nitrate and water. EXAFS studies have shown that structures to be found in the solid
state are also present in the liquid phase. The principal implication for solvent extraction, however, is
that there are several possible candidate species that may move across the aqueous and organic solvent
interface during extraction or stripping [8]. In each of the above structures the metal is bound to ligand
molecules (solvent extraction reagent) but in addition there are nitrate anions and/or water molecules.
However, since these species have high rates of self exchange compared with polydentate groups,
there is the possibility that a wide range if structural types are formed during solvent extraction.
Therefore, the above Guideline 4 is pertinent throughout partitioning studies – particularly so if the
species may have very different rates of transfer across the aqueous/organic interface.
2.2 Partitioning – enclosing the primary co-ordination sphere with only the ligand molecules
When the metal cation is completely surrounded by the ligand molecule(s) the rates of
self-exchange of the ligands are minimised because the more mobile ligands such as water and nitrate
are excluded from the co-ordination sphere of the metal. Thus it might be expected that there are
enhanced solvent extraction properties when this is the case. BTPs are reagents in which the
co-ordination sites of the metals may be satisfied by only the ligand. These reagents include
2,6-bis(5,6-dipropyl-1,2,4-triazin-3-yl)-pyridine and one general method of synthesis of these reagents
is shown in Figure 3. For R = propyl, the ligand gave D Am values of between 22 and 45 and SF Am/Eu of
131-143 when 0.034 M of the ligand in modified TPH was used to extract from 0.9 -0.3 M HNO 3 and
different amounts of NH 4 NO 3 [5]. Other BTPs give separation factors between 50 and 150, which
values are also far in excess of those obtained with many other ligands containing just carbon,
hydrogen, nitrogen and oxygen atoms [5,6]. In addition, these BTP ligands do not require the use of a
synergist such as 2-bromodecanoic acid, which is frequently necessary when extractions are carried
out with other nitrogen heterocycles.
275

Figure 3. One method of the synthesis of BTP-type reagents
O
N H 2
N H 2
N
N
N
NH 2
NH 2
R
O
R
CH 2
Cl 2
, MgSO 4
r.t. or reflux
R
R
N
N N
N
N N
N
R
R
The initial solvent extraction studies indicated that the formula of the extracting species for Eu
was possibly ML 3 (NO 3 ) 3 .HNO 3 . The co-ordination of three tridentate heterocyclic aza-aromatic
ligands to an actinide (III) or a lanthanide (III) ion in the presence of a co-ordinating anion such as
nitrate was unprecedented and we initiated structural studies in order to verify the composition of the
complexes formed with the 2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)-pyridines. Whereas for the larger
lanthanides (La-Sm), centrosymmetric dimers of the form [M 2 L 6 (NO 3 ) 6 ] were confirmed for the
smaller lanthanides (Sm-Lu) [ML 3 ] 3+ cations were found with a variety of complex anions. These
results contrast markedly with the types of complexes formed by lanthanum nitrates with other
terdentate nitrogen ligands. For example, we have recently studied complexes formed between the
lanthanide nitrates and the ligand 2,6-bis(5-methyl-1,2,4-triazol-3-yl) pyridine [8] and those formed
with the terpyridyl ligand. The terpyridyl ligand also forms 1:1 complexes of several formulations but
with excess ligand it is possible to form complexes with two ligands e.g. [M(NO 3 ) 2 (terpy) 2 ]
[M(NO 3 ) 4 (terpy)], where M = La, Nd, Sm, Tb and Dy. Our work with other terdentate ligands has
established similar structural features where the majority of complexes showed M:L ratios of 1:1 but
for the larger lanthanides M:L ratios of 1:2 were occasionally observed.
The structure of [Yb(L) 3 ] 3+ where L is the BTP molecule with propyl groups is shown in Figure 4.
Thus, we can conclude that our structural studies provide the first solid state evidence for the unique
type of complex, which is involved in the extraction process using these new solvent extraction
reagents. This structure is also to be found in the with the elements from Sm-Lu and include the
complex cation [Eu(L) 3 ] 3+ cation. The structure shows that the nitrogen atoms in the tridentate cavity
are bound to the central metal cation. Thus the co-ordination sphere, which results in a co-ordination
number nine is completely satisfied by the ligand molecules with no nitrate or water in the primary
co-ordination sphere (Guidelines 2 and 4). The rates of exchange of the ligand(s) are minimised so that
as far as possible there is one extracted species.
2.3 Intermolecular interactions
When the ligand or solvent extraction reagent is bound to the metal ion, there is the possibility to
have a hydrophilic or hydrophobic “external” surface presented to the solvent. In the case where there
is extensive hydrogen bonding, the interaction between the neighbouring extracted species may be too
high. For the [Yb(L) 3 ] 3+ cation, the counter anion may be nitrate or a complex anionic species. In spite
of this, the intermolecular interactions are minimised owing to the hydrophobic nature of the outer
regions of the ligand. These would be further minimised if a neutral species could be extracted.
Perhaps anionic groups could be built into the ligand.
276

2.4 Implications for partitioning
The above guidelines are able to provide general ideas for the design and implementation of new
solvent extraction reagents with enhanced properties. Of course, in one way the guidelines may be
called a “wish-list” of desirable properties. Consideration of the guidelines does, however, indicate
that there is still a lack of fundamental knowledge particularly with respect to the comparison between
the chemistries of the lanthanides and the actinides.
Figure 4. The structure of [Yb(L) 3 ] 3+ where L is the BTP molecule with propyl groups [9]
(Note how the ligands completely occupy the co-ordination sphere of the metal ion
with Yb-N distances ranging from 2.450(11) to 2.499(11)Å)
Acknowledgements
We are grateful for the financial support by the European Union Nuclear Fission Safety
Programme (Contract FI41-CT-96-0010). We would also like to thank the EPSRC and the University
of Reading for funding of the image-plate system for the X-ray structures.
277

REFERENCES
[1] Gabriel Y.S. Chan, Michael G.B. Drew, Michael J. Hudson, Peter B. Iveson, Jan-Olov Liljenzin,
Mats Skålberg, Lena Spjuth and Charles Madic, Solvent Extraction of Metal Ions From Nitric Acid
Solution Using N,N – Substituted Malonamides. Experimental and Crystallographic Evidence for
Two Mechanisms of Extraction, Metal Complexation and Ion-pair Formation, J. Chem. Soc.,
Dalton Trans, 1997 6 649-660.
[2] Michael J. Hudson, Michael G.B. Drew, Mark L. Russell, Peter B. Iveson and Charles Madic,
Experimental and Theoretical Studies of a Triazole Ligand and Complexes Formed With the
Lanthanides, J. Chem. Soc. Dalton, 1999, 2433-2440.
[3] Michael G.B. Drew, Michael J. Hudson, Peter B. Iveson, Jan-Olov Liljenzin, Lena Spjuth and
Charles Madic, Crystal Structures of Two Different Ionic Complexes Formed Between Protonated
Terpyridine Cation and Lanthanide Nitrates, Polyhedron, 1998 18(17) 2845-2849.
[4] Michael J. Hudson, I. Hagström, L. Spjuth, Å. Enarsson, J.O. Liljenzin, M. Skålberg,
Peter B. Iveson, Charles Madic, Pierre-Yves Cordier, Clement Hill and N. Francois, Synergistic
Solvent Extraction of Trivalent Americium and Europium by 2-bromodecanoic Acid and Neutral
Nitrogen-containing Reagents, Solvent Extraction and Ion-Exchange, 178, 221-242 (1999).
[5] Zdenek Kolarik, U. Müllich and F. Gassner, Ion. Exch. Solvent Extr., 1999 17 23; Z. Kolarik,
U. Müllich and F. Gassner, Ion. Exch. Solvent Extr., 1999, 17, 1155.
[6] Michael J. Hudson, Michael G.B. Drew, Peter B. Iveson, Charles Madic and Mark L. Russell,
A Study of Lanthanide Complexes Formed With the Terdentate Nitrogen Ligand 4-aminobis(2,6-(2-pyridyl))-1,3,5-triazine,
J. Chem. Soc., Dalton Trans., 2000, 2711-2720.
[7] Patent (Application) PCT/GB96/01700: Separation of Lanthanides and Actinides with
Heterocyclic Ligands.
[8] Michael J. Hudson, Michael G.B. Drew, Peter B. Iveson and Charles M. Madic, Comparison of the
Extraction Behaviour and Basicity of Some Substituted Malondiamides, Solvent Extraction and Ion
Exchange, 18, 2000 1-23.
[9] Michael J. Hudson, Michael G.B. Drew, Denis Guillaneux, Mark L. Russell, Peter B. Iveson and
Charles Madic, Lanthanide(III) Complexes of a Highly Efficient Actinide(III) Extracting Agent –
2,6-bis(5,6-dipropyl-1,2,4-triazin-3-yl)-pyridine, Inorg. Chem. Commun. 2000, in press.
278

SEPARATION OF MINOR ACTINIDES FROM A GENUINE MA/LN FRACTION
B. Sätmark, O. Courson, R. Malmbeck, G. Pagliosa, K. Römer, J.P. Glatz
European Commission, Joint Research Centre, Institute for Transuranium Elements
Hot Cell Technology, Postfach 2340, 76125 Karlsruhe, Germany
Abstract
Separation of the trivalent Minor Actinides (MA), Am and Cm, has been performed from a genuine
MA(III) + Ln(III) solution using BisTriazinePyridine (BTP) as organic extractant. The representative
MA/Ln fraction was obtained from a dissolved commercial LWR fuel (45.2 GWd/tM) submitted
subsequently too a PUREX process followed by a DIAMEX process. A centrifugal extractor set-up
(16-stages), working in a continuous counter-current mode, was used for the liquid-liquid separation.
In the nPr-BTP process, feed decontamination factors for Am and Cm above 96 and 65, respectively
were achieved. The back-extraction was more efficient for Am (99.1% recovery) than for Cm (97.5%).
This experiment, using the BisTriazinePyridine molecule is the first successful demonstration of the
separation of MA from lanthanides in a genuine MA/Ln fraction with a nitric acid concentration of
ca. 1M. It represents an important break through in the difficult field of minor actinide partitioning of
high level liquid waste.
279

1. Introduction
Radioactive by-products are unavoidably generated during normal reactor operation. Some of
these by-products are very long-lived and radiotoxic elements which must be separated from the
biosphere for a very long time. The potential harmfulness of the wastes generated by reprocessing are
primarily due to the presence of Minor Actinides (MA) and they are of special concern regarding
partitioning and transmutation.
Current reprocessing technology is based on the aqueous PUREX process in which uranium and
plutonium are recovered. The technique can also be extended for the recovery of neptunium, but
americium and curium cannot be separated directly in this process. The partitioning of MA is instead
done by advanced reprocessing of High Level Liquid Waste (HLLW) generated by the PUREX
process, and different systems based on liquid-liquid extraction have been proposed world-wide.
Due to the difficult separation of trivalent MA from trivalent lanthanides (Ln) especially at high
acidities, two step processes are at present considered. In the first step, at high acidity, a group
separation of MA and lanthanides is carried out, followed by a separation of MA from lanthanides at
lower acidity, see Figure 1.
Figure 1. Main routes for the separation of MA from HLLW
dissolved fuel
PUREX
1 cycle process
HLLW
2 cycle process
Ln / MA separation
(at high acidity)
FP
FP (Ln)
MA separation
(at high acidity)
Ln / MA
MA separation
(at low acidity)
Ln
MA
MA
Transmutation
In the French DIAMEX (DIAMide EXtraction) process the minor actinides are directly extracted
from the PUREX raffinate together with fission lanthanides using the completely combustible
DiMethyl-DiButyl-TetraDecyl MalonAmide (DMDBTDMA). The MA(III)+Ln(III) mixture generated
after this first step is low-acidic to facilitate the second process, the SANEX process, which concerns
the separation of the MA from the lanthanides. This process is based on the BTP, which belongs to a
new family of extractants, the Bis-Triazinyl-Pyridine developed by Z. Kolarik et al, and is very
efficient for a selective extraction of MA(III) at high acidity [1] and shows good capabilities in
centrifugal extractors.
In the present work, the 2,6-Bis(5,6-alkyl-1,2,4-Triazin-3-yl)Pyridine (nPr-BTP) process has been
tested in continuous counter-current extraction experiments, using a centrifugal extractor battery
280

installed in a hot cell. The feed was a genuine MA(III)+Ln(III) mixture obtained from small scale
PUREX/DIAMEX reprocessing of commercial LWR fuel (45.2 GWd/tM) [2].
2. Experimental
2.1 Reagents
The nPr-BTP compound was obtained from CEA Marcoule [3]. It was dissolved, using an
ultrasonic bath, in Hydrogenated TetraPropene (TPH) and 30vol% of octanol obtained from
PANCHIM (France) and MERCK (Germany), respectively. The solvent with a final concentration of
0.04M nPr-BTP was directly used as organic phase in the centrifugal extractor experiment.
All reagents and chemicals were of the analytical reagent grade. MQ grade water (18 MΩ/ cm) was
used for all dilutions.
2.2 Continuous experiments using a genuine MA/Ln fraction
The centrifugal extractor equipment installed in the hot cells, see Figure 2, is described elsewhere
[4,5]. For the nPr-BTP process, 16 extractors were used, with 5 extraction stages, 3 acid scrubbing
stages and 8 strip stages. The continuous counter-current centrifugal extraction scheme is shown in
Figure 3. This flow-sheet was optimised on the basis of preliminary data obtained from batch tests
with spiked solutions [6]. The genuine MA/Ln solution obtained as a product in the DIAMEX process
was used as feed solution and adjusted to 1 M HNO 3 with concentrated nitric acid.
Figure 2. Photograph showing the 16 stage continuous
counter-current extractor battery installed in the hot cell
281

Figure 3. Flowsheet for the hot experiment using the nPr-BTP and centrifugal extractors
Organic phase
0.04M nPr-BTP
in TPH / 30vol% octanol
33ml/h
Organic
out
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16
Raffinate
Feed
An/Lnfraction
1M HNO 3
44ml/h
Acid wash
0.1M HNO 3
14ml/h
An fraction
Strip solution
0.05M HNO 3
53ml/h
After the system reached steady-state conditions (ca. 3 h), the outgoing fractions were collected
for about 30 min. At the end of the experiment the centrifuges and the pumps were switched off
simultaneously and samples were taken from mixing chambers (well sampling) of each centrifuge.
All concentrations in the aqueous samples were determined using a quadrupole ICP-MS
(Perkin-Elmer, ELAN250).
3. Results and discussion
The aqueous concentration profiles of actinides (Np, U, Pu, Am and Cm) in µg of isotope per g of
solution are shown in Figure 4.
Figure 4. Aqueous profiles of actinides in the wells
100
Extraction
Scrub
Strip
10
243 Am
Concentration [µg/g]
1
0.1
0.01
244 Cm
239 Pu
238 U
237 Np
0.001
Organic phase
Aqueous phase
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16
Stage
Similar extraction behaviour of Am and Cm is observed and the concentrations of those elements
decrease by several orders of magnitude. Np and U are efficiently washed out in the scrubbing section
but Pu is co-extracted. In the strip section all actinides are back-extracted. It should be mentioned that
282

the concentrations of U, Pu and Np in a MA/Ln feed originating from optimised PUREX/DIAMEX
process schemes will be insignificant.
In Figure 5 the aqueous concentration profiles of lanthanides are shown.
Figure 5. Aqueous profiles of lanthanides in the wells
1000
Extraction
Scrub
Strip
100
Concentration [µg/g]
10
1
0.1
153 Eu
152 Sm
156 Gd
146 Nd
0.01
140 Ce
141 Pr
Organic phase
139 La
Aqueous phase
0.001
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16
Stage
As expected, the lanthanides are not extracted and the acid scrubbing efficiently reduces the
co-extraction. However, the scrubbing efficiency significantly decreases with the increase of the
element number. Higher lanthanides (Eu, Gd) are to some extent transported to the organic phase and
efficiently back-extracted together with the MA. This co-extraction can be prevented by addition of
more scrubbing stages.
The aqueous concentration profiles of some lighter fission products are shown in Figure 6.
283

Figure 6. Aqueous profiles of fission products in the wells
100
Extraction Scrub Strip
10
Concentration [µg/g]
1
0.1
105 Pd
89 Y
99 Tc
98 Mo
0.01
0.001
Organic phase
102 Ru
Aqueous phase
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16
Stage
Extraction can be seen for Tc, Pd and Mo. The acid scrubbing efficiently removes Ru and Y from
the organic phase, especially Ru is almost completely washed-out. For Tc the number of scrubbing
steps is not enough, and Pd is even extracted in the scrubbing section. In the strip section only Y is
well back-extracted. The other elements Tc, Mo and Pd are accumulated in the organic phase.
In Table 1 the process decontamination factors, DF, is shown. They were calculated according to
Equation 1, where C and V are aqueous component mass concentration (µg/g) and total volume (mL),
respectively.
C
DF =
C
feed
raff
⋅ V
⋅ V
feed
raff
(1)
Table 1. Decontamination factors of the feed (in µg /g)
DF DF DF DF
89 Y 1.02 105 Pd 300
144 Nd 1.00
237 Np 1.00
98 Mo 13
99 Tc 11
139 La 1.00 152 Sm 1.00
140 Ce 1.00 153 Eu 1.01
243 Am 122
244 Cm 64
101 Ru 1.00 141 Pr 1.00 156 Gd 1.04
In spite of the high acidity of the process (1M HNO 3 ) high decontamination factor is achieved for
the MA elements. Lanthanides are not extracted except for Eu and Gd showing DF of 1.01 and 1.04,
respectively. In parallel the DF of Eu, Cm and Am were also determined by α and γ spectrometry to be
1.01, 151 and 112 respectively.
284

Table 2 shows the recovery in the raffinate and in the MA fraction obtained after 3.5 hours of
experiment. The small amounts of co-extracted lanthanides are efficiently back-extracted in the
scrubbing section, as can be seen in Figures 5 and 6. To decrease the amount of co-separated higher
lanthanides the number of scrubbing stages has to be increased. This is also the case for some lighter
fission products such as Y, Mo and Tc.
Raffinate
MA
fraction
Table 2. Recovery (% feed)
Raffinate
MA
fraction
Raffinate
MA
fraction
Y 98.2 1.8 Ce >99.99 – Np 99.9 0.05
Mo 7.5 6.0 Pr >99.99 – Pu 52 48
Tc 9.2 8.5 Nd >99.9 – Am 0.8 99.1
Ru 99.8 0.2 Sm 99.9 0.1 Cm 1.6 97.5
Pd 0.3 3.9 Eu 99.9 0.1
La >99.99 – Gd 96 4
4. Conclusion
The process reported, using the BisTriazinePyridine molecule, is the first successful
demonstration of MA separation from lanthanides in a genuine MA/Ln fraction with a nitric acid
concentration ~1 M. In the experiment, carried out in a centrifugal continuous counter-current set-up, a
MA fraction almost free of lanthanides was obtained. Due to an efficient MA extraction and backextraction
a reasonably good recovery of Am was achieved. However, the process scheme has to be
improved to increase the recovery of Cm, to decrease the co-extraction of lanthanides, and to prevent
the Pd accumulation in the organic phase. Nevertheless, this result represents an important break
through in the difficult field of minor actinide partitioning of high level liquid waste.
Acknowledgements
This work was partially funded by the European Commission in the framework of its R&D
programme “Nuclear Fission Safety (1994-1998)”, Project: “NEWPART: New partitioning techniques”,
contract FI4I-CT96-0010.
285

REFERENCES
[1] Z. Kolarik, U. Müllich, F. Gasner, Selective Extraction of Am(III) Over Eu(III) by
2,6-ditriazolyl- and 2,6-ditriazinylpyridines, Solv. Extr. Ion Exch. 17(1), (1999) 23.
[2] R. Malmbeck, O. Courson, G. Pagliosa, K. Römer, B. Sätmark, J.P. Glatz, P. Baron, Partitioning
of Minor Actinides From HLLW Using the DIAMEX Process. Part 2 – “Hot” Continuous
Counter-current Experiment, Radiochim. Acta in press (2000).
[3] C. Madic, M.J. Hudson, J.O. Liljenzin, J.P. Glatz, R Nannicini, A. Facchini, Z. Kolarik, R. Odoj,
New Partitioning Techniques for Minor Actinides – Final Report, EU Nuclear Science and
Technology (EUR19149, 2000).
[4] O. Courson, M. Lebrun, R. Malmbeck, G. Pagliosa, K. Römer, B. Sätmark, J.P. Glatz,
Partitioning of Minor Actinides From HLLW Using the DIAMEX Process,
Part 1 – Demonstration of Extraction Performances and Hydraulic Behaviour of the Solvent in
a Continuous Process, Radiochim. Acta in press (2000).
[5] J.P. Glatz, C. Song, X. He, H. Bokelund, L. Koch, Partitioning of Actinides From HAW in a
Continuous Process by Centrifugal Extractors, Special Symposium on Emerging Technologies
in Hazardous Waste Management – Atlanta (D.W. Tedder, ed.), ACS, Washington DC 1993.
[6] P. Baron, Personal communication.
286

PARTITIONING ANIONIC AGENTS BASED ON 7,8-DICARBA-NIDO-UNDECABORATE
FOR THE REMEDIATION OF NUCLEAR WASTES
Clara Viñas, Isabel Rojo, Francesc Teixidor
Institut de Ciència de Materials de Barcelona, CSIC,
Campus UAB, Bellaterra 08193, Spain
Abstract
[3,3’-M(1,2-C 2 B 9 H 11 ) 2 ] - (M = Co 3+ , Fe 3+ , Ni 3+ ) anionic compounds perform similarly in PVC
membranes as Cs + sensors in ion selective electrodes. Their behaviour is very similar, but the higher
stability of [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - makes it the more interesting for extraction at low pH’s. Species
[1,1’-(PPh 2 ) 2 -3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - , [2] - , [1,1’-(OPPh 2 ) 2 -3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - [3] - , [3,3’-Co(1-
CH 3 -2-(CH 2 ) n OR-1,2-C 2 B 9 H 9 ) 2 ] - ([4] - : n = 3, R = -CH 2 CH 3 ; [5] - : n = 3, R = -(CH 2 ) 2 OCH 3 ; [6] - : n = 3,
R = -(CH 2 ) 3 CH 3 ; and [7] - : n = 6, R = -(CH 2 ) 3 CH 3 ), were tested for 137 Cs, 90 Sr and 152 Eu in extraction.
Permeability tests on Supported Liquid Membranes with H[6], and H[7] have shown that these
compounds present the highest values reported so far for this sort of radionuclides transport
experiments.
287

1. Introduction
Nuclear waste reprocessing operations produce both high level and medium level activity liquid
wastes (HLW/MLW). The major nuclides in these radioactive wastes are those with long half-lives,
mainly β/γ emitters or α emitters such as transuranium elements. This is why great efforts have been
devoted throughout the world to propose harmless storage of these wastes. The burial of vitrified
reprocessed HLWs (containing fission products and α emitters) has been considered as the safest
method for their permanent disposal, whereas MLWs are treated by evaporation in order to
concentrate their radioactivity into the smallest possible volume. This treatment nevertheless leads to
large volumes of concentrates composed of active and inactive salts (mainly: NaNO 3 , 4 mol.l -1 and
HNO 3 , 1 mol.l -1 as the matrix). The greater part of these concentrates has to be disposed off in
geological formations after embedding due to their activity in long-lived radionuclides (actinides,
strontium, caesium, etc.). Therefore it would be desirable to remove these long-lived radionuclides
from the contaminated liquid wastes before embedding. These would allow a large part of these wastes
to be directed to a subsurface repository, and a very small part containing most of the long-lived
radionuclides to be disposed off, after conditioning, in geological formation [1].
The field of metallacarborane chemistry was initiated by Hawthorne in 1965 [2]. Since that time,
metallacarboranes from all areas of the periodic table have been prepared using the dicarbollide ligand
[3] [C 2 B 9 H 11 ] 2- (Figure 1). These derivatives have become of increasing interest with regard to their
solubility [4], isolation, separation and characterisation of organic bases, radiometal carriers [5],
electron acceptor molecules [6], among other areas. One of these organometallic complexes,
[3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - ([1] - ), has attracted the most attention because of its robustness, its stability in
the presence of strong acid (HNO 3 ), at relatively high temperatures and under a very high radiation
[7]. This stability allows it to be considered for nuclear waste remediation. Its hexachloro protected
analogue, [1-Cl 6 ] - , is remarkable as an extractant (Figure 2). The large size to charge ratio and the
hydrophobic nature of [1] - and [1-Cl 6 ] - allows extraction of caesium and strontium ions from an
aqueous phase to an organic phase, leaving other alkaline and higher-valent metals behind [4,7,8]. The
ions 137 Cs and 90 Sr are used for thermoelectric generators and sterilisation of medical equipment,
among other areas, making the possibility of recycling them very attractive [9].
Figure 1 Figure 2. [1-Cl 6 ] -
Cl
Cl
Cl
Cl
CH
BH
Cl Cl
Co
CH
B or BH
The [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - , [1] - anion, has been used as a highly selective 137 Cs sequestering
agent in extraction processes, in the presence of Na + , with nitrobenzene [4] as the receiving phase.
However, nitrobenzene is an ecologically unacceptable solvent, so that other receiving phases are
required for environmental applications. Thus, other solvents are needed. On the other hand, additional
metallacarboranes similar to [1] - are available, namely these with Fe 3+ , and Ni 3+ . In an effort to decide
288

which of these, the Co 3+ , Fe 3+ and Ni 3+ metallacarboranes was the most adequate, their Cs + salts were
implemented in ion selective electrodes (ISE’s). As the potentiometric performance of an ISE can be
viewed to be similar to the transport process in one membrane their study as Cs + sensors would provide
reliable and important information on the stability and extracting capacity of these metallacarboranes,
hence permitting to decide which one was more adequate. The study was also intended to discern which
of the three anions was more selective towards Cs + , in order to choose one for the subsequent studies.
The results obtained, however, indicated that the three anions [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - , [3,3’-Fe(1,2-
C 2 B 9 H 11 ) 2 ] - , and [3,3’-Ni(1,2-C 2 B 9 H 11 ) 2 ] - were comparable in their behaviour in ISE’s [10]. Since the
Co 3+ complex provides easier synthetic routes, higher yield and higher stability, it was chosen for the
studies on the extraction of radionuclides.
Organic compounds incorporating oxygen in the molecule mainly as ethers, and phosphine oxide
derivatives have been also tested for this kind of radionuclides waste removal [11]. Accordingly
several ether C-substituted cobaltacarboranes [12,13] were prepared and great effort has been
dedicated more recently to phosphine oxides, and fluorinated compounds in order to improve the
efficiency showed by [1] - and [1-Cl 6 ] - .
In this paper we report on the synthesis and extracting possibilities offered by these C-substituted
cobaltabisdicarbollide species, along with some comments on the B-substitution.
2. Discussion
2.1 What benefits can be expected from the polyhedral anions?
Anions with low nucleophilicity, good solubility and weak coordination capacity have recently
experienced great interest in areas of major commercial importance such as olefin polymerization [14],
lithium battery technology [15], and the radionuclides extraction mentioned above. In 1986 the
-
carborane anion CB 11 H 12 was introduced as a candidate for the least co-ordinating anion [16].
Paradoxically, carboranes combine a versatile functionalization chemistry with unparalleled inertness.
One of their main characteristics, which is relevant to this work, is the delocalization of the anionic
charge throughout its volume, thus producing the anions with lowest charge density. It has been
demonstrated [17] that the delocalized charge on these large anions tends to make them nearly ideal
spectator ions with little opportunity to perturb the structure of the cation. In addition to the 12-vertex
ions, CB 11 H - 12 and B 12 H 2- 12 there are related classes of anions based on the 10-vertex B 10 H 2- 10 and
CB 9 H - 10 ion, as well as the above mentioned M 3+ bisdicarbollides. The cobaltabisdicarbollide anion,
whose hexachloroderivative molecular structure is depicted in Figure 2 consists of a Co(III) ion
sandwiched by two dicarbollide moieties. Each dicarbollide (C 2 B 9 H 11 ) 2- bears two negative charges,
overall producing a mononegative species. This anion presents a great chemical resistance, e.g. it
withstand in HNO 3 2M and in concentrated HCl for several days without apparent decomposition.
Thus, these anions do not encounter a parallel in current inorganic or organic areas of chemistry. They
are very relevant and can hardly be replaced by other anions. They offer other possibilities derived
from their elemental nature as neutron scavengers or glass forming elements, both important for
nuclear waste remediation.
289

Figure 3
Cs
2.2 Possibilities of modification
As can be seen from Figure 2, two different sort of reacting points are possible in [3,3’-Co(1,2-
C 2 B 9 H 11 ) 2 ] - , the BH’s and the CH’s. Not all BH’s sites are equally reactive, those that have permitted
substitution by Cl are the more reactive. This matter has been developed by the Boron Chemistry
group at Rez near Prague, and their results shall be attributed to S. Hermaneck, J. Plesek and
B. Grüner [18]. One excellent example for extraction of Cs + is given by bisphecosan represented in
Figure 3. Excellent recent results derive from the dioxane substituted Co(1,2-C 2 B 9 H 11 ) 2 ] - complex,
8-C 4 H 8 O 2 -3,3’-Co(1,2-C 2 B 9 H 10 ) (1’,2’-C 2 B 9 H 11 ). This is neutral and strongly susceptible to
nucleophilic attack by anionic nucleophiles, being a precious and highly versatile starting material to
produce a rich variety of anionic extracting agents for actinides, the result being only limited to the
availability of anionic nucleophiles. These results have been produced in the frame of contract IC15-
CT98-0221 and shall be attributed to the same authors from Rez plus Dr. J. Baca.
2.3 Synthesis of C-substituted derivatives of [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] -
A general approach to the synthesis of C-substituted derivatives of cobalt bis(1,2-dicarbollide)
consists in the preparation of the corresponding substituted o-carboranes, their degradation into the
nido-7,8-dicarbaundecaborates, followed by deprotonation and reaction with cobalt(II) chloride [19].
This synthesis is extremely time consuming and always shall produce [3,3’-Co(1-R-1,2-C 2 B 9 H 10 ) 2 ] -
derivatives with identical substituents in each dicarbollide moiety.
However, a more versatile method starting from [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - was needed as it would
permit a more rapid way to produce derivatives at carbon. A recent approach by Chamberlain et al., to
[3,3’-Co(1-R-1,2-C 2 B 9 H 11 ) (1,2-C 2 B 9 H 10 )] - anionic mono-derivatives consists in the treatment of
[3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - with n-butyllithium followed by the reaction with alkyl halides [20].
This procedure was a good alternative as [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - was a readily available starting
material. Some years before we had attempted the same procedure but we had got no conclusive
results [21]. According to Chamberlain et al. the synthetic procedure is straightforward. A violet
colour due to the deprotonated [3,3’-Co(1,2-C 2 B 9 H 10 ) (1,2-C 2 B 9 H 11 )] 2- dianion is produced upon the
addition of 1 equivalent of BuLi to the orange initial solution of [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - , in THF.
290

This dianionic species must be extremely basic, thus very reactive towards weak acids. We had
abandoned this reaction in 1994 since following an equal procedure, the product always reverted to the
orange colour of the cobalt bis(1,2-dicarbollide). However, the interesting report [20] from Los
Alamos National Laboratory, brought back the possibility of this, otherwise, extremely interesting
reaction.
The reaction is very tricky and does not proceed precisely, in our hands, as the authors say. By
using the same reagents as they did we were able to get partial conversion, usually of the order of
30-40%. The rest was unreacted [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - . Now, after having reinvestigated this
reaction we believe that this “unreacted [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - ” is in fact a “reverted [3,3’-Co(1,2-
C 2 B 9 H 11 ) 2 ] - ”.
3. Preliminary results
The solvent initially used was dimethoxyethane, CH 3 OCH 2 CH 2 OCH 3 , however after noticing the
strong basicity of the deprotonated [3,3’-Co(1,2-C 2 B 9 H 10 ) (1,2-C 2 B 9 H 11 )] 2- , this was replaced by THF.
The initial reagents used were Cl(CH 2 ) 3 Br and ClPPh 2 . For this last one, a minor product was obtained
which suggested that some reaction, although in very low extension, had taken place. The rest was
[3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - . This led us to think that [3,3’-Co(1,2-C 2 B 9 H 10 ) (1,2-C 2 B 9 H 11 )] 2- was a strong
base but a weak nucleophile. Thus, TMDA was added to the reaction pot, (TMDA =
(CH 3 ) 2 NCH 2 CH 2 N(CH 3 ) 2 ), as it is known that butyllithium aggregates show a marked increase in the
reactivity when co-ordinating solvents or reagents are added. This explains the initial use of
dimethoxyethane. However, neither at room temperature nor under refluxing conditions the result did
improve.
Other reagents were BrCH 2 CH 2 OCH 2 CH 3 , alone or in the presence of AgBF 4 . The result was as
unsuccessful as previously. The low nucleophilicity of [3,3’-Co(1,2-C 2 B 9 H 10 ) (1,2-C 2 B 9 H 11 )] 2- had to
be overcome by a good leaving group on the second reagent, as was obvious when using Cl(CH 2 ) 6 I. In
this case the 11 B{ 1 H} NMR of the reaction crude indicated that there was something else besides
[3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - , but this ended up to be a mixture of several compounds with very similar
properties which did not permit an adequate separation, but the interesting point was that it seemed to
be necessary the existence of a good leaving group in the reagent.
Figure 4
BuLi
ClPPh 2
dme
PPh 2 +
PPh
PPh 2
2
Although results seemed to be improving the method was not convenient as cumbersome
separation procedures were needed. However from the data we had gathered the following conclusions
could be inferred: 1) The existence of weakly acidic hydrogens in the reagent was responsible for the
291

protonation of [3,3’-Co(1,2-C 2 B 9 H 10 ) (1,2-C 2 B 9 H 11 )] 2- , thus reverting to [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - ; 2)
Good leaving groups are necessary to facilitate attack by [3,3’-Co(1,2-C 2 B 9 H 10 ) (1,2-C 2 B 9 H 11 )] 2- ; and
3) A high excess of reagent is convenient to lower the ratio of reverted [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - .
Considering these points it was initially decided that I 2 was a very good candidate. It did not carry
any hydrogen, is susceptible to nucleophilic attack and it can be removed from the flask reaction by
sublimation at room temperature. As it was described in the introduction, I 2 has been adequate to
produce electrophilic substitution on boron atoms in [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - . Contrarily, in this
project, the formation of a C-I bond is sought as a versatile candidate for C-C bond formation, by
using appropriate Grignard reagents for coupling reactions.
4. Synthesis of diphenylphosphine cobaltabisdicarbollide derivatives
The reaction is schematised in Figure 4. Reaction of [Co(C 2 B 9 H 11 ) 2 ] - with BuLi in
dimethoxyethane followed by chlorodiphenylphosphine yields the disubstituted species [2] - and only
minor amounts of the monophosphine. Concerning the disubstituted species, a mixture of isomers
could be possible: the meso form (C s symmetry) and a racemic mixture (C 2 symmetry). The first
contains a mirror plane (σ h ) while the second contains a C 2 axis. These possible species are
schematically represented in Figure 5. The 13 C-, and 1 H-NMR confirms the existence of only one of
the two possible species. To determine which one of the two had been obtained, crystals suitable for
X-ray diffraction were grown [22]. The structure demonstrated that the racemic form was the one
produced. These phosphine compounds are basic as a consequence of the negative charge, and of the
phosphines themselves. These should not be compatible with the strong acid conditions found in the
radioactive waste. The phosphine oxides would be more convenient for the purposes of this project.
Figure 5
PPh 2
PPh 2
meso
PPh 2 Ph 2 P
PPh 2
Ph 2 P
racemic mixture
5. Synthesis of diphenylphosphine oxide cobaltabisdicarbollide derivatives
The reaction is schematised in Figure 6. Reaction of [1,1’-(PPh 2 ) 2 -3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - , [2] -
with H 2 O 2 in acetone yields the expected phosphine oxide.
292

Figure 6
Ph 2 P
acetone
PPh 2 H2 O 2
Ph 2 P
O
O
PPh 2
As the initial diphosphine was in the racemic form, upon oxidation, it was required that the
resulting phosphine oxide maintained this isomerism. Indeed this is what happened. This was fully
confirmed by the X-ray structural determination of suitable crystals. The acid stability of both [1,1’-
(PPh 2 ) 2 -3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - and [1,1’-(OPPh 2 ) 2 -3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - [3] - was studied by
dissolving the Cs + salts of these two species in 3M HNO 3 and HCl. The resulting solutions were then
neutralized with 3M NaOH. The possible modifications were then followed by 31 P-NMR. It was found
that the oxide was recovered fully, while the initial phosphine could be recovered in a 80%. These
results demonstrated the stability of the phosphine oxides versus the starting phosphines. The [1,1’-
(OPPh 2 ) 2 -3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - oxide is one cluster anion bearing two OPPh 2 - units, similar to the
anion shown in Figure 7. This would allow to study the role of the cluster. Both are monoanionic
species and both contain two OPPh 2 - units.
Figure 7
O
PPh 2
PPh 2
O
6. Introducing bridging fragments
C,C’ bridged species could be relevant as they hinder rotation of the two halves of the molecule.
Interestingly, extraction results found in the former project CIPA-CT93-0133 (EUR 18217 EN) by the
IIC-, NRI (Rez) and CEA (Cadarache) upon derivatives of [Co(C 2 B 9 H 11 ) 2 ] - bridged at the
8,8’- positions, both being boron positions, e.g. PHECOSAN stimulated research on these species.
Similarly, bridged [Co(C 2 B 9 H 11 ) 2 ] - on the carbon atoms, have been produced now. Contrarily to those
described formerly, the bridge in these cases would be between carbon atoms.
Syntheses have been performed as described in Figure 4. Cl 2 PPh and perfluorinated benzene have
been used as reagents. The bridging phosphine has been oxidized following the H 2 O 2 /acetone
procedure shown in Figure 6.
293

Figure 8
F
O
PPh 2
F
F
F
[3] -
6.1 Protecting the cluster with halogen groups
It was long reported that hexachlorinated [Co(C 2 B 9 H 5 Cl 6 ) 2 ] - and dibrominated [Co(C 2 B 9 H 10 Br 2 ) 2 ] -
derivatives of [Co(C 2 B 9 H 11 ) 2 ] - withstand acid conditions better than the parent compound. The fact
that the hexachloro and the dibromo protected species had a comparable stability in acid conditions
indicated that the crucial point was substitution on 8,8’- as these are the only common protected
positions in both compounds. Considering that the low performance of these and other previously
reported [Co(C 2 B 9 H 11 ) 2 ] - derivatives at [H + ]>0.1 M in Eu, Sr, and actinides had to be due to the
reactivity of BH’s at the 8,8’- positions, a project leading to protect these positions is under way.
This may be addressed starting from the already protected 8,8’- positions either with halogen or
carbon atoms. A second alternative would be to protect the phosphine or phosphine oxide product.
Results have shown that starting from the [8,8’-Cl 2 -Co(C 2 B 9 H 19 ) 2 ] - is better than starting from [Cl 6-
Co(C 2 B 9 H 8 ) 2 ] - . This may have to do with steric hindrance.
On the other hand, when the protection reaction was performed on the [1,1’-(PPh 2 ) 2 -3,3’-Co(1,2-
C 2 B 9 H 11 ) 2 ] - anion, results concerning halogen substitution at boron atoms were unsuccessful. Reaction
with Br 2 in glacial acetic acid led to borates.
More encouraging seem to be the results obtained from the phosphine oxide [1,1’-(OPPh 2 ) 2 -3,3’-
Co(1,2-C 2 B 9 H 11 ) 2 ] - . Reactions have been directed to get 8,8’- disubstitution (protection) as these seem
to be sufficient to improve the stability of the cluster in acid media.
7. Extractants derived from [Co(C 2 B 9 H 11 ) 2 ] -
The series of compounds derivatives of [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - derivatives incorporating ether
groups, [3,3’-Co(1-CH 3 -2-(CH 2 ) n OR-1,2-C 2 B 9 H 9 ) 2 ] - , which are presented graphically in Figure 9 ([4] - :
n = 3, R = -CH 2 CH 3 ; [5] - : n = 3, R = -(CH 2 ) 2 OCH 3 ; [6] - : n = 3, R = -(CH 2 ) 3 CH 3 ; [7] - : n = 6,
R = -(CH 2 ) 3 CH 3 ) [12,13] have been studied following the general approach indicated in “Synthesis of
C-Substituted Derivatives of [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - ”.
294

Figure 9
H 3 C
RO(CH 2 )n
CH 3 (CH 2 )nOR
[4] - n=3, R= CH 2 CH 3
[5] - n=3, R= (CH 2 ) 2 OCH 3
[6] - n=3, R= (CH 2 ) 3 CH 3
[7] - n=6, R= (CH 2 ) 3 CH 3
The bis(dicarbollide) derivatives [4] - , [5] - , [6] - , [7] - , have been tested in the liquid-liquid
extraction of 137 Cs, 90 Sr and 152 Eu from the aqueous HNO 3 phase to the organic nitrophenyl hexyl
ether.
For the extraction of 137 Cs, the polyether substituted compounds show a very high extraction
efficiency at pH 3 (D>100) regardless of the nature of the exocluster chain. This efficiency is expected
to be lower if the acidity of the medium is increased. This behaviour is displayed by [4] - , whose D
value decreases from >100 to 3, just by varying the pH value from 3 to 1. However, [6] - , which is very
similar to [4] - , but with the longest alkyl chain next to the oxygen atom, maintains an excellent
efficiency for the extraction of 137 Cs even at pH 1. Anion [6] - shows again the best performance for the
extraction of 90 Sr but the D value decreases strongly with decreasing pH. For the extraction of 152 Eu, it
seems that the larger the exocluster chain the better the performance of the extracting agent. Therefore,
[7] - shows the best performance in the extraction of 152 Eu. The lower efficiency in the extraction of
90 Sr and 152 Eu shown by compound [6] - and the excellent results obtained in the extraction of 137 Cs,
should permit a selective extraction of Cs + from a mixture containing all the radionuclides in solution.
This result led us to perform some transport experiments at CEA (Cadarache) by using Supported
Liquid Membranes (S.L.M.) with NPHE (nitrophenylhexyl ether) as the membrane solvent.
Preliminary results carried out with compounds H[4], H[5], H[6] and H[7] showed that the best
transport performance was for compound H[6]. Compound [7] - , displaying the best efficiency in the
extraction of 152 Eu, was used for its transport studies.
At pH 3, transport of 137 Cs was very efficient with compound [6] - , and a permeability of 30.6
cm/h was obtained. An extraction of 93% of Cs + in 1 hour was achieved.
In the case of the transport of europium, transport with compound H[7] is very rapid at pH 3,
showing a permeability of 8.9 cm/h, i. e. an extraction of 31.2% after 1 hour or 91.3 after 3.5 h. For
comparison, permeabilities ranging from 1 to 4 cm/h have been measured for several “carriers” such
as calix[4]arenes crown 6, CMPO (carbamoylmethylphosphine oxides) or diphosphine dioxides under
comparable conditions. Improved permeabilities were achieved with calixarenes incorporating CMPO
moieties (4-7 cm/h) [23]. Thus, generally speaking, the cobalta(dicarbollide) carriers are considerably
faster transport agents than others well-recognised as doing this job, such as those indicated above
[12,13].
As expected H[2] does not display a good extracting capacity at 3≥pH due to the basicity of the
phosphine groups, enhanced by the negative charge of the anion. This extracting capacity is very much
improved when the phosphine oxides are utilised, however it diminishes abruptly at pH = 1. It is
295

emarkable the Sr affinity of the fluorinated bridged species shown in Figure 8. The existence of the
bridging monophosphine oxide does not influence favourably the extraction.
8. Conclusion
Metallacarborane complexes of formula [3,3’-M(1,2-C 2 B 9 H 11 ) 2 ] - (M = Co, Fe, No) have been
tested as Cs sensors. The results have demonstrated a similar behavior, however due to the large
stability of the Co complex and the better yield this seem to be the anion to be studied.
Synthesis of [3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - derivatives can be made from the o-carborane, producing
the wanted substitution , removing the boron connected to both carbon atoms, and the proton. The
dianion, thus formed, is reacted with anhydrous cobalt chloride. An alternative is to start directly from
[3,3’-Co(1,2-C 2 B 9 H 11 ) 2 ] - . In this paper compounds obtained by both ways are prepared, however the
second provides a more accessible route to these extracting agents.
The largest drawback that these anions suffer is the lost of activity at low pH, specially below 1.
This will require protection at the boron 8,8’- positions. The phosphine oxide extracting agents seem
to be adequate for radioactive waste treatment once the pH dependence problem is solved.
On the other hand the anionic character and low charge density of these extracting agents along
with the high boron contents and specially the results already obtained by us and other groups
participating in the EU project make them extremely suitable candidates to solve the present problem
of partitioning.
REFERENCES
[1] (a) R.E. Berlin and C.C. Stanton, Radioactive Waste Management, John Wiley and Sons Ed.,
USA, 1989; (b) Les Déchets Nucléaires (Dossier Scientifique, Société Française de Physique),
R. Turlay, Les Éditions de Physique, Les Ulis, 1997.
[2] (a) M.F. Hawthorne, D.C. Young and P.A. Wegner, J. Am. Chem. Soc., 87 (1965) 1818;
(b) M.F. Hawthorne and T. D. Andrews, J. Chem. Soc., Chem. Commun., (1965) 443.
[3] A.K. Saxena, J.A. Maguire and N. S. Hosmane, Chem. Rev., 97 (1997) 2421.
[4] J. Plesek, Chem. Rev., 92 (1992) 269.
[5] (a) K. Shelly, C.B. Knobler, and M.F. Hawthorne, New J. Chem., 12 (1988) 317;
(b) M.F. Hawthorne, Pure Appl. Chem., 63 (1991) 327.
[6] P.A. Chetcuti, W. Hofherr, A. Liégard, G. Rihs, G. Rist, H. Keller and D. Zech, Organometallics,
14 (1995) 666.
296

[7] J. Rais and P. Selucky, Nucleon, 1 (1992), 17.
[8] S.D. Reilly, C.F.V. Mason and P.H. Smith, Cobalt(III) Dicarbollide: A Potential 137 Cs and 90 Sr Waste
Extraction Agent, Report LA-11695, Los Alamos National Laboratory: Los Alamos, NM (1990).
[9] P.K. Hurlburt, R.L. Miller, K.D. Abney, T.M. Foreman, R.J. Butcher and S.A. Kinkead, Inorg.
Chem., 34 (1995) 5215.
[10] C. Viñas, S. Gomez, J. Bertran, J. Barron, F. Teixidor, J.F. Dozol, H. Rouquette, R. Kivekäs and
R. Sillanpää, J. Organomet. Chem., 581 (1999) 188.
[11] (a) J.F. Dozol, V. Böhmer, A. McKervey, F. Lopez Calahorra, D. Reinhoudt, M.J. Schwing,
R. Ungaro and G. Wipf, New Macrocyclic Extractants for Radioactive Waste Treatment: Ionizable
Crown Ethers and Functionalised Calixarenes, EU Report EUR 17615 EN, Luxembourg (1997);
(b) R. Ungaro, F. Casnati, F. Ugozzoli, A. Pochini, J.F. Dozol, C. Hill and H. Rouquette, Angew.
Chem. Int. Ed. Engl., 33 (1994) 1506.
[12] C. Viñas, S. Gomez, J. Bertran, F. Teixidor, J.F. Dozol and H. Rouquette, Chem. Commun.,
(1998) 191.
[13] C. Viñas, S. Gomez, J. Bertran, F. Teixidor, J.F. Dozol and H. Rouquette, Inorg. Chem., 37
(1998) 3640.
[14] A.M. Thayer, Chem. Eng. News 1995, 73, 15.
[15] K. Seppelt Angew. Chem., Int. Ed. Engl. 1993, 32, 1025.
[16] K. Shelly, C.A. Reed, Y.J. Lee, W.R. Scheidt. J. Am. Chem. Soc. 1986, 108, 3117.
[17] Z. Xie, R. Bau, C. A. Reed. Inorg. Chem. 1995, 34, 5403.
[18] F. Teixidor, B. Casensky, J.F. Dozol, S. Hermanek, H. Mongeot, J. Rais., New Trends in the
Separation of 137Cs, 90Sr and Transplutonium Elements from Radioactive HLW by Borane and
Heteroborane Anions. European Commission, Nuclear Science and Technology, 1998, EUR 18217, 2.
[19] a) Viñas C., Pedrajas, J. Bertran, J. Teixidor, F. Kivekäs, R. Sillanpää, R. Inorganic Chemistry,
36 (1997) 2482. b) Viñas, C., Gomez, S., Bertran, J., Teixidor, F., Dozol, J-F. Rouquette,
H. Inorganic Chemistry, 37 (1998) 3640. c) Viñas C.; J. Bertran, J. Gomez, S. Teixidor,
F. Dozol, J-F. Rouquette, H. Kivekäs, R. Sillanpää, R Journal of Chemical Society, Dalton
Transactions (1998) 2849.
[20] R.M. Chamberlin, B.L. Scott, M.M. Melo, K.D. Abney, Inorganic Chemistry 1997, 36, 809.
[21] C. Viñas, S. Gomez, J. Bertran, F. Teixidor, Unpublished data.
[22] C. Viñas, I. Rojo, F. Teixidor, R. Kivekäs, R. Sillanpää, Unpublished work.
[23] (a) C. Hill, J.F. Dozol, H. Rouquette, S. Eymard and B. Tournois, J. Membrane Sci., 114 (1996) 73;
(b) F. Arnaud-Neu, V. Böhmer, J.F. Dozol, C. Grüttner, R.A. Jakobi, D. Kraft, O. Mauprivez,
H. Rouquette, M.J. Schwing-Weill, N. Simon and W. Vogt, J. Chem. Soc., Perkin Trans. II,
(1996) 1175; (c) H.J. Cristau, P. Mouchet, J.F. Dozol and H. Rouquette, Heteroatom Chem., 6
(1995) 533.
297

SESSION III
PARTITIONING
SUB-SESSION III-B:
DRY REPROCESSING
299

PYROCHEMICAL PROCESSING OF IRRADIATED TRANSMUTER FUEL
James J. Laidler
Argonne National Laboratory
9700 South Cass Avenue, Argonne, Illinois 60439-4837, USA
James C. Bresee
US Department of Energy,
NE-20, Forrestal Building, 1000 Independence Avenue, SW, Washington, D.C. 20585, USA
Abstract
The US accelerator transmutation of waste program is directed toward the destruction of transuranic
elements and long-lived fission products present in spent light water reactor fuel. Initial separation of
these materials from the light water reactor spent fuel will be accomplished by conventional aqueous
processing methods. The transuranic elements will be incorporated in blanket fuel assemblies that will
be irradiated to burn-ups in the range of 30 atom percent, and the fuel assemblies will be processed to
recover and recycle unburned transuranic elements and newly-generated long-lived fission products.
The accelerator driven transmutation system will use a fuel type much different from light water
reactor fuel, however, and the fuels under consideration are amenable to pyrochemical processing.
Two different pyrochemical processing methods are described in this paper, both involving an initial
chlorination of the irradiated transmuter fuel.
301

1. Introduction
The present concept for partitioning and transmutation of selected radionuclides in the US
accelerator transmutation of waste (ATW) program calls for the extraction of transuranic elements and
certain long-lived fission products from spent light water reactor (LWR) fuel and their subsequent
destruction in an accelerator-driven sub-critical system. Transmutation will be carried out in an
accelerator-driven sub-critical assembly, providing an intense flux of high-energy neutrons produced
by spallation reactions resulting from the impingement of high-energy protons on a liquid metal target.
The technology development roadmap [1] for this program called for approximately 1 450 tons of
spent LWR fuel to be treated per year, corresponding to some 14 tons of transuranics per year. The
form of the targets to be used for long-lived fission product (i.e. 99 Tc and 129 I) transmutation is
tentatively set as metallic technetium and sodium iodide, and the selection of the form of the
transuranic-bearing blanket fuel elements seems to be converging on a design that comprises a
dispersion of metallic transuranic elements in a metallic zirconium matrix. Uranium is excluded from
the fuel in order to preclude the formation of additional transuranic elements by neutron absorption.
The initial version of the fuel contains 23 weight percent transuranics and 77 weight percent
zirconium. More recent studies suggest that a 50-50 composition may be preferred. An alternative to
the metal dispersion/metal matrix fuel is a dispersion of transuranic nitrides or oxides in an inert
matrix such as molybdenum, stainless steel, or zirconium nitride. Regardless of fuel type selected, the
burn-up target for the transmuter blanket fuel is on the order of 300 MWd/kgHM, or approximately
30 atom percent burn-up of the TRU elements; the fuel must thus be processed to extract the
significant concentrations of newly-generated fission products and to recover the unfissioned
transuranics (as well as iodine and technetium) for recycle to the transmuter system. At this burn-up
level, throughput requirements for the processing of the transmuter blanket fuel are in the range
4-35 tons (heavy metal) per year, depending on the fuel composition, transmuter operating cycle, and
plant deployment scheme ultimately chosen.
2. Process selection
In the case of the metal dispersion/metal matrix fuel, the low throughput rate and the large
concentration of zirconium in the fuel material, combined with a desire to minimize the generation of
high-level radioactive wastes and secondary wastes, favor the use of a non-aqueous processing
method. Conventional electrorefining methods, as applied to metallic fast reactor fuels, were
considered inappropriate in this application because the high zirconium concentration would tend to
reduce throughput to an unacceptably low level and require extensive replication of equipment to
achieve throughput goals. The focus of process development has thus been on volatility processes,
specifically chloride volatility processes that retain the non-volatile transuranic elements in a chloride
salt that can be dealt with by means already developed for actinide extraction.
3. Primary chloride volatility process
The process concept under study, shown in Figure 1, involves the chopping of irradiated fuel
elements, followed by chlorination of the constituents of the fuel in a salt bath containing added
cadmium chloride that is formed by sparging chlorine gas through the liquid metal. The cadmium
chloride reacts with the metallic constituents of the fuel by reactions such as:
2CdCl2 + Zr = ZrCl4 + 2Cd,
3CdCl2 + 2Pu = 2PuCl3 + 3Cd, etc.
302

The metallic cadmium product of the chlorination reaction drops back to the bottom of the
chlorination vessel for subsequent reaction with chlorine. The temperature of the chlorination vessel is
then increased to volatilize ZrCl 4 . The primary purpose of the volatilization step is to remove the
matrix zirconium, in order to facilitate the subsequent extraction of the transuranic elements.
Experiments have shown the ready volatilization of ZrCl 4 at modest operating temperatures
(500-700 o C) whilst transuranic elements and the noble metal fission products are retained in the salt.
Initial tests with a LiCl-KCl salt were not successful because the ZrCl 4 tended to complex with KCl,
forming a non-volatile K 2 ZrCl 6 compound. Removing the KCl from the bath resulted in full
volatilization of the zirconium without loss of any of the transuranic elements. The recovered ZrCl 4
can be reduced to form zirconium metal, which can then be recycled to the fuel fabrication process.
The liquid cadmium at the bottom of the chlorination vessel contains the noble metal fission
products (NMFP). The cadmium can be separated from the residual salt by freezing the metal and
drawing off the salt bearing the transuranic chlorides and other fission product chlorides. Cadmium
can then be recovered for recycle by distillation, sending the noble metal fission products to an
extraction step where technetium is selectively removed and sent to transmutation target fabrication.
The other noble metal fission products are immobilized by alloying with zirconium and iron to
produce a highly durable metal waste form. This waste form will effectively immobilize technetium as
well, should it prove disadvantageous to extract and transmute technetium.
The residual salt containing the transuranic chlorides, strontium and caesium chlorides and
lanthanide chlorides is transferred to an electrowinning vessel, where the transuranic elements are
extracted by electrowinning. It appears that the iodine can also be extracted by electrowinning, but that
step has yet to be proven. The transuranic elements are sent to the fuel fabrication step, and the iodine
to transmutation target fabrication after converting it to sodium iodide. The recovery of TRU elements
has been demonstrated and it is planned to conduct experiments with iodine extraction during the year
2001.
Because the thermodynamic properties of technetium are not well known, it is not clear that the
technetium will be volatilized along with the zirconium, but the expectation is that it will remain in the
salt. Experiments are necessary to validate this assumption. The fate of iodine in this process is also
not clear, and there may be multiple routes for recovery of iodine for subsequent transmutation.
Testing of the complete flow-sheet in 2001 will serve to answer many of these questions.
4. Alternative fuel processing
Figure 2 shows a schematic flow-sheet for a process to treat those alternative transmuter blanket
fuels incorporating an inert matrix material that will not form a volatile chloride. This process would
be applicable to oxide or nitride fuel dispersions, with the added restriction in the case of nitride fuels
that the nitrogen may be fully enriched in 15 N, to preclude the formation of excessive amounts of 14 C
by the (n,p) reaction with the more common isotope of nitrogen, 14 N. This necessitates recovery and
recycle of the enriched nitrogen. The process would also be useful in treating LWR fuels designed
with inert matrices for burning plutonium. The initial part of the flow-sheet is much the same as that
for treatment of TRU-Zr alloy fuel described above, with the exception that no species are volatilized
in the course of the chlorination step. The noble metal fission products will not be chlorinated, and
they will be sent to the metal waste form along with the inert matrix material (molybdenum or
stainless steel), which will also not be chlorinated by CdCl 2 . The transuranic chlorides, rare earth
fission product chlorides, strontium and caesium chlorides, and the LiCl-KCl carrier salt are then
contacted with a dilute alloy of lithium in cadmium in a series of centrifugal contactors operating well
above the melting temperature of the salt mixture. By appropriate control of the lithium activity, the
303

transuranics will be reduced to the metallic state and transferred into the cadmium phase, while the
fission product chlorides remain in the salt phase. The fission product bearing salt is directed to the
ceramic waste form production process, resulting in a glass/sodalite composite waste form that has
been shown to be highly resistant to leaching by groundwater.
The transuranic elements in the cadmium phase are oxidized by sparging chlorine through the
melt, and the resultant transuranic chlorides (together with a trace amount of rare earth chlorides
formed from the small amount of rare earth fission products that extract with the transuranics in the
reductive extraction step) are separated from the metal phase. They are then placed in a LiCl-KCl salt
bath, where the transuranics are reduced with metallic lithium and recovered as metals. The metallic
transuranic elements are then sent to the fuel fabrication process for recycle to the transmuter. The
trace amount of rare earth contamination of the transuranics will not cause any neutronics problems in
the high neutron energy spectrum of the transmuter.
Figure 1. Schematic flow-sheet for the chloride volatility process for treatment of irradiated
transmuter blanket fuel of the general composition 30-50% TRU, 70-50% Zr
Irradiated Fuel
Chlorination
(in LiCl)
Zr chloride; TRU chlorides; Sr, Cs chlorides;
RE chlorides; NMFP (metals)
Volatilization
ZrCl 4
Reduction
Cd
Residual salt
+ cadmium
Cl 2
Recycle to
chlorination
step
Metal
Waste Form
Production
Cd
Other
NMFPs
Cadmium
Distillation
NMFPs
Technetium
Separation
Cd +
NMFPs
Iodine
Salt/Cadmium
Separation
Electrowinning
and
Iodine
Extraction
TRU (metallic)
LiCl
TRU chlorides
Sr, Cs chlorides
RE chlorides
Salt + FP chlorides
Zr metal
(to fuel fab.)
Tc
Target Fabrication
Fuel
Fabrication
Ceramic
Waste
Form
Production
304

Figure 2. Schematic flow-sheet for the chlorination process for treatment of irradiated transmuter
blanket fuel containing inert matrix material that will not form a volatile chloride. This process
could be used for treatment of fuels with, for example, stainless steel or ceramic matrices.
Irradiated
Fuel
TRUs, FPs,
matrix
Chlorination
TRU
Chlorides
FP Chlorides
LiCl - KCl
NMFP
Recovery
TRU
Chlorides
RE Chlorides
CsCl, SrCl 2
LiCl - KCl
Cd/Li Alloy
Reductive
Extraction
RE Chlorides,
CsCl, SrCl 2
LiCl - KCl
Ceramic
Waste
Form
Zeolite,
Glass
Off-gas
NM FPs
(including
Tc)
Cd Recycle
TRUs in Cd
Off-gas Iodine
Tc
TRU
Cl 2
Recovery
Recovery
Oxidation
Iodine
Transmutation
Target
Fabrication
Tc
NM FPs
Metal
Waste
Form
LiCl-KCl
Recycle
Li / K
TRU
Reduction
TRU Chlorides
trace RE Chlorides
LiCl-KCl
TRU Metal
trace RE
ATW Fuel
Material
Preparation
5. State of process development
Both processes described here have been demonstrated with simulated irradiated fuel containing
representative transuranic elements and non-radioactive fission product elements. The recovery of
technetium and iodine has not, however, been shown, and remains to be accomplished with the next
year. Experiments with irradiated fuels are necessary for final validation of process chemistry, but; the
absence of an operating fast reactor test facility in the United States imposes a delay on access to
irradiated fuel samples. In the meantime, tests will be conducted with fuel samples irradiated in
thermal reactors. Even though the burn-up levels in thermal reactors will be somewhat less than would
be achieved with fast reactor irradiations, the initial experiments will comprise a reasonable test of
process designs and should be adequate for flow-sheet adjustment and discrimination among
competing processes. Scale-up to prototype equipment sizes will occur within the next three to five
years, with initial tests to be performed with simulated irradiated fuel.
Acknowledgements
The authors wish to thank Dr. Karthick Gourishankar, Dr. James Willit, and Dr. Mark Williamson
for their contributions to the development of the processes described here. This work was performed at
Argonne National Laboratory, Argonne Illinois, operated by the University of Chicago for the
United States Department of Energy under Contract W-31-109-Eng-38.
305

REFERENCE
[1] A Roadmap for Developing Accelerator Transmutation of Waste (ATW) Technology – A Report
to Congress, DOE/RW-0519, October 1999.
306

R&D OF PYROCHEMICAL PARTITIONING IN THE CZECH REPUBLIC
Jan Uhlir
Nuclear Research Institute R ] plc
250 68 - R ], Czech Republic
Abstract
The Czech national research and development programme in the area of “Pyrochemical Partitioning”
is directed primarily on the development of the “front-end” part of the fuel cycle technology for the
molten salt reactor systems with a liquid fuel based on fluoride melts. The present research is directed
particularly on the development of suitable fluoride separation technology based on “fluoride volatility
method” the target of which is the removal of the uranium component from spent nuclear fuel and on
the research of the electroseparation procedures and further on the development of appropriate
construction materials and equipment for the fluoride molten salt technologies.
307

1. Introduction
Nuclear waste, especially spent fuel from nuclear reactors containing long-lived radionuclides
represented mainly by actinides and long-lived fission products, presents a common problem for all
countries operating nuclear power plants. Satisfactory solution of this problem is a factor limiting to a
considerable extent the further nuclear power industry development, namely in developed countries.
The present technical and technological sum of knowledge indicates that this problem could be largely
resolved by using the transmutation technology so that the nuclear power industry may become widely
acceptable for the public.
The use of transmutation reactors with fluoride salts based liquid fuel might be one of the
possible answers. In the near future, these systems might be conceived as critical ones and
successively as sub-critical accelerator driven reactor systems. In addition to the nuclear burning of
plutonium and minor actinides produced in the U-Pu cycle these reactor systems might consecutively
operate within the U-Th cycle as well.
The advantage of the molten salt transmutation reactors (MSTR) demonstrates itself above all in
connection with a continuous or at least quasi-continuous chemical separation process.
For such a compact coupling of MSTR with chemical reprocessing it will be very appropriate to
keep fuel in one chemical form, as far as possible, in the course of the entire fuel cycle. Accordingly,
if the MSTR fuel will be based on fluoride melt then the separation processes should also be based on
separation techniques from fluoride melt media. Pyrochemical and pyrometallurgical technologies
comply generally with this requirement [1].
2. Czech research and development programme
The Czech research and development programme in the field of pyrochemical and
pyrometallurgical separation is based first of all on the experience acquired in the past in the
development and realisation of a pilot-plant fluoride technology for the reprocessing of spent fuel from
the Russian BOR-60 fast reactor [2]. At present, this experience is utilised for the development of
suitable separation processes and technologies for the fluoride based MSTR fuel cycle.
Experimental and theoretical studies in the field of pyrochemical technology development for the
MSTR fuel cycle are oriented in particular to the following areas:
• Technological research in the field of the “Fluoride volatility” method directed at the
suitability verification of a technology for thermal or fast reactor spent fuel reprocessing,
which may result in a product the form and composition of which might be applicable as a
starting material for the production of liquid fluoride fuel for MSTR. Consequently, the
objective is a separation of a maximum fraction of uranium component from Pu, minor
actinides and fission products. Integral part of this research also is the flowsheeting research –
working out a proposal for a suitable technological flowsheet for treating spent fuel into a
form fitted for MSTR including the separation procedures before transmutation (front-end)
and separation processes after passage of fuel through the transmutor (back-end).
• Research on material and equipment for fluoride salts media connected with the experimental
programme on ADETTE technological loops. Objective of the programme is verification of
the new developed construction material for fluoride melts, development of selected devices
308

(first of all pumps) for fluoride melts and acquirement of practice in fluoride melts handling
and manipulation in greater amounts.
• Laboratory research on electro-separation methods in fluoride melts media in relation to the
study of their properties. The effort in this field is aimed first of all to the determination of
optimum conditions for uranium and fission product separations and to the selection of
suitable electrolyte composition based on fluoride salt mixture.
At present, the greatest attention is aimed to the realisation of the pilot-plant technology for the
separation of the uranium component from other spent fuels components by the modified “Fluoride
volatility” method. The other two areas mentioned are also intensively studied.
3. Technological research in the area of “fluoride volatility” method
Technological research of the “fluoride volatility” method can be ranked into the “front-end” area
within the MSTR fuel cycle. It is a separation process for reprocessing spent fuel from thermal or fast
reactors into a form suitable for the MSTR.
The main operations of this process are:
• Removal of the cladding material from spent fuel.
• Conversion of the fuel into a powder form (oxides) of a granulometric composition suitable
for the fluorination reaction.
• Fluorination of the fuel (the purpose of this operation is the separation of the uranium
component from plutonium, minor actinides and most of fission products).
• Uranium component purification.
The essential research activities are centred upon the technological verification of items 3 and 4
leading to the removal of the main portion of the uranium component from spent fuel. The proposed
technology is based on the spent fuel fluorination with gaseous fluorine in a flame fluorination reactor,
where the volatile fluorides are separated from the non-volatile ones, and on the subsequent
purification of the component by using technological operations of condensation and distillation and
also of sorption. As the fluorination reaction is suggested as a flame reaction, the necessary size of the
flame fluorination reactor (Figure 1) is a critical parameter influencing dimensions of the entire
experimental technological line.
According to the process design, up to 95-99% of uranium in the form of volatile UF 6 will be
removed from the non-volatile fluorides of plutonium (PuF 4 ), minor actinides and majority of fission
products. In this way, the component representing the greatest share of the spent fuel will be removed.
The technological research is closely associated with the “flowsheeting research”, the aim of
which is also to act as an unifying framework of the individual research activities, in addition to the
working out of the technological flowsheet. In Figure 2 a simplified flowsheet is presented of the
designed MSTR fuel cycle, the “front-end” of which is based on the “Fluoride volatility” method.
309

Figure 1. Experimental flame fluorination reactor
Figure 2. Simplified scheme of fuel cycle of MSTR
310

4. Research on material and equipment for fluoride technologies
Development of material and equipment for molten fluoride salts based technologies is connected
first of all with the ADETTE technological loop programme [3] (Figure 3). The main objectives of the
ADETTE loops experiments are:
• Testing of construction materials – corrosion research including the stress corrosion.
• Testing of welds.
• Research of fluoride melts thermohydraulics.
• Testing of pumps and valves for fluoride melt medium.
• Testing of measuring sensors and of the methods of measurement and control.
• Collection of data for the development and design of apparatuses for ADS technology with
fluoride melts.
Special attention has been paid to the testing of the new corrosion resistant alloy MONICR
SKODA. This high nickel content alloy is designated as a structural material for fluoride experimental
loops for the operation temperatures of approx. 700 o C.
The MONICR material is also intended for the construction of specific components of the loops
like molten salt pump (impeller vertical pump with a flange-mounted electric motor), control and
closing valves, molten salt storage tanks, heat exchangers etc.
Figure 3. Experimental molten salt loop ADETTE-0
311

5. Laboratory research on electroseparation methods
Laboratory research in the area of electro-separation methods is directed first of all at the
determination of optimum conditions for residual uranium and fission product separation from fluoride
melt and further on the selection of a suitable composition of the electrolyte based on fluoride salts
mixture. As the fluoride melt should be able to dissolve sufficient amounts of plutonium and minor
actinide elements, mixtures of LiF–NaF and LiF–NaF–KF type are in the foreground of interest. The
assumed ability of sodium and potassium fluorides to form co-ordination compounds with
transuranium element fluorides and so to significantly increase their solubility in the melt was the
reason for choosing the mixtures mentioned.
The research programme in this area is further directed to the determination and study of selected
physicochemical properties of fluoride melts, particularly to the:
• Solubility of lanthanides in molten fluorides.
• Standard redox potentials of individual elements.
• Melting points of the molten salt mixtures.
• Density and viscosity of the supporting fuel matrices.
• Data on corrosion resistance of structural materials.
6. Conclusion
Research and development in the area of “Pyrochemical Partitioning” is carried out in the
Czech Republic as a component of the national P&T programme within the framework of the
“Transmutation” consortium. The programme is funded first of all by the Ministry of Industry and
Trade and by the Radioactive Waste Repository Agency. The participants of the programme are, in
addition to the NRI R ], SKODA Nuclear Machinery, Energovyzkum Brno, Nuclear Physics Institute
of the Czech Academy of Sciences and Faculty of Nuclear Sciences and Physical Engineering of the
Czech Technical University Prague.
Experimental research work related to the development of pyrochemical technologies is
concentrated mainly in the Fluorine Chemistry Department of the NRI R ].
The Czech national conception in the area of P&T research issues from the national power
industry programme and from the Czech Power Company intentions of the extensive utilisation of
nuclear power in our country. The Czech Republic, as a relatively small country, has an
understandable interest in a wide integration into the solution of problems associated with spent
nuclear fuel in the international context, first of all in co-operation with the EU member countries.
Involvement into the 5th Framework Programme EC is an evidence of it, viz. in the area of
“Pyrometallurgical Partitioning” where, for example, the PYROREP project is complementing very
well with the research activities within the national programme framework.
312

REFERENCES
[1] Uhlir J., Pyrochemical Reprocessing Technology and Molten Salt Transmutation Reactor Systems,
OECD/NEA Workshop on Pyrochemical Separations, Avignon, France, March 14-16, 2000.
[2] Uhlir J., An Experience on Dry Nuclear Fuel Reprocessing in the Czech Republic, Proc.
5th OECD/NEA Int. Information Exchange Meeting on Actinide and Fission Product
Partitioning and Transmutation, Mol, Belgium, November 25-27, 1998, EUR 18898 EN,
OECD/NEA (Nuclear Energy Agency), Paris, France, 1999.
[3] Chochlovsky I et al., Molten Fluoride Salt Loop ADETTE-0 Experimental Programme,
ADTTA99 – 3rd Int. Conf. on Accelerator Driven Transmutation Technologies and
Applications, Prague, Czech Republic, June 7-11, 1999.
313

DEMONSTRATION OF PYROMETALLURGICAL
PROCESSING FOR METAL FUEL AND HLW
Tadafumi Koyama, Kensuke Kinoshita, Tadashi Inoue
Central Research Institute of Electric Power Industry (CRIEPI)
2-11-1 Iwado-kita, Komae, Tokyo 201-8511, Japan
Michel Ougier, Jean-Paul Glatz, Lothar Koch
European Commission Joint Research Center
Institute for Transuranium Elements (JRC-ITU)
Postfach 2340, 76125 Karlsruhe, Germany
Abstract
CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the
capability of this type of process for separating actinide elements from spent fuel and HLW.
Experiments on pyro-processing of un-irradiated metal alloy fuel (U-Pu-Zr or U-Pu-MA-RE-Zr) by
molten salt electrorefining and molten salt/liquid metal extraction will be carried out. The necessary
equipment is installed in a new experimental set-up at JRC-ITU. The stainless steel box equipped with
telemanipulators is operated under pure Ar atmosphere and prepared for later installation in a hot cell.
In first electrorefining tests, U (about 10 g) and Pu (about 5 g) were deposited on a solid and a liquid
Cd cathode respectively. Preliminary experiments on molten salt/liquid metal extraction in countercurrent
batch extraction systems with REs were conducted in CRIEPI. The results showed good
separation efficiency in 3 batch extraction stages.
315

1. Introduction
The increasing interest in pyrometallurgy after the selection a few decades ago of oxide fuel and
aqueous reprocessing as the fuel cycle reference can be attributed to the drastic change of boundary
conditions around the nuclear fuel cycle in the world. The former mission of the fuel cycle was to
recover Pu as an important fissile material for the fast breeder reactor, however at present positive
credit from recovered Pu can hardly be expected. Today’s main emphasis is put on a maximal cost
reduction of the fuel cycle. Furthermore recovery of long-lived nuclides becomes a new requirement,
since geological disposal of high level waste (or once-through fuel) is facing large difficulties to get
public acceptance. Recovery of long-lived nuclides means to use of various reactor systems for
transmutation, resulting in new requirements to reprocess different fuel types e.g. MOX, metal fuel,
nitride fuel, high burn-up fuel, etc. These new requirements may result in a different choice for future
fuel cycle technology. Pyrometallurgical processing is one of the most attractive alternatives to meet
these requirements. The requirement for product purity being much less stringent, the recovery of
minor actinides (MA: Np, Am, Cm) will take place simultaneously with plutonium due to the
thermodynamic properties of molten salt media. The recovery of MA allows the reduction of TRU
wastes, and decrease at the same time the risk of nuclear proliferation. The molten salt media also
have two important advantageous properties as a solvent material in nuclear processing. The radiation
stability of molten salt allows the processing of spent fuels of high radioactivity (e.g. spent fuel with
short cooling time) without any increase of solvent waste. Since molten salt is not a neutron moderator
such as water is, comparatively large amount of fissile material can be handled in the process
equipment, i.e. experimental facilities are compact and economical.
The Central Research Institute of Electric Power Industry (CRIEPI) investigated these promising
features in pyroprocessing according to an information exchange with US-EPRI in 1985. The
feasibility of pyrometallurgy to separate/recover actinides from spent nuclear fuel or high level waste
(HLW) has been started on 1986 [1,2,3]. As a joint study with US-DOE, CRIEPI had participated in
the Integral Fast Reactor (IFR) Program of Argonne National Laboratory (ANL) from 1989 to 1995 in
order to study the pyrometallurgical technology development [4,5] and to demonstrate the pyroprocess
of spent metal fuel [6]. In parallel, the measurement of thermodynamic properties of actinides as well
as pyrometallurgical partitioning of TRUs from simulated HLW had been carried out by CRIEPI in
collaboration with the Missouri University and Boeing North American [7,8]. In the course of the
study, the feasibility of pyrometallurgical processing to recover/separate actinides from spent metal
fuel or HLLW was confirmed by the results of experiments with unirradiated TRU materials and
theoretical calculations based on measured thermodynamic properties. The demonstration of TRU
recovery from spent metal fuel (or even from unirradiated ternary alloy fuel) could however, not be
realised because of a sudden cancellation of the IFR program. The Institute for Transuranium
Elements (JRC-ITU) has studied since many years the capacities of aqueous processing as for the
separation of TRUs from HLW [10]. The CRIEPI and JRC-ITU collaboration to study metal target
fuels for the transmutation of TRU [9] has led to a new joint study on pyrometallurgical processing.
This study will demonstrate the feasibility of pyrometallurgical processes for separating actinide
elements from real spent fuel and HLW, also in view of a rational evaluation of future fuel cycle
technology. Furthermore this study has also been included in a project of the European 5th Framework
program, where CRIEPI and ITU are a joint partner in an international network. In this paper, the
current status as well as the whole test plan of this project will be reported.
2. Experimental plan of the joint study
The first phase of the joint study will be carried out from 1998 to 2004, where three different
stages are to be carried out as shown in Figure 1.
316

Figure 1. Experimental test plan of the joint study
1998 `
Design & Construction
MA: mainor actinides
HLLW
Alloy fuel
U-Pu-MA-RE-Zr
Electrorefining
recovered products
U,Pu,MA-Cd
sim.
HLLW
Denitration
products
Spent molten salt
(MA contained)
products
Pu,MA
{Cd
Chlorination
products
Molten Salt Extraction
Molten Salt Extraction
simulated HLLW
chlorinated salt
Cd Distillation
Cd distillation
Stage
Development and
Installation of
Experimental Apparatus
Development of metal fuel
reprocessing
U,Pu,MA
U,Pu,MA
Recovered Products
Recovered Products
Demonstration of TRU recovery from HLLW
The first stage is the development and installation of the experimental apparatus. In this stage, an
argon atmosphere hot cell equipped with an electrorefiner dedicated for pyrometallurgical experiments
is developed. The second stage is the development of metal fuel reprocessing, where recovery of
actinides from unirradiated metal alloy fuel such as U-Pu-Zr and U-Pu-MA-RE-Zr are to be carried
out. The metal alloy fuel is first submitted to an electrorefining step followed by a reductive extraction
process of the molten salt electrolyte to recover residual actinides and to separate them from
lanthanides. The recovered TRU-Cd metal will be treated by distillation to separate Cd from TRUs.
The third stage is the demonstration of TRU recovery from HLW where pyrometallurgical partitioning
is to be demonstrated on actual HLW. In this stage, reductive extraction, Cd distillation and
chlorination are first tested with simulated materials. For the experiments on actual HLW, the whole
system will be moved inside a lead shield. The actual HLW will be converted into oxide and
afterwards into chlorides in a Cl 2 gas flow. The obtained chloride salt mixture will be used for the
reductive extraction process described previously. The reprocessing of metal fuel irradiated in the
French PHENIX reactor will be carried out in the last phase of the project.
3. Development and installation of the Ar-atmosphere hot cell system
The experimental apparatus was newly designed and fabricated for this study and was conceived
for later installation in the hot cell system. The apparatus consists of a stainless steel box to be
installed in a 15 cm-thick lead shielding. The stainless steel box was first installed in an alpha
laboratory Figure 2.
317

Figure 2. Stainless steel box with Ar purification unit
The box is operated in a pure Ar gas atmosphere, continuously purified. The airlock system for
the introduction or extraction of material is separately flushed by Ar. A so called “La Calhène”
container is used for the transport. The box is equipped with a vertical heating well (150 mm in
diameter, 600 mm in depth) consisting of an inconel liner and a stainless steel tube. The well sited on
the bottom of the box is heated from outside by a cylindrical resistance heater which can be heated up
to 1 273 K. Double sleeves with intermediate Ar flushing were employed for the telemanipulators in
order to reduce diffusion of oxygen. After many modifications, the stainless steel box is now in an
operational condition, with an oxygen and moisture concentration less than 10 ppm.
4. Electrorefining process
4.1 Development of electrorefiner
The electrorefiner is a key part of pyrometallurgical reprocessing, since the fuel dissolution as
well as the actinide refining is to be done in this step. An electrorefiner was newly designed and
fabricated in CRIEPI based on experience gained in various types of experiments. It should be noted
that the design concept of this electrorefiner is to demonstrate the separation yield and recovery yield
of metal fuel reprocessing. The electrorefiner consists of three electrodes and a liquid Cd pool covered
by a molten LiCl-KCl eutectic mixture. It was shipped to JRC-ITU, and installed in the stainless steel
box as shown in Figure 3.
318

Figure 3. Electrorefiner installed in a stainless steel box (lift-up position)
The electrorefiner cell of 100 mm × 130 mm is hung on a metal flange equipped with cathode,
anode, stirrer, reference electrode, sampling etc. Metal alloy fuel previously fabricated at ITU in a
joint study with CRIEPI on transmutation of TRU targets is charged into a metal basket working as an
anode. The pool of liquid Cd below the molten salt works as an anode, or just as a receiver for the
noble elements. The cathode assembly of the electrorefiner uses either a solid iron cathode for
U recovery or a liquid metal cathode for TRU recovery. The solid iron cathode (∅18mm) with a spiral
groove can be rotated during electrodeposition to achieve a better recovery [2]. The liquid metal itself
will be stirred by means of a ceramic stirrer submerged in the liquid metal cathode (surface
area = 8 cm 2 ) in order to avoid formation of U dendrites that will hamper the deposition of Pu [2]. The
electrode potentials are monitored by a Ag/AgCl reference electrode known for its reliability. The
concentration of relevant elements in each phase will be measured by the chemical analyses.
4.2 Electrorefining experiments with U and Pu
The electrorefiner was loaded with approximately 1 000 g of LiCl-KCl eutectic salt and 500 g of
cadmium. The whole system was heated up to the operation temperature of 773 K to melt both phases.
Depleted U metal was then charged in the anode basket followed by addition of CdCl 2 to oxidise some
U metal to UCl 3 , necessary to facilitate the electrotransportation in the molten salt electrolyte. At the
concentration of U 3+ of about 1 wt%. in the LiCl-KCl salt, notable polarisation was not observed at a
current of 1A. Hence electrodeposition of uranium on the solid iron cathode as carried out at a
constant current of 500 mA and 1A, respectively. Figure 4 shows the dendritic uranium deposit
obtained at 500 mA.
319

Figure 4. U deposit on solid cathode (12g)
The expected amount of uranium metal calculated from the coulomb passed was 12 g. The
amount of U deposited will be determined from the chemical analyses of the metal deposit. The
electrodeposition of uranium on the liquid Cd cathode was also carried out at a constant current of
200 mA for 1.5 hours resulting in the deposition of about 1 g U in 80 g Cd. After these experiments,
the residual U in the molten salt was recovered by the “drow-down electrorefining” where U 3+ is
reduced to metal form at the cathode while metal Ce is oxidised at the anode. The chemical analysis of
the treated salt is still under way, but the colour of molten salt turned from purple (U 3+ ions) to white
after drow-down electrorefining indicating a high efficiency of the reaction.
About 45g of plutonium metal was then charged with new 1 000g LiCl-KCl eutectic salt in the
crucible. About 1 mol% of PuCl 3 was formed in the salt by adding the equivalent amount of CdCl 2 .
Then electrodeposition of Pu in liquid Cd cathode was carried out at a constant current of 500 mA for
4 hours. About 5 g of Pu was recovered in the liquid Cd cathode of 85 g. Figure 5 shows the liquid Cd
cathode just after the electrodeposition.
Figure 5. Pu deposit in liquid Cd cathode (~5 g)
The deposit was easily removed from LCC crucible as shown in Figure 6.
320

Figure 6. Recovered Cd-Pu ingot and AlN crucible
In the next step, unirradiated U-Pu-Zr metal alloy fuels will be charged in the anode basket.
Plutonium recovery into liquid Cd cathode will be performed after several fuel treatments to recover
only U onto solid iron cathodes. FP simulating elements such as lanthanides will next be added to the
system, and electrorefining of U-Pu-MAs-REs-Zr will be performed to simulate the processing of
irradiated metal alloy fuel. The TRU recovery into the liquid Cd cathode will then be carried out after
U recovery on the solid cathode. The cathode products will be analysed to determine the recovery
yields and the decontamination factors. The results will provide data on the up to now unknown
behaviour of TRUs during electrorefining as well as the operation sequence to maximise the TRU
recovery, that will be crucial for the experiments with real irradiated fuel.
5. Reductive-extraction process
5.1 Process description
In the pyro-partitioning process a reductive-extraction technique in LiCl-KCl/Bi or LiCl-KCl/Cd
system is the key step for the recovery and separation of actinides. As explained in the former report [8],
a multiple batch extraction experiment was successfully carried out with a high recovery yield of TRUs
and high separation efficiency between TRUs and FPs. On the other hand, the separation efficiency by
means of the countercurrent extraction will be much higher than that by means of the multiple batch
extraction. Hence reductive extraction by means of a counter-current batch method should be tested with
TRUs. In this joint research, the reductive-extraction experiments will be carried out after the
electrorefining experiments.
5.2 Preliminary experiments with lanthanides
Before starting experiments using TRUs, the experiment of three stages counter-current
extraction was carried out by batch system using REs. Ce, Gd and Y were used as substitution
elements for U, Am (TRUs) and Nd (REs), because the relationship of distribution coefficients among
these three RE in LiCl-KCl/Bi system were roughly similar to that of U, Am and Nd. Figure 7 shows
the schematic flow of the counter-current batch extraction method.
321

Figure 7. Schematic flow of the countercurrent bath extraction method
Stage #1 Stage #2 Stage #3
Step #n
Step #n+1
Step #n+2
Step #n+3
Step #n+4
Salt with
TRUs, REs
TRUs
recovered
into Bi
Salt with
TRUs, REs
TRUs
recovered
into Bi
Bi-Li
Treated salt
Bi-Li
Treated salt
An initial salt phase is introduced into stage #1 and recovered as waste from stage #3. The Bi
phase with Li is introduced into stage #3 and recovered as product from stage #1.
Figure 8 shows the separation factors of Ce and Gd against Y obtained during three stages at
773K.
Figure 8. Separation factors of Ce and Gd vs. Y
1.00E-02
Ce
Gd
1.00E-03
1.00E-04
1.00E-05
1.00E-06
1 2 3 4 5
Step number
322

The separation factor (SF(M)) of metal M (Ce or Gd), is defined by:
SF(M) = {X(M)/Y(M)}/{X(Y)/Y(Y)}
where X(M), X(Y) are the concentrations of MCl 3 and YCl 3 in the salt at stage #3, respectively, and
Y(M), Y(Y) are the concentrations of M and Y in Bi at stage #1, respectively.
The separation factors of Ce and Gd vs. Y in the single stage extraction previously measured are
1.4 × 10 -3 and 1.7 × 10 -2 , respectively [12], while those in the counter-current stages obtained here
were given as 5.9 × 10 -6 and 3.4 × 10 -4 , showing far better separation between TRUs and REs.
Figure 9 shows the recovery yield of each RE.
Figure 9. Recovery yields of Ce, Gd and Y
100
99
98
97
96
95
94
93
92
91
90
Ce
Gd
Y
1 2 3 4 5
Step number
10
9
8
7
6
5
4
3
2
1
0
The recovery yield (RCM) of metal M (Ce, Gd or Y), is defined by:
R(M) = W 1,Bi (M)/{W 1,Bi (M) + W 3,Salt (M)}
where W 1,Bi (M) is the amount of M recovered in Bi at stage #1 and W 3,Salt (M) is the amount of M
recovered in salt at stage #3.
The obtained recovery yields R for Ce, Gd and Y were >99.9%, 98% and 99.9%, >99% and

actinide elements from spent (metallic) fuel and HLW and should provide important data in view of a
rational selection of future nuclear options.
A successful installation of the equipment in a new experimental set-up was achieved at JRC-
ITU. Electrorefining tests on U and Pu using solid cathodes and a liquid Cd cathode prove the
operational capabilities of the facility. The next experiments will be on electrorefining of unirradiated
metallic U, Pu, Zr fuels containing MA and REs.
Furthermore reductive extraction of TRUs from molten salt will be demonstrated in countercurrent
batch extraction systems developed in CRIEPI. The preliminary experiments using Ce, Gd and
Y as stand-in elements for U and Am showed good separation efficiency from REs (Nd) in 3 batch
extraction stages.
REFERENCES
[1] T. Inoue, T. Sakata, M. Miyashiro, H. Matsumoto, M. Sasahara, N. Yoshiki, Development of
Partitioning and Transmutation Technology for Long-lived Nuclides, Nucl. Technol., Vol. 93,
206 (1991).
[2] T. Koyama, M. Iizuka, H. Tanaka, M. Tokiwai, An Experimental Study of Molten Salt
Electrorefining of Uranium Using Solid Iron Cathode and Liquid Cadmium Cathode for
Development of Pyrometallurgical Reprocessing, J. Nucl. Sci. Technol., Vol. 34(4), 384-393 (1997).
[3] T. Inoue, H. Tanaka, Recycling of Actinides Produced in LWR and FBR Fuel Cycle by Applying
Pyrometallurgical Process, Proc. Global’97, Yokohama, Japan, Oct. 5-10, 1997.
[4] Y.I. Chang, The Integral Fast Reactor, Nucl. Technol., Vol. 88,129 (1989).
[5] T. Koyama, T.R. Johnson and D.F. Fischer, Distribution of Actinides between Molten Chloride
Salt/Liquid Metal Systems, J. Alloys Comp., Vol. 189, 37 (1992).
[6] M. Lineberry, H.F. McFarlane, and R.D. Phipps, Status of IFR Fuel Cycle Demonstration, Proc.
Global’93, Seattle WA, Sept. 12-17, 1993, p. 1066.
[7] Y. Sakamura, T. Hijikata, T. Inoue, T.S. Storvick, C.L. Krueger, J.J. Roy, D.L. Grimmet,
S.P. Fusselman, and R.L. Gay, Measurement of Standard Potentials of Actinides (U, Np, Pu,
Am) in Licl-Kcl Eutectic Salt and Separation of Actinides from Rare Earths by Electrorefinig, J.
Alloy. Comp., Vol. 271-273, 592-596 (1998).
[8] K. Kinoshita, T. Inoue, S.P. Fusselman, D.L. Grimmett, J. Roy, R.L. Gay, C.L. Krueger,
C.R. Nabelek, and T.S. Storvick, Separation of Uranium and Transuranic Elements from Rare
Earth Elements by Means of Multistage Extraction in Licl-Kcl/Bi System, J. Nucl. Sci. Technol.,
36, 189-197 (1999).
324

[9] T. Inoue, M. Kurata, L. Koch, J.-C. Spirlet, C.T. Walker and C. Sari, Characterisation of Fuel
Alloys with Minor Actinides, Trans. Am. Nucl. Soc., Vol. 64, 552 (1991).
[10] O. Courson, R. Mambeck, G. Pagliosa, K. Roemer, B. Saetmark, J.P. Glatz, P. Baron, C. Madic,
Separation of Minor Actinides from Genuine HLLW Using the DIAMEX Processes, Proc.5th
International Information Exchange Meeting, Mol, Belgium, 25-27 Nov. 1998, EUR 18898 EN,
p. 121, OECD Nuclear Energy Agency, Paris, France, 1999.
[11] K. Kinoshita, M. Kurata, K. Uozumi, T. Inoue, Estimation of Material Balance in
Pyrometallurgical Partitioning Process for Trus from HLLW, Proc. 5th International Information
Exchange Meeting Mol, Belgium, 25-27 Nov. 1998, EUR 18898 EN, p. 169, OECD Nuclear
Energy Agency, Paris, France, 1999.
[12] M. Kurata, Y. Sakamura, T. Hijikata, K. Kinoshita, J. Nucl Mter., 1995, 227, 110.
325

DEVELOPMENT OF PLUTONIUM RECOVERY PROCESS
BY MOLTEN SALT ELECTROREFINING WITH LIQUID CADMIUM CATHODE
Masatoshi Iizuka, Koich Uozumi, Tadashi Inoue
Central Research Institute of Electric Power Industry
2-11-1, Iwado-kita, Komae, Tokyo 201-8511, Japan
Takashi Iwai, Osamu Shirai, Yasuo Arai
Japan Atomic Energy Research Institute
Oarai, Higashi-Ibaraki, Ibaraki 311-1394, Japan
Abstract
The effects of electrochemical conditions on the behaviour of plutonium and adequate conditions for
recovery at liquid cadmium cathode (LCC) used in pyrometallugical reprocessing were studied with
small, not stirred electrodes. Cathodic current density adequate for plutonium collection at LCC was
considered to be controlled by diffusion plutonium ion in molten salt and proportional to its
concentration. It was shown that plutonium collected at the LCC beyond saturation formed
intermetallic compound PuCd 6
and accumulated at the bottom of the LCC. This behaviour of
coexisting americium was reasonably explained by the local equilibrium model between plutonium
and americium at the surface of the LCC. The plutonium collection rate in practical electrorefining
equipment estimated by extrapolation of experimental results was satisfactorily high in designing
practical equipment and process.
327

1. Introduction
Metallic fuel cycle which consists of a metal (U-Zr or U-Pu-Zr) fuelled fast reactor and
pyrometallurgical reprocessing has been proposed originally by Argonne National Laboratory (ANL)
as an innovative nuclear fuel cycle technology [1]. The metallic fuel cycle has an excellent safety
potential aspect originating from high thermal conductivity of the metal fuel [2]. It also has economic
advantage because a pyrometallurgical reprocessing plant is estimated to be smaller than conventional
aqueous reprocessing plants due to fewer steps and smaller equipments [3].
The main step in the pyrometallurgical process is molten salt electrorefining [4], where the
actinide elements are recovered and decontaminated from the fission products. Figure 1 shows a
schematic flow of the normal operation of this electrorefining step. The spent fuel is cut into small
pieces, loaded in a steel basket, and immersed into molten chloride electrolyte. Almost all of the
actinide elements in the spent fuel are anodically dissolved. Noble metal fission products are left in
the anode basket by controlling the anode potential. Chemically active fission products such as alkali,
alkaline earth, and rare earth metals exchange with the actinide chlorides in the electrolyte and
accumulate in the molten salt in the form of their chlorides. Two kinds of cathodes are used to obtain
different streams of products. One is a solid cathode made of iron and the other is a liquid cadmium
cathode (LCC). At the solid cathode, uranium is selectively collected because the free energy change
of chloride formation for uranium is negatively less than those of the other actinide elements. On the
other hand, free energy changes of the actinide elements are close to each other at LCC because the
transuranium elements (plutonium, neptunium, americium and curium) are stabilised in the LCC due
to their very low activity coefficients in liquid cadmium [5,6]. Therefore, transuranium elements can
be collected at LCC together with uranium.
Figure 1. Schematic flow of routine operation of the electrorefining step
The use of LCC is the most important technology in the pyrometallurgical process, where
plutonium is recovered, roughly separated from uranium, and decontaminated from fission products.
Because performance of LCC significantly influences the feasibility of the pyrometallurgical
reprocessing, ANL studied plutonium recovery with LCC in depth with laboratory scale equipment
[7-10]. Central Research Institute of Electric Power Industry (CRIEPI) has also reported on the LCCs,
especially focusing on the formation of dendritic uranium deposit [11,12]. In those studies, it was
shown that stirring in cathode cadmium with vertical paddles is effective to restrain growth of the
328

uranium dendrite and that uranium can be collected into LCCs at a cathodic current density of
0.2 A/cm 2 up to about 10 wt% in the cathode without dendrite formation [12]. In addition to the
uranium studies, we have launched a joint research program with Japan Atomic Energy Research
Institute (JAERI) on pyrometallurgical processes for the actinide elements. In 1999, a plutonium
electrorefining apparatus equipped with a LCC assembly was fabricated and installed in a glove box.
In this study, fundamental plutonium electrotransport experiments were carried out in order to
understand the effects of electrochemical conditions on the behaviour of plutonium at LCC preceding
investigation of engineering factors like stirring method.
2. Experiment
2.1 Apparatus
All the experiments were carried out in a high purity argon atmosphere glove box. Both oxygen
and moisture levels in the atmosphere were kept less than two ppm during the tests. Figure 2 is a
schematic view of the experimental apparatus. Inner diameter and depth of the container for molten
salt were 124 mm and 120 mm, respectively. The amount of lithium chloride-potassium chloride
(LiCl-KCl) eutectic mixture loaded in this container was about 1 200 grams. Under the molten salt
electrolyte, a liquid cadmium layer was placed and used as an anode which supplied plutonium in
electrotransport experiments. The amount of the anode cadmium was about 1 400 grams. The salt and
anode cadmium were heated with an electric furnace and the temperature of the system was kept to
773 ± 1 K.
Figure 2. Schematic view of experimental apparatus
329

The electrorefining apparatus and the cathode assembly were originally designed to accommodate a
LCC of 50 mm outer diameter, which would be stirred to facilitate the mass transfer of plutonium. In
this study, however, much smaller cathodes were used because the study aimed to understand the effects
of fundamental electrochemical conditions on the behaviour of plutonium at LCC proceeding to
investigate engineering factors like stirring method. The size of the cathode crucible used in this study
was 9 mm in diameter and 16 mm in depth. About 3 to 5 grams of cadmium was loaded in the crucible.
A silver-silver chloride (1 wt% AgCl in LiCl-KCl) reference electrode contained in a thin Pyrex glass
tube was used as a reference electrode.
2.2 Chemicals
The chlorides (LiCl-KCl, CdCl 2
and AgCl) were purchased from Anderson Physical Laboratory.
Because their purity was no less than 99.99% and their moisture content was extremely low, they
were used with no additional purification procedure. Cadmium metal of more than 99.9999% purity
for the anode and the cathode was purchased from Rare Metallic Corporation. Because the cadmium
had been packed under vacuum just after production to avoid oxidation by the air, it was not washed
or polished before use. PuO 2
used in this study contained about 2% of americium which was
generated by (n, γ) reaction of Pu 239 and β-decay.
PuCl 3
was prepared in the following two steps. (a) carbothermic reduction of PuO 2
to produce
PuN [13], and (b) exchange reaction between PuN and cadmium chloride (CdCl2) in LiCl-KCl. Pu in
the liquid cadmium anode layer was prepared by reduction of PuCl 3
by addition of Cd-Li alloy. After
these procedures were completed, concentrations of plutonium in the molten salt and in the liquid
cadmium anode were 2.28 wt% and 1.72 wt%, respectively.
2.3 Analytical procedures
EG&G Princeton Applied Research potentio/galvanostat Model 273A and EG&G 270/250
Research Electrochemistry Software were used for both electrochemical measurement and constantcurrent
electrotransport. The concentrations of plutonium and cadmium in the molten salt were
determined by inductively coupled plasma-atomic emission spectroscopy (ICP-AES) of the samples.
Cathode products were analysed by scanning electron microscope (SEM) and electron probe
microanalyser (EPMA). An X-ray diffract meter (XRD) was also used to determine the chemical
form of the cathode deposit.
3. Results and discussion
In order to understand the relationship between the behaviour of plutonium at LCC and its
reduction rate, electrotransport experiments were carried out at various cathodic current densities.
Major results are summarised in Table 1 with experimental conditions.
330

Table 1. Conditions and results of Pu electrotransport experiments with LCCs
Run
no.
Pu concentration
in molten
(wt %)
Cathodic current
density
(mA/cm 2 )
Electrotransporttime
(s)
Electricity
passed in
experiment
(C)
Initial amount
of cathode
(g)
Increase of
cathode
weight
(g)
Collection
efficiency
(%)
Final Pu
concentration
in cathode
(wt %)
1 2.28 33 12 000 240 4.036 0.1983 100 4.68
2 2.11 41 11 870 297 2.918 0.245 105 7.75
3 2.28 50 7 800 234 4.899 0.1555 80.6 3.08
4 2.11 66 6 500 260 3.406 0.0023 1.07 0.07
5 4.6 66 7 200 288 4.0287 N/A N/A N/A
331
6 4.6 82 5 400 270 4.0056 N/A N/A N/A
7 4.6 100 5 400 324 4.024 N/A N/A N/A

3.1 Time course of LCC potential and plutonium recovery efficiency
Changes of LCC potential in the electrotransport tests at plutonium concentration of about 2 wt%
in the molten salt are shown in Figure 3. At cathodic current density of 33 to 41 mA/cm 2 , cathode
potential was kept between -1.4 V and -1.55 V after a slight shift to the lower direction at the
beginning. In this range of the potential, reduction of plutonium followed by dissolution to liquid
cadmium or formation of an intermetallic compound is expected to occur from the result of the CV
measurement [14]. The moderate change of the cathode potential indicated that plutonium was
smoothly collected in the LCC without abrupt growth of solid phase at the interface. Collection
efficiencies for plutonium calculated from increase of cathode weight and electric charge passed
between the electrodes were nearly 100% in these conditions. This result supports the above
consideration. Figure 4 is a photograph of the cathode cadmium ingot taken out of the crucible in
Run 2 where plutonium was collected up to 7.75 wt% in cadmium at cathodic current density of
41 mA/cm 2 . Although there was a little inequality on the surface of the LCC, no growth of dendritic
deposit was found.
Figure 3. Change of LCC potential in Pu electro-transport tests
Figure 4. Cathode Cd ingot obtained after Pu electro-transport test
at cathodic current density of 41 mA/cm 2
332

Cathode potential went down to -1.65 V at cathodic current density of 50 mA/cm 2 . The solidified
salt on the top of the cathode cadmium in Run 3 was white although the bulk salt containing about
2 wt% of plutonium is usually light blue in colour. Collection efficiency for plutonium was about
80%, a little lower than in the preceding case. These results indicate that lithium in the electrolyte was
reduced at the LCC at -1.6 V and that the reduced lithium reacted with plutonium tri-chloride near the
cathode after the electrotransport. Although lithium forms a very stable chloride which has more than
0.6 V lower standard potential than that of plutonium at an inert electrode, its metal is stabilised in
liquid cadmium due to the very low activity coefficient [15]. Figure 5 shows a CV measured for blank
LiCl-KCl with a liquid cadmium electrode. It can be seen that reduction current for lithium increases
from about -1.6 V, suggesting the validity of the above consideration.
Figure 5. Cyclic voltammogram for blank LiCl-KCl with liquid cadmium electrode
When cathodic current density was increased to 66 mA/cm 2 , cathode potential descended to
-1.7 V at first and subsequently ascended in two steps. After the experiment, the cathode was visually
inspected. The lower part of the alumina insulator sheath of the electric lead for the LCC had turned
black and a deposit with metallic gloss was found on that region. XRD analysis showed that the major
portion of this deposit was PuCd 6
. At such low potential and higher cathodic current density, it is
expected that the reduction rate and the LCC surface concentration of lithium were increased and that
the lithium reacted with the alumina sheath. It is very likely that the alumina sheath was wetted much
more easily with liquid cadmium due to the reaction with lithium. This is considered the reason why
the alumina sheath worked as a thin LCC and PuCd 6
was deposited there. The very low collection
efficiency for plutonium (25%) in Run 3 should be due to the PuCd 6
formation out of the LCC.
333

3.2 Plutonium concentration dependence of optimum cathodic current density
Electrotransport experiments were carried out at higher plutonium concentration in molten salt in
order to investigate the effect of plutonium concentration on the reduction behavior of plutonium at
LCC. The concentration of plutonium in the molten salt was adjusted to 4.6 wt% by the procedure
described above. The results were compared with those at lower concentrations in Figure 6(a) to (c).
In Figure 6(a) and (b), it is clear that the overall trends of the charts at approximately same ratio
between plutonium concentration in the molten salt and cathodic current density can be closely
correlated. It indicates that cathodic current density at which plutonium can be smoothly collected
into LCC is proportional to the plutonium concentration in molten salt at least in the range of this
study. A distinct difference was found in two charts in Figure 6(c). In Run 3, cathode potential went
down to -1.65 V and lithium in the solvent was considered to be reduced. In Run 7, on the other hand,
cathode potential was kept higher than -1.55 V at which it was expected that plutonium was
selectively reduced at the LCC. As mentioned above, the ratio between cathodic current density and
the plutonium concentration in molten salt was a little higher in Run 3 (24 mA/cm 2·wt%-Pu)
compared to Run 7 (22 mA/cm 2·wt%-Pu). It is thought that a limitation in the mass transfer rate of
plutonium by diffusion in molten salt in a not stirred system lies between those conditions.
Conversely, selective and smooth plutonium reduction at the LCC would be expected at a cathodic
current density proportional to the concentration of plutonium in molten salt at a ratio of
22 mA/cm 2·wt%-Pu at least.
Throughout the plutonium electrotransport experiments with LCCs in this study, the highest
cathodic current density at which plutonium was recovered selectively and stably was 100 mA/cm 2 at
plutonium concentration of 4.6 wt% in the molten salt.
334

Figure 6. Change of LCC potential in electrotransport tests
(effect of cathodic current density and Pu concentration in molten salt)
3.4 Behaviour of plutonium and americium in LCC
The LCC ingot recovered after the electrotransport was analysed in order to evaluate the
behaviour and distribution of plutonium in the cathode. Figure 7 is a SEM image of the intersection of
the LCC ingot obtained in Run 2, where plutonium was collected into the cathode up to 7.75 wt% at
cathodic current density of 41 mA/cm 2 . There is a layer near the bottom of the LCC containing a
335

crystallized phase in high density. Figure 8 is a characteristic X-ray image of plutonium of this layer.
It is clearly shown that the crystallised phase in this region contains a high concentration of
plutonium and that only a small amount of plutonium exists in the bulk. The plutonium-rich phase
was identified to be PuCd 6
by quantitative EPMA analysis.
Figure 7. SEM image of the LCC ingot shown in Figure 4 (near the bottom)
Figure 8. Characteristic X-ray image of Pu at bottom region of LCC ingot
From these results, it seems most likely that plutonium reduced at the LCC beyond its solubility
limit in liquid cadmium instantly forms PuCd 6
at the surface of the LCC and settled down to the
bottom of the cathode. It is still possible, however, that the segregation of PuCd 6
was caused by
vertical temperature gradient in the LCC, because it was cooled very slowly after the experiments.
Further tests are needed to elucidate the mechanism of PuCd 6
accumulation at the bottom of the LCC.
336

Figure 9. Relation between Pu concentration in LCCs and γ-ray from cathode ingots
The exposure dose rate of γ-ray from Am 241 in the LCC ingots was plotted in Figure 9 versus the
concentration of plutonium collected in the cathodes. The dose rate was measured for both top and
bottom of the ingots by a GM survey meter placed outside of the glove box at a distance of about
2 mm from the ingots. These plots have a distinctive tendency. At low concentrations of plutonium in
the LCCs, the dose rate at either top or bottom of the ingots increased according as the
electrotransport proceeded. When the concentration of plutonium in the LCCs reached its solubility
limit, however, the increase of the dose rate simultaneously stopped. In our previous LCC study with
uranium and lanthanide elements, similar behaviour was observed [16]. While the concentration of
uranium in the LCC increased linearly to the electricity, deposition of gadolinium and neodymium
stopped before uranium saturation and their concentrations remained almost constant. Such behaviour
of americium and lanthanides can be explained by the following consideration based on a local
equilibrium model. Assume that electrode reactions of plutonium and americium at the LCC are
reversible, that is, a local equilibrium relationship between the two elements at the cathode
cadmium/molten salt interface described in equation in Figure 10 is established at every moment.
Activity of plutonium in the LCC increases with its concentration before it reaches solubility limit.
After saturation, plutonium forms intermetallic compound PuCd 6
. Because PuCd 6
is solid at 773 K,
the activity of plutonium in the LCC does not change although a larger amount of plutonium may be
collected beyond its solubility limit. Under this condition, deposition of americium would be
restrained so that the local equilibrium would be maintained.
337

Figure 10. Concept of Pu activity change in liquid cadmium and
local equilibrium at the surface of LCC
3.5 Expectation of plutonium recovery rate at a practical electrorefiner
From the results of the electrotransport experiments with LCCs, it was found that lithium was
reduced after exhaustion of plutonium in the salt at higher cathodic current densities, and that
cathodic current density adequate for smooth plutonium collection is proportional to its concentration
in the molten salt at least in the range of this study. Therefore, it is reasonable to assume that
plutonium reduction current at the LCC is controlled by diffusion of plutonium ion and is
proportional to its concentration in the molten salt. It is also proper to assume that plutonium
reduction current is proportional to the surface area of the LCC, although this relationship should be
significantly influenced by the geometric design of the electrorefining equipment.
Based on the above consideration, plutonium collection rate at LCC in practical electrorefining
equipment was estimated as follows. The sum of the concentrations of all actinides in the molten salt
is planned to be adjusted to 2 mol% (about 8 wt%) in the practical operation of electrorefining step
[16]. In LCC operation, the plutonium / uranium ratio in molten salt will be set considerably high in
order to avoid formation of uranium dendrite. This ratio was assumed to be 7/1 in this calculation.
Cathodic current density adequate for smooth plutonium collection was assumed to be proportional to
the concentration of plutonium in molten salt at a ratio of 20 mA/cm 2·wt%-Pu based on the
338

consideration described in the preceding section. The inner diameter of a practical scale LCC was
supposed to be 30 cm. Consequently, reduction current for plutonium at one LCC was evaluated as:
0.02 (A/cm 2·wt%) * 7 (wt%) * 15 2 π(cm 2 ) = 99.0 (A)
This is equivalent to a collection rate of 294 grams of plutonium per hour. This performance is
considered high enough in designing a practical electrorefiner and pyrometallurgical process.
4. Conclusion
Plutonium was smoothly collected into a LCC even without cathode stirring. At plutonium
concentration of 2.11 wt% in molten LiCl-KCl and cathodic current density of 41 mA/cm 2 , the
collection efficiency of plutonium was nearly 100% and maximum plutonium loading into the LCC
was 7.75 wt%. At higher cathodic current densities, lithium and plutonium metals were generated at
the surface of the LCC and reacted with ceramic LCC parts. Collection efficiency was decreased due
to these reactions.
Cathodic current density adequate for smooth plutonium collection was proportional to its
concentration in molten salt at a ratio of about 20 mA/cm 2·wt%-Pu at least in the range of this study.
At plutonium concentration of 4.6 wt% in molten salt, cathodic current density of 100 mA/cm 2 was
attained without any trouble such as solid deposit growth or descent of cathode potential indicating
reduction of lithium.
It was considered that plutonium collected into the LCC after saturation formed intermetallic
compound PuCd 6
and accumulated at the bottom of the LCC based on EPMA analysis of the LCC
ingot. It is still possible, however, that segregation of PuCd 6
was caused by a vertical temperature
gradient in the LCC in the course of the slow cooling process.
Increase of γ-ray count from Am 241 in the LCC ingots stopped coincident with saturation with
plutonium. This behaviour was reasonably explained with the local equilibrium model between
plutonium and americium at the surface of the LCC.
Plutonium collection rate in practical electrorefining equipment was estimated to be 294 grams
per hour for one LCC based on the assumption that the collection rate is proportional to the plutonium
concentration in the molten salt and the surface area of the LCC. This performance is considered
sufficient in designing a practical electrorefiner and pyrometallurgical process.
Acknowledgements
This research was carried out under the joint program Fundamental Study on Molten Salt
Electrorefining between the Japan Atomic Energy Research Institute (JAERI) and the Central
Research Institute of Electric Power Industry (CRIEPI). The authors would like to thank
Mr. Shiozawa, JAERI, for chemical analysis of the samples. We would like to express special thanks
to Mr. Sasayama, JAERI. We also appreciate Dr. Suzuki, JAERI, for continuing encouragement and
helpful advice. Finally, we gratefully acknowledge all the staff at the Plutonium Fuel Research
Facility in the Oarai Research Establishment, JAERI for their warm support.
339

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[2] T. Yokoo, M. Kawashima and Y. Tsuboi, Proceedings of International Conference on the
Physics of Reactors: Operation, Design and Computation, PHYSOR, 1990, Marseille, France,
Vol. 4, p. III-41.
[3] M. Tokiwai, T. Kobayashi, T. Koyama, M. Tsunashima, S. Horie, T. Kawai, I. Kakehi,
H. Matsuura, K. Yanagida and M. Shuto, Proceedings of International Conference on Fast
Reactors and Related Fuel Cycles, FR’91, October 1991, Kyoto, Japan, Vol. II, 12.7-1.
[4] T. Koyama, R. Fujita, M. Iizuka and Y. Sumida, An Experimental Study of Molten Salt Electrorefining
of Uranium Using Solid Iron Cathode and Liquid Cadmium Cathode for Development
of Pyrometallurgical Reprocessing, Nucl. Technol., 1995, 110, 357-368.
[5] I. Johnson, M.G. Chasanov and R.M. Yonco, Pu-Cd System: Thermodynamics and Partial
Phase Diagram, Trans. Metallurg. Soc. AIME, 1965, 233, 1408-1414.
[6] J. Roy, L. Grantham, D. Grimmett, S. Fusselman, C. Krueger, T. Storvick, T. Inoue,
Y. Sakamura and N. Takahashi, Thermodynamic Properties of U, Np, Pu, and Am in Molten
LiCl-KCl Eutectic and Liquid Cadmium, J. Electrochem. Soc., 1996, 143, 2487-2492.
[7] Annual Technical Report for 1993, Chemical Technology Division, Argonne National
Laboratory Report ANL-94/15 (1994).
[8] Annual Technical Report for 1991, Chemical Technology Division, Argonne National
Laboratory Report ANL-92/15 (1992).
[9] Annual Technical Report for 1992, Chemical Technology Division, Argonne National
Laboratory Report ANL-93/17 (1993).
[10] Annual Technical Report for 1994, Chemical Technology Division, Argonne National
Laboratory Report ANL-95/24 (1995)
[11] T. Koyama, M. Iizuka, Y. Shoji, R. Fujita, H. Tanaka, T. Kobayashi and M. Tokiwai, An
Experimental Study of Molten Salt Electrorefining of Uranium Using Solid Iron Cathode and
Liquid Cadmium Cathode for Development of Pyrometallurgical Reprocessing, J. Nucl. Sci.
Technol., 1997, 34, 384-393.
[12] T. Koyama, M. Iizuka, N. Kondo, R. Fujita, H. Tanaka, Electrodeposition of Uranium in
Stirred Liquid Cadmium Cathode, J. Nucl. Mater., 1997, 247, 227-231.
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[13] Y. Arai, S. Fukushima, K. Shiozawa and M. Handa, Fabrication of (U, Pu)N Fuel Pellets, J. Nucl.
Mater., 1989, 168, 280-289.
[14] O. Shirai, M. Iizuka, T. Iwai, T. Suzuki and Y. Arai, Electrode Reaction of Plutonium at Liquid
Cadmium in Licl-Kcl Eutectic Melts, J. Nucl. Electroanal. Chem., 2000, 490, 31-36.
[15] M. Lewis, T. Johnson, A Study of the Thermodynamic and Reducing Properties of Lithium in
Cadmium at 773 K, J. Electrochem. Soc., 1990, 137, 1414-1418.
[16] T. Koyama, M. Iizuka, H. Tanaka, Proceedings of International Conference on Nuclear
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341

SESSION IV
BASIC PHYSICS, MATERIALS AND FUELS
S. Pilate (BN) – H. Takano (JAERI)
343

SESSION IV
BASIC PHYSICS, MATERIALS AND FUELS
SUB-SESSION IV-A:
BASIC PHYSICS
345

NUCLEAR DATA MEASUREMENTS FOR P&T AND FUTURE PLANS IN JNC
H. Harada, S. Nakamura, K. Furutaka, T. Baba
Japan Nuclear Cycle Development Institute
4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1194, Japan
Abstract
Measurements of thermal neutron capture cross-sections (σ 0
) and resonance integrals (I 0
) of some
important fission product (FP) nuclides, performed at JNC for partitioning and transmutation (P&T)
studies, are presented. Method of the measurements and the results are reviewed, and possible reasons
for discrepancies between the present data and that obtained by other researchers are discussed.
Future plans on nuclear data measurements for P&T studies are presented.
347

1. Introduction
The reduction of the environmental loads is one of the important issues of the countries all over
the world. In the field of nuclear energy production, the amount of radioactive nuclear wastes should
be reduced. To reduce the amount, some methods have to be designed to transform these radioactive
nuclides into stable ones.
One of the ways to transform these radioactive nuclides is the transmutation using the reactor
neutrons. In order to study schemes of nuclear transmutation using the reactor neutrons, it is essential
to know precise values of neutron cross-sections of these radioactive nuclides. Looking at the nuclear
data of neutron reactions for these radioactive FP nuclides, the data are rather scarce, and the existing
data are old and sometimes poor in accuracy. In this point of view, we have performed measurements
of thermal (2 200 m/s) neutron capture cross-sections (σ 0
) and resonance integrals (I 0
) of some
important radioactive FP nuclides and some the surrounding stable nuclides, using an activation
method. These nuclides include 133,134,135,137 Cs, 90 Sr, 99 Tc and 127,129 I.
In this paper, our experimental method to determine σ 0
and I 0
of these radioactive FP nuclides,
and the obtained results are reviewed. Then, our future plans on nuclear data measurements are
presented.
2. Nuclear data measurements for P&T by JNC from 1990 to 1999
JNC has organised some researches on nuclear data measurements by several universities in
Japan. These researches include:
• Measurements of fast neutron induced fission cross-section of americium isotopes
(Department of Quantum Science and Energy Engineering, Tohoku University).
• Neutron capture cross-section measurement of 237 Np with lead slowing-down spectrometer
(Research Reactor Institute, Kyoto University).
• Preliminary experiment of neutron capture cross-section of 99 Tc with lead slowing-down
spectrometer (Research Reactor Institute, Kyoto University).
• Measurement of neutron capture cross-sections of 99 Tc (Research Laboratory for Nuclear
Reactors, Tokyo Institute of Technology).
• Measurement of fission cross-section and fission neutron spectrum of 237 Np by an advanced
technique (Department of Quantum Science and Energy Engineering, Tohoku University).
At the same time, JNC has continued their own effort on measurements of neutron capture crosssections
of long-lived FP nuclides (LLFP) using some research reactors in Japan, from 1990 until now
[1-9]. This paper concentrates on the latter topic, and describes our experimental methods and results.
The experimental procedure to determine σ 0
and I 0
is based on an activation method, in which
samples are irradiated with neutrons and then γ rays are measured which are emitted during deexcitations
of the daughter of the capture products, to determine reaction rate R of the capture
reaction. For the γ-ray measurements, a high purity Ge detector with a large volume is used. This
enables determination of γ-ray yield in an efficient and reliable manner and thus the reaction rate R is
determined precisely in case the precise values of emission probabilities of the γ-rays are available. In
348

order to determine σ 0
and I 0
, at the same time, irradiations and measurements are also done for
samples with a Cd shield.
The procedure to determine σ 0
and I 0
from the obtained R is based on Westcott’s convention [10].
It was already described elsewhere [1], and only a brief summary is given here:
In the Westcott’s convention, the reaction rate R in well-moderated neutron fields is expressed as:
R = n υ 0
σ eff
where in the convention nυ 0
is the “neutron flux” with neutron density n including thermal and
epithermal neutrons and with velocity υ 0
=2 200 m/s, and σ eff
an effective cross-section. The σ eff
is
written as:
σ eff
= σ 0
[g G th
+ r(T/T 0
) 1/2 s 0
G epi
]
where σ 0
is the reaction cross-section for 2 200 m/s neutrons and g the measure of deviation of the
cross-section from the 1/υ law in the thermal energy region. In the analysis the g is assumed to be
unity. The quantity r(T/T 0
) 1/2 gives the fraction of epithermal neutrons in the neutron spectrum, and s 0
is defined as:
s 0
= 2I’ 0
/((π) 1/2 σ 0
)),
with I’ 0
the reduced resonance integral, i.e. the resonance integral after subtracting the 1/υ
component. The G th
and G epi
are self-shielding factors for thermal and epithermal neutrons,
respectively. The above equations are combined to read:
R/σ 0
= G th
φ 1
+ s 0
G epi
φ 2,
where φ 1
and φ 2
represent simplified flux factors. The φ 1
and φ 2
can be determined by using flux
monitors whose cross-sections and resonance integrals are already determined precisely. As flux
monitors, we use Co and Au, which differ in sensitivities to thermal and epithermal neutrons. By
using two flux monitors with different sensitivities to thermal and epithermal neutrons, φ 1
and φ 2
can
be determined unambiguously. The self-shielding factors G th
and G epi
are usually almost unity and can
be calculated by considering geometries of irradiations.
From the obtained reaction rates R and flux factors φ 1
, φ 2
for irradiations with and without Cd
shield the cross-sections σ 0
and the reduced integrals I’ 0
are deduced. The resonance integral, I 0
, is
deduced from I’ 0
using the following relation:
I 0
= I’ 0
+ 2 σ 0
(E 0
/E Cd
) 1/2
where E 0
and E Cd
are neutron energy at 2 200 m/s and Cd cut-off energy.
In Table 1, the results obtained by the present authors are summarized [1-9] along with the data
previously published by other research groups [11-18]. The table includes results of the neutron
capture cross-sections and resonance integrals for long-lived FP nuclides (LLFP) as well as those for
their stable isotopes: the latter are also important because these stable nuclides absorb neutrons and
affect transmutation rates of LLFP, and also these stable nuclides can be transformed into radioactive
ones by absorbing neutrons.
Some of the data obtained by the present author do not differ significantly from those obtained
previously. For example, for 134 Cs nuclide, the effective cross-section obtained by Bayly et al. [17]
349

agrees ours within limits of errors. Also, the thermal neutron cross-section and the resonance integral
of 129 I nuclide obtained by the present authors are close to those published by Eastwood et al. [14].
On the other hand, for some nuclides, the results obtained by the present authors differ
considerably from others. An example of the discrepancy is the result for 99 Tc nuclide. Although the
thermal neutron cross-section does not differ much, the result of the reduced resonance integral
obtained by the present authors [4] are about twice as large as that obtained by Lucas et al. [13], as
depicted in Figure 1. The origin of this discrepancy may be ascribed to the characteristics of their
irradiation: their analysis was based on the same convention as ours which is valid only for well
moderated neutron spectrum, but at one of their irradiation positions, the index for the epithermal
neutrons, r, is as large as 0.15. Even with our data, the number of existing data of resonance integral
for 99 Tc nuclide is only two. It should be stressed that, in order to be certain that correct values of σ 0
and I 0
are obtained, at least two different types of measurements have to be done. This is also true for
other radioactive FP nuclides such as 90 Sr and 137 Cs.
It should also be noted that the data presented in [4] do not include the error of γ-ray emission
probabilities: because of its short life, γ-ray emission probabilities of 100 Tc nuclide are determined
with an error of as large as 17% [22]. In order to obtain σ 0
and I 0
of 99 Tc nuclide more accurately,
accurate values of γ-ray emission probabilities of 100 Tc nuclide are needed.
3. Future plans on nuclear data measurements for P&T in JNC
Following the decision of the Atomic Energy Commission in Japan that the basic study on P&T
should be continued JNC resumes the nuclear data measurement under a basic study scheme for P&T
from 2000. Now we are planning to extend our area of nuclear data measurements over capture crosssections,
fission cross-sections and decay data for important LLFP and MA for the energy region
from thermal to a few MeV. The plan includes the following researches and developments:
• More precise determination of the capture cross-sections of nuclides such as 99 Tc and 129 I.
• Development of prompt γ-ray spectroscopic method for the determination of the neutron
capture cross-sections of LLFP.
• Development of a new spectroscopic method to measure neutron cross-sections for energy
range from thermal to a few MeV.
3.1 More precise determination of the capture cross-sections
As already mentioned above, γ-ray emission probabilities of 100 Tc nuclide are not determined with
enough accuracy because of its short life. To obtain more precise values for capture cross-sections of the
99
Tc nuclide, the γ-ray emission probabilities of 100 Tc should be determined more accurately. In order to
achieve this, a β-γ coincidence measurement system has been developed for the determination of γ-ray
emission probabilities of short-lived nuclides [23]. An experiment has been already performed using the
system to precisely determine γ-ray emission probabilities of 100 Tc nuclide.
350

3.2 Prompt γ-ray spectroscopy
For the determination of capture cross-sections of nuclides whose capture products are stable, a
conventional activation method can not be applied in which de-excitation γ-rays are observed of
daughter nuclides of the capture products. These include some important long-lived FP nuclides such
as 93 Zr, 79 Se and 107 Pd. In order to determine capture cross-sections of such nuclides, a prompt γ-ray
spectroscopic method is being developed, in which complete level schemes are constructed by inbeam
γ-γ coincidence measurements using thermal neutron beam and then γ-ray emission
probabilities are determined. By using the obtained emission probabilities, neutron capture crosssections
are determined.
This method is also applicable to nuclides whose capture products are not stable ones:
measurements in this method will confirm the results that are already obtained by using other
methods such as an activation technique.
3.3 Development of a new spectroscopic method for neutron cross-sections in a wide energy region
In order to efficiently determine neutron cross-sections of LLFP and minor actinides over a
broad energy range from thermal to MeV region, some new experiments will be required. The present
authors are planning to start the international collaborations from 2001 Japanese fiscal year.
Table 1. Neutron capture cross-sections at 2 200 m/s neutron energy and resonance
integrals for some important fission product nuclides, obtained
by the present authors as well as other researchers
Nuclide
Half-life (year)
137
Cs 30
90
Sr 29
Previous data (barns)
(Authors and published year)
σ eff
= 0.11 ± 0.03
(Stupegia 1960 [11])
σ eff
= 0.8 ± 0.5
(Zeisel 1966 [12])
99
I’ 0
= 186 ± 16
Tc 2.1 × 10 5 σ 0
= 20 ± 2
(Lucas 1977 [13])
σ 0
= 27 ± 2
129
I 1.6 × 10 7 I 0
=36 ± 4
127
I
(stable)
(Eastwood 1958 [14])
σ 0
= 4.7 ± 0.2
I 0
=109 ± 5
(Friedmann 1983 [15])
135
I 0
= 61.7 ± 2.3
Cs 2.3 × 10 6 σ 0
= 8.7 ± 0.5
(Baerg 1958 [16])
134
Cs 2
133
Cs
(stable)
σ eff
=134 ± 12
(Bayly 1958 [17])
σ 0
= 30.4 ± 0.8
I 0
= 461 ± 25
(Baerg 1960 [18])
Data obtained by JNC
(barns)
σ 0
= 0.25 ± 0.02
I 0
= 0.36 ± 0.07
(1990 [1], 1993 [2])
σ = (15.3 + 1.3 - 4.2) × 10 -3
I ≤0.16
0
(1994 [3])
σ 0
= 22.9 ± 1.3
I = 398 ± 38 (I’ = 388 ± 38)
(1995 [4])
σ 0
= 30.3 ± 1.2
I 0
= 33.8 ± 1.4
(1996 [5])
σ 0
= 6.40 ± 0.29
I 0
= 162 ± 8
(1997 [6])
σ 0
= 8.3 ± 0.3
I 0
= 38.1 ± 2.6
(1997 [7])
σ eff
= 141 ± 9
(1999 [8])
σ 0
= 29.0 ± 1.0
I 0
= 298 ± 16
(1999 [9])
351

Figure 1. Thermal neutron capture cross-sections (upper)
and resonance integrals (lower) of 99 Tc nuclide
352

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S. Motoishi, J. Nucl. Sci. Tech., 36 (1999), pp. 635-640.
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[14] T.A. Eastwood, A.P. Baerg, C. B. Bigham, F. Brown, M.J. Cabell, W.E. Grummitt, J.C. Roy,
L.P. Roy, R.P. Schuman, in Proceedings of the International Conference on the Peaceful Uses
of Atomic Energy, (United Nations: Geneva, 1958), 54-63.
[15] L. Friedmann, D.C. Aumann, Radiochimica Acta, 33 (1983), pp. 183-187.
[16] A.P. Baerg, F. Brown, M. Lounsbury, Can. J. Phys., 36 (1958), pp. 863-870.
[17] J.G. Bayly, F. Brown, G.R. Hall, A.J. Walter, J. Inorg. Nucl. Chem., 5 (1958), pp. 259-263.
353

[18] A.P. Baerg, R.M. Bartholomew, R.H. Betts, Can. J. Chem., 38 (1960), pp. 2147-2531.
[19] Neutron Cross-section, BNL-325 and Suppl. 1, (1955 and 1957).
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Energy, Geneva, Vol. 16, p. 44 (1958).
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354

NEW DATA AND MONTE CARLO SIMULATIONS ON
SPALLATION REACTIONS RELEVANT FOR THE DESIGN OF ADS
J. Benlliure
Universidad de Santiago de Compostela
15706 Santiago de Compostela, Spain
Abstract
The main European experimental programs to characterise spallation reactions used as neutron sources
are reviewed. The neutron production is described in terms of the multiplicities, spatial and energy
distributions. Experiments to determine the residual nuclei production in the spallation target are also
discussed. These data are used to benchmark existing nuclear model calculations.
355

1. Introduction
Nowadays it is well established that spallation reactions constitute an optimum neutron source to
feed a sub-critical reactor in an accelerator driven system (ADS). However, the present knowledge
about this reaction mechanism is not accurate enough for any technical application. Two main aspects
will play a major role in the design and construction of the target assembly of the spallation neutron
source used in an ADS: the neutron yields and the residual nuclei produced in the reaction.
The neutron production should be characterised in terms of the neutron multiplicity and their
spatial and energy distributions. The neutron multiplicity will determine the intensity of the protondriver
accelerator while their energy and spatial distribution should be considered to design the
geometry of the spallation target and the shielding to high-energy neutrons.
Spallation reactions do not only produce neutrons but also residual nuclei. Most of these nuclei
are radioactive, therefore, activation problems should be considered in the design of the target. In
Figure 1 we report an example of the simulated activity induced in a cylindrical lead target by a 1 mA
proton beam. As it is showed in the figure, both the cooling time and the total activity induced in the
target are not negligible. In addition the residual nuclei will contribute to the corrosion of the target
and to the radiation damages in the target, accelerator window and structural materials.
Figure 1. Calculated radioactivity induced in a cylindrical lead target
(120 cm long and 46 cm diameter), by a 1 GeV proton beam of 1 mA after
one year of irradiation. Calculations done with the Lahet Code system by D. Ridikas [1].
Although spallation reactions are understood qualitatively, they are not known with the degree of
accuracy needed for any technical application. In this sense, most of the existing codes used to
describe these reactions have a limited predictive power. Therefore a large experimental program has
been initiated in Europe few years ago in order to improve our knowledge on these reactions. These
experiments will provide accurate data to benchmark more reliable model calculations. In the
following sections we will describe some of these experiments.
356

2. General considerations on spallation reactions
Spallation reactions are collisions induced by light-energetic projectile on a heavy-ion target.
These reactions can be described as a two-stage process. First the incoming projectile induces quasifree
nucleon-nucleon collisions with the nucleons of the target nucleus. These collisions lead to the
prompt emission of few neutrons and protons. A fraction of the kinetic energy of the incoming
projectile will be transferred to the target nucleus as excitation energy, e.g. a 1 GeV proton deposits on
average 200 MeV in the target nucleus. The rest of the energy will be shared between the prompt
emitted nucleons. This emission of fast nucleons will play an important role in the development of an
inter-nuclear cascade process inside the target.
In a second step the residual nuclei produced in the collisions will de-excite by evaporation of
low energy protons and neutrons or fissioning. In principle, neutron evaporation is favoured since to
evaporate protons or to fission the system needs extra energy to overcome the Coulomb barrier. The
energy of the evaporated nucleons is determined mainly by the temperature reached by the residual
nucleus in the collisions and will be in the range of a few MeV.
To describe the full interaction of a relativistic projectile with a target material we should
consider that the most probable interaction of this projectile with the target material will be governed
by electromagnetic processes. The main consequence of the electromagnetic interaction of the
projectile with the electrons of the target material will be the slow down of the projectile and heat load
of the target.
Figure 2. Range of protons in lead as a function of their energy
The nuclear interaction between the projectile and the target is determined by the total reaction
cross-section. In the case of the reaction proton on lead at 1 GeV the reaction cross-section corresponds
to a mean free path of protons on lead around 15 cm. In contrast, the mean free path for electromagnetic
interaction is much shorter, consequently the incoming projectile will be slowed down before any
nuclear interaction. The electromagnetic interaction can be characterised in terms of the range of the
incoming particle in the traversed medium. In Figure 2 we represent the range of protons in lead as a
function of their energy. As can be seen the range of a proton with 1 GeV in lead is around 55 cm.
357

In order to describe the inter-nuclear cascade inside the target we should estimate the energy
balance in the interaction of the projectile with the target. If we consider that on average the nuclear
interactions take place at 15 cm, the mean energy loss of the incoming projectile before the reaction will
be 200 MeV. In addition the energy dissipated in the first spallation reaction is around 200 MeV. This
excitation energy leads to a large population of different residual nuclei. The remaining kinetic energy
§ 600 MeV will be shared between the four or five prompt nucleons emitted during the first stage of the
reaction. These nucleons will lead to secondary reactions in the target (inter-nuclear cascade).
The maximum energy of the prompt emitted nucleons is expected to be lower than 300 MeV.
According to Figure 2, at this energy the range of protons in lead is few centimetres, therefore most of
the secondary protons will be stopped before they induced any secondary reactions. The inter-nuclear
cascade will be then induced mainly by neutrons with energies lower than 300 MeV. At this energy
the spallation reaction is less violent and only few nucleons will be produced with an energy range
lower than 100 MeV. Consequently the final reaction residues will be very close in mass and atomic
number to the target nucleus.
In summary we can conclude that an incoming proton at 1 GeV on a lead target will induce on
average two spallation reactions. The first one at high energy will determine mostly the residual nuclei
produced in the target. The second reaction at lower energy will contribute to the multiplication and
moderation of the neutrons.
3. Neutron production in spallation reactions
The neutron yield produced in spallation reactions will depend strongly on the projectile-target
combination. In principle the heavier the target nucleus the larger the neutron excess leading to a
larger neutron yield. Nevertheless the gain factor between heavy and light targets is not larger than a
factor of five. In contrast, the radiotoxicity induced in the spallation target can be drastically reduced
when using lighter targets as discussed in [2].
In addition to the neutron yields, reliable information on the energy and spatial distributions of
the neutrons is required. This information can be used in the design of the spallation-target assembly
geometry or the shielding to high-energy neutrons.
Neutron detection is not an easy task since neutrons only feel the strong interaction. This is the
reason why different experimental devices are needed to characterise the neutron production in
spallation reactions. In the following we will consider two examples.
3.1. Measurement of neutron yields
Neutron multiplicities can be investigated using liquid-scintillator based detectors with a large
angular acceptance. A clear example is the detectors BNB (Berlin Neutron Ball) [3] and ORION [4]
used by the NESSI collaboration (Berlin-Ganil-Jülich). This collaboration has performed a large
experimental program to determine the neutron yields produced in thin and thick targets for a large
range of primary projectiles and energies. To fulfil this programme experiments were done at GANIL
(France) [4], Jülich (Germany) [3] and CERN (Switzerland) [5].
358

Figure 3. Average neutron multiplicity per incident proton as a function of target thickness and
beam energy for Pb, Hg and W materials obtained by the NESSI collaboration [3]
Figure 3 shows some representative results obtained by this collaboration at Jülich with the BNB
detector. This figure represents the measured average neutron multiplicity per incident proton as a
function of target thickness and beam energy for Pb, Hg and W materials. As can be seen, for the
different target materials, the neutron multiplicity saturates at a given target thickness which increase
with the proton energy. This saturation corresponds to the previous picture where every incident
proton originates on average two collisions with a mean free path of 15 cm.
3.2. Energy and spatial distribution of neutrons
Specific experimental set-ups are needed to measure the spatial and energy distribution of the
neutrons produced in spallation reactions. A clear example is the experiments performed by the
“transmutation” collaboration at Saturne (France). These measurements use two different experimental
techniques to cover the full energy range of the neutrons produced in the reaction. The detection of
neutrons with energies lower than 400 MeV was based in a measurement of their time of flight
between the incident proton beam, tagged by a plastic scintillator, and a neutron-sensitive liquid
scintillator [6]. Neutrons with higher energies were measured using (n,p) scattering on a liquid
hydrogen converter and reconstruction of the proton trajectory in a magnetic spectrometer [7]. An
additional collimation system allowed determining the angular distribution of the neutrons.
This experimental technique was used to investigate the neutron production in reactions induced
by protons with energies between 0.8 and 1.6 GeV on thin and thick lead targets. In Figure 4 we report
some of the results obtained with a 1.2 GeV proton beam on a two centimetres thick lead target. This
359

kind of measurements allow to characterise the spallation process. High energy neutrons emitted at
low angles are representative of the first stage of the collision while low energy neutrons emitted
isotropically correspond to the evaporation phase. Measurements done with thicker targets are
representative of the inter-nuclear cascade leading to the multiplication and moderation of neutrons.
Figure 4. Neutron production double-differential cross-sections measured in 1.2 GeV induced
reactions on a 2-cm thick Pb target [8]. The histograms represent calculations using the Bertini INC
Code [9] while the dotted lines corresponds to calculation done with the Cugnon INC Code [10].
4. Residue production in spallation reactions
Residue production in spallation reactions can be investigated using two different experimental
approaches. In the standard one, the reaction is induced in direct kinematics, the light-energetic
projectile hits a heavy target. In this case the recoil velocity of the residues produced in the reaction is
not sufficient to leave the target and -spectroscopy or mass spectrometry techniques are used to
identify those residues. The main limitation of this technique is that for most of the residues the
measurement is done after decay and consequently only isobaric identification is possible.
Better suited seems to be the measurement of the spallation residues in inverse kinematics. In this
case the heavy nucleus is accelerated at relativistic energies and impinges a light target. Due to the
kinematical conditions, the reaction residues leave easily the target and can be identified in a short
time using the appropriate technique.
360

4.1 Measurement of residue production in inverse kinematics
One of the most outstanding experiments are the ones performed by a German-Spanish-French
collaboration at GSI. The technique used in these experiments takes advantage of the inverse
kinematics and the full identification in mass and atomic number of the reaction residues by using a
magnetic spectrometer.
The experiments have been performed at the SIS synchrotron at GSI. Primary beams of 197 Au,
208
Pb and 238 U accelerated up to an energy of 1 A GeV impinged on a liquid hydrogen or deuteron
target. The achromatic spectrometer FRS [11] equipped with an energy degrader, two position
sensitive scintillators and a multisample ionisation chamber allowed to identify in atomic and mass
number all the reactions products with half lives longer than 200 ns and with a resolving power of
A/ A § 400. Figure 5 represents an example of the resolution achieved with this experimental
technique. The final production cross-sections are evaluated with an accuracy around 10%. In
addition, the magnetic spectrometer allows determining the recoil velocity of the reaction residues.
This information is relevant for the characterisation of the damages induced by the radiation in the
accelerator window or the structural materials. More details about these experiments can be found in
[12-15].
Figure 5. Example of identification matrix obtained with the Fragment Separator at GSI [13]
361

Figure 6. Two-dimensional cluster plot of the isotopic production cross-sections of all the spallation
residues measured at GSI in the reaction 208 Pb(1 A GeV) + p shown as chart of nuclides [15]
In Figure 6 we present in a chart of the nuclides all the residues measured in the reaction
208
Pb(1 A GeV) + p. More than 1 000 different spallation residues were identified in this reaction. As
can be seen in this figure, the spallation residues populate two different regions of the chart of the
nuclide. The upper region corresponds to the spallation-evaporation residues which populate the socalled
evaporation-residue corridor. The second region populates medium-mass residues produced in
spallation-fission reactions. Both reactions mechanism, fission and evaporation, should be considered
to describe the production of spallation residues in these reactions.
The measured isotopic production cross-sections for some selected elements are presented in
Figure 7. This figure shows clearly the quality of the measured data that can be used to benchmark any
model calculation.
362

Figure 7. Isotopic production cross-sections for some of the elements produced in the reactions
208
Pb + p at 1 A GeV [15]. The data are compared with two model calculations, the dark line correspond
to the results obtained with the Lahet Code [16] while the hell line was obtained with the intra-nuclear
cascade model of Cugnon [10] coupled to the evaporation-fission Code ABLA from GSI [17,18].
4.2 Measurement of residue production in direct kinematics
Although this method only allows isobaric identification after -decay, for some shielded isotopes
it is possible to determine their primary production cross-sections. In principle this experimental
technique is less beam time consuming than the inverse kinematics. Therefore full excitation functions
can be established for selected isotopes as shown in Figure 8. In addition this method can be applied to
thin and thick targets.
From the results shown in Figure 8 we can conclude that the low energy reactions produced
mainly residues close to the target nucleus, while most of the reaction residues populating a large part
of the nuclear chart are produced by energetic particles. The two most important experimental
programs in Europe using this technique are the ones performed by the group of R. Michel at the
University of Hanover [19] and Y.E. Tiratenko at the ITEP in Moscow [20].
363

Figure 8. Excitation functions for some selected isotopes produced
in the interaction of protons with lead measured with γ-spectroscopy techniques [21]
5 Reactions in the 20-200 MeV energy range
Reactions induced by neutrons and light-charged particles in the energy range between 20 and
200 MeV are representative of the inter-nuclear cascade in the spallation target. This reactions will
play a major role in the multiplication and moderation of the neutrons. The energy dissipated in these
reactions leads to the emission of few particles and consequently only residual nuclei close in mass
and atomic number to lead will be produced.
These experiments intend to measure the double-differential production cross-sections of
neutrons and light-charged particles. It is out to the scope of this work to review on all the
experimental programs investigating these reactions. Most of them contribute to the Hindas project of
the Fifth Framework Programme of the European Commission. The experiments take advantage of a
large network of European facilities delivering protons and neutrons in the investigated energy range:
KVI (Netherlands), Louvain-la-Neuve (Belgium) and Uppsala (Sweden). More detailed information
about this program can be found in the contributions of N. Marie, F.R. Lecolley and J.P. Meulders to
this conference.
6. Model simulations
Most of the existing models to simulate spallation reactions describe the first stage of the collision
in terms of semi-classical nucleon-nucleon collisions (intra-nuclear cascade) and a statistical deexcitation
of the hot residue. The main inputs of the intra-nuclear cascade are the elastic and inelastic
nucleon-nucleon cross-sections and the distribution of the nucleons in the target nucleus in position
and momentum space. The statistical evaporation of particles is generally based in the Weisskopf
364

formalism while fission can be describe according to the prescription of Bohr. In this case the main
parameters are the description of the level density and the Coulomb barriers for charged-particles
emission or fission. Another critical parameter is the coupling time between the intra-nuclear cascade
and the evaporation.
The last model intercomparison done by NEA [22] revealed important deficiencies in most of the
existing codes to describe spallation reactions. In fact these deficiencies can be understood due to the
lack of experimental information. The new data provided by the present experimental programs will
help to improve this situation. In Figure 7 we compare the measured isotopic production cross-sections
for some of the elements produced in the reactions 208 Pb + p at 1 A GeV at GSI [15] with two model
calculations. In this figure, the dark line correspond to the results obtained with the Lahet Code
(Bertini + Dresner) [16] while the hell line was obtained with the intra-nuclear cascade model of
Cugnon [10] coupled to a new the evaporation-fission Code ABLA from GSI [17,18]. As can be seen,
the new models provide a much better description of the experimental data.
7. Conclusions
Spallation reactions are considered as an optimum neutron source to feed an ADS. However this
reactions are not known with the degree of accuracy needed for the design of such devices. This is the
main justification for a large experimental program initiated in Europe few years ago to collect high
quality data about neutron and residual nuclei production in these reactions. This experimental
program takes advantage of most of the existing heavy-ion facilities in Europe in order to cover the
full energy range involved in the interaction of light-energetic projectiles with heavy-ion targets. Most
of these programs are supported by different programs of the European Commission like Hindas or the
European Spallation Source (ESS).
365

REFERENCES
[1] D. Ridikas, thesis, University of Caen, 1999.
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[3] A. Letourneau et al., Nucl. Instr. and Methods B 170 (2000) 299.
[4] B. Lott et al., Nucl. Instr. and Methods A 414 (1998) 117.
[5] D. Hilscher et al., Nucl. Instr. and Methods A 414 (1998) 100.
[6] F. Borne et al., Nucl. Instr. and Methods A 385 (1997) 339.
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[8] X. Ledoux et al., Phys. Rev. Lett. 82 (1999) 4412.
[9] H.W. Bertini et al., Phys. Rev. 131 (1963) 1801.
[10] J. Cugnon et al., Nucl. Phys. A 620 (1997) 475.
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[13] J. Benlliure et al., Nucl. Phys. A, in print.
[14] F. Farget et al., Nucl. Phys. A, in print.
[15] T. Enqvist et al., Nucl. Phys. A, in print.
[16] R.E. Prael et al., Los Alamos National Laboratory, Report LA-UR-89-3014.
[17] A.R. Junghans et al., Nucl. Phys. A 629 (1998) 635.
[18] J. Benlliure et al., Nucl. Phys. A 628 (1998) 458.
[19] R. Michel et al., Nucl. Instr. and Methods B 129 (1997) 153.
[20] Y.E. Titarenko et al., Nucl. Instr. and Methods A 414 (1998) 73.
[21] M. Gloris et al., Nucl. Instr. and Methods B (2000), in print.
[22] R. Michel, P. Nagel, International Codes and Model Intercomparison for Intermediate Energy
Activation Yields, NSC/DOC(97)-1, NEA/P&T No. 14.
366

THE MUSE EXPERIMENTS FOR SUB-CRITICAL NEUTRONICS
VALIDATION AND PROPOSAL FOR A COMPUTER BENCHMARK
ON SIMULATION OF MASURCA CRITICAL AND SUB-CRITICAL EXPERIMENTS
R. Soule
CEA-Cadarache, DRN/DER/SPEx
Building 238, 13108 Saint-Paul-Lez-Durance, France
E. Gonzalez-Romero
CIEMAT, FACET Project
Av. Complutense 22, 28040 Madrid, Spain
On behalf of the MUSE collaboration
Abstract
Accelerator driven systems (ADS) are being explored in France in the frame of the research
programme on radioactive waste management options. Besides studies aimed to clarify the
motivations for ADS, a significant programme has been started to validate experimentally the main
physics principles of these systems. This experimental programme was initiated at CEA-Cadarache in
1995, with the sponsorship of EdF and Framatome. Since 1997, the CNRS has joined the programme,
which is now a common CEA-CNRS-EdF-Framatome programme, open to external partners, in
particular since October 2000 the European Community in the frame of the 5th FW Programme.
367

1. Introduction
Since 1991, the Commissariat à l’Énergie Atomique (CEA) MASURCA (MAquette
SURgénératrice de CAdarache) has studied the physics of hybrid systems, involving a sub-critical
reactor coupled with an accelerator.
These studies are being explored in France in the frame of the research programme on radioactive
waste management options.
The potential of this kind of systems is to be found in:
• The concentration of waste in a limited number of dedicated facilities.
• The sub-criticality of such a system, which is a particularly attractive argument in favour of the
safety of such concepts and which, more particularly, allows for the introduction of new fuels.
Besides studies aimed to clarify the motivations for ADS, a significant programme has been
started to validate experimentally the main principles of these systems, in terms of the physical
understanding of the different phenomena involved and their modelling, as well as in terms of
experimental validation of coupled systems, sub-critical environment/accelerator.
This validation must be achieved through mock-up studies of the sub-critical environments
coupled to a well-known source of external neutrons which represents the spallation source. The
experimental investigations on the physics of sub-critical external source-driven systems are
performed at the CEA Cadarache MASURCA facility, in the frame of the MUSE (MUltiplication of
an external Source Experiments) programme.
2. Neutronic validation of source-driven sub-critical systems
2.1 Principles
Neutronic studies of fast critical systems have been largely performed in the past and the
associated calculations tools (including both recommended nuclear data and calculation tools) and bias
factors developed for the predictions of such systems have been mainly based on integral experiments
in critical facilities. Validated experimental techniques have also been developed, directly applied to
power critical operating systems.
In order to validate the physics characteristics of a source-driven sub-critical multiplying system,
the main original idea has been to separate the experimental validation of the sub-critical
multiplication phenomena of the external neutron source, from the validation of the external source
characteristics. This can be done using a well-known (in energy spectrum and geometrical position)
external source to drive the multiplying sub-critical core.
The neutron source (e.g. a spontaneous fission source or a fixed energy neutron generator), can be
surrounded by a “buffer” medium, simulating the diffusing properties of a spallation source. The
leakage neutrons through the “buffer” zone are then used as an external source but with a modified
energy spectrum. The source neutrons, after having travelled approximately one mean-free path in the
multiplying medium, become distributed, in energy and space, as the neutrons generated by fission in
the multiplying medium.
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The experimental programmes allow to validate both nuclear data and calculation methods used
to describe the sub-critical core, in terms of sub-critical reactivity level, spatial flux distributions,
neutron spectra, spectrum indexes and source neutron worth (the ϕ* parameter) [1]. If the source can
be used in continuous and pulsed modes, static and dynamic reactivity measurements are possible.
This point is of relevance, since the experimental investigation of the different techniques to monitor
the sub-criticality level during operation of an ADS is still an open question.
2.2 The MUSE experiments
This validation experimental programme was started at CEA (with the sponsorship of EdF and
Framatome) in 1995 with the short exploratory MUSE-1 experiment [1], providing some insight into
the physical behaviour of the neutron population in the sub-critical system. The MUSE-2 experiment
[2], (2 months in 1996) was devoted to the experimental study of diffusing materials (sodium and
stainless steel) placed around the external source to modify the neutron importance of the external
source.
Since 1997, the French Scientific Research Committee (the CNRS) has joined this programme,
which is now a common CEA-CNRS-EdF-Framatome programme in the frame of the joint research
programme GEDEON (Waste management with innovative options programme).
In 1998, during three months the MUSE-3 experiments have been performed [3]. The external neutron
source of about 1.0 E+08 n/s was produced by a commercial neutron generator based on the (d,t) reaction.
This neutron generator operating in both continuous and pulsed modes allowed to complete the study of
diffusing materials (sodium and pure lead) and to explore the dynamic behaviour of the multiplying
medium for different sub-criticality levels (≈-0.16, -1.6, -3.2, -4.7 $ respectively). The expected dependence
of monitors responses in function of the sub-criticality has been observed.
The MUSE programme is entered a new phase starting in 2000.
First of all, the installation at MASURCA of a ad-hoc deuton accelerator (the GENEPI),
especially developed and built at the CNRS/IN2P3/ISN Grenoble, with improved performances (in
terms of the quality of the neutron pulse and source intensity), and the use of both (d,d) and (d,t)
reactions, will enable to explore different neutron spectra, different source worths and their ratios to
the fission neutron worth (the ϕ* parameter). Accurate dynamic measurements based on the pulsed
mode operation of the GENEPI will allow new experimental reactivity determination of the subcritical
multiplying media.
Secondly, the MUSE experiments have been opened to the international collaborations via the
5th Framework Programme of the European Community and also via bilateral collaborations between
CEA and Argonne National Laboratory (USA) and JAERI (Japan) respectively.
The future MUSE experiments will investigate several sub-critical configurations loaded in the
MASURCA facility driven by the GENEPI external neutron sources. The foreseen configurations will
have MOX fuel (with ≈25% enrichment in Pu) with sodium, gas or lead coolant. Physical presence of
a spallation source will be simulated by surrounding the GENEPI neutron source with a pure lead zone
(see Figure 3). Several levels of sub-criticality will be investigated (from -0.2 to -16 $). Foreseen
measurements concern the sub-criticality levels by classical Source Multiplication Method but also via
dynamic and noise methods, the neutron spatial distributions, the neutron spectra, the effective delayed
neutron fraction and the neutron source importance parameter. Extensive cross-comparisons of codes
and nuclear data are foreseen.
369

Experimental reactivity control techniques, related to sub-critical operation, will be developed
and inter-compared. In particular, in the field of reactivity control related to sub-critical operation,
development, inter-comparison and improvement of experimental techniques will be performed.
Description of experimental conditions, techniques and associated results with uncertainties will be set up.
Complementary experiments (the SAD experiments) will be performed at Dubna (Russian
Federation) in the frame of a sub-contract of the 5th FWP of the European Community, studying
different spallation neutron sources (Pb, Pb-Bi, W targets) produced by the 660 MeV protons of the
Dubna synchrotron. These experiments will allow the experimental characterisation of the spallation
neutrons propagation into materials (target, fuel and structural materials) encountered in ADS.
The analysis of the whole experiments allows to develop a reference calculation route (including
both recommended nuclear data, validated calculation tools and associated residual uncertainties) for
the design of ADS and for the deep spallation neutrons transport penetration to optimise the neutron
shielding, with a special attention to a “forward” direction (behind the target area).
3. The MASURCA facility
The MASURCA facility is dedicated to the neutronic studies of fast reactor lattices. The materials
of the core are contained in cylinder rodlets, along with in square platelets. These rodlets or platelets
are put into wrapper tubes having a square section (4 inches) and about 3 meters in height. These tubes
are hanged vertically from a horizontal plate supported by a structure of concrete. The core itself can
reach 6 000 litres. To build such cores the tubes are introduced from the bottom in order to avoid that
the fall of a tube corresponds to a positive step in reactivity.
The reactivity control is fulfilled by absorber rods in varying number depending of core types and
sizes. The control rods are composed of fuel material in their lower part, so that the homogeneity of
the core is kept when the rods are withdrawn. The core is cooled by air and is surrounded by a
biological shielding in heavy concrete allowing operation up to a flux level of 10 9 n/cm 2 /sec. Core and
biological shielding are inside a reduced pressure vessel, relative to the outside environment. The
limited maximum operating power of the facility is limited to 5 kW th
.
4. The GENEPI accelerator
The GENEPI (GÉnérateur de NEutrons Pulsé Intense) accelerator has been especially formed for
the MUSE experiments in the MASURCA facility for brief neutron injections with a very fast
intensity decrease.
It will produce a pulsed neutron beam of about 1 µs during a maximum relative time of 5 . 10 -3 s,
that is a maximum frequency of about 5 000 Hz.
In this way, deuton impulses are created, then focalised, accelerated and guided to a deuterium or
tritium target. The (D,D) or (D,T) nuclear reactions produce neutrons of about 2.67 MeV or 14.1 MeV
respectively. For incident deutons of about 250 keV, the neutron yield is greater for the (D,T) reaction
than for the (D,D) reaction.
This accelerator is a classical electrostatic one with a lower mean neutrons production than the
same type of accelerators. The main originality of GENEPI concerns its operating mode based on high
ions peak current (50 mA) and a decreasing time of the neutron impulse of some 100 µs.
370

The GENEPI accelerator is mainly composed of:
• A deuton source.
• The extraction and focusing electrodes.
• The 250 keV electrostatic accelerator.
• The mass separator.
• The deuterium or tritium target , as indicated in the Figure 1.
The main characteristics of the ion beam are indicated in the Table 1.
Table 1. Deuton beam characteristics
Beam energy
Peak current
Repetition rate
Minimum pulse duration
Mean beam current
Spot size
Pulses reproducibility
140 to 240 keV
50 mA
10 to 5000 Hz
700 nanoseconds
(200 µA (for a duty cycle of 5 000 Hz)
Diameter ≈20 mm
Fluctuations at the 1% level
The characterisation of the neutron production by both deuterium and tritium targets has been
performed by the ISN Grenoble team. The characterisation of the neutron source intensity is based on
the activation of a Si detector by:
• The recoiled protons produced by the (d,d) reaction on the deuterium target and in the magnet
chamber due to deuterium implantation.
• The recoiled protons and alpha particles produced by the (d,t) reaction on the tritium target.
This alpha monitoring gives a neutron pulse shape very similar to this obtained by ionic current as
indicated in Figure 2.
The characterisation of the neutron production spectrum is based on the activation analysis of 58 Ni
foils. For the 2.67 MeV neutrons produced by the D(d,n) 3 He reaction, the 58 Ni(n,p) 58 Co is used. The
14 MeV neutrons spectrum produced by the T(d,n) 4 He reaction is determined by both the
58
Ni(n,2n) 57 Ni and 58 Ni(n,np) 57 Co reactions representative of the neutrons with an energy higher than
13 MeV. For a natural Ni target of 20 mm diameter (corresponding to a mass of about 580 mg)
irradiated during 14 hours at 2 000 Hz with a pulse width of 700 nanoseconds FWHM, the
neutrons/pulse intensities, indicated in the Table 2 have been measured.
371

Table 2. Neutron intensities
Target characteristics* Nuclear reaction Neutrons/pulse
D in 1mg/cm 2 Ti deposit (Φ 30mm) D(d,n) 3 He 4.0 E+04
T (1Ci) in 0.25 mg/cm 2 Ti deposit (Φ 25mm) D(t,n) 4 He 1.7 E+06**
T (10 Ci) target idem Expected: 3.0 to 9.0 E+06
*D/Ti or T/Ti atomic ratio is close to 1.5.
**Measurement done after a 50% decrease of the tritium content of the target.
An accurate monitoring of the external neutron source in term of intensity and pulse form is of
prime importance for a good and accurate understanding of the dynamic measurements. The target
beam current, the proton and alpha + proton spectroscopy signals and the alpha + protons time
distribution referenced to the neutron source pulse will be available for the physicists during the future
MUSE experimental campaigns.
5. The MUSE-4 experiments
As the MUSE experiments are based on a parametric approach, the MUSE-4 configurations are
based on the ZONA2 fuel cell (see Figure 2), representative of a Pu fast burner core (Pu enrichment of
≈25% with ≈18% content of 240 Pu) with sodium coolant. The fuel zone is radially and axially reflected
by a stainless steel/sodium (75/25) shielding. The GENEPI deuteron guide is horizontally introduced
at the core mid-plane and the deuterium or tritium target is located at the core centre (see Figure 3). To
compensate the spatial effect due to the presence of the GENEPI beam guide in the north part of the
loading, the south symmetric part will be loaded with pure lead. To simulate the physical presence of a
spallation source, a pure square (20 cm thick) lead zone will be introduced around the GENEPI target
(see Figure 3).
Six different experimental configurations will be studied:
• A critical one, the GENEPI being shut off, in which all the safety and neutron flux level and
spectrum measurements will be performed. In this configuration the reactivity scale will be
experimentally determined by classical pilot rod shutdown measurement.
• Three sub-critical configurations (k eff
being successively of about 0.994: the SC1
configuration, 0.97: the SC2 configuration and 0.95: the SC3 configuration). These three
configurations will be obtained by replacing radially some peripheral fuel cells by stainless
steel/sodium cells. The west/east symmetry along the beam guide axis will be preserved.
• Two complementary asymmetrical sub-critical configurations, with k eff
of about 0.95 and
0.93, obtained from the reference critical one and from the above SC1 sub-critical
respectively, by complete insertion of the same safety rod. These two last configurations will
be of interest in the frame of studying the decoupling effects and the excitation of high order
flux harmonics by the external source.
A very extensive experimental programme has been planned for one year, including the active
participation of the different partners as indicated in the Table 3. In support to the transmutation
studies of minor actinides, fission rates of 232 Th, 233 U, 237 Np, 240 Pu, 241 Pu, 242 Pu, 241 Am, 243 Am and 244 Cm
will be measured using fission chambers. Transmutation of some long-lived fission products will be
372

also experimentally determined using activation foils such as 197 Au, 115 In, 160 Dy, natural Mn,
representative of the LLFP’s of interest in term of capture cross-sections.
6. Benchmark on computer simulation of MASURCA critical and sub-critical experiments
The study of the neutronic of accelerator driven systems, in which an intense external neutron
source maintains a stationary power level, requires the extension and validation of appropriated
computational tools to solve steady-state and time–dependent problems, from the standard codes and
nuclear data libraries developed for critical reactors. The MASURCA nuclear assembly used in the
MUSE experiment, that has been and will be configured as a critical and sub-critical reactor, offers a
unique opportunity for test and validation of the available and new computational tools. For these
purposes we propose to organise in collaboration with the OCDE Nuclear Energy Agency a
benchmark on computer simulation of MASURCA critical and sub-critical experiments particularly
concentrated on the MUSE-4 experiments. The fact that the results can be compared with already
available experimental data and data to be obtained in very short time will allow to go beyond the
simple observation of the coincidence and discrepancy between codes or nuclear data libraries.
The benchmark model would be oriented to compare simulation predictions based on available
codes and nuclear data libraries between themselves and with experimental data related to: TRU
transmutation, criticality constants and space and time evolution of the neutronic flux following source
variation, in the framework of liquid metal fast sub-critical systems.
The benchmark could be divided in three steps:
• First step will allow understanding the simulation methods of the different groups and tuning
of the simulations programmes with the experimental data of one already measured critical
configuration (COSMO).
• In the second step, the MUSE-4 reference configuration is proposed for simulation of the
different reactor parameters (criticality constant, flux distribution...) in a nearly critical
configuration, critical-2$.
• The third step is oriented to the simulation of reactor time response to the external source in
the sub-critical reference configuration.
To allow the use of the widest range of simulation codes to participate in the benchmark the
geometry and material compositions will be described in detail but homogenised at the tube level. The
errors introduced by the homogenisation approximation have been checked by the MUSE
collaboration, and they are very small, typically from less than 0.1% in k eff
to a maximum of 8% in the
absolute flux at the worst tube (k eff
= 0.995). Detailed figures and tables will clarify this geometrical
description of the MASURCA configurations and of the reference points for requested calculations.
Special attention will be paid to insure that most of the requested calculations can be compared with
directly measurable parameters.
In the case of the COSMO critical MASURCA configuration the requested calculations will
include: the criticality constant, k eff
, 235 U fission rate as a function of the position at the available
experimental channels (horizontal and vertical); spectral index from the reaction rates in the available
detectors and activation foils and the energy dependence of the neutron spectrum in a few
characteristic positions (to clarify discrepancies between codes).
For the critical-2$ MUSE-4 reference configuration calculations the requested parameters should
include: the criticality constant, k eff
, 235 U fission rate as a function of the position at the available
373

experimental channels (horizontal and vertical); spectral index from the reaction rates in the available
detectors and activation foils and the energy dependence of the neutron spectrum in a few
characteristic positions; the 235 U fission rate in the available experimental positions as a function of the
time after the deuteron-tritium source pulse and the neutron mean lifetime.
Finally for the sub-critical MUSE-4 reference configuration calculations the requested parameters
should include: the 235 U fission rate as a function of the position at the available experimental channels
(horizontal and vertical); spectral index from the reaction rates in the available detectors and activation
foils and the energy dependence of the neutron spectrum in a few characteristic positions; the 235 U
fission rate in the available experimental positions as a function of the time after the deuteron-tritium
source pulse; the neutron mean lifetime; the change in neutron multiplication from the critical-2$ to
the sub-critical configuration and the difference between k eff
and k source
for the sub-critical
configuration.
The probable situation that some of the calculations will be made before the experiments are
performed is the best warranty for making blind simulations and to understand the potentialities and
accuracy of the different computational tools.
7. Conclusions
From the year 2000, the MUSE experiments begin an international test stand for the intercomparison
and development of specific experimental techniques and for the validation of a reference
calculation route, including recommended nuclear data, validated calculation tools and associated
residual uncertainties related to the neutronics specificities of the accelerator driven systems. During
the MUSE-4 experiments in year 2001, the coupling between the GENEPI accelerator and a MOX fuel
with sodium coolant will be studied. During the two following years, the GENEPI accelerator will be
coupled with a MOX fuel with gas coolant representative of Fast Gas Cooled sub-critical system. A
small MOX fuel zone with lead coolant will be also investigated.
A first important conclusion of the European collaboration during the definition of the MUSE-4
critical and sub-critical configurations concerns the important discrepancy observed between
deterministic code and stochastic codes using a priori the same nuclear libraries. The understanding of
this discrepancy should be obtained via an international calculation benchmark based, in a first step on
very simplified experimental configurations, in terms of geometrical description and material
compositions. In a second step, real critical and sub-critical configurations studied during the MUSE-4
experiments will be proposed.
Acknowledgements
The authors wish to thank all the partners of the MUSE collaboration with a special mention for
their fruitful participation to Mrs. C.A. Bompas (CEA-Cadarache), Mrs. Billebaud, Messrs Giorni,
Loiseaux, Brissot and Wachtarczyk (ISN-Grenoble), Mr. Plaschy (PSI, Villingen), Mr. Villamarin
(CIEMAT, Madrid) and Mr. Seltborg (5KTH Stockholm).
374

REFERENCES
[1] M. Salvatores, M. Martini, I. Slessarev, R. Soule, J.C. Cabrillat, J.P. Chauvin, P. Finck,
R. Jacqmin, A. Tchistiakov in Proceedings of the Second International Conference on
Accelerator Driven Transmutation Technologies and Applications, (ADTTA’96), Kalmar,
Sweden, June 3-7, 1996.
[2] R . Soule, M. Salvatores, R. Jacqmin, M. Martini, J.F. Lebrat, P. Bertrand, U. Broccoli, V. Peluso
in Proceedings of the International Conference on Future Nuclear Systems, (Global’97),
Yokohama, Japan, October 5-10, 1997, Vol. 1, pp. 639-645.
[3] J.F. Lebrat, R. Soule, M. Martini, J.P. Chauvin, C.A. Bompas, P. Bertrand, P. Chaussonnet,
M. Salvatores, G. Rimpault, A. Giorni, A. Billebaud, R. Brissot, S. David, D. Heuer, J.M. Loiseaux,
O. Meplan, H. Nifenecker, J.B. Viano, J.F. Cavaignac, J.P. Longequeue, J. Vergnes, D. Verrier in
Proceedings of the Third International Conference on Accelerator Driven Transmutation
Technologies and Applications, (ADTTA’99), Prague, Czech Republic, June 1999.
375

376

Figure 3. XY loading (at the core mid-plane)
of the MUSE-4 reference critical configuration (provisional)
Figure 4. Neutron pulse time spectrum
0.2
0.175
Ionic Current
particles
0.15
Arbitrary Unit
0.125
0.1
0.075
0.05
0.025
0
1000 800 600 400 200 0 200 400 600 800 1000
Time (ns)
377

Table 3. Planned experimental programme during the MUSE-4 experiments
378
SC1 SC2 SC3 SC3 SC2
REF OFF (d,t) OFF (d,t)
(d,d)
SYM ASY
SYM ASY SYM ASY SYM ASY
OFF (d,t) (d,d) (d,d)
Operating
Rod worth X X X X X
Monitor calibration X X X X X
Reactor calibration X X X X X
Chamber inter-calibration X
GENEPI monitoring X X
Target control study X X X X X X
Statics
Source multiplication X X X X
Radial traverses X X X X X X X X X
Axial traverses X X X X X X X X X
Spectrum indices X X X
Foil activation X X X X X X
3
He spectrum X X X X X X
252
Cf source importance X X X X
GENEPI source importance X X X X X X X X
Dynamics
Reactor noise X X
Transfer function X X X X X X
Frequency modulation X X X X X X
Pulsed source methods X X X X X X X X
Rossi- & Feyman-α methods X X X X X X X X X X
• For each sub-critical configuration (SC1, SC2 and SC3) the GENEPI will be shut OFF/ON with deuterium target and ON with tritium target.
• SYM configurations correspond to “clean” configurations.
• ASYM configurations correspond to the above “clean” configurations, but with BC2 safety rod completely inserted.

OECD/NEA BENCHMARK CALCULATIONS FOR ACCELERATOR DRIVEN SYSTEMS
M. Cometto
CEA/PSI
5232 Villigen, Switzerland
B.C. Na
OECD/NEA
Le Seine Saint-Germain, 12, blvd des Iles, 92130 Issy-les-Moulineaux, France
P. Wydler
5452 Oberrohrdorf, Switzerland
Abstract
In order to evaluate the performances of the codes and the nuclear data, the Nuclear Science
Committee of the OECD/NEA organised in July 1999 a benchmark exercise on a lead-bismuth cooled
sub-critical system driven by a beam of 1 GeV protons. The benchmark model is based on the ALMR
reference design and is optimised to burn minor actinides using a “double strata” fuel cycle strategy.
Seven organisations (ANL, CIEMAT, KAERI, JAERI, PSI/CEA, RIT and SCK•CEN) have
contributed to this exercise using different basic data libraries (ENDF/B-VI, JEF-2.2 and JENDL-3.2)
and various reactor calculation methods. Significant discrepancies are observed in important
neutronic parameters, such as k eff
, reactivity swing with burn-up and neutron flux distributions.
379

1. Introduction and benchmark specification
Recognising a need for code and data validation in the area of accelerator driven systems (ADS),
the OECD/NEA Nuclear Science Committee organised in 1994 a benchmark on a sodium cooled subcritical
system with a tungsten target and minor actinide (MA) and plutonium nitride fuel [1]. In that
benchmark, considerable differences in calculated initial k eff
and burn-up reactivity swing indicated a
need for refining the benchmark specification and continuing the exercise with a wider participation
[2,3]. The present benchmark was therefore launched in July 1999 to resolve the discrepancies
observed in the previous exercise and to check the performances of reactor codes and nuclear data for
ADS with unconventional fuel and coolant. The choice of lead-bismuth as a coolant and target
material reflects the increased interest in this technology.
The ADS is designed to operate as a MA burner in a “double strata” strategy, featuring a fully
closed fuel cycle with a top-up of pure MA. Two fuel compositions are prescribed in accordance with
this strategy. In the start-up core, MAs are mixed with plutonium from UOX-fuelled LWRs. In the
equilibrium core, the fuel represents an asymptotic composition reached after an indefinite number of
cycles. Both fuel compositions differ strongly from the usual U-Pu mixed oxide (MOX) composition.
The fuel is diluted with zirconium as an inert matrix for the core to give a k eff
of about 0.95 at BOL.
Since the emphasis is on code and data validation in the energy region below 20 MeV, a predefined
spallation neutron source, produced with HETC assuming a proton energy of 1 GeV and a beam
radius of 10 cm, was provided to the participants.
The R-Z benchmark model, shown in Figure 1, comprises four regions: a central lead-bismuth
target zone, a void zone in the beam duct region, a multiplying region which consists of homogenised
fuel, cladding and lead-bismuth coolant and, finally, an outer reflector zone. The reactor operates at a
nominal power of 377 MW th
and the core has a residence time of 5 years. To simulate a load factor of
0.85, the power is reduced to 320 MW th
in the burn-up calculations. At EOL the fuel reaches an
average burn-up of approximately 200 GWd/t HM
.
Figure 1. R-Z model of ADS
200
Void
Reflector
Z T
= 150
Height (cm)
Core
(Fuel region)
50
Target
0
0
20 92 142
Radius (cm)
380

The choice of adopting the ALMR reference system as a basis for the benchmark model has the
advantage that a detailed plant concept is available and the characteristics of the plant with normal
cores has already been analysed in great detail, including transient and beyond design basis
behaviour. The present benchmark model is therefore also suited for transient benchmarks.
As a follow-up to the present benchmark, a transient benchmark dealing with the beam trip
problem of the ADS is currently being defined. Preliminary results of the current benchmark exercise
were presented in November 1999 at the OECD/NEA Workshop on Utilisation and Reliability of
High Power Accelerators in Aix-en-Provence [4].
2. Results and discussion
Seven institutions participated in the benchmark, using nuclear data mainly from ENDF/B-VI,
JEF-2.2 and JENDL-3.2. For the core analysis, both deterministic and Monte Carlo methods were
applied. The list of participants, basic libraries and codes used are summarised in Table 1.
Table 1. List of participants, basic data libraries and codes
Organisation Basic library Codes used Method
ANL
(USA)
CIEMAT
(Spain)
KAERI
(Korea)
JAERI
(Japan)
PSI/CEA
(CH/France)
RIT
(Sweden)
SCK•CEN
(Belgium)
ENDF/B-VI
ENDF/B-V for lumped FP
MC 2 -2, TWODANT, REBUS-3 Deterministic
JENDL-3.2
EVOLCODE system
Monte Carlo
ENDF/B-VI for fission yields (NJOY, MCNP-4B, ORIGEN-2.1)
JEF-2.2
TRANSX-2.15, TWODANT, Deterministic
JENDL-3.2 for Pb and 242m Am DIF3D-7.0, REBUS-3
JENDL-3.2
ATRAS (SCALE, TWODANT, Deterministic
BURNER, ORIGEN-2)
ERALIB I (JEF-2.2 based) ERANOS, ORIHET Deterministic
JEF-2.2
JEF-2.2
ENDF/B-VI for Pb and 233 U
NJOY, MCNP-4B, MCB,
ORIGEN-2
NJOY97.95, MCNP-4B,
ORIGEN-2
Monte Carlo
Monte Carlo
In addition to their reference solution, some participants provided additional results obtained
with different methods or basic libraries. In particular, RIT provided neutron flux distributions for the
equilibrium core obtained with ENDF data and k eff
results for the start-up core based on the JENDL
library.
In the following, we summarise the most important results of the benchmark exercise. In
particular, we focus on the one-group microscopic cross-sections, the neutron spectrum, the neutron
flux distributions, the integral parameters k inf
and k eff
as well as the k eff
variation with the burn-up.
Other important parameters such as the external neutron source strength and safety parameters are
also discussed.
In the analysis of the results, it is necessary to consider how the participants accounted for the
nuclear power in the system. Whereas ANL, CIEMAT, KAERI and PSI/CEA took into account the
energy coming from both fissions and captures, JAERI and SCK•CEN considered only the energy
coming from fissions and RIT neglected both the energy coming from captures and from delayed
381

neutrons. Since the neutron flux is normalised to the given reactor power of 377 MW, neglecting the
contribution of some reactions to the power leads to an overestimation of the flux. The effect can be
estimated to be about 4% for JAERI and SCK•CEN and about 9% for RIT; it therefore has an
important influence, especially on the fuel burn-up and on the neutron flux distributions.
The main neutronic characteristics reported by the participants are summarised in Tables 2 and 3.
Table 2 refers to the start-up core and Table 3 refers to the equilibrium core.
Table 2. Main neutronic characteristics of the start-up core
Organisation
Parameters ANL CIEMAT JAERI KAERI PSI/CEA RIT SCK•CEN
Library ENDF JENDL JENDL JEFF JEFF JEFF JEFF
k inf
1.15894 1.13732 1.15920 1.13256 1.13141 1.149 1.14729
k eff
at BOL 0.98554 0.9570 0.9650 0.94546 0.94795 0.9590 0.9590
P/A ratio * 1.307 1.245 1.253 1.226 1.228 1.220 1.241
Source (n/s)-BOL 6.1E17 1.65E18 1.25E18 4.11E18 2.26E18 2.54E18 2.29E18
Source (n/s)-EOL 4.12E18 3.51E18 2.88E18 7.33E18 3.96E18 4.83E18 4.54E18
Neutron’s median 210 212 162 222 214 220 220
energy (keV)
Coolant void
reactivity (pcm) **
3 161
2 433
3 905
3 214
3 813
3 048
3 686
2 596
2 870
1 655
2 904
1 863
2 896
1 681
Fuel Doppler effect
(pcm) ***
0
13
38.2
323.7
20.2
31.9
16.5
27.2
6.2
12.4
48
53
11
48.7
β eff
at BOL (pcm) 156 246 173.5 – 184.0 195 –
Table 3. Main neutronic characteristics of the equilibrium core
Organisation
Parameters ANL CIEMAT JAERI KAERI PSI/CEA RIT SCK•CEN
Library ENDF JENDL JENDL JEFF JEFF JEFF JEFF
k inf
1.14420 1.11629 1.14192 1.13366 1.13165 1.150 1.14884
k eff
at BOL 0.96895 0.9370 0.9494 0.94174 0.94374 0.9570 0.95509
P/A ratio * 1.308 1.241 1.258 1.258 1.260 1.245 1.274
Source (n/s)-BOL 1.39E18 2.54E18 1.94E18 4.49E18 2.55E18 2.70E18 2.47E18
Source (n/s)-EOL 3.18E18 2.38E18 2.00e18 5.80E18 2.86E18 3.39E18 2.64E18
Neutron’s median 185 188 152 181 179 188 183
energy (keV)
Coolant void
reactivity (pcm) **
3 318
2 154
4 511
2 582
4 138
2 821
3 902
2 034
2 732
1 925
3 045
1 605
3 144
1 681
Fuel Doppler effect
(pcm) ***
20
12
17.1
277.6
30.4
45.8
23.0
43.7
4.2
5.8
45
49
98
103
β eff
at BOL (pcm) 116 221 145.2 – 155.9 188 –
* Ratio between production and absorption reaction rates in the heavy metals.
voided ref
** Calculated as keff
− keff
for the BOL (first row) and EOL (second row).
980K
1580K
980K
1580K
*** Calculated as ( keff
− keff
)/( keff
⋅ keff
) for the BOL (first row) and EOL (second row).
382

2.1 One-group microscopic cross-sections and k inf
The one-group microscopic cross-sections provided by ANL, KAERI, PSI/CEA and SCK•CEN
are obtained by means of a cell calculation; i.e. the fundamental mode neutron spectrum of the fuel
cell is used for averaging the cross-sections. CIEMAT, JAERI and RIT derived one-group
microscopic cross-sections from a reactor calculation; the latter cross-sections represent averages
over the core fuel zone and, therefore, differ from those obtained from cell calculations. As shown by
additional calculations made by PSI/CEA and SCK•CEN, the differences due to the averaging method
are between 4.5% and 13% for the one-group capture cross-sections and less than 6% for the onegroup
fission cross-sections.
The figures at the end of this section show microscopic one-group capture cross-sections of the
most relevant isotopes in the equilibrium core. Figure 2 compares cell averaged and Figure 3
compares core averaged one-group cross-sections.
The cross-sections show some correlation with the basic data used: the JENDL-based crosssections
are in good agreement for all the isotopes. Some discrepancies appear between the JEFFbased
cross-sections: when comparing the core averaged data, RIT and SCK•CEN give very close
results which, however, differ from the PSI/CEA results for the majority of isotopes. When
comparing the cell averaged data, the results provided by KAERI, PSI/CEA and SCK•CEN generally
differ from one another.
A direct comparison of both cell- and core-averaged cross-sections requires caution.
Nevertheless, the following general conclusions can be made: when comparing one-group crosssections,
based on different basic libraries, good agreement can be observed only for uranium
isotopes (not presented in the figures). Large discrepancies are observed for plutonium isotopes,
particularly for the capture cross-sections. Compared to other libraries, ENDF gives a higher value for
238
Pu and lower values for the other isotopes. The values obtained with JEFF and JENDL are closer,
except for the isotopes 238 and 241.
Considerable discrepancies between cross-sections based on different libraries are observed also
for neptunium, americium and curium. For example, JENDL gives by far the highest capture crosssections
for 243 Cm, 246 Cm and 247 Cm, and ENDF gives significantly lower fission cross-sections for
242
Cm, 243 Cm and 245 Cm (not presented in this paper).
The k inf
results (presented in Tables 2 and 3) show quite a spread. The maximum differences are
2.5% for the start-up core and 3.0% for the equilibrium core. Interestingly, no clear correlation with
the nuclear data used can be observed. Significant discrepancies are observed between the two
JENDL-based results (about 2.0%). The four JEFF-based results can be grouped into two classes,
characterised by high (RIT and SCK•CEN) and low (KAERI and PSI/CEA) values. Another
interesting effect is that JEFF predicts similar k inf
values for both cores, whereas ENDF and JENDL
predict a k inf
difference of about 1.5% between the start-up and the equilibrium cores.
383

Figure 2. Microscopic capture cross-sections for the equilibrium core (cell averaged)
2
1.8
1.6
1.4
ANL (ENDF)
KAERI (JEF)
PSI/CEA (JEF)
SCK-CEN (JEF)
1.2
XS (Barn)
1
0.8
0.6
0.4
0.2
0
Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Cm-243 Cm-244 Cm-245 Cm-246
Figure 3. Microscopic capture cross-sections for the equilibrium core (core averaged)
2
1.8
1.6
1.4
CIEMAT (JENDL)
JAERI (JENDL)
RIT (JEF)
SCK-CEN (JEF)
PSI/CEA (JEF)
1.2
XS (Barn)
1
0.8
0.6
0.4
0.2
0
Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243 Cm-243 Cm-244 Cm-245 Cm-246
384

2.2 k eff
at beginning of life and k eff
variation
The k eff
values at beginning of life (see Tables 2 and 3) show a maximum discrepancy of 4% for
the start-up core; the spread in the k eff
values is slightly reduced for the equilibrium core. The k eff
values do not show a clear correlation with the library used, indicating that the sensitivity of the
results to the data processing route and/or the neutron transport approximation also has to be
investigated. 1% of the discrepancies arise from the two ENDF-based calculations and 0.8% from the
three JENDL-based results. The four JEFF-based results can be grouped into two classes
characterised by high (RIT and SCK•CEN) and low (KAERI and PSI/CEA) k eff
values. However,
comparing these results is difficult because only RIT used the complete JEF-2.2 library. PSI/CEA
used the library ERALIB1 (adjusted from JEF-2.2) and KAERI and SCK•CEN used data for Pb from
JENDL and ENDF respectively.
A systematic sensitivity analysis performed by KAERI using a simplified 1-D model of the start-up
core enables us to estimate the impact of the nuclear data on k eff
[5]. The results show that the k eff
values are
lower by 2 250 pcm for JEF and by 1 160 pcm for JENDL, when all the important actinides are substituted
from the reference ENDF-based calculation. Overall, the k eff
obtained using JEFF is 2 070 pcm lower than
the reference ENDF-based result; the differences arise mainly from heavy metals (-2 250 pcm), lead
(+680 pcm) and 15 N (-420 pcm) while the contribution of bismuth is small (-73 pcm). JENDL also gives a
lower k eff
value (-2800 pcm) mainly due to the contribution of lead (-1 100 pcm) and heavy metals (-1 160
pcm). The contribution of 15 N and bismuth (-450 and -77 pcm) is similar to that for JEFF.
From the reaction rates provided by the participants, the production over absorption (P/A) ratio is
calculated for all the heavy nuclides and it is presented in Tables 2 and 3. ANL (ENDF) predicts by
far the highest values for both core configurations, whereas the other results are closely grouped. The
two JENDL-based results are similar and, for the four JEFF-based results, SCK•CEN gives the
highest value and RIT the lowest value for both core compositions. The ratios of production to
absorption are similar for the start-up and the equilibrium cores when using ENDF and JENDL data.
However, all four JEFF-based results give larger values (+2.5%) for the equilibrium core than for the
start-up core. Interestingly, the k eff
values are not correlated in a systematic way with P/A ratios as one
would expect. In particular, all four JEFF-based solutions give a lower P/A ratio but a higher k eff
in the
start-up core. ANL, CIEMAT and JAERI calculated similar P/A values but different multiplication
factors for both cores.
The multiplication factors for the two cores do not exhibit consistent biases. This may be due to
the fact that the two cores are not dominated by the same producers and absorbers. From the
submitted reaction rate balance components, it can be deduced that the production rate is dominated
by 239 Pu, 241 Am and 241 Pu in the start-up core and by 238 Pu and 245 Cm in the equilibrium core. The
absorption rate is dominated by 241 Am, 239 Pu and 243 Am in the start-up core and by 243 Am and 238 Pu in
the equilibrium core.
The k eff
variations with burn-up are shown in Figures 4 and 5 and the respective burn-up
reactivity drops, k BOC
-k EOC
, including decomposition of the global reactivity drop into actinide and
fission product components, are given in Tables 4 and 5. In addition to the solution obtained with the
JEFF library, RIT presented, for the start-up core, an additional solution obtained with JENDL data.
385

Figure 4. k eff
variation in the start-up core
0.99
0.98
0.97
0.96
0.95
ANL
CIEMAT
KAERI
PSI/CEA
JAERI
RIT
SCK-CEN
RIT (JENDL)
0.94
0.93
0.92
0.91
0.90
0 1 2 3 4 5
Burn-Up (years)
Figure 5. k eff
variation in the equilibrium core
0.98
0.97
0.96
0.95
0.94
ANL
CIEMAT
KAERI
PSI/CEA
0.93
JAERI
RIT
SCK-CEN
0.92
0 1 2 3 4 5
Burn-up (years)
386

The burn-up reactivity drop values, k BOC
-k EOC
, range from 0.036 to 0.069 in the start-up core and
from -0.004 to 0.036 in the equilibrium core. ANL (ENDF) predicts, in both cases, by far the highest
reactivity drop. The total reactivity drop values in the start-up core are close for five participants
(CIEMAT, KAERI, JAERI, JEFF-based RIT and SCK•CEN) with values between 3 900 and
4 100 pcm but this result hides compensating effects between fission products and heavy metals
contributions and is fortuitous. The results are more spread for the equilibrium core.
Neither the actinide nor the fission product components are well correlated with the nuclear data
library used. Only the contribution of the actinides gives similar results for three of the four JEFFbased
results. Other possible causes of discrepancies can be related to the treatment of fission
products. In particular, it is questionable whether lumped fission products generated for U and Pu can
be representative for a system where a significant fraction of the fissions arise from minor actinides
with a higher mass number, such as Am and Cm. However no obvious correlation can be observed
related to the use of individual or lumped fission products.
Table 4. Start-up core: end of cycle reactivity drop components (in units of 10 3 k)
k from ENDF JENDL JEFF
ANL CIEMAT JAERI RIT KAERI PSI/CEA RIT SCK•CEN
Actinides 28 28 14 22 3 8 4 16
FP’s 41 13 27 34 37 28 35 25
Total 69 41 41 56 40 36 39 41
Table 5. Equilibrium core: end of cycle reactivity drop components (in units of 10 3 •k)
k from ENDF JENDL JEFF
ANL CIEMAT JAERI KAERI PSI/CEA RIT SCK•CEN
Actinides -11 -18 -27 -27 -26 -27 -17
FP’s 45 14 30 43 35 44 22
Total 36 -4 3 16 9 17 5
2.3 Neutron spectrum
Neutron spectra are calculated for both cores at R = 56 cm and Z = 100 cm; this point
corresponds to the centre of the fuel region where the neutron spectrum is dominated by the fission
neutrons. From the submitted neutron spectra, median energies were calculated (see Tables 1 and 2).
Good agreement is observed for most of the participants except for JAERI, which predicts a
clearly softer spectrum (its median energy is about 20% lower than the others). The spectra provided
by the other six participants show a good agreement especially for the energy range above 5-10 keV
that covers approximately 95% of the neutrons. It is only in the lower resonance region (between
100 eV and 1 keV) that the differences between the results become pronounced. Finally, it is
interesting to notice that the spectrum in the equilibrium core is softer than that in the start-up core.
387

2.4 Neutron flux distribution
One radial flux distribution corresponding to the mid-plane and two axial flux distributions
corresponding, respectively, to the centre of the target and the fuel zone were requested. Considerable
differences between the participants are observed, especially for the radial neutron flux distribution
and for the axial flux distribution in the target zone. In particular, ENDF-based results (ANL and RIT)
estimate by far the lowest neutron flux through the target. The discrepancies are also significant in the
fuel region, where differences in flux level and in peak flux position appear. These discrepancies can
be partially explained by the different level of sub-criticality: a system with a lower k eff
needs more
external neutrons in order to maintain the chain reaction. This results in a more peaked axial flux in
the target, at the interface with the duct, and a displacement of the axial flux peak in the fuel towards
the upper part of the core.
An additional study was performed with ERANOS in order to assess the impact of the k eff
value
on the neutron flux distribution. By modifying appropriately the ν values for reproducing the k eff
values submitted by the participants, neutron flux distributions were recalculated for each participant.
From these distributions, spatial dependent correction factors were obtained. Finally, the neutron flux
distributions supplied by the participants were rescaled for the reference k eff
value of 0.95 using these
spatial dependent correction factors. The rescaled neutron flux distributions are presented in Figures
6, 7 and 8 and refer to the equilibrium core.
Figure 6. Axial flux distribution in the target region; results scaled to a k eff
value of 0.95
1.6E+16
1.4E+16
Neutron Flux
1.2E+16
1E+16
8E+15
6E+15
ANL
CIEMAT
KAERI
PSI/CEA
JAERI
RIT (JEF)
RIT (ENDF)
SCK-CEN
4E+15
2E+15
0
0
10
20
30
40
50
60
70
80
90
100
110
120
130
140
150
160
170
180
190
200
Axial position (cm)
388

Figure 7. Axial flux distribution in the fuel region; results scaled to a k eff
value of 0.95
4.5E+15
4E+15
Neutron Flux
3.5E+15
3E+15
2.5E+15
2E+15
ANL
CIEMAT
KAERI
PSI/CEA
JAERI
RIT (JEF)
RIT (ENDF)
SCK-CEN
1.5E+15
1E+15
5E+14
0
0
10
20
30
40
50
60
70
80
90
100
110
120
Axial position (cm)
130
140
150
160
170
180
190
200
Figure 8. Radial flux distribution in the mid-plane; results scaled to a k eff
value of 0.95
9E+15
8E+15
Neutron Flux
7E+15
6E+15
5E+15
4E+15
ANL
CIEMAT
KAERI
PSI/CEA
JAERI
RIT (JEF)
RIT (ENDF)
SCK-CEN
3E+15
2E+15
1E+15
0
0
5
10
15
20
25
30
35
40
45
50
55
60
65
70
75
80
85
92
100
110
120
130
142
Radial position (cm)
In the fuel region, the shape of the axial neutron flux distributions is in good agreement for most
participants. However, the peak position reported by KAERI is shifted to the core centre
(Z = 100 cm). The spread in the absolute value of the flux is still significant and the difference
389

etween the highest and the lowest value (obtained by RIT and PSI/CEA respectively) is about 20%.
As expected, the discrepancies in the axial neutron distributions in the target region become
considerably smaller after the adjustment. The shapes of the flux distributions have a similar trend
and the values are less spread. Similar considerations apply to the radial neutron flux distributions
corresponding to the mid-plane. Referring to the normalisation of flux to power mentioned earlier, the
neutron flux values provided by RIT, SCK•CEN and JAERI are overestimated with respect to the
other participants and should be reduced correspondingly (9% for RIT, 4% for JAERI and
SCK•CEN).
Even with the adjustment taking into account the k eff
effect, there still remain differences in the
flux distributions, especially in the absolute value of the flux and its shape. Large discrepancies are
observed in the shape of the axial flux in the target calculated by CIEMAT and KAERI. The
CIEMAT result shows a distinct feature in the lowest part of the target region, due to the geometry
model used: the lowest part of the target (the first 50 cm) was replaced by a void region. The KAERI
result has a vanishing flux in the void zone; that is probably due to the diffusion approximation.
2.5 Source strength
The source strength, i.e. the number of neutrons per second that the ADS needs in order to
maintain the chain reaction, is an important parameter because it is directly proportional to the
required accelerator power. Its value is given by the following equation [6]:
Pth
⋅ν
k ⎛ 1 ⎞ 1
N = ⋅⎜
−1⎟ ⋅
E ⎝ k ⎠ ϕ *
f
where N is the number of neutrons/s, P th
the thermal power, ν k
and E f
, respectively, the average
number of neutrons and the average energy released per fission in the fuel, k the multiplication factor
of the system without source and ϕ* the importance of spallation neutrons relative to fission neutrons.
The results, presented in Tables 2 and 3, for the beginning and for the end of irradiation, show
quite a spread: at BOL, the ratio between the highest (KAERI) and the lowest (ANL) values is about 7
for the start-up core and 3 for the equilibrium core. The other three JEFF-based results lie in a similar
range (maximum difference of about 10%) and the two JENDL-based results show a difference of
about 30% for both core configurations.
As the above equation indicates, the required number of external neutrons is strongly dependent
on the multiplication factor. It may be interesting to isolate that effect by dividing the source strength
value by (1/k-1) in order to remove the differences due to the multiplication factor (at least partially,
knowing that ϕ* is also dependent on k). The numerical value obtained in that way is dependent on
P th
, ν k
, E f
and ϕ* and therefore should be close for all the participants. The values obtained are quite
discrepant, KAERI presenting by far the highest value. Interestingly, the correlation between the three
other JEFF-based results turns out to be fortuitous; RIT and SCK•CEN results are closer and the
PSI/CEA value is similar to that obtained by ANL. The two JENDL-based results lie in a similar
range.
2.6 Isotopic composition at end of irradiation
From the submitted results (not shown in this paper), significant discrepancies in the isotopic
390

composition at the end of irradiation are observed. The results seem only partially correlated with the
nuclear data libraries used.
Before analysing the isotopic composition of the irradiated fuel, it is interesting to calculate the
fraction of heavy metals that fissioned after five years of irradiation. Surprisingly, the values are
strongly different and can be grouped into two classes characterised by a high “burn-up” (KAERI,
RIT and SCK•CEN, all using the JEFF library) and a low “burn-up”. The former values are between
21.3% and 22% and the latter between 18.5% and 18.8%. The discrepancy in the total number of
heavy metals at EOL is therefore considerable (more than 15%) and is not fully understood.
As for the isotopic compositions at the end of irradiation, a clear dependence on the “burn-up” is
observed. In comparison with the other participants, RIT, JAERI and SCK•CEN (high burn-up) report
lower values for burned-up isotopes during the irradiation, such as 237 Np, 241 Am and 243 Am and higher
values for build-up isotopes (in the start-up core only), such as 238 Pu, 242 Cm and 244 Cm.
The discrepancies are important not only for the minor actinides such as 237 Np, 241 Am and 243 Am
(up to 10% relative to the average) but also for the Pu isotopes. Large differences are observed
especially for 238 Pu, 240 Pu and 241 Pu where the results of SCK•CEN strongly deviate from the others.
Some discrepancies in the results are also related to differences in the branching ratios or to
different treatment of some reactions. For example, the differences observed in the 242 Cm and 242m Am
concentrations are due to discrepancies in the branching ratios used for the (n,γ) reaction of 241 Am.
Most participants used the 0.2/0.8 values, whereas RIT used 0.225/0.775, PSI/CEA 0.15/0.85 and
SCK•CEN 0.09/0.91. Consequently RIT gives the highest and PSI/CEA and SCK•CEN the lowest
concentration of 242m Am. As expected, an opposite effect is observed on the 242 Cm. Discrepancies are
observed in the concentrations of 236 U and 235 U in the KAERI and SCK•CEN results. In the start-up
core, very low concentrations of these isotopes are reported, whereas in the equilibrium core their
results are in acceptable agreement with the other participants. This probably indicates a different
treatment of the (n, 2n) reaction for 237 Np, which should be thoroughly investigated.
2.7 Safety parameters
Coolant void reactivity calculations are traditionally difficult. Only integral values for the
coolant void reactivity from k eff
difference calculations are available. The JENDL-based calculations
by CIEMAT and JAERI give similar results (agreement within about 10%). Assuming that the
Monte Carlo calculations by CIEMAT and RIT can be considered as reference calculations which are
only sensitive to differences in nuclear data, the decrease in the BOL coolant void reactivity arising
from the substitution of JENDL by JEFF is about 30% (26% for the start-up core and 32% for the
equilibrium core). For the other JEFF-based coolant void reactivity predictions, one observes a
maximum deviation of 30% with respect to the RIT prediction. It is interesting to notice that, in the
ANL and JAERI case, the voided core becomes supercritical.
A general observation on the calculated Doppler reactivity is that it represents an almost “zero
effect” on the system. Knowing that the magnitude of the Doppler reactivity in fast spectrum systems
was demonstrated with an uncertainty of ±15%, it is difficult to make a comparison of these small
values dispersed around zero. Nevertheless, since the Doppler reactivity comes essentially from
capture reactions in the thermal energy region, a more thorough insight into the calculation
procedures used by the participants, especially energy self-shielding treatment in their calculations,
would be necessary to understand the origin of discrepancies among them.
391

The isotope specific β eff
values calculated from the two libraries (ENDF/B-VI and JEF-2.2) are in
good agreement. However, ANL (ENDF/B-VI) gives a total β eff
value smaller than that of JAERI
because the contributions of 242m Am, 243 Am and Cm isotopes were not taken into account (the delayed
neutron data for these isotopes are not available in ENDF/B-VI used by ANL). JAERI used the
delayed neutron data of 242m Am, 243 Am and 245 Cm isotopes from JENDL-3.2. PSI/CEA took into
account the contribution of all the isotopes; its results therefore give the largest value among the
results from ANL, JAERI and PSI/CEA. The Monte Carlo calculations give larger β eff
values. RIT
used the delayed neutron data based on ENDF/B-VI, but did not consider 242m Am, 243 Am and Cm
isotopes, whereas CIEMAT used the delayed neutron data based on JENDL-3.2 in its calculations. In
general, it is seen that the results of the β eff
calculations depend on the accuracy of the delayed neutron
data used.
3. Conclusions
Seven organisations contributed to this benchmark exercise using different basic data libraries
and reactor analysis codes and applying both deterministic and Monte Carlo methods. The analysis of
the results shows significant discrepancies in important neutronic parameters, such as k inf
, k eff
and
burn-up reactivity swing. Strong discrepancies appear also in the estimation of the external neutron
source, i.e. a parameter which determines the requested accelerator power.
As demonstrated by a separate parametric study, the impact of the different basic nuclear data on
these integral parameters is important but it is not sufficient to fully explain the discrepancies
observed in the results. In future benchmark exercises which may be based on an experimental result,
attention should therefore be given to both the data processing route and the neutron transport
approximations. Concerning the burn-up calculations, attention should be given to the treatment of
the fission products and to the actinide decay chains noting that in minor actinide burner cores
different isotopes are involved compared to MOX-fuelled cores.
Acknowledgements
The authors express their gratitude to all the participants who devoted their time and effort to this
benchmark exercise.
392

REFERENCES
[1] OECD/NEA NSC Task Force on Physics Aspects of Different Transmutation Concepts, JAERI
Proposal of Benchmark Problem on Method and Data to Calculate the Nuclear Characteristics
in Accelerator-based Transmutation System with Fast Neutron Flux, NEA/NSC/DOC(1996),
10 April 1996.
[2] H. Takano et al., Benchmark Problems on Transmutation Calculation by the OECD/NEA Task
Force on Physics Aspects of Different Transmutation Concepts, Proc. Int. Conf. on the Physics
of Nuclear Science and Technology, 5-8 October 1998, Long Island, USA, 1462.
[3] Calculations of Different Transmutation Concepts: An International Benchmark Exercise,
OECD/NEA report, NEA/NSC/DOC(2000)6, February 2000.
[4] B.C. Na, P. Wydler and H. Takano, OECD/NEA Comparison Calculations for an Acceleratordriven
Minor Actinide Burner: Analysis of Preliminary Results, NEA-NSC Workshop on
Utilisation and Reliability of High Power Accelerators, 22-24 November 1999, Aix-en-
Provence, France, (To be published).
[5] J.D. Kim and C.S. Gil, KAERI, Private Communication.
[6] H. Takahashi and H. Rief, Concepts of Accelerator Based Transmutation Systems, Proc. of the
Specialists’ Meeting on Accelerator-based Transmutation, 24-26 March 1992, PSI Villigen,
Switzerland, 2-26.
393

SESSION IV
BASIC PHYSICS, MATERIALS AND FUELS
SUB-SESSION IV-B:
MATERIALS
395

STAINLESS STEEL CORROSION IN
LEAD-BISMUTH UNDER TEMPERATURE GRADIENT
Dolores Gómez Briceño, Fco. Javier Martín Muñoz, Laura Soler Crespo,
Federico Esteban Ochoa de Retana, Celia Torres Gurdiel
Nuclear Fission Department, CIEMAT
Av. Complutense 22, Madrid 28040, Spain
Abstract
Austenitic steels can be used in ADS in contact with liquid lead-bismuth at temperatures below 400ºC.
At higher temperatures, martensitic steels are recommended. However, at long times, the interaction
between structural material and eutectic leads to the dissolution of some elements of the steel in the
liquid metal. In a non-isothermal loop, the material dissolution takes place at the hot leg and, due to
mass transfer, deposition occurs at the cold leg. Formation of oxide layers on structural materials
improves its performance. F82Hmod. and 2¼ Cr-Mo steels have been tested in a small natural
convection loop built of austenitic steel (316L), which has been operating for 3 000 hours. During all
the operation, a gas with 10 ppm oxygen content has been bubbling in the hot area. The obtained
results show that an oxide layer is formed on the samples introduced in the loop at the beginning of the
operation and this layer increases with time. However, the samples introduced at intermediate times
are not protected by oxide layers and present different levels of attack.
397

1. Introduction
Due to its excellent physic-chemical and nuclear characteristics, lead-bismuth has been proposed
both as a coolant and a spallation target for hybrid systems, so called accelerator driven systems
(ADS) [1]. However, heavy liquid metals, and particularly lead-bismuth, present a high corrosivity to
most of structural materials.
Austenitic steels may be used in a hybrid system in contact with liquid lead-bismuth if the operating
temperature is not beyond 400ºC. For higher temperatures, martensitic steels are recommended [2].
However, with long operation times, the interaction between the structural material and the eutectic
leads to the solution of some elements of the steel (Ni, Cr and Fe, mainly) in the liquid metal. In a nonisothermal
lead bismuth loop, the material dissolution takes place at the hot leg of the loop and, due to
mass transfer, deposition occurs at the cold leg. The available experience, proceeding from the Former
Soviet Union, shows that one of the possible ways to improve the performance of structural materials
in lead-bismuth is the formation and maintenance of a protective oxide layer, which would constitute a
barrier between the liquid metal and the steel.
2. Experimental
Tests have been performed in a small thermal convection loop built of austenitic stainless steel
type 316, containing a lead-bismuth volume of 1.2L. A scheme of the loop is shown in Figure 1. The
maximum temperature is of 550ºC with a temperature gradient between 50 and 100ºC. Thermocouples
placed in several points of the loop, embedded in the lead-bismuth, were used to control the loop
temperatures. The eutectic and the gas for controlling the atmosphere were introduced into the loop
from an expansion tank placed on top of it.
Figure 1. Thermal convection loop
Pb-Bi
Ar + O 2
Cold leg
Corrosion
specimens
Cold leg
Hot leg
Hot leg
Drain tank
Cylindrical specimens (10 mm length and 7 mm of diameter) of martensitic steel F82Hmod. and
low alloy steel 2¼ Cr-Mo, named P22, have been tested. The materials composition is shown in
398

Table 1. Specimens were inserted and removed at several times from the test zone placed at the hot leg
(500ºC) during loop operation, without stopping the flowing of lead-bismuth, according with the
scheme shown in Figure 2. In this scheme, samples A, B, C and D correspond to F82H steel and P, Q,
R and S correspond to P22 steel. Some samples have been in the loop during all the operation (A 3
and
P 3
) and others have lived different times and periods of loop life. During the tests, a flow of 40 cc/min
of argon with 10 ppm of oxygen was bubbling in the hot area of the loop.
Table 1. Materials composition
% weight F82Hmod 2¼ Cr-1Mo
Fe Bal. Bal.
Cr 7.750 2.250
Ni 0.015
Al 0.004
Mo 0.010 1.000
C 0.100 0.090
Si 0.230 0.200
Ta 0.005
Ti 0.004
Mn 0.160 0.450
Nb

Figure 2. Tests scheme
0 340 h. 1 030 h. 1 990 h.
3 022 h.
A 1 , P 1
A 2 , P 2
A 3 , P 3
340 h.
1 030 h.
3 022 h.
B, Q
690 h.
C 1 , R 1
C 2 , R 2
960 h.
1 992 h.
D, S
1 032 h.
Figure 3. Temperature evolution during loop life
A 1
P 1
A 2
P 2
A 3
P 3
P 1 Q
Q
A 2
A 1 B
P 2
B
C 1
R 1
C 2
R 2
R 1 S
S
A 3
P 3
C 2
R 2
C 1 D
D
600
550
500
450
400
350
300
250
200
0
83
164
243
329
399
483
565
647
727
809
890
971
1051
1133
1217
1300
1406
1486
1566
1644
1727
1832
1913
1957
2038
2119
2201
2287
2371
2455
2541
2622
2760
2846
2934
3013
Te mperature (ºC)
T1
T2
T3
T4
T5
T6
T7
Time (hours)
3. Results
The temperature evolution in different points of the thermal convection loop is represented in
Figure 3. Specimens insertion and removing times are also indicated in this figure. According with the
temperature registration, loop operation can be divided in four steps. During all the steps, the
400

thermocouples placed in the hot zone of the loop present uniform values with a light decrease over
time. However, the cool zone temperature oscillates during the first step and stabilises at the beginning
of the second step. From this level, it decreases 50ºC, and it soars to a higher level than the existent at
the beginning of the second step. All this process occurs in a time bracket of 1 000 hours of operation.
During the third step, the temperature decreases again, reaching lower minimum values than in the
previous case, after 1 700 hours of operation. During the last step, the temperatures stabilise again
until the end of the loop operation, keeping a temperature gradient between the hot and the cool zones
of 80ºC aprox. Temperature decreases may be due to slags formation hindering lead-bismuth flowing.
Sudden minimum temperature increases in the cool zones may be due to slag dissolution. A constant
flow of gas was kept during all the operation.
As mentioned above, three different steels were examined, the steel samples (F82H and P22) and
the structural material (316L austenitic steel). All the samples were examined by optical and scanning
electron microscopy and, in some cases, Auger spectroscopy was carried out to obtain the depth
profile composition of the oxide layers formed on the sample surfaces.
In general, the specimens of martensitic steel F82H mod. have shown a slightly better behaviour
than the 2¼ Cr-Mo specimens although the evolution is the same for both steels. This means that both
types of steel formed an oxide layer when the samples were tested from the beginning of the operation
and both presented attack at intermediate states.
The six specimens (type A and P) incorporated into the loop at the beginning of the operation
show a good corrosion resistance. For F82Hmod., all the specimens are covered by an homogenous
oxide layer, with minor spalling areas in samples A 2
and A 3
. Type A 1
specimens, tested for 340 hours,
present a thin oxide layer of 2.5-3.5 µm, formed by two sub-layers. The outer layer, of 1-1.5 µm
thickness, is iron oxide and the inner layer is formed by iron, chromium and oxygen with a chromium
concentration higher than its value in the alloy, as can be seen in the Auger depth profiles, Figure 4. A
slight depletion of chromium in the underlying alloy was detected. For 2¼ Cr-Mo, P 1
specimens
present thicker oxide layers with a wrinkled aspect that let the lead-bismuth penetrate and be placed
between the oxide layer and the steel (Figure 5). Due to the thickness of oxide layers formed on
2¼ Cr-Mo specimens, it was not possible to carry out Auger analysis in most of the samples.
401

Figure 4. Auger depth profile concentrations.
A 1
specimens, 340 hours
Figure 5. Oxide layers on P 1
specimens, 340
hours
X 100
100
120
90
80
area 1
Fe
100
A.
C.
%
70
60
50
O
80
60
40
30
Fe
40
20
Cr
Cr
10
Pb
O
0
0 1 2 3 4 5 6 7
DEPTH, microns
20
0
X 500
100
90
area 2
Fe
100
90
80
80
A.
C.
%
70
60
50
40
30
Fe
O
70
60
50
40
30
20
10
Pb
Cr
Cr
20
10
0
0 1 2 3 4 5 6 7
0
DEPTH, microns
Type A 2
specimens (F82Hmod.), tested for 1 030 hours, show an oxide layer very similar in
composition and thickness to the ones detected in A 1
specimens. However, A 3
specimens (F82Hmod.),
tested for 3 022 hours, from the beginning to the end of the loop operation, present an oxide layer with
a thickness of 20 µm but with similar composition to A 1
and A 2
specimens. Figure 6 shows that in A 3
specimens the outer layer is formed by iron and oxygen (magnetite), whereas the inner layer has iron,
chromium and oxygen ,
with the higher concentration of chromium placed in the interface between the
inner layer and the alloy. In all A specimens, the outer layer composition corresponds to magnetite
whereas inner layer is an Fe(Fe 2-x
)Cr x
O 4
. Lead-bismuth eutectic (named lead in all the figures) is
incorporated to the outer layer in all A specimens. All type P specimens (2¼ Cr-Mo) present an oxide
layer with the same characteristics as P 1
sample. Figure 7 shows the comparison between A 3
and P 3
specimens.
402

Figure 6. Auger depth profile concentrations.
A 3
specimens, 3 022 hours
Figure 7. Appearance of A 3
and P 3
specimens, 3 022 hours
A 3 specimens
P3specimens
× 10
× 10
100
e
10 0
90
area 1
90
80
80
70
60
O
70
60
A. C. %
50
40
Fe
50
40
30
30
20
10
Pb
Cr
Cr
20
10
0
0
0 5 10 15 20 25 30
DEPTH, m ic rons
× 200
× 200
10 0
90
ar ea 2
Fe
100
90
80
80
70
70
A. C. %
60
50
40
30
20
Fe
60
50
40
30
20
× 100
× 100
10
Pb
Cr
Cr
10
0
0
0 5 10 15 20 25 30
DEP TH, mi c ro ns
Both present a quite high corrosion resistance, but P 3
present a thicker and slightly less protective
oxide layer.
Type B (F82Hmod.) and Q (2¼ Cr-Mo) specimens were introduced into the loop after 340 hours
from the operation beginning, and removed 690 hours later. These specimens present a general
solution in most of the surface with a depth attack up to 50 µm for F82H sample and 60 µm for
2¼ Cr-Mo. Small areas covered by a very thin oxide layer have been detected. Auger analysis shows a
single oxide layer formed by chromium and oxygen. Type C (F82Hmod.) and R (2¼ Cr-Mo)
specimens inserted into the loop after 1 030 hours from the operation beginning were removed
960 hours (C 1
and R 1
specimens) and 1992 hours afterwards (C 2
and R 2
specimens). C 1
samples present
a deeper attack than R 1
, which present only a slight attack. C 2
samples also present a worse behaviour
than R 2
, but both show general corrosion with a depth attack up to 140 µm .
Small surface areas covered
by a thin oxide layer are still visible. The thickness and the composition of this oxide layer are similar
to the observed in type B specimens, Figure 8.
Type D (F82Hmod.) and S (2¼ Cr-Mo) specimens were introduced into the loop 2 000 hours
after the beginning of the operation and removed at the end of the tests together with A 3
and P 3
specimen. No dissolution was observed in these samples. Auger analysis shows a single thin layer of
less than 1 µm covering the entire specimen, with a high chromium concentration and without iron,
Figure 9. In this sample, lead is incorporated to the oxide layer rich in chromium. In this case,
S specimen behaves slightly better than D.
403

Figure 8. Auger depth profile concentrations. C 2
specimens, 1 992 hours
A. C. %
100
Fe
90 area 2
80
70
60
50
40
30
A. C. %
100
90
80
70
60
50
40
30
20
Cr
20
10
10
b
0
0
0 0,2 0,4 0,6 0,8
DEPTH, mi crons
100
100
90 are a 1
Fe
90
80
80
70
70
60
60
50
50
40
40
30
30
Cr
20
20
Cr
10
b
O
10
0
0
0 0,2 0,4 0,6 0,8
DEPT H, m icro ns
Figure 9. Auger depth profile concentrations. D specimens, 1 032 hours
A. C. %
100
100
Fe
90 ar ea 1
90
80
80
70
70
60 O
60
50
50
40
40
30 Pb
30
20
20
Cr
Cr
10
10
O
0
0
0 1 2 3 4 5 6 7 8 9 10 11 12
DEPTH, microns
100
90
area 2
Fe
100
90
80
80
A. C . %
70
60
50
40
O
70
60
50
40
30 Cr
30
20
10
Pb
Cr
20
10
0
0
0 1 2 3 4 5 6 7 8 9 10 11 12
DEPTH, microns
After 2 000 hours from the beginning of the operation a lead-bismuth sample was taken out from
the loop in order to measure impurities. This sample had 265 ppm nickel, 5.4 ppm chromium and
17 ppm iron. Iron concentration is higher than iron solubility in eutectic (3.6 ppm), while nickel and
404

chromium concentration values are lower than their solubility values at 530ºC (30 212 and 13.8 ppm
respectively).
After the end of the tests, the loop was cut to be destructively examined. The structural austenitic
steel presented material solution, especially at the hottest areas. Figure 10 shows a cut of the coldest
area of the loop in which particle deposition was observed. This deposition was also detected at the
upper corner of the cold leg. Two different kinds of particles were identified. The round particles
closer to the steel wall are formed by chromium, iron and oxygen and the particles with geometrical
shape are an intermetallic compound of nickel and bismuth.
Figure 10. Slags deposition detected in the cold leg loop
Cold leg
Ar + O 2
Cold leg
Hot leg
Hot leg
Wall loop, 316L SS
Samples of solidified lead–bismuth from the hot and cold zones were taken to measure oxygen by
LECO. A heterogeneous oxygen distribution was observed in the samples. At the core, 2 ppm oxygen
were observed in the hot zone, and 1 ppm oxygen in the cold zone. Near the walls, 9 ppm oxygen were
measured in the hot zone and 6 ppm oxygen in the cold zone.
4. Discussion
Liquid metal corrosion depends on the solution rate and the solubility value of the solid metal in
the liquid metal. Lead alloys, and in particular lead-bismuth eutectic, show a higher agressivity to the
structural materials than the alkali liquid metals. In a static system, solution of the solid metal occurs
until the solubility value of the main elements of the alloy is reached. In a dynamic system with
temperature gradient, a mass transfer process occurs. The dissolved material at the hot zone is
405

deposited at the cold zone, and the elements concentration in the liquid metal represents a steady–state
balance between the rates of solution and precipitation in different zones of corroding systems [3].
For austenitic stainless steels, it is well documented (for example, Gorynin et al. [2]) that corrosion
resistance is determined by oxygen thermodynamic activity in lead and lead alloys. For oxygen
concentrations lower than the concentration at which the dissolved oxygen is in equilibrium with the
spinel formation on the steel surface materials, dissolution occurs. This equilibrium value is about
5 × 10 -8 wt% at 550ºC. For concentrations higher than this value, materials oxidation appears. Steel
composition has also a significant influence in the corrosion/protection of structural materials. In
general, it is accepted that in absence of oxygen, carbon steels present a higher corrosion resistance
than high chromium steels, which suffer attack [4]. However, results obtained by Gorynin [13] point
out that at 460ºC, in flowing Pb-Bi with oxygen concentration less than 10 -7 wt%, 1Cr-MoV alloy
presents higher corrosion rate than 18Cr-10Ni-Ti alloy, whereas 1Cr-2Si-Mo is more resistant to
corrosion than both the previous ones. For high oxygen activity the corrosion of iron–chromium alloys
is higher than for iron whereas for low oxygen activity the alloy presents higher resistance corrosion.
Formation of oxide layer on the structural materials can prevent alloy dissolution.
A general consideration of the results seems to point out that the oxygen content in the loop is not
enough to form, in general, protective oxide layers on the specimens placed in the hot zone of the
loop. In fact, specimens B, C (F82H) and Q, R (2¼ Cr-Mo) present material dissolution and no oxide
layers. However, in the type A (F82H) and P (2¼ Cr-Mo) specimens tested from the beginning of the
operation, an appreciable growth of the oxide layer was detected. It seems that A and P samples are in
oxidation condition whereas the rest of the specimens would be in dissolution conditions. This
observation would mean that, although the gas is bubbling during all the operation time with the same
flow rate and the same oxygen content (10 ppm), the oxygen concentration in the liquid lead-bismuth
varies along the operation time and so does the corrosion behaviour of the samples.
Type A specimens (F82H) inserted into the loop at the beginning of the operation are covered by
an homogeneous double oxide layer. The thickness of this oxide layer reaches 20 µm after
3 022 hours. The growth of this oxide layer is almost inappreciable between 340 and 1 030 hours, and
then it grows significantly up to 3 022 hours. These results seem to be in accordance with the general
behaviour accepted for the oxidation of iron-chromium alloys and stainless steels [5]. After an initial
protective period, a sudden rate increase occurs once the break-away time has been reached. This stage
is often followed by a further rate reduction by a self-healing process. This last step has not been
observed in our tests, probably due to their short duration. The oxide layers of type A specimens show
similar characteristics. The inner oxide layer, Fe (Fe 2-x
)Cr x
O 4
, seems to be protective enough to prevent
lead-bismuth penetration. No eutectic is incorporated to this layer contrary to the observed in iron
oxide outer layer. Eutectic concentration in the outer layer decreases and disappears in the interface
outer/inner oxide layer.
Type P specimens (2¼ Cr-Mo) inserted into the loop at the beginning of the operation are covered
by a non-adherent oxide layer with a wrinkled aspect. Labun et al. [6] describe a layer with these
characteristics for a 3Cr-Fe steel oxidised in dry oxygen at 700-800 ºC. They say that the wrinkling is
due to a substantial growth of this layer in the lateral direction relative to the underlying layers. In our
case, lead-bismuth penetrates through the external oxide layer and is place between it and the steel.
The layer formed on 2¼ Cr-Mo specimens is much thicker in all the cases that the formed on type A
specimens. Fedirko et al. [7] tested Armco iron, Fe-16Cr and Fe-16Cr-1Al in stagnant liquid lead at
600ºC with some amount of lead oxide to get a concentration of 10 -5 wt% oxygen and they mention in
this work that the formed oxide layer is thicker for the Armco iron. The information on the oxidation
behaviour of stainless steels in liquid lead-bismuth is very scarce. However, it is accepted than the
available information on stainless steels oxidation in molten lead can be useful to analyse the materials
behaviour in lead-bismuth.
406

For high oxygen activity, alloys with a high chromium content present a better behaviour than
low chromium alloys. This is due to the different characteristics of the protective oxide layer based on
formation of oxide rich in chromium, which is hindered for the low chromium alloys. At the same
time, free and low chromium alloys are able to form thicker oxide layers, which are accepted to be less
protective [14]. For low oxygen activity in the melt, alloys with low chromium content present a
higher corrosion resistance [4], since chromium has a higher solubility than iron in lead-bismuth. This
seems to be the case of Q, R (2¼ Cr-Mo) specimens comparatively to B, C (F82H) specimens
although the behaviour of both steels is not conclusively different.
Fedirko et al. [8] also found that spinel rich in chromium can protect iron-chromium alloys in
molten lead. After 1 000 hours of testing, four different zones could be observed on the transverse
micro-sections of Fe-16Cr-1Al. The first zone was a porous oxide layer with a high concentration of
iron and lead and a low concentration of chromium and aluminium. This zone was formed by
magnetite and lead. Lead concentration decreased with distance from the surface, but on the interface
of the first and the second zone lead was present. The second zone was a continuous oxide film
enriched with chromium and aluminium. The composition of this film was a Fe 3
O 4
-FeCr 2
O 4
solid
solution with the spinel structure. Lead was not present in the interface between the second and third
zone. A process of internal oxidation with chromium and aluminium oxides mainly along grain
boundaries was detected in the third zone. Finally, the fourth zone had the matrix composition but
grain boundaries also contained oxide compounds. On the contrary to the observed in oxygen
saturated melts, spinel rich in chromium is not permeable to lead in solutions with low oxygen
activity.
In addition, in static tests performed in oxygen saturated molten lead at 520ºC, Benamati et al. [9]
observed the formation of continuous layers of reaction products with an average thickness of 20 µm
after 2 000 hours and of 40 µm after 3 700 hours in martensitic steel F82H mod. specimens. The
reaction product layers were formed by two distinct sub-layers of Me 3
O 4
. Only the inner one contained
chromium in addition to iron. Lead was detected in the outer sub-layer of all the product layers. These
authors point out that oxide layers formed in oxygen saturated liquid lead containing an additional
source of oxygen in the form of lead oxide are not protective, since the oxide layer thickness increases
along with the exposure time.
On the other hand, Müller et al. [10] found oxide layers of similar characteristics to the observed
in type A specimens in samples of a martensitic steel with 9.99 Cr (OPTIFER) tested in stagnant liquid
lead at 550ºC under controlled Ar-H 2
/H 2
O atmosphere containing 8 × 10 –6 at% oxygen. Müller
observed a corrosion attack with three different zones. The outer layer consisted of magnetite without
appreciable chromium concentration. The middle layer was formed by Cr-Fe spinel, with a chromium
concentration lower than the detected in the A 3
specimens. Finally, an internal oxygen diffusion zone
in which oxides precipitate along the grain boundaries could be observed. After 3 000 hours, the
thickness of magnetite and spinel were 20 and 15 µm respectively whereas the depth of diffusion zone
was 10 µm.
According to Fedirko et al. [7], in liquid lead with low activity of oxygen the formation of the
external oxide layer based on Fe 3
O 4
is very low and significant amounts of active alloying elements
are accumulated in more internal layers. The lack of mobility of chromium through the magnetite
lattice explains the formation of the spinel layer. Faster diffusing elements like manganese and iron
will pass through to the outer layer, while slower diffusing elements like chromium will be oxidised
without movement and remain in the inner layer. Both magnetite and spinels M 3
O 4
are based in a
close-packed cubic sub-lattice of oxygen ions in which the metals ions are the faster moving species
and determine the oxidation rates. Thus, in martensitic steels, chromium moves more slowly than iron.
The outer layer is formed by magnetite and the inner layer consists of iron, chromium spinels. The
407

inner layer grows inward from the original metal surface. The oxygen transport through this layer can
take place via pores within the inner layer, since solid state diffusion of oxygen through the magnetite
lattice is too low [11]. The Auger analysis results on F82H specimens point out that the composition of
the outer oxide layer fits well with magnetite composition, and the inner oxide layer corresponds to a
spinel Fe(Fe 2-x
)Cr x
O 4
. Spinel rich in chromium is the typical oxide layer on stainless steels with Cr
concentration lower than 13% when they are oxidised at high temperature. Müller et al. [10] consider
that no main differences exist between the oxide layers formed in lead and in an oxidation process in a
controlled furnace atmosphere at the same temperature.
The thin oxide layers detected on sample D (F82Hmod.) and S (2¼ Cr-Mo) are formed by oxygen
and chromium with a minor amount of iron. No chromium depletion was observed in the underlying
alloy. In spite of having a high chromium concentration, these layers allow the eutectic penetration,
whose concentration is maximum in the oxide/alloy interface and that apparently penetrates in the base
alloy underneath the oxide/alloy interface. No external iron oxide was detected in these samples. The
lack of an external iron oxide layer can be a consequence of two causes, not necessarily exclusive. On
the one hand, oxygen activity in lead-bismuth may be lower than the necessary to form magnetite but
high enough to form oxide of more noble elements like chromium. The low oxygen activity is
supported by the dissolution process observed in the specimens inserted into the loop at intermediate
times (B, Q and C, R specimens) and by the loop structural material dissolution and is questioned by
the growth of oxide in type A and P specimens placed into the loop at the beginning of the test. On the
other hand, an iron concentration higher than its solubility in lead-bismuth was measured in an eutectic
sample taken out during the loop operation. This high iron concentration can hinder iron diffusion and
prevent magnetite formation.
Regarding loop operation, nickel, chromium and iron dissolved in the hot area and deposited at
the cold zone play an important role in the loop operation and may also be of significant importance
with respect to changes of oxygen concentration in lead-bismuth along the loop life. Horsley et al. [11]
point out that if the concentration of iron in the melt is higher than the solubility of iron in bismuth at
the maximum temperature of the loop, then the rate of mass transfer is controlled by the precipitation
rate in the cold region. On the one hand, the dissolution of these steel elements into the melt provokes
the beginning of a plug formation at the coldest zones of the loop and explains the decreases of
temperatures in these areas. On the other hand, these elements take the oxygen dissolved in the melt,
since the particles found at the cold corners of the loop are formed by iron, chromium and oxygen.
This would explain the behaviour of samples at intermediate times, which present attack, due to a
insufficient oxygen concentration to form a protective oxide layer. As it was shown in Figure 3, at
certain times of loop operation, especially after 2 000 hours, a sudden increase of temperatures of the
cold zone is observed. This may be due to slag dissolution and may provoke oxygen liberation, which
could explain the formation of a thin oxide layer on D and S specimens. In spite of this hypothesis, at
present we do not have a plausible explanation for the observed behaviour of similar specimens tested
for the same period at different loop operation times. However, the growth of oxide layer in A and P
specimens simultaneously with the dissolution in B, Q and C, R specimens point out that lower
oxygen activity is necessary for oxide layer growth than for oxide layer formation.
408

6. Conclusions
The results obtained in this experimental work point out that oxide layer protection of martensitic
steels in lead-bismuth under temperature gradient is possible in determined conditions. A double oxide
layer composed by a magnetite outer layer and a spinel Fe(Fe 2-x
)Cr x
O 4
inner layer, very rich in
chromium, is formed. The spinel is non-permeable to lead-bismuth and constitutes a barrier between
base material and liquid metal. Higher oxygen activity seems to be necessary for oxide formation than
for oxide layer growth.
In the tested conditions the corrosion resistance of martensitic steel F82Hmod. and of low alloy
steel 2 ¼ Cr-Mo have been very similar, in spite of the lower chromium concentration of the later.
However, the oxide layers formed on F82Hmod. seem to be no-permeable to lead-bismuth whereas the
ones formed on 2 ¼ Cr-Mo are thicker and permeable to the eutectic
It is possible to operate a natural convection loop for long times up to 3 000 hours. However,
impurities in the melt, coming from structural material dissolution, interfere with oxide layer
formation on the samples tested, and lead to a difficult interpretation of the obtained results.
REFERENCES
[1] J.J. Park et al., Review of Liquid Metal Corrosion Issues for Potential Containment for Liquid
Lead and Lead-bismuth Eutectic Spallation Targets as a Neutron Source, Nuclear Engineering
and Design, 2000, 196, pp. 315-325.
[2] I.V. Gorynin, G.P. Karzov, V.G. Markov, V.A. Yakovlev, Structural Materials for Atomic
Reactors with Liquid Metal Heat-transfer Agents in the Form of Lead or Lead-bismuth Alloy,
Metal Science and Heat Treatment, 1999, Vol. 41, No. 9-10.
[3] J.R. Weeks and C.J. Klamut, Liquid Metal Corrosion Mechanisms, Proceedings of the Conference
on Corrosion of Reactor Materials International Atomic Energy Agency, Vienna 1962.
[4] Ning Li, Active Control of Oxygen in Molten Lead-bismuth Eutectic Systems to Prevent Steel
Corrosion and Coolant Contamination, 1999, Report from Los Alamos National Laboratory.
[5] G.C. Woods, The Oxidation of Iron-chromium Alloys and Stainless Steel at High Temperature,
Corrosion Science, 1961, Vol. 2, pp. 173-196.
[6] P.A. Labun et al., Micro-structural Investigation of the Oxidation of an Fe-3 Pct Cr Alloy,
Metallurgical Transactions A, 1982, Vol. 13ª, pp. 2103-2112.
[7] V.M. Fedirko, O.I. Eliseeva, V.L. Kalyandruk and V.A. Lopushans’kyi, Effect of Admixtures of
Oxygen on the Oxidation of Iron and Fe-Cr Alloys in Lead Melts, Materials Science, 1977,
Vol. 33, No. 3.
409

[8] V.M. Fedirko, O.I. Eliseeva, V.L. Kalyandruk, V.A. Lopushans’kyi, Corrosion of Armco Iron
and Model Fe-Cr-Al Alloys in Oxygen- Containing Lead Melts, 1977, Materials Science,
Vol. 33, No. 2.
[9] G. Benamati, P. Buttol, V. Imbeni, C. Martini and G. Palombarini, Behaviour of Materials for
Accelerator Driven Systems in Stagnant Molten Lead, Journal of Nuclear materials 279, 2000,
pp. 308-316.
[10] G. Muller, G. Schumacher and F. Zimmermann, Investigation on Oxygen Controlled Liquid
Lead Corrosion of Surface Treated Steels, Journal of Nuclear Materials, 2000, 278 pp. 85-95.
[11] J. Robertson, High T Corrosion of Stainless Steel, Corrosion Science, 1991, Vol. 32, No. 4,
pp. 443-465.
[12] G.A. Horsley and J.T. Maskrey, The Corrosion of 21/4% Cr-1%Mo Steel by Liquid Bismuth,
Journal of the Iron and Steel Institute, 1958, pp. 139-148.
[13] I.V. Gorynin, G.P. Karzov, V.G. Markov, V.S. Lavrukhin,V.A. Yakovlev, Structural Materials
for Power Plants with Heavy Liquid Metals as Coolants, Conference on Heavy liquid Metals
Coolants in Nuclear Technology, HLMC-98, Obninsk 1998.
[14] G.C. Wood, The Oxidation of Iron-chromium Alloys and Stainless Steels at High Temperatures,
Corrosion science, 1961, Vol. 2, pp. 173-196.
410

ACCUMULATION OF ACTIVATION PRODUCTS IN
PB-BI, TANTALUM, AND TUNGSTEN TARGETS OF ADS
A.S. Gerasimov, G.V. Kiselev, A.I. Volovik
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
Data on new radionuclide production in three types of target, Pb-Bi, tantalum, and tungsten target of
ADS are presented in this paper. The irradiation by neutrons produced in blanket and in the target
itself do not take into account proton irradiation. The change of isotopic composition, accumulation
of new radionuclides, and radiation characteristics (activity, radiotoxicity is water, and radiation dose
power) are calculated.
411

1. Introduction
One of the main parts of the accelerator driven system (ADS) is the neutron-producing target. It
can be made of solid heavy metal such as tantalum or tungsten or liquid metal such as lead-bismuth.
All these materials are typical for a neutron-producing target. ADS target is irradiated by accelerated
protons, by high-energy neutrons from the target itself, and by low energy neutrons from the subcritical
blanket surrounding the target. The average energy of neutrons from the target itself is about
several MeV. A flux density of neutrons from the blanket on the target is of the order of 10 13 cm -2 s -1
for common-type power blanket with thermal neutrons and can reach several units times 10 15 cm -2 s -1
for high flux blanket. Under influence of target irradiation, there are nuclide conversions causing the
change of target isotopic composition and radioactive nuclei production.
The change of isotopic composition, accumulation of new radionuclides, and radiation
characteristics (activity, radiotoxicity by water, and radiation dose power) caused by external
neutrons from blanket and by internal neutrons from the target itself are calculated. The influence of
protons on side nuclide production should be calculated separately and is not considered in the paper.
In calculating nuclide conversions by thermal neutrons, reaction rates A i
are taken using values
of thermal neutron cross-sections σ i
and resonance integrals I i
of nuclides [1], A i
= (σ i
+γI i
)Φ, where Φ
is neutron flux density, γ is neutron spectrum hardness showing a ratio of epithermal to thermal
neutrons. Value γ = 0.4 (spectrum typical for light-water thermal-neutron blanket) is considered. For
irradiation by internal neutrons from target, monoenergetic neutrons with energies 10 MeV are
considered [2,3]. The average high-energy neutron flux density is over the volume of a target about
10 15 cm -2 s -1 at energy of protons from accelerator 1 GeV and beam current 10 mA. Corresponding
thermal neutron flux from blanket is about 10 14 cm -2 s -1 . It was accepted that the target has the form of a
continuous cylinder with a diameter of 50 cm. Specific activity Q (Ci/g), radiotoxicity RT (litre/g)
and radiation dose power QΓ (R⋅cm 2 /g⋅hr) are defined by sums on all radioactive nuclides included in
a target:
Q = Σ Q i
RT = Σ RT i
, RT i
= Q i
/ MPA i
,
QΓ = Σ QΓ i
where MPA i
– maximum permissible activity of the given nuclide i in water determined by the
modern Russian radiation safety standard [4], Γ i
– gamma-constant of the nuclide i. They are referred
to 1 gram of a target.
2. Pb-Bi target irradiation
The initial target is Pb-Bi eutectic containing 44.5% Pb and 55.5% Bi with natural isotopic
composition. In Tables 1 and 2, nuclide concentration and radiation characteristics of a target are
presented. For low energy neutrons, three Φ values and γ = 0.4 are considered. Concentrations of
nuclides are normalised by 0.445 nuclei of lead and 0.555 nuclei of bismuth in initial target. Only the
most important nuclides are submitted in the tables. Radiation characteristics – activity Q, radiotoxicity
RT and radiation doze power QΓ are referred to 1 gram of a target. Irradiation time T = 1 year.
In irradiation by thermal neutrons, capture cross sections of nuclides in target are very small. So,
effects of thermal self-blocking of Pb-Bi target are not essential. Initial nuclide burning is negligible.
412

An important radionuclide determining the radiation characteristics of a target is 210 Po. Because of
alpha-decay of this nuclide, radiotoxicity is high. However, radiation doze power is not so great as for
other target materials because of low gamma-radiation of 210 Po. In high-energy neutron irradiation,
210
Po is the main radioactive nuclide, and 204 Tl gives also small contribution to radiation
characteristics.
Table 1. Nuclide concentration and radiation characteristics of Pb-Bi target
irradiated by external low energy neutrons from blanket
Nuclide
Initial
T = 1 yr
concentrations Φ=10 13 cm -2 s -1 Φ =10 14 cm -2 s -1 Φ=10 15 cm -2 s -1
204
Pb 0.0064 6.4-3 6.4-3 6.1-3
205
Pb 0 2.95-6 2.94-5 2.88-4
206
Pb 0.107 0.107 0.107 0.107
207
Pb 0.0983 0.0983 0.0981 0.0959
208
Pb 0.233 0.233 0.233 0.235
209
Bi 0.555 0.555 0.555 0.553
210
Po 0 6.32-6 6.31-5 6.30-4
Q, Ci/g – 0.0289 0.289 2.88
RT, litre/g – 8.90 + 9 8.90 + 10 8.88 + 11
QΓ, (R cm 2 /g hr) – 1.54-3 0.0154 0.153
Table 2. Nuclide concentrations and radiation characteristics
of a target irradiated by 10 MeV neutrons
Nuclide
Initial
T = 1 yr
concentrations ϕ = 10 15 cm -2 s -1 ϕ = 10 16 cm -2 s -1
204
Hg 0 2.38-10 2.38-9
204
Pb 0.0064 0.0064 0.0064
204
Tl 0 1.11-7 1.11-6
206
Pb 0.107 0.107 0.107
207
Pb 0.0983 0.0983 0.0983
208
Pb 0.233 0.233 0.233
209
Bi 0.555 0.555 0.555
210
Po 0 1.61-5 1.61-4
Q, Ci/g – 0.0735 0.753
RT, litre/g – 2.27 + 10 2.27 + 11
QΓ, (R cm 2 /g hr) – 4.19-3 4.19-2
The comparison of radiation characteristics caused by neutrons from the target itself and
neutrons from external blanket is based on the assumption that a high energy neutron flux density of
10 15 cm -2 s -1 corresponds to a neutron flux from external blanket of 10 14 cm -2 s -1 . The radiotoxicity caused
by neutrons from the target itself is about 3 times less, and radiation dose power is 4 times less than
the same characteristics caused by neutrons from an external blanket. This result is important as it
shows a rather high role of neutrons from the target itself in the process of accumulation of those
radionuclides which define the main radiation characteristics of the irradiated target.
413

3. Tantalum target irradiation
The initial target is made of natural tantalum. In Table 3, nuclide concentration and radiation
characteristics are presented for low energy neutron irradiation. For Φ = 10 14 cm -2 s -1 , value γ = 0.1 and
for Φ = 10 15 cm -2 s -1 value γ = 0 are considered with thin target, so effects of self-blocking are weak.
Radiation characteristics are determined by 182 Ta and, at some extend, by 185 W.
Table 3. Nuclide concentration and radiation characteristics of tantalum target
irradiated by external low energy neutrons from blanket
Nuclide
Initial
T = 1 yr
T = 0.5 yr
concentrations Φ =10 14 cm -2 s -1 Φ=10 15 cm -2 s -1
181
Ta 1.0 0.761 0.724
182
Ta 0 4.40-3 1.04-3
182
W 0 9.56-3 1.13-3
183
W 0 0.210 0.252
184
W 0 0.015 0.021
185
W 0 2.62-5 1.37-4
Q, Ci/g – 27.8 7.84
RT, litre/g – 1.1+10 2.7+10
QΓ, (R cm 2 /g hr) – 1.85+5 4.37+4
4. Tungsten target irradiation
In Table 3, partial nuclide introduction to radiation characteristics is shown for Φ = 10 14 cm -2 s -1 .
In Table 4, concentration of nuclides and radiation characteristics of a target are presented for low
energy neutron irradiation. Concentrations of nuclides are normalised by one nucleus of natural
tungsten, radiation characteristics by 1 gram of a target. For neutron flux Φ = 10 14 cm -2 s -1 , value
γ = 0.4 is considered, and for Φ = 10 15 cm -2 s -1 , γ = 0 and T = 0.5 years is taken. Target diameter 50 cm.
Only the most important nuclides are submitted.
Table 3. Radiation characteristics of tungsten target
irradiated by low energy neutrons from blanket with Φ=10 14 cm -2 s -1
Nuclide Q, Ci/g RT, litre/g QΓ, (R cm 2 /g hr)
181
W 0.703 1.4 + 7 150
185
W 5.68 6.5 + 8 1.6
187
W 21.9 – 5.3 + 4
188
W 0.297 1.7 + 8 3.3
182
Ta 0.703 2.8 + 8 470
183
Ta 0.0486 1.6 + 7 80
186
Re 1.46 5.8 + 8 140
188
Re 2.30 – 730
Total 32.4 1.7 + 9 5.5 + 4
414

Nuclide
Table 4. Nuclide concentration and radiation characteristics of
tungsten target irradiated by external neutrons from blanket
Initial
concentrations
T = 1 yr
T = 0.5 yr
Φ = 10 14 cm -2 s -1 Φ = 10 15 cm -2 s -1
180
W 0.126-2 9.6-4 1.3-3
181
W 0 1.2-4 6.6-6
182
W 0.263 0.26 0.26
183
W 0.143 0.14 0.14
184
W 0.306 0.31 0.31
185
W 0 6.0-4 7.1-5
186
W 0.286 0.28 0.28
187
W 0 3.1-5 2.4-5
188
W 0 2.9-5 2.0-7
182
Ta 0 1.1-5 4.0-9
183
Ta 0 3.5-7 3.6-10
186
Re 0 7.8-6 6.7-8
188
Re 0 2.3-6 3.5-7
Q, Ci/g – 32.4 18.1
RT, litre/g – 1.7 + 9 8.5 + 7
QΓ, (R cm 2 /g hr) – 5.5 + 4 4.1 + 4
Nuclide
Table 5. Nuclide concentrations and radiation characteristics
of a tungsten target irradiated by 10 MeV neutrons
Initial
concentrations
T = 1 yr
T = 0.5 yr
ϕ = 10 15 cm -2 s -1 ϕ = 10 16 cm -2 s -1
180
W 1.3-3 1.4-3 2.6-3
181
W 0 4.2-3 0.030
182
W 0.263 0.26 0.25
183
W 0.143 0.15 0.17
184
W 0.306 0.29 0.24
185
W 0 4.0-3 0.03
186
W 0.286 0.27 0.22
187
W 0 3.0-8 2.5-7
182
Ta 0 1.0-7 1.9-6
186
Re 0 3.2-7 1.0-6
188
Re 0 4.7-12 2.0-10
Q, Ci/g – 62.2 486
RT, litre/g – 5.1 + 9 4.1 + 10
QΓ (R cm 2 /g hr) – 3.9 + 3 9.9 + 4
In low energy irradiation, self-blocking effects are very high. For this reason, production of new
nuclides is low in purely thermal spectrum even in high flux 10 13 cm -2 s -1 . For neutron spectrum typical
for light water blanket, with γ = 0.4, nuclide production is higher because of epithermal neutrons.
Radiotoxicity at γ = 0.4 is determined by all radioactive nuclides. At γ = 0, 185 W gives a major part to
radiotoxicity. Radiation doze power is determined by short-lived 188 W.
415

At 10-MeV neutron irradiation, radiotoxicity is determined by 181 W, 185 W, 184 Re, and radiation
dose power by 184 Re.
Comparison of these data shows that a radiotoxicity caused by neutrons born in a target is
3 times greater and radiation dose power is 14 times less than the same characteristics caused by low
energy neutrons from external blanket. However, it is necessary to mention that radiation
characteristics caused by neutrons from an external blanket are defined by short-lived nuclides 187 W,
188
Re with half-life periods about 1 day, whereas at irradiation by neutrons born in target, main
contribution come from nuclides 185 W, 181 W, and 184 Re with half-life periods from 38 up to 121 days. If
we consider only radionuclides with half-life periods not less than several tenths of days, then it
appears that high-energy neutrons born in a target make radiation dose power 6 times greater than
neutrons from an external blanket.
5. Conclusion
Comparison of radiation characteristics produced in irradiation by low energy neutrons from
surrounding blanket and high energy neutrons from a target shows a rather high role of neutrons from
the target itself in the process of accumulation of those radionuclide which define the main radiation
characteristics of the irradiated target.
Absolute values of radiation characteristics allow estimating necessary modes of irradiated target
management. For tantalum and tungsten targets, radiation dose power is rather high and decreases
slowly at cooling. Radiotoxicity value for all considered targets is close to that of radioactive waste of
nuclear reactors. It should be recommended to store irradiated targets after some cooling while taking
measures as for middle and high radioactive waste management.
REFERENCES
[1] S. Mughabghab, Neutron Cross-sections. Vol. 1, Part B. Academic Press, N.Y.-London, 1984.
[2] V. McLane et al., Neutron Cross-sections, Vol. 2, Neutron Cross-section Curves, National
Neutron Data Centre, BNL. N.Y., Acad. Press, 1988.
[3] ENDF/B-VI, Revision 2.
[4] Radiation Safety Standards (NRB-99), Minzdrav of Russia, Moscow, 1999.
416

THERMAL AND STRESS ANALYSIS OF HYPER TARGET SYSTEM *
T.Y. Song, N.I. Tak, W.S. Park
Korea Atomic Energy Research Institute
P.O. Box 105 Yusung, Taejon, 305-600, Republic of Korea
J.S. Cho, Y.S. Lee
Department of Nuclear Engineering, Seoul National University
Shinlim-dong 56-1, Kwanak-gu, Seoul, 151-742, Republic of Korea
J.H. Choi, M.K. Song
Department of Mechanical Engineering, Gyeongsang National University
Kajoa-dong 900, Chinju, Kyongnam, 660-701, Republic of Korea
Abstract
HYPER (HYbrid Power Extraction Reactor) is the accelerator driven transmutation system which is
being developed by KAERI (Korea Atomic Energy Research Institute). We plan to finish the
preliminary design of HYPER by 2001. Pb-Bi is used as the coolant and target material of HYPER.
One of the issues related to the HYPER target system is the thermal and mechanical loads imposed on
the Pb-Bi and the beam window. We used LCS (LAHET Code System) to calculate heat generation.
FLUENT was used for thermal-hydraulic calculation, and finally stress calculation was performed by
ANSYS. A beam condition such as current varied. The initial velocity of Pb-Bi also varied.
*
This work has been supported by the Korea Ministry of Science and Technology (MOST).
417

1. Introduction
HYPER (HYbrid Power Extraction Reactor) is the accelerator driven transmutation system
designed by KAERI (Korea Atomic Energy Research Institute) [1]. An accelerator driven system
provides the possibility of reducing plutonium, minor actinides, and environmentally hazardous
fission products from the nuclear waste coming from the conventional nuclear power plant. In
addition, it can be used to produce electricity. HYPER is designed to transmute TRU and fission
products such as 99 Tc and 129 I.
Because an accelerator driven system is a sub-critical reactor, external neutrons should be
provided by a target system inside the reactor. HYPER adopts Pb-Bi as the coolant and target
material, which are not separated. Some key issues related to developing target system are window
and Pb-Bi cooling, corrosion, radiation damage etc. Corrosion and radiation damage degrade the
performance of the beam window, and an experimental study is necessary to understand the change
due to those damages. In this paper, we use simulation codes to determine the target geometry and
beam conditions under which HYPER target system can be operated with stability before corrosion
and radiation damage affect the beam window.
We studied the basic thermal hydraulic characteristics of the target system using FLUENT code,
and we also used ANSYS code [2] to calculate the stress of the beam window. The heat generation
inside beam window and Pb-Bi was calculated using LCS (LAHET Code System) [3].
2. Double window target
Figure 1 shows the structure of the target area and beam window geometry. HYPER beam
channel is cylindrical and located at the centre of the reactor with a 50-cm diameter. The window is
designed to have 2-mm thick steel layers, and Pb-Bi coolant flows between the two layers for window
cooling. The gap width of the coolant channel is about 4 mm. The cross-section of the beam tube is
30 × 30 cm 2 and the window has a cylindrically curved profile.
Figure 1. Target area and beam window geometry
Pb-Bi
Beam
Pb-Bi
9Cr-2WVTa
30 cm
15 cm
Pb-Bi
418

The target Pb-Bi is coming from the bottom of the beam channel and the beam is injected from
the top. The Pb-Bi flow is slowed just below the centre of the beam window. Therefore Pb-Bi is
forced to flow from left to right between windows.
For the beam window material, 9Cr-2WVTa was chosen since advanced martensitic/ferritic
steels are better in Pb-Bi corrosion than austenitic steels and do not show a DBTT problem [4]. The
yield strength of 9Cr-2WVTa is about 600 MPa at 400 o C.
3. Calculation conditions
The cylindrical forced convection target system is set to be 50 cm in diameter and 100 cm in
height. The bottom of the beam window is located 25 cm below the top. The initial temperature of
Pb-Bi is set to be 340 o C for both target and window cooling Pb-Bi. The initial velocity of window
cooling Pb-Bi is 6 m/s. We separated the double window calculation from the single window
calculation to simplify the calculation geometry.
The heat deposition in Pb-Bi and window is calculated using the LAHET Code System. The
beam is assumed to have a circular shape with a diameter of 10 cm and a parabolic density
distribution. The result shows that about 52% of the total beam energy are deposited as heat in the
target zone. Figure 2 shows the heat deposition as a function of the radius from the beam centre and
distance from the target surface for Pb-Bi and the window in the case of a 20 mA beam.
Figure 2. Heat deposition rate for a 20mA beam
Proton Beam
Injection
(× 10 9 W/m 3 )
10cm 9cm 8cm 7cm 6cm 5cm 4cm 3cm 2cm 1cm
10cm 0.02 0.03 0.06 0.12 0.26 1.94 4.86 7.14 8.73 9.64
10cm 0.05 0.07 0.12 0.19 0.42 1.39 3.01 4.30 5.23 5.68
10cm 0.05 0.07 0.10 0.19 0.39 0.92 1.63 2.26 2.71 2.98
10cm 0.04 0.06 0.10 0.18 0.31 0.55 0.82 1.08 1.29 1.36
10cm 0.04 0.06 0.09 0.14 0.22 0.32 0.42 0.50 0.60 0.64
Proton Beam
Injection
(×10 9 W/m 3 )
10cm 9cm 8cm 7cm 6cm 5cm 4cm 3cm 2cm 1cm
2mm 0.004 0.006 0.009 0.019 0.071 1.90 5.29 7.20 9.09 9.64
4. 2-D calculation
The general CFD code FLUENT was used to simulate the two-dimensional thermal and flow
distribution of the liquid target. Calculation analyses are performed in two-dimensional axi-symmetry
cylindrical geometry. The calculation parameter is the liquid Pb-Bi inlet velocity which vary from
1.1 m/s to 2.0 m/s. The beam current and velocity of cooling Pb-Bi are fixed to 20 mA and 6 m/s
respectively. In this calculation, the surfaces are set to be adiabatic boundaries. For the calculation of
419

this study, we used an orthogonal co-ordinate transformation. This transformation is performed using
the grid generation components of the FLUENT code.
Figure 3 shows the calculation geometry of the FLUENT code. The centre of the Pb-Bi beam
channel is narrowed to increase the velocity of up-coming Pb-Bi so that the efficiency of window
cooling is maximised.
In the single window calculation, the maximum temperature of the window is 2 277 o C for an
inlet velocity of 1.1 m/s and 1 808 o C for an inlet velocity of 2.0 m/s in the steel beam window region.
These temperatures exceed the beam window melting temperature. Therefore, this is not allowable for
the steel window. Figure 4 shows the temperature distribution and the velocity vector profile of the
target and the window region for an inlet velocity of 2.0 m/s. Table 1 shows the calculation result of
the maximum temperature for each case. Based on these calculations, the beam window can be
damaged by high temperature and thermal stress. So, the independent cooling system for the beam
window must be considered.
Figure 3. FLUENT calculation geometry for the single and double window
420

Figure 4. Temperature and velocity distribution of the single window
Table 1. Maximum temperature for the single window
Bottom inlet velocity
(m/s)
Max. temp.
( o C)
1.1 1.35 1.5 2.0
2 277 2 086 2 003 1 808
The coolant flowing direction is x-directional co-ordinate and the vertical direction of the coolant
flowing is the y-directional co-ordinate. The spacing of the y-directional cells is 1 mm. We used the
same heat generation rates for the inner and outer windows as shown in Figure 2. The heat transfer of
the lower surface of lower window is treated by the heat transfer coefficient obtained by the
calculation of the single window.
Maximum temperatures in the double windows are presented in Table 2. The maximum
temperature in the inner window reaches 1 580 o C. This value is higher than the window melting
temperature. The inlet velocity of the Pb-Bi coolant flowing in the narrow channel must be larger than
6 m/s. As a result, the maximum temperature in the lower window reaches 1 077 o C for a 1.1 m/sec
bottom inlet velocity of the single window and 927 o C for a 2.0 m/s bottom inlet velocity.
Table 2. Maximum temperature for the double window
Bottom inlet velocity
(m/s)
Upper window
( o C)
Lower window
( o C)
1.1 1.35 1.5 2.0
1 580 1 580 1 580 1 580
1 077 1 027 997 927
421

Figure 5. Temperature distribution of the double window case
5. 3-D calculation
Figure 6 shows target geometry for FLUENT 3-D calculations. The beam channel is assumed to
be rectangular to simplify the calculation. We first calculated the temperature of the bottom Pb-Bi and
the single window part and then we calculated the double window part separately. In 3-D calculations,
the velocity of up-coming Pb-Bi is fixed to be 2 m/s and 3 different beam currents are used, which are
2 mA, 10 mA and 20 mA. The velocity of Pb-Bi flowing between windows is 6 m/s.
422

Figure 6. Target geometry for 3-D calculation
3URWRQÃ%HDP
2XWOHW
2 XWOHW
%HDP Ã: LQGRZ
3E%LÃ7 D UJ HW
,QOHW
The maximum temperature of the single window was calculated to be 571 o C in the case of a
2 mA beam. In the same case, the maximum temperature of the upper and lower window are found to
be 464 and 429 o C, respectively. Tables 3 and 4 show the maximum temperatures of the single and
double window for 3 different beam currents.
Table 3. Maximum temperature for the single window
Beam current
(mA)
Max. temp
( o C)
2 10 20
571 1 497 2 657
Table 4. Maximum temperature for the double window
Beam current
(mA)
Upper window
( o C)
Lower window
( o C)
2 10 20
464 958 1576
429 767 1187
After using FLUENT to produce a 3-D temperature distribution for the window, the result is
transferred to ANSYS to calculate the thermal stress of the upper and lower window. Figure 7 shows
the results of the ANSYS calculation. The maximum Von Mises thermal stresses are respectively
about 185 and 97 MPa for the upper and lower beam window. In the stress calculation, the beam
shape is assumed to be rectangular for the simplicity of calculation.
423

Figure 7. Thermal stress distribution of the upper window for a 2 mA beam
6. Conclusion
The thermal hydraulic and stress analysis of the liquid Pb-Bi target and beam window have been
presented in this paper. Based on 2-D and 3-D analysis, temperature and velocity distributions were
studied using FLUENT. We used ANSYS to calculate the stress of the beam window. A double
window system was introduced to enhance the window cooling. The velocity of Pb-Bi flowing
between two windows is set to be 6 m/s. When the beam current and velocity of up-coming Pb-Bi are
2 mA and 2 m/s respectively, the maximum temperature and thermal stress of the beam window were
calculated to be 464 o C and 185 MPa. Our target system is not a separate system, but a part of the
whole sub-critical reactor. Therefore, the environment of the target should be considered to finalise
the temperature and stress distribution. We are also considering the case of single window with a
shape of hemisphere.
REFERENCES
[1] W.S. Park et al., Development of Nuclear Transmutation Technology, KAERI/RR-1702/96, 1996.
[2] ANSYS User’s Manual for Revision 5.0.
[3] R.E. Prael et al., User Guide to LCS: The LAHET Code System, LA-UR-89-3014, 1989.
[4] Y. Dai, Proceedings of the International Workshop on the Technology and Thermal Hydraulics
of Heavy Liquid Metal, 6.27-6.39, 1996.
424

SESSION IV
BASIC PHYSICS, MATERIALS AND FUELS
SUB-SESSION IV-C:
FUELS & TARGETS
425

FUEL/TARGET CONCEPTS FOR TRANSMUTATION OF ACTINIDES
A. Fernández, D. Haas, R.J.M. Konings, J. Somers
European Commission, Joint Research Centre, Institute for Transuranium Elements
P.O. Box 2340, 76125 Karlsruhe, Germany
Abstract
Four different concepts for fuels and targets for transmutation of (minor) actinides are discussed in the
present paper. These include thorium-based mixed oxides, inert matrix mixed oxide, and composites
based on mixtures oxide powders (CERCER) or mixtures of oxide and metal powders (CERMET).
Fabrication methods have been investigated, especially taking account of the specific requirements for
handling significant quantities of minor actinides (dust-free processes, remote handling). The
processes tested at ITU are based on sol-gel and infiltration (INRAM) techniques or combination
thereof. The processes are being validated first using cerium and then plutonium as simulants for the
minor actinides, before the actual fabrication of Am- and Cm-containing materials begins in earnest
following the completion of the construction of specially designed shielded cells (the MA-lab).
427

1. Introduction
Various fuel cycle concepts for partitioning and transmutation (P&T) of actinides are under
discussion at present. In an evolutionary strategy, fast reactors are introduced in which the minor
actinides are mixed with the plutonium in a mixed oxide, either uranium-based or eventually thoriumbased.
In such a multiple recycle scenario, the possibility to reprocess the fuel is of key importance. In
a radial strategy, dedicated “transmuters” such as accelerator driven systems are introduced, with the
aim to eliminate plutonium and minor actinide in a separate “second stratum” [1]. Dedicated fuel types
are considered for this second stratum, which are characterised by a high minor actinide content and a
high extent of transmutation. The latter can be achieved best if uranium is omitted from the fuel, as
breeding of transuranium elements is avoided. This is especially important when the second stratum is
a once-through process.
For the uranium-free fuels a wide variety of alternative matrix materials are considered, true inert
material and “quasi” inert materials based on thorium. The actinide phase and the matrix can be
combined in a homogeneous fuel form in which the actinides form a solid solution with the matrix,
well known from the uranium-plutonium mixed oxide fuels. However, most inert-matrix mixed oxides
of this type are generally characterised by a relatively low thermal conductivity. To overcome this,
composite fuel forms are considered in which the matrix (a ceramic or a metal) improves the thermal
properties of the fuel.
At the Institute for Transuranium Elements (ITU) some of these fuel and target options for
transmutation of actinides are being studied, with emphasis on clean and, thus, dust-free fabrication
methods for minor actinides (specifically americium and curium). A process consisting of a
combination of SOL-GEL and infiltration techniques [2,3] is being developed and used for the
fabrication of the following oxide-based fuel/target forms in a pellet-type fuel packing:
• Thorium-based mixed oxide (THOMOX), with the actinides incorporated as a solid solution
in a ThO 2
matrix.
• Inert-matrix mixed oxide (IMMOX), where the actinides are incorporated in an yttriastabilised
zirconia (YSZ) solid solution.
• Ceramic-ceramic composite (CERCER), in which the actinides are in (near) spherical YSZ
particles which are dispersed in a MgO matrix.
• Ceramic metal composite (CERMET), again with the actinides in spherical YSZ particles,
which are dispersed in Mo, Fe or Zr matrices.
Currently these processes are being tested and validated with cerium and plutonium. The actual
fabrication of Am- and Cm-containing materials will be performed in shielded cells, which are under
construction at present. These cells, which form a complete fabrication chain (the MA-Lab), are
equipped with gamma- and neutron shielding in the form of lead (50 mm) and water (500 mm).
2. Thorium-based mixed oxide
The fabrication of thorium based mixed oxide fuels and targets for transmutation can be achieved
directly using the sol-gel method to give the (Th,MA)O 2
product. This process has been validated at
the ITU for the Uranium analogues in the SUPERFACT and TRABANT irradiation experiments in
Phénix and HFR Petten, respectively [4,5]. The particular advantage of the process lies in the wide
range of actinide content, which can be obtained. In the SUPERFACT experiment, for example,
(U 0.55
Np 0.45
)O 2
and (U 0.6
Np 0.2
Am 0.2
)O 2
targets were prepared, while in the TRABANT1 experiment
428

(U 0.55
Pu 0.4
Np 0.05
)O 2
fuels were prepared. The main disadvantage of the sol-gel method lies in the
aqueous liquid wastes produced.
If only relatively small quantities of minor actinide are required (up to 20 mol %), the infiltration
(INRAM) procedure, described in detail below for yttria-stabilised zirconia (YSZ) targets, offers an
interesting alternative to the sol-gel technique. Due to its partial solubility in weak acid solutions, UO 2
does not readily satisfy one of the basic requirements of the INRAM process. In contrast, ThO 2
is
ideally suited, and, given its low activity, microspheres thereof can be produced with limited operator
shielding. In a further variation of the process, (Th,Pu)O 2
spheres can be manufactured in conventional
glove boxes and infiltrated with minor actinide nitrate solutions to give (Th,Pu,MA)O 2
products. At
the ITU initial investigations have been performed on the production of such (Th,Pu)O 2
microspheres.
These investigations will be continued and the process tested with the infiltration of uranyl or cerium
nitrate solutions, before being validated in the minor actinide laboratory.
3. Inert-matrix mixed oxide
The fabrication process for the inert-matrix mixed oxide fuel/targets is shown schematically in
Figure 1. Yttria-stabilised zirconia spheres are produced by a sol-gel process. Feed solutions with a
determined Zr/Y ratio are prepared from Zr and Y oxychloride salts. Following addition of a surface
active agent and an organic thickener, the solution is dispersed into droplets by a rotating cup
atomiser. The droplets are collected in an ammonia bath, where gelation occurs. After ageing, the
resulting spheres are washed, dried using azeotropic distillation procedure, and calcined at 1 123 K.
These spheres have a polydisperse size distribution in the 40 and 150 µm range, and their porosity is
about 80%. X-ray diffraction indicates that they have a cubic crystal structure with a measured lattice
parameter of 514.0 ± 0.3 pm, which is in agreement with the published value for (Zr 0.85
Y 0.15
)O 1.93
(513.9
± 0.1 pm)[6].
Figure 1. Schematic representation of the fabrication process of the
inert-matrix mixed oxide fuels/targets using the infiltration method.
A photograph of the polydisperse spheres is shown on the left.
Zr,Y nitrate solution
Actinide solution
Droplet to particle
conversion
Calcination
40-150 µm
microspheres
Solution infiltration
Thermal treatment
Pressing
429
Sintering
Once calcined the spheres are then infiltrated with a lanthanide (as simulant) or actinide solution.
They are thermally treated to convert the infiltrated phase to the corresponding oxide. If higher

quantities of lanthanide/actinide are required this sequence of infiltration – calcination steps can be
repeated. The metal content can be determined simply by gravimetric analysis of the beads before
infiltration and after the calcination step. The resulting beads are free-flowing (at least for metal
contents up to 40 wt%), and can be pressed directly into pellets, following addition of zinc stearate as
lubricant. First tests have been made with Ce-nitrate solutions with concentrations of 200 and 400 g·l -1 ,
and second tests with Pu-nitrate solutions with a concentration of 200 g·l -1 , which for Pu is the
maximum obtainable without risk of polymerisation and precipitation. In the case of Am, however,
concentrations of up to 400 g·l -1 can be obtained also.
The results of the studies with cerium nitrate (200 g·l -1 ) indicate that about 40 wt% cerium can be
infiltrated into the beads if 5 consecutive infiltration steps are used. X-ray diffraction measurements of
the materials after sintering at 1923K for 6 hours in air showed that they have a cubic crystal structure,
with the lattice parameter increasing with increasing CeO 2
content as described by Vegard’s law
(Figure 2).
Figure 2. Left: lattice parameter of the (Zr,Y,Ce)O 2-x
spheres as a function of the Ce content;
the solid line represent the Vegard’s law for the (Zr,Y)O 2-x
-CeO 2
solid solution.
Right: geometric density of the pellets as a function of the infiltrated metal concentration.
Lattice Parameter (pm)
524
520
516
Density (gcm -3 )
6.5
6.0
5.5
5.0
Theoretical
Geometrical (Air)
Geometrical (Ar/H 2
)
512
0 0.1 0.2 0.3 0.4
0 10 20 30 40 50
Mole Fraction CeO 2
Ce Content (Wt%)
The density of the pellets manufactured in this way decreases with increasing CeO 2
content (for
constant compaction pressure), whereas the theoretical density increases (Figure 2). A ceramograph of
a (Zr,Y,Ce)O 2-x
pellet with 10wt% Ce (Figure 3) indicates it is devoid of cracks or other defects. The
experiments with Pu, which are presently in progress, show comparable results. The density decrease
may be related to the known change in mechanical properties of stabilised zirconia with increasing
extent of stabilisation. It might be overcome by modifying the compaction pressure and the sintering
atmosphere; such studies have been initiated.
430

Figure 3. Left: Ceramograph of a (Zr,Y,Ce)O 2
pellet with 10% Ce.
Right: Ceramograph of a (Zr,Y,Ce)O 2-x
-MgO composite;
the volume fraction of (Zr,Y,Ce)O 2-x
is 20%.
1000 µm
4. Ceramic-ceramic composite
CERCER composite pellets of (Zr,Y,Ce)O 2
-MgO were fabricated by mixing (Zr,Y,Ce)O 2
spheres
with MgO granules, using Zinc stearate as lubricant. The pellets were sintered at 1923K for 6 hours in
air. The (Zr,Y)O 2
spheres were fabricated by the rotating cup, as described above, and the 80-90 µm
and 125-140 µm fractions were selected, by sieving, for infiltration with a Ce-nitrate solution with a
concentration of 400 g·l -1 . A commercial MgO powder (CERAC M-1017) was granulated (size
fraction 50200 – 80 µm) and either used directly or mixed with the original fine powder. Tests on
MgO pellets without YSZ macrospheres showed that by pressing the different mixtures of granules,
densities of greater than 95% of the theoretical value were obtained in all cases.
The resulting products in these experiments did not prove satisfactory, which is in contrast to
previous work on (Zr,Y,Ce)O 2
- MgAl 2
O 4
composite pellets [7]. The (Zr,Y,Ce)O 2
-MgO CERCERs
always exhibit many cracks in the pellets, predominantly between the spheres (Figure 3). To
understand this behaviour, dilatometer (Netsch DIL 402) measurements of the sintering behaviour of
on (Zr,Y)O 2
, (Zr,Y,Ce)O 2
spheres, MgO and MgAl 2
O 4
were performed. The results shown in Figure 4
indicate distinct differences in the densification of the various materials. It is clear that MgO densifies
much more (DL/L = 22.0%) than the (Zr,Y,Ce)O 2-x
spheres (DL/L = 14.7%) and, moreover, the
densification starts at a somewhat lower temperature. In contrast, the extent of densification of
MgAl 2
O 4
is significantly less (DL/L = 18.0%) and starts at temperatures above that of the (Zr,Y,Ce)O 2-
x spheres. This implies that the sintering behaviour of the two powders in the (Zr,Y,Ce)O 2 -MgO
CERCER are so distinct so that cracks are difficult to avoid. This would require an extensive
investigation to manipulate the properties of both powders to match another one. Tests on calcining
the (Zr,Y,Ce)O 2
at lower temperature proved unsuccessful, as it appeared that the organic additives
were not sufficiently removed from the spheres. Preliminary tests on calcining the MgO granules at
higher temperatures (up to 1273 K) were also unsuccessful as the sintered pellets show multiple
cracks. A solution thus must be found by modification of the sintering properties of the MgO powder.
Alternatively, one could consider the reduction of the size of the beads, which will reduce the stresses
during sintering also. This will, however, have a penalty with respect to the undamaged volume
fraction in the composite during its irradiation in a nuclear reactor.
431

Figure 4. Densification of MgAl 2
O 4
, (Zr,Y,Ce)O 2-x
, (Zr,Y)O 2-x
,
and MgO as a function of temperature.
0
MgAl 2
O 4
(Zr,Y,Ce)O 2
(Zr,Y)O 2
MgO
DL/L
-10
-20
700 900 1100 1300 1500 1700
Temperature (°C)
5. Ceramic-metal composite
CERMET composite pellets of (Zr,Y,Ce)O 2
-Mo were fabricated by compaction of a mixture of
the cerium infiltrated sieved fraction (112 – 125 µm) of the beads and commercial Mo powder (Merck)
with addition of zinc stearate as lubricant. The pellets were sintered at 1923K in Ar/H 2
.
The densities of the (Zr,Y,Ce)O 2
-Mo CERMETS were typically in the order of 90-92% of the
theoretical value. Visual inspection of the pellet surface and the ceramographs of sectioned samples
show that they have perfect cylindrical geometry and excellent integrity without macro or micro
cracks (see Figure 6). Nevertheless due to the fact that there is a very distinct difference in hardness of
the two materials, many spheres were pulled out of the matrix during polishing, resulting in “apparent”
porosity in the pellet. A typical ceramograph of a (Zr,Y,Ce)O 2
-Mo is shown in Figure 6. It should also
be noted that there is excellent physical contact between the Mo matrix and the macrospheres so that
the maximum benefit of the high thermal conductivity of the Mo is achieved.
432

Figure 6. Left: ceramograph of a(Zr,Y,Ce)O 2-x
-Mo CERMET; the amount of (Zr,Y,Ce)O 2-x
is about
20 vol%. Right: Detailed image of the sphere-matric interface in a (Zr,Y,Ce)O 2-x
-Mo CERMET.
6. Conclusions and outlook
The results presented in this paper have demonstrated the feasibility of the fabrication of
homogeneous zirconia- and thoria-based fuels for transmutation of minor actinides, using liquid
processes such as sol-gel and infiltration. Though our experiments have been made with cerium and
plutonium as simulants of americium, the previous experience with these techniques obtained in the
SUPERFACT and EFTTRA experiments [2,4] gives confidence that they can be extended to
americium without difficulties. The sol-gel technique offers the highest flexibility with respect to the
actinide content in the material, but it produces significant amount of active liquid waste, which could
be recycled in an industrial process. The active liquid waste is minimal when the infiltration technique
is used, but the actinide content that can be obtained is limited. Moreover, the properties of the
infiltrated YSZ powder change with increasing amount of infiltrant, resulting in an increase of the
porosity with increasing infiltrated metal content. Though this is favourable to manage the helium
accumulation typical for MA fuels [8], it will lead to a decrease of the thermal conductivity. Especially
for ZrO 2
-based material, where the thermal conductivity is already a limiting factor for its application,
this may be unacceptable. Experiments to investigate the cause of these low densities have therefore
been started.
The results presented in this paper for the (Zr,Y,An)O 2
-MgO composite fuels are not promising. It
seems technically difficult, if not impossible, to obtain fault-free composite pellets when the dispersed
phase consists of spherical particles with a size of greater than 100 µm, which is required, if the
radiation damage to the matrix is to be minimised [9]. The effect of reducing the size of the dispersed
particles on the fabrication process will be tested. Smaller particles would lead to greater damage of
the matrix during irradiation, but this penalty could be tolerated, if the improvement of the thermal
behaviour is the decisive requirement.
Promising results have been obtained for (Zr,Y,An)O 2
-Mo composite fuels, but the Mo matrix is
not the prime candidate for CERMET fuel. Therefore the study of steel-based CERMET fuels will be
initiated.
433

REFERENCES
[1] R.J.M. Konings and J.L. Kloosterman, A View of Strategies for Transmutation of Actinides.
Progr. Nucl. Energy, in press (IMF’6 Special Issue).
[2] A. Fernandez, K. Richter, J. Somers. Preparation of Spinel (MgAl 2
O 4
) Spheres by a Hybrid Solgel
Technique. 9th Cimtec- World Ceramics Congress. Advanced in Science and Technology
15. Ceramics: Getting into the 2000’s. Part C (1999).
[3] K. Richter, A. Fernandez, J. Somers, Infiltration of Highly Radioactive Materials: A Novel Approach
to the Fabrication of Transmutation and Incineration Targets. J. Nucl. Mater., 1997, 249, 121.
[4] N. Chauvin and J.-F. Babelot, Rapport de synthèse commun CEA/ITU sur l’expérience
SUPERFACT 1, CEA Cadarache, Note Technique SDC/LEMC 96-2028 (1996).
[5] J. Somers, J.P. Glatz, D. Haas, D.H. Wegen, S. Fourcaudot, C. Fuchs, A. Stalios, D. Plancq,
G. Mühling, Status of the TRABANT Irradiation Experiments, in Proceedings of the International
Conference on Future Nuclear Systems, Global’99, Jackson Hole, Wyoming, 1999.
[6] JCPDS, International Centre for Diffraction Data, 30-1468, (1996).
[7] N. Boucharat, A. Fernández, J. Somers, R.J.M. Konings, D. Haas, Fabrication of Zirconiabased
Targets for Transmutation, Progr. Nucl. Energy, in press (IMF’6 Special Issue).
[8] R.J.M. Konings, R. Conrad, G. Dassel, B. Pijlgroms, J. Somers, E. Toscano, J. Nucl. Mat., in
press (also published as EUR Report 19138 EN (2000)).
[9] N. Chauvin, R.J.M. Konings, H.J. Matzke, Optimization of Inert Matrix Fuel Concepts for
Americium Transmutation. J. Nucl. Mat., 1999, 274, 105-111.
434

AMERICIUM TARGETS IN FAST REACTORS
S. Pilate 1 , A.F. Renard 1 , H. Mouney 2 , G. Vambenèpe 2
1
Belgonucléaire, Brussels
2
Électricité de France, Paris & Lyon
Abstract
It would be advantageous to irradiate americium targets up to very large burn-ups, so as to throw
them away to waste without any further reprocessing. Former studies considered the recycling of
americium in conventional LWRs. But even after long irradiation times the incineration rate of
americium, of the order of 50%, was not sufficient to avoid reprocessing and re-fabrication of targets.
In fast reactors, americium targets could be placed in special assemblies filled with a neutron
moderator material which enhances Am incineration. Recent calculations show that a burn-up of 90%
in the targets could be reached upon affordable irradiation times.
A major problem remains the choice of the inert support material of the targets. So far, the support
materials currently envisaged have exhibited too large swelling rates. A type of support material
virtually non swelling would be in the form of coated particles, like those irradiated without damage
up to very high burn-ups in high temperature reactors. Studies are being launched on the feasibility of
such Am targets with coated particles in fast reactors.
435

1. Introduction
It is a hard challenge in the nuclear fuel cycle to simultaneously insure a sustainable energy
supply, and reduce the long-term toxicity of nuclear waste.
The first goal implies to reprocess the spent fuel and to recycle it to obtain the maximum energy
generation. This leads to recycle plutonium up to a high burn-up, many successive times. The reactor
system has to be (at least) self-sustaining in plutonium.
But the irradiation of plutonium gives rise to a build-up of americium, which in turn produces
curium, which will later partly decay to plutonium again. The long-term toxicity of nuclear waste is
dominated, for storage times ranging from 300 years to 100 000 years, by Pu, Am and Cm isotopes.
In first approximation, the sum of the quantities of Pu, Am and Cm discharged from the reactor,
related to the electrical energy produced, may be taken as an indicator of the long-term radio-toxicity.
The reactor type which allows best to extract energy from plutonium is the fast neutron reactor; it
also limits to a large extent the Pu isotopic degradation and the formation of Am and Cm. However
some minor actinides are still created and should be minimised. In particular, curium is an intense alpha
and neutron emitter, so that its presence is a heavy burden in fuel cycle operations (reprocessing,
transport, re-fabrication, handling).
In this paper, the following approach is favoured:
• Preference is given to a heterogeneous recycling of minor actinides, separated from
plutonium and embedded in special target pins, with a support material inert to neutrons; in
comparison, the homogeneous admixing of these actinides in the basic MOX fuel is more
cumbersome and penalising.
• It would be an advantage to recycle americium only, leaving curium to the waste, provided
however that a long once-through irradiation of these Am targets makes sense, with the aim
to avoid if possible any spent target reprocessing.
This paper presents in part 2 the types of fuels and reactors considered. In part 3, different
scenarios of energy generation either by PWRs or by LMFRs are considered for recycling, Pu, Am
and Cm; a “double-strata” scenario with the introduction of accelerator driven systems (ADS) is also
included for comparison.
Part 4 shows how a mixed scenario (PWRs+LMFRs) can be adapted to recycle Pu in the basic
fuel, and incinerate Am and Cm, or Am only, in targets. A discussion of all these scenarios is
conducted in part 5, with an hint at the penalties incurred in fuel cycle costs and electricity generation
costs. Conclusions are drawn in part 6.
The paper makes use of calculations run at Belgonucléaire [1,2] and at CEA [3,4]. The results
were sometimes normalised to improve their consistency.
2. Types of fuels and reactors
The paper is centred on oxide fuels, either in thermal or fast reactors.
436

The reference case for all comparisons is a 1 500-MWe PWR loaded with UO 2
, irradiated to an
average discharge burn-up of 60 GWd/t, which is the burn-up target for the years to come.
Pu recycling is being practised in PWRs nowadays. But so far, Pu was rarely recycled more than
once, because of its important isotopic degradation and production of minor actinides under
irradiation with thermal neutrons.
A way to allow Pu multi-recycling in PWR is the so-called MIX concept, in which the MOX Pu
content is stabilised by adding enriched UO 2
to the MOX. In this way, Pu multi-recycling can proceed
further in PWRs and reach an equilibrium in which Pu consumption equals Pu production. According
to calculations, about 10 successive recycles are necessary to reach the equilibrium. Nevertheless,
such a solution is costly, as nearly the same UO 2
enrichment is needed as in the reference UO 2
case.
The fast reactor concept considered is the EFR of 1 500 MWe, which was extensively studied by
the EFR Associates as a successor to Superphenix [5]. The EFR cores retained, (140 GWd/t average
burn-up with the design limit of 200 dpa damage in steel), are either self-sustaining in Pu with a thin
fertile blanket, or moderately burning Pu when all fertile blankets are removed. The average Pu
enrichment is about 20% in the first case and 23% in the second one.
CAPRA-type EFR cores with very high Pu enrichments of about 40% are not considered here:
these strong Pu burners are producing a too large amount of minor actinides.
Table 1 gives the main fuel cycle characteristics and the types of reactors retained. The
equilibrium is reached between production and consumption of Pu, or of Pu+Am+Cm.
Table 1. Fuel characteristics and reactor types (Equilibrium cores)
Reactor type
Main fuel
enrichment
%
Fuel mass
in reactor
Residence time Actinides created (+)
or destroyed (-)
(kg/TWhe)
(t/GWe) (efp days) Pu Am Cm
PWR (UO 2
), 60 GWd/t 4.9 (U5) 86 1 780 +26 +1.65 +0.25
PWR (MIX), 60 GWd/t
- Pu recycling only
- Pu, Am, Cm recycling
4.5 (U5), 2.1 (Pu)
4.5 (U5), 2.7 (Pu)
76
76
1 560
1 560
0
0
+4.5
0
+2.24
0
LMFR: EFR, 140 GWd/t
- Pu recycling only
- Pu, Am, Cm recycling
- Pu burner
20 (Pu)
20
23
28.5
28.5
28.5
1 700
1 700
1 700
0
0
-20
+3.5
0
0
+0.28
0
0
3. Scenarios of Pu, Am and Cm homogeneous recycling
Calculations were first done with Pu only being recycled in these reactors, while Am and Cm are
thrown to waste. Plutonium is supposed to be recovered at reprocessing up to 99.9%, a performance
already achieved today. The quantities of Pu, Am and Cm, related to the electricity generation, in
kg/TWhe, going to wastes, give an indicator of waste toxicity over storage times varying from 1 000
to 100 000 years.
437

When recycling Pu but not the minor actinides, PWR and LMFR reactors allow reducing the
actinide waste by a factor of 4 or 7, respectively. To reduce them further requires to also recycle
minor actinides.
Calculations were thus mainly run for the simultaneous recycling, in the form of a homogeneous
MOX fuel, of Pu, Am and Cm. One assumes that, while 99.9% of Pu is recovered at reprocessing,
99% of Am and Cm can be recovered. The respective losses of 0.1 and 1% are coherent targets; as
Am + Cm represent less than 10% of Pu in the reference case, it makes sense to assume 1% losses.
The core variants of PWR and FR have slightly larger enrichments than when recycling Pu alone,
as they need to be self-sustaining in Pu, Am and Cm.
An interesting mixed scenario is built, combining PWRs, and LMFRs; the exact share results
from mass balance calculations. The PWRs fuelled with UO 2
produce the actinides while the LMFRs
incinerate them. The LMFR reactor type is an EFR without blankets, with an enrichment of about
23% Pu.
A further scenario is added for the sake of comparison. It follows the double strata principle. The
major electricity generation is still provided by a mixed reactor system, PWR / EFR, but the latter
recycles Pu only, while special dedicated reactors are assumed to burn minor actinides (they need
some plutonium, too). Being fuelled mostly with minor actinides in a dedicated U-free fuel, they will
have deteriorated safety coefficients (lower Doppler, larger coolant void reactivity), and this is the
reason why they would preferably be accelerator driven systems (ADS), i.e. fast reactors with
Keff ~ 0.95, and a neutron spallation source.
Calculations showed that it was possible to minimise the (expensive) second stratum to some 5%
of the electricity production.
Table 2 gives the actinide throughputs, in kg/TWhe, for all these scenarios of homogeneous
recycling of Pu+Am+Cm.
The first observation is that both all-PWR (MIX) and all-LMFR (EFR) scenarii offer equivalent
actinide waste reduction factors, about 130. One noticeable difference is that the all-LMFR strategy
recycles more Pu, but 5 to 6 times less Cm.
A mixed scenario with about 34% PWRs fuelled with UO 2
and 66% LMFRs of the EFR type
without blankets can give a larger reduction factor, mainly because the recycled Pu quantities are
lower. According to the EFR studies [5], the kWhe production cost is about 10% larger for EFR than
for a PWR, so that a mixed reactor park is better from the economical point of view.
Surprisingly enough, a double strata strategy with 5% ADS is not better to reduce the actinide
wastes. This is explained by the relative accumulation of minor actinides: on multiple recycling the
quantities of minor actinides to be loaded remain relatively important. Of course the uncertainty on
such quantities at equilibrium is quite large, so that the efficiency could be quoted roughly
comparable.
438

It should be underlined that this conclusion depends on the assumptions made on the recovery
rates at reprocessing: 0.1% for Pu and 1% for Am and Cm. If they were taken to be 0.1% for Pu and
all Am and Cm isotopes (as many scientists do assume), the comparison would give different results,
like:
• A factor 180 for the all-FR scenario.
• A factor 290 for the mixed PWR/EFR scenario.
• A factor 250 for the introduction of 5% ADS in a double strata strategy.
Nevertheless the trends observed above remain the same, with a slight advantage to the mixed
scenario.
Table 2.Homogeneous recycling of Pu, Am, Cm.
Actinides throughputs (kg/TWhe).
Scenarii
Core throughputs
(kg/TWhe)
Pu Am Cm TRU
Actinide waste
(kg/TWhe)
U+
TRU Pu Am Cm Total
Waste
reduction
factors
100% PWR UO 2
0 0 0 0 0 26 1.65 0.25 27.9 Ref.
All-PWR: 56 6.5 8.9 71 2 050 0.056 0.065 0.089 0.210 130
All-EFR: 143 5.7 1.6 150 705 0.143 0.057 0.016 0.216 130
Mixed
PWR(UO 2
)-EFR:
98 5.9 2.1 106 463 0.098 0.059 0.021 0.178 160
Double strata:
Mixed
PWR(UO 2
)-EFR(Pu)
+ 5% ADS
99 10.5 3.65 113 381 0.099 0.105 0.036 0.24 120
4. Scenarios with minor actinide targets (heterogeneous)
Instead of polluting all MOX fabrication plants with Am and especially Cm, strong emitter of
neutrons and of alphas (heating), it is preferable to handle the minor actinides in a distinct fabrication
chain with reinforced shielding where minor actinides could be embedded in targets with a support
material inert to neutrons.
Further simplifications and cost savings could be obtained:
a) By incinerating the targets in a single long irradiation, up to say, 90% burn-up, and by
rejecting the spent targets to waste without reprocessing, so as to avoid the target
reprocessing costs (target reprocessing has not been demonstrated so far).
b) By incinerating americium only, and rejecting curium at all stages.
To judge the pros and cons, cost savings are to be put in regard of the actinide waste reduction
factors. It is clear that a) and b) lead to an imperfect incineration.
439

A priori, actinide targets could be inserted in thermal reactors. Calculations were done [2] with
the following assumptions about the target pins:
• Loading onto corner positions of MOX assemblies in PWR reactors with the MIX concept
(use of enriched 235 U to stabilise the Pu enrichment).
• Irradiation time 3 times longer than for the basic fuel pins.
But even with such prolonged irradiations, the results were disappointing, as 50% of the minor
actinides loaded had not yet been fissioned at discharge.
For that reason, actinide targets are better loaded in fast reactors; they would be placed in special
assemblies of the core, filled with a moderator material (B 11 C, ZrH and CaH have been considered).
4 2 2
The presence of moderator improves the efficiency of actinide transmutation. More details on these
calculations are given in [4].
Table 3 gives, for the mixed PWR/EFR scenario, the actinide mass balances with the use of
targets in EFR, for three different assumptions:
a) Multiple irradiation with intermediate reprocessing of Am+Cm targets, for which a 1% loss
will be assumed, in coherence with the homogeneous cases above.
b) One irradiation only of Am+Cm targets up to a 90% burn-up; the spent targets are thrown to
waste without reprocessing.
c) One irradiation of Am targets up to a 90% burn-up.
The second case was explicitly calculated [4]. Reaching 90% burn-up was shown to be possible
by the insertion of 42 target assemblies in the EFR core which normally contains 388 fissile positions,
and by the addition of a complete outer row of 78 target assemblies replacing the radial blanket. The
residence time of the 42 inner target assemblies would reach 10 years, to be compared with the
6 years residence time of the EFR fissile assemblies.
Thanks to the introduction of these target assemblies, some important safety parameters are
improved, like the Doppler effect (increased) and the coolant void reactivity (decreased);
simultaneously, the limit in terms of steel damage (200 dpa) can still be guaranteed.
The results of Table 3 refer to a mixed PWR(UO 2
)/EFR scenario: according to the calculations,
44% of the energy would be supplied by PWRs and 56% by FRs of the EFR type without blanket
(These are preliminary results, the respective shares of the reactors could still slightly change).
It can be observed that a mixed PWR(UO 2
)/EFR scenario with reprocessing gives about the same
actinide waste reduction factor (160) in the homogeneous or in the heterogeneous option.
An imperfect target recycling leads to smaller reduction factors of, about respectively:
• 50 if (Am,Cm) targets are incinerated to 90% and not reprocessed anymore.
• 30 if targets loaded with Am only are incinerated to 90% and not reprocessed anymore.
440

Cases
Table 3. Heterogeneous recycling of Am, Cm in targets.
Actinide streams (kg/TWhe).
In-core streams
(kg/TWhe)
Pu Am Cm TRU
Actinide wastes
(kg/TWhe)
U +
TRU Pu Am Cm 1) Total
Scenario
44% REP UO 2
and
2) 89 89 394 0.089 0.035 0.005
56% EFR
+ + +
Targets:
a) Am, Cm reprocessed
and recycled 3.5 0.8 0.014 – 0.029
b) Am + Cm up to
90% burn-up
c) Am alone to
90% burn-up
3.5 0.5 0.125 0.038 0.235
3.5 – 0.086 0.032 0.690
1) In case a, for Cm: 0.465 are rejected from the fuel and 0.230 from the targets.
2) Figures on this line correspond to the core basic fuel; figures of the following lines to the targets.
Wastes
reduction
factor
0.172 160
0.527 50
0.937 30
6. Discussion
The discussion will concern successively: the envisaged scenario, the heterogeneous option, the
reduction of waste toxicity, the flexibility of target irradiation and the way to effectively reach 90%
burn-up in targets.
6.1 The envisaged scenario
The mixed PWR/FR scenario envisaged above obviously relies on a revival of the fast reactor
option. This assumes that fast reactors will be largely deployed, at the time uranium resources will
become scarce, and thus more expensive. This was the subject of many studies in the past, and will
not be discussed here again.
It will simply be recalled that for the introduction of fast reactors of the EFR type, cost estimates
were published by the EFR associates [5]. They essentially set the EFR kWhe cost at a level about
10% higher than for the PWR, and the part of the EFR kWhe cost due to the fuel cycle was estimated
to be also about 10%. Such estimates were supposed to apply when EFR type reactors would have
largely been deployed.
6.2 Homogeneous or heterogeneous option
Inserting Am or Am + Cm into the MOX fuel itself of the FR in the homogeneous option brings
additional difficulties in the re-fabrication plant, so that the fuel cycle cost will increase.
441

The deliberate insertion of Am in the MOX is an extrapolation from the present fabrication
conditions of MOX fuel from an aged Pu. On this effect a comparison of dose rates was made in [6]. But
the presence of Cm in the fuel would bring a severe penalty, which can hardly be estimated at present.
The heterogeneous option has the clear advantage to disconnect the fabrication routes of Pu
(treated in MOX fuel as presently) and Am + Cm, which would be placed in target rods in a separate
smaller fabrication facility. Being smaller, it is easier to shield.
According to the figures of Tables 2 and 3, the streams of minor actinides (in kg/Twhe) are
reduced from:
Am = 5.9 Cm = 2.1 in the homogeneous case
to Am = 3.5 Cm = 0.8 in the heterogeneous case.
The cumbersome Cm streams have been significantly reduced.
It is clear that two sources of extra-costs can be avoided by, first, renouncing to target
reprocessing, and especially by recycling Am only and not Cm. Does it make sense?
6.3 The associated reductions of waste toxicity
Is it useful to reduce the waste toxicity due to actinides by a factor of 30 or 50?
The recent studies on partitioning and transmutation were often setting a waste toxicity reduction
factor of 100 as a good target. While the toxicity of the spent LWR fuel comes back to the “natural
level”, i.e. that of the uranium ore initially used, after about 200 000 years, a reduction by a factor 100
means that this would be after about 2 000 years. The risk associated with human intrusion is
obviously minimised.
A reduction by a factor 30 to 50 means that the doses associated with a human intrusion in the
waste storage become comparable to those of a human intrusion into a uranium ore layer, not after
2 000 years, but after some 10 000 years.
Reduction factors by 30 to 50 thus appear to make sense.
On the other hand, such reductions would also be favourable for what concerns the heat release
of the waste storage. This aspect deserves further studies.
6.4 The flexibility of target irradiation
The great flexibility of target irradiation in moderated assemblies of the fast reactor can be
underlined. Indeed this option could be deployed progressively:
• In a first step, the targets would contain Am only, irradiated in one run, and thrown to waste
after discharge: the actinide masses in wastes could be reduced by a factor 30, in a mixed
PWR(UO 2
)/EFR strategy, for a moderate increase of the kWhe cost only.
442

• A later step could be to add Cm to the Am targets, and burn them in the same way; the waste
reduction factor would somewhat increase to 50, but the cost would also increase.
• Later on, and if the need is recognised, the reduction factor might be improved progressively
up to about 160, provided that the targets of minor actinides are reprocessed and recycled.
Such versatility is attractive. The needed R&D programme in support, engaged step by step, remains
at a moderate level.
6.5 Ways to reach 90% burn-up in targets
Research has been engaged on Am targets. Many European laboratories have started fabrication
and irradiation of such targets. In particular, the EFFTRA-T4 experiment with fabrication made at
ITU and irradiation in the HFR reactor at Petten, supported by other partners, has reached an Am
burn-up of 28% [7]. A problem was however raised by the large swelling of the spinel matrix, which
would not allow much longer irradiation. The EFFTRA partners are searching for improvements.
Among the possible solutions, a promising one is offered by the technology of Pu coated particle
fuel, as successfully experienced in the DRAGON experimental reactor [8]: this particle fuel did not
swell even after a 60% burn-up corresponding to almost complete depletion of the Pu. The adaptation
of this process to Am targets deserves therefore careful feasibility verifications.
Studies have just been started at BELGONUCLEAIRE and EDF to assess the possibility of an
adaptation of this type of fuel to Am targets, to make them resistant to the important build-up of gas
pressure related to helium and fission products, the goal being to effectively burn 90% of the Am
loaded.
7. Conclusions
This paper has underlined how attractive is the concept of putting Am, or Am + Cm targets on
special moderated positions of an EFR core. Their irradiation up to 90% burn-up seems feasible in a
long but still affordable irradiation.
It was shown by calculations that, if reprocessed and recycled, this target concept in a fast reactor
could lead, in a reactor park made of a mix of PWRs (UO 2
) and LMFRs, to an actinide waste mass
reduction factor of 100 or more.
If the targets can reach a 90% burn-up and are not further reprocessed, this factor decreases to
about 50 (for the case of Am + Cm targets) or 30 (targets loaded with Am alone). Such reduction
factors, though moderate, already represent sensible improvements of the waste storage conditions.
The flexibility of the concept is an advantage. The research might be first focused on Am target
irradiation without reprocessing. It could progressively encompass the addition of Cm, and later the
target reprocessing.
A problem remains the integrity of targets with a burn-up as high as 90%. Inert support materials
irradiated so far exhibited a large swelling rate. The concept of particle coated fuel, virtually
non-swelling, deserves to be examined for application to Am targets. Studies on this concept are starting.
443

REFERENCES
[1] Th. Maldague et al., Core Physics Aspects and Possible Loading for Actinide Recycling in
Light Water Reactors, Global’95 Conf., Versailles, Sept. 1995.
[2] Th. Maldague et al., Recycling Schemes of Americium Targets in PWR/MOX Cores, 5th Information
Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, Mol, Belgium,
Nov. 1998, EUR 18898 EN, OECD/NEA (Nuclear Energy Agency), Paris, France, 1999.
[3] J.Y. Doriath et al., Scenario with Inert Matrix Fuel: a Specific Study, 6th Inert Matrix Fuel
Workshop, Strasbourg, May 2000.
[4] C. De Saint Jean et al., Optimisation of Moderated Targets for the Incineration of Minor
Actinides in a Fast Reactor in the Framework of Scenario Studies, to be submitted to Global
2001 Conf., Paris.
[5] J.C. Lefevre et al., (EFR Associates): European Fast Reactor: Outcome of Design Studies, 1998.
[6] A. Renard et al., Fuel Fabrication Constraints when Recycling Americium, Global’97 Conf.,
Yokohoma, Oct. 1997.
[7] R. Konings et al., Transmutation of Americium and Technetium: Recent Results of EFFTRA,
5th Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, Mol, Belgium, Nov. 1998, EUR 18898 EN, OECD/NEA (Nuclear Energy
Agency), Paris, France, 1999.
[8] H. Bairiot et al., Plutonium Coated Particles Development, Nuclear Technology, Vol. 23,
Sept. 1974.
444

RESEARCH ON NITRIDE FUEL AND
PYROCHEMICAL PROCESS FOR MA TRANSMUTATION
Y. Arai, T. Ogawa
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan
Abstract
Research on nitride fuel and pyrochemical process for transmutation of long-lived minor actinides
(MAs) in JAERI is summarised, focusing on the recent results following those presented at the last
conference in Mol. Fabrication of MAs nitride, irradiation tests of nitride fuel and development of
nitride/pyrochemical process have been carried out in JAERI based on the double-strata fuel cycle
concept, in which MAs are transmuted to short-lived or stable nuclides by sub-critical accelerator
driven system (ADS) with nitride fuel.
445

1. Introduction
The partitioning and transmutation (P&T) study in Japan, so-called OMEGA programme, is
entering the second phase after the C&R by Atomic Energy Committee (AEC) conducted in 1999.
Japan Atomic Energy Research Institute (JAERI) has proposed the transmutation of long-lived MAs
such as Np, Am and Cm using sub-critical ADS with nitride fuel based on the double-strata fuel cycle
concept. Besides the construction of high-energy proton accelerator and design study of
transmutation plant, the technological development of separation of MAs from high-level waste
(HLW), fabrication of MAs nitride fuel and reprocessing of the spent fuel is the important subject
investigated hereafter.
In the double-strata fuel cycle concept, commercial fuel cycle and MAs transmutation fuel cycle
are designed and operated independently. The former insists on economy and reasonable utilisation
of Pu in both LWR and FBR cycles, while the latter focuses on effective transmutation of hazardous
MAs. Nitride is suitable for the fuel material for MAs transmutation from the viewpoint of supporting
hard neutron spectrum and heat conduction ability. In addition, actinide mononitride with NaCl-type
structure will have a mutual solubility leading to the flexibility of accommodating variable
composition in the fuel. Pyrochemical process is used for the reprocessing of spent fuel, since it has
several advantages over the wet process in case of treating MAs concentrated with large decay heat
and fast neutron emission. One of the drawbacks of nitride fuel is that we must use nitride fuel with
15
N enriched nitrogen in this case. But the pyrochemical reprocessing has the practical feasibility of
recycling expensive 15 N.
In this paper the research on nitride fuel and pyrochemical process for MAs transmutation in
JAERI is summarised, focusing on the recent results following those presented at the last conference
in Mol [1]. Fabrication of MAs and Pu bearing nitride is described next. The present status of the
irradiation programme of nitride fuel in JAERI is introduced in the third part. The fourth part
concerns the subjects related to pyrochemical reprocessing of nitride fuel. Finally, summary and
future subjects are given in concluding remarks.
2. Fabrication of MA nitride
2.1 AmN and (Cm,Pu)N
Carbothermic reduction was applied to the preparation of AmN and (Cm,Pu)N for the first time
[2,3]. The experiments were carried out in JAERI’s hot cells of Waste Safety Testing Facility
(WASTEF). Starting materials were 243 AmO 2
and 244 CmO 2
powders obtained from Federal Science
Centre of Russia and Oak Ridge National Laboratory, respectively. For the latter oxide, however, a
considerable amount of 240 Pu has accumulated by the decay of 244 Cm during storage for about
30 years. The present composition of the oxide was determined to be (Cm 0.40
Pu 0.60
)O 2
by alpha
spectrometry.
The molar mixing ratios of C/Am for carbothermic reduction were chosen at 4.65 and 1.59, while
those of C/(Cm+Pu) were 3.2 and 1.6. The mixtures of AmO 2
+C and (Cm,Pu)O 2
+C were heated in a
molybdenum crucible at 1 573 and 1 773 K in N 2
gas stream, respectively. The CO gas release was
monitored continuously by an infrared spectroscope. After the release of CO gas subsided, the flowing
gas was changed to N 2
-4%H 2
mixed gas. The reason that we lowered the heating temperature for AmN
than for (Cm,Pu)N by 200 K is to avoid the loss of Am by vaporisation [4,5]. On the other hand, heating
446

temperature for (Cm,Pu)N followed the case for PuN since metallic Pu and Cm almost have the same
vapour pressures. Characteristics of the products of carbothermic reduction were examined by X-ray
diffraction analysis.
The formation of AmN and (Cm,Pu)N with NaCl-type structure was confirmed in all cases after
carbothermic reduction. However, oxide phases were also identified in case the initial mixing C/M
ratios (M = Am or Cm + Pu) were smaller than 2.0 as anticipated. The remaining oxides were
monoclinic Am 2
O 3
, and the mixtures of monoclinic Cm 2
O 3
and hyperstoichiometric bcc Pu 2
O 3
in the
respective cases. This result might be related to thermodynamic stability of sesquioxide in
transplutonium elements compared with their dioxide. Conditions of the carbothermic reduction and
results of X-ray diffraction analysis are summarised in Table 1.
The lattice parameter of AmN with initial C/Am ratio of 4.65, 499.8 pm, almost agreed with the
value of AmN prepared by metal or hydride route. On the other hand, the lattice parameter of
(Cm 0.40
Pu 0.60
)N with initial C/(Cm + Pu) ratio of 3.2 almost agreed with the value estimated from
Vegard’s law between CmN and PuN. This result confirmed the mutual solubility of CmN and PuN,
which is one of the advantages of nitride fuel for transmutation of MAs. Apparatus for chemical
analysis are under installation in the hot cell for examining chemical purity of the nitrides.
Table 1. Conditions of carbothermic reduction and
results of X-ray diffraction analysis for AmN and (Cm,Pu)N
Starting
composition
Temperature
(K)
Flowing gas
(Time)
Phases
identified
Lattice
parameter
(pm)
AmO 2
+ 4.65C
1 573
N (1.7 h) + N -4%H (1.7 h)
AmN
499.8 ± 0.1
AmO 2
+ 1.59C
(Cm,Pu)O 2
+ 3.2C
1 573
1 773
N 2
(5 h)
N 2
(4 h) +N 2
- 4%H 2
(4 h)
AmN
Am 2
O 3
(Cm,Pu)N
500.3 ± 0.1
494.8 ± 0.1
(Cm,Pu)O 2
+ 1.6C
1 743
N 2
(5 h)
(Cm,Pu)N
Cm 2 O 3 , bcc Pu 2 O
497.4 ± 0.2
2.2 Np(C,N)
Carbonitride is an intermediate product of carbothermic reduction for synthesising mononitride.
In carbothermic reduction, an excess amount of carbon is usually added to dioxide and the residual
carbon is removed by subsequent heating in N 2
-H 2
mixed gas stream as in the above case. It is known
that the residual carbon exists in carbonitride solid solution or as free carbon, depending on nitrogen
partial pressure and temperature. Here, a thermodynamic consideration is given to Np(C,N) in order
to investigate the reasonable condition for synthesising high-purity mononitrides.
Starting materials were 237 NpO 2
and reactor-grade graphite powders obtained from Harwell
laboratory of UK and Graphitwerk Kropmühl of Germany, respectively [6]. At first, two-phase
specimen of + was prepared by heating 237 NpO 2
+ 2.8C mixtures in N 2
stream at
1 773 K. Then it was heated to equilibrium in different temperatures and nitrogen partial pressures; at
447

1 723, 1 823 and 1 923 K in N 2
, N 2
/Ar = 1/1 and N 2
/Ar = 1/99 streams. X-ray diffraction analysis was
carried out in order to determine the lattice parameter of Np(C,N) and to confirm that the products
were still constituted by the two phases of + . The composition of Np(C,N) was
calculated from the lattice parameter assuming Vegard’s law between NpC and NpN. The
reasonableness of the present experimental manner was confirmed by the preceding tests using
U(C,N) and Pu(C,N) solid solutions, for which the thermodynamic properties were reported
previously [7,8].
The equilibrium compositions of Np(C,N) determined from the present experiments were plotted
in Figure 1. Since the specimen heated at 1 923 K in N 2
/Ar = 1/99 stream deviated from the twophase
region, it was excluded in the analysis. It is seen in Figure 1 that the equilibrium composition
of Np(C,N) shifts to nitride rich side as decreasing heating temperature and increasing nitrogen
partial pressure during heating. On the other hand, the equilibrium composition was evaluated from
thermodynamic calculation based on the following equilibrium:
[NpN] NpC
+ = [NpC] NpN
+ 0.5(N 2
) (1)
where [NpN] NpC
and [NpC] NpN
indicate one mole of NpN and NpC dissolved in Np(C,N), and free
carbon and nitrogen gas are indicated as and (N 2
), respectively. At first an ideal solid solution
model was applied to Np(C,N) and the solid lines in Figure 1 shows the calculation results based on
this assumption. In calculation, Gibbs energy of formation of NpN was cited from a recent
vaporisation experiment [9] and that for NpC from the table recommended by IAEA [10].
It is shown in Figure 1 that the results of experiments and thermodynamic calculation agree with
each other within an experimental error in the present case. This agreement suggests that Np(C,N)
solid solution could be treated as ideal one as is the case of Pu(C,N). Further, it was found from the
present experiments that the soluble amount of carbide in mononitride decreases in order of UN, NpN
and PuN; at 1 823 K under 1.05 × 10 5 Pa of nitrogen partial pressure, the equilibrium compositions of
carbonitride coexisting with free carbon are U(C 0.14
N 0.86
), Np(C 0.05
N 0.95
) and Pu(C 0.01
N 0.99
), for example
[11]. This tendency was caused by relative thermodynamic instability of monocarbide in higher
actinides. In the case of higher actinides, an increase in carbon to dioxide mixing ratio will not lead to
increase of carbon impurity contents in the products, since excess carbon existing as free carbon is
likely to be removed with relative ease compared with the case of carbonitride. The evolution of CO
gas during the initial stage of carbothermic reduction is also promoted by increasing carbon to
dioxide mixing ratio in general.
448

Figure 1. Temperature dependence of equilibrium composition of Np(C 1-X
N X
)
solid solution under N 2
, N 2
/Ar = 1/1 and N 2
/Ar = 1/99 streams. Experimental
results are compared with thermodynamic calculation assuming an ideal solution model.
Temperature (K)
1 973
1 873
1 773
N 2N2/Ar=1/1
í N 2/Ar=1/99
1 673
0.6 0.7 0.8 0.9 1
X inNp(C1-XNX)
2.3 Nitride with inert matrix
Nitride fuel used in ADS will contain Pu besides MAs in order to control a core efficient
multiplication coefficient nearly constant at ~0.95 during operation period. In addition, so-called inert
matrix nitrides are added as a diluting material. From a material-science viewpoint, the requirement
of the fuel for ADS is structural stability under operating temperature and high radiation fluence, high
heat conduction ability, compatibility with cladding material and reprocessing technology and so on.
But there has been little information on the behaviour of nitride fuel containing inert matrix for the
moment. The purpose of the present study, fabrication of nitride containing inert matrix, is to provide
basic information on the fabrication and feasibility of nitride fuel containing inert matrix. Here, ZrN,
TiN and YN are arbitrarily chosen among the candidates of inert matrix. PuN pellets containing ZrN
and TiN were fabricated by classical mechanical blending manner [12], while (Am,Y)N solid
solution was prepared by carbothermic reduction of AmO 2
+ Y 2
O 3
mixture [13].
ZrN and TiN powders on the market and PuN prepared by carbothermic reduction of PuO 2
were
used as starting materials. Classical mechanical blending manner was applied in this case, where Pu
content was adjusted at 40 and 60 wt% for PuN + ZrN and 50 wt% for PuN + TiN pellets. The mixed
powders were pressed into thin disk, heated in N 2
-8%H 2
stream at 1 673 K and crushed into powders
again. This procedure was repeated by three times, followed by pressing into green pellets and
sintering in Ar stream at 2 003 K. The final heat treatment was carried out in N 2
-8%H 2
stream at
1 673 K for control of stoichiometry.
X-ray diffraction pattern of PuN + ZrN pellets showed an almost single phase of NaCl-type
structure. The lattice parameters almost agreed with the value estimated from Vegard’s law between
PuN and ZrN, which suggested the formation of (Pu,Zr)N solid solutions. On the other hand, two
separate NaCl-type phases were identified for PuN + TiN pellet. The lattice parameters did not
change from those of PuN and TiN, which suggested that the amount of PuN dissolved in TiN and
that of TiN in PuN were negligibly small under the present experimental conditions. This contrast
result between PuN + ZrN and PuN + TiN pellets could be explained by the relative lattice parameter
difference (RLPD) speculated by Benedict [14].
449

A contrast result was also found for the density of the sintered pellets. PuN + ZrN pellets were
sintered to higher density than 90% T.D. but only 76% T.D. was attained for PuN + TiN pellet under
the same sintering condition. It is suggested that the formation of solid solution promotes sintering to
high density and leads to toughness of pellets in this case. Microstructure of PuN + ZrN pellet (Pu,
40 wt%) is shown in Figure 2 compared with that of pure PuN pellet. It is seen that they have a
similar grain size of 7-8 µm. The single phase is not a strict requirement for the fuel of ADS, but the
control of microstructure becomes important if the fuel is the mixture of multi phases. Further, some
consideration is needed to attain high density leading to toughness for PuN + TiN pellet.
On the other hand, (Am 0.1
Y 0.9
)N solid solution was prepared by direct carbothermic reduction of
the mixture of 0.1 243 AmO 2
+ 0.45Y 2
O 3
. The molar C/(Am+Y) ratio was chosen at 1.98, which was
higher than the stoichiometric value of 1.55. The mixed powder was compacted into a disk and
heated in N 2
stream at 1 573 K for reducing AmO 2
at first. Then temperature was raised to 1 773 K in
N 2
stream for reducing Y 2
O 3
and formation of (Am,Y)N, followed by removal of excess carbon in N 2
-
4%H 2
stream at 1 773 K. It was found from X-ray diffraction pattern that (Am 0.1
Y 0.9
)N solid solution
without any oxide phases was prepared. The lattice parameter was 490.14 pm, which was close to the
value assumed by Vegard’s law between AmN and YN. No significant loss of Am by vaporisation
was observed.
Figure 2. Microstructure of (Pu,Zr)N pellet (Pu; 40 wt%) (right)
compared with PuN pellet (left).
20 µ m 20 µm
3. Irradiation test of nitride fuel
3.1 Irradiation of (U,Pu)N fuel
The irradiation test of (U,Pu)N fuel has been carried out in JAERI since 1990. For the moment,
the post irradiation examinations (PIEs) of 4 He-bonded fuel pins irradiated at Japan Materials
Testing Reactor (JMTR) have been completed and basic information on the fuel behaviour has been
clarified [11]. On the other hand, the irradiation of two He-bonded (U,Pu)N fuel pins at fast test
reactor JOYO was finished in 1999 based on the joint research JAERI and Japan Nuclear Cycle
Development Institution (JNC). After cooling for a few months, PIEs were started in the end of last
year.
As for the fuel pins irradiated at JOYO, some preliminary results have been obtained from the
non-destructive PIEs carried out at JNC’s hot cells. The maximum linear power and burn-up were
evaluated at 78 kW/m and 39 000 MWd/t, respectively. Any failure of fuel pins was not observed.
450

The difference of two fuel pins exists in diametrical gap width between the (U,Pu)N fuel pellet and
cladding tube of austenitic stainless steel, i.e. 0.17 and 0.32 mm, namely smear density of fuel pin,
i.e. 82 and 78% T.D. According to the results of profilometry, a larger increase of diameter was
observed for the higher smear-density fuel pin. However, the maximum increase was 0.04 mm
(∆d/d = 0.5%) at most, which would not affect the fuel performance. Fission gas release was
evaluated at about 5 and 3% for the respective fuel pins, which were much smaller than those of
MOX fuel irradiated under the similar condition. After the completion of non-destructive PIEs,
destructive PIEs are carried out at both JAERI and JNC’ hot cells.
3.2 Irradiation of U-free nitride
A candidate fuel for ADS is a mononitride solid solution containing MAs and Pu besides
diluting inert matrix. However, there is little information on the irradiation behaviour of nitride fuel
containing MAs or inert matrix. So a capsule irradiation test of U-free nitride fuel is planned in
JAERI from 2002 at JMTR. According to a preliminary design, 2 He-bonded nitride fuel pins are
encapsulated and possible fuel compositions are (Pu,Zr)N, (Pu,Y)N, PuN + TiN containing Pu of
about 20 wt%. Cladding material is austenitic stainless steel and the linear power ranges from 30 to
50 kW/m according to the design study of ADS [15]. Following the detailed design of capsule and
licensing procedure, the preparation of fuel pellets and fabrication of fuel pins are carried out in 2001.
Acquiring new information on basic irradiation behaviour of U-free nitride fuel is expected. Further,
irradiation tests of MAs nitride fuel are near-future subjects.
4. Pyrochemical reprocessing of nitride fuel
4.1 Preparation pyroprocess database
In the reprocessing stage of spent fuel, pyrochemical process based on the electrorefining in
LiCl-KCl eutectic melt developed for metallic fuel [16] is applied to nitride fuel of ADS. In order to
support the development of nitride/pyrochemical process, some basic research has been carried out in
JAERI [17]. It includes high-temperature spectrophotometry, molecular dynamics and EXAFS for
chloride systems besides electrochemical measurements in the LiCl-KCl eutectic melt.
Thermodynamic calculation of LiCl-KCl-MCl x
-CdCl 2
-MN-Cd-M-N-C-O system (M, actinide
element) is also carried out using the free energy minimiser, Chemsage. The output should be
consistent with reliable experimental data from electrochemical measurements. Further, JAERI is
constructing “pyroprocess database” under the joint research with JNC. The user will be able to select
data on free energy formation of super-cooled liquid chlorides, activity coefficients in the LiCl-KCl
eutectic melt, stability diagram of M-N-Cl system and so on in the database.
4.2 Electrolysis of NpN and PuN
Anodic dissolution of NpN and PuN in W cage in the LiCl-KCl eutectic melt and subsequent
recovery of Np and Pu metals at solid cathode were investigated [18,19]. Cyclic voltammograms of
NpN and PuN were taken in the LiCl-KCl melt containing small amounts of NpCl 3
and PuCl 3
.
Typical result for NpN is shown in Figure 3. Anodic current caused by the following equation was
observed in the voltammogram:
NpN = Np 3+ + 3e - + 0.5N 2
(2)
451

Cathodic current was also observed in the reverse potential sweep of the voltammogram. Equation (2)
was a slow reaction and considered as a rate-determining step of the electrodissolution of NpN, since
the current did not increase in proportion to square root of the potential scanning rate, and the
cathodic peak seemed to shift to more negative potential as increasing scanning rate.
On the other hand, the equilibrium potential of NpN in case of using Ag/AgCl electrode as a
reference was determined as -0.779, -0.773 and -0.766 V at 723, 773 and 823 K, respectively, from
electromotive force measurements. The equilibrium potential was interpreted by comparison with the
theoretical redox potential of NpN. Since Ag/AgCl electrode was used as a reference electrode, the
overall reaction could be expressed as:
NpN + 3AgCl = NpCl 3
+ 0.5N 2
+3Ag (3)
So the theoretical redox potential of NpN, E NpN-Ag/AgCl
could be derived from Gibbs energy of formation
of NpCl 3
, NpN and AgCl, activity of NpCl 3
in the LiCl-KCl eutectic melt and that of AgCl in the LiCl-KCl
eutectic melt of the reference electrode and the partial pressure of N 2
. For example at 773 K, by use of
reported thermodynamic values and some assumptions, E NpN-Ag/AgCl
can be expressed as:
E NpN-Ag/AgCl
= -0.730 + 0.0111 ln p N2
(4)
where p N2
denotes the partial pressure of N 2
gas in the LiCl-KCl eutectic melt. The first term of the
right side of Equation (4), -0.730 V, corresponds to the standard redox potential of NpN at 773 K.
This value is comparable with the equilibrium potential obtained by electromotive force
measurement, -0.773V. Assuming that the difference between observed and theoretical potential is
only caused by the nitrogen partial pressure in the LiCl-KCl eutectic melt, p N2
is calculated at 2.1 kPa.
Although the measurement was carried out under high purity Ar gas atmosphere, N 2
gas generated by
the electrodissolution of NpN would be present in the LiCl-KCl eutectic melt. However, the exact
contribution of p N2
to the redox potential should be further investigated, since p N2
in the LiCl-KCl
eutectic melt was not determined in the present study and the equilibrium potential might be
influenced by the impurity and surface condition of NpN. The similar results were also obtained for
the electrolysis of PuN.
Figure 3. Cyclic voltammograms of NpN in the LiCl-KCl eutectic melt at 723, 773 and 823 K.
Concentration of NpCl 3
and scan rate are 0.53 wt% and 0.01 V/s, respectively.
0.15
0.10
723K
773K
823K
urrent (A)
0.05
0
-0.05
-0.10
-1.2 -1.0 -0.8 -0.6 -0.4
Potential (V) vs. Ag/AgCl
452

In addition to the voltammetric studies, the recovery of Np metal at solid cathode was carried out
by the electrolysis of NpN by both potential-controlled and current-controlled method. In the
potential-controlled method, the constant potential of -1.800 V was applied between Mo and
reference electrodes with monitoring anode potential and flowing current. On the other hand, in the
current-controlled method, the constant current of -0.020 A was applied during the electrolysis with
monitoring cathode and anode potentials. It was suggested from the variation of the potentials that Np
metal was successfully recovered at Mo cathode, while NpN being dissolved at anode.
The results of ICP-AES indicate that the concentration of NpCl 3
in the eutectic melt was kept
constant during the electrolysis of NpN. Further, the amount of Np dissolved at anode was nearly
equal to that of deposited Np at cathode estimated from the accumulated electric current. These
results suggest that neptunium nitride chloride, NpNCl, would be scarcely formed unlike the
formation of UNCl observed in the electrolysis of UN [20]. In the electrolysis of PuN, any formation
of insoluble nitride chloride was not observed either. These results were possibly caused by relative
high stability of trivalent ions of transuranium elements in the chloride eutectic melt.
Since the electrodeposites at cathode were the mixtures of Np metal and the eutectic melt, they
were heated at 1 073 K for 3 600 s. in Ar gas stream in order to separate them. A fraction of the
products was subjected to X-ray diffraction analysis. Some salt components such as LiCl, KCl and
NpCl 3
, however, were still identified in the diffraction pattern in addition to alpha-Np phase.
4.3 Electrode reaction of Np and Pu in liquid metal cathode
In the transmutation fuel cycle of double strata concept, Pu and MAs are recovered together at
liquid Cd cathode. It is known that such recovery becomes possible since the free energy change of
electrochemical reduction comes close at liquid Cd due to the decrease of activity coefficients of
transuranium elements and rare earth metals. So it is inevitable that electrodeposites at liquid Cd are
more or less contaminated by rare earth elements. On the other hand, Bi is another candidate of liquid
cathode. The separation of transuranium elements from rare earth elements is easier at liquid Bi
cathode than at liquid Cd. An advantage of Cd is that the recovered actinides would be easily
separated by distillation of Cd. Here, the electrode reaction of Np and Pu at liquid Cd and Bi cathodes
was investigated electrochemically [21].
Cyclic voltammograms of Np 3+ /Np couple in the LiCl-KCl eutectic melt at liquid Cd and Bi
electrodes are shown in Figure 4 compared with that at Mo electrode. The peaks corresponding to
electrodissolution and electrodeposition can be found in the figure. The redox potential of Np 3+ /Np
couple at Mo electrode was obtained from electromotive force measurement. There is a difference
between the redox potential of Np 3+ /Np couple at Mo electrode and those at liquid Cd and Bi
electrodes. The cathodic peaks at liquid Cd and Bi electrodes at 723 K appeared at more positive
potential by about 0.2 and 0.5 V, respectively, compared with at Mo electrode. We have speculated
that the potential shift was caused by thermodynamic stabilisation of actinides due to the formation of
intermetallic compounds as mentioned below.
Redox potential of Np 3+ /Np couple at solid electrode can be expressed as:
0
E Np3+/Np
= E Np3+/Np
+ (RT/3F) ln [Np 3+ ]/[Np] (5)
0
where E Np3+/Np
denotes standard potential vs. Ag/AgCl electrode, [Np 3+ ] and [Np] their activities and F
the Faraday constant.
453

On the other hand, the following equation for formation of intermetallic compound was assumed
at liquid Cd cathode.
Np + n Cd = NpCd n
(6)
Using Gibbs energy of formation of NpCd n
,∆G NpCdn
, redox potential of Np 3+ /Np couple at liquid
Cd electrode can be written as:
E Np3+/Np-Cd
= E Np3+/Np
+ ∆G NpCdn
/3F – (RT/3F) ln [NpCd n
] + (nRT/3F) ln [Cd] (7)
where [NpCd n
] and [Cd] denote their activities. It seems reasonable to suppose activity coefficients of
NpCd n
and Cd at the liquid cathode and concentration of NpCd n
are close to unity locally. Taking into
the above assumptions, the potential difference between solid Mo and liquid Cd electrodes are written
by:
∆E = -∆G NpCdn
/3F + (nRT/3F) ln [Cd] (8)
Experimental results agreed with the potential difference estimated from Equation (8) when the
formation of NpCd 11
was assumed at 723 K and formation of NpCd 6
was assumed at 773 and 823 K.
Using the same manner, it was suggested that the formation of NpBi 2
was most likely at the liquid Bi
electrode.
Almost the similar results were obtained in the case of Pu 3+ /Pu couple. In this case the formation
of PuCd 6
was assumed and this speculation was also confirmed by the micro-probe analysis of the
liquid Cd cathode after the recovery of Pu as mentioned below.
Figure 4. Cyclic voltammograms of Np 3+ /Np couple at Cd, Bi and Mo electrodes at 723 K.
Concentration of NpCl 3
and scan rate are 0.465 wt% and 0.01 V/s, respectively.
Cd electrode
Mo electrode
Bi electrode
Current (A)
Potential (V) vs. Ag/AgCl
4.4 Recovery of Pu into liquid Cd cathode
This study was carried out under the joint research of JAERI and Central Research Institute of
Electric Power Industry (CRIEPI) and the experimental details are described in another paper in this
conference [22]. Electrorefining was carried out in the LiCl-KCl eutectic melt containing about
454

2 wt% of Pu at 773 K. For the moment, the recovery of ten-gram scale of Pu into liquid Cd cathode
has been demonstrated with a concentration of Pu higher than 10 wt%, which is much higher than the
solubility limit of Pu in liquid Cd at 773 K.
The surface of the liquid Cd after recovery of Pu was smooth and it was easily separated from
the crucible. The crucible made of AlN seemed reusable under the present condition. After cutting
and polishing, the liquid Cd cathode was subjected to electron probe microanalysis. It was confirmed
that the recovered Pu seemed to have accumulated at the bottom of liquid Cd cathode. Further, the
formation of PuCd 6
phase was confirmed as anticipated from the thermodynamic speculation
mentioned above.
5. Concluding remarks
The recent results on nitride fuel and pyrochemical process development carried out in JAERI
were presented. In addition to UN, NpN, PuN and their solid solutions, small amounts of AmN and
(Cm,Pu)N were prepared by carbothermic reduction using 243 Am and 244 Cm nuclides. U-free nitride
fuel diluted by inert matrix such as ZrN, TiN and YN was prepared and characterised for the first
time. The physical and chemical property and irradiation behaviour shall be examined hereafter. The
irradiation of 2 He-bonded (U,Pu)N fuel pins at fast test reactor JOYO was completed and PIEs are
underway. As for pyrochemical reprocessing of nitride fuel, electrolysis of NpN and PuN, electrode
reaction of Np and Pu at liquid Cd and Bi electrodes, and recovery performance of Pu into liquid Cd
have been investigated.
The irradiation behaviour and physical and chemical properties of MA nitrides should be
investigated hereafter. The irradiation test of U-free nitride is scheduled for 2002 at JMTR. Possible
composition of the fuel is (Pu,Zr)N, (Pu,Y)N or PuN+TiN pellets. Electrorefining of burn-up
simulated nitride fuel, nitride fuel fabrication from the liquid Cd cathode and its characterisation are
the important subjects for the development of nitride/pyrochemical process.
In addition, a Module for TRU High Temperature Chemistry (TRU-HITECH) having three hot
cells and one glovebox comes into construction stage under the joint research with the Japan Atomic
Power Company (JAPC). Ten-gram order of 241 Am, hundred-milligram order of 243 Am and tenmilligram
order of 244 Cm can be handled in TRU-HITECH besides U, Np and Pu. The research on Am
and Cm in TRU-HITECH will start from 2002.
455

REFERENCES
[1] Y. Suzuki, T. Ogawa, Y. Arai, T. Mukaiyama, Recent Progress of Research on Nitride Fuel
Cycle in JAERI, in Proc. 5th Information Exchange Meeting on Actinide and Fission Product
P&T, Nov. 25-27, 1998, Mol, Belgium, EUR 18898 EN, pp. 213-221, OECD/NEA (Nuclear
Energy Agency), Paris, France, 1999.
[2] M. Takano, A. Itoh, M. Akabori, T. Ogawa, S. Kikkawa, H. Okamoto, Synthesis of Americium
Mononitride by Carbothermic Reduction Method, in Proc. Global’99: Nuclear Technology –
Bridging the Millennia (Jackson Hole, Wyoming, Aug. 29-Sept. 3, 1999) (CD-ROM).
[3] M. Takano, A. Itoh, M. Akabori, T. Ogawa, M. Numata, H. Okamoto, Carbothermic Synthesis
of (Cm,Pu)N, 10th International Symposium on Thermodynamics of Nuclear Materials
(STNM-10) (Halifax, Nova Scotia, Aug. 6-11, 2000), to be published in J. Nucl. Mater.
[4] T. Ogawa, T. Ohmichi, A. Maeda, Y. Arai, Y. Suzuki, Vaporisation Behaviour of (Pu,Am)N, J.
Alloys and Compounds, 224 (1995) 55-59.
[5] T. Ogawa, Y. Shirasu, K. Minato, H. Serizawa, Thermodynamic of Carbothermic Synthesis of
Actinide Mononitride, J. Nucl. Mater., 247 (1997) 151-157.
[6] Y. Suzuki, Y. Arai, K. Okamoto, T. Ohmichi, Preparation of Neptunium Mononitride by
Carbothermic Reduction, J. Nucl. Sci. Technol., 31 (1994) 677-680.
[7] M. Katsura, T. Sano, The Uranium-carbon-nitrogen System, J. Nucl. Sci. Technol., 3 (1996)
194-199.
[8] P.E. Potter, A Note on the Plutonium-carbon-nitrogen System, UK Report, AERE–M2156 (1969).
[9] K. Nakajima, Y. Arai, Y. Suzuki, Vaporisation Behaviour of Neptunium Mononitride, J. Nucl.
Mater., 247 (1997) 33-36.
[10] C.E. Holley, Jr., M.H. Rand, E.K. Storms, in: The Chemical Thermodynamics of Actinide
Elements and Compounds, Part 6: The Actinide Carbides, International Atomic Energy
Agency, Vienna (1984) p.49.
[11] Y. Arai, T. Iwai, K. Nakajima, O. Shirai, Y. Suzuki, Experimental Research on Nitride Fuel
Cycle in JAERI, in Proc. Global’99: Nuclear Technology – Bridging the Millennia (Jackson
Hole, Wyoming, Aug. 29-Sept. 3, 1999) (CD-ROM).
[12] Y. Arai, K. Nakajima, Preparation and Characterisation of PuN Pellets Containing ZrN and
TiN, J. Nucl. Mater., 281 (2000) 244-247.
456

[13] M. Akabori, M. Takano, A. Itoh, T. Ogawa, M. Numata, M. Shindo, Fabrication of Americium
Nitride Targets/Fuels for MA Transmutation, 30iemes Journées des Actinides (Dresden,
May 4-6, 2000).
[14] U. Benedict, The Solubility of Solid Fission Products in Carbides and Nitrides of Uranium and
Plutonium, Part II: Solubility Rules Based on Lattice Parameter Differences, EUR 5766 EN (1977).
[15] T. Takizuka, T. Sasa, H. Tsujimoto, H. Takano, H. Takano, Studies on Accelerator Driven
Transmutation Systems, in Proc. 5th Information Exchange Meeting on Actinide and Fission
Product P&T, Nov. 25-27, 1998, Mol, Belgium, EUR 18898 EN, pp.383-392, OECD/NEA
(Nuclear Energy Agency), Paris, France, 1999.
[16] Y.I. Chang, The Integral Fast Reactor, Nucl. Technol., 88 (1989) 129-138.
[17] T. Ogawa and Y. Arai, Nitride/Pyroprocess for MA Transmutation and Fundamental Database,
OECD/NEA Workshop on Pyrochemical Separation, Avignon, France, March 14-15, 2000.
[18] O. Shirai, T. Iwai, K. Shiozawa, Y. Suzuki, Y. Sakamura, T. Inoue, Electrolysis of Plutonium
Nitride in LiCl-KCl Eutectic Melts, J. Nucl. Mater., 277 (2000) 226-230.
[19] O. Shirai, M. Iizuka, T. Iwai, Y. Suzuki, Y. Arai, Recovery of Neptunium by Electrolysis of
NpN in LiCl-KCl Eutectic Melts, J. Nucl. Sci. Technol., 37 (2000) 676-681.
[20] F. Kobayashi, T. Ogawa, M. Akabori, Y. Kato, Anodic Dissolution of Uranium Mononitride in
Lithium Chloride-Potassium Chloride Eutectic Melt, J. Am. Ceram. Soc., 78 (1995) 2279-2281.
[21] O. Shirai, M. Iizuka, T. Iwai, Y. Suzuki, Y. Arai, Electrode Reaction of Plutonium at Liquid
Cadmium in LiCl-KCl Eutectic Melts, J. Electroanalytical Chem., 490 (2000) 31-36.
[22] M. Iizuka, K. Uozumi, T. Inoue, T. Iwai, O. Shirai, Y. Arai, Development of Plutonium Recovery
Process by Molten Salt Electrorefining with Liquid Cadmium Cathode, 6th Information Exchange
Meeting on Actinide and Fission Product P&T, Madrid, Spain, Dec. 11-13, 2000, EUR 19783 EN,
OECD/NEA (Nuclear Energy Agency), Paris, France, 2001.
457

TRANSMUTATION STUDIES IN FRANCE,
R&D PROGRAMME ON FUELS AND TARGETS
M. Boidron 1 , N. Chauvin 1 , J.C. Garnier 1 , S Pillon 1 , G. Vambenepe 2
1 Direction des Réacteurs Nucléaires, CEA, Cadarache 13108 Saint Paul Lez Durance Cedex, France
2
EdF, 12-14 Av. Dutrievoz, 69628 Villeurbanne Cedex, France
Abstract
For the management of high level and long-lived radioactive waste, a large and continuous research
and development effort is carried out in France, to provide a wide range of scientific and technical
alternatives along three lines, partitioning and transmutation, disposal in deep geological formations
and long term interim surface or subsurface storage.
For the line one, and in close link with the partitioning studies, research is carried out to evaluate the
transmutation potential of long-lived waste in appropriate reactors configurations (scenarios) relying
on current technologies as well as innovative reactors. Performed to evaluate the theoretical feasibility
of the Pu consumption and waste transmutation from the point of view of the reactor cores physics to
reach the equilibrium of the material fluxes (i.e. consumption = production) and of the isotopic
compositions of the fuels, these studies insure the “scientific” part of the transmutation feasibility.
For the technological part of the feasibility of waste transmutation in reactors, a large programme on
fuel development is underway. This includes solutions based on the advanced concepts for plutonium
fuels in PWR and the development of specific fuels and targets for transmutation in fast reactors in the
critical or sub-critical state.
For the waste transmutation in fast reactors, an important programme has been launched to develop specific
fuels and targets with experiments at various stages of preparation in different experimental reactors
including Phénix. Composite fuels as well as particle fuels are considered. This programme is presented
and recent results concerning the preparation of the experiments, the characterisation of the compounds
properties, the thermal and mechanical modelling and the behaviour of U free fuels are given.
459

1. Introduction
For the management of high level and long live radioactive waste and in the frame of the first line
of research identified in the French Law of December 91, the potential of a partitioning and
transmutation strategy to reduce the quantity and the toxicity of the waste is to be evaluated (along with
the other alternatives that are disposal in deep geological formations and long term interim surface or
subsurface storage [1]). The research is centred on minor actinides (americium, curium, neptunium)
which represent the majority of the long-term radiotoxic elements in the waste, once plutonium has been
extracted, and certain fission products with a very long-lived isotope, relatively abundant and potentially
mobile (technetium, iodine and caesium). The objective of the partitioning studies is to develop chemical
processes to obtain advanced partitioning of radionuclides to complete the partitioning of uranium and
plutonium. The development of extracting molecules and the validation of the basic concepts, which
corresponds to the stage of scientific feasibility of this research, is currently underway, and the process
validation (technical feasibility) is to be achieved for 2006.
Within the same time, the objectives of the transmutation programme are to evaluate the
transmutation potential of long lived waste in appropriate reactor configurations (scenarios) relying on
current technologies (Pressurised Water Reactor and Fast neutrons Reactors) as well as innovative
reactors (with dedicated systems such as the accelerator driven systems) and also to study the
materials to be used for the new type of fuels suitable for transmutation in order to define the first
elements of adequate solutions.
After a general presentation of what is considered in the R&D programmes on transmutation and
a short review of the scenarios studies, emphasis is led on the presentation of the research carried out
on the materials needed for the future fuels necessary for a transmutation strategy.
2. Strategy for a long-term work programme
Three steps can be identified before a possible industrial development of the transmutation
strategy, the first being the stage of scientific feasibility in which the possibility of transmutation is
evaluated on the basis of the reactor cores physics, the second including detailed studies to obtain
elements of technical feasibility in terms of fuel cycle impacts, safety and economic considerations
and fuel development, and the third dealing with the industrial feasibility with a stage of
demonstration in representative conditions of the chosen technologies.
Studies on transmutation deal with the two first steps with a special emphasis on the fuels studies,
the development of fuels being one of the key points to reach the objectives of waste transmutation.
Feasibility of the transmutation have been considered in different parks of reactors taking into
account in the first part of the scenarios the reactors of current technology (PWR and FR) ensuring
electricity production as well as incineration of waste, and in the second part, the reactors dedicated to
transmutation with high content of waste that can be either critical or ADS. The third part to be
considered are the other innovative technologies and cycles like molten salt reactors and
pyrochemistry that can be alternatives to the other ways. A first review of the results including
scientific and technological feasibility is planned in 2001 for the scenarios dealing with reactors of
current technology, in 2003 for the innovative reactors and in 2005 for the alternative ones.
Since plutonium is both a recyclable energy material and the main contributor to potential longterm
radio toxicity, the start point of all the scenarios is the management of plutonium in the fleet of
reactors and the research on transmutation is connected to the studies linked to the plutonium
consumption (links identified in the Capra Cadra programme [2]), the development of advanced
460

concepts for plutonium consumption [3,4], and particle fuels [5] and the research on new nuclear
technologies for the future. In addition to competitiveness, the new types of reactors will have to
present marked progresses in terms of minimisation of natural resources, safety and reduction of waste
production. These requirements induce research to develop fuels with good thermal characteristics,
able to sustain high temperatures, to reach high specific power and high burn ups. The solutions to be
found for the transmutation have to be connected to these developments
3. Transmutation scenarios
The scenario studies were performed to insure the theoretical (or “scientific”) feasibility of the Pu
consumption and waste transmutation from the point of view of the reactor core physics to reach the
equilibrium of the material fluxes (i.e. consumption = production) and of the isotopic compositions of
the fuels.
Taking the open cycle as a reference (Reference case in Table 1), five families of scenarios have
been considered in agreement with the French National Commission of Evaluation; three scenarios
rely on existing technology (PWR of the Franco-German European Pressurised Reactor–EPR type)
using plutonium and optionally ensuring the incineration of minor actinides (Case 1); fast neutron
reactors of the European Fast Reactor-EFR type ensuring the multirecycling of plutonium and minor
actinides (Case 2) (optionally the mono-recycling of minor actinides); a combination of PWR (UOX
and MOX) and fast neutron reactors (Case 3) to burn the plutonium and incinerate, according to the
variety, the minor actinides and some long-lived fission products). The other two cases use innovative
technologies (combination of PWR [UOX] reactors and dedicated systems such as the ADS); a double
component park considers PWR and dedicated fast reactors (Case 4), and in the “double strata” system
(Case 5), the first stratum contains PWR and fast neutron reactors that multirecycle the plutonium, and
in the second stratum, the hybrids transmute the minor actinides and long-lived fission products.
The characteristics of the various reactor fleets considered are summarised in Table 1.
Table 1. Description of scenarios
Scenario PWR UO 2
PWR MIX EFR EFR Dedicated FR U free
Pu+Am+Cm
(double component)
ADS (U free)
(Mainly Am-Cm
recycling)
(double strata
scenario)
Ref. 100%
1 100%
2 100%
3 44% 56% (1)
4 79% 21%
5 46% 49% (2) 5%
(1) Incineration in moderated targets.
(2) Pu recycling only.
461

The scientific feasibility of plutonium management and waste transmutation have been
established for the different cases [2,6]:
• With homogeneous recycling of Pu and MA in the EPR reactors (Case 1), in the case of the
homogeneous multirecycling with a 235 U enriched fuel (MIX) and in the EFR reactors (Case 2)
that allow additional transmutation of LLFP in moderated targets.
• In the EPR – EFR park (Case 3) with homogeneous recycling of Pu and Np, and transmutation
of Am and Cm in targets placed in a moderated neutron spectrum in the fast core.
• In the double component hypothesis (Case 4), around 20% of dedicated systems are needed to
ensure Pu and MA consumption.
• In the double strata (Case 5), around 50% of the first stratum are EFR reactors burning Pu,
5% of dedicated systems assuming MA transmutation.
In terms of reduction of the radio toxicity of the ultimate waste in the case of ingestion, the results
are roughly of the same order for the different scenarios, depending on the elements considered, with a
reduction factor of 3 to 10 for Pu consumption alone, and a reduction factor of 100 for Pu and Minor
Actinides management, by comparison to the open cycle (reduction factor of one).
Nevertheless, the scenarios are not equivalent if one considers the amount of recycled masses,
Scenario 1 leading for example to large amounts of Pu and MA and especially of curium which have
to be taken into account when considering the technical feasibility and especially the impact on the
cycle.
These results lead to consider the development of MA fuels for PWR (with multi-recycling of
Pu), for fast reactors (with either mono-recycling of MA and LLFP in moderated targets or multirecycling
in quasi standard fuels) and also for dedicated reactors. The R&D programme is presented
below according to these three items, after a review of the available results.
4. R&D programme on materials for transmutation
The aim of the R&D programme is to obtain elements of technical feasibility for the fuels to be
used in the different strategies to contribute to the evaluation of the scenarios in 2006.
Aside the homogeneous recycling (in PWR and FR) which requires standard fuels with a low
content in minor actinides, the heterogeneous recycling (in FR and ADS) leads to consider fuel with a
high content of minor actinides but without uranium to prevent the formation of “new” actinides. This
type of fuels can be either solid solutions of actinides or composite fuels with an actinide compound in
a matrix support that must be as inert as possible towards neutrons to be stable under irradiation and
able to reach very high fission rates up to 90%, far above the standard ones (see Table 2).
462

Table 2. Objectives for transmutation in fast neutron spectrum
Targets
Standard fuel in FR
Composition Inert matrix ~1-7g.cm -3 MA (U, Pu)O 2
Fast fluence (n.m -2 )
10-40 10 26 (moderated or fast 20 10 26
spectrum)
Linear power (W/cm) Min.: 10, max: 400 400
Temperature range 500-2 000 2 200
Fission rate (%) 30% ->>90% 17.5%
Prod. helium
36 (FR = 85%, AM~1g/cm 3 ) 0.15
(cm 3 .g -1 of fissile phase)
Prod. fission gases
20.6 (FR = 85%, AM~1g/cm 3 ) 3.6
(cm 3 .g -1 of fissile phase)
Dose on the cladding in dpa 200 150
For these U free fuels, that represent a technological discontinuity with regards to U and Pu
oxides, development is needed in different areas, first with the characterisation of the basic properties
of the fuels components, and for the elaboration of the fabrication process, then with the realisation of
experimental irradiations and post irradiation examinations to obtain elements of the behaviour under
irradiation and also in term of simulation to prospect the behaviour of new concepts.
As the different phases require the use of shielded nuclear installations, the time needed to define
a solution usable for the industrial level, will be around 15 years and more. For the specific case of fast
neutron reactors, a first step will be reached before 2010 with the identification of the performance
potential of the tested solutions. This will allow the definition of a second step for 5 to 10 years to
reach the ultimate objectives fixed to the selected concepts.
This leads to privileged generic and basic research and, along with the present experimental
programme, to develop the simulation of irradiated elements and fuels (including specific irradiation
tools) and to share this development in international collaborations.
The programme detailed below covers a large fields of applications and is presented according to
the technology considered with in first, the hypothesis of transmutation in PWR for which the R&D is
to be connected to the projects under consideration to burn plutonium, and in second the actions linked
to the use of fast neutrons reactors which offer determinate advantages for a transmutation strategy
(ratios of fission to capture more favourable than in thermal flux, availability of neutrons) and have
proven their capacity to use plutonium. Before the presentation of the experimental programme in FR,
a status of the knowledge of the behaviour of the composite fuels is made, to point out the fields of
research. In third the research for the fuels to be used in ADS is starting with work beginning on the
characterisation of the elements of interest.
5. Fuels studies for transmutation in pressurised reactors
The solution considered in the calculations is the homogeneous mode on the basis of the MIX
fuel [4], with a fuel composition of 2.7% plutonium, 0.3% americium, 0.4% curium and a uranium
enrichment of 4.5%, in a standard UO 2
EPR fuel rod geometry.
463

Another option using a basis of standard fuel rod geometry and UO 2
fuel is the Corail concept [3]
in which around 30% of MOX type rods are set with enriched UO 2
rods in a standard PWR assembly.
To prevent the formation of 239 Pu from 238 U captures, two other options are under investigations
using an inert matrix in which the plutonium and actinides can be dispersed. This concept is
investigated considering:
• Standard geometry for the rods that are of two kinds, one including UO 2
fuel and others
containing composite fuel (Duplex concept).
• A modified geometry with annular rods and composite fuel.
This last concept is studied to develop an Advanced Plutonium Fuel Assembly (APA) [4] and
recent developments [7] have shown the possibility to integrate actinides in the fuel. The objective of
stabilising the plutonium inventory is reached assuming 29% of APA EPR, 36% APA EPR being
needed to stabilise the plutonium inventory together with Am and Cm transmutation.
The R&D work is concentrated on the development of the appropriate concept for Pu
consumption in PWR, the fuel to be developed for this purpose will have to take into account the
possibility to burn also minor actinides. Boiling water reactors that constitute a growing part of the
reactors in the world and that have the capacity to use plutonium will also be considered in order to
assess their ability to recycle plutonium and to transmute the waste when compared to the EPR one.
The high temperature reactors characterised by high thermal efficiency and the capacity to offer
inherent safety may be used to burn plutonium in a complementary way. Their possibilities are under
investigation [5] and their contribution regarding the objectives of the fuel cycle will be assessed.
6. Fuel for transmutation in fast reactors – elements of behaviour under irradiation
In conventional reactors (PWR, FR) the addition of limited amounts of Am, Np, Cm in the
standard fuel in the whole core (homogeneous mode) is not supposed to affect deeply the fuel
behaviour. For the UPu type of fuel of fast reactors, the fuel behaviour is not too much affected by less
than 2 wt% of minor actinide (MA) addition as was confirmed by the SuperFact [8] experiment in the
fast reactor Phénix where (U, Pu, Np)O 2
and (U, Pu, Am)O 2
were successfully irradiated until a fission
rate of 7% (32% transmutation rate).
The main problems are concentrated in the “heterogeneous recycling”, in which MA targets are
loaded with a high content of actinides in some areas of the core. These so called targets are U-free
fuels in order to reduce waste production and the support of the MA compound is a matrix such as a
ceramic or a metal as inert as possible regarding neutron interaction.
6.1 Requirement and design of MA targets
The aim of such inert matrix fuels (IMF) is to be efficient for the transmutation and to allow a
good level of safety in case of incident or accident the requirement for a transmutation strategy being
fission rate up to 90% with a MA content of ~1-2 g cm -3
(low part of the range compared to the
requirement of 7-8 g cm -3 for ADS fuels with fission rate around 30%).
The fissile atoms and the support matrix can either form a solid solution or be integrated in a
composite fuel like ceramic inclusion in ceramic matrices (Cercer) or ceramic inclusions in metal
(Cermet), the respective composition and concentration of the different parts of these composites
464

eing adjusted to take into account the different effects induced by irradiation. The main requirements
are good thermal and mechanical properties for the matrix and chemical stability in the course of its
evolution for the actinide phase. Possible ceramic or metal candidate materials have been selected with
criteria concerning their basic properties (thermal and mechanical properties, activation with neutrons,
chemical compatibility with neighbouring materials,…) and their behaviour under irradiation.
The criteria prevailing for the selection have been initially considered on the basis of the available
data [9] that concern essentially out of pile behaviour, data being rather scarce in the field of the
behaviour under irradiation in the adequate neutron energies and fluxes.
Under irradiation in reactor, three main sources of damage have to be considered for the IMF: fast
neutrons interaction, effects of fission fragments and alpha decay products (alpha particle + heavy
recoil atom). These energetic particles produce damage through electronic and atomic interactions and
the consequences, that depend on the material, may affect significantly bulk properties: changes of
lattice parameter, phase changes, amorphization, swelling, evolution of thermal and mechanical
properties.
Furthermore, MA fuels for transmutation will have a specific behaviour under irradiation when
compared to standard fuels: the power evolution history will not be constant and will vary with a
factor of 10 (or more), the total quantity of gases (fission gases + helium) will be higher of factors of
some hundreds, and the maximum burn-up level to be reached is above 90% of the Fissile Initial Metal
Atoms which is far above the usual levels of standard fuels (see Table 2 for comparison).
In order to design specific IMF concepts adapted to the objectives of transmutation, the R&D
programme must cover first the basic materials qualification (properties, fabricability, behaviour under
irradiation) and also the in pile test of the concept itself. For the material qualification under
irradiation, the experiments are designed to study the different effects:
• Fast neutron fluence with experiment like Matina (see below).
• Fission products with ion irradiations in accelerators, this part must be developed after the
first experiments on the spinel [10].
• Alpha interaction with helium implantation experiments started in the frame of the Efttra
programme.
• Effect of fast neutrons and fission products in irradiations like Matina and Efttra T3, T4 ter.
Global experiments are then performed to test the in pile behaviour of the concept i.e. the
evolution of the different elements and of the composite itself.
6.2 Irradiation results and qualification of the concepts
The main results in this field come from the irradiation programme that was conducted in the
frame of the European collaboration Efttra (where three main field were identified, materials for
transmutation, either matrices and actinides compounds, and test of target concepts) [11] and also from
experiments performed in the Siloe and Phénix reactors. The main results are synthesised below.
465

6.2.1 Inert matrices
6.2.1.1 MgAl 2
O 4
If spinel behaves very well under high fast neutron fluence [12,13] (more than 22 10 26 n.m -2 ),
fission products recoil or alpha decays product severe damage in the matrix. The large programme on
this material has given results for different irradiation conditions. The Efttra T4 (effect of alpha decay
due to Am + fission) [14] and Tanox and Thermhet irradiations [15,16] (effects of fission only)
illustrate the swelling and the modification of the material (Figures 1a and 1b). This quite
unsatisfactory behaviour have not been observed in the case of fuels based on macromasses concept
(Thermhet fuel with fissile inclusions of 100 to 300 µm in diameter, Figure 1c) and operating in a
different temperature range (above 1 000°C) like Thermhet and Matina [17].
Due to the complexity of the different effects and of their respective interactions, behaviour of
spinel under irradiation is not fully understood up to now. Nevertheless, the use of the spinel matrix
for transmutation may still be a solution in the case of a concept of macro dispersed fuel tailored to
take into account the irradiation damage and fuel swelling therefore operating at a sufficiently high
temperature to favour defects recovery and gas diffusion.
Figure 1a. Clad diameter
change of Efttra T4
Figure 1b. Micro dispersed
fuel of Thermhet
Figure 1c. Macro masses of
Thermhet
6.2.1.2 MgO
The Matina experiment allowed also to test magnesia up to 1.4% FIMA and 2 10 26 n.m -2 [17] of
fast fluence (effect of neutrons without fissions) and the pellets appearance was very closed to the
fresh ones. This lead to consider MgO as a good candidate for fast reactors conditions. Its behaviour
will be tested in the future Ecrix experiment in the micro dispersed form and also in the Camix
experiment with macro dispersed americium oxide.
6.2.1.3 Al 2
O 3
The average swelling of 1.9% in volume measured in the Efttra T2bis irradiation for only 0.46
10 26 n/m² of fast neutron fluence and the one of 28% for a fast fluence of 17 10 26 n/m² in Santenay in
Phenix confirm the poor interest of alumina for transmutation since fast neutrons induce extensive
dislocation-loop formation and swelling.
466

6.2.1.4 ZrO 2
Due to its low heat conductivity, the element was not identified at the start as a reference, but its
fairly good structure stability under irradiation (in its stabilised form), makes it a possible candidate
that will be tested in the Camix experiment (see below).
6.1.2.5 Other matrices, CeO 2
, Y 3
Al 5
O 12
, Y 2
O 3
, TiN, W, Nb, V, Cr
Most of these materials have been tested under neutron irradiation [11,15,17] but the examination
being still underway, their potential as matrices remain to appreciate.
6.2.2 Actinide compounds
6.2.2.1 NpO x
For the experiments planned in the Super Phénix reactor, pellets have been successfully
fabricated under the form of standard fuel loaded with some 2 wt% of Np to test the homogeneous
way and NpO 2
inclusions were dispersed in MgAl 2
O 4
, Al 2
O 3
and MgO for the heterogeneous recycling.
The decision to stop the Super Phenix operation brought the irradiation projects to an end in 97 and the
new programme planned in Phénix have been concentrated on the main problems, the efforts being
centred on americium compounds.
6.2.2.2 AmO x
Americium oxide has a very complex phase diagram and the compound may show a high oxygen
potential or a high chemical reactivity towards its environment (like the matrix in contact or other
elements like sodium) [18], furthermore the thermal conductivity of AmO x
is very low. This simple
oxide form has been considered in the first experiments (Efttra T4, Ecrix) and an alternative is now
proposed with a solid solution of AmO 2
and ZrO 2
.
6.2.2.3 (Am, Zr)O x
Some zirconia based solid solutions present very attractive properties and analogy with UZrO 2
suggests a good behaviour towards radiation. (Am 0.5
, Zr 0.5
)O 2
and the pyrochlore form, (Am 2
, Zr 2
)O 7
,
were both considered and characterised [19] in the Oak Ridge National Laboratory in the frame of the
CEA/ORNL collaboration that may be extended to the same targets with curium in the place of
americium. This type of compound will be tested in the Camix experiment.
6.2.3 Target concepts
The first targets were based on a concept of a micro dispersion of the actinide phase in the host
matrix and a lot of experiments have been performed based on two types of matrices: MgO + AmO x
,
MgAl 2
O 4
+ AmO x
.
Magnesia based targets have been fabricated for the Ecrix experiments in Phénix, and samples
were used for properties measurements of the composite (thermal characteristics, oxygen potential,
melting point, heat capacity, thermal expansion and diffusivity) as well as specific tests like
compatibility with sodium in collaboration between ITU and CEA. The irradiation will give indication
on the behaviour of the magnesia-based concept as a candidate for americium transmutation.
The first fabrication of this type of fuel was made by the impregnation process and then irradiated
in the High Flux Reactor in Petten in two irradiations: Efttra T4 and T4bis [11,14] where fission rates
467

eached respectively 38.5% and 70% FIMA with a transmutation rate closed to 100%. This gives
indication that technical feasibility of transmutation is possible but the pellet have swollen
considerably, as a consequence of radiation damage and gases accumulation (Figure 1a).
Figure 2a. Calculated effect of
fission and alpha particles on
the damaged volume of the matrix
Figure 2b. Macro
particles of UO 2
dispersed
in a spinel matrix
Figure 2c. Macro
particles of UO 2
dispersed in MgO
vol% of damaged matrix
100
10
1
α particle
Fission Product
0 50 100 150 200 250 300
Size of fissile particles (microns)
100 µm
200 µm
These results and data of other experiments indicate that the target concept is to be improved to
reach the ambitious objectives for transmutation scenario (fission rate >90%) namely in taking into
account the radiation damage and gas production. Improvement of the dispersion is researched with
the introduction of macro masses (100-300 µm diameter) [20] to concentrate radiation effects due to
fission fragments or alpha decay in a small shell (Figure 2a). On a spinel-based matrix the product
answers the requirements (puerility and size of the particles, homogeneity of the target, absence of
cracks) as shown in Figure 2b, but the process is still under development for magnesia to obtain an
acceptable composite (see cracks in Figure 2c) with other innovative options of the fissile particles.
The experiments Thermhet and Efttra T3 were the first tests of this concept that will be
introduced in the Camix experiment.
6.3 Modelling
The irradiation behaviour of fuels for homogeneous recycling is simulated with the usual code
Germinal used for standard fuel, completed with specific models for helium production or evolution of
minor actinide isotopes. For targets, a heterogeneous modelling (see Figures 3a and 3b) is used with a
finite elements code (Castem) to calculate thermal and mechanical behaviour taking into account the
fissile-bearing phase and the matrix.
The experiments Thermhet [21,22] and Efttra T4ter [23] were calculated with a good agreement
measurement/calculation.
468

Figure 3a. 3D idealised meshing of a pellet
with a periodic distribution of inclusions
Figure 3b. 2D meshing deduced
from a metallography
6.4 Conclusion on the behaviour under irradiation
The work performed up to now, have brought numerous results in terms of fabrication of this new
types of fuels, characterisation of the basic properties, behaviour under irradiation of the different
components and of composite fuels.
With a start point based on simple chemical form of the fissile elements (AmO 2
particle) directly
mixed with the matrix to obtain a micro dispersed composite, the different results obtained have lead
to an optimisation of the concept with a greater complexity of the various parameters, the optimisation
elements being [9,20]:
• The choice of a spherical host phase with size ranging between 50 to 300 µm in diameter.
• A MA compound in a stabilised phase of the type AmZrYO 2
.
• A matrix material leading to a composite with acceptable thermal and mechanical
characteristics and that can sustain a not too complex fabrication process. In this domain the
possible choices rely on spinel, MgO and zirconia.
• The management of matrix damage and gas production, the later being still a matter of
discussion the choice going from complete retention to complete release of fission gases.
Calculations are still to be made to evaluate the possibilities of new ways of porosity
repartition (porous matrix or “jingle” concept) and of coated particles.
In addition to the experimental programme described below, an important work remains to be
done, in close relation with the reactor choice, in order to collect the necessary data. Figure 4 shows
that after the present experimental phase will be completed and new designs are defined, a second
phase will be necessary to reach the final objectives fixed to the fuels.
469

Figure 4. Experiments for MA transmutation
100
90
Objective
for MA once-through in fast reactor.
80
70
EFTTRA T4bis
fission rate (at%)
60
50
40
30
EFTTRA T4
ECRIX-H
CAMIX-COCHIX
ECRIX-B
Multi - recycling in fast reactors
and ADS
Objective for 1 cycle
20
EFTTRA T3
10
MATINA 2-3
TANOX
MATINA1A SUPERFACT
MATINA1
0
THERMHET
0 5 10 15 20 25 30 35 40
fast neutron fluence (10 26 n/m 2 )
7. Irradiation programme for fuel development in Phénix
The irradiation programme is designed to cover various objectives related to the behaviour under
irradiation of the fuel rod, and more generally of the different materials entering in the concept
(actinide or long-life fission product targets, moderators...):
• For MA transmutation in the homogeneous mode and considering the major acquired
knowledge on the fuel of fast neutron reactors, no major problem is expected and the
objective of the experiments is to evaluate the performance of the fast neutron reactor fuel in
the presence of a low percentage of minor actinides.
• For MA in heterogeneous mode, the developments of actinide target-fuels is at its beginning.
The objectives of the first experiments are to test the various possible solutions for the
elements (inert matrix, americium compound...) and for the composite (dispersion fraction
and mode of the americium composite...).
• Similarly, for the LLFP and the solid moderating materials selected in the incineration
concepts, the behaviour under irradiation of the composites with the most interesting physical
and chemical properties is to be studied.
7.1 Incineration of MA in the homogeneous mode
For the homogeneous mode, in addition to the experience already acquired with SuperFact [8]
based on a fast neutron reactor standard UPuO 2
fuel, the performances of a metallic fuel with a low
percentage of minor actinides (neptunium + americium + curium) and rare earth, will be studied in the
Metaphix experiments conducted in the scope of a contract with ITU on behalf of the Japanese
CRIEPI. Metaphix 1,2,3 are three 19-rod type rigs capsules with three experimental rods. The target
burn-up are 2, 7 and 11 at%, respectively. The rods are at Phénix dimensions, cladded in AIM1 with a
fuel constituted of a metallic alloy UPuZr (+minor actinides) and a sodium bond. Each rig will contain
a reference rod (without minor actinides), a rod with a MA content of 2% and a rod with a MA content
470

of 5% [24]. The target irradiation conditions for the metallic fuel rods at the beginning of their
irradiation are a maximum heat rating of 350 W/cm, a nominal TNG cladding temperature of
ca. 580°C. The 3 capsules will be introduced in the reactor in the year 2001.
7.2 Incineration of MA in the heterogeneous mode
For the heterogeneous mode and to optimise the irradiation possibilities of Phenix, several aspects
will be studied simultaneously:
• Damage of the matrices by irradiation separating the effects of neutron effects, fission
products and alpha particles.
• Behaviour under irradiation of the various americium composites and/or of the composite and
matrix arrangement.
7.2.1 Selection of the matrix
Matina 1 allows the irradiation up to 2 10 26 n.m -2 (fast flux) of various matrices (MgO, MgAl 2
O 4
,
Al 2
O 3
, Y 3
Al 5
O 12
, TiN, W, V, Nb, Cr), some of them (MgO, MgAl 2
O 4
, Al 2
O 3
) with UO 2
micro inclusions
to study the effects of fission products. Dismantled in 96 at the end of the 49 th
operating cycle,
Matina 1 supplied 2 rods subjected to destructive examinations to provide elements for the selection of
the matrix for the Ecrix H and B experiments.
The remaining rods were re-introduced in Phénix awaiting for the continuation of the irradiation
Matina 1A up to 6 10 26 n.m 2
In addition to Matina 1A, the experimental base on the reference matrix MgO will be broaden:
• On the one hand, by taking into account the microstructure of the composite as a parameter to
be optimised (the “macrodispersed” concept).
• On the other hand, by irradiations up to more significant fast fluences.
Furthermore, the investigation field will be enlarge to include matrices that were not (or slightly)
present in Matina 1.The matrices considered for this purpose are stabilised zirconia ZrO 2
, tungsten as a
metallic matrix, the manufacturing and characterisation work on these materials will possibly be
shared with other R&D entities.
Both objectives are taken into account in the Matina 2-3 irradiation project i.e. a 19-rod rig with
ca. a dozen experimental rods containing Cercer and Cermet composites planned to begin its
irradiation in 2003 up to a fast fluence of ca. 10 10 26 n.m -2 .
7.2.2 Selection of the americium composite – composite optimisation
The Ecrix programme includes two irradiations, identical as regards the material constituting the
americium target but different as regards the target irradiation conditions: the irradiation neutron
spectrum will be moderated by two different materials: 11 B 4
C for Ecrix B and CaH x
for Ecrix H. The
two experiments have required the development of two specific irradiation rigs planned to be available
beginning of 2001 and specific calculations (Figures 5 and 6a) to take into account the flux
modifications [25].
471

Figure 5a. Ecrix B rig
Figure 5b. Ecrix H rig
moderator 11 B 4 C
moderator CaH x
target pin
target pin
steel
steel
sodium
sodium
The fuel of the two Ecrix rods is a composite target with americium oxide “micro-dispersed” in
the MgO matrix. The objective is to reach a fission rate of 30 at%. The required duration to reach this
objective is 700 Effective Full Power Days for Ecrix B and 450 for Ecrix H with respective maximum
linear power of ca. 50 and 70 W/cm without the uncertainty. The introduction in the reactor of the two
experimental capsules is planned for the year 2001.
The manufacturing of the pellets has been completed in Atalante (Figure 6b) [26], and the two
rods will be ready by end of 2000. Compared to the other experiments, the simultaneous
implementation of a fast neutron reactor flux, a neutron spectrum converter and an americium target
gives these experiments a prototypic aspect that goes beyond the sole objective of studying the
behaviour of the target
Figure 6 a. Evolution of the volumic power
in Ecrix B and H (top)
Figure6 b. Micrograph of the Ecrix fuel
600
500
400
Puissance volumique dans l'aiguille américiée
Porosities in black
AmOx white
MgO grey
W/cm3
300
ECRIX-H
200
ECRIX-B
100
250 µm
0
0 JEPN 60 JEPN 100 JEPN 200 JEPN 300 JEPN 400 JEPN 500 JEPN
Duré e
To open the field of investigation, a further irradiation is planned to select the optimised
composite and the americium targets. A research axis will concern the stabilisation of the americium
oxide in a cubic structure thanks to the addition of a stabilising element with composites of the
(Am,Zr,Y)O 2-x
type. The Camix irradiation (Composites of AMericium in PHÉNIX) will cover the
optimisation of the actinide compound. A second axis of the investigation will concern the dispersion
mode of the americium compound in the inert matrix considering the macro dispersed concept (with
(Am,Zr,Y)O 2-x
in ZrYO 2
and MgO). Such a process applied to different matrices is to be operational
472

for the Cochix irradiation (optimised target in Phénix) that should start in end 2002 with the objective
of a fission rate equivalent to that of Ecrix i.e. ~30 at%.
7.3 Incineration of long life fission products
In this domain, and after the first results obtained in the Efttra T1 (Tc,I) and T2 (Tc) experiments
[11], studies are concentrated on iodine and technetium (transmutation of caesium requiring both
chemical and isotopic separation, the reference strategy is direct disposal).
Two experiments are planned in Phénix. The first experiment Anticorp 1 includes 3 rods
containing technetium 99 in the metallic form, to study the behaviour of the material under irradiation
for a transmutation atomic rate of 15% corresponding to an irradiation of 350 EFPD in a CaH x
moderated device planned to start its irradiation beginning of 2002.
A second LLFP transmutation experiment focused on iodine transmutation is planned in 2003
using the natural isotope (the use of 129 I is still considered) to study the behaviour under irradiation of
various possible iodine compounds (effect of the neutron damage and chemical evolution of the
compound). This programme will be conducted in close partnership with NRG.
7.4 Experiment on moderators
The need of a locally moderated spectrum in fast neutron reactors leads to the design of two
specific devices for the Phénix irradiations using the first moderators of acquired fabrication and so
easily available 11 B 4
C and CaH x
. Research on other moderators will continue in the Modix irradiation
(MODerator in PHÉNIX) planned in 2002, to test hydride-moderating materials (CaHx and possibly
YHx) up to a significant fast fluence of 10 10 26 n.m -2 .
The temperature stability of the compounds is the subject of an experimental out of pile
programme (studies of the dissociation temperature in different atmospheres).
7.5 Experiments linked to nuclear data
In order to improve the knowledge of the cross-sections of the various nuclei involved in the
corresponding transmutation chains, under neutron flux conditions representative of the considered
recycling mode (fast or locally moderated spectrum), two specific experiments will be launched each
containing rods with a large number of minor actinide and long-life fission product isotopes separated
into small quantities of several milligrams: Profil R in an otherwise standard sub assembly in the fast
flux of the internal core and Profil M using a moderated 11 B 4
C device.
8. Experiments in other reactors
CEA is involved in several R&D international collaborations aimed to develop technologies for
transmutation.
First of all, the Efttra collaboration including CEA and other European organisations (ITU, NRG,
EDF, IAM, FZK) working jointly on P&T issues. A valuable set of data has already been gained from
the experiments performed in the HFR reactor of Petten [11]. Various inert matrices have been tested
with and without inclusions of fissile material together with experiments on LLFP.
473

A CEA/Minatom work programme is underway in Russia. The Bora Bora irradiation in Bor60 is
designed to test various fuel concepts developed in the frame of Capra Pu management programme
with the test of pelleted PuO 2
-MgO that will bring data on the behaviour of this composite. The
Amboine project – in its initiating phase - deals with the feasibility of the Vipac concept for Am
transmutation based on AmO 2
MgO. The irradiation of such a target is planed in a second phase.
The determination of fundamental proprieties of zirconium pyrochlore as the host phase for Am-
Cm in the composite target is underway with the contribution of Oak-Ridge National Laboratory [19].
On the mid-term, new realisation in Japan, in the frame of bilateral collaboration agreements with
JNC and JAERI are under discussion, exchanges being considered on oxide compounds for the
homogeneous mode and also on the properties of nitrides.
A general view of the various components of the experimental programme is given in Table 3.
Table 3
Homogeneous
dispersion of MA
SuperFact; Trabant 1
pin 2 (oxide);
Metaphix 123 (metal)
Done, under preparation
Matrices
behaviour,
effect of
neutrons
Santenay,
T2,2bis,
Matina 1,
1A, T3
Effect of neutrons
and FP,
composite study
Matina 1,1A,
Tanox,T3,T4ter,
Thermhet, Bora
Bora
Effect of n + FP + alpha
particles, Am
compound and
composite study
T5, Ecrix B H, Camix
Cochix, AmO2Vipac
PFVL studies
T1,T2bis,
Anticorp 1 2
General data
Profil R M
(crosssections)
Modix
(moderators)
9. Transmutation in dedicated systems
The impact of the fuel design on the transmutation in dedicated reactors has been largely
investigated firstly in the frame of the European Capra Cadra project [2], then in the frame of the
“Fuel and Fuel Processing (FFP)” sub-group of the European Technical Working Group on ADS [27].
Taking into account the innovative specifications of such “dedicated” fuels, a systematic analysis
of the different actinide compounds have been done, in order to select the most promising candidates
with the current knowledge. Presently, an R&D programme proposal is being submitted to the
European Commission in the frame of the 5th framework programme. It aims at extending the basic
properties knowledge, assessing the synthesis and reprocessing processes and optimising the design of
such innovative fuels.
9.1 Specifications
Compared to conventional fuels (UOX or MOX), dedicated fuels are distinguished by:
• The absence of fertile uranium (U-free fuels or Pu-based fuels) to enhance the transmutation
efficiency (no new actinide formation). In that case, inert matrices may be considered as
support of the actinide phase in replacement of uranium as in the case of transmutation in fast
reactors. Considered in a strategy of multi recycling and reprocessing, these fuels may
include plutonium which can be interesting for the choice of the compounds and that lead to
more stable irradiation conditions than in FR.
474

• Their high minor actinides (Np, Am, Cm) content (3 to 4 times the one of composite fuels for
FR) with the ratio Pu/MA varying from 1:5 to 5:1. That leads to enhance the radioactivity
level of the virgin fuel compared to conventional fuels and to request for the fabrication step
remote handling and special protection to shield gamma and neutron radiation.
• High burn-up to decrease the fuel cycle cost, but less than in the case of transmutation in FR
(30% against up to 90%). As for the “once through” fuels, dedicated fuels must accommodate
a large fission gas and helium production and a high level of radiation damage. The impact on
fuel design and materials choice is thus very significant but to the difference of FR, the
reactor parameters are not fixed and that leads to some additional possibility in the choice.
9.2 Actinide compounds selection
According to the fuel specifications above, a classification of the different fuel types, from the
less promising to the most one, is proposed taking into account the current knowledge on minor
actinide compounds, which is unfortunately very sparse and poor.
• Metallic fuels, based on metal actinide alloys, are considered as the less interesting
candidates. because of the low melting point of the major constituents (Np and Pu melts at
640°C), the expected limited mutual solubility of the actinides and the risk of stainless steel
clad-fuel eutectic reaction at low temperature (410°C). Even if some improvements may be
put forward such as a large Zr addition to enhance the fuel margins to the melting,
considerable uncertainties remain on the actinides alloys metallurgy and on the fabrication
processes (because of the high volatility of Am and Pu).
• Carbide fuels are known to have good thermal and mechanical properties and have shown in
the past relatively good performance. However, the complex phase relations, especially
between the sesqui- and monocarbides and the highly pyrophoric nature of these compounds
make them less interesting than the other classes of refractory compounds (nitrides and
oxides).
• Nitride fuels are attractive because of their expected good thermal properties and their ability
to form solid solution whatever the minor actinide content. The major uncertainties concern
the risk of dissociation at high temperature, which could be a critical issue in case of severe
accident and the americium nitride vaporisation, which could complicate the fabrication step
and limit the running temperature in pile. Large swelling under irradiation is also a specific
feature of the nitride fuels, which should involve technical developments to accommodate it.
If improvements to meet all the requirements can be considered like operation at low
temperature, the development of such fuels must have to overcome the problems of
temperature stability and also the technical and economical problem of the nitrogen 15
enrichment to avoid the 14 C formation.
• Oxide fuels are probably the most promising candidates since they offer a logical extension of
the current MOX fuel technology. Although the thermal properties of actinide oxides are not
so favourable compared to nitrides, they should be improved by using support matrices (oxide
or metal) with good thermal properties. The experience developed in Europe on composite
targets could be directly applied to dedicated fuels. Other engineering solutions (e.g. annular
pellets, annular pins with internal cooling, specific coated particles) could be also developed.
Finally, in spite of thermal weakness, improved oxide fuels should be considered not only as
the “safest” solution if we take into account all the basic knowledge accumulated for 30 years
on MOX fuels and the great synergy with other programmes, but also the best compromise if
one consider the entire fuel cycle (in terms of fabrication and reprocessing command, reactor
safety approach,…).
475

9.3 Programme for dedicated fuels
From this analysis based on the current knowledge on actinide compounds, nitrides and oxides
are thought to be the most attractive support to the transuranium elements in ADS. However a large
R&D programme is needed to enlarge our knowledge on such compounds and to optimise the fuel
design to make it sure and safe.
The “Confirm” programme [28], which started in September 2000 in the frame of the 5th PCRD,
is the first one devoted to the nitride fuels. The (Pu/Am, Zr)N compounds will be synthesised and
characterised and (Pu, Zr)N will be irradiated in the Stüdvisk reactor. In parallel, modelling
development is performed to predict the performance of such fuels.
The “Future” programme, which will be proposed to the European Commission in January 2001,
will be focused on the oxide fuels. The (Pu, Am; (Zr))O 2
will be synthesised and characterised and
composite fuels based on (Pu, Am; (Zr))O 2
will be studied. Modelling codes for the homogeneous and
heterogeneous fuels will be developed.
Both programmes should allow by 2004 to collect information enough to judge the feasibility of
such fuels and to influence the next R&D programmes.
10. Conclusion
The efforts made since several years for the development of the fuels necessary to insure a
transmutation strategy lead to acquire knowledge on the basic data for the different materials needed,
the fabricability of the composites, and the first elements about their behaviour under irradiation for
MA and FP compounds.
The experimental programme planned in Phenix have been consolidated and the preparation work
have reached marked milestones like the irradiation devices fabrication, the realisation of the first
experimental pins with AmO 2
MgO composite, the various calculations and safety files necessary for
the start of the experiments planned in 2001 (Metaphix 1,2,3; Ecrix B,H; Profil R). The first
experimental results together with the knowledge acquired during the preparation of this first phase
allow the definition of an optimisation of the targets that will be introduced in a second phase in 2003.
Results from this programme together with those coming from other irradiations and of international
programmes will be use to elaborate the first elements of technical feasibility of the fuels for
transmutation in 2006.
Definition of fuels answering all the needs will require an additional research of long duration
integrating the choices made in the scenarios taking into account external parameters such as cost,
industrial feasibility, reactor park evolution and public acceptance. In this context, the necessity to
cover a large domain of parameters, will lead to favour international co-operation and to privilege
analytical experiments owing a better understanding of the phenomena together with some
technological ones. The identified concepts will need to be tested in experimental reactors before
considering an industrial phase.
If the necessary developments for an homogeneous transmutation using the present fuels as a
start, appear available in around ten years time, the development of composite fuels in technological
discontinuity when compared to the present fuels, needs continuous efforts on matrices, actinide
compound and fuel conception. The development of dedicated fuels will require specific
characterisation of the potential fuels adapted to the working parameters of ADS and taking into
account the evolution of the strategy in this domain. The use of different cycles like molten salts must
be appreciated in the course of the other actions made in these domains on a long-term schedule.
476

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Spinel MgAl 2
O 4
Radiation Effect in Insulators, Proc. Nucl. Instrum., Method in Phys. Research.
Jena, July 18-23, 1999, Paper P 03.28.
[11] D. Haas et al., The EFTTRA European Collaboration for the Development of Fuels and Targets
for Transmutation: Status of Recent Development, Conference Global’99.
[12] E.A.C. Neeft et al., Neutron Irradiation of Polycrystalline Yttrium Aluminate Garnet, Magnesium
Aluminate Spinel and A-alumina. Journal of Nuclear Materials, 1999, 274, 78-83.
[13] K. Fukumoto et al., Journal of Nuclear Materials, 32 (1995) 773-778.
477

[14] R.J.M. Konings et al., The EFTTRA T4 Experiment on Americium Transmutation, Technical
Report EURATOM 1999.
[15] A. Mocellin, Ph. Dehaudt, Composite Fuels Behaviour Under and After Irradiation, Technical
Committee Meeting AIEA Moscow, 1-4 oct.1996.
[16] V. Georgenthum et al., Influence of the Microstructure for Inert Matrix Fuel Behaviour.
Experimental Results and Calculation, Global’99.
[17] N. Chauvin, T. Albiol et al., In-pile Studies of Inert Matrices with Emphasis on Magnesia and
Magnesium Aluminate Spinel. Journal of Nuclear Materials, 1999, 274, 91-97.
[18] S. Casalta, H.J. Matzke, C. Prunier, A Thermodynamic Properties Study of the Americium-oxygen
System, Global’95.
[19] Ph. Raison, D. Haire, T. Albiol, Materials Relevant for Transmutation of Americium and
Experimental Studies on the Pseudo-ternary System, AmO 2
-ZrO 2
-Y 2
O 3
, Journal of Nuclear
Materials, to be published.
[20] N. Chauvin, R.J.M. Konings, H.J. Matzke, Optimisation of Minor Actinide Fuels for Transmutation
in Conventional Reactors (PWR, FR), Conference Global’99.
[21] V. Georgenthum et al., Experimental Study and Modelling of the Thermo-elastic Behaviour of
Composite Fuel in Reactor – Emphasis on Spinel Based Composites, Paper accepted for
Progress in Nuclear Energy.
[22] V. Georgenthum, CEA/EDF PhD thesis, Université de Poitiers 2000.
[23] R.J.M. Konings, K. Bakker et al., Journal of Nuclear Materials, 1998, 254, 84-90.
[24] Fabrication of UPuZr Metallic Fuel Containing Minor Actinide, Global’97 Conference,
October 5-10, 1997, Yokohama, Japan.
[25] J.C. Garnier, N. Schmidt et al., The ECRIX Experiments, Global’99 Conference, 29 Aug.-3 Nov. 1999,
Jakson Hole, USA.
[26] Y. Croixmarie et al., The Ecrix Experiment, Global’99.
[27] M. Salvatores, Transmutation: a Decade of Revival. Issues, Relevant Experiments and Perspectives,
6th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation,
Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN, OECD/NEA, Paris, France (2001).
[28] J. Wallenius et al., 6th Information Exchange Meeting on Actinide and Fission Product Partitioning
and Transmutation, Madrid, Spain, 11-13 Dec. 2000, EUR 19783 EN, OECD/NEA, Paris, France
(2001).
478

FISSION PRODUCT TARGET DESIGN FOR HYPER SYSTEM
Won S. Park, Yong H. Kim
Korea Atomic Energy Research Institute
P.O. Box 105, Yusong, Taejon 305-600, Republic of Korea
E-mail: wonpark@kaeri.re.kr
Jong Seong Jeong
Department of Nuclear Engineering
Seoul National University, Republic of Korea
Abstract
The design optimisation of FP Target is performed to maximise the transmutation of 99 Tc and 129 I in the
HYPER system without causing any core safety concerns. The localised thermal flux is obtained by
inserting some moderators such as CaH 2
. Many types of target configurations are investigated. The
configuration that 99 Tc is loaded as a plate type in the outer-most region and 129 I is loaded as NaI rods
mixed with CaH 2
rods in the inner region is believed to be the most optimum in terms of transmutation
rate and core power peaking. The designed FP target configuration is estimated to have the
transmutation rate of 6.41%/yr and 13.88%/yr for 99 Tc and 129 I, respectively. The maximum pin power
peaking is 1.232 that is within the acceptable range. In addition, the configuration is expected to make
the core coolant void coefficient more negative but the Doppler coefficient less negative.
479

1. Introduction
An accelerator driven sub-critical system named HYPER (HYbrid Power Extraction Reactor) is
being developed within the framework of the national long-term nuclear research plan. Many types of
transmutation systems were investigated in terms of transmutation capability, safety, and the
proliferation resistance of the related fuel cycles. A simple stratum shown in Figure 1 is supposed to
be the most reasonable for transmutation in terms of the proliferation issues. The HYPER system
utilising fast neutron spectrum is believed to have excellent compatibility with this single stratum.
The whole development schedule for the HYPER system is divided into three phases. The basic
concept of the system and the key technical issues are derived in Phase I (1997-2000). Some
experiments will be performed to confirm the key technical issues in Phase II (2001-2003). A thermal
hydraulic test for the Pb-Bi, an irradiation test for the fuel and a spallation target test are the major
experiments that KAERI is considering. In Phase III (2004-2006), a conceptual design for HYPER
system will be finished by completing the development of design tools based on the experiments
99
Tc and 129 I are being considered to be transmuted among long-lived fission products. The fission
product targets are loaded in the middle ring of the core in order to make the support ratio of fission
product similar to that of TRU. The FP target region is designed to have very localised thermal
neutrons for the efficient burning of 99 Tc and 129 I.
The preliminary results of the basic material studies have shown that a pure metallic form is the
most desirable one for the incineration of 99 Tc; a fabrication route for casting the technetium metal has
been developed and irradiation experiments did not show any evidence of the swelling or
disintegration of the metal [1]. On the other hand, an elemental form was found not to be acceptable
for iodine because of its volatility and chemical reactivity. Thus, metal iodides are being considered.
Sodium iodide (NaI) and calcium iodide(CaI 2
) are the desirable forms. Sodium iodide is expected to
have melting problems when the sodium is liberated from iodine due to the transmutation.
In this study, a design optimisation of the FP target is performed to maximise the transmutation of
99
Tc and 129 I without causing any core safety problems in the HYPER system.
Figure 1. Material flow in the HYPER system
: DVWH
+

2. General description of the core
The HYPER core adopts a hexagonal type fuel array to render the core compact and to achieve a
hard neutron energy spectrum by minimising neutron moderation. Table 1 represents the design
parameters of the HYPER core. In order to keep the radial assembly power peaking within the design
target value of 1.5, the core is divided into three zones. A low TRU fraction fuel is designed to be
loaded in the innermost zone and a high TRU fraction fuel is loaded in the outermost region. The
refuelling is to be performed based on scattered loading with 3 batches for each zone. The core
configuration is shown in Figure 2. The HYPER core has a relatively small amount of fertile nuclides.
This raises two problems in terms of core neutronic behaviours. The first is small Doppler coefficient
that contributes to making the fuel temperature coefficient negative. Preliminary calculations show a
Doppler coefficient of about -0.36 pcm/K. The coolant void and temperature coefficients were also
found to be negative though they are very small. The homogeneous void coefficient for BOC is about
~-140 pcm/%void. However, the local void coefficient in the central region of the core was evaluated
as slightly positive. The coolant temperature coefficient is about -2.1 pcm/ o C.
The second problem is the relatively large reactivity swing in the core. As the TRU burns up, the
reactivity inside the core is reduced and more accelerator power is needed to maintain constant power.
However, the reactivity runs down so quickly that the system cannot be operated effectively with a
desirable cycle length (~1 year). To avoid such a large amount of reactivity change and minimise the
fluctuation of the required accelerator beam power due to the TRU burn-up, the burnable absorber is
employed. 90% enriched B 4
C is used as a burnable absorber [2]. Two different loading types of
burnable absorber are being investigated. The first one coats the inside of the cladding with B 4
C to a
thickness of 0.002 cm. The second replaces some of the TRU fuel rods with burnable absorber rods.
The coating method is evaluated to reduce the reactivity swing by about 38% compared to the nonburnable
absorber cases for the depletion period of 180 days. In addition, the introduction of burnable
absorber is believed to make the core neutron energy spectrum much harder. The HYPER core is to
transmute about 370 kg of TRU a year. This corresponds to a support ratio of ~5.
Either TRU-Zr metal alloy or (TRU-Zr)-Zr dispersion fuel is considered as a blanket fuel for the
HYPER system. In the case of the dispersion fuel, the particles of TRU-Zr metal alloy are dispersed in
a Zr matrix. A blanket rod is made of sealed tubing containing actinide fuel slugs in columns. The
blanket-fuel cladding material is ferritic-martensitic steel. It is expected that the dispersion fuel will
generally withstand significantly higher burn-up than alloy fuel. If the fuel particles are separated
sufficiently, the areas damaged by fission fragments will not overlap and remains a continuous metal
phase which is essentially undamaged by fission fragment. This relatively undamaged metal matrix
can withstand higher burn-ups without significant swelling than is possible with alloy fuel. As a result,
the dispersion fuel does not need as much gas plenum as the alloy fuel needs. In addition, it is not easy
to control the vaporisation of americium nuclides in the fabrication process of an alloy type fuel rod.
Pb-Bi is employed as a coolant for the HYPER system. According to Russian results, the
maximum allowable temperature of Pb-Bi coolant is approximately 650 o C. The lower limit of the
coolant temperature can be started from the Pb-Bi melting point, 125 o C. For safe operation, Pb-Bi
temperature must be sufficiently above 125 o C. Therefore 125 and 650 o C can be the basic temperature
limits of Pb-Bi coolant. The core inlet and outlet temperature of Pb-Bi coolant were determined to be
340 and 510 o C, respectively. This temperature range marginally satisfies the basic temperature limits.
Resulting core flow rate in order to cool 1 000 MW thermal power is 46569 kg/s. Coolant velocity of
primary cooling system can also cause a design constraint. Coolant velocity affects the integrity of
structural materials and the pumping load. The primary cooling system of the HYPER should be
designed with low coolant velocity as long as it can satisfy another design requirements. Since Pb-Bi
does not significantly absorb or moderate neutrons, it allows the use of a loose lattice that favours
lower coolant velocity. The P/D (Pitch-to-Diameter) of the HYPER’s core is chosen to be 1.5 and the
481

corresponding Pb-Bi velocity is 1.1 m/s, which is a relatively low coolant velocity compared to that of
typical power reactors. Instead of wire spacer commonly used for tight lattice, grid spacers are suitable
to ensure proper separation of the fuel rod. A loop type configuration was selected for the preliminary
design of the HYPER system and three-loop system was chosen as the optimal one for the HYPER.
The number of loop is determined by considering the coolant velocity and pressure drop across the
loop.
Table 1. Major system design parameters
Parameter (unit) Values Parameter (unit) Values
System
- Core thermal power (MW)
- Active core height (m)
- Effective core diameter (m)
- Total fuel mass (TRU-Kg)
- System multiplication factor
- Accelerator beam power (MW)
- Ave. discharge burn-up (%at)
- Transmutation capability (Kg/yr)
- Number of fuel assembly
- Ave. linear power density (KW/m)
1 000
1.2
3.8
2 961
0.97
~6
~25
380
183
13.5
- Ave. neutron energy (keV)
- Ave. neutron flux (n’s/sec-cm 2 )
Assembly
- Ass. pitch (cm)
- Flow tube outer surface flatto-flat
distance (cm)
- Tube thickness (cm)
- Tube material
- Rods per assembly
600
6 × 10 15
19.96
19.52
0.3556
HT-9
331
Figure 2. Core configuration
The core coolant is also used as the spallation target. Pb-Bi comes from the bottom of the reactor
and encounters the beam window before going out of the top of the reactor. A single beam window is
adopted so that there is no independent window cooling system. There are some design goals for the
stable and safe operation of the target and reasonable lifetime of the beam window. We set the
maximum allowable temperature and stress of the beam window at 700 o C and 200 MPa, respectively.
The temperature of Pb-Bi is set to be less than 600 o C and the lifetime of the beam window is set to be
1 year.
482

3. Basic characteristics of FP transmutation
The transmutation rate is a function of the neutron flux level and the cross-section(Φnσ a
). The
introduction of moderator slows down the neutron energy. The moderation reduces the neutron flux
level but increase the absorption cross-section of fission product. The preliminary studies were
performed to evaluate the effectiveness of local moderator. They were based on the constant volume
model (moderator volume + FP volume = FP target volume = constant) because the total volume of a
FP target can not be larger than the volume of a TRU assembly. The evaluation showed that the
transmutation rate could be improved considerably for 129 I while there was not much difference in the
transmutation of 99 Tc (Figure 3). In addition, a graphite and calcium hydride were evaluated in terms
of their effectiveness for moderation. The calculational results showed that a calcium hydride is much
better for the production of localised thermal neutrons [3].
In order to decide the way of loading moderator into the FP target, two basic types of FP target
configuration were studied in terms of their impact on core power peaking factor. The first one
(defined as the outer moderating) is to install the moderating material at the outer region of the FP
target. This kind of configuration resulted in unacceptably high core power peaking factors at the TRU
assemblies surrounding FP target. The second one (defined as the inner moderating) is to install the
moderating material at the central region of the FP target assembly. In this case, the power peaking
could be reduced to a certain acceptable level. However, the TRU assemblies around 129 I showed
relatively high power peaking compared to those of 99 Tc (Figure 4). The loaded 129 I (NaI) target
assembly was found to be less effective to screen out thermal neutrons because of its low density and
chemical form.
Figure 3. Total reaction rate
Figure 4. Pin power peaking factor
2.0
Total Reaction ( x a
x V) x10 18 /sec
1.5
1.0
0.5
Tc
I
Relative Power
3.0 Rod surrounding Tc
Rod surrounding I
2.5
2.0
1.5
1.0
0.0
0 1 2 3 4 5
Moderator Thickness (cm)
0 1 2 3 4 5
Inner Moderator Thickness (cm)
Based on the preliminary investigations, the configuration that 99 Tc is loaded in the outer most
region and 129 I is loaded in the inner region of target is suggested to be the desirable one.
4. Effect of 99 Tc thickness
99
Tc is to be loaded in the outer most region of target to cut off the streaming of thermal neutron
into the surrounding TRU assemblies. The plate type loading was believed to be the best form to
minimise the leakage of thermal neutrons into the surrounding TRU assemblies. An investigation to
estimate the variation of pin power peaking of TRU assemblies as a function of 99 Tc plate thickness
was performed using MCNAP based on ENDF-B/VI [4]. All surrounding TRU assemblies were
described by pin-by-pin model. Figure 5 shows the cross-sectional view of the calculational model for
483

the fission product target. The rods (calcium hydride or 99 Tc) are installed using triangular array with
the P/D ratio of 1.19 in FP target.
Table 2 represents the variation of pin power peaking in the TRU assemblies surrounding FP
target. As expected, the increase of 99 Tc plate thickness reduces the pin power peaking. The pin power
for Model #1 can be used as a reference value. The thickness more than 2.4 cm is believed to be not
necessary in terms of pin power control. The transmutation rate of 99 Tc is increased from 4.86%/yr to
6.09%/yr by using the localized thermal neutrons.
Figure 5. Calculational model
Table 2. Pin power peaking variation vs 99 Tc plate thickness
Model No.
Thickness
(cm)
No. of moderator
rings
Transmutation
rate (%/yr)
Pin power
peaking
1 All Rods are Tc 0 4.86 1.158
2 0.5 13 – 3.763
3 1.5 11 – 1.685
4 2.4 10 6.09 1.193
5. Optimum configuration of FP assembly
Two types of target configurations were investigated on whole core basis as shown in Figure 6.
129
I is loaded as a plate type of NaI at the just inner side of 99 Tc plate in Type A. On the other hand, 129 I
is loaded as a rod type of NaI mixed with the moderating rods in Type B. Type A was designed to
increase the loading amount of 129 I while Type B was to reduce the self-shielding effects. Some precalculations
decided the thickness of 1.9 cm and 1.3 cm for 99 Tc and 129 I plate, respectively.
484

Figure 6. FP target configuration
a) Plate Type for I-129 b) Rod Type for I-129
Table 3 shows the results of evaluations. The higher transmutation rate for both 99 Tc and 129 I are
achieved in Type B than Type A. The loading amount of 129 I for Type B is reduced by 30% compared
to that of Type A. However, there is not much difference between two types in terms of the total
amount of the transmuted 129 I. In addition, the pin power peaking is kept within an acceptable range in
both types.
Table 3. Performance of FP Targets in the HYPER System
Target
configuration
Type A
Initial loading
(kg)
Transmuted
(kg/yr)
Transmutation
rate(%/yr)
99
Tc 901.8 54.2 6.01
129
I 129.72 13.12 10.09
Pin power
peaking
1.211
Type B
99
Tc 901.8 57.8 6.41
129
I 93.8 13 13.90
1.232
As mentioned before, the support ratio of the HYPER is expected to be about ~5. In terms of
material balance, the HYPER system is desired to have the support ratio for 99 Tc and 129 I similar to that
of TRU. The optimised configuration Type B can transmute 57.8 kg, 13 kg of 99 Tc and 129 I,
respectively. In general, LWR (1.0 GWth) produces about 8.826 kg of 99 Tc and 2.721 kg of 129 I a year.
The HYPER system itself generates about 7.9 kg and 2.3 kg of 99 Tc and 129 I due to the TRU
transmutation. As results, the net support ratios of the HYPER are estimated to be 5.7 and 4.0 for 99 Tc
and 129 I, respectively. The fission product support ratios are very close to that of TRU in the HYPER
system.
6. Safety factor evaluation
As mentioned, the localised thermal flux increases the pin power peaking of TRU assemblies.
The target configuration is optimised to minimise such an impact. Figures 7 and 8 represent the
neutron energy spectrum for the target configuration of Type A and B. The energy spectrum for the
485

fuel rods are at the TRU assembly the most far away from the core centre among the TRU assemblies
surrounding FP target. It can be seen that the neutron energy spectrum of the fuel rods near by FP
target become much softer due to the incoming thermal neutron from FP target region. The neutron
energy spectrum for 99 Tc and fuel rods are very similar in both cases. However, Type B has slightly
softer spectrum than Type a for NaI region. In addition, the resonance absorptions of 99 Tc in
epithermal region are detected in both cases.
The effects of FP target loading on the coolant void and Doppler coefficients were investigated.
The loading amount of TRU was adjusted to make the initial k eff
of the core about 0.97. Monte Carlo
Code, MCNAP with ENDF-B/VI was adopted for the evaluation. Table 4 shows the results of the
evaluation. The designed FP target makes the coolant void coefficient more negative but Doppler
coefficient less negative though the change is very small. As results, the optimised FP targets with a
localised thermal flux are expected to cause no severe core safety problems.
Table 4. Void and Doppler coefficient change due to FP target
Core type
No void
k eff
(std)
10% void
k eff
(std)
Void coeff.
(pcm/%void)
Fuel temp
300 K
k eff
(std)
Fuel temp
1 100 K
k eff
(std)
Doppler
coeff.
(pcm/K)
Reference
(No. FP)
Type A FP
loading
Type B FP
loading
0.96897
(0.00050)
0.96075
(0.00049)
0.96355
(0.00047)
0.95619
(0.00048)
0.94707
(0.00051)
0.95015
(0.00051)
-138 0.96897
(0.00050)
-150 0.96075
(0.00049)
-146 0.96355
(0.00047)
0.96628
(0.00049)
0.95929
(0.00051)
0.96157
(0.00044)
-0.36
-0.14
-0.26
7. Summary
99
Tc and 129 I are selected to be transmuted in the HYPER system. A study has been performed to
develop an optimum configuration for the fission product target. The introduction of moderator to
generate a local thermal flux is concluded to considerably increase the transmutation rate of fission
product without causing any severe core safety problems. The configuration that 99 Tc is loaded as a
plate type in the outer most region and 129 I is loaded as a NaI rod mixed with calcium hydride rod in
the inner region of fission product target is estimated to be one of the best configurations.
The support ratios of the HYPER system for 99 Tc and 129 I are estimated to be 5.7 and 4.0, respectively.
The support ratio for 99 Tc is slightly larger while that for 129 I is less compared to the support ratio for
TRU. Some minor consideration has to be given for the adjustment. In addition, the cooling problems
for the FP target will be investigated in near future.
486

Figure 7. Neutron energy spectrum
for Type A configuration
Figure 8. Neutron energy spectrum
for Type B configuration
0.1
0.1
0.01
0.01
Neutron Fraction
1E-3
1E-4
1E-5
1E-6
Moderator region
NaI region
Tc region
Fuel Rod(near)
Fuel Rod(far)
Neutron Fraction
1E-3
1E-4
1E-5
1E-6
Moderator+NaI region
Tc region
1E-7
1E-7
Fuel Rod(near)
Fuel Rod(far)
1E-8
1E-10 1E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 0.01 0.1 1 10
Energy(MeV)
1E-8
1E-101E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 0.01 0.1 1 10
Energy(MeV)
Acknowledgements
The authors greatly appreciate the financial supports of Ministry of Science & Technology
(MOST) for this study.
REFERENCES
[1] D.W. Wootan et al., Irradiation Test of Tc-99 and I-129 Transmutation in the Fast Flux Test
Facility, ANS Trans. 64, 125, 1991.
[2] Y.H. Kim et al., A Study on Burnable Absorber for a Fast Sub-critical Reactor HYPER,
6th Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, Madrid, Spain (2000), EUR 19783 EN, OECD Nuclear Energy Agency, Paris
(France), 2001.
[3] J.S. Chung, Y. H. Kim and W.S. Park, Design of the Fission Product Assembly in the Subcritical
System HYPER, Proceedings of Korean Nuclear Society Spring Meeting, 2000.
[4] H.J. Shim et al., Monte Carlo Depletion Analysis of a PWR with the MCNAP, M&C 99 Int.
Conf. on Mathematics and Computation, reactor physics and Environmental Analysis in
Nuclear Applications, Sept. 1999, Madrid, Spain, Vol. 2, 1026-1035, 1999.
487

SESSION V
TRANSMUTATION SYSTEMS AND SAFETY
Y. Arai (JAERI) – W. Gudowski (KTH)
489

SAFETY ANALYSIS OF NITRIDE FUELS IN
CORES DEDICATED TO WASTE TRANSMUTATION
Jan Wallenius, Kamil Tucek, Waclaw Gudowski
Department of Nuclear and Reactor Physics
Royal Institute of Technology
100 44 Stockholm, Sweden
Abstract
We have investigated safety characteristics of nitride fuels in critical and sub-critical cores dedicated
to waste transmutation. It is shown that decomposition of actinide nitrides followed by escape of
nitrogen gas from the core will not lead to positive reactivity feedback, provided that a) 15 N enriched
nitrogen is used, and b) pin pitches are sufficiently large. Hence the reason for nitride fuels not being
licensed for use in Phenix is not valid in the context of P&T, where use of 15 N enriched nitrogen in
nitride fuel fabrication is a prerequisite, and neutron economy is much less constrained than in FBRs.
Consequently, the main safety concerns related to use of nitride fuels can be eliminated by proper
core and fuel design.
491

1. Introduction
Since the 50s, nitride fuels have been considered as an alternative to oxides for use in fast
neutron reactors [1]. In comparison with oxides, the higher actinide density of (U,Pu)N fuels enables
shorter doubling times, which was an important objective in fast reactor development until the end of
the 80s. While nitride fuel pins have been fabricated and irradiated in both the United States, Western
Europe and Japan [2,3,4], the largest effort was undertaken in the Russian Federation, where the
BR-10 reactor for 15 years was operating on UN fuel [5]. (U,Pu)N fuel is also to be used in
BREST-OD-300, a prototype lead cooled reactor planned for construction in Beloyarsk [6].
During the last decade, the focus of fast reactor development has shifted towards utilisation for
plutonium and minor actinide burning. Safety requirements however limit the minor actinide
concentration in large cores to less than 2.5% [7]. Excess concentration of americium in TRU
inventories arise due to decay of 241 Pu if Pu is not recycled, and due to neutron capture on 242 Pu if Pu is
recycled in thermal reactors [8]. Hence sub-critical operation of dedicated minor actinide burners was
proposed by a number of authors, starting with Foster et al. in 1974 [9].
In the context of partitioning and transmutation, a renewed interest in nitride fuels has arisen,
due to additional attractive features of this fuel type, namely:
• Sufficient solubility of plutonium nitride in nitric acid for PUREX reprocessing to be
applicable. Nitrides are as metal and oxide fuels reprocessable by pyrometallurgical
methods, but the latter still have to prove ability to provide recovery fractions above 99% in
large-scale facilities.
• High thermal conductivity, enabling operation at higher linear power. Hence, the number of
minor actinide containing fuel pins to be fabricated, irradiated and reprocessed can be
significantly reduced.
For these and other reasons, a number of fast reactor designs based on the use of nitride fuels,
critical and sub-critical, were elaborated during the nineties [10-13]. As appropriate, they were not
left unchallenged. It was argued that the decomposition of plutonium and americium nitride into
metal and nitrogen gas taking place at temperatures below the melting point could cause unacceptable
safety problems. The observation of PuN decomposition in the NILOC irradiation [3] may have
triggered the cancellation of the NIMPHE program in Phenix. Subsequently, the stability of nitride
fuelled cores in beyond design basis accidents was questioned [14,15].
Obviously, if one proposes to use nitride fuels in any reactor design, one has to prove that either
fuel temperatures will never exceed the decomposition limit, or if decomposition would indeed occur,
that the consequences are of acceptable character. It is the purpose of the present paper to study the
latter case. We start by displaying nitrogen void worths in a CAPRA type of core [16], and then
proceed by analysing the behaviour of JAERI’s nitride fuelled sub-critical core design in more detail.
2. Nitrogen void worths in CAPRA cores
A fully three dimensional pin by pin model of the sodium cooled CAPRA core was set up for the
continuous energy Monte Carlo simulation code MCNP. The oxide fuel of the reference core was
substituted with (U,Pu)N fuel, having a degraded Pu vector. Following the high burn-up CAPRA core
design, the diluent spinel and steel pins were substituted with 11 B 4
C pins to obtain a softer spectrum.
Pin pitches were varied in order to cover the range from 1.2 to 1.8 times pin diameters. The fraction
of moderator pins was adjusted in order to retain a k-eigenvalue equal to unity. The nitrogen void
492

worth was calculated by removing all nitrogen from the fuel pins, corresponding to a hypothetical
scenario where loss of pin integrity at BOL leads to escape of pin bonding, followed by complete
decomposition of both UN and PuN, and escape of all nitrogen gas formed from the core region. The
resulting change in reactivity as function of pin pitch is displayed in Figure 1. As can be seen, the
voiding of natural nitrogen leads to a significant increase in reactivity, mainly due to the absence of
(n,p) reactions on nitrogen in the voided state. As expected, the void worth decreases with increasing
pin pitch, corresponding to an increase in the probability of neutron leakage in the non-voided state.
Note, however the significantly smaller void worth pertaining to the use of 15 N for fabrication of
the nitride fuel. 15 N has a full neutron shell and is therefore neutronically as transparent as 16 O. 15 N
void worths hence typically are smaller than those of 14 N by more than 1 000 pcm. Increasing pin
pitches up to 1.7 times pin diameters it even becomes negative for the sodium cooled CAPRA core
here investigated.
Figure 1. Nitrogen void worths in a (U,Pu)N fuelled CAPRA core
From Figure 1, one can infer that the use of natural nitrogen based nitride fuels in existing FBR
configurations rightfully has been questioned. However, when designing new reactor types for P&T
purposes the situation is quite different. Firstly, the use of nitrogen enriched in the 15 N isotope is
anyway foreseen in order to avoid production of 14 C. Second, the constraints on neutron economy are
much less severe for fuels containing high fractions of plutonium, allowing to increase pin pitches
without too large penalty in terms of excessive neutron leakage.
3. Void worths in sub-critical cores of JAERI design
In JAERI, sub-critical minor actinide burners have been studied since the beginning of the
OMEGA program. A nitrided fuelled core design emerged in the second half of the nineties, adopting
a Pu to TRU ratio of about 40% in order to minimise the reactivity swing over a large number of
burn-up cycles [12]. The primary coolant option considered by JAERI is sodium, even though
exploratory calculations on a Pb-Bi cooled core has been made. The fuel is diluted with zirconium
nitride and the total core power is 800 MW th
.
493

A three dimensional model of a sub-critical core similar to the JAERI design was made for
MCNP in order to evaluate the nitrogen void worth. Liquid lead-bismuth was used as spallation target
instead of solid tungsten with sodium cooling. The radius of the target was set to 20 cm. The subassembly
duct pitch was fixed to 16 cm. An average linear rating of 32 kW/m was adopted, and the
number of sub-assemblies of the core was increased with increase in pin pitches in order to maintain a
total core power of 800 MW th
. Equal molar fractions of zirconium nitride and transuranium nitride
were assumed, and the concentration of Pu in the fuel was adjusted to obtain a k-eigenvalue of the
core equal to 0.95. The Pu and MA vectors used in the simulation correspond to those of LWR
discharges after 5 years of cooling. Figure 2 displays the resulting coolant void worths for sodium and
lead-bismuth, respectively, adopting 99% 15 N enriched nitrogen for the fuel. As is well known, leadbismuth
gives a smaller void worth for P/D less than 2.0, but the strongly negative worth reported by
JAERI [12] is only present in the case of voiding upper plenum in addition to the core. Note further
that the difference in void worths between the coolants decreases for large pin pitches, going down
from 3 500 pcm at P/D ~1.5 to 1 000 pcm at P/D ~2.0.
Figure 2. Change in k-eigenvalue when voiding the core of coolant for sodium (left)
and lead-bismuth (right) cooled cores. The lower lines gives •k when voiding
upper plenum in addition to the core.
In Figure 3, the change in k-eigenvalue when voiding the core from nitrogen gas formed after
decomposition of actinide nitrides is shown. It was assumed that zirconium nitride does not undergo
decomposition. For the sodium-cooled core, the 15 N void worth becomes negative for large pin
pitches, as in the case of the CAPRA core. For the lead-bismuth cooled core, however, void worths
remain positive, with values exceeding 1 000 pcm. This is apparently due to the better reflective
properties of lead-bismuth, leading to lower leakage. Natural nitrogen void worths are about 750 pcm
larger than 15 N worths, which is a smaller difference than for the CAPRA core. It can be understood
from the fact that only 50% of the nitrogen inventory is lost in the assumption for the JAERI core.
Considering that uranium free cores in general have very little Doppler feedback, one would like
to operate minor actinide burners like the JAERI core on sub-criticality levels sufficiently deep for
super-criticality to be excluded in all cases. With sodium cooling and standard fast reactor pitches
(P/D

component increases with pin pitch, leading to a total void worth being smaller than the nitrogen
worth for P/D>1.8!
Figure 3. Change in k-eigenvalue when assuming decomposition of
transuranium nitrides and escape of nitrogen gas from
cores cooled by sodium (lower lines) and lead-bismuth (upper lines).
When comparing the safety margins of sodium and lead-bismuth cooled sub-critical core
designs, one should also take into account the reactivity losses appearing due to burn-up. The better
neutron economy of Pb-Bi leads to lower requirement of Pu concentration to obtain a certain
k-eigenvalue at a given pin pitch. For instance, a Pu fraction of 40% is required to obtain k = 0.95 at
P/D = 1.5 in the sodium cooled JAERI core. The same Pu fraction is sufficient to attain the same
eigenvalue at P/D = 1.8 in the lead-bismuth cooled version. Remembering that a 40% Pu fraction at
BOL is optimal in order to minimise reactivity losses over a series of burn-up cycles [12], a Pb-Bi
cooled core with the same reactivity loss as a Na cooled core thus features significantly lower void
worths, and can be operated at higher k-eigenvalues. For P/D = 1.8 the present study indicates that
k = 0.97 would yield sufficient margins to unprotected super-criticality using lead-bismuth as coolant.
The void worths here presented are of course upper limits to reactivity changes, at least as long
as core geometries remain intact. The high boiling temperature of lead-bismuth will lead to clad and
steel structure melting before coolant boiling, (TRU,Zr)N pellets would thus float to the surface of the
liquid metal pool where increased leakage decreases reactivity. The scenario of loss of coolant due to
tank rupture would void upper plenum before the core, again increasing neutron leakage. The
possibility of fission gas and helium leakage from fuel pins leading to gas bubbles passing through
the core should not be discarded, but the impact of such events should be of fairly local character.
495

Figure 4. Change in k-eigenvalue for simultaneous voiding
of lead-bismuth coolant and 15 N, compared to the separate void worths.
A more likely accident scenario in nitride fuelled cores is the partial loss of nitrogen due to
decomposition of americium nitride. While the dissociation temperature of AmN is not exactly
known, it is expected to be lower than that of PuN. In Figure 5, the insertion of reactivity due to
decomposition of AmN followed by escape of nitrogen gas from the core is displayed. As seen, it
remains below 1 000 pcm for all pin pitches in the lead-bismuth cooled core.
Figure 5. Change in k-eigenvalue for voiding of 15 N gas forming after decomposition of AmN,
compared to the full void worth of transuranium nitride nitrogen. PbBi cooling was assumed.
496

4. Conclusions
Having calculated the changes in reactivity resulting from voiding nitride fuelled critical and
sub-critical burner cores from nitrogen, we make the following conclusions:
• Use of 15 N in fabrication of nitride fuels diminishes nitrogen void worths by up to 2 000 pcm,
comparing with natural nitrogen void worths. Sodium cooling yields lower nitrogen void
worths than lead-bismuth (provided core geometry remains intact), due to larger neutron
leakage into upper and lower plena. For large pin pitches and sodium cooling, 15 N void
worths can become negative.
• Voiding both coolant and nitrogen, the resulting change in reactivity almost equals the sum
of void worths for the separate events for small pin pitches, but becomes smaller than the
nitrogen void for larger pitches (P/D>1.75 in the case of Pb-Bi cooling).
• With proper core and fuel design, i.e. using 15 N for fabrication of nitride fuels and large pin
pitches, nitrogen and coolant void worths can be maintained within reasonable limits (e.g.
2 000 pcm). The use of lead-bismuth coolant allows to minimise reactivity losses for large
pin pitches, and a variant of the JAERI core design based on (Zr0.5, Pu0.2, MA0.3) 15 N fuel,
P/D = 1.8 and lead-bismuth coolant appears to be possible to operate at a BOL eigenvalue in
the vicinity of 0.97.
The issue of excessive cover gas pressure resulting from nitride fuel decomposition in core
disruptive accidents has not been addressed in the present paper, but it should be noted that the
nitrogen inventory in the JAERI core design is less than half, and the actinide nitride inventory of
nitrogen is less than one quarter of that in the core investigated by Umeoka [16].
Acknowledgements
The authors would like to thank G. Ledergerber and W. Maschek for encouraging our work on
nitride fuels. The present study was performed as a part of the CONFIRM project within the
5th European Framework Programme on Partitioning and Transmutation. Financial support from the
European Commission, Svensk Kärnbränslehantering AB, and Kärntekniskt centrum is gratefully
acknowledged.
REFERENCES
[1] H. Matzke, Science of Advanced LMFBR Fuels, North-Holland, 1986.
[2] A.A. Bauer and V.W. Storhok, Irradiation Studies of (U,Pu)N, in Proc. 4th Int. Conf.
Plutonium and Other Actinides, p. 532, New Mexico, 1970.
[3] H. Blank, Fabrication of Carbide and Nitride Pellets and the Nitride Irradiations Niloc 1 and
Niloc 2, EUR-13220 EN, European Commission, 1991.
497

[4] Y. Arai et al., Experimental Research on Nitride Fuel Cycle in JAERI, In Proc. Int. Conf.
Future Nuclear Systems, Global’99 ANS 1999.
[5] B.D. Rogozkin et al., Mononitride U-Pu Mixed Fuel and its Electrochemical Reprocessing in
Molten Salts, IAEA Advisory Group Meeting, RDIPE, Moscow, October 2000.
[6] A. Filin, Current Status and Plans for Development of NPP With BREST Reactors, IAEA
Advisory Group Meeting, RDIPE, Moscow, October 2000.
[7] J. Tommassi et al., Long-lived Waste Transmutation in Reactors, Nucl. Tech. 111 (1995) 133.
[8] S.L. Beaman, Actinide Recycling in LMFBRs as Waste Management Alternative, in Proc.
1st Int. Conf. Nuclear Waste Transmutation, p. 61, University of Texas, Austin, 1980.
[9] D.G. Foster et al., Review of PNL Study on Transmutation Processing of High Level Waste,
LA-UR-74-74, Los Alamos National Laboratory, 1974.
[10] E. Adamov, The Next Generation of Fast Reactors, Nucl. Eng. Des. 173 (1997) 143.
[11] K. Ikeda et al., Safety Analysis of Nitride Fuel Core Toward Self-consistent Nuclear Energy
System, in Proc. Int. Conf. Future Nuclear Systems, Global’99, ANS 1999.
[12] T. Takizuka et al., Studies on Accelerator Driven Transmutation Systems, in Proc. 5th Int.
Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, Mol, Belgium, 1998, EUR 18898 EN, OECD/NEA (Nuclear Energy Agency),
Paris, France, 1999.
[13] J. Wallenius et al., Application of Burnable Absorbers in an Accelerator Driven System, Nuc.
Sci. Eng. 137 (2001)1.
[14] W. Maschek, D. Thiem and P. Lo Pinto, Core Disruptive Accident Analysis for Advanced
CAPRA Cores, in Proc. 4th Int. Conf. Nuclear Engineering, ICONE-4, p. 237, ASME 1996.
[15] T. Umeoka et al., Study of CDA Driven by ULOF for the Nitride Fuel Core, in Proc. Int. Conf.
Future Nuclear Systems, Global’99, ANS 1999.
[16] A. Languille et al., CAPRA Core Studies, the Oxide Reference Option, in Proc. Int. Conf.
Future Nuclear Systems, Global’95, p. 874, ANS 1995.
498

ASPECTS OF SEVERE ACCIDENTS IN TRANSMUTATION SYSTEMS
Hartmut U. Wider, Johan Karlsson, Alan V. Jones
Joint Research Centre of the European Commission
21020 Ispra (VA) Italy
E-mail: hartmut.wider@jrc.it,
johan.karlsson@jrc.it,
alan.jones@jrc.it
Abstract
The different types of transmutation systems under investigation include accelerator driven (ADS) and
critical systems. To switch off an accelerator in case of an accident initiation is quite important for all
accidents. For a fast ADS the grace times available for doing so depend strongly on the total heat
capacity and the natural circulation capability of the primary coolant. Cooling with heavy metal Pb-Bi
has considerable advantages in this regard compared to gas cooling. Moreover it allows passive exvessel
cooling with natural air or water circulation. In the remote likelihood of fuel melting, oxide fuel
appears to mix with the Pb-Bi coolant. Fast critical systems that are cooled by Pb-Bi will
automatically shut off if the flow or heat sink is lost. Reactivity accidents can be limited by a low total
control rod worth. High temperature reactors can achieve only incomplete burning of actinides. If an
accelerator is added to increase burn-up, a fast spectrum region is needed, which has a low heat
capacity.
499

Nomenclature
ADS:
ATW:
GT-MHR:
LOF:
LOHS:
LBE:
Pb:
RVACS:
TRISO:
TRUs:
Accelerator driven system.
Accelerator driven transmutation of waste.
Gas turbine modular high temperature reactor.
Loss-of-flow accident.
Loss-of-heat-sink accident.
Lead bismuth eutectic (MP 123ºC).
Lead (MP 327ºC).
Reactor vessel auxiliary cooling system.
Coated particle with three layers pyrocarbon, siliconcarbide and again pyrocarbon.
Transuranium elements: neptunium, plutonium, americium, curium.
1. Introduction – Some ADS designs
Important for the acceptability of nuclear power is a strong reduction in the long-lived higher
actinides and soluble fission products in the nuclear waste is. And thus, it is imperative for keeping the
nuclear option open for the future. Of course, the transmutation reactors should also meet modern
safety criteria such as the one of Generation IV [1] which require future reactors to be demonstrably
safe and deterministically free of catastrophic behaviour. Furthermore, transmuters should not only
lead to the reduction of the waste but also to a new generation of reactors for an economical and clean
energy generation.
The first proposal to use an accelerator driven sub-critical system for burning nuclear waste was
made in 1986 by Bonnaure, Mandrillon, Rief and Takahashi [2]. The first realisation of this concept
and the first preliminary design for a sub-critical waste burner was presented by Prof. Rubbia et al.
[3,4]. It is a fast sub-critical system (k = 0.97) with a thermal power of 1 500 MW. The proposed
accelerator is a cyclotron with proton current of 15 mA. This pool-type ADS features natural
circulation Pb cooling in a 30 m tall vessel. It has an overflow device for passively blocking the beam
and emergency decay heat removal by passive ex-vessel air cooling (RVACS).
The next ADS design, which is already quite advanced, is the Ansaldo demonstration facility of
80 MWt [5]. The sub-criticality is also about 0.97 and it has a cyclotron that delivers a 3 mA proton
current. It features LBE cooling using an “enhanced natural circulation” by the addition of argon
bubbles above the core and gas removal from the upper plenum. This allows the reduction of the
primary pool height to 8 m and provides good control of the primary flow. Moreover, the coolant flow
path is rather simple, passing from the core up through the riser section, then through the heat
exchangers that are in the downcomer and further down to the core inlet. This allows arrangement
such a good natural circulation that the full power can be removed even if the injection of gas bubbles
fails. The secondary coolant is a diathermic fluid with low vapour pressure. For emergency decay heat
removal a new type of RVACS is proposed, schematically shown in Figure 5.
Another LBE-cooled ADS with three proton beams has been presented by FZK [6]. This relies on
mechanical pumps. This means that the coolant, after having passed through the heat exchangers, has
to get back up to the inlet of the pumps, a fact which degrades the natural circulation capability. The
full power can rather certainly not be removed without the pumps running, but the decay heat and the
emergency decay heat can be easily removed due to the good natural circulation capability of the LBE
coolant.
Framatome has presented a gas-cooled fast ADS demonstrator [7]. It has a thermal power around
100 MWt. The sub-criticality is 0.95 and the proton beam has a current of 10 mA. The design is a
500

direct cycle gas-cooled fast spectrum reactor with the vessel and the gas turbine as in the GT-MHR
reactor. However, this ADS uses fast reactor fuel pins and the helium flow in the core is upwards. In
case of loss of the forced circulation of helium, the decay heat can be removed by helium natural
circulation using the heat exchangers of the shutdown cooling system. In the case of loss-of-pressure,
blowers (an active system) and the intermediate helium/water heat exchangers of the shutdown
cooling system remove the decay heat. The same approach is used during handling operations of the
shutdown ADS.
Another ADS that is proposed by General Atomics [8,9] is the 600 MWt Integrated Thermal –
Fast Transmuter. It works as a gas cooled HTR with TRISO fuel that contains LWR waste and erbium
poison in order to get an extended burn-up. After three years of critical operation, a horizontal proton
beam is used to drive the inner transmutation region that occupies about 15% of the active region and
is surrounded by a graphite reflector. It operates in the fast energy neutron spectrum and contains
tungsten rods that house TRISO particles already transmuted before in the thermal region. Since the
TRUs will only be burnt by a little more than 80%, a three-year further burn-up follows in fast gascooled
ADS (as in the Framatome approach above) in which the 50-mm tall compacts containing the
already burnt up TRISO particles will be inserted. This approach is somewhat complex. But it requires
only the reprocessing of the LWR waste. However, TRISO particles of 500-µm diameter for HTR
have only been licensed for 80 000 MWd/t burn-up. The validation of the accident behaviour for
longer burn-up and plutonium/neutron absorber containing 200 µm TRISO particles is still necessary
[10].
Several critical LBE or Pb cooled critical designs have recently been proposed that can at least
burn a considerable amount of the plutonium isotopes. The amount of minor actinides in such a core
should not be too high because this would reduce the delayed neutron fraction, which has safety
implications. If larger amounts of minor actinides should be burned, an ADS is more appropriate. In
this conference a paper describes the burn-up of nuclear waste by fast critical systems. The most
prominent recent announcement was by Minatom, Russian Federation, to build the 300 MWe Brest-
300 reactor [11] within 10 years. It is claimed that it will burn waste, be “naturally safe” and
proliferation proof. There are also LBE- cooled designs proposed by IPPE Obninsk, Russian
Federation – the SVBR-75 reactor [12] and the Tokyo Institute of Technology proposes a compact Pb-
Bi cooled reactors with long-lived (12 years) fuel. A very economical 300 MWt LBE-cooled design is
proposed by ANL, US that features natural circulation cooling [13,14]. The University of Berkeley
proposes a small proliferation-proof reactor for which the entire core can be removed [15]. The South
Koreans are proposing the PEACER reactor that can burn considerable amounts of waste and is
claimed to be very safe [16].
A potentially important way of reducing the excess plutonium with existing water-cooled reactors
is the use of plutonium fuel with a thorium matrix. Galperin and Raizes [17] have shown that a large
PWR can burn more than 1 000 kg of plutonium per year. The 233 U that is bred in the process can be
separated from the thorium and could replace some of the 235 U enrichment necessary for LWR fuel.
Proliferation problems can be avoided by denaturing the 233 U with 238 U.
2. Problems if the accelerator is not switched off following an accident initiation in an ADS
It can rather certainly be assumed that regulators will want to know what happens if the
accelerator is not switched off when one of the generic accident initiators occurs. These include the
loss-of-flow (LOF) accident (also called loss-of-forced circulation accident for gas-cooled systems);
loss-of-heat sink (LOHS) accident – e.g. due to loss of feedwater; a depressurisation in a gas-cooled
system; reactivity insertion accidents; a new type of accident for the ADS is the beam power increase
accident and for LBE-cooled systems inlet blockages due to crud formation have to be considered
501

Regarding switching off the beam, one accident type does not have to be considered – the station
blackout accident. This is because the accelerator will be automatically switched off when the
electricity supply is interrupted.
Quite a few scoping analyses on the behaviour of LBE-cooled ADS in accidents without beam
shut off have been performed earlier [18,19,20]. They generally show that negative reactivity effects
such as the Doppler effect, axial fuel expansion or even molten fuel sweepout cannot bring the power
much below nominal as long as the beam is on. On the other hand, positive feedbacks due to the
introduction of reactivity, even at a fast rate do not lead to a power burst but to an overpower condition
of a few tenths of percent above nominal. The latter will lead to some fuel pin failures after several
tens of seconds. The resultant sweepout will bring the power back to near nominal. The same type of
behaviour occurs when the beam power is increased (a 50% increase and a doubling of the beam
strength was investigated). At any rate, the reactor coolant will be contaminated (and could possibly
be cleaned afterwards) but there will be no major problem.
Analyses with the STAR-CD code [21] of an LBE-cooled ADS undergoing a major coolant
disturbance such as a LOF or LOHS gave the following results: In the Ansaldo design with its
excellent natural circulation capability, the gas injection can be shut off (LOF) and the heat generated
by the full power can still be removed. However, the outlet temperature is about 80 K hotter and this
should not be maintained for long periods. ADS with mechanical pumps and a worse natural
circulation capability may have considerably greater outlet temperature increases. But this will depend
on the specific design and on the thermal power of the ADS.
For the LOHS accident with the beam on and the argon injection working, there is a grace time of
about 40 min in the 80 MWt Ansaldo ADS before the 900ºC limit for vessel creep is reached. This is
due to the large heat capacity of the heavy metal coolant. The long decrease of the temperature is due
to the ex-vessel cooling with a PRISM type RVACS. For a combined LOHS and LOF (but not a
station blackout, which would shut off the accelerator), there is only a 30 min grace time. This is
because a map of hot LBE collects near the top of the vessel and is only intermittently removed by the
natural circulation. If the beam is still not switched off after this grace time it can be assumed that the
beam pipe will rupture before the vessel fails. This would flood the beam pipe with LBE so that the
spallation source could be removed from the core. Later in the section on beam shut off, it will be
shown that a deliberate weak spot in the beam pipe (a so-called melt-rupture disk) increases the grace
time significantly.
Figure 1. LOHS in 80 MWt ADS –
beam on for 40 min
Figure 2. LOHS + LOF with
beam on for 30 min
1300
0.32
1300
0.30
Temperature, K
1200
1100
1000
900
800
700
600
Wall temperature
Core outlet temperature
Core outlet velocity
0.30
0.28
0.26
0.24
Velocity, m/s
Temperature, K
1200
1100
1000
900
800
700
600
Wall temperature
Core outlet temperature
Core outlet velocity
0.25
0.20
0.15
0.10
0.05
0.00
Velocity. m/s
500
0.22
500
0 10 20 30 40
-0.05
0 10 20 30 40
Time, hours
Time, hours
502

If an inlet blockage occurred in an LDE-cooled ADS it could lead to some fuel melting. However,
an accident in an early LBE-cooled Russian submarine showed that molten fuel (at least oxide fuel)
disperses in the heavy metal coolant in a coolable manner. To detect such a blockage in order to avoid
a longer term blockage propagation may require instrumentation at each subassembly outlet although
coolant activity measurements might be sufficient.
For a gas-cooled ADS hand calculations have shown that the core will melt in a few minutes if
the beam is not switched off in a LOHS accident [22]. The grace times in depressurisation accidents
may be even shorter; the one in a loss of circulation accident is probably somewhat longer. The
melting of a sub-critical fast core can lead to a re-criticality. In reactivity or beam power increase
accidents it is not clear whether an efficient molten fuel sweepout from the core is likely. Otherwise
there will be fuel blockage formations that may propagate.
For an HTR-ADS, the inner fast zone will also have a low heat capacity. Thus it will also have
short grace times for LOHS, loss-of-circulation or depressurisation accidents without beam shut off.
3. Beam shut off possibilities
In principle it is simpler and faster to switch off an accelerator than to insert shutdown rods in a
critical system. The manual switching off of the accelerator based on increased temperatures (which
will occur in all the important ADS accident scenarios) will remain an important option. There should
also be an automatic interruption based on high temperature readings. If these methods fail, a meltrupture
disk in the wall of the beam pipe that would fail and flood this vacuum tube with heavy liquid
metal would be useful as a last resort [23]. The STAR-CD calculations of an LOHS and a combined
LOHS and LOF in the Ansaldo design show the effect of the melt-rupture disk failing after different
heat ups of the coolant. It can be seen that for the combined LOHS and LOF the triggering (i.e. the
melting of the solder material around this disk) should occur sooner. This is a difficult natural
circulation problem with a 3 MW heat source in the upper part of the primary pool together with the
decreasing core decay heat and the ex-vessel air cooling. The latter can only remove the entire heat
when the vessel temperature is around the creep limit. But this is only reached after nearly 2 days in
the LOHS accident. In the unlikely case of LOF + LOHS accident without beam shut off a map of hot
coolant will collect in the upper part of the vessel due to the loss of forced circulation. After about
7 hours the wall temperature will surpass the creep limit. When a core with a higher thermal power
and a stronger spallation source is used in the same vessel, the grace time gets shorter [23]. In a gascooled
ADS this passively activated beam blocking is not possible. We are presently investigating
further approaches for passively switching off the accelerator in heavy metal cooled systems. These
are based on the thermal and electrical conductivity of liquid metals.
503

Figure 3. Beam blocking
10 min (200 K) after LOHS initiation
Figure 4. Beam blocking
3 min (60 K) after LOHS + LOF
1300
0.31
1400
0.4
1200
1100
0.30
0.29
1200
0.3
Temperature, K
1000
900
800
Wall temperature
Core outlet temperature
Core outlet velocity
0.28
0.27
0.26
Velocity, m/s
Temperature, K
1000
800
0.2
0.1
Velocity, m/s
700
600
0.25
0.24
600
Wall temperature
Core outlet temperature
Core outlet velocity
0.0
500
0.23
0 10 20 30 40 50
Time, hours
400
0 10 20 30 40
Time, hours
-0.1
4. Emergency decay heat removal
Emergency decay heat removal is necessary after beam shut off in the case of a station blackout
accident or a LOHS e.g. due to the lack of feedwater. Liquid metals with their good heat conductivity
allow ex-vessel cooling by natural air (RVACS) or water circulation or in – vessel cooling by direct
reactor auxiliary cooling systems (DRACS). Ansaldo [5] is proposing a new approach for an RVACS
(see Figure 5). This new design forms an additional barrier and would prevent fission product releases
even in the remote eventuality of a guard vessel failure. Moreover, it could still cool a disintegrated
core. Since this design consists of many U-shaped pipes, even the failure of a few of them would not
be a problem. Calculations with the STAR-CD code have shown that the decay heat can be easily
removed for the 80 MWt design
Figure 5. Schematic of new Ansaldo RVACS
Figure 6. Coolant temperatures and velocities
at the top of the core during a station blackout.
The initial temperature decrease is due
to the large momentum of the LBE flow
660
0.4
650
0.3
Temperature, K
640
630
Temperature
Velocity
0.2
0.1
Velocity, m/s
620
0.0
610
0 10 20 30
-0.1
Time, hours
504

Another innovative approach is part of the BREST-300 design [11]. A thick concrete wall that
contains pipes through which water is circulated surrounds the main vessel.
The emergency decay heat removal in gas-cooled systems can also be done passively for systems
not much larger than 600 MWt. However, in a depressurisation accident, diesel-driven blowers are
needed to remove the decay heat.
All heavy metal cooled critical reactors can also use the above mentioned passive means for
removing the emergency decay heat.
5. Conclusions
It has been shown that heavy-metal cooling of accelerator-driven system has considerable
advantages regarding the behaviour of an ADS in severe accident conditions and in particular when
the proton beam is not switched off during an accident initiation. Heavy metal cooling also allows
passive approaches for the beam blocking or switching off the accelerator. An equally important
aspect is the possibility of passive emergency decay heat removal systems that can also be used for
critical reactors with heavy metal cooling. In contrast to sodium heavy metals do not burn and react
mildly with water.
However, it should also be mentioned that the functionality of heavy metal cooling in normal
operation is not yet well established in Western countries. Russian Federation has a considerable
advantage in this regard because of its earlier experience with lead/bismuth cooling in submarine
reactors. But considerable research on the corrosion behaviour and thermal hydraulics is now also
underway in Western countries.
On the other hand gas-cooling is well understood and if one wanted to build an ADS in the near
future, the functionality of a gas-cooled system would be more assured.
Once Pb-Bi or Pb-cooling is well established, critical reactors with heavy metal cooling that are
also very safe, could be built for clean energy generation. These systems would benefit strongly from
the research on heavy metal cooling for accelerator driven systems. A possible future scenario using
both critical and accelerator-driven systems is shown below. In the nearer future one could start using
thorium based fuels in LWRs to reduce excess plutonium and to avoid the generation of higher
actinides.
505

Figure 7. A possible scenario for nuclear power development
Time
LWR
U/Pu
LWR
Pu/Th
233
U +
ADS
+
Pb/Bicooled
+Electricity
+Heat
+H2
+Desalination
Critical
Pb/Bicooled
Deep
repository
Purex Thorex
Pyro- or simplified
pyro-processing
Limited
repository
506

REFERENCES
[1] US DOE, Generation IV Workshop, Bethesda, MD, USA, May 1-3, 2000.
[2] Bonnaure P., Mandrillon P., Rief H., Takahashi H., Actinide Transmutation by Spallation in the
Light of Recent Cyclotron Development, ICENES86, Madrid, July 1986.
[3] Rubbia C. et al., (1995), Conceptual Design of a Fast Neutron Operated High Power Energy
Amplifier, CERN/AT/95-44 (ET).
[4] Rubbia C., Buono S., Kadi Y., Rubio J.A., Fast Neutron Incineration in the Energy Amplifier as
Alternative to Geologic Storage: the Case of Spain, CERN/LHC/97-01(EET).
[5] Ansaldo Nucleare, (1999), Energy Amplifier Demonstration Facility Reference Configuration –
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[6] Cheng X., Knebel J.U., Hofmann F, Thermal Hydraulic Design of an ADS with Three
Spallation Targets, ADTTA’99, Prague, Czech Republic.
[7] Carluec B., Proposal for a Gas-cooled ADS Demonstrator, ADTTA’99, Prague, Czech Republic.
[8] Rodriguez C., Baxter A., Transmutation of Nuclear Waste Using Gas-cooled Reactor Technologies,
ICONE8, Baltimore, April 2000, paper not on CDROM, contact Carmelo.Rodriguez@gat.com.
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507

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Driven Systems, Annual Meeting on Nuclear Technology 2000, May 2000, Bonn, Germany.
508

A SIMPLE MODEL TO EVALUATE THE NATURAL CONVECTION IMPACT ON
THE CORE TRANSIENTS IN LIQUID METAL COOLED ADS
A. D’Angelo, G. Bianchini and M. Carta
ENEA – C.R. Casaccia
via Anguillarese 301, 00060 S. Maria di Galeria (Roma), Italy
P. Bosio, P. Ravetto, M.M. Rostagno
Dipartimento di Energetica, Politecnico di Torino
Corso Duca degli Abruzzi 24, 10129 Torino, Italy
Abstract
A simple model has been developed at ENEA Casaccia to preliminarily evaluate the primary-coolant
natural convection impact on core-dynamics of an 80 Mw energy amplifier demonstration facility
(EADF) fuelled by U-Pu mixed oxides and cooled by a molten lead-bismuth eutectic. The model has
been already coupled with the Tieste-Minosse “point dynamics” code elaborated at ENEA Casaccia,
and in the near future will be easily coupled to the codes that are being developed at the Politecnico di
Torino in the frame of the cooperation with ENEA on multi-dimensional investigations of solid fuelled
ADS core dynamics. After the model formulation, some preliminary results on the primary-coolant
impact on the EADF core dynamics are presented in this paper.
509

1. Introduction
In the EADF design [1,2], the primary-coolant flow is assured by natural convection and
enhanced by a particular system of cover gas injection into the bottom part of the riser. The aim of this
work is to present a simple model that has been recently developed at ENEA Casaccia to preliminarily
investigate the capability that the primary-coolant natural convection has to mitigate typical core
transients in ADS. The model, that allows to quickly evaluate the primary-coolant velocity variation
induced by the natural convection during the core transients, has been already implemented in the
Tieste-Minosse code [3,4] recently developed at ENEA to preliminarily investigate core transients in
solid fuelled ADS [5]. Results obtained by taking into account the primary-coolant velocity variations
evaluated by means of the simple natural convection model are also presented and compared with the
corresponding transient trends obtained by considering a constant primary-coolant flow. In particular,
the present paper results concern transients induced by: 1) the proton beam interruption [6-9] or short
duration beam trips [5,10]; 2) the proton beam jumps [5,11,12,13]; 3) the loss of the primary-flow due
to the failure of the active system of convection enhancement.
2. The simple model
We face the subject of the natural convection process, in the attempt to derive a simplified model
to be easily used in core dynamics codes for solid fuelled ADS. We aim to obtain a preliminary
evaluation of the natural convection impact on liquid metal cooled ADS dynamics. In order to make
this quick evaluation, we assume the possibility of reducing a three-dimensional configuration (the
actual plant) into a one-dimensional model. Practically, this assumption leads to neglect the recirculation
phenomena into the vessel pool.
Moreover, we will take into account the heat exchange from the fuel to the coolant into the core
and from the primary to the secondary loop coolant into the heat exchangers, but we will neglect the
heat exchange between the primary coolant and the loop walls. Finally, we will consider the coolant
movement, but we will neglect the heat conduction along the coolant.
To look for the hydraulic solutions of a single-phase fluid, flowing in a supposed onedimensional
loop, in the following we will indicate with:
ν the mean velocity of the primary-coolant
ρ the primary-coolant density
σ the generic section area of the primary loop
Q V
the volume flow rate≡ ρ σ
Q m
the mass flow rate ≡ Q σ = ρσ ν
V
G the specific flow rate ≡ Q m
/ σ.
As a consequence of the mass conservation, assuming a generic volume V inside two normal
sections of a stream-tube, we can write the following equation of continuity:
∂
∫ ρ ( t
∂ t
)
V
dV
=
Q
m
− Q ′
m
(1)
510

That in steady state conditions becomes the mass flow rate conservation law:
mono-dimensional problems):
ρ σ
v′
= v
ρ′
σ ′
Q = Q′
i.e. (for
where ν can be considered a reference velocity, while apostrophes indicate quantities relevant to a
different loop section.
m
m
(2)
2.1 Pressure drops
and
A throttling inside the loop induces local and linear pressure drops:
∆
P
=
f
c
l
d
∆
2
2
G Qm ρ
P = K = K = K
2
2
G
2 ρ
2ρ
=
f
c
l
d
2ρσ
Q
m
2 ρσ
2
v
2
2
2
=
f
c
l
d
ρ
v
2
where: K is a local pressure drop coefficient, l is a channel length, d is an equivalent diameter and f c
is
a linear pressure drop coefficient.
Usually, linear pressure drops prevail in the main components of the cooling loop. Therefore, the
main contributes to the pressure drop can be approximated by the following sum on different loop
segments:
∆ P
i
i
2
i
i
i
i
2
m
2
iσ
i
li
G
l
i
Q
= ∑ f
ci
= ∑ f
ci
=
d 2 ρ
d 2 ρ
∑
i
f
ci
l
i
d
i
ρ
i
v
2
2
i
2
(3)
(4)
(5)
By using the continuity law (2) to obtain ν i
as a function of the reference velocity ν, the equation
(5) becomes:
⎛
l
⎞
⎜
i
ρ
(6)
2 2
∆ P = ρ σ v
⎟
∑ f
ci
2
⎝ i 2 d ρ
i
σ
i
i ⎠
As a first approximation, the total pressure drop could be considered to be proportional to ρ ν 2 .
Besides considering constant the geometrical characteristics as (σ i
, l i
, σ j
), we have other two
significant approximations to assume constant the term between brackets:
• Considering constant the Reynolds number functions (f ci
,K j
coefficients).
• Considering close to unity the density ratios in different loop positions:
ρ
≈ 1 ; ≈ 1
ρ
ρ (7)
i
ρ j
511

Equation (6) can be improved by considering how the pressure drop coefficients f ci
actually
depend on the Reynolds number. In our case, owing to the turbulent flow condition into the (core and
heat exchangers) smooth channels, the following Blausius correlation [14] can be used:
−0.25
= 0.32
e
f c
R
were
e
≡
REYNOLDS NUMBER
=
D e
ρ v
µ
=
Q
m
D e
S µ
e
(8)
and µ is the fluid viscosity, S e
is the equivalent flow area of the considered channel or component and
D e
is the corresponding equivalent diameter.
By considering the (8) Blausius correlation, the equation (6) becomes:
1.75
m
Q
∆P
∝
ρ
= ρ
0.75
v
1.75
µ
0.25
(9)
The improved model (9), besides the hypothesis (7), needs only the following further
approximation:
0.25
µ i
0.25
µ
≈ 1
2.2 Gravitational pull and energy balance in steady conditions
In steady state convection, the kinetics energy dissipated by pressure drop is compensated by the
work made by the gravity field along the entire loop. By considering unit volumes, we can write this
balance in terms of pressure drops:
∆P g
= ∆P
where the gravitational pull ∆P g
is defined as:
∆
P g
≡
∫
r r
ρ g • d s
(10)
512

Moreover, the a leg of the assumed loop (see figure on side)
will be at constant temperature T (outlet and inlet refers to the
core flow) and b has to be at almost (within 1%) constant
temperature T inlet
to optimise the loop efficiency.
Thus, the ∆P evaluation can be particularly easy under some
additional hypotheses on the heat exchanger height and the
temperature distribution. In literature, the same height (and
therefore h a
= h b
) and linear temperature behaviour is often assumed
in the two components. Generally, we will not need such an
approximation for temperatures, nevertheless it can be worthwhile
to recall that it would lead to:
∆P
= g = gh
g
b
( −ρ
) = −ρgh
β T −T
)
ρ (11)
b
a
b
(
out in
where β is the volumetric dilatation coefficient of the coolant.
Actually, if we know the coolant temperature (i.e. density)
distributions along the components, the steady state equation (11)
can be substituted by more precise formulations, for instance the
following one:
∆ P t = 0)
= g
g
( h ρ + h ρ −h
ρ −h
ρ )
( (12)
exc
exc
b
b
core
core
a
a
2.3 The transient
Under time-dependent conditions, besides pressure drops and gravitational pull variations, also
kinetics energy variations must be considered. The general method is the classical one relevant to the
mechanics: infinitesimal work relevant to external forces equals the infinitesimal kinetics-energy
variation. However, practically, a detailed knowledge of the loop is needed. In fact, it is easy to verify
that, in any loop volume comprised between two sections and having length l and flow area σ, an
infinitesimal kinetics energy variation can be written as:
2
⎛ 1 2 ⎞ ⎛ 1 Q ⎞
m l
(13)
d⎜
mv ⎟ = d
⎜ l = QmdQm
ρσ
⎟
⎝ 2 ⎠ ⎝ 2 ⎠ ρσ
Owing to the fact that we are dealing with liquid coolant and we will apply our model to slow
temperature transients, the time derivative of equation (1) can be neglected to evaluate current flow
rates. Equation (13) shows that, if Q m
does not depend on the considered circuit section, the
infinitesimal variation of kinetics energy will depend on the considered section area σ. Practically, this
leads to the fact that in a loop characterized by several portions l i
having different flow area (σ i
), the
kinetics-energy variation has to be expressed as a summation of terms depending on the loop
geometry:
Qm
ρσ
( ∆P g
−∆P ) dV ≡σ
( ∆P
g
−∆P ) vdt=
∑li
dQm
= v dQm
∑li
= vLdQm
i
ρ σ
i
i
i
ρ σ
i
i
(14)
513

where L is obtained by a summation of different portion lengths, weighted by numerical coefficients
that can be approximated to be constant in the time:
ρσ v
(15)
i
L = ∑ li
= ∑li
ρ σ v
i
i
i
i
It is worthwhile to note that in pool cooling systems a unique definition of the loop length is not
easy to give. This matter rises because we are supposing a mono-dimensional problem, which actually
is at least two-dimensional. In any case, if we are able to evaluate the “effective length” L, equation
(14) leads to:
dQ
L
dt
m
dv
= σ ( ∆Pg
− ∆P)
i. e ρL
= ∆Pg − ∆P0
dt
(16)
Finally, if we remember the pressure drop formulation (9) and the initial steady state condition,
we can write:
ρL
dv
dt
= ∆P
g
− ∆P
g
ρ
0.75
v
1.75
0
ρ v µ
0.75 1.75
0 0
µ
0.25
0.25
0
(17)
that can be solved numerically in the following way:
∆v
k
1 ⎛
= P
L
⎜∆
ρ ⎝
gk
− ∆P
g 0
ρ
ρ
0.75
k−1
0.75
0
v
v
1.75
k−1
1.75
0
µ
µ
0.25
k−1
0.25
0
⎞
⎟ ∆t
⎠
k
(18)
i.e. by adding and subtracting ∆P g0
∆v
k
1 ⎡
= ⎢
ρL
⎢⎣
⎛
⎜
⎝ ρ
0
0.75 1.75 0.25
ρ
k −1
vk
−1
µ
k −1
( ∆Pgk
− ∆Pgo
) − ∆P
⎜
⎟
g 0
−1
∆tk
v
0.75 1.75
0
µ
0.25
0
⎞⎤
⎟⎥
⎠ ⎥ ⎦
where k indicates the generic time step relevant to the transient thermal-hydraulic solution.
In order to avoid numerical oscillations, we could also impose the limit of the asymptotic solution
to the evaluation of the resistant strength (per unit of volume), i.e. to the value of the pressure drop per
length unit. We will impose that the resistant strength will never exceed the gravitational pull. In
particular, a reduction of velocity will never occur as a consequence of a gravitational pull increasing
or an increase of velocity will never occur as a consequence of a gravitational pull decrease (as a
consequence of bad estimations of the resistant strength):
Practically, if ((∆P g k
>∆P g k-1
and ∆ ν k

2.4 The gravitational pull calculation
In order to complete the present model definition, we have to mention how to calculate the
gravitational pull ∆P g0
and ∆P gk
to be used in equation (18).
In steady state we use equation (12). For the further time steps, we have to directly apply
equation (10). Practically, we have to follow (every time step) the distribution of the coolant density
and velocity in the core, in the heat exchanger and in both the a and b loop legs. To do that, we have
also to apply a heat exchanger model 1 to evaluate the primary coolant variations of the heat exchanger
outlet temperature.
3. Results
The natural convection model presented above has been already coupled with the Tieste-Minosse
“point dynamics” code elaborated at ENEA Casaccia. Some Tieste-Minosse results, that can be useful
to preliminarily evaluate the natural convection impact on the EADF core dynamics, are presented in
this section.
3.1 The proton beam interruption and the short duration beam trips
Figure 1 shows the trends of the outlet coolant temperature induced by short duration beam trips
in the EADF. The neutron source transients have been performed either considering or neglecting the
inlet coolant velocity variations due to lead-bismuth natural convection.
It can be easily seen that the impact of the primary coolant natural convection on the outlet
coolant temperature trends is not significant for the EADF core dynamics. The main reason of this
result can be related to the presence of an active system of coolant flow enhancement in the EADF
design. In particular, the pull due to the system of cover gas injection into the bottom part of the riser
is about five times the gravitational pull due to the coolant natural convection. Practically, while this
active system of flow enhancement is working, the most of the pull that moves the coolant remains
constant and the gravitational pull variations due to the natural convection transient do not induce
significant effects.
1 The definition of a specific Heat Exchanger Model will be the object of a further paper and therefore is not
defined in the present one.
515

Figure 1. The natural convection impact on short duration trips and a definitive beam
interruption: Demo Facility Average Assembly outlet coolant temperatures.
673
663
temperature (K)
653
643
633
623
613
603
593
583
Red lines: the natural
convection has been
neglected by assuming
constant the coolant flow
rate (nominal value).
1s trip
2s trip
4s trip
6s trip
beam interruption
573
0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20
time (s)
3.2 The proton beam jumps
The full power proton beam is assumed to be suddenly dumped into the reactor during proton
beam jump events. If the EADF active system of natural convection enhancement is assumed to be
working during the transients, also the beam jumps will evolve without severe consequences and the
impact of the primary coolant natural convection on the temperature trends does not modify
significantly the EADF core dynamics. As in the beam trip cases, the natural convection does not
induce significant effects because the active system of flow enhancement (assumed to be regularly
working) ensures a predominant and constant pull during the transients.
On the contrary, the transient mitigation due to the lead-bismuth natural convection can be
important if the system of flow enhancement is pessimistically assumed not to be available during the
beam jump from low power event (start-up accident). Figure 2 shows the results obtained under this
pessimistic assumption, either considering or neglecting the inlet coolant velocity variations due to
lead-bismuth natural convection: in this case, the impact of the natural convection mitigation is
evident. In particular, if the lead-bismuth natural convection is not taken into account, the calculation
indicates that the transient results in severe consequences: the fuel clad melts in about 40 seconds. But,
if the inlet coolant velocity variations due to the natural convection are taken into account during the
transient, the results indicate that even the worst source transient can evolve without severe
consequences. Results show that the lead-bismuth natural convection can stabilise the core
temperatures well below the clad meting level. Figure 3 shows that, while the full power proton beam
is driving the EADF, the natural convection can be able to pull the lead-bismuth flow to about 50% of
the nominal flow. The pressure loss in the two components and along the loop in this new asymptotic
steady state is foreseen to be about one third of the nominal one (about 10 kPa instead of 30 kPa) and
the difference between outlet and inlet coolant temperatures about double of the nominal one (about
185° instead of 100°).
516

Figure 2. The natural convection impact on the unprotected beam jump from low power
(start-up accident), assuming that the coolant flow enhancement system is unavailable:
temperatures at the Demo Facility Average Assembly active fuel top.
1673
temperature (K)
1473
1273
1073
873
673
473
Trends obtained by assuming
constant the coolant flow rate:
temperatures rise up to the clad
melting level in about 40 s.
Trends obtained by taking into account the natural convection mitigation
0 10 20 30 40 50 60 70
time (s)
Clad melting temperature
Fuel-pellet centre
Fuel-pellet boundary
Inner Clad
Outlet Coolant
Figure 3. Unprotected beam jump from low power (start-up accident) assuming that the coolant
flow enhancement system is unavailable: trend of the Demo Facility coolant flow.
100
coolant flow (% of the nominal value)
90
80
70
60
50
40
30
20
10
coolant flow
no natural convection
0
0 10 20 30 40 50 60 70
time (s)
3.3 The unprotected loss of flow
The results presented above point out the natural convection capability to mitigate the
consequences of a possible failure of the EADF active system of coolant flow enhancement. In the
present section we further investigate this mitigation capability. In particular we assume the cooling
system failure at nominal conditions without any beam power variation. Practically we preliminarily
investigate the natural convection mitigation of an Unprotected Loss of Flow (ULOF) accident.
517

Figure 4 shows the temperature trend results during and after the assumed linear reduction of the
active system pull (from about 25 kPa (nominal pull) to zero in 10 seconds).
Figure 4. The natural convection impact on the unprotected loss of flow:
temperatures at the Demo Facility Average Assembly active fuel top.
1673
temperature (K)
1473
1273
1073
873
673
Trends obtained by assuming a linear
coolant flow reduction in 10s:
temperatures rise up to the clad
melting level in about 40 s.
Trends obtained by taking into account the natural convection mitigation
Clad melting temperature
Fuel-pellet centre
Fuel-pellet boundary
Inner Clad
Outlet Coolant
Inlet coolant
473
0 10 20 30 40 50 60 70
time (s)
Figure 4 results clearly show a particularly benign evolution of the ULOF event in the EADF:
core temperature variations remain limited to about 60-90 degrees if the natural convection mitigation
is taken into account. The main reasons for these limited temperature variations are the following: 1)
the results in Figure 4 confirm that the natural convection can largely mitigate the consequence of a
possible failure of the EADF active system; 2) as it is well known, in sub-critical systems the feed
back effects are much lower with respect to a critical systems.
4. Conclusions
A simplified model for the natural convection of the primary-coolant of solid fuelled ADS has
been proposed and used to preliminary evaluate the impact of the natural convection on core transient
behaviors. This mono-dimensional transient model neglects re-circulation phenomena and needs the
definition of the “effective length” of the loop.
In the EADF design, the natural convection pull is about 20% of the pull due to the active system
based on cover gas injection. Preliminary results confirm that while the active system of cover gas
injection is working, the impact of the primary coolant natural convection on EADF transient
behaviours is not significant. On the contrary, the natural convection mitigation of temperature
transients becomes clearly significant if the active system of cover gas injection is assumed to be
unavailable or to fail. The system of cover gas injection has been pessimistically assumed to be
unavailable during the transient induced by a beam jump from low power event. Moreover, an
Unprotected Loss of Flow transient has been analysed.
518

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520

COMPARATIVE STUDY FOR MINOR ACTINIDE TRANSMUTATION
IN VARIOUS FAST REACTOR CORE CONCEPTS
S. Ohki
Japan Nuclear Cycle Development Institute (JNC), Oarai Engineering Center
4002, Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki-ken, 311-1393, Japan
Abstract
A comparative evaluation of minor actinide (MA) transmutation property was performed for various
fast reactor core concepts. The differences of MA transmutation property were classified by the
variations of fuel type (oxide, nitride, metal), coolant type (sodium, lead, carbon dioxide) and design
philosophy. Both nitride and metal fuels bring about 10% larger MA transmutation amount compared
with oxide fuel. The MA transmutation amount is almost unchanged by the difference between sodium
and lead coolants, while carbon dioxide causes a reduction by about 10% compared with those. The
changes of MA transmutation property by fuel and coolant types are comparatively small. The effects
caused by the difference of core design are rather significant.
521

1. Introduction
Research and development of the fast reactor have been carried out in several nations with mixed
oxide fuel and sodium coolant chosen as standard components of the reactor core. In addition, the
transmutation of minor actinide (MA) nuclides using a fast reactor core has been investigated
extensively from the viewpoint of reducing the environmental burden of long-lived radioisotopes, that
is considered as one of the main features of the fast reactor.
Meanwhile alternative core concepts have been proposed to utilise liquid heavy metal or gas as a
coolant, and the design works of these cores are performed widely to exploit the merit of each coolant.
Concerning the fuel material, both mixed nitride and metal fuels are considered as the feasible
candidates. The MA transmutation properties have also been investigated for these core concepts made
from various fuel and coolant materials, and a good performance has been reported for each concept as
a result of its hard neutron spectrum.
In order to provide basic information for selecting candidates of the commercial-use fast reactor
core, this study presents a comparative evaluation of MA transmutation property for various fast
reactor core concepts employing the same analytical method and nuclear data. The differences among
oxide, nitride and metal fuels were examined, besides the three types of coolants, sodium, lead and
carbon dioxide were compared.
2. Calculation conditions
2.1 Definition of investigated cores
Various fast reactor cores investigated in this study are described below. All of them were
selected in terms of the practical feasibility prospected by conceptional design works.
Table 1 shows the three sodium-cooled large cores that examine the difference among oxide,
nitride and metal fuels (referred as Na-MOX, Na-MN and Na-Metal cores). In order to extract the
difference of fuel type the basic reactor performances (i.e. reactor power, cycle length, average fuel
burn-up, coolant pressure drop, etc.) were not changed. Maximum linear heat rate was also conserved
among these cores in the same level (limited up to about 430 W/cm). Fuel pin diameter, pin pitch and
number of sub-assemblies were adjusted in order to satisfy these conditions.
It seems that the nitride and metal-fuelled cores have better core characteristics than that of the
oxide-fuelled core. Their high heavy metal density causes a reduction in Pu enrichment as well as
burn-up reactivity loss. It increases also the breeding ratio.
The reference oxide fuelled core was designed and investigated as a feasible candidate for the
commercial-use fast reactor by JNC-Japan [1]. It is a conventional FBR core having homogeneous two
fuel enrichment zones surrounded by fertile blanket. One remarkable feature is that automatically
dropping absorber using a curie-point magnet is installed inside each of inner core sub-assembly. The
number of fuel pins in inner core sub-assembly is reduced for placing a guide tube of the safety
device. Nevertheless this does not influence on the MA transmutation property.
The investigated cores to examine the difference of coolants between sodium and lead are listed
in Table 2. The BREST-300 reactor proposed by RDIPE-Russia [2] was selected as a reference leadcooled
power reactor. It has a safety-oriented core concept such as almost zero burn-up reactivity, low
coolant pressure drop and low coolant void reactivity, which is totally different from a conventional
FBR design. The superior core characteristics are achieved by means of high heavy metal density of
522

nitride fuel, reflection effect of lead coolant and a sub-assembly without duct tube. BREST-300 has no
fertile blanket but large heavy metal inventory in other words it can be said that the blanket is included
into the active core.
Table 1. Core design parameters for the fast reactors compared in this study (1)
– Oxide fuel vs. nitride and metal fuels –
Na-MOX
Large-sized core
Na-MN
core
Na-Metal
core
Reactor power [MW th
] 3 800 Ä Ä
Operation cycle length [EFPD] 540 Ä Ä
Fuel exchange batch 5 Ä Ä
Average fuel burn-up [GW th
d/t] 150 Ä Ä
Core height [cm] 120 Ä Ä
Coolant Na Ä Ä
Coolant temperature [ o C] (outlet / inlet) 550/395 Ä Ä
Coolant pressure drop [kg/cm 2 ] ~3 Ä Ä
Fuel type (U, Pu)O 1.98
(U, Pu) 15 N U-Pu-10Zr
Pu isotopic vector [wt%]
3/52/27/9.5/7/1.5 Ä Ä
( 238 Pu/ 239 Pu / 240 Pu / 241 Pu / 242 Pu / 241 Am)
Fuel pin diameter [mm] 9.7 8.40 9.02
Smeared density [%TD] 82 80 75
Pin pitch/pin diameter 1.15 1.20 1.17
Number of fuel pins per sub-assembly
234/271 Ä Ä
(IC/OC)
Sub-assembly pitch [mm] 195.4 178.2 185.9
Number of sub-assemblies (IC/OC) 246/216 252/210 252/210
Heavy metal inventory (core) [t] 63 63 64
Pu enrichment [wt%] (IC/OC) 17.8/19.8 14.3/16.3 14.9/16.9
Burn-up reactivity loss [%dk/kk’] 2.9 1.4 1.9
Breeding ratio 1.04 1.14 1.11
523

Table 2. Core design parameters for the fast reactors compared in this study (2)
– Sodium coolant vs. lead coolant –
524
Na-MOX
Medium-sized core
Pb-MOX
core
Pb-MOX core
(Low pressure
drop)
Pb-MN core
(Low pressure
drop)
BREST-300
(Ref. [2])
(*IC/MC/OC)
Reactor power [MW th] 700 Ä Ä Ä Ä
Operation cycle length
540 Ä Ä Ä ~300
[EFPD]
Fuel exchange batch 5 Ä Ä Ä Ä
Average fuel burn-up
[GW thd/t]
150 Ä Ä Ä 60
Core height [cm] 120 Ä Ä Ä 110
Coolant Na Pb Ä Ä Ä
Coolant temp. [ o C]
(outlet/inlet)
550/395 Ä Ä Ä 540/420
Coolant pressure drop
~3 Ä ~1 Ä Ä
[kg/cm 2 ]
Fuel type (U, Pu)O 1.98
Ä Ä (U, Pu) 15 N Ä
Pu vector [wt%]
3/52/27/9.5/7/1.5/0/0 Ä Ä Ä 0.5/64/28/3.1/
( 238 Pu/ 239 Pu/ 240 Pu/ 241 Pu/ 242 Pu/ 241
1.7/2.1/0.1/0.5
Am/ 242m Am/ 243 Am)
Fuel pin diameter [mm] 9.7 9.7 9.7 8.87 9.1/9.6/10.4*
Pin pitch / pin diameter 1.15 1.27 1.40 1.45 1.49/1.42/1.31*
No. of fuel pins per S/A
271 Ä Ä Ä 114
(IC/OC)
S/A pitch [mm] 195.4 216.0 238.2 226.3 150
No. of S/As (IC/OC) 30/54 30/54 30/54 30/54 57/72/56*
Heavy metal inventory
(core) [t]
12 12 12 12 16
Pu enrichment [wt%]
18.8/23.9 19.0/24.7 21.1/26.8 17.8/22.6 14.0/14.0/14.0*
(IC/OC)
Burn-up reactivity loss
[%dk/kk’]
2.1 2.2 2.4 2.0 ~0
Breeding ratio 1.11 1.08 1.02 1.09 ~1

Table 3. Core design parameters for the fast reactors compared in this study (3)
– Sodium coolant vs. carbon dioxide coolant –
Na-MOX
Large-sized
core
Na-MOX core
(Equivalent to
ETGBR)
ETGBR
(Ref. [3])
Reactor power [MW th
] 3 800 Å 3 600
Operation cycle length [EFPD] 540 Å 344
Fuel exchange batch Å Å 5
Average fuel burn-up [GW th
d/t] 150 Å 120
Core height [cm] 120 Å 150
Coolant Na Ä CO 2
Coolant temperature [ o C] (outlet/inlet) 550/395 Ä 525/252
Coolant pressure drop [kg/cm 2 ] ~3 Ä ~3?
Fuel type Å Å (U, Pu)O 1.98
Fuel pin diameter [mm] 9.7 7.84 8.2
Pin pitch/pin diameter 1.15 1.30 1.55
Number of fuel pins per sub-assembly 234/271 271 169
(IC/OC)
Sub-assembly pitch [mm] 195.4 185.2 180.6
Number of sub-assemblies (IC/OC) 246/216 183/161 334/216
Heavy metal inventory (core) [t] 63 49 50
Pu enrichment [wt%] (IC/OC) 17.8/19.8 17.5/21.5 18.7/26.7
Burnup reactivity loss [%dk/kk’] 2.9 2.5 2.4
Breeding ratio 1.04 1.08 1.08
In the present investigation, a series of lead cooled cores (Pb-MOX, Pb-MN cores) was prepared
starting from a conventional medium-sized Na-MOX core as shown in Table 2. These cores provide
the effects of a replacement of sodium to lead, a reduction of coolant pressure drop and an
employment of nitride fuel under keeping the reactor performance in the same level. It is found that
employment of the lead coolant and the reduction of coolant pressure drop make the core
characteristics worsen while the nitride fuel improves them. The rest of the differences between
Pb-MN core and BREST-300 are to be considered all together as a difference of design philosophy.
The sub-assembly specification of the medium-sized Na-MOX core is the same as the large
Na-MOX core appeared in Table 1, excepting that the safety devices are not placed inside the inner
core sub-assemblies.
Table 3 presents the cores prepared for comparing coolant effects between sodium and gas. The
ETGBR designed by NNC-United Kingdom [3] was chosen as a feasible concept of gas cooled fast
reactor. It has a two-region homogeneous core employing MOX fuel and carbon dioxide gas coolant.
Its active core height is larger than that of the reference large-sized Na-MOX core because there is no
need to concern about the coolant void reactivity, this is one of the merits of gas cooled fast reactor.
The other design parameters (reactor power, cycle length and average fuel burn-up) are also different.
To compare these cores under the same condition, a Na-MOX core equivalent to ETGBR was
prepared as shown in Table 3. Comparison among the three cores can classify the observed differences
into the effects of coolants and design parameters. Almost the same burn-up characteristics are
obtained for both the equivalent Na-MOX core and ETGBR except that a fairly high Pu enrichment is
needed in the outer core of ETGBR.
525

2.2 Representation of MA transmutation
The MA treated in this study was assumed to come from LWR spent fuel with five-year cooling
time before reprocessing. The isotopic composition of the MA is shown in Table 4, which was
calculated by the ORIGEN2 code [4]. It consists mainly of fertile MA nuclides such as 237 Np, 241 Am
and 243 Am.
Table 4. Compositions of minor actinides from LWR waste*
Nuclide
Composition (wt%)
237
Np 49.14
241
Am 29.98
242m
Am 0.08
243
Am 15.5
242
Cm 0.0
243
Cm 0.05
244
Cm 4.99
245
Cm 0.26
* Discharged from PWR (35 GWd/t) and cooled for 5 years before reprocessing.
MA nuclides were homogeneously distributed into all the core fuel in replacement of heavy metal
nuclides. A content of MA was considered up to 5wt% of the fuel heavy metal amount. The Pu
enrichment was adjusted to assume the same minimum-required reactivity all through the operation
cycle.
Following net MA transmutation amount per cumulative power was used in this study as a
quantity representing the transmutation property:
MA transmutation amount
[kg/GW th
/year]
= (MA inventory at BOC [kg] – MA inventory at EOC [kg])
/Reactor power [GW th
]/Operation cycle length [year],
where the MA transmutation amount was divided by reactor power and cycle length in order to
compare various reactor cores of different specifications. Note that the above MA transmutation
amount has a dimension of transmuted amount per energy emission from a reactor.
2.3 Method of calculation
Neutron flux and depletion calculations were carried out by the CITATION code [5] in diffusion
approximation where core geometry was modeled in two-dimensional RZ representation and a
7-group energy structure was used. Seven-group effective cross sections were collapsed from the
adjusted cross section library ADJ98 [6] based on the evaluated nuclear data JENDL-3.2 [7]. The
depletion calculation was performed until the core compositions settled down to a state of fuel cycle
equilibrium. Neutron flux was normalised at each burn-up step using the values of fission energy
emission recommended by Sher [8], where the contributions from capture gamma heat were also taken
into account.
526

2.4 Comparison of neutron spectra
The neutron spectra for the fast reactor cores investigated in this study are shown in Figure 1.
Transmutation of MA includes capture reactions ( 237 Np, 241 Am), fissile-type fission reactions ( 242m Am,
245
Cm) and threshold fission reactions ( 237 Np, 241 Am, 243 Am, etc.). One-group cross-sections for these
transmutation reactions depend on the shape of neutron spectrum. The former two types of reaction are
enhanced by the flux in keV energy region, the latter one is determined by the flux in MeV energy
region.
Concerning the fuel types, a harder neutron spectrum is observed in the order of metal, nitride and
oxide fuels (Figure 1(a)). For nitride and metal fuels, it should be noted that the decrease of flux in
MeV energy region occurs after the normalisation due to the lack of resonance of oxygen around
500 keV.
Figure 1. Comparison of neutron spectra at core center
(a) Oxide fuel vs. nitride and metal fuels
1D02;
1D01
1D0HWDO
1HXWURQÃ(QHUJ\Ã>H9@
(b) Sodium coolant vs. lead coolant
(c) Sodium coolant vs. carbon
dioxide coolant
1D02;Ã0HGLXPVL]HG
3E02;
3E02;Ã/RZÃSUHVVXUHÃGURS
3E01Ã/RZÃSUHVVXUHÃGURS
%5(67
1D02;
1D02;Ã(TXLYDOHQWÃWRÃ(7*%5
(7*%5
1HXWURQÃ(QHUJ\Ã>H9@
1HXWURQÃ(QHUJ\Ã>H9@
527

Figure 1(b) shows the cases for the comparison of sodium and lead. It is found that the lead
coolant reduces the neutron flux in MeV energy region, which is caused by inelastic scattering of lead.
Neutron hardening by lead coolant seems smaller than that obtained by nitride fuel. In addition, the
differences of reactor size, design parameters as well as the design philosophy do not drive any
significant change on the neutron spectrum. It is possible to say that the fuel and coolant types mainly
determine the neutron spectrum.
Carbon dioxide makes the neutron spectrum harder compared with sodium (Figure 1(c)). The
comparison also shows that the difference of design parameters in the two Na-MOX cores does not
alter the neutron spectrum so much.
3. Results and discussion
3.1 MA transmutation properties
The results of calculation for MA transmutation characteristics are indicated in Figure 2 below.
Figure 2. Comparison of MA transmutation properties (LWR discharged MA)
(a) Oxide fuel vs. nitride and metal fuels
1D02;
1D01
1D0HWDO
0$+0 ZWÈ
*UDGLHQWÃRIÃWKHÃOLQHÃ>\HDU@
ÃÃ1D02;ÃÃÃÃÃÈ
ÃÃ1D01ÃÃÃÃÃÃÈ
ÃÃ1D0HWDOÃÃÃÈ
0$+0 ZWÈ
0$Ã,QYHQWU\ÃDWÃ%2&Ã>NJ*:WK@
(b) Sodium coolant vs. lead coolant
1D02;Ã/DUJHVL]HGÃFRUH
1D02;Ã0HGLXPVL]HGÃFRUH
3E02;
3E02;Ã/RZÃSUHVVXUHÃGURS
3E01Ã/RZÃSUHVVXUHÃGURS
%5(67
0$+0 ZWÈ
0$+0 ZWÈ
(c) Sodium coolant vs. carbon
dioxide coolant
1D02;
1D02;Ã(TXLYDOHQWÃWRÃ(7*%5
(7*%5
0$+0 ZWÈ
*UDGLHQWÃRIÃWKHÃOLQHÃ>\HDU@
1D02;Ã/DUJHVL]HGÃFRUHÃÈ
1D02;Ã0HGLXPVL]HGÃFRUHÃÈ
3E02;ÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÈ
3E02;Ã/RZÃSUHVVXUHÃGURSÃÈ
3E01Ã/RZÃSUHVVXUHÃGURSÃÃÈ
0$+0 ZWÈ %5(67ÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÈ
0$Ã,QYHQWU\ÃDWÃ%2&Ã>NJ*:WK@
*UDGLHQWÃRIÃWKHÃOLQHÃ>\HDU@
Ã1D02;ÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÈ
Ã1D02;(TXLYDOHQWÃWRÃ(7*%5ÃÈ
Ã(7*%5ÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÃÈ
0$+0 ZWÈ
0$Ã,QYHQWU\ÃDWÃ%2&Ã>NJ*:WK@
528

The vertical and horizontal axes represent the MA transmutation amount defined in the previous
section and MA inventory at BOC, respectively. The gradient of the line stands for MA transmutation
rate.
It is found that the nitride and metal-fueled cores bring about 10% larger MA transmutation
amount compared with the oxide-fueled core (Figure 2(a)). In addition the effect of reduction of Pu
enrichment appears as an upward transfer of the graphic lines at the point of MA/HM=0wt%.
From a comparison of sodium and lead cooled cores in Figure 2(b), it is shown that the MA
transmutation amount is almost unchanged by the substitution of sodium to lead, reduced by the
decrease of the coolant pressure drop, and increased by the utilization of nitride fuel. However the
differences caused by a change of reactor power and the unique core concept of BREST-300 are rather
significant. Especially, large heavy metal inventory of BREST-300 enables to obtain a larger MA
transmutation amount. The transmutation rate of BREST-300 is almost same as that of Pb-MN core,
there might be a compensation effect arise from the differences of reactor specifications such as lead
reflector, Pu isotopic composition, cycle length, fuel burn-up, etc.
Figure 3(c) presents the difference between sodium coolant and carbon dioxide coolant. It is
found that the MA transmutation amount decreases about 10% from sodium to carbon dioxide when
compared in an equivalent reactor condition. It should be noted that the difference in core specification
between the two Na-MOX cores has caused considerable changes.
As a result, it is found that the differences in MA transmutation property arising from the
variation of fuel and coolant types are comparatively small, that is within ±10%. The effects caused by
the core design difference are rather significant. Since it is possible to construct various types of
reactor core even using the same fuel and coolant material, the reactor design philosophy might be
more important factor which changes the MA transmutation property. If we put more priority on MA
transmutation efficiency rather than core characteristics, it is possible to say that nitride and metal
fuels have more potential for MA transmutation.
3.2 Breakdown of the changes in MA transmutation
More detailed examination of the changes in MA transmutation more showed some interesting
results. The difference of MA transmutation amount can be divided into the effects of neutron flux
level, neutron spectrum and MA inventory. When we compare the MA transmutation rate, only the
factors of flux level and spectrum are to be considered. Result of the analysis is shown in Table 5,
where the effects are extracted from comparing the cores in the equivalent reactor condition.
Table 5. Analysis of the changes in MA transmutation amount
(LWR discharged MA, MA/HM = 5wt%)
Fuel type Reactor power Coolant type
MOXÅMN MOXÅMetal
LargeÅ
Medium
NaÅPb NaÅCO 2
Total flux level +23% +31% -27% +10% -14%
Neutron spectrum* -3% -22% -1% -13% -9%
MA inventory
per reactor power
-8% -2% +17% +1% +10%
Net +12% +8% -10% +2% -13%
* Including the effect of Pu enrichment change.
529

It is found that the higher MA transmutation amount observed in nitride and metal-fueled cores is
the consequence of the increase in neutron flux level.
When the reactor power is reduced, the worsened neutron economy causes the decrease in flux
level and the increase in MA inventory. This can also be seen in the substitution from sodium to
carbon dioxide, which seems to be resulted from the relatively poor neutron economy due to high
neutron leakage.
Thought the net MA transmutation properties are the same between sodium and lead cooled cores,
there exits a cancellation of the flux level and the neutron spectrum effects.
The effect of neutron spectrum depends on MA nuclide composition. Then a decomposition of the
neutron spectrum effect into individual reaction process was carried out as shown in Figure 3.
Figure 3. Effect of neutron spectrum change for individual MA transmutation process
(LWR discharged MA, MA/HM=5wt%)
Transmutation by
capture reaction
Transmutation by
fission reaction
Creation from Pu
Net transm utation
Np237→ Pu238
Am241→ Pu242
Am241→ Cm242→ Pu238
Np237
Am241
Am242m
Am243
Cm244
Cm245
Pu→ Am
MOX→ Metal
Na→ Pb
Na→ CO2
-30 -20 -10 0 +10 +20 +30
Change in MA transmutation amount (%)
It turns out that the spectrum effect on the transmutation of LWR discharged MA consists mainly
from the capture reactions of 237 Np and 241 Am, as well as the creation of Am from Pu. Especially for the
replacement of oxide fuel by metal fuel, the large negative contribution of capture reaction processes
is compensated by the large positive contribution from a reduction of MA creation from Pu.
The tendency of the variation of each reaction process is consistent with the changes of the
neutron spectrum and the Pu enrichment. For most nuclides, transmutation components by capture and
fission reactions are getting reduced by the spectrum hardenings, excepting the increase in the
threshold fission reactions by the substitution from sodium to carbon dioxide. It can be said that a
hardening of neutron spectrum does not always give a positive contribution to the MA transmutation
performance.
3.3 Effects on the core characteristics
Effects on the core characteristics by the MA loading were reviewed. Changes of core
characteristics by 5% of MA introduction are shown in Table 6. By the role of fertile material of MA
there occur the decreases in Pu enrichment and burn-up reactivity loss. However the coolant void
reactivity and the Doppler constant are changed towards not-safe direction due to the change of direct
530

and adjoint flux. Relatively large increase of the coolant void reactivity on BREST-300 is caused by a
larger amount of lead in the core. Since the tendency for the changes in core characteristics looks
similar for every type of the cores, it is possible to say that neither fuel nor coolant types brings a
prominent penalty to the core characteristics induced by MA loading.
Table 6. Changes in the core characteristics by MA loading
(LWR discharged MA, MA/HM=0.5wt%)
Na-MOX
largesized
core
Na-MN
core
Na-Metal
core
Pb-MOX
core
BREST-300
(*IC/MC/OC)
ETGBR
Change in Pu
enrichment [wt%]
(IC/OC)
Change in burn-up
reactivity [%dk/kk’]
Change in breeding
ratio
Change in coolant
void reactivity
[%dk/kk’]
Change in Doppler
constant [10 -3 Tdk/dT]
-1.2/-1.4 -2.0/-1.5 -1.7/-1.4 -1.1/-1.0 -0.1/+0.1/+0.2* -1.1/-0.4
-1.7 -2.2 -2.0 -1.4 -0.7 -1.1
+0.01 +0.04 +0.03 0.00 -0.06 -0.01
+0.47 +0.49 +0.44 +0.55 +0.81 Not
available
+1.8 +1.6 +1.1 +1.0 +1.6 +1.3
4. Conclusions
MA transmutation properties for various fast reactor cores were compared in the equivalent
condition as a power reactor. The observed differences were classified into the effects of fuel, coolant
type and design philosophy separately. It is concluded that there shows no significant difference in
MA transmutation amounts and rates arising from the variation of fuel types (oxide, nitride, metal) and
coolants (sodium, lead, carbon dioxide). The effects caused by the difference of core design are rather
important. If we stand on the viewpoint that core characteristics should be compensated for improving
the MA transmutation efficiency, it is possible to say that nitride and metal-fueled cores have more
potential for MA transmutation than oxide fueled core. By means of breaking down the MA
transmutation amount into the effects of flux level, spectrum and MA inventory, the differences of
MA transmutation property were analyzed more in detail.
531

REFERENCES
[1] T. Ikegami, H. Hayashi, M. Sasaki, T. Mizuno, K. Kawasima, N. Kurosawa, Y. Sakashita and
M. Naganuma, Design Study on the Core Characteristics of Sodium Cooled Fast Reactor –
Results in FY1999, JNC TN9400 2000-068, (March 2000).
[2] E.O. Adamov, V.V. Orlov, V.S. Smirnov, A.I. Filin, V.S. Tsykunov, A.G. Sila-Novitsky and
V.N. Leonov, Progress in Lead-cooled Fast Reactor Design, Proc. Int. Conf. on Design and
Safety of Advanced Nuclear Power Plants (ANP’92), Vol. 2, p. 16.6-1 (October 1992).
[3] T.A. Lennox, D.M. Banks, J.E. Gilroy and R.E. Sunderland, Gas Cooled Fast Reactors,
ENC’98, Transactions Vol. IV, p. 60 (October 1998).
[4] A.G. Croff, A User’s Manual for the ORIGEN2 Computer Code, ORNL/TM-7175 (1980).
[5] T.B. Fowler, D.R. Vondy and G.W. Cunningham, Nuclear Reactor Core Analysis Code:
CITATION, ORNL-TM-2496, Rev. 2 (1970).
[6] K. Yokoyama, K. Numata and M. Ishikawa, Development of the Adjusted Nuclear Cross-section
Library Based on JENDL-3.2 for Large FBR, JNC TN9400 99-042, (April 1999).
[7] T. Nakagawa, K. Shibata, S. Chiba, T. Fukahori, Y. Nakajima, Y. Kikuchi, T. Kawano,
Y. Kanda, T. Ohsawa, H. Matsunobu, M. Kawai, A. Zukeran, T. Watanabe, S. Igarasi,
K. Kosako and T. Asami, Japanese Evaluated Nuclear Data Library Version 3 Revision-2,
JENDL-3.2, J. Nucl. Sci. Technol., 32, 1259 (1995).
[8] R. Sher, Fission Energy Release for 16 Fissioning Nuclides, Proc. Specialists’ Mtg. Nuclear
Data Evaluation and Procedures, Upton, New York, 1980, BNL-NCS-51363.
532

STUDY ON A LEAD-BISMUTH COOLED
ACCELERATOR DRIVEN TRANSMUTATION SYSTEM
Hideki Takano, Kenji Nishihara, Kazufumi Tsujimoto, Toshinobu Sasa,
Hiroyuki Oigawa, Kenji Kikuchi, Yuichiro Ikeda, Takakazu Takizuka, Toshitaka Osugi
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 Japan
Abstract
Transmutation of minor actinides (MA) and iodine was studied by using a lead-bismuth cooled ADS
with 800 MWt, and it is shown that amount of MA and iodine produced from nine LWRs per year can
be simultaneously transmuted. The mass flows of MA and iodine are investigated in a future
symbiosis system for transmutation consisting of UO 2
/MOX-LWRs, FBRs and ADSs. Lead-bismuth
technologies are discussed. Current activities of design studies for the experimental facilities for ADS
technology demonstration are reviewed.
533

1. Introduction
The management of the high-level radioactive wastes (HLW) is one of the important key issues in
nuclear society at present. Various concepts for transmuting long-lived radioactive nuclides contained
in HLW to shorter-lived or stable nuclides have been proposed to reduce the risk from long-term
toxicity. Reactors with a hard neutron spectrum have capability to burn minor actinides (MA) such as
neptunium, americium and curium which dominate the long-term toxicity of spent fuels.
Recycling the all actinides and some long-lived FPs into fast breeder reactor (FBR) and the
accelerator driven system (ADS) to close the fuel cycle from a view point of actinide confinement is,
therefore, one of the promising options to be considered in solving the problem.
A new concept of nitride fuel cycle system based on pyrochemical reprocessing has been
proposed, and excellent core performance of the lead cooled nitride FBR could provide design
flexibility of reactor systems for energy production and/or reactors for burning or transmuting
long-lived radioactive nuclides [1,2]. For utilising the nitride fuel, the effect of 15 N enrichment on
nuclear characteristics and the evaluation of toxicity of 14 C generated from 14 N was appeared, and
excellent performance for the minor actinide (MA) transmutation was shown [3]. Furthermore, the
current status for the liquid heavy metal Pb-Bi technologies were investigated [4].
The symbiosis concept of the fuel cycle system based on nitride fuelled fast reactors consists of
base load reactors. The base load reactor produces electric power and extra plutonium fuel if needed,
and the MA (about 1 wt%) generated by themselves are recycled [2]. The MA transmuter designed in
the present study has simultaneously a role of incineration for Pu, MA and 129 I generated from UO 2
and
MOX fuelled LWRs, and FBRs systems. A part of the recovered Pu and minor actinides by
pyrochemical reprocessing are recycled into the ADS transmuters to adjust to the excess or remained
plutonium and to transmute the MA and iodine.
Furthermore, accumulation and transmutation of MA and 129 I based on future symbiosis recycle
system are investigated for introducing the accelerator driven system (ADS) with 800 MWt. It is
shown that in the scenario of nuclear plant capacities for maximum 140 GWe, which consists of
LWRs and FBRs, the introduction of ADS can play a significant role as “Transmuter” in the back-end
of fuel cycle.
A conceptual design study based on the experimental program for development and
demonstration of accelerator driven transmutation technology under the project plan of the High
Intensity Proton Accelerator and the OMEGA Programme at JAERI has been done. And the current
activities of the design studies for the experimental facilities for ADS technology demonstration are
reviewed.
2. Design study of ADS
Transmutation of MA and iodine was studied by using a lead-bismuth cooled ADS with
800 MWt. MA (Am, Cm and 237 Np) are most dominant contributors for long-term potential hazard in
spent fuel. On the other hand, long-lived fission product nuclide 129 I will be recovered as AgI
formation in reprocessing system for spent fuels. And the iodine is soluble in water and one of the
most troublesome nuclide on the geological disposal technology, though its potential hazard is smaller
than those of MA Thus transmutation for this nuclide is strongly expected as one of troublesome
isotopes from a viewpoint of waste disposal. The iodines are loaded axially and radially with the form
of NaI around the MA-fule core in ADS, and it was shown that an ADS can transmute the MA and
iodines generated from 9 or 10 units of LWR with 33 GWd/t per year as shown in Table 1. The
conceptual design of the ADS is shown in Figure 1.
534

Proton beam-trip analysis: In an ADS, a beam trip causes a abrupt drop of core thermal power
very similar to the case of a scram in a critical reactor. The beam trips would be much more frequent
than reactor scrams. The main source of beam trips occurring at the existing accelerator facilities is
failure of RF system. Majority of the beam trips is of short duration, say within few minutes, and the
beam recovers from them automatically. For this reason, the ADS has to maintain full flow
performance until a recovery of beam power.
Figure 1. Concept of Pb-Bi cooled ADS for transmuting MA and iodine
Proton Beam
Steam
Generator Main Pump
Primary
Vessel
11 m
5 m
Core
φ 6.4 m
Core Support Structure
0 m
φ 9.9 m
The analysis dealt with one single loop, under the assumption of the equal operating condition
among all the four symmetric loops. The lead-bismuth primary system and the secondary water/steam
system were modeled with a simple one-dimensional flow network.
535

Table 1. Core performance of MA and I transmutation ADS
with proton beam power of 30 MW (1.5 GeV and 20 mA)
Core
Thermal power
K eff
Core height/diameter (cm)
Fuel compositions
Initial heavy metal loading (kg)
Transmutation of MA (kg/300 days) 250
Blanket
Thickness (cm)
NaI/(NaI+ZrH) (%)
Initial iodine loading (kg)
Transmutation of I (kg/300 days)
800 MW
0.95
100/115
60%MA + 40%Pu
4 000
Axial
25
50
765
35
Radial
16
40
839
22
An analysis of beam trip transient was made for the accelerator-transmutation plant [5]. Transient
of the primary coolant temperature, the water/stream temperature, the water/steam pressure, the
turbine flow rate were required at a time of 380 s after beam trip to prevent from overcooling. The
maximum temperature swing was 185°C in lead-bismuth, and 82°C for in water/steam for the case
when beam recovered at a time of 370 s. The temperature change during beam trip transient is shown
in Figure 2.
3. Scenario studies of ADS introduction in future fuel cycle
The effect of ADS on MA transmutation: transmutation performance by ADS has been studied
by assuming the existence of MA of 100 tons. As the MA transmuter, we selected the ADS system
with 800 MW thermal power with a 1.5 GeV and 20 mA proton beam, and with the core having an
effective neutron multiplication factor of 0.95. With the assumption of a load factor of 80%, the net
MA transmutation rate becomes approximately 250 kg/y. The system inventories of MAs in ADS fuel
cycle will be reduced from around 80 tons in 2050 to 20 tons in 2100, and will be minimised in 2150.
MA of 100 tons , that is, will be transmuted during 120 years.
Dependence of MA and iodine build-up on fuels, burn-up and cooling times: the amount of the
MAs produced in LWRs with UO 2
/ MOX fuels and an FBR with MOX fuel have been investigated
for various conditions such as fuel compositions, burn-up and cooling time duration as shown in
Table 2. For the UO 2
-LWRs, MA build-up increases slightly with increasing burn-up. The MA buildup
quantities for full MOX fuelled LWR are more than three times for the UO 2
fuelled LWRs. In
addition, MOX-FBR produces about two times MA comparative with UO 2
-LWR.
536

Figure 2. Temperature change in primary lead-bismuth system during beam trip transient
Pb-Bi Temperature (°C)
Core Inlet
Core Outlet
SG Inlet
SG Outlet
Time t (s)
Table 2. Comparison of build-up for MA and iodine in various reactors
Reactor PWR PWR PWR PWR FBR
Fuel UO 2
UO 2
MOX MOX MOX
Burn-up(GWd/t) 33 60 33 60 140
Cooling(y) 5 5 5 5 3
MA build-up(kg/y/plant) 22.2 26.3 91.8 88.1 39.1
Iodine build-up(kg/y/plant) 5.7 5.5 7.5 7.3 8.6
The effect of ADS introduction: An estimation of the nuclear plant capacities introduced in
future will be depend on a requirement from CO 2
reduction scenario decided in the COP3 conference
at Kyoto and limitation of uranium resources. Here, based on these conditions in Japan, the nuclear
plant capacities are decided as maximum 140 GWe as shown in Figure 3. The mass flows of MA and
iodine are investigated in a future symbiosis system for transmutation consisting of UO 2
/MOX-LWRs,
FBRs and ADSs. The accumulations of MA and iodine become about 150 and 20 tons in 2040 and
900 and 200 tons in 2200, respectively, as shown in Figure 4. When the ADS will be introduced from
2040, the system inventories of MA with Pu and iodine in the ADS cycle will become about 250 and
50 tons during a time period of 2200. The mass for Pu, MA and iodine are balanced by introducing
22 units of ADS until 2160.
537

Figure 3. Nuclear electricity capacity composed of
UO 2
/MOX-LWRs and FBRs introduced based on CO 2
reduction scenario in Japan
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3:502; %:502;
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0$ ,RGLQH
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0$RULRGLQHDPRXQWWRQ
Figure 4. Comparison of MA and iodine accumulations
and transmutation by introducing ADS from 2040
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538

4. Lead-bismuth technology
Lead-bismuth is the first candidate material for liquid metal target and coolant of fuelled blanket
system in ADS researched at JAERI. Advantages of the lead-bismuth utilisation are non-active
material, very low capture cross-section, low melting point of 125°C and high boiling point of
1 670°C, and beside coolant void reactivity become negative. But potential problems are due to the
high corrosivity to most of the structural materials. At this moment the corrosive data are scarcity.
Solution is how to control material performance in corrosive environment. Here, corrosivity, reaction
with water, thermal-hydraulics etc. are studied by investigating some facilities utilised and researched
really for lead or lead-bismuth. And, furthermore, polonium evaporation rate and bismuth resource are
investigated. Main results obtained are as follow [4]:
• In a refinery of Japan, there are enough employment experience for liquid Pb-Bi in period of
about 17 years and not corrosion for the thermal conductive materials (1Cr-0.5Mo steel) used
under the condition of natural convection with temperature around 400°C.
• In Russia, extensive experience in the use as Russian submarines and in R&D during about
50 years are available. And as a result, it will be able to lead approximately zero corrosion for
Cr-Si materials by adjusting oxygen film with oxygen concentration control between 10 -7 to
10 -5 % mass. However, the corrosion data are not enough systematically collected involving
them in radiation dose field.
• In liquid-dropping experiment, it is shown that interaction between water and high
temperature liquid Pb-Bi is reduced steeply with rising of atmosphere pressure. But, in order
to design the second circuit removal model of ADS, the interaction should be evaluated by
water continuous injection experiment.
• Polonium forms PbPo in Pb-Bi, and the evaporation rate become less three factor than that of
Po, and furthermore, the rate decreases in the atmosphere. The effects of Po on employee and
environment will not be dominant in comparison with those of fission products.
• In Bi-resource, a confirmed amount will be 260 000 tonnes and an estimated amount will
become ten times of the confirmed ones by including resources in Russia. This shows there
are enough amounts for ADS developments.
5. Experimental facilities for the ADS technology demonstration
There are several technical challenges unique to the accelerator-driven transmutation system. The
major areas of technology to be tested and demonstrated are sub-critical reactor physics, system
operation and control, transmutation, thermal-hydraulics, and material irradiation [6,7,8].
Reactor Physics Experimental Facility “PEF”: Table 3 shows the purposes and items of
experiments at the PEF. Data based on the FCA experiments using a 252 Cf neutron source and a
14 MeV D-T pulsed neutron source will be used in designing the PEF facility and planning the
experimental program.
539

Table 3. Experimental items at PEF
Purpose
Neutronics of fast sub-critical systems
driven by a spallation source
Demonstration of controllability of a
hybrid system
Validation of transmutation rate of MA
and long-lived FP (LLFP)
Experimental items
• Power distribution in deep sub-critical systems
• Effective source strength and multiplication factor
• Effect of high energy particles
• Feedback control by beam adjustment
• System behaviour for beam trip and restart
• Energy gain
• MA fission rate
• LLFP reaction rate in moderated region
• Reactivity worth of MA and LLFP samples
A typical sub-critical core configuration at the PEF is shown in Figure 4. The structure of the PEF
is based on that of the FCA facility with flexible structure to carry out various experiments. The
effective multiplication factor is in the range of 0.90-0.98. The maximum proton beam power and the
core thermal power are limited to 10 W and 500 W, respectively, due to the heat removal limitations
by forced circulation of air. In Table 4, the proton beam and core power specifications are shown for
Step 1 and Step 2. Experiments at the critical state will be also carried out as the step for the precise
measurement of reactivity and neutron multiplication factor.
In Step 1, reactor physics experiments of the sub-critical core and demonstration of the principle
of ADS are to be performed, and it may be the first demonstration of the ADS concept in the world.
Table 4. Beam and power specification for sub-critical experiments at PEF
Step 1
Step 2
Proton beam
0.6 GeV, ~16nA,
10 W, 25Hz pulse
1.0 GeV, ~10nA,
10 W, 50 Hz pulse
Core thermal power
~500 Wt
~500 Wt
Engineering Experimental Facility “EEF”: The reference target design assumes to have a
hemispherical beam window made of chromium-molybdenum steel cooled by flowing lead-bismuth.
One of the high priority issues is degradation of structural material in a lead-bismuth coolant at high
proton and neutron fluxes and high temperatures. Design of the beam window, in particular, represents
the highest technical challenge since it will suffer radiation damage, thermal stress, differential
pressure load and corrosion.
540

Figure 4. Schematic cross-section of the sub-critical experimental ADS
The major objective of the ADS engineering experimental facility (EEF) is to accumulate data for
the design of a spallation target system. In Step 1, the 600-MeV linac is to deliver a 0.3 mA proton
beam (200 kW) to EEF. Radiation damage data of window materials for high intensity protons and
neutrons is of critical importance in the ADS development. Step-1 experiments will mainly be devoted
to irradiation of window materials. The preliminary estimate of damage indicates that several tens of
dpa (displacement per atom) per year can be achieved with a reasonable proton current density at
200 kW beam. Other important data of material corrosion/erosion, etc., will be obtained under various
flow conditions with lead-bismuth at high temperatures.
In Step 2, the beam power to EEF will be increased to utmost 2 MW. The 2 MW target will be a
mock-up of a full-scale ADS target. The main goals are to verify the target and window design
concept, to verify the reliability and safety, and to demonstrate the structural integrity of the spallation
target and beam window.
6. Concluding remarks
It was shown that in the future scenario of nuclear plant capacities for maximum 140 GWe, which
consists of LWRs and FBRs, the introduction of ADS of MA and iodine transmutation can play a
significant role as “Transmuter” in the back-end of fuel cycle. The R&D of Pb-Bi technologies will be
necessary, though the employment experiences in a refinery, submarines etc. were useful.
Experimental programme for the ADS technology demonstration was shown.
541

REFERENCES
[1] Takano H., Akie H., Handa M. et al., A Concept of Self-completed Fuel Cycle Based on Nitride
Fuel Lead-cooled Fast Reactor, Proc. 7th International Conf. on Emerging Nuclear Energy
Systems, ICENES’93, p. 308, World Scientific Press (1993).
[2] Osugi T., Takano H., Ogawa T. et al., A Conceptual Design Study of Self-completed Fuel Cycle
System, Proc. International Conf. on Evaluation of Emerging Nuclear Cycle Systems,
Global’95, Sept. 11-14, Versailles, France, 1995, Vol. 1, p. 181 (1995).
[3] Takano H. and Osugi T., A Concept of Nitride Fuel Actinide Recycle System Based on
Pyrochemical Reprocessing, presented to Japan-Russian Seminar on Fast Breeder Reactor,
11-15 Dec. 1995, Oarai, Japan (1995).
[4] Takano H., Takizuka T. and Kitano T., Investigation of Corrosion, Water Reaction, Polonium
Evaporation and Bismuth Resource in Liquid Metal Lead-bismuth Technology, JAERI-Review
2000-014, (2000), in Japanese.
[5] Takizuka T. et al., Design Study of Lead-bismuth Cooled ADS Dedicated to Nuclear Waste
Transmutation, Susono Seminar, 2000.
[6] Mukaiyama T. et al., Review of Research and Development of Accelerator-driven System in
Japan for Transmutation of Long-lived Nuclides, to be published in Progress of Nuclear
Analyses, 2001.
[7] Joint Project Team of JAERI and KEK, Joint Project for High Intensity Proton Accelerators,
JAERI-Tech. 99-056, 1999.
[8] Ikeda Y., Kikuchi K., Oigawa H. and Sasa, T., private communication.
542

TRANSURANICS ELIMINATION IN AN
OPTIMISED PEBBLE-BED SUB-CRITICAL REACTOR
Pablo T. León 1 , José María Martínez-Val 1 , Emilio Mínguez 1 ,
José Manuel Perlado 1 , Mireia Piera 2 , David Saphier 3
1 E.T.S.I.Industriales, Universidad Politécnica de Madrid, Spain
2 E.T.S.I.Industriales, UNED, Spain
3 Soreq N.R.C., Israel
Abstract
In a nuclear energy economy the nuclear waste is a big burden to its further development and
deployment. The possibility of eliminating the long-term part of the waste presents an appealing
opportunity to the sustainability and acceptance of a better and cleaner source of energy. It is shown
that the proposed pebble-bed transmutator has suitable characteristics to transmute most of the
isotopes that contribute to the long-term radioactivity. This proposed reactor presents also inherent
safety characteristics, which is a necessary element in a new reactor design to be accepted by the
society. Throughout this paper, we will characterise the new reactor concept, and present some of the
neutronics and safety characteristics of an accelerator driven pebble-bed reactor, (ADS) for
transuranics elimination.
543

1. Introduction and background
The sustainability of nuclear energy depends on the capability to decrease the long-term
radiotoxicity of nuclear waste [1], as the present situation in which toxical waste has a lifetime of
10 5 years and more, is totally unacceptable. The main contributors to the radiotoxicity of the spent
nuclear fuel of a light water reactor (LWR.), which is the most common design in the world nuclear
park, are depicted in Figure 1.
Figure 1. Total toxicity of the spent nuclear fuel from LWR
TOTAL TOXICITY
1.00E+09
1.00E+08
EFFECTIVE COMMITED DOSE (Sv/ton)
1.00E+07
1.00E+06
1.00E+05
1.00E+04
1.00E+03
1.00E+02
1.00E+01
1.00E+00
1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
TIME(YEARS)
Activation P. Actínides Fision P Total
As can be seen the activation products have the smallest contribution, and will decay in a period
of less than 50 years. (At this time they reach a radiotoxicity similar to natural uranium). Fission
products are the main short-term (less than 100 years) contributors, they reach the radiotoxicity of the
uranium ore in a period of less than 400 years, which is an acceptable period of time that our
technology can control. As we can see the actinides have different characteristics, it takes an important
contributor in the short term, but the only contributors in the long term. It takes approximately one
million years for the radiotoxicity of MA to reach the uranium ore level, so that the main radiotoxic
heritage problem are these isotopes. To identify exactly which actinides are the most dangerous in the
nuclear waste, Figure 2 presents the LWR waste isotope composition as a function of time after
removal from the reactor.
544

Figure 2. Main actinides radiotoxicity
FIGURE2. PRINCIPLE ACTINIDES RADIOTOXICITY
1.00E+08
Pu238
Am241
Pu240
1.00E+07
Pu239
EFECTIVE COMMITTED DOSE (Sv/tU)
1.00E+06
1.00E+05
1.00E+04
1.00E+03
Np237
1.00E+02
1.00E+01
10 100 1000 10000 100000 1000000
TIME(YEARS)
PB210 PO210 RA226 TH229 TH230 PA233 U233 U234 U235
U236 U237 U238 NP237 PU238 PU239 PU240 PU241 PU242
AM241 AM242M AM242 AM243 CM242 CM244 CM245
In this figure, we can see that the Pu isotopes (in particular 238 Pu, 239 Pu, 240 Pu and 241 Pu via their
disintegration to 241 Am and this to 237 Np) are the main contributors. The disposal of the nuclear fuel
without any treatment (open cycle) in a deep storage facility (DSF) has several problems: we must
consider nuclear proliferation, the criticality during the isotopic evolution of the fuel, the heat released
by decay of these isotopes and the uncertainties associated to the enormous period of time (for
example, the geometry in the DSF might change) [2-6]. The elimination of the actinides, in particular
the Pu isotopes and the minor actinides (specially the Am and Cm isotopes) will reduce the burden of
DSF in a substantial way.
2. Pebble-bed transmutator
Recently a new concept of sub-critical nuclear reactors is being considered, namely the
accelerator driven systems (ADS) [6-12]. The principal characteristics of this kind of reactors are:
Improved safety characteristics. In a sub-critical reactor, the shut-down of the external neutron
source (something easily achievable), results in the instant shut-down of the reactor power. If we can
guarantee the sub-criticality of the reactor under any situation, the risk of reactivity accidents, as
happened in Chernobyl, are inherently null.
Flexibility in the reactor burn-up. Since the main goal of a transmutator is to get rid of all the fuel
(mainly fissile) isotopes, the k eff will decrease very rapidly during the burn-up. In a normal critical
reactor, this is an insurmountable problem. In a sub-critical reactor, we have the additional freedom of
the external neutron source. With a suitable recycling strategy, and a somewhat variable external
545

neutron source, the reactor can maintain the neutron flux level and power during the life cycle of the
fuel, and an almost constant k eff .
Flexibility of the isotopic composition of the fuel. In the case of a critical reactor, the necessity to
obtain a k eff equal to one during the burn-up of the fuel is a constrain to the fuel composition. In the
case of the proposed open cycle PB reactor, the isotopic composition is fixed by the spent fuel
discharged from the nuclear park in each country. The advantage of the ADS is that the external
neutron source can maintain the neutron flux at the desired level, so that the composition of the fuel is
less restricted than in a conventional critical reactor.
The Pebble-bed transmutator [23] studied in this paper is an ADS, with an external spallation
neutron source, created by interactions of high energy protons with a heavy metal, such as Pb. The fuel
surrounding the target is contained inside of 3 cm of radius graphite spheres (the pebbles), having two
regions [13-15]. The inner region contains carbide fuel kernels surrounded by a porous carbonaceous
layer, called buffer. The buffer function is to absorb the gaseous fission products. It is coated with two
layers of high density pyrolitic graphite and a layer of silicon carbide in between (TRISO), and the
external part of the pebble is made of high purity graphite. The TRISO micro-spheres have a diameter
of 0.9 mm [16-19].
In the neutronic calculations carried out in the present study, a homogeneous mixture of carbon
and fuel atoms was assumed in the fuel region. We can do this [24] because the mean free path of the
neutrons is larger than the size of the TRISO coated fuel particles, so for a neutron it is an
homogeneous zone. The external radius of the pebble is 3 cm, and is a fixed parameter in these
neutronic studies. An important characteristic of pebble-bed reactors is that it is possible to change the
neutron spectra in the reactor by varying the radius of the inner region of the pebble assuming a
constant mass of fuel. In the calculations below, 2 g per fuel sphere were assumed.
The isotopic composition of the pebbles is given in Table 1. This isotopic composition
corresponds to the actinides discharged from a LWR after an average burn-up of 30 MWd/kg-Metal.
Only isotopes with a higher contribution to the radiotoxicity of the spent fuel have been chosen. The
reactor is cooled by CO 2 , and the average temperature of the gas inside the core is 550 K, and the
nominal outlet temperature is 800 K. This value is well below the disassociation threshold of CO 2 . The
main characteristics of the PBT prototype studied are described in Table 2.
The pebble-bed high temperature gas cooled reactors (HTGCR) has been designed already in the
60s and advanced concepts in the 70s. Some analysis on fast gas-cooled critical reactors [25] pointed
out that decompression accidents (loss of coolant) would imply reactivity insertions leading to slightly
supercritical states in a very short time (shorter than the estimated scram time, and without a
significant Doppler effect in a core loaded with minor actinides). From the point of view of nuclear
safety, such a type of critical transmutator would be impossible to license. However, in the case of an
ADS, with an adequate margin of reactivity a transmutator of this type could be possible. The reactor
will have continues refuelling and discharge of pebbles. It is important to notice that there is no
intention of reprocessing the pebbles. That is, a once through transmutation scenario is imposed.
546

Table 1. Fuel composition
Isotope
Discharged mass LWR
(gr/ton U)
Isotopic
composition
236 Np 5.312E-04 4.575E-08
237 Np 6.514E+02 5.610E-02
238 Pu 2.277E+02 1.961E-02
239 Pu 5.912E+03 5.092E-01
240 Pu 2.593E+03 2.233E-01
241 Pu 6.823E+02 5.867E-02
242 Pu 5.983E+02 5.153E-02
244 Pu 4.176E-02 3.597E-06
241 Am 7.651E+02 6.590E-02
Isotope
Discharged
mass LWR
(g/t U)
Isotopic
composition
242m Am 2.452E+00 2.112E-04
243 Am 1.446E+02 1.245E-02
242 Cm 5.933E-03 5.110E-07
243 Cm 4.326E-01 3.726E-05
244 Cm 3.090E+01 2.661E-03
245 Cm 2.339E+00 2.014E-04
246 Cm 3.165E-01 2.726E-05
247 Cm 3.656E-03 3.149E-07
248 Cm 2.440E-04 2.101E-08
Table 2. Main specifications of a cylindrical prototype core
Pebble-bed sub-critical prototype data
Thermal power: 100 MW Porosity: 0.396
Fuel sphere diameter: 0.06 m Total gas volume: 0.835 m 3
Inner fuel zone diameter = 0.03 m Mean radius of the neutron source channel = 0.29 m
Number of pebbles: 11 200
Active core height = 1.2 m
Thermal power per pebble = 9 000 W Reactor core radius = 0.8 m
Pebble outer surface: 113 cm 2 Reactor cross-section area = 1.74 m 2
Total outer surface of pebbles:125 m 2 Average gas flow cross-section: 0.6975 m 2
(for an active height of 1.2 m)
Average heat flux: 8 × 10 5 W/m 2 Graphite reflector inner radius = 0.8 m
Total pebble volume: 1.255 m 3 Reactor vessel inner radius = 1.00 m
(20 cm thick graphite reflector)
Total reactor volume: 2.09 m 3 Reactor vessel outer radius = 1.07 m
3. Effective cross-sections and transmutation rates
The calculation model was made up from an infinite array of hexagonal channels. Inside each
channel the pebble-bed spheres are stacked from bottom to top, in an infinite array. Each sphere
contained an inner fuel region of variable diameter containing a mixture of graphite and TRISO coated
fuel kernels. For the calculations, the fuel zone was assumed to be homogeneous. MCNPX code was
used for all the neutronic calculations using ENDF-B/VI libraries. The average effective microscopic
cross-sections obtained are given in Table 3.
We can observe significant differences for different fuel radii. The average cross-section depends
on the neutron spectra shown in Figure 3, which depends on the mass ratio between carbon and fuel
547

atoms. For fuel region Rf = 2.5 cm, so we are going to have a high build up of 241 Pu due to the higher
value of capture microscopic cross-section of 240 Pu than the absorption of 241 Pu, with the appearance of
241 Am and 237 Np as result of the decay, which have a high radiotoxicity and are more difficult to
destroy. On the other hand, with Rf = 0.5 cm, all the cross-sections are small and it is not possible to
achieve the desired burn-up, in particular for 242 Pu. The neutron spectrum in this case is hard. The best
transmutation results are obtained with fuel regions having a radius of 1.5-2 cm and therefore the
1.5 cm was chosen as the basis for the present design. The transmutation of Pu isotopes for 1.5 cm is
shown in Figure 4.
548

Table 3. Average microscopic cross-section for an infinite array of pebbles. Case of 2 grams of fuel per pebble
Rfuel = 2.5 cm Rfuel = 2 cm Rfuel = 1.5 cm Rfuel = 1 cm Rfuel = 0.5 cm
Capture Fission Capture Fission Capture Fission Capture Fission Capture Fission
237 Np 43.4594907 0.2431753 44.6410243 0.2495093 45.4299176 0.2677779 44.0585301 0.3206016 32.5835109 0.5329437
238 Pu 23.0414352 2.0672338 26.1688234 2.1537858 31.1980560 2.2972491 38.1362471 2.5198135 40.8787764 2.7071440
239 Pu 37.0417824 64.5387731 39.4894265 70.2675055 41.269663 76.2579395 40.2015133 79.5149652 31.3709412 69.7655902
240 Pu 184.515625 0.39656019 153.332879 0.39417989 114.638993 0.39956794 76.9422796 0.43448781 40.9676545 0.6160064
241 Pu 25.2185185 75.9947917 27.6275723 83.1830142 30.7858063 92.9785454 33.8962673 103.39741 31.805035 98.4604738
242 Pu 61.6064815 0.2074265 59.0063594 0.21262737 53.442918 0.22794492 41.5126133 0.27269487 17.7652483 0.45292418
241 Am 99.0798611 0.83390046 102.948982 0.85475025 105.002587 0.87309424 100.267171 0.88392484 74.4904723 0.94933279
243 Am 91.3512731 0.28546412 89.1338323 0.28893663 84.5810905 0.30078927 73.9322169 0.33711723 44.2873282 0.49500373
549
242 Cm 6.66435185 0.25365278 6.64541172 0.27884109 6.59329833 0.324211 6.993233 0.40709087 5.11853893 0.5936618
244 Cm 31.1480903 0.87228009 30.2226087 0.86749186 28.4144073 0.86336317 24.5326305 0.87094279 13.8792891 0.95424777
Cnat 0.00025757 0 0.00028724 0 0.00033754 0 0.00043462 0 0.000654 0

Figure 3. Neutron flux spectra for different fuel radii
Figure 4. Transmutation of Plutonium isotopes for Rf = 1.5 cm
ATOMIC CONCENT.(at/cm^3)
RESIDUAL FRACTIONS
3
2
1
Pu238:+
Pu239:-
Pu240:*
Pu241:o
Pu242:--
0
0 2 4 6 8 10
FLUENCE
x 10 22
2 x 1020 Pu238:+
Pu239:-
1.5
Pu240:*
Pu241:o
1
Pu242:--
0.5
0
0 2 4 6 8 10
FLUENCE
x 10 22
Assuming that the residual Pu fraction should be 0.001, then the other Pu isotopic concentration
will be as shown in Table 4, where the fluence needed to achieve this degree of burn up is also shown.
As can be seen from the table the chosen design of 1.5 cm results in the best transmutation. There is
still more than 20% of 242 Pu left, however this Pu isotope has a low radiotoxicity.
550

Table 4. Residual fractions of the fuel isotopes when
the residual fraction of 239 Pu isotope is 0.001. Infinite array cell assumed.
Fuel radius.
238 Pu
240 Pu
241 Pu
242 Pu Fluence
Rf = 2.5 cm 0.0985 0.0006 0.0119 0.0677 9.23⋅10 22
Rf = 2 cm 0.0914 0.0009 0.0146 0.1131 8.45⋅10 22
Rf = 1.5 cm 0.0767 0.0022 0.0254 0.2277 7.67⋅10 22
Rf = 1 cm 0.0545 0.0143 0.0692 0.5361 7.16⋅10 22
Rf = 0.5 cm 0.0313 0.0862 0.1658 1.4177 7.95⋅10 22
The desired source intensity to achieve the transmutation levels in Table 4 as well as other core
parameters are shown in Table 5.
Table 5. Source intensity and proton beam requirements for a 10 MWt prototype,
as a function of the radius of the fuel region, for two values of the proton energies
Rf k eff Average flux
(n/cm 2 ⋅s)
Source (n/s)
Proton beam intensity (mA)
E = 0.45 GeV
Yield = 10
E = 1 GeV
Yield = 30
2.5 cm 0.6643 1.39E+14 4.10E+17 6.60 2.18
2 cm 0.71093 1.31E+14 3.20E+17 5.15 1.70
1.5 cm 0.76362 1.26E+14 2.43E+17 3.90 1.28
1 cm 0.79526 1.29E+14 1.98E+17 3.17 1.06
0.5 cm 0.69828 1.77E+14 3.36E+17 5.38 1.80
In this table we can see the relation of the k eff value to the source needed to obtain a specific
thermal power. The neutron population in a sub-critical reactor is given by equation:
S⋅
l
N=
1-k eff
Where “N” is the neutron population inside the reactor, “S” is the neutron source, and “l” is the
averaged neutron lifetime. The closer the k eff is to unity, the smaller is the external neutron source
needed. However safety characteristics must also be considered in choosing the level of sub-criticality.
An optimisation of the k eff value is needed for a final reactor design.
4. Problems with burn-up
An important aspect during the burn-up of the fuel is the decrease of k eff. In the fuel loaded, there
is only a small amount of fertile 240 Pu, consequently the k ∞ of the pebble decreases significantly with
burn-up, as can be seen in Figure 5. In order to maintain a constant power (and k eff) in the core, an
appropriate recycling strategy has to be developed.
551

A possible way to maintain the k eff with the burn-up is the increase of the fuel content in a pebble.
However, an increase of the mass concentration can give a lower k ∞ , as we have seen in the case of
Rf = 0.5cm. We can see this results in Figure 6, which shows the needed increase of the fuel in the
pebble to obtain higher values of k ∞ .
Figure 5. Variation on the k ∞ with burn-up
Figure 6. Variations on the k ∞ with fuel mass
per pebble
The effective microscopic cross-sections changes significantly with the mass charged per pebble.
If the mass is very high, the neutron spectra become hard, and the effective microscopic cross-sections
are low. That means that we can not vary the mass loaded per pebble with Rp = 3 cm, it must remain
at 2 g. Another possibility is to change the pebble size and increasing the fuel sphere radius to 5 cm
(with the internal fuel region 2.5 cm) and 9.26 g of fuel per pebble. A much higher k eff of the core is
obtained as shown in Figure 6. The total quantity of fuel in the core is the same as with 3 cm pebbles,
and higher big burn-up can be achieved. The possibility of employing larger fuel spheres permits more
flexibility in designing the core, the k eff and the refuelling strategy. However, one should note that
until today there is no experience with pebbles larger than 6 cm.
These results open additional options. Maintain the pebbles radius in Rp = 3 cm and try to obtain
a suitable recycling strategy to achieve a nearly constant k eff during the burn-up of the fuel, or change
the original design, to permit variations in the fuel contents per pebble, so as to increase the k eff , with a
suited recycling strategy. Further studies to optimise the pebble size and its content will be performed.
Another key point is the power distribution in the reactor during the burn-up. The pebbles with a
higher burn-up will produce less power, so the axial power distribution inside the reactor will vary
significantly from top to bottom. Another objective of the recycling strategy is to try to obtain a power
distribution as uniform as possible. In the case of ADS, we have and additional parameter design,
which is the location of the external neutron source inside the reactor, which can be used to improve
the system performance.
To achieve a high destruction rate of TRU the fuel sphere has to withstand high burn-up and
fluence. From the previous work on pebble-bed reactor at FZA (Schenk) it was concluded that
600 MWd/kg was easily endured by the pebbles. In order to understand it and to explain how such
high burn-ups are achievable, note that in a 6 cm diameter pebble there will be more than 240 g of
carbon and two grams of TRU. In terms of nuclei, it means 1.2⋅10 25 of C and 5⋅10 21 of TRU, i.e. the
552

number of C nuclei is near 2 400 times larger than the number of TRU nuclei. Taking into account that
in the neutron cycle 3 neutrons are born per each TRU nuclei eliminated by fission, including source
neutrons. Some of them escape in the moderation process, but an average of 80 elastic collisions with
carbon are suffered by a neutron. The lethargy gain per collision is 0.1577 in this case and a slowing
down from 3 MeV to 1 eV, 14.9 lethargy units must be passed. Totally there will be about
240 collisions in C for each fission. If it is taken into account that the number of C nuclei is about
2 400 times larger than the number of TRU nuclei, the number of C nuclei affected in a neutron cycle
is one tenth of the number of TRU affected nuclei.
A total estimate of collisions per C atom along a fluence can be obtained including the very
important burn-up effect. Let σ f be the average microscopic fission cross-section. The neutron fluence
to achieve a residual fraction r, is given by:
r =
Na(t)
Na(0)
=
t)
e ( −σf
⋅Φ⋅
where Na is the number of actinide (TRU) nuclei. On the other hand, the number of fission neutrons
per carbon atom is:
σ
⎛ ⎞
s
⋅ Nc
ln r
Tc = ⋅ Φ ⋅ t = σ ⋅
⎜ −
⎟
s
Nc
⎝ σf
⎠
For instance, for σ s = 4.5 b (scattering microscopic cross-section), an effective σ f = 70 b, for
r = 0.05 (95% of burnt-up TRU) 0.19 collisions per C atom are obtained, which seems to be a
moderate value that could be withstood by the graphite matrix.
5. Safety characteristics
Calculations have been performed to asses the safety characteristics of the proposed design. The
accidents studied included: water ingress, changes in reactor pressure and changes in the void fraction.
In all cases the reactor remains sub-critical, except during water ingress. In a sub-moderated reactor, as
the PBT presented in this paper, the ingress of water is accompanied by a large increase in reactivity.
The good moderation properties of water lead to higher k eff values. Consequently, in the proposed
design, water ingression must be excluded. As the fuel loaded has a small amount of fertile material,
the Doppler effect will have little importance. At the beginning of cycle, the presence of 240 Pu with a
very high capture cross-section, and a broad resonance region at high energies leads to a small
Doppler effect with a reactivity coefficient of ρ = -0.52⋅10 -5 ∆K/K. During the burn-up of the fuel this
coefficient will change.
6. Conclusions and future work
A preliminary conclusion drawn from the present analysis is that it is theoretically feasible to
eliminate a large fraction of transuranics by means of an ADS pebble-bed transmutator, without the
need of chemical reprocessing. Such reprocessing would be almost impossible to do, because of the
resistance of the graphite matrix. In fact, the objective is to make TRU-fuelled pebbles from the spent
LWR fuel, and to burn them in successive burn-up cycles by unloading and reloading the pebbles in an
appropriate mixture with fresh fuel. Further work must be done in the recycling studies and in the
optimisation of the size and TRU load of the pebbles. The objective is to reach an elimination fraction
553

higher than 99% of 239 Pu, and higher than 95% in the rest of the offending nuclei. An overall goal is to
eliminate about 95% of the TRU.
The final answer to the question of the capability of the pebble to withstand the high neutron
fluence has to be of an experimental nature, but it is important to asses that the sought value of fuel
burn-up does not result in an unbearable number of collisions per C atom. This point supports the
feasibility of massive TRU elimination by an ADS pebble-bed transmutator.
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[7] Rubbia C. et al., A Realistic Plutonium Elimination Scheme with Fast Energy Amplifiers and
Thorium-plutonium Fuel, CERN/AT/95-33(et) (1995).
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Transmutation by Adiabatic Resonance Crossing in Accelerator Driven Systems, Physics
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Systems for Transmutation or Power Production USING Thorium or Uranium Fuel Cycles,
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[11] Venneri F. et al., Accelerator Driven Transmutation of Waste (ATW) Technical Review at MIT,
LANL Report, LA-UR-98-608 (1998).
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[12] Salvatores M., Spiro M., The INCA Project: Radwaste Incineration Using Accelerators,
Nuclear Europe Worldscan, 7-8 (1997) 116.
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555

TRANSMUTATION OF NUCLEAR WASTES WITH GAS-COOLED PEBBLE-BED ADS
A. Pérez-Navarro, A. Abánades, A. Castro, M.P. Gimeno, D. Iñiguez
LAESA, Zaragoza, Spain
R. B. Pérez, J.T. Mihalczo
Oak Ridge National Laboratory, Oak Ridge, TN, USA
J.L. Muñoz-Cobo, Y. Rugama
U.P.V., Valencia, Spain
J.M. Martínez-Val, E. Mínguez, J.M. Perlado
U.P.M., Madrid, Spain
M. Piera
U.N.E.D., Madrid, Spain
Abstract
Transmutation of nuclear wastes is being explored for its application to waste management, a
fundamental issue for nuclear industry. Several concepts are under consideration, mainly fast breeder
reactors and accelerator driven systems (ADS). Inside this second category, we are analysing a
helium-cooled graphite moderated sub-critical assembly, which uses as fuel units a small amount of
transuranics diluted, in the form of TRISO coated particles, in graphite pebbles. This configuration
(PBT) allows for neutron spectra that, taking advantage of the existence of huge capture resonances in
the epithermal region, increase in a substantial factor the system transmutation efficiency.
Neutronic studies to determine transmutation performance and thermal behaviour are presented and
discussed together with an analysis of the additional studies to address before going into detailed
design activities.
557

1. Introduction
Accelerator driven transmutation is a promising method to alleviate the environmental impact
associated with the final disposal of the spent nuclear fuel from Light Water Reactors (LWR). According
to such method, nuclear cascades, initiated by spallation on heavy materials by medium energy (few
hundred MeV) protons, are used in a sub-critical assembly to transmute the unwanted wastes into less
harmful species.
Previous designs of these transmuters have been essentially derived from the Energy Amplifier
concept [1] initially intended to produce energy by using the Thorium cycle. Here, we are instead
focusing on a pure transmuter based on the use of the Adiabatic Resonance Crossing (ARC) method
and specifically oriented to the most effective elimination of nuclear wastes, especially plutonium that
is the most worrisome and abundant element in those wastes.
In this paper we are presenting the main results from the studies being carrying out by LAESA
and collaborators on a Pebble Bed Transmuter (PBT) based on the above-mentioned philosophy. PBT
is a gas-cooled sub-critical nuclear core filled with graphite-fuel pebbles and coupled to an
accelerator. A small amount of transuranics is diluted inside the graphite pebbles in form of TRISO
coated particles (a few grams of TRU in each 6-cm diameter pebble). The lethargy gain per scattering
and the small TRU concentration make the neutron slowing-down to follow the ARC scheme. In
addition, this system could take advantages of the technology already developed in the seventies for
the High Temperature Gas Reactor (HTGR) in Germany and USA
2. Resonance enhanced transmutation
Nuclear waste transmutation reactions are based on neutrons as inducing particles. Neutron
energy spectrum can be classified in three main types:
• Thermal spectrum, neutron energy in the range of 0.1-1 eV, depending on the moderator
temperature.
• Fast spectrum, with energies in the order of 1-10 MeV.
• Iso-lethargic spectrum, intermediate to the previous ones with neutron energy over the nuclei
resonance region.
ARC moderation is produced when high-energy neutrons are injected in a large, diffusive
medium with negligible absorption and large elastic cross section values, such as lead or carbon. ARC
moderation generates an iso-lethargic spectrum based on the tiny energy loss steps of the neutron in
its way down to thermal energies. Hence, the neutron crosses in its moderation process all the energy
range, having a great chance to find the cross-section resonances of the material in the medium. With
regard to the elimination of actinides, this spectrum has the advantage, compared to the thermal case,
of a bigger fission to capture ratio, while compared to the fast spectrum it is more efficient both for
fission and capture. This method was tested and proved in the TARC experiment at CERN [2].
3. The pebble-bed transmuter
PBT, helium-cooled graphite moderated sub-critical assembly using as fuel small amounts of
transuranics diluted in graphite pebbles, is a device optimised for nuclear waste transmutation, in
particular TRU burning, although fission products elimination is also envisaged. Its main parameters are
shown in Table 1. Figure 1 shows a schematic view of the system.
558

Figure 1. PBT conceptual view
Table 1. 10 MW PBT parameters
Main PBT components are:
Parameter Value Unit
Thermal power 10 MW
Beam power 3.8 MW
Criticality constant 0.75
Mean power density 4.7 W/g
Core mass 2.19 Tm
Initial TRU load 14.6 Kg
Fuel
Graphite pebbles + TRU
Proton beam energy 380 MeV
Beam current 10 mA
• The accelerator system, to provide a medium-energy high-intensity proton beam.
• The spallation target, which couples the accelerator proton beam to the nuclear system,
providing the neutron source needed to sustain the sub-critical system.
559

• The sub-critical nuclear core, whose fuel is mainly composed of the offending materials to
transmute.
Two cyclotrons in cascade compose the accelerator system proposed for the PBT. The main one
is a booster cyclotron with six separated sectors and four accelerating cavities capable to reach
energies up to 380 MeV and beam intensities in the order of 10 mA. Four RF cavities operating at a
frequency of 70.4 MHz are used in order to get sufficient turn separation at the extraction radius. A
small injector provides the necessary beam, with energy of 20 MeV, at the entrance of the booster.
This injector is a superconducting 40 MeV H +
2 cyclotron that generates the required 20 MeV protons
by stripping phenomena.
The spallation target, based on liquid lead-bismuth, has a geometrical design (Figure 2) oriented
to optimise two main features: a) broadening of the source, and b) strengthening of the separation
window. The system proposed for the PBT is conical in shape, with a length of nearly 1 meter, which
will extend the neutron generation almost all over the core axis. Another subject to be carefully
considered is the energy deposition along the target. As a result of the relatively low proton energy
(380 MeV), this deposition is very close to the solid window, increasing the difficulties in the design
of the cooling system. The ionisation losses in the structural material of the window are also relevant
in our case. This target is under study by a LAESA/CRS4 working group [3] to establish material
working conditions and structural damage, including neutronics, thermal-hydraulics and radiological
hazard.
Figure 2. PBT spallation target: conceptual view and
energy deposition calculated with FLUKA [4] code
The core is a cylinder containing the fuel pebbles. The central part is occupied by the spallation
target, which acts as neutron source. The core is filled with graphite pellets containing the nuclear
fuel. The proposed fuel pellet that is proposed for the PBT is similar to the one that has been
developed for the HTGR reactors. The fuel is confined in 3-cm-radius pebbles. The external layer of
the pebble is made of pyrolytic graphite with a thickness of 5 mm, while the inner 2.5-cm-radius
volume is filled with 1-mm-diameter TRISO micro-spheres containing the fuel material. The main
advantages for these pebble bed cores are the possibility of continuous refuelling and that pebbles are
560

very tight holders for fission fragments and produced radioactivity for temperatures up to almost
2 000ºC. The reflector is made of a 60 cm carbon wall.
4. Transmutation performance
Evaluation of the efficiency of PBT in transmuting the actinides present in the PWR waste has
been addressed by simulating the neutronic behaviour of the device with Monte Carlo techniques. We
have used in these simulations the codes LAHET [5], NJOY, MCNP-4B [6] and ORIGEN-2.1 [7].
In those simulations, the core is divided in 10 horizontal layers. We have not considered the
materials homogeneously distributed inside the core. By the contrary, our simulations respect the real
distribution of materials. Homogeneous distribution is only assumed inside the pebble (except the
coating) by using the corresponding fractions of carbon, silicon and fuel. We consider that the microparticles
are imbedded in a carbon matrix, and that all the volume not occupied by fuel or silicon
carbide is filled by carbon with a density of 1.7 g/cm 3 . The fuel is TRU-oxide, with the composition
given by the TRU from a PWR discharge after 10 years of cooling-down. This fuel would include not
only Pu, but also Np, Am and Cm. Cross sections for the most relevant isotopes (Carbon in particular)
have been processed taking into account thermal working conditions and resonance broadening.
Initially, every layer of the core is filled with the same fuel. Afterwards, we make a series of
cycles: every 99 days we reload the upper level with fresh fuel, extract the lower one, and shift one
position down the rest. In this scheme, the fuel is exposed to a total burn-up of 990 days between the
insertion up to the extraction times. The data here summarised are referred to the system under
equilibrium conditions, namely 990 days after the beginning of operation of the machine.
In the equilibrium state, the 10 MW PBT gives an average mean power density of 6 W/cm 3 in the
core, or 8 MW/cm 2 on the surface of the balls. However in the upper level that power density is more
than twice that value while in the lower one it is less than one third of it. Then, the cooling will be
enhanced by a downward flow. The power release in the core has been used for the thermal analysis
of the PBT.
Regarding transmutation performance of the system. Plutonium isotopes suffer a considerable
mass reduction (Figure 3 and Table 2). In particular, the most abundant isotopes ( 239 Pu, 240 Pu and 241 Pu)
are drastically reduced. 242 Pu and 238 Pu increase slightly their masses, but the radio-toxicity of 242 Pu is
very small. The mass of 237 Np is also considerably reduced. 241 Am is eliminated while 243 Am mass
increases. The quantity of curium also increases, but it remains much smaller than the initial mass of
actinides.
Concerning the activity of the final waste stream in the long term (Figure 4), the creation of
fission products implies an increment of the radioactivity during the first hundred years, but the
smaller actinides mass produces a reduction in the long term. The minor actinides and their
descendant ( 244 Cm, 243 Am) produce a very important part of the radioactivity. We are exploring
possibilities to reduce these final minor actinides, either by reinsertion into the system waiting for a
stabilisation of their masses (as proposed in [8]), or by using a system with multiple spallation targets.
561

Figure 3. TRU elimination vs. irradiation time in the PBT
Figure 4. Activity comparison of the radwaste stream before and after burning in the PBT
Table 2. Initial and final TRU masses after irradiation in the 10 MW PBT
Isotope Initial mass (g) Final mass (g)
237
NP 655 190
238
PU 203 346
239
PU 7 512 185
240
PU 3 481 201
241
PU 1 165 199
242
PU 713 1 484
241
AM 752 37
243
AM 134 610
242
CM 0 68
244
CM 24 522
245
CM 0 9
562

5. Cooling system
A preliminary design has been done to determine the pressure losses and working temperature at
each part of the core during steady-state operation. He and N 2
have been taken into account as coolant
candidates. The reference configuration that we have adopted after a parametric study is shown in
Table 3. The increase in the mean gas temperature when passing through the core has been set up to
250ºC, resulting in a required pumping power between ~110 (He) and ~230 kW (N 2
) at a gas working
pressure of 20 bar. In these conditions, the maximum temperature reached in the fuel pellet is 982ºC,
very well within the safety requirements for this kind of fuel.
A closed gas cycle has been considered for energy production in a PBT system. The cycle has the
usual main components: high-pressure and low-pressure turbines, recuperator, compressor and
condenser. The high-pressure turbine feeds the compressor to maintain the compression ratio between
the working values. The low-pressure turbine is coupled to a generator connected to the external grid
(or to the accelerator electrical power system). In Table 4 we summarise the general features of the
system, for which we have chosen He as more favourable coolant. The required coolant flow to carry
the 10 MW thermal power is 7.7 kg/s. The total efficiency in the cycle makes achievable to produce
almost 4 MW of electrical power, what will be enough to supply the accelerator needs in steady-state
conditions and under the established PBT design parameters. The scheme of the gas cycle and its TS
diagram is shown in Figure 5.
Table 3. Cooling system reference parameters
Parameter
Value
Coolant He N 2
Pumping Forced Forced
Inlet temperature (°C) 500 500
Average outlet temperature (°C) 750 750
Pressure (bar) 20 20
Flow area/ball (cm 2 ) 0.5 0.5
Mass flow (kg/s) 7.7 34
Average velocity (m/s) 13.4 9
Pressure drop (Pa) 16072 48899
Pumping power (kW) 112 230
Mean mass density (kg/m 3 ) 1.1 7.2
Thermal power (MW) 9.7 9.7
563

Table 4. General characteristics of the PBT gas cycle
Compression ratio 2.3
He mass flow (for 10 MWth) (kg/s) 7.7
Polytropic efficiency of the turbine 0.89
Polytropic efficiency of the compressor 0.88
Efficiency coupling HP-compressor 0.95
Recuperator efficiency 0.95
Installation efficiency 0.38
Figure 5. He gas cycle for the PBT and TS diagram
6. Future R&D
Activities in progress are oriented to complete the above-mentioned studies and close a
conceptual concept for PBT before going into design work. In particular, neutronics calculations are
addressing: a) the feasibility to burn long-life fission products (as 99 Tc); b) the possibility to improve
minor actinides elimination, and c) the optimisation of the power level in order to achieve
transmutation rates suitable for industrial application.
Stability studies include reactivity coefficients analysis and transient studies and control
methods, in special sub-criticality. Current work is in progress in two areas of research: the
determination of the coolant gas void reactivity coefficient and on the effect of temperature and
pressure transients on the system operating conditions. To measure the multiplication constant of the
Pebble system we are addressing two methods have been successfully tested in Monte Carlo
simulations performed with LAHET and MCNP-DSP [8,9]. The first one can be used with the proton
source on and allows to perform an on line determination of the CPSD (w) function between the
proton source and one neutron detector located in the system. This CPSD can be easily obtained
experimentally, and Monte Carlo simulations with one detector in the system show that this method
gives good values of the multiplication constant. Second method is based on the Mihalczo ratio of
564

spectral densities [10,11] and can be used with the source turned off and a Californium source to
excite the system.
PBT power level in these studies has been selected looking at the minimum required for a
meaningful experimental device. Next step in our studies is the optimisation of that power level from
an industrial point of view.
7. Conclusion
PBT device is devoted to nuclear waste transmutation, in particular Pu burning, although fission
products elimination is also envisaged. Its transmutation efficiency is very high for most of the Pu
isotopes in a LWR discharge. It is possible, in a single pass through PBT an almost complete
elimination of 239 Pu, 240 Pu and 241 Pu with some residual long-lived and much less radiotoxic 242 Pu and
some modest Cm and Am production. Such single pass-“once through” procedure is fast and can be
accomplished in a time which is comparable to a single fuel cycle of a standard LWR. Additional
technical and optimisation studies are in progress.
REFERENCES
[1] Rubbia C. et al., Conceptual Design of a Fast Neutron Operated High Power Energy Amplifier,
(1995) CERN/AT/95-44 (ET).
[2] Arnould H. et al., Experimental Verification of Neutron Phenomenology in Lead and
Transmutation by Adiabatic Resonance Crossing in Accelerator Driven Systems, Phys. Lett. B
458 (1999) 167-180.
[3] Abánades A., Buono S., Castro A., Maciocco A., Moreau V., Status Report of the Preliminary
Design of the Proton Target Window for a Low-power ADS Prototype, (1999) Internal report,
LAESA-UI/99-16/23-VI.
[4] A. Fassò et al., FLUKA92, Proc. of the Workshop on Simulating Accelerator Radiation
Environments, Santa Fe, 11-15 January 1993.
[5] Prael R.E., Lichtenstein H., User Guide to LCS: The LAHET Code System, 1989. Group X-6.
MS B226, Los Alamos National Laboratory.
[6] Briesmeister J.F., Editor, MCNP. A General Monte Carlo N-Particle Transport Code, 1997.
LA-12625-M, Version B.
[7] Croff A.G., A User’s Manual for the ORIGEN2 Computer Code, 1980, ORNL/TM-7175, July 1980.
[8] Rubbia C., Resonance Enhanced Nuclear Waste Transmutation, University of Pavia, May 1999.
565

[9] Rugama Y., Muñoz-Cobo J.L., Valentine T. (2000), Proceedings of the MC2000 (Monte Carlo
2000, Lisbon), in press.
[10] Muñoz-Cobo J.L., Rugama Y., Valentine T.E., Mihalczo J.T., Pérez R.B., Sub-critical Reactivity
Monitoring in Accelerator Driven Systems, submitted for publication to Annals of Nuclear Energy
(2000).
[11] Mihalczo J.T., Nuclear Science and Engineering, Vol. 53, pp. 393-414 (1974).
566

MYRRHA, A MULTI-PURPOSE ADS FOR R&D AS FIRST STEP
TOWARDS WASTE TRANSMUTATION – CURRENT STATUS OF THE PROJECT
H. Aït Abderrahim, P. Kupschus, E. Malambu,
Ph. Benoit, K. Van Tichelen, B. Arien, F. Vermeersch, P. D'hondt
SCK•CEN, Boeretang 200, Mol 2400, Belgium
E-mail: haitabde@sckcen.be or myrrha@sckcen.be
Y. Jongen, S. Ternier, D. Vandeplassche
IBA, Chemin du Cyclotron 3, Louvain-la-Neuve 1348, Belgium
E-mail: jongen@iba.be
Abstract
SCK•CEN, the Belgian Nuclear Research Centre, in partnership with IBA s.a., Ion Beam
Applications, is designing an ADS prototype, MYRRHA, and is conducting an associated R&D
programme. The project focuses primarily on research on structural materials, nuclear fuel, liquid
metals and associated aspects, on sub-critical reactor physics and subsequently on applications such as
nuclear waste transmutation, radioisotope production and safety research on sub-critical systems. The
MYRRHA system is intended to be a multipurpose R&D facility and is expected to become a new
major research infrastructure for the European partners presently involved in the ADS Demo
development. Ion Beam Applications is performing the accelerator development. Currently the
preliminary conceptual design of the MYRRHA system is under way and an intensive R&D
programme is assessing the points of greatest risk in the present design. This work will define the final
choice of characteristics of the facility. In this paper we will report on the status of the pre-design
study as of June 2000 as well as on the methods and results of the R&D programme.
Keywords: Accelerator Driven System (ADS), Partitioning & Transmutation, Irradiation Technology
567

1. Introduction
SCK•CEN, the Belgian Nuclear Research Centre, and IBA s.a., Ion Beam Applications, are
developing jointly the MYRRHA project, a multipurpose neutron source for R&D applications on the
basis of an Accelerator Driven System (ADS). This project is intended to fit into the European strategy
towards an ADS Demo facility for nuclear waste transmutation.
The R&D applications that are considered in the future MYRRHA facility can be grouped in
three blocs:
• Continuation, and extension, towards ADS of the ongoing R&D programmes at SCK•CEN in
the field of reactor materials, fuel and reactor physics research.
• Enhancement and triggering of new R&D activities such as nuclear waste transmutation,
ADS technology, liquid metal embrittlement.
• Initiation of medical applications such as proton therapy and PET production.
The present MYRRHA concept, as described below, is determined by the versatility of the
applications it would allow. Further technical and/or strategic developments of the project might
change the present concept.
The design of MYRRHA needs to satisfy a number of specifications such as:
• Achievement of the neutron flux levels required by the different applications considered in
MYRRHA:
−
−
−
Φ>0.75 MeV = 1.0 × 1 015 n/cm².s at the locations for minor actinides (MA)
transmutation.
Φ>1 MeV = 1.0 × 1 013 to 1.0 × 1 014 n/cm².s at the locations for structural material and
fuel irradiation.
Φth = 2.0 to 3.0 × 1 015 n/cm².s at locations for long-lived fission products (LLFP)
transmutation or radioisotope production.
• Sub-critical core total power: ranging between 20 and 30 MW.
• Safety: k eff ≤0.95 in all conditions, as in a fuel storage, to guarantee inherent safety.
• Operation of the fuel under safe conditions: average fuel pin linear power

hexagonal assemblies of 122 mm plate-to-plate. The central hexagon position is left free for housing
the spallation module. The core is made of 18 fuel assemblies of which 12 have a Pu content of 30%
and 6 a Pu content of 20%.
The MYRRHA design is determined by the requirement of versatility in applications and the
desire to use as much as possible existing technologies. The heat exchangers and the primary pump
unit are to be embedded in the reactor pool. The accelerator is to be installed in a confinement building
separated from the one housing the sub-critical core and the spallation module. The proton beam will
be impinging on the spallation target from the top.
Figure 1. Global view of the present design of MYRRHA
Thermal neutron Island
Proton Beam Line
Spallation Target Loop
Fast Core
2.1 Accelerator
IBA, a company that has designed the world reference cyclotron for radioisotope production and
other machines, is in charge of the design of the accelerator. The accelerator parameters presently
considered are 5 mA current at 350 MeV proton energy. The positive ion acceleration technology is
envisaged, realised by a two-stage accelerator, with a first cyclotron as injector accelerating protons up
569

to 40 to 70 MeV and a booster further accelerating them up to 350 MeV (Figure 2). This option is not
yet frozen: a trade-off of higher proton energy against current is being explored. Other designs, to go
in one step from the ion source energy injection up to the 350 MeV desired energy, or accelerating H 2
molecules with stripping at the final energy stage for beam extraction, are in the assessment phase. For
more details, see Section 3.1.
Figure 2. Lower half of the magnets, with the acceleration electrodes
of a 350 MeV MYRRHA booster cyclotron
2.2 Spallation target
The spallation target is made of liquid Pb-Bi. The Pb-Bi is pumped up to a reservoir from which it
descends, through an annular gap (∅ outer 130 mm), to the middle of the fast core. Here the flow is
directed by a nozzle into a single tube penetrating the fast core (∅ outer 80 mm). At about the position of
the nozzle a free liquid metal surface is formed, which will be in contact with the vacuum of the
proton beam guideline. No conventional window is foreseen between the Pb-Bi free surface and the
beam in order to avoid difficulties in engineering this component and to keep the energy losses at a
minimum. When the Pb-Bi has left the fast core region, it is cooled and pumped back to the reservoir.
The MYRRHA windowless spallation module is given special attention in the present pre-design
phase because of its particular features, as illustrated in [2] and summarised in Section 3.2.
2.3 Sub-critical system
The design of the sub-critical assembly will have to yield the neutronic performances and provide the
irradiation volumes required for the considered applications. In order to meet the goals of material
studies, fuel behaviour studies, radioisotope production, transmutation of MA and LLFP, the subcritical
core of MYRRHA must include two spectral zones: a fast neutron spectrum zone and a thermal
spectrum one.
2.3.1 Fast zone description
The fast core will be placed centrally in a liquid Pb-Bi or Pb pool, leaving a central hexagonal
assembly empty for housing the spallation target. It consists of hexagonal assemblies of MOX FR-type
fuel pins with a Pu-content, Pu/(Pu+U), ranging from 20% to 30%, arranged in a triangular lattice with
a pitch of 10 mm. The fuel pins have an active fuel length of 50 cm (but could be increased to 60 cm
570

to achieve the requested performances) and their cladding consists of 9% Cr martensitic steel. The fuel
pins are arranged in typical FR fuel hexagonal assemblies with an assembly dimension of 122 mm
plate-to-plate. The fast zone is made of 2 concentric crowns, the first one consisting of 6 highly
enriched fuel assemblies (with 30% Pu content) and the second one of 12 fuel assemblies of which 6
are 30% enriched and 6 are 20% enriched.
Neutronic calculations coupling the high energy transport code HETC and the lower energy
neutron transport deterministic code DORT have been carried out for simulating typical configurations
of the fast core and led to encouraging results showing that the targeted performances could be
achieved. Table 1 illustrates the preliminary results we obtained for a particular configuration with an
active length of 50 cm but in which the fuel assemblies were not well simulated [3].
Table 1. Achievable performances in the MYRRHA sub-critical core
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2.3.2 Thermal zone description
The initial design, with a water pool surrounding the fast core zone and housing the thermal
neutron core zone, has been completely changed for evident safety reasons (water penetration into the
fast zone). In the present approach the thermal zone will be kept at the fast core periphery, but it will
consist of various In-Pile Sections (IPS) to be inserted in the Pb-Bi liquid metal pool from the top of
the reactor cover. Each IPS will contain a solid matrix made of moderating material (Be, C, 11 B 4 C) on
which a total leakage flux of 1 to 3 10 15 n/cm².s will impinge. Local boosters made of fissile materials
can be considered depending on the particular performance needed in the thermal neutron IPS. Black
absorbers settled around the IPS could ensure the neutronic de-coupling of the thermal islands from
the fast core.
In addition to the spallation target, the fast core and the thermal islands, the pool will contain
other components of a classical reactor such as heat exchangers, circulation pumps, fuel loading and
handling machines, and emergency-cooling provisions.
571

2.4 Confinement building
Parallel to the core and the spallation module design, attention is given to the confinement
building where the MYRRHA sub-critical reactor including the spallation module will be located. The
accelerator will be kept in a separate confinement building to facilitate the maintenance and inspection
procedures.
For the sub-critical reactor building, three options are being assessed:
• Re-using an existing confinement building where the operators are not allowed to enter
during the operation of the system, as illustrated in Figure 3.
• Re-using an existing confinement building where the operators are allowed to enter during
the operation of the system, which means that the dose exposure is less than 10 µSv/h. A
preliminary assessment showed that lateral shielding of 1 m steel followed by 2 m heavy
concrete would be necessary for achieving such a radiation level due to the very high neutron
leakage. These preliminary estimates are based on analytical estimates as well as on MCNP
modelling [4]
• Designing a completely new building with the 2 options considered above.
Figure 3. MYRRHA in an inaccessible confinement building, during operation
572

3 MYRRHA associated R&D programme
For the period 1999-2000 the MYRRHA project team is performing a detailed conceptual design
and is completing the needed R&D effort to assess the main technical risks of this design for the
accelerator and the spallation source, the most important parts of the system, as outlined below.
3.1 Accelerator
IBA is conducting preliminary design studies on the accelerator required for MYRRHA.
The present design of the sub-critical core requires the accelerator to deliver a 350 MeV, 5 mA
proton beam. This 1.75 MW CW beam has to satisfy a number of requirements, some of which are
unique in the world of accelerators up to now. At this level of power it is compulsory to obtain an
extraction efficiency above 99.5% and a very high stability of the beam, but on top of that the ADS
application needs a reliability well above that of common accelerators, bringing down the beam trip
frequency (trips longer than a few tenths of a second) to below 1 per day. The design principles are
based on the following lines of thought:
• Statistics show that the majority of beam trips is due to electric discharges (both from static
and RF electric fields). Hence the highest reliability requires minimising the number of
electrostatic devices, which favours a single stage design.
• In order to obtain the very high extraction efficiency, two extraction principles are available:
through a septum with well-separated turns, or by stripping.
• The beams are dominated by space charge. Therefore one needs careful transverse and
longitudinal matching at injection, and avoiding of cross talk between adjacent turns (by an
enhanced turn separation) if a separated turn structure is required for the extraction
mechanism.
The space charge dominated proton beam needs a 20-mm turn separation at 350 MeV if a septum
extraction has to be implemented. This solution requires the combination of a large low-field magnet
and of very high RF acceleration voltages for realising such a large turn separation, and also an
electrostatic extraction device. In view of what precedes, this solution is not well suited for very high
reliability operation. Extraction by stripping does not need separated turns. It may be obtained by the
acceleration of H - ions, but the poor stability of these ions makes them extremely sensitive to
electromagnetic stripping (and hence beam loss) during acceleration. The use of H - would, therefore,
lead to the use of an impractically large magnetic structure. The other solution is to accelerate 2.5 mA
of HH + ions up to 700 MeV, where stripping transforms them into 2 protons of 350 MeV each, thus
dividing the magnetic rigidity by 2 and thereby allowing to extract. This solution reduces the problems
related to space charge since only half the beam current is accelerated. However, the high magnetic
rigidity of a 700 MeV HH + beam imposes a magnetic structure with a pole radius of almost 7 m,
leading to a total diameter of the cyclotron of close to 20 m. The cyclotron would consist of
4 individual magnetic sectors, each of them spanning 45 degrees.
At the present stage of R&D the last option appears to be the most appropriate one.
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3.2 Spallation source
The choice of a windowless design was influenced by the following considerations:
• At about 350 MeV, an incident proton delivers 7 MeV kinetic energy per spallation neutron.
Almost 85% of the incident energy exits the target in the form of “evaporation” energy of the
nuclei. The addition of a window would diminish the fraction of the incident energy delivered
to the spallation neutrons [5].
• A windowless design avoids vulnerable parts in the concept, increasing its reliability and
avoiding a very difficult engineering task.
• Because of the very high proton current density (>130 µA/cm²) and the low energy proton
beam we intend to use, a window in the MYRRHA spallation module would undergo severe
embrittlement.
The project team has identified the three following main risks to be assessed for this windowless
design.
3.2.1 Need for basic spallation data
Since the flux characteristics in an ADS are determined by the spallation neutron intensity and since
there is a lack of experimental spallation data in the proton energy range considered, SCK•CEN is
assessing, in collaboration with Paul Scherrer Institute (PSI-Switzerland) and Nuclear Research Centre
Soreq (NRC-Soreq, Israel), basic spallation reaction data when bombarding a thick Pb-Bi target with
protons at energies close to the values that are considered for MYRRHA (E p = 350 to 590 MeV). A joint
team from the three institutes conducts the experimental programme at the PSI proton irradiation facility
(PIF). The programme started in December 1998 and is due to finish by the end of May 2000 for the
experimental part. The analysis of the data is still going on and expected to be finalised by the end of
2000. The expected data from this programme are:
• Neutron yield or amount of spallation neutrons per incident proton (n/p yield).
• Spallation neutron energy spectrum.
• Spallation neutron angular distribution.
• Spallation products created in the Pb-Bi target.
3.2.2 Feasibility of the windowless design
The design of the windowless target is very challenging: a stable and controllable free surface
needs to be formed within the small space available in the fast core (∅ outer 140 mm). This free surface
will be bombarded with protons, giving rise to a large and concentrated heat deposition (1.75 MW)
dispersed over a 15 cm depth starting from the surface for a proton energy of 350 MeV. This heat
needs to be removed to avoid overheating and possible evaporation of the liquid metal.
To gain confidence and expertise in the possibility of creating a stable free surface, SCK•CEN is
conducting an R&D program in collaboration with the thermal-hydraulics department of the
Université Catholique of Louvain-la-Neuve (UCL, Belgium). Within this R&D program, water
experiments on a one-to-one scale are performed. Water is used because of its good fluid-dynamic
similarity with Pb-Bi. This programme has been complemented by velocity field measurements in
574

collaboration with Forschungszentrum Rossendorf (FZR, Germany) using ultrasonic velocity profile
and hot-wire techniques. Currently, the design of the spallation target is being fine-tuned and adapted
to the latest geometrical constraints imposed by the neutronics of the fast core.
A confirmation experimental programme making use of Hg as a fluid is in progress at the
Institute of Physics at the University of Latvia (IPUL) at Riga. As a final confirmation, we will run
experiments with the real fluid at the actual temperatures in collaboration with Forschungszentrum
Karlsruhe (FZK, Germany) where the MYRRHA spallation target head will be inserted in their
KALLA Pb-Bi-loop, which has a working temperature of about 250°C.
In parallel with the experiments, numerical simulations using Computational Fluid Dynamics
codes are performed, aimed both at reproducing the existing experimental results and giving input for
the optimisation of the head geometry in the experiments. The CFD calculations will also be used to
investigate the flow pattern and temperature profile in the presence of the proton beam, which cannot
be simulated experimentally at this stage. At SCK•CEN the CFD modelling is performed with the
FLOW-3D code which is specialised for free surface and low Prandtl number flow. This effort is
being backed up at Université Catholique of Louvain-la-Neuve (UCL, Belgium) using the Fluent code.
Moreover, a collaboration agreement with Nuclear Research Group - Petten (NRG, The Netherlands)
is set up for more CFD calculations with the Star-CD code. Details on this R&D associated
programme can be found in Reference [2].
3.2.3 Compatibility of the windowless free surface with the proton beam line vacuum
As the free surface of the liquid metal spallation source will be in contact with the vacuum of the
proton beam line, SCK•CEN is concerned about the quantitative assessment of emanations from the
liquid metal. These can lead to the release of volatile spallation products, Pb and Bi vapours and of Po,
which will be formed by activation of Bi. These radioactive and heavy metal vapours can contaminate
the proton beam line and finally the accelerator, making the maintenance of the machine very difficult
or at least very demanding in terms of manpower exposure.
In order to assess the feasibility of the coupling between the liquid metal of the target and the
vacuum of the beam line and to assess the types and quantities of emanations, SCK•CEN is preparing
the VICE experiment (Vacuum-Interface Compatibility Experiment), studying the coupling of a
vacuum stainless steel vessel containing 130 kg Pb-Bi, heated up to 500°C, with a vacuum tube
(10 -4 ~ 10 -6 Torr) simulating the proton beam line. A mass spectrometer will measure the initial and
final out-gassing of light gasses and the metal vapour migration. To protect the vessel from liquid
metal corrosion, the possibility of Mo and W coating is currently being investigated. The full
experiment will be commissioned during the third quarter of 2000. First results are expected in early
2001.
4. MYRRHA international collaborations
The MYRRHA project, in its design phase, during its construction and also in its future
operational stage, is an international collaboration project. Agreements have already been signed and
collaborations are in progress with:
• Nuclear Research Centre (NRC, Israel): basic spallation data.
• Paul Scherrer Institute (PSI, Switzerland): basic spallation data, MEGAPIE.
575

• Ente per le Nuove tecnologie, l’Energia e l’Ambiente (ENEA, Italy): spallation source thermal
hydraulics, core dynamics.
• Université Catholique of Louvain-la-Neuve (UCL, Belgium): spallation source design.
• Ion Beam Applications (IBA, Belgium): cyclotron design and construction.
• Forschungzentrum Rossendorf (FZR, Germany): instrumentation for the spallation target.
• Forschungzentrum Karlsruhe (FZK, Germany): spallation source testing with Pb-Bi.
• Nuclear Research Group (NRG, The Netherlands): CFD modelling and system safety
assessment.
• Commissariat à l'Énergie Atomique (CEA, France): sub-critical core design, MUSE
experiments, system studies and window design for the spallation target.
• Institute of Physics of University of Latvia, Riga (IPUL, Latvia): spallation source testing
with Hg.
Contacts that may lead to additional collaborations exist with:
• RIT, Sweden : participation in MYRRHA.
• International Science and Technology Centre (ISTC), Contact Expert Group of the Project
559, IPPE Obninsk, Russia : PbBi target design.
• LANL et al., USA: Accelerator Transmutation of Waste (ATW) project.
• AEKI, Hungarian Nuclear Energy Institute: modelling of the spallation source.
• Belgonucléaire (Belgium): fuel and core design, fuel loading policy and fuel procurement.
• Tractebel Energy Engineering (Belgium): confinement building and auxiliary systems.
5. Funding sources for the MYRRHA project
An accurate evaluation of the needed investment to build MYRRHA and an analysis of the
potential sources of funding is expected to be completed by the end of the pre-design phase end 2000-
mid-2001.
Since the MYRRHA project is likely to be attractive for several types of scientific and industrial
groups at the regional, national and international level, SCK•CEN and IBA will explore funding
possibilities such as:
• Nuclear waste management agencies and producers of long-lived radioactive waste (at the
Belgian level: NIRAS/ONDRAF and the electric utilities).
• Governmental authorities at the regional, national and international level in charge of
scientific policy, energy policy or industrial development. Since MYRRHA is proposed as a
first technical step in the development of a large-scale demonstration model for the
transmutation of radioactive waste in Europe, proposals for support of MYRRHA will be
submitted to the European Union or to specific member states.
576

• Industrial partners, in particular engineering companies challenged by innovative
technologies, for which participation in the development of MYRRHA can be an important
reference. These industrial opportunities may also attract public and private venture capital.
6. Conclusions
Accelerator Driven Systems can become an essential and very viable solution to the major
remaining problems of nuclear energy production. The MYRRHA system would provide the
indispensable first ADS step towards a European ATW installation without forcing to freeze all
options of ADS (liquid Pb-Bi versus gas, pool versus loop, sub-criticality level, mitigating tools for
reactivity effects, etc.).
MYRRHA is an innovative project that will trigger different research and industrial activities in
fields such as accelerator reliability, nuclear waste management (transmutation), development of new
materials, environmental medicine, structural material corrosion and embrittlement, and safety of
nuclear installations. Increasing knowledge and know-how in these fields will contribute to aspects of
sustainable development and offer a potential for industrially applicable spin-offs.
REFERENCES
[1] H. Aït Abderrahim, P. Kupschus, Y. Jongen, S. Ternier, SCK•CEN Report, BLG 841, (2000).
[2] K. Van Tichelen, P. Kupschus, H. Aït Abderrahim, J.M. Seynhaeve, G. Winckelmans,
H. Jeanmart, Proceedings of ICENES-2000, pp.130-142 , Petten, The Netherlands, September
25-28, 2000, (2000).
[3] E. Malambu, SCK•CEN Report R-3438 (2000).
[4] M. Coeck, Th. Aoust, F. Vermeersch, H. Aït Abderrahim, to be published in the Proceedings of
MC2000 Conference – Advanced Monte Carlo for Radiation Physics, Particle Transport
Simulation and Applications, Lisbon, Portugal, October 23-26, 2000, (2000).
[5] W. Wacquier, Master Degree Dissertation, Universiteit Gent/Katholieke Universiteit Leuven,
Belgium (1997).
577

ADS: STATUS OF THE STUDIES PERFORMED BY THE EUROPEAN INDUSTRY
Bernard Carluec
Framatome ANP Direction Novatome
10, rue Juliette-Récamier, 69456 Lyon CEDEX 06, France
Fax: +33 4 72 74 73 30
E-mail: bcarluec@framatome.fr
Luciano Cinotti
Ansaldo Nuclear Division
C.so Perrone, 25, 16161 Genova, Italy
Fax: +39 10 655 8400
E-mail: cinotti@ansaldo.it
579

1. Introduction
The transmutation of most of the long-lived nuclear waste is a promising solution, which could
play a substantial role in the safety of the fuel cycle. The maximisation of the transmutation of minor
actinides is obtained with a fast neutron spectrum. Due to the neutronic characteristics, a core
dedicated to the fission of the minor actinides would have to be operated in a sub-critical state and
controlled by an external neutron source. The accelerator driven systems (ADS) allow this request.
The feasibility of transmutation on an industrial scale, in the ADS has to be evaluated.
On 1998, the Research Ministers of France, Italy and Spain have established a Technical Working
Group (TWG) including R&D organisations and industrial companies in charge of reactor and
accelerator studies, in order to identify the crucial technical issues for which R&D is needed. The
recommendations of the TWG indicate the needs to design and operate an eXperimental ADS (XADS)
facility at a sufficiently large scale to become the precursor of the industrial, practical-scale
transmuter.
Based on the recommendations of the TWG, the industrial companies, grouped in a European
Industrial Partnership, have proposed preliminary concepts of the XADS.
The candidate cooling media of a fast neutron sub-critical core are liquid metals and gas. Sodium
is well known and the most validated among the cooling media for fast neutron reactors. The concepts
using both lead or lead-bismuth eutectic (LBE), and gas need significant R&D activities. Preliminary
design studies of these concepts will be performed in the frame of the Fifth European Framework
Programme. These studies will be performed on a common basis in order to be capable to make a
consistent comparison of the concepts.
The purpose of the paper is to present the status of the design studies of the LBE- and gas-cooled
concepts proposed respectively by Ansaldo and Framatome.
2. The need for an experimental facility
The need for an experimental facility has been clearly stated in the “Interim Report of the
Technical Working Group on accelerator driven sub-critical systems” issued on October 12, 1998;
from this document the following basic guidelines have been extracted.
The XADS programme should be of a sufficiently large breadth to permit to explore and
eventually master most of the critical issues associated with the technology of the ADS concept.
However, the XADS is not yet the prototype of the industrial device, although most of the problems of
the latter should be explored separately and solved in realistic conditions by the XADS. The
realisation of a XADS is deemed inevitable to make real progress in the field. In addition, the practical
realisation of a unique European XADS constitutes the fastest and most cost effective way to
conclusively assess the potentialities and the feasibility of a full-scale industrial programme based on
ADS.
One of the main parameters of XADS is the maximum produced power of the sub-critical core.
The optimum value is a compromise between the minimum value requested to validate the
technological options (components, procedure, performance) at an industrial level; the minimum value
for performing experimental irradiation; and the investment and operation costs. A preliminary value
has already been set tentatively by the TWG to be of the order of 100 MW.
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There is no specific need at the first XADS stage to make use of the heat produced that can be
dissipated to the environment. A relatively low operating temperature can be used for the fuel and in
the successive heat extraction process.
An important part of the experimental programme is the investigation of the transmutation
capabilities of the XADS for minor actinides and long lived fission fragments, in particular 99 Tc and
129 I. An appropriate area at the periphery of the core should be planned and operated in parallel with
the rest of the facility.
The mission of the XADS can be summarised as follows:
• Operability of the accelerator/spallation target/sub-critical assembly complex in realistic
conditions and with a sufficient power (100 MW). The initial fuel could be an existing U-Pu
oxide, like for instance the second, fresh core of Superphénix or an equivalent high
enrichment fuel of commercial availability.
• Innovative fuels with a high minor actinide enrichment and fuel cycle qualification and
operation.
• Demonstration of the capability to transmute various actinides fuels.
• Assessment of the capacity to transmute long-lived fission products at an industrial level.
3. XADS cooled by the Pb-Bi eutectic
Since early 1998, the Italian ENEA, INFN, CRS4 and Ansaldo have set up a team, led by
Ansaldo, to design an 80 MWth XADS, a key-step towards the assessment of the feasibility and
operability of an ADS prototype. The results obtained so far, though preliminary and not exhaustive,
allow outlining a consistent XADS configuration. The main issues investigated and the associated
solutions proposed (see Table 1 and Figure 1, in fine) are concisely described here below.
3.1 The accelerator drive
The process of selection of the accelerator type, a cyclotron or a linac or a combination of both all
types presenting advantages and drawbacks is continuing at present. The results of investigations
carried out so far support the confidence that the required facility can be obtained scaling up by a
factor of 2 to 4 the power of existing facilities, such as the cyclotron installed at the PSI or the linac
installed at Los Alamos. In particular the cyclotron-based facility has been analysed in detail and the
modifications identified.
The INFN-LNS of Catania has carried out a scoping study aimed at identifying a compact
accelerator system based on an upgraded design of currently operating facilities. This work has
screened out a solution based on cyclotrons arranged in series and capable to supply a proton beam
power in excess of 3 MW.
The main issues to be solved are stability and reliability of the present accelerator technology.
In an ADS, the ideal proton beam to be supplied to the sub-critical core, must be reliable and
stable in time as required by the nuclear safety and investment protection.
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Existing accelerators have been designed for purposes other than for use in ADS and it is
apparent that they would not be suitable for operation in an ADS.
Considering that recorded unscheduled shutdowns of modern LWR’s are only a few per reactor
per year, it appears necessary that the respective designers reach a compromise. More precisely, new
accelerator-driven reactors should be designed to tolerate much more transients and accelerators
should be substantially improved to become more stable and reliable.
3.2 The target
Target eutectic is kept separate from the reactor primary coolant by means of a retrievable target
unit. Two target configuration concepts have been investigated, which differ in the separation barrier
adopted at the interface between vacuum pipe and target lead-bismuth eutectic.
The “window” target configuration features a mechanical barrier of a material transparent to the
largest possible extent to neutron and proton irradiation and engineered to withstand pressure and
thermal loads, the eutectic circulates under natural circulation, cooled in the upper part of the target
unit by the diathermic fluid of an auxiliary system.
In the “windowless” target configuration, the proton beam from the accelerator impinges directly
on the target eutectic, that circulates driven by a stream of cover gas, according to the same gas lifting
principle adopted for the primary system and is cooled by the reactor coolant in the heat exchanger
located in the bottom part of the target unit.
3.3 The core
The basic fuel Sub-Assembly (SA) is a boxed hexagonal cluster of ninety highly enriched (about
20% Pu) Superphénix-like MOX fuel pins. The pin diameter is the same as in Superphénix, whereas
the active length is slightly shorter (87 vs 100 cm) and the ratio of pin pitch to diameter is larger than
in Superphénix.
The SA’s are arranged in an annular array of four rows surrounding the target cavity. Because six
additional SA's couples have been added at the periphery of the core, the number of SA’s amounts to
120. The core multiplication factor results 0.97 at beginning of life and reduces to 0.94 at end of cycle,
at full power.
k eff = 0,97 is sufficiently low to ensure the safe operation of the reactor without control and
shutdown rods; twelve absorber radially positioned by means of the refuelling machine, operating
without target unit displacement, can bring the k eff below 0.95 at refuelling conditions (200°C, zero
power, target unit vertically displaced).
The core is surrounded by an outer region of four rows of dummy assemblies, which are empty
duct structures. This offers a continuous fast-to-thermal neutron flux region, useable for burning tests
of minor actinides and long life fission fragments SA’s.
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3.4 The primary coolant and the reactor configuration
The eutectic lead-bismuth has been chosen instead of lead for the XADS, because, while
behaving neutronically like lead, it allows a lower operating temperature of the reactor and its
chemistry, in particular for the corrosion protection of the structural steels, can benefit from the
Russian experience on reactors for the submarine propulsion. The preliminary tests made by ENEA at
the Brasimone facility confirm the compatibility of known steels with LBE at the envisaged XADS
operating temperature.
The major drawback of this eutectic is the formation of polonium from bismuth, in addition to the
relative scarcity (and therefore high price) of this element. The pool-type, instead of the loop-type
configuration, has been chosen for the reactor, because of the possibility to contain within the main
vessel all the primary coolant with and of the large experience acquired with the design and operation
of sodium-cooled, pool-type reactors. The loop-type configuration would additionally suffer of a
major disadvantage, because molten lead should be pumped at low speed in order to limit
corrosion/erosion and this would lead to larger diameter piping for a given volumetric flowrate, with
high linear specific weight and difficulties and cost associated with the design of the seismic supports.
The design experience of sodium-cooled, pool-type reactors has been used extensively for the
case of in-vessel and ex-vessel fuel handling machines and the rotating plug in the reactor roof.
Whenever this experience did not appear applicable to the specific tasks, however, solutions have
been proposed, that, though being innovative in the nuclear field, are not new to the industrial practice.
It is the case of the primary coolant circulation and of the choice of the secondary coolant.
The combination of more permeable fuel elements and lower average specific power of the core,
can reduce the primary coolant pressure loss to few tenths of a bar, i.e. about one tenth of the pressure
loss of the primary circuit of a sodium-cooled fast reactor.
Natural circulation allows a simple primary system configuration. This is apparent by comparison
with the bulky pumping system of the primary coolant of a sodium-cooled fast reactor, that consists of
primary pumps, designed to deliver some bar of pressure, of a pressure-plenum upstream of the core,
and of interconnecting piping. The primary sodium is fed at high speed in order to reduce the piping
diameter. Besides the fact that high speed in case of lead as a primary coolant should be avoided,
owing to the associated erosion of the structures, some of the space made available on the reactor roof
by the superfluous mechanical pumping system has been conveniently used to accommodate the
proton-beam pipe and auxiliaries system, that is peculiar to an ADS reactor.
Natural circulation of the primary coolant presents, however, some drawbacks or design
constraints as follows:
• The reactor vessel height must be increased by the head required to drive the natural
circulation.
• The requirements of low-pressure loss through the core and the heat exchanger.
• The reduced controllability of the primary coolant flowrate, that would inherently limit the
range of operating conditions of the reactor itself. This is a particularly important drawback,
because a test campaign is essential part of the scope of the XADS.
As a conclusion, both the simple primary system configuration typical of the coolant circulation
by natural convection and the merits of forced circulation are appealing to the designer. This fact have
583

suggested to look for an innovative primary coolant circulation concept capable of featuring the basic
advantages of both natural and forced circulation outlined above, while keeping the capability of full
decay heat removal by natural circulation.
In the configuration of the XADS being designed by the Italian team (Figure 1), the lead-bismuth
eutectic circulation is enhanced by a flow of about 100 Nl/s cover gas, injected into the bottom part of
the twenty four, 0.2 m ID, identical pipes arranged in circle, which make up the riser. The natural
draught alone provides the circulation needed for the safety-grade decay heat removal, as first step in
the heat transfer route towards the reactor vessel air cooling system. This proposal [1] would combine
the uncompromising reliability required by the safety function, ensured by the natural circulation, with
the advantages of reactor compactness and operational flexibility characteristic of the forced
circulation.
A simulation of the gas lifting process with air and water has been carried out early 1998 by
means of a real size test rig installed in the Ansaldo’s own test facility in Genova. The test results have
been used for the preliminary estimate of the gas flowrate capable to generate the required pressure
differential between riser and downcomer in isothermal flow.
Looking at Figure 1, it will be noted that the XADS-Intermediate Heat Exchanger (IHX) is hung
at the reactor roof as in conventional pool-type reactors, but has been installed free in the downcomer,
i.e. without the usual physical separation between hot plenum and cold plenum, an inner structure
(“redan” in French), that “forces” the coolant to flow through the IHX. With this XADS layout, the
coolant flowrate route in the downcomer is substantially determined by the natural convection taking
place within the IHX, that “forces” the coolant to flow through the IHX, while keeping the coolant
quasi stagnating outside the IHX. In fact, the interface between hot plenum and cold plenum, that
forms outside the IHX, may gently move in steady-state operation along its current shell, depending
on occasional fluctuations of the IHX power level. This engineering choice gives the opportunity to
illustrate another key-aspect of the design approach of the Italian team, i.e. simplification to the largest
possible extent of equipment and fixed internals immersed in the lead-bismuth eutectic, in order to
minimise the risks of their failure and the requirements of in-service inspection, that is difficult in
liquid metals. The elimination of the primary pumps is an example of this design approach. The
elimination of the redan structure is a second major example, with the relatively minor associated
drawback of the need of larger-diameter IHX. This layout configuration has been adopted for the
XADS, with the reserve that it shall be confirmed by the results of the thermal-mechanical analysis on
the reactor vessel/cylindrical inner vessel assembly, and also it shall be compatible with the corrosion
protection techniques that could be envisaged.
3.5 The secondary system
The secondary system is constituted by two safety-related loops, that in normal operation
dissipates the heat generated by the reactor to the atmosphere.
Each secondary loop is made up of two IHX’s arranged in parallel and of three Air-fin Heat
Exchangers (AHX) arranged in series, a circulation pump, and of the interconnecting piping. The
system, as it has been designed, could re-use the six AHX’s belonging to the RSR circuit of the PEC
reactor.
The thermal cycle temperatures, 320°C for the hot leg, and 280°C for the cold leg, are consistent
with the choice of a synthetic diathermic fluid as the coolant, owing to the low vapour pressure of
584

these fluids and the insurance of no fast chemical reactions, in case of leak, with the lead-bismuth
eutectic or the air.
Diathermic fluids do not have so good heat transfer properties as molten metals (their thermal
conductivity λ oil ≅ 0.1 W/m 2 K is quite low with respect to lead, λ Pb ≅ 15). Nevertheless, the heat
transfer coefficients achieved with these fluids in association with innovative-design IHX’s are only
slightly lower than in the case of lead. The still slightly better performance of lead is outbalanced,
however, if the comparison is extended to the whole secondary circuit. In fact, the diathermic fluid has
about 17 times as much heat capacity as lead and can be pumped at higher speed, so that the required
circulating mass of the diathermic fluid is about 50 times less than the mass of lead.
3.6 Refuelling
Handling systems are similar, however not equal to the homonymous systems designed for
sodium cooled reactors, because the SA’s are lighter than lead-bismuth eutectic and rise, unless
constrained.
The SA handled in lead-bismuth eutectic must be guided into position and locked to the rotor and
the diagrid and vice-versa.
Fuel charge on the diagrid is done by means of:
• A rotor lift combined with a flask as the link between in-vessel and secondary fuel handling.
• A fixed-arm charge machine for in-vessel fuel transfer.
The SA’s can be winched down from the flask and locked to the rotor, because forced to sink by
the guided gripper pushed down by ballasts.
The fixed-arm charge machine grabs the head of the SA by means of a cylindrical-shaped
constraint, puts the SA into position, locks its foot to the diagrid and, by the same kinematics link,
unlocks the SA head.
Secondary fuel handling equipment transfers the SA into a flask, and eventually into a cask as
intermediate storage. More precisely:
• The SA is lowered from the flask into the encapsulator and tight-sealed in a canister.
• The SA-bearing canister is placed in a storage rack immersed in a water pool.
• After sufficient decay time, the SA-bearing canister is placed in a cask and dry-stored away.
585

Table 1. Main lead-bismuth eutectic XADS data by plant area
Plant area
Plant power
Reference solution
80 MWth sub-critical system controlled by a 600 MeV,
6 mA proton beam
Target/Window Two Options: a) Proton Window
b) Windowless Target
Core
0.97 (at beginning of cycle ) < k eff < 0.94 (at end of cycle),
at full power
Fuel
U and Pu MOX
Primary system Pool configuration with four integrated IHXs
Primary coolant
circulation
Circulation enhanced by gas injection in a natural-circulation reactor
configuration
Secondary system Two low vapour pressure organic diathermic fluid loops rejecting
heat by means of air coolers
Thermal cycle
300°C at core inlet, 400°C at core outlet
Reactor roof
Metallic plate
Main vessel and safety Hung from a cold annular beam
vessel
Structural materials Vessels and internals: 316L
Target and fuel SA’s: 9Cr 1Mo
In-vessel fuel handling One rotating plug, one fixed arm, one rotor lifting machine
Secondary fuel handling Flask, encapsulator, canister, lifting and translating equipment,
water pool
Nuclear island
Common basement on anti-seismic support
Plant safety
Full passive system
4. XADS cooled by gas
The studies performed in France related to the fuel cycle, in particular in the frame of the research
group GEDEON, have demonstrated the potential of the ADS concept for the reduction of the
radiotoxicity of the nuclear wastes.
The need to develop a first experimental facility has been recognised for the demonstration on an
industrial scale of the feasibility of the ADS concept.
The main options for the XADS have been defined in 1998 by a French working group leaded by
the Ministry of Research and grouping CEA, CNRS, EDF and Framatome. The main technical options
are as follows:
• A proton beam with energy between 400 MeV and 1 GeV, impacting a heavy metal spallation
target.
• A sub-critical core in a fast neutron spectrum.
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• A solid fuel for the transmutation of the radioactive wastes.
• A maximal power for the sub-critical core lower than 200 MW thermal.
• A physical separation (“window”) between the accelerator and the spallation target.
• A physical separation between the spallation target and the reactor housing the sub-critical core.
Based on these main options, a XADS concept has been proposed by France at the European
TWG. The concept is still preliminary and studies should be performed in the frame of the Fifth
European Framework Programme to consolidate the proposed design.
Gas has been chosen as cooling medium of the sub-critical core. It had been judged that this
option should be investigated in order to propose an alternative to the liquid metal concepts using
sodium, lead or lead-bismuth eutectic. Compared to liquid metal concepts, the gas has intrinsic
advantages. Mainly:
• Much less chemical interactions, and corrosion.
• Easier in-service inspection and repair, thanks to the transparency of the gas and the
shutdown temperature which is close to the ambient temperature.
In addition, gas has been chosen because of the important experience feedback of this cooling
fluid in various nuclear plants.
Helium has been preferred due to its thermal characteristics, and because the risk of chemical
interactions, radiolyse and radioactive activation can be intrinsically excluded.
4.1 The accelerator drive
The French accelerator experts have concluded that the linac concept is the most suited to be used
in an industrial ADS transmuter. This is due to the higher capabilities of linac related to the beam
power and the beam availability which should be easier to obtain with linac than cyclotron. Therefore,
in order to get experience, it is recommended to develop the linac techniques at the stage of the
XADS. For these reasons, an oversized linac accelerator has been chosen. Moreover, the studies on the
spallation efficiency have shown that the energy of about 1 GeV allows to optimise the neutron
produced per proton and per GeV. The maximum beam intensity is fixed at 10 mA.
4.2 The target
Due to its high spallation efficiency combined with its thermal characteristics, the lead-bismuth
eutectic has preliminarily been chosen as spallation target.
The spallation target is located in the centre of the core and, in order to maintain the core
symmetry, the proton beam is introduced vertically from the top of the primary circuit. The leadbismuth
eutectic circulates in a circuit also located on the axis of the core, around the proton beam
pipe. The circulation is driven by an electromagnetic pump; the lead-bismuth eutectic is cooled by a
heat exchanger. The lead-bismuth velocity in the circuit is limited at 2 m/s.
The size of the window is defined assuming a maximum proton beam density of 30 µA/cm²
allowing to limit the window damages.
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4.3 The core
The basic fuel sub-assembly is a boxed hexagonal cluster of thirty-seven U-Pu oxide pins (Pu
enrichment lower than 35%). The pin diameter is thirteen millimetres. In order to optimise the
efficiency of the neutron source, the length of the active core is 1.5 meter.
The SA’s are arranged in an annular array surrounding the target cavity. At the periphery of the
core radioactive protection are implemented. An internal storage for fuel elements is located into the
protection. In addition, special locations are foreseen for experimental devices.
The sub-critical level of the core is 0.95. This value is determined by a preliminary safety analysis
assuming that the core has to be maintained in a sub-critical level taking into account all the credible
events capable to lead rapidly to reactivity insertion. Additionally, in order to increase the margins
during the shutdown states, particularly during handling states, and to avoid criticality, an absorber
system is introduced in the core during the shutdown states.
4.4 The reactor configuration
The primary circuit consists of two pressure vessels, the core vessel and the power extraction
vessel. The reactor vessel is shown in the Figure 2. The pressure of helium is 6 MPa. This value is
lower than the pressure used in the High Temperature Reactor (HTR) plants.
The gas circulation in the core is preliminary defined from the bottom to the top. The core inlet
temperature is 200°C, a low value but higher than the melting point of the lead-bismuth eutectic
(125°C). The core outlet temperature is 450°C, low value that allows to avoid creep damages for the
stainless steel structures.
The hot lead-bismuth eutectic temperature is the same as the core outlet temperature, 450°C. The
cold lead-bismuth temperature is 300°C, which allows a significant margin compared to the melting
point.
In order to eliminate the criticality risk due to water ingress in the core, no steam generator is
implemented. The power of the sub-critical core is extracted by a direct cycle using a turbocompressor.
The heat sink is achieved by a heat exchanger. The secondary fluid is liquid water at low
pressure. An alternator produces electricity, at least for supplying the accelerator needs.
In shutdown states, the core decay heat is removed by the shutdown reactor cooling system. Two
redundant blowers and two redundant heat exchangers are implemented in the top of the reactor
vessel. The system is fully redundant and electrically supplied. This system is also used in accidental
conditions. In case of loss of internal and external electrical supply, the system is capable to remove
the decay heat by natural circulation of the primary helium and the secondary circuit. In case of loss of
the helium pressure, the decay heat removal is achieved by the redundant blowers electrically
supplied.
5. Conclusion and design studies proposed at the Fifth European Framework Programme
The 21 st century is coming with a number of challenges to sustainable growth. In particular the
perspective of a growing energy demand satisfied primarily through the burning of fissile fuels, as is
the case today, has a limited future with regard to resource management and an increasing awareness
of the risk of climatic change. The nuclear fission power should take a substantial share, provided that
588

its extended use will not become a challenge to future generations, mainly with respect to the closure
of the fuel cycle.
The practicability of transmutation on an industrial scale requires operating an experimental
accelerator driven system, which will demonstrate the coupling of the accelerator, the neutron
producing target and the sub-critical core.
These objectives have been recognised by the European Atomic Energy Community (Euratom).
Therefore, in the frame of the Fifth Framework Programme of Euratom for research and training in the
field of nuclear energy, the European leading nuclear industrial companies and research centres,
propose to join together for performing the design studies of the different XADS concepts in order to
assess and compare them on a common basis and to recommend the development of the most adequate
concepts.
For this purpose, the general specifications for the European XADS will be more precisely
defined, a common safety approach based on the European safety requirements for future nuclear
plants will be elaborated, the research and development needs supporting the development of the
concepts will be identified, the technical feasibility concerns of each concept will be assessed, and a
preliminary cost assessment of each concept will be done.
REFERENCE
[1] L. Cinotti, G. Corsini, 1997, A Proposal For Enhancing The Primary Coolant Circulation in the EA,
International Workshop on Physics of Accelerator Driven Systems for Nuclear Transmutation and
Clean Energy Production, Trento, Italy.
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Figure 1. Experimental accelerator driven system assembly drawing of
an 80 MW facility cooled by Pb-Bi
REACTOR C
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Figure 2. Experimental accelerator driven system
assembly drawing of a 100 MW facility cooled by gas
591

HELIUM-COOLED REACTOR TECHNOLOGIES FOR
ACCELERATOR TRANSMUTATION OF NUCLEAR WASTE
Alan Baxter, Carmelo Rodriguez, Matt Richards, Jozef Kuzminski
General Atomics
2237 Trinity Drive, Los Alamos, NM 87544, USA
E-mail: Alan.Baxter@gat.com, Carmelo.Rodriguez@gat.com,
Matt.Richards@gat.com, Kuzminski@gat.com
Abstract
Helium-cooled reactor technologies offer significant advantages in waste transmutation. They are ideally
suited for use with fast, thermal or epithermal neutron energy spectra. They can provide a relatively hard
thermal neutron spectrum for transmutation of fissionable materials such as 239 Pu using ceramic-coated
particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of
transmutation with minimal or no intermediate reprocessing, (2) facilitate passive decay heat removal
via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly
efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly
resistant to corrosion for long times.
They also provide the hardest possible fast neutron environment since the helium coolant is
essentially transparent to neutrons and does not degrade neutron energies. This facilitates
transmutation of actinides that have low fission-to-capture ratios in the thermal neutron energy range.
In this paper, we report work on the development of two concepts using helium-cooled reactor
technologies for transmutation. Both concepts make use of thermal and fast energy spectra. One
concept (thermal-fast) may be more attractive for transmutation of nuclear waste in a once-through
mode, without reprocessing after initial removal of fertile uranium and fission products from the
waste. It uses a single type of transmuter to eliminate essentially all weapons-useful material in the
waste and achieve a significant reduction in total toxicity. It also has the potential to be economically
attractive by generating substantial amounts of electricity.
The other concept (two strata) may be more flexible and attractive to achieve deeper levels of
transmutation. In this system, the thermally fissile isotopes are destroyed in a critical reactor operation in
the passively safe Gas-Turbine Modular Helium Reactor, or GT-MHR, followed by a deep burn-up
phase in an accelerator-driven GT-MHR. Then the discharge material is processed into fast reactor fuel,
and irradiated in an accelerator driven Gas-Cooled Fast Assembly. The processing of the thermally
irradiated fuel would not require burning off the graphite. Instead, fuel compacts would be mechanically
extracted from the graphite fuel blocks, and the coated particles would be mechanically separated from
the compact binder material. The particles would then be chemically processed to separate the remaining
transuranics and produce fast reactor fuel. This is a similar process to that being considered for the
multiple-pass liquid metal transmutation process.
The gas-cooled fast assembly provides the hardest neutron spectrum for minor actinide transmutation,
and hence, maximum transmutation per pass. This minimises the number of reprocessing steps required
to reach a given degree of transmutation.
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1. Introduction
Nuclear waste can be stored in geologically stable repositories and allowed to decay for long
times. However, technologies are becoming available that have the potential to add significant value
to the disposal process. They provide for transmutation of nuclear waste into more stable materials
that decay relatively fast and are not attractive for use in nuclear weapons, thus reducing long term
toxicity and proliferation risks in the repositories.
Transmutation of nuclear waste could have profound benefits for world political stability and the
environment. It could drastically reduce the availability of weapons materials in the world, and reduce
waste disposal requirements in terms of space and safeguarding time.
2. The problem
Nuclear fuel production begins with uranium ore, which is not without hazard as it contains some
natural fission products and daughter elements that are created as uranium naturally decays to lighter
elements. However, this naturally occurring material provides a useful benchmark for evaluating the
nuclear waste that is ultimately produced. Uranium ore goes through several processing steps,
including an enrichment step in which the fraction of the lighter 235 uranium is boosted from 0.7% to
higher values (typically 3% for LWR reactor fuel) relative to the more naturally abundant 238 U. As a
fuel, the uranium is usually converted into oxide form (uranium oxide) and encased in a metal rod for
use in LWRs.
In addition to the production of fission products, neutron capture in both 235 U and 238 U leads to the
creation of plutonium, as well as minor actinides, (i.e. isotopes of elements with atomic number
greater than 92), including neptunium, americium, and curium. The fuel is used until the 235 U content
becomes too low to sustain the chain reaction, around 0.8%. It is then moved to a spent-fuel water
pool, where the hundreds of radioactive isotopes generated by the fission process begin to naturally
decay to stable, and generally harmless forms. After ten years of decay, spent nuclear reactor fuel is
composed of the materials listed in Table 1.
Table 1. Spent nuclear reactor fuel after 10 years decay
Actinides
Fission products
Uranium 95.6% Stable or short-lived 3%
Plutonium 0.9% Caesium & strontium 0.3%
Minor actinides 0.1% Iodine & technetium 0.1%
The table shows that about 98.6% of the spent-fuel inventory is not a concern for long-term
disposal. The uranium can be separated from the other materials and disposed of as class C waste, or
can be recycled. The stable and short-lived fission products are also of little concern, except for very
small quantities that may be classified as hazardous or mixed waste. The remaining 1% that is
composed of plutonium and minor actinides, as well as the 0.4% that is composed of caesium,
strontium, iodine, and technetium, need to be dealt with.
Plutonium is the unique waste component. It is fissionable, and capable of releasing significant
amounts of energy. It is also a hazardous material, particularly if inhaled in particulate form. Because
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of its potential use in nuclear weapons, there is great sensitivity about isolating plutonium from other
components of the nuclear waste stream. Minor actinides contain potential energy proportionate to
plutonium, although their generally small thermal cross-sections make them much more difficult to
fission.
If the 1% of the reactor discharge that consists of plutonium and minor actinides is transmuted, the
resultant waste stream would contain nearly 90% stable or short-lived fission products, and about 10%
caesium, strontium, iodine, and technetium. This is a favourable trade-off, as a significant amount of
energy would be produced, equal to about 18% of the energy that was produced in the reactor.
Therefore, in transmuting the plutonium and minor actinides (i.e. Np, Am, and Cm) from the
100 nuclear power plants in operation in the US, one would be generating the equivalent to another
18 power plants worth of electric power.
But what of the 0.4% of problem fission products? The caesium and strontium problem is caused
by a couple of isotopes having half-lives of about 30 years, which are hard to transmute. However,
with 30-year half-lives, the inventory drops by a factor of nearly 10 every century, so four centuries of
decay would drop the inventory by a factor of nearly 10 000. We can trust containers to provide
isolation for that long, and the need for isolation thereafter is greatly diminished. In contrast, the
technetium and iodine isotopes of concern are very long-lived and are primarily of concern well after
the containers have ceased to be effective. Fortunately, the iodine and technetium isotopes of concern
can be converted to stable isotopes if one has enough neutrons available. In that respect, we are
fortunate that the transmutation of the plutonium and the minor actinides can provide many available
neutrons.
The potential impact of removing and transmuting the plutonium and actinide wastes is
illustrated in Figure 1. The data for this figure are taken from the 1989 CURE report. Since that time,
however, the DOE ingestion toxicity hazard indices have been increased significantly. The net result
is that the toxicity levels illustrated in the figure have increased by about two orders of magnitude.
Taking this into account, even after a million years, untreated nuclear reactor waste is still
significantly more hazardous than natural uranium ore, and continued isolation from the environment
is still important. Furthermore, significant amounts of plutonium still remain in the untreated waste.
In contrast, the step of transmuting the actinides offers the potential to make the waste stream less
hazardous than uranium ore within three to four centuries, and reduce the need for isolation and
safeguarding. However, although this looks very attractive, it should be recognised that the real
impact may be somewhat less dramatic than the figure shows since even some minor residual
amounts of plutonium and minor actinides will tend to make the treated waste toxicity curve cross the
natural uranium ore line farther to the right.
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Figure 1. Impact of removing & transmuting actinides
Impact of Removing & Transmuting Actinides
Relative Ingested Toxicity
1000
100
10
1
0.1
Natural Uranium Ore
10 4 10 100 1000 10 4
Nuclear Reactor Waste
Reactor Waste without Actinides
0.01
Time (Years)
3. Gas-cooled systems for transmutation of nuclear waste
As stated above, if the 1% of the waste stream that is plutonium and minor actinides is
transmuted, and the iodine and technetium isotopes of concern are also converted to stable isotopes,
one has the basic elements of a solution to the nuclear waste problem.
Let us consider some of these elements. In the transmutation process, the materials to be
transmuted are irradiated in a neutron flux. Neutrons (1) can fission the atoms of the irradiated
materials creating atoms of lower atomic weight that are generally more stable, or (2) are absorbed by
the irradiated atoms. In the latter case, new heavier atoms are formed, which may in turn fission when
hit by other neutrons. For the long-lived fission products, neutron absorption transmutes them to
stable or short lived species.
Gas-cooled reactor technologies offer significant advantages in accomplishing this transmutation
process. They are ideally suited for use with thermal neutron spectra since they allow operation at
high temperatures and neutron energies that produce plutonium fission without the need for fertile
material as a burnable control poison. In addition, they also are ideally suited for use with fast neutron
spectra since they provide the hardest possible fast neutron environment for transmutation of higher
actinides, which are more inclined to fission in the fast neutron energy spectra. This is due to the fact
that the gas coolant is essentially transparent to neutrons and does not degrade the energy spectrum as
is the case with other coolants.
3.1 Thermal neutron systems
Generally both capture and fission cross-sections for thermal neutrons are an order of magnitude
larger than in a fast neutron spectrum. Thus, a suitably designed thermal system is a most effective
tool to destroy essentially all the proliferation-offensive Plutonium isotopes ( 239 Pu and 241 Pu). A
helium-cooled, graphite moderated, thermal neutron energy spectrum assembly using ceramic-coated
fuel, and operating as a critical system or as an accelerator-assisted sub-critical system, is an
attractive choice for this fission function for several reasons.
596

Figure 2. Neutron flux distribution and flux sections of
plutonium and erbium in gas-cooled assembly
As shown in Figure 2, a Modular Helium Reactor (MHR) type assembly produces a relatively
large flux in the thermal regime where fission cross-sections are quite high. This promotes fission. In
addition, the assembly operates in a temperature range (shown in Figure 2) in which the capture
cross-section of erbium has a resonance at a neutron energy such that it can be used as a burnable
poison to produce a strong negative temperature coefficient of reactivity. The lack of interaction of
the helium with neutrons means that temperature feedback is the only significant contributor to the
power coefficient. This provides for a stable operation of the reactor or sub-critical assembly. In
addition, it does not require 238 U as a burnable poison so no additional plutonium is produced in the
process.
Another feature of great importance in the thermal gas-cooled reactor or sub-critical assembly is
the use of ceramic-coated “TRISO” fuel particles. The ceramic materials are stable at high
temperatures, and have very high melting points. This provides large thermal margins to ensure fuel
integrity during loss of coolant events. Moreover, the coated particles are nearly spherical, and
include large gas expansion volumes within the coated particles. The expansion volumes are able to
accommodate the production of fission gas products within the coated particles with lower resultant
internal pressures. In addition, the spherical shape is better able to withstand the mechanical stresses
due to these pressures. The composite effect is that the particles can tolerate high levels of irradiation,
and allow deeper levels of transmutation (burn-up) without reprocessing. This capability has been
demonstrated in multiple reactor irradiations.
An important advantage of the ceramic coatings is that they are much more durable than metallic
coatings. Extrapolated corrosion test results indicate that the incremental waste exposure in the
597

epository due to corrosion of the ceramic coatings is expected to be negligible for hundreds of
thousands of years (Figure 3). We believe these particles offer the only practical chance to achieve the
toxicity reductions illustrated in Figure 1.
Figure 3. Particle integrity
Given these important features, the use of gas-cooled, graphite-moderated ceramic-coated-fuel
thermal reactors, or accelerator-driven sub-critical assemblies of the same type for destroying
weapons grade plutonium has been studied in detail [1]. The studies have led to the conclusion that an
assembly operating as a critical system can transmute about 90% of 239 Pu, and 65% of a total load of
Pu in a three-year pass. Then, if the 3-year irradiated load is further irradiated in an accelerator driven
sub-critical assembly for one more year, the destruction of 239 Pu and total Pu increases to 99.9% and
87% respectively, with no intermediate reprocessing.
3.2 Fast neutron systems
Fast reactor systems are typically much smaller than thermal reactors since they need a high
neutron flux and no moderator, and typically have higher fuel densities. This leads to higher power
densities, more demanding cooling requirements, and more complex cooling and cooling control
systems. The smaller delayed neutron fractions and complex reactivity feedback effects in fast
neutron systems can potentially produce severe reactivity and heating effects.
The fission cross-sections in the fast-neutron region are smaller than in the thermal region;
however, fission-to-absorption ratios are higher in the fast-neutron region than they are in the thermal
region. So, even though many minor actinides are hard to fission at any energy level because of their
small cross-sections, their relative destruction rates can be better than in thermal reactors if a high
neutron flux is provided.
Calculations show that a fast sub-critical assembly cooled with helium gas allows a destruction
rate of actinides of approximately 26% (in weight) per year. This is somewhat better than has been
reported in other studies for liquid metal cooled systems, possibly because the gas coolant allows the
production of a harder neutron energy spectrum, and consequently, a higher fission to absorption
598

atio. Given this rate of destruction, one could irradiate the actinides for several years and then store
the discharge in a geological repository where further natural decay would take place. So, if one
desires to destroy certain minor actinides, there is merit in using fast neutrons after plutonium is
transmuted in a thermal spectrum. And if one wishes to do so, it helps that the amount of actinides is
a very small portion of the initial waste (0.1%) since the fast transmutation assemblies end up being
smaller than they would have to be to transmute plutonium as well. These findings form the basis for
a transmutation scheme utilising both fast and thermal neutron spectra as discussed in the next
section.
4. Burn-up possibilities
Based on the above considerations, we have explored the process of using (1) thermal neutrons
(near the cross-section resonance peak) to do what they do best, i.e. fission plutonium, and (2) fast
neutrons to fission minor actinides. One advantage of using this process is that most of the
transmutation of plutonium is done in the thermal regime where technologies are more mature and
development risks are lower. Since the amounts of minor actinides found in the waste material are
significantly lower than the amounts of plutonium, the fast assemblies needed can be significantly
smaller or fewer than the thermal assemblies.
The fuel cycle that we have studied for this scheme is as described in the previous section: three
years of transmutation of plutonium and minor actinides in a thermal neutron spectrum assembly
operating in the critical mode, followed, without reprocessing, by one year of transmutation in the
same thermal neutron spectrum assembly operating as an accelerator-assisted sub-critical system. At
this point, essentially all fissionable materials are burned up. What remains is mainly non-fissionable
minor actinides, which are moved to a fast neutron spectrum assembly operating as a sub-critical
system.
Burn-up calculation results for this fuel cycle are shown in Figure 4 for an initial 1 000 kg charge
of weapons-grade plutonium. As the figure indicates, most of the Pu transmutation is accomplished in
a thermal critical regime. When this is followed by a one-year irradiation step in a thermal sub-critical
regime (accelerator driven), essentially all 239 Pu is gone. At this time, a three-year step of
transmutation in a fast subcritical regime leads to Point C in the chart, when most of the initial charge
is gone. If this remaining material is placed in a repository for 200 years, only 60 out of 1 000 kg of
the initial charge are left.
599

Figure 4. Burn-up of plutonium using thermal and fast neutron spectra
1 200
1 000
800
Point A
Total Pu
239 Pu
Total MA
Kilograms
600
400
Point B
200
Point C
0
Initial charge
After
thermal,
critical
After
thermal,
sub-critical
After
fast,
sub-critical
In
repository
Residence times
Thermal, critical: 3 years
Thermal, critical: 1 year
Fast, sub-critical: 3 years
Repository: 200 years
The significance of Point B in the above figure is that it is equivalent to the level of
transmutation demonstrated in the Peach Bottom 1 reactor. In that test, Plutonium fuel particles
coated in Silicon Carbide were irradiated to levels exceeding 700 000 MW-days per Metric ton of Pu
fuel, which transmuted over 95% of 239 Pu, and corresponds to the Point A-to-Point B trajectory in
Figure 4. Two irradiated “TRISO” particles from this test are shown in Figure 5. The second (lighter)
layer (from the outside in) is the silicon carbide coating. The two gray layers on each side of the
silicon carbide layers are pyrocarbon layers that provide mechanical protection and pre-compression
for very high structural strength margins in the silicon carbide. The central part of the particle
contains the transmuted material. The dark areas within the silicon carbide layer are empty spaces
occupied by fission gasses.
600

Figure 5. Irradiated TRISO particles
1 200
Burn-up of LWR TRU
1 000
Kilograms
800
600
400
Total Pu
239
PU
Total MA
200
0
Initial
charge
After
thermal
critical
After
thermal
sub-critical
After fast
sub-critical
After
multiple
fast passes
The conclusions from the burn-up levels shown in Figure 4 and the durability of the silicon
carbide “TRISO” particles illustrated in Figure 3 are that (1) transmutation can reduce plutonium and
minor actinides waste by about two orders of magnitude, and (2) the transmuted material will remain
isolated from the repository environment for hundreds of thousands of years.
Similar preliminary calculations have been performed for LWR transuranic waste. In this case,
the isotopic composition of the plutonium material to be transmuted is somewhat different than
weapons-grade plutonium. In addition, there are other minor actinides present in the mix.
Nevertheless, the results are equally encouraging, as shown in Figure 6.
Figure 6. Burn-up of LWR TRU waste using thermal and fast neutron spectra
Pu Oxide
747,000 MW-days/tonne
>95% 239 Pu, and
>65% all Pu transmuted
Th-Pu Oxide
183,000 MW-days/tonne
>95% 239 Pu transmuted
Consider now the manner in which the thermal and fast neutron energy spectra could be
packaged together. There are two concepts that seem attractive for different reasons. One is based on
the use of a single type of transmuter, running part time in the critical mode, and part time in the
601

accelerator-driven sub-critical mode. We call it the thermal-fast concept. The other uses two separate
types of transmuters, one thermal and the other fast. We call this the two-strata concept.
5. The thermal-fast concept
The single transmuter-type, thermal-fast concept is illustrated in Figures 7 and 8. Referring to
these figures, the transmutation assembly consists of a steel vessel housing, inside of which there is
an annular nuclear transmutation region that operates in the thermal neutron energy regime. In this
annular region, plutonium and minor actinides from LWR waste are fissioned together in TRISO
particles. Most of the mixture, about 90%, is plutonium. The remaining 10% is minor actinides.
Fission neutrons in this annular region are thermalized in graphite blocks in which the TRISO
particles are contained as shown in Figure 9. Surrounding the annular thermal region there is an inner
and an outer graphite neutron reflector. This thermal region operates in the critical mode for 75% of
its cycle time, followed by operation in an accelerator-assisted sub-critical mode for the remaining
25%.
Still referring to Figures 7 and 8, in the centre of the inner reflector there is a cylindrical region,
approximately 15% of the size of the active thermal region, that operates in the fast energy neutron
energy regime. This region consists of tungsten tubes that house TRISO particles already transmuted
in the thermal region. Therefore, they contain mainly minor actinides. The main motivation for
including this fast assembly inside of the thermal assembly is to take advantage in the fast fission
process of the large heat storage and conduction heat removal capabilities of the thermal assembly.
Figure 7. Thermal-fast transmuter elevation
Figure 8. Thermal-fast transmuter cross-section
Beam
Fast Region
Thermal Region
Transmutation:
Pu, MA (thermal spectrum)
MA (fast spectrum)
Tc, I (thermal spectrum)
Spallation Target
Connection to
Power Conversion System
Transmutation: MA
Tc, I
Pu
Beam window flange
Beam
Power Conversion
System Duct
The fast cylindrical region is designed so that, by itself, it is sub-critical. However, neutrons
reaching it by travelling from the thermal region through the reflector can cause fission and get
amplified, thus creating sub-critical transmutation.
As discussed above, the transmuter operates in the critical mode for approximately three years,
which corresponds to 75% of its cycle time. In this mode, the fission process is driven by the critical
reaction in the thermal region. After that, the thermal region becomes sub-critical, and is then driven
for a fourth year to cause deep levels of transmutation by neutrons generated in a spallation target
located in the centre of the fast region. The target is driven by the proton beam illustrated in Figures 7
and 8. Deep levels of transmutation can be achieved with no reprocessing thanks to the encapsulation
602

in ceramic-coated microspheres of the materials to be transmuted, which accommodate the production
of fission gas products within internal expansion volumes.
A beneficial anti-proliferation effect of including the fast assembly within the thermal assembly
is that the neutron economy in the integrated assembly cannot support breeding.
Going back to the operating sequence, the fact that the transmuter needs the proton beam for only
a part of its operating time makes the entire process more economical because the accelerator can be
time-shared by several transmuters in the plant configuration such as that illustrated in Figure 10.
Referring again to Figures 7 and 8, there is shown a coaxial duct in the lower part of the
transmuter. The outer part of the duct brings in cold cooling helium to remove fission heat from the
transmuter. The helium then flows upward in an annular space between the inside of the vessel and
the outside of a steel barrel that contains the thermal-fast assembly.
Figure 9. TRISO coatings and graphite are excellent engineering barriers
for normal operation, severe accidents, and permanent disposal
Helium then flows downward through cooling channels in the fission regions of the transmuter,
and carries the heat at a temperature of 850°C through the central part of the coaxial duct to a directcycle
gas-turbine-generator system that generates electricity. The high operating temperatures and the
603

characteristics of this direct (Brayton) power conversion system allow electric generation with a high
net thermal efficiency of approximately 47%.
This high power conversion efficiency, the fact that 75% of the transmutation is done in a critical
operating mode, and the fact that the proton accelerator is time-shared by four transmuters, leads to a
favourable revenue-cost balance and the potential to attract investment for the deployment of these
units.
Roughly, the cost of the plant configured as shown in Figure 10 (four 600 MW th
transmuters, and
one 15 MW beam accelerator) would be expected to be in the $1.5B to 2.0B range. This would
translate into an annual cost of $190M assuming interest on and return of capital of 8.8%, plus typical
nuclear plant operations and maintenance (O&M) costs increased by 50% to account for accelerator
O&M. It also includes an allowance for decommissioning the plant, but assumes, however, that the
fuel (to be transmuted) is government-supplied.
Figure 10. Thermal-fast plant representative configuration
Operating Sequence
Ti me Tr ansm itter 1 Transmuter 2 Transm uter 3 Transmuter 4
t o
Crit ical Crit ical Cr itical Su bcritical
t + 1 yr o
Crit ical Crit ical Subcrit ical Crit ical
t + 2 yrs o
Crit ical Subcritical Crit ical Crit ical
t + 3 yrs o
Su bcritical Critical Critical Critical
th erm al re gi o n
fast region
Af te r t he 1 y ea r of subcr itica l oper at ion:
é Fr esh fue l g o es i nt o th er ma l se cto r
é Fu el com p acts i rr ad i ated for 4 year s in
ther mal sector are moved to fast sector
1000 MeV, 15 mA Beam
t o + 3 yr
t o
+ 2 yr
Transmuter 1
Tr ansm uter 2
t o + 1 yr
t o
Transmuter 3
Transmuter 4
The revenues, assuming 4 cents per kWe-hr, 75% plant availability, 47% transmuter thermal
efficiency, and an accelerator efficiency (beam to electric power ratio) of 32% would be
approximately $270M per year.
These estimates suggest that transmutation plants of this type have the potential to be
economically viable and possibly attract investment.
From the safety standpoint, there are two important considerations. Criticality and cooling.
Criticality safety in this respect is ensured by the use of erbium in the thermal assembly. Erbium has a
neutron capture cross-section that increases with temperature, and peaks a higher temperature than the
fission cross-section of 239 Pu. Erbium and plutonium quantities can be selected that provide a strong
negative temperature coefficient of reactivity during the entire fuel cycle.
The other important safety consideration is cooling. In this respect, the geometry of the assembly
(tall and thin, and annular thermal configuration) has been shown to provide for passively safe
conduction cool-down of a 600 MW th
thermal spectrum-only assembly, even in a loss of coolant
604

(LOCA) event. The effect on this feature by the inclusion of a fast fission region is currently being
studied. However, preliminary conservative calculations suggest that including a fast region may still
allow passive conduction cooling in a LOCA event if thermal region power is lowered so that total
power of the transmuter is kept below 600 MWt.
Thus, it appears that levels of safety comparable to those that are encountered in gas-cooled
reactors may also be achievable in a thermal-fast gas-cooled transmuter.
Important work is still needed, however, to further advance the design of the thermal-fast
transmuter. Preliminary analyses seem to indicate, for example, that the fast flux generated in the
central region during critical operation of the thermal region can be relatively low. This would mean
that transmutation of minor actinides during this time would also be low, and that more reliance on
the accelerator to generate a fast flux would be needed. However, there are potentially more attractive
ways to obtain a higher fast flux during these conditions that need to be evaluated. These include
reducing the density of the graphite immediately surrounding the fast region.
6. The two-strata concept
In the two-strata concept, the thermal and the fast neutron spectra are separated. The system is
illustrated in Figure 11. As shown in this figure, the system includes the front-end separation step in
which uranium in the waste stream is separated and recycled to the commercial fuel cycle, or
disposed of as low-level (Class C) waste.
Figure 11. The two strata process
To the Fast Step
605

The plutonium and minor actinides are then converted to silicon carbide-coated fuel particles and
assembled in the three graphite-moderated thermal spectrum critical reactors shown in the figure. The
fuel fabrication stage is shown as step 4 in the figure, and the thermal reactors are shown as step 5.
The GT-MHR operates at a sufficiently high coolant temperature that it can be coupled to a direct
cycle gas turbine for electricity generation at approximately 47% efficiency [2]. A cross-section of
the MHR core is shown in Figure 12.
Figure 12. GT-MHR cross-section
The discharge from the reactors is further burned-up in an accelerator driven MHR, or AD-MHR.
This is step 6 in Figure 11. The AD-MHR core is very similar to the GT-MHR core, the primary
difference being the provision for an accelerator target for neutron production by spallation, and a
surrounding pressure vessel, both located in the central graphite reflector. This graphite reflector
serves to moderate the spallation neutrons and produce a thermal flux spectrum in the fuel annulus.
The beam enters at the top of the vessel, and is directed onto the target in the middle of the core
assembly to produce high-energy neutrons. However, methods are being evaluated to allow a straight
side beam entry into the core assembly. This would eliminate the need for a 90-degree beam turn
from its normal horizontal orientation to the vertical entry.
Following the GT-MHR and AD-MHR there would be a sub-critical fast neutron energy
assembly used mainly to finish the job of burning minor actinides.
An elevation section of this modular, fast helium-cooled assembly (AD-FMHR) is shown in
Figure 13. A similar design has been proposed by Framatome [3] also to burn Minor Actinides. In this
concept, the proton beam enters at the top of the vessel and is directed onto the target in the middle of
the core assembly to produce high-energy neutrons. However, as for the AD-MHR, methods are being
evaluated to allow a straight side beam entry into the core assembly.
606

Figure 13. Elevation AD-FMHR
The fast helium-cooled assembly (AD-FMHR) shown in Figure 13 is based on the gas-cooled
fast reactor (GCFR) developed by General Atomics in the 1970s with US DOE support. In the GCFR
design, the elements in the core and blanket assembly are externally similar, and each element is
hexagonally shaped. The overall length of the elements is 118.25 inches The structural material for
the GCFR element was 20% cold worked type 316 stainless steel. Other structural materials, such as
Inconel 718 and tungsten, are being considered for the AD-FMHR application, to ensure that a loss of
flow or pressure accident does not result in fuel damage or release of fission products. The fuel and
blanket elements are clamped rigidly and pre-loaded at their upper end into the grid plate. This
preload force reacts on the upper face of the grid plate through a compression tube.
In the AD-FMHR, the fuel section of the elements would be 150 cm high, with 100-cm top and
bottom blankets. The helium coolant would flow in the upward direction. The inlet temperature would
be 300° C and the outlet temperature would be at least 530°C or higher, based on the use of hightemperature
fuel coatings.
The assembly would use the same fuel form as envisioned for the liquid metal options that are
also being considered. Obviously, no moderator material would be used. However, the use of
ceramic-type high-temperature fuel coatings that would result in wider safety margins is being
607

investigated. Most of the fuel in this assembly would consist of minor actinides, which would allow
the assembly to be fairly small. Sizes in the order of 60 to 100 MWt would be attractive. Fission
products, such as 99 Tc and 129 I, could be placed in the outer blanket for transmutation.
The energy in the hot helium exiting the fuel would be available to either generate steam or
directly drive a gas turbine and generate electricity.
The use of helium coolant in this accelerator-driven fast-spectrum sub-critical assembly offers
important potential advantages, such as the following:
• Potential use of coated fuel with its capability to 1) withstand higher temperatures than other
fuels, 2) retain radionuclides in the event of accidents, and 3) provide an extra long-life
barrier for the retention of radioactivity in the repository.
• No metal-air or metal-water chemical reactions.
• No generation of mixed waste coolant.
• No leakage of highly toxic metal.
• Potential elimination of high pressure steam generators by using a direct-cycle gas turbine.
• Potentially harder neutron energy spectrum for more effective burning of actinides.
• Much more viable in-service inspection. In the gas-cooled assembly, the integrity of the fuel,
fuel supports, etc. is easily observable, which provides for greater safety assurance.
Earlier analyses performed for the development of the GCFR showed that the natural circulation
of helium would provide adequate passive cooling following a loss of forced circulation for reactor
ratings up to 840 MWt. This capability needs to be explored beyond design basis events.
7. Conclusions
Gas-cooled nuclear reactor technologies offer the potential to eliminate essentially all weaponsuseful
material in nuclear waste, and achieve more than two orders of magnitude reduction in the
amount of high-level waste. Repository heat loads and the toxicity of the waste are also significantly
reduced. The process provides a durable transmuted waste form that is highly resistant to corrosion,
without generating mixed waste.
The process uses thermal and fast neutron energy spectra. 239 Pu and other fissionable materials
have large fission cross-sections in the thermal spectrum. Thus, they are fissioned in a thermal fission
region of the transmuter. Minor actinides are more inclined to fission in a fast neutron energy
spectrum. Thus, they are fissioned in a fast fission region of the transmuter.
The process has the potential to transmute about 75% of the waste in a nuclear critical mode.
Then, use is made of a proton accelerator to generate spallation neutrons and drive the fission process
in a sub-critical mode to deep levels of burn-up. Most importantly, these deep burn-up levels are
achieved with no plutonium reprocessing. This is made possible by encapsulating the waste to be
transmuted in ceramic-coated microspheres that accommodate large amounts of fission products in
spherical expansion volumes.
608

Deep burn-up of 239 Pu and fissionable materials with no plutonium reprocessing, and the possible
use of a fast neutron region within the thermal region in the transmuter, which precludes breeding, are
important proliferation-resistance features of the proposed process.
Preliminary calculations suggest that the unique reactivity and cooling safety features offered by
gas-cooled nuclear reactors can also be implemented in the proposed transmuter.
The use of a direct-cycle gas turbine-generator power conversion system with the proposed
transmuter would lead to conversion efficiencies of approximately 47% when the transmuter is
operating in the critical mode. This, along with the fact that the accelerator may only be needed for
the 25% deep burn-up phase of the cycle, leads to a relatively high overall efficiency and low cost.
Preliminary economic analyses suggest that the proposed transmutation process has the potential to be
economically viable and attract investment for deployment.
REFERENCES
[1] D. Alberstein, A. Baxter, and W. Simon, The Plutonium Consumption Modular Helium
Reactor, IAEA Technical Committee on Unconventional Options for Plutonium Disposition,
November 1994, Obninsk, Russia.
[2] Gas Turbine-Modular Helium Reactor (GT-MHR) Conceptual Design Description Report,
General Atomics report 910720, Rev. 1, July 1966.
[3] B. Carluec and P. Anzieu, Proposal for a Gas-cooled ADS Demonstrator, 3rd International
Conference on Accelerator Driven Transmutation Technologies and Applications, Praha, Czech
Republic, June 7-11, 1999.
609

POSTER SESSIONS
611

POSTER SESSION
PARTITIONING
M.J. Hudson (University of Reading)
613

STUDIES ON BEHAVIOUR OF SELENIUM AND ZIRCONIUM IN PUREX PROCESS
J.A. Suárez, G. Piña, A.G. Espartero, A.G. de la Huebra
CIEMAT, Dpto. de Fisión Nuclear, Avda. Complutense, 22, 28040 Madrid, Spain
Abstract
The studies about the behaviour of 79 Se and 93 Zr allow identifying the PUREX process streams in
which these radionuclides remain. In this way, in further investigations, other specific separation
techniques could be applied in order to get targets pure enough for their possible transmutation. The
studies showed that the extraction of Se decreases when the acidity of the aqueous phase increases
from 0.5 M to 8.0 M, obtaining D values from 1.1E-2 to 3.4E-4 respectively and the uranium
concentration has not influence in the Se extraction up to 130 g/L. In the case of Zr, its distribution
coefficent value increases when the acidity of the medium increases from 1.96E-2 to 3.84E0 and
3.1 M to 8.0 M respectively, but the study about the influence of the uranium concentration in the Zr
extraction showed that the Zr distribution coefficient decreases down to 1.0E-2 when the organic
phase is 100% saturated in uranium (130 g/L). Then it can be concluded that 79 Se and 93 Zr remain in
the raffinate of the first step of PUREX process.
615

1. Introduction
The study of the behaviour of some fission products such as 79 Se, 93 Zr, 107 Pd and 126 Sn about their
possible liquid-liquid extraction, together with uranium and plutonium, by the tributhyl phosphate (TBP)
in the main steps of the process for U and Pu separation and purification in the irradiated nuclear fuel
reprocessing (PUREX process), is considered one of the main lines of the project Long-lived
Radionuclides Separation by Hydrometallurgic Processes developed within the framework of
CIEMAT-ENRESA agreement.
The main steps of the PUREX process consist on a jointly separation of U and Pu from the
dissolution of a spent fuel (Figure 1), which are separated in a later step, by consecutive re-extraction
processes, being necessary to reduce the oxidation state of Pu(IV) to Pu(III) and to use a weak acidic
medium for U re-extraction. Then, both elements are purified independently by liquid-liquid extraction
processes.
The aim of this paper is to show the studies carried out about the behaviour of 79 Se and 93 Zr, in
order to identify the PUREX process streams in which these radionuclides remain. In this way, in
further investigations, other specific separation techniques could be applied in order to get targets pure
enough for their possible transmutation.
Figure 1. Simplified flow diagram of PUREX process
Solvent Feed Scrub
TBP 30%
n-Dodecano
U - Pu
HNO 3
3M
HNO 3
3 M
HNO 3
0.2 M
+ Reducing
Pu Strip
U Strip
HNO 3
0.01 M
Raffinate
Pu Product
U- Product Used Solvent
Co-decontamination Pu Re-extraction U Re-extraction
2. Experimental
2.1 Reagents
• Analytical grade TBP, n-dodecane, nitric acid, Na 2 SeO 3·5H 2 O and ZrOCl 2·8H 2 O.
• Uranium of nuclear purity prepared from nitric digestion of UO 3 .
• Beta-gamma tracers of 75 Se and 95 Zr (Amersham UK).
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2.2 Chemical speciation
The oxidation states of selenium in nitric acid medium, considering the conditions in which the
irradiated nuclear fuel is dissolved in the first step of PUREX process, can be (IV) and (VI), although
Se(IV) is the most stable oxidation state in aqueous solution, and its chemical form in acid nitric
medium is H 2 SeO 3 [1].
The chemical form of zirconium is difficult to evaluate due to its strong tendency to hydrolyse.
Depending on the medium acidity, zirconium forms complexes and generates a great variety of colloid
species. The tetra-positive ion is considered the dominant zirconium specie in the first step of PUREX
process, where the nitric acid concentration is high [2].
Depending on the acidity of the medium, zirconium can be extracted by TBP-dodecane in the form
of di-, tri- or tetra nitrate complex [3-6]: Zr(OH) 2 (NO 3 ) 2 (TBP) 2 , ZrOH(NO 3 ) 3 (TBP) 2 , Zr(NO 3 ) 4 (TBP) 2 .
2.3 Reference irradiated nuclear fuel
To calculate the concentration of Se and Zr present in an irradiated nuclear fuel, a burnt up of
40 000 MWd/tU, 3.5% enrichment and a cooling time of 5 years were considered. This reference fuel
element is the same used by ENRESA in the calculation and assumptions concerning the project Spent
Fuel Characterisation and Behaviour under Relevant Repository Conditions [7].
2.4 Experimental conditions and equilibrium diagrams
The studies were carried out considering a mass concentration of 67.5 g Se/tU and 4 284 g Zr/tU.
The uranium concentrations considered were 25, 50, 100, 150, 200, 250 and 300 g/L.
Se(IV) and Zr(IV) equilibrium diagrams were obtained in the following conditions, which are
those of the first extraction and scrubbing step of the PUREX process [8,9]:
• HNO 3 concentration: 3, 4 and 5 M.
• TBP in dodecane concentration: 20, 25 and 30% (v/v).
• Phase ratios (Or/Aq): 1, 2 and 3.
• Se(IV) concentration: 1.70, 3.40, 6.80, 10.2, 13.6, 17.0 and 20.4 mg/L
• Zr(IV) concentration: 0.11, 0.22, 0.43, 0.64, 0.86, 1.07 and 1.28 g/L.
The stable Se(IV) and the tracer 75 Se were prepared in the same chemical form Se O 3 2- . This
specie was obtained by flow back boiling Na 2 SeO 3·5H 2 O and 75 Se in nitric acid 3, 4 and 5 M.
The stable Zr(IV) as ZrO 2 Cl 2 and the tracer 95 Zr were treated successively with HNO 3 18 M to
obtain the specie ZrO 2 (NO 3 ) 2 .
The influence of the uranium concentration was studied considering those nitric acid
concentrations in which the extraction of both, Se(IV) and Zr(IV), was maximum. The uranium
concentrations tested were ranged from 25 g/L to 300 g/L.
617

Plutonium was simulated with the non active chemically analogous element Ce(IV) being the
total mass 2 861 gCe/tU.
Se(IV) and Zr(IV) were tested independently and the equilibria were carried out with
TBP-dodecane as organic phase. The equilibria between nitric and TBP phases were reached by
30 minutes of mechanical shaking at 950 u/min and the separation was performed after 60 minutes of
decantation time. 75 Se, 95 Zr, free [H + ] and uranium determinations were carried out in both phases.
Results are detailed in Figures 2 to 7 and Tables 1 to 4.
2.5 Analytical procedures
• Determination of Se concentration: it is performed by gamma spectrometry using the net peak
area at 136 keV line of 75 Se.
• Determination of Zr concentration: it is performed by gamma spectrometry using the net peak
area at 724 keV line of 95 Zr.
• Determination of free [H + ]: the cations present in the sample are complexed or precipitated by
an excess of oxalic/oxalate buffer at pH 7.0. The solution is potentiometric titrated with KOH
0.1 M until pH 7.0.
• Determination of U concentration (Method I): gravimetry as U 3 O 8 .
• Determination of U concentration (Method II): it is performed by gamma spectrometry using
the net peak area at 186 keV line of 235 U.
3. Results and discussion
3.1 Extraction of Se(IV)
The extraction equilibrium diagrams of Se(IV) show the typical isothermal curves that are
generally obtained in these kind of studies (Figure 2). For all TBP concentrations studied, the
extraction of Se(IV) decreases when the acidity of the aqueous phase increases. The maximum value
of the distribution coefficient (D), obtained for the lower acid media in the steady state (1.83 M) is
4.5E-3 (Table 1), which indicates the low Se(IV) extraction in these conditions.
618

Figure 2. Isothermal equilibrium curves of Se(IV) with different
HNO 3 (M) and TBP-dodecane concentrations. Or/Aq phase ratio 1
0.11
0.10
[H + ] aq 3M TBP 30%
[Se(IV)]or (mg/L)
0.09
0.08
0.07
0.06
0.05
0.04
0.03
0.02
0.01
[H + ] aq 3M TBP 25%
[H + ] aq 4M TBP 30%
[H + ] aq 5M TBP 30%
[H + ] aq 3M TBP 20%
[H + ] aq 4M TBP 25%
[H + ] aq 5M TBP 25%
[H + ] aq 4M TBP 20%
[H + ] aq 5M TBP 20%
0.00
0 5 10 15 20 25
[Se(IV)] aq (mg/L)
H + (M)
Aq. phase
Ratio
Or/Aq
Table 1. Distribution coefficients of Se(IV) by TBP 30%
in dodecane. Initial Se(IV) concentration 20.4 mg/L.
H + (M)
Aq. phase
U (g/L)
Aq. phase
U (g/L)
Or.phase
Steady state
Se(IV)
(mg/L)
Aq. phase
Se(IV)
(mg/L)
Or. phase
Distribution
coefficient
(D)
2.99 3 1.83 0 0 21.9 0.0977 4.46 E-3
2.99 2 2.12 0 0 22.0 0.0882 4.04 E-3
2.99 1 2.49 0 0 21.3 0.0733 3.44 E-3
3.90 3 2.53 0 0 21.9 0.0683 3.12 E-3
3.90 2 2.87 0 0 21.5 0.0563 2.62 E-3
5.27 3 3.43 0 0 22.8 0.0488 2.14 E-3
3.90 1 3.46 0 0 20.8 0.0450 2.16 E-3
5.27 2 3.90 0 0 22.0 0.0376 1.71 E-3
5.27 1 4.21 0 0 22.4 0.0296 1.32 E-3
3.02 3 1.93 < 0.5 9 21.9 0.204 9.32 E-3
3.00 3 1.92 1 17 21.6 0.348 1.61 E-2
2.94 3 2.00 1 36 21.5 0.453 2.11 E-2
3.14 3 2.24 1 52 22.4 0.178 7.96 E-3
3.03 3 2.30 2 68 22.8 0.105 4.60 E-3
3.07 3 2.47 6 83 23.4 0.0941 4.02 E-3
3.10 3 2.55 35 125 21.7 0.0621 2.86 E-3
3.09 3 2.60 170 138 22.2 0.0511 2.30 E-3
619

The influence of uranium and plutonium concentration in the distribution coefficient of Se(IV),
was tested in the conditions in which the Se extraction is maximum (TBP 30% and HNO 3 3 M).
The obtained results (Table 1 and Figure 3) show that uranium concentration has not influence in
the extraction of Se(IV), because the D value increment is not significant. For this, it can be
established that Se(IV) is not extracted in the first U and Pu co-extraction step of the PUREX process.
Figure 3. Influence of U concentration (g/L) on distribution
coefficient of Se(IV), TBP-dodecano 30% HNO 3 5 M
Distribution coefficient (D)
1.0E-01
1.0E-02
1.0E-03
0 50 75 100 125 150
[U] or (g/L)
In order to complete the study about the influence of the aqueous nitric acid concentration in the
Se(IV) extraction, samples with acid nitric concentrations between 0.5 M and 3 M and between 6 M
and 8 M were analysed considering the maximum extraction conditions (TBP 30% and Se(IV)
concentration 20.4 mg/L).
Figure 4. Influence of HNO 3 concentration (M) on extraction
coefficient of Se(IV) by TBP-dodecane 30%
1.0E-02
Distribution coefficient (D)
1.0E-03
1.0E-04
0 1 2 3 4 5 6 7
[HNO 3 ] aq (M)
Table 2 and Figure 4 show that the tendency of the Se(IV) distribution coefficient is the same
observed before, nevertheless the maximum D value obtained is 1.0E-2 that means Se(IV) is not
extracted by TBP, so it can be concluded that 79 Se remains in the raffinate of the first step of PUREX
process.
620

Table 2. Influence of aqueous acid nitric concentration less than 3 M
and higher than 5 M, in the distribution coefficient of Se(IV)
by TBP 30% in dodecane. Initial Se(IV) concentration 20.4 mg/L.
H + (M)
Aq. phase
Ratio
Or/Aq
H + (M)
Aq. phase
Se(IV) (mg/L)
Aq. phase
Steady state
Se(IV) (mg/L)
Or. phase
Distribution
coefficient (D)
0.524 3 0.39 21.0 0.223 1.06 E-2
1.06 3 0.69 20.9 0.191 9.12 E-3
2.13 3 1.31 21.4 0.149 6.98 E-3
6.17 3 4.56 22.3 0.0285 1.28 E-3
7.22 3 4.97 23.6 0.0228 9.67 E-4
8.20 3 5.99 22.9 0.0078 3.43 E-4
3.2 Extraction of Zr(IV)
The extraction equilibrium diagrams of Zr(IV) show the typical isothermal curves obtained in this
kind of studies (Figure 5). Although the maximum extraction is obtained when the aqueous nitric acid
and TBP concentrations are high, Zr(IV) is not extracted by TBP in a significant quantity because the
maximum distribution coefficient value is 0.55 (Table 3) obtained for an acidity in the steady state of
4.7 M.
Figure 5. Isothermal equilibrium curves of Zr(IV)
with different HNO 3 (M) concentration and TBP-dodecane % concentrations
0.45
[H + ] aq 5M TBP 30%
[H + ] aq 5M TBP 25%
[Zr(IV)]or (g/L)
0.30
[H + ] aq 5M TBP 20%
[H + ] aq 4M TBP 30%
0.15
[H + ] aq 4M TBP 25%
[H + ] aq 4M TBP 20%
[H + ] aq 3M TBP 30%
[H + ] aq 3M TBP 25%
[H + ] aq 3M TBP 20%
0
0 0.2 0.4 0.6 0.8 1.0 1.2 1.4
[Zr(IV) aq (g/L)
621

The influence of uranium and plutonium concentration in the distribution coefficient of Zr(IV)
was tested in the conditions in which the extraction is maximum, TBP 30% and HNO 3 5 M.
The obtained results (Table 3 and Figure 6) show that when uranium concentration increases the
Zr(IV) extraction decreases. The value of the distribution coefficient is very low (0.01) when the
organic phase is 100% saturated with uranium (130 g/L). This effect shows that Zr(IV) is not extracted
in the first cycle of co-extraction of U and Pu in the PUREX process.
Figure 6. Influence of U (g/L) concentration on distribution coefficient of Zr(IV),
TBP-dodecane 30%, HNO 3 5 M
Distribution coefficient (D)
1.0E+00
1.0E-01
1.0E-02
0 50 100 150
[U] or (g/L)
Due to the extraction of Zr(IV) by TBP increases as the nitric acid concentration increases
(Figure 5), it is necessary to study the behaviour of Zr(IV) when the nitric acid concentration, in the
aqueous phase is higher (from 5 M to 8 M) than the range considered in the first studies and within the
maximum extraction conditions TBP 30% and Zr(IV) concentration 1.28 g/L.
H + (M)
Aq. phase
Ratio
Or/Aq
Table 3. Distribution coefficient of Zr(IV) by TBP 30%
in dodecane. Initial Zr(IV) concentration 1.28 g/L
H + M
Aq. phase
U g/L
Aq. phase
Steady state
U g/L
Or. phase
Zr(IV) g/L
Aq. phase
Zr(IV) g/L
Or. phase
Distribution
coefficient
(D)
3.08 3 1.86 0 0 1.31 0.0256 0.0196
3.08 2 2.13 0 0 1.29 0.0395 0.0307
3.08 1 2.49 0 0 1.25 0.0626 0.0203
4.07 3 2.55 0 0 1.14 0.0580 0.0510
4.07 2 2.92 0 0 1.16 0.0951 0.0823
5.06 3 3.20 0 0 0.993 0.119 0.120
4.07 1 3.48 0 0 1.14 0.163 0.144
5.06 2 3.70 0 0 0.935 0.192 0.205
5.06 1 4.39 0 0 0.966 0.362 0.375
5.50 1 4.70 0 0 0.827 0.452 0.547
4.89 1 4.68 < 0.1 52 1.12 0.206 0.183
4.97 1 4.82 < 0.1 97 1.26 0.0762 0.0605
4.88 1 4.88 23 128 1.34 0.0248 0.0185
4.91 1 4.88 76 135 1.34 0.0159 0.0119
4.99 1 4.90 118 138 1.33 0.0150 0.0113
4.89 1 4.90 168 137 1.36 0.0140 0.0103
622

Figure 7. Influence of HNO 3 concentration (M) on
distribution coefficient of Zr(IV) by TBP-dodecane 30%
istribution coefficient (D)
5.0
4.0
3.0
2.0
1.0
0.0
1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0
[HNO 3 ] aq (M)
Table 4. Influence of nitric acid concentration higher than 5 M on the distribution coefficient of
Zr(IV) by TBP 30% in dodecane. Initial Zr(IV) concentration 1.28 g/L.
H + (M)
Aq. phase
Ratio Or/Aq
H + (M)
Aq. phase
Zr(IV) (g/L)
Aq. phase
Steady state
Zr(IV) (g/L)
Or. phase
Distribution
coefficient (D)
5.50 1 4.70 0.827 0.452 0.547
6.10 1 5.24 0.641 0.635 0.990
7.03 1 6.22 0.421 0.854 2.030
7.95 1 7.26 0.259 0.994 3.840
Results from Table 4 and Figure 7 show that the distribution coefficient values of Zr(IV) increase
when the HNO 3 concentration increases, being the distribution coefficient higher than 1.0 for HNO 3
concentration higher than 6 M and close to 4.0 for acid concentrations of 8 M. This behaviour could be
useful for a possible separation of 93 Zr from the raffinate of the first cycle of PUREX process, in
which it has been demonstrated that Zr remains totally.
623

REFERENCES
[1] I.I. Nazarenko, A.M. Yermakov, Analytical Chemistry of Selenium and Tellurium, Analytical
Chemistry of the Elements Series, 1971, Russian Academy of Sciences Publ., 1-32.
[2] W.B. Blumenthal, The Chemical Behaviour of Zirconium, 1958, Van Nostrand, Princenton N.J.
[3] G.F. Egorov, Solvates of Zirconium and Hafnium Nitrates with TBP, Russ. J. Inorg. Chem., 1960,
5, 503-505.
[4] A.E. Levitt, H. Freund, Extraction of Zirconium by TBP, J. Am. Chem. Soc., 1956, 78(8),
1545-1549.
[5] O.A. Sineergibrova, G.A. Yagodin, Zirconium-hafnium Separation, 1966, At. Energy Rev., 4, 93-99.
[6] A.S. Solovkin, Thermodynamics of the Extraction of Zirconium, Present in the Monomeric
State, From HNO 3 Solutions by TBP, Russ. J. Inorg. Chem., 1970, 15, 983-984.
[7] ENRESA-2000, Elemento Combustible de Referencia Irradiado: Inventario Másico y Radiactivo,
Potencia Térmica Residual, Espectro Fotónico e Inventario Radiotóxico, 1999 Clave
49-1PP-L-02-02.
[8] Y. Koma, T. Koyama, Y. Tomaka, Recovery of Minor Actinides in Spent Fuel Reprocessing
Based on PUREX Process, RECOD’98, Niza, 1998, I, 409-416.
[9] O. Courson, R. Malmbeck et al., Separation of Minor Actinides from Genuine HLLW Using the
Diamex Process, 5th International Information Exchange Meeting, Mol, 1998, EUR 18898 EN,
OECD/NEA (Nuclear Energy Agency), Paris, France, 1999.
624

SOLUBILIZATION STUDIES OF RARE
EARTH OXIDES AND OXOHALIDES. APPLICATION OF
ELECTROCHEMICAL TECHNIQUES IN PYROCHEMICAL PROCESSES
C. Caravaca, P. Díaz Arocas, J.A. Serrano, C. González
CIEMAT, Dpto. de Fisión Nuclear, Avda. Complutense, 22, Madrid 28040, Spain
E-mail: c.caravaca@ciemat.es
R. Bermejo, M. Vega, A. Martínez and Y. Castrillejo
Universidad de Valladolid, Dpto Química Analítica, F. de Ciencias,
Prado de la Magdalena s/n, 47005 Valladolid, Spain.
E-mail: ycastril@qa.uva.es
Abstract
Chemical and electrochemical properties of rare earths (La, Ce, Pr and Y) chloride solutions in the
eutectic LiCl-KCl and the equimolar CaCl 2 -NaCl mixture were studied at 450 and 550 0 C respectively.
The stability of the oxidation states of rare-earths and the standard potential of the different redox
couples have been determined. The solubility product of oxides and oxychlorides were calculated, the
differences observed on pKs values between the two molten media demonstrate the different oxoacidic
properties of both molten baths. All these data have been summarised in E-pO 2- diagrams which
displays the stability domains of rare earth compounds on each melt.
Gaseous HCl was used as chlorinating agent during the solubilization tests of the corresponding rare
earth oxides and oxychlorides, efficiencies close to 100%.
The electrochemical behaviour of rare earth solutions has been studied at W and Mo electrodes using
different electochemical techniques, observing that Me electrodeposition could be complicated by
alkaline co-deposition (Li or Na). Mass transport towards the electrode is a simple diffusion process,
and the diffusion coefficients of Me(III) were obtained. In LiCl-KCl, nucleation and crystal growth of
the rare earth metal seems to be the controlling step in most cases, while in CaCl 2 -NaCl this
phenomenon has not been observed.
625

1. Introduction
The long-term radiological hazard of spent nuclear fuel is determined by the transuranium
elements (TRU: Pu, Am, Cm, Np) and some long-lived fission products (LLFP: Cs, I, Tc). If these
elements could be separated efficiently from the spent fuel and be transformed into short-lived or
stable ones, a significant positive on the overall performance repository will be achieved [1].
Over the last years, a renewed interest on pyrochemical separation processes in molten salt media
has been shown mainly due to the progress in the assessment of new concepts for transmutation and
the corresponding fuel cycles [2]. Pyrochemical processes are considered to have potential advantages
over aqueous processes to reduce the inventory of actinides and long lived fission products in the
nuclear wastes. In order to assess its feasibility, several processes have been developed for the
recovery of actinides from spent metallic, nitride, oxide nuclear fuels, and high level radioactive liquid
wastes [3].
Some of the main advantages of the pyrometallurgical process are that the purity of the product is
less stringent, the recovery of minor actinides takes place simultaneously with plutonium due to the
thermodynamic properties in molten salt. The recovery of minor actinides allow the reduction of TRU.
On the other hand, the radiation stability of molten salt enables to process spent fuels of high
radioactivity without increasing the secondary waste, and since molten salt does not act as neutron
moderator comparatively higher amount of fissile material can be handled in the processes equipment
than in the aqueous processes [4].
Since the separation behaviour of actinides and rare-earths is essential for designing the
pyrochemical processes, much effort has been made to study actinide and rare-earth chemistry in
molten salt media in order to have a reliable data base [3]. As a part of a wider UE project that is
focused on separation of actinides from LLFP from oxide nuclear fuels, the present work presents a
study of the chemical and electrochemical properties of several rare-earths in two different molten
chloride baths.
Separation prediction can be made from thermodynamic data by means of the so called
generalised Pourbaix type diagrams (GPTD), E-pO 2 , for rare earth (i.e. La, Ce, Pr and Y)-O
compounds and the chlorinating gaseous mixtures in two molten chloride mixtures of different
intrinsic acidities, the LiCl-KCl eutectic melt and the CaCl 2 -NaCl equimolar mixture, which enables to
propose the main lines for the pyrochemical separation process. Thermodynamic data of metal oxides
are often available in the literature but unfortunately, in most cases, data for most of the pure metal
oxychlorides are not available. Therefore, the stability of these species has to be experimentally
determined by using specific tools such as the yttria-stabilized zirconia membrane electrode (ysze).
Electrochemical techniques provide an efficient tool to investigate the reaction mechanisms. The
establishment of modern electrochemical technologies requires early engineering evaluation of the cell
behaviour with reliable procedures. To reach a better view of the feasibility of the process, the kinetic
parameters of the reaction steps are measured from transient techniques taking into account the
diffusion’s contribution of electroactive species, electron transfer, kinetics and additionally adsorption
or crystallisation, the last one generally is controlled by the rate of nucleus formation and the diffusion
of active species.
626

2. Experiment details
Cyclic voltammetry and other pulse techniques were performed. The working electrodes (WE)
used were tungsten or molybdenum wires of 1mm diameter, as counter electrode tungsten was used.
The active surface area of the WE was determined by measuring the depth of immersion.
The reference electrode consisted of a silver wire (1 mm diameter) dipped into a silver chloride
solution (0.75 molKg -1 ) in the CaCl 2 -NaCl or LiCl-KCl molten mixture contained in a quartz tube.
Potentials were measured by reference to the AgCl/Ag couple.
The pO 2- indicator electrode used is a tube of yttria-stabilised zirconia, filled with molten
CaCl 2 -NaCl or LiCl-KCl and oxide and silver ions (3 × 10 -2 and 0.75 molkg -1 respectively) in this
mixture a silver wire was also immersed (inner reference Ag + /Ag).
The chloride mixtures CaCl 2 -NaCl or LiCl-KCl (analytical-grade) were melted under vacuum,
next raised to atmospheric pressure using dry argon, and then it was purified by bubbling HCl through
the melt for at least 30 minutes, and then kept under argon atmosphere [5-7]. Working temperature
was measured by a thermocouple introduced into the melt. Salt handling was carried out in a glove
box under argon atmosphere.
Solutions of the electroactive species was prepared by direct addition of MeCl 3 . In order to remove
any traces of oxide ions, the solutions were purified by gaseous HCl bubbling.
2.1 pK s
determinations
pKs determinations were performed using a ysze electrode placed into the melt. A known amount of
metal ion introduced as chloride was potentiometrically titrated either with Na 2 CO 3 or BaO. Continuous
stirring with dried argon was required.
3. Results and discussion
3.1 Determination of the stable oxidation states of rare earths
The electrochemical properties of dilute solutions of rare earth ions (Ce(III), La(III), Pr(III) and
Y(III)) in the equimolar CaCl 2 -NaCl and eutectic LiCl-KCl, at 550 and 450 o C respectively, were
studied.
Figure 1 shows voltammograms obtained with a solution of CeCl 3 in the eutectic LiCl-KCl melt
using a tungsten electrode. The process is characterised by one cathodic peak well defined, associated
with the corresponding sharp re-oxidation peak, (anodic dissolution), which is characteristic of the
formation of a product that remains adhered to the electrode. This fact has been confirmed by
examination of the voltammograms, the ratio of the anodic to the cathodic current was higher than
unity, the ratio between the total anodic to cathodic charge was close to unity and independent of the
scan rate (Figure 2). In the anodic region, there was not observed the electrochemical system
Ce(IV)/Ce(III), in none of the melts, which indicate that the standard potential of that system is out the
range accessible in these melts, and that Ce(IV) is a powerful oxidizing agent which oxidizes the
chloride ions of the melt according to the reaction:
Ce(IV) + Cl - ↔Ce(III) +1/2 Cl 2 (g) (1)
627

Ce Ce 3+ Cl - Cl 2
The voltammograms obtained for other rare earth trichlorides solutions,in both chloride melts,
behaved in a similar way except for a positive or negative shift of the peak potential value compared to
that of the CeCl 3 . Moreover, it has been observed that the peak potential values obtained in the melt
CaCl 2 -NaCl are slightly less cathodic than those obtained in the eutectic LiCl-KCl.
Figure 1. Cyclic voltammograms of LiCl-KCl with
CeCl 3 (1.57.10 -4 mol/cm 3 ), W E : tungsten 0.28 cm 2
0,2
0,15
Li Li +
0,1
I/A
0,05
0
-0,05
Ce 3+ Ce
-0,1
Li + Li
-0,15
-3 -2 -1 0 1 2
E/V
Figure 2 (a). Voltammogram that proves the formation of a solid product at the tungsten
surface. Reduction step Me(III)+3e ⇔Me and the subsequent anodic dissolution of the deposit
0,06
0,04
Ce Ce 3+
(a)
0,02
I/A
0
-0,02
Ce 3+ Ce
-0,04
-0,06
-2,4 -2,2 -2 -1,8 -1,6 -1,4 -1,2 -1
E/V
628

Figure 2 (b). Comparison between the cathodic and anodic charge
for the deposition and subsequent reoxidation of solid cerium
0,06
0,04
(b)
0,02
0
I/A
-0,02
-0,04
Qc
Qa
-0,06
-0,08
9 9,5 10 10,5 11 11,5
t/s
Square wave voltammetry (SWV) to study the stable oxidation states of rare earths in both melts was
used. According to Baker et al. [8] and Osteryoung et al. [9] the width of the half-peak, W 1/2 , depends on
the number of electrons exchanged and on the temperature as follows:
W 1 / 2 =
3.52
RT
nF
(2)
The n-values obtained with all the MeCl 3 solutions were close to 3.
Similar results were obtained by chronopotentiometry. This type of transients show the existence
of a potential plateau in the range –2,20 V (v.s. Ag + /Ag). After this plateau a rapid decrease was
observed, and then, the electrode potential reaches a limiting value corresponding to the deposition of
alkaline metals (Figure 3).
The reduction of rare earth trichlorides (La(III), Ce(III), Pr(III) and Y(III)) to metal takes place in
a single step, according to the reaction:
Me(III) + 3e ↔ Me(s) (3)
629

Figure 3. Chronopotentiograms for the reaction Pr 3+ + 3e↔ Pr in LiCl-KCl
-1,9
-2
-2,1
-2,2
E/V -2,3
-2,4
-2,5
-2,6
-2,7
0 1 2 3 4 5
t/s
3.2 Standard potential of Me(III)/Me(0), and activity coefficient γ(MeCl 3
) determination
The standard potential of the redox couples Me(III)/Me was determined by measuring the
equilibrium potential of a tungsten wire covered with an electrodeposit of Me(0), obtained by
coulometry at a constant potential value, avoiding any alkaline deposition.
For each rare earth the e.m.f values were measured for several metal chloride concentrations.
Based on these measurements the variations of the e.m.f. is given by the Nernst equation. The plots of
the e.m.f. versus the logarithm of the Me(III) concentration were linear with slopes in agreement with
the theoretical values for a three-electron process. The standard potentials were deduced from linear
extrapolation of the plots at a MeCl 3 concentration equal to 1 mol/kg, (see Table 1). The apparent
standard potentials are very close, and in the order:
• Eutectic LiCl-KCl: La > Ce > Y, Pr.
• Equimolar CaCl 2 -NaCl: La > Ce > Pr.
According to Equation (4) the standard potentials can also be deduced from the peak potential
values of the voltammetric reduction of Me(III) [10].
E
p
= E
o
+
RT
2.3
nF
log C
o
− 0.849
RT
nF
(4)
This equation is valid for conditions where the electrochemical reaction is diffusion controlled. The
E 0 values derived, (see Table 1, c.v. values), were several mV more negative than those obtained by
e.m.f. measurements. This is probably due to the quasi-reversible behaviour of the electrochemical
system, and/or to the influence of nucleation and crystal growth phenomena, since Equation (4) does
not take into account any of those phenomena.
630

Activity coefficients of MeCl 3 in the melts, γMeCl3 which take into account the free enthalpy of
formation of MeCl 3 (s) and of solvated MeCl 3 (dissolved) were calculated by the ∆E corresponding to
the reaction:
Me +3/2 Cl 2 ↔ MeCl 3 (dissolved) (5)
and related with the ∆E* of the same reaction between pure compounds, by means of the equation:
log γ MeCl3 = (∆E*-∆E)3F/2.3RT (6)
∆E* was derived from the literature [11] and ∆E from previously recorded experimental data. The
values obtained are given in Table 1.
Table 1. Standard potential values and activity coefficient
of some rare-earths chlorides in LiCl-KCl and CaCl 2 -NaCl
LiCl-KCl
CaCl 2 -NaCl
Redox couple
Ce(III)/Ce
La(III)/La
Pr(III)/Pr
Y(III)/Y
* Preliminary results.
Standard Potential/V
Molality scale
e.m.f.-3.155
c.v. .-3.201
e.m.f.-3.160
c.v. -3.254
e.m.f.-3.150*
c.v. -2.985*
e.m.f.-3.152*
c.v. –3.305
∆E*/V log γ MCl3 Standard Potential/V
Molality sacale
3.034 -2.53 e.m.f.-3.034
c.v. .-3.074
3.100 -1.26 e.m.f.-3.138
c.v.-3.174
3.032 -2.47* e.m.f.-3.020*
c.v. –3.007*
2.774 -7.91* –
c.v. –3.023*
∆E*/V log
γ MCl3
2.949 -1.56
3.023 -2.11
2.952 -1.25*
2.698 –
The activity coefficient of rare earth chlorides gives information about the complexation of the
cations by the melt. The complexation power depends both on the nature of the cations of the molten
salts and on the working temperature. In the melt CaCl 2 -NaCl at 550°C, which is a more oxoacidic
than the eutectic LiCl-KCl, the activity coefficients values obtained are higher than in LiCl-KCl,
except for lanthanum. This corresponds to a lower complexation of the MeCl 3 by the chloride ions of
the melt, which indicates more stable complex ions formation.
3.3 Stability of Me(III)-O compounds. pk s
determination
Identification of the rare earth oxides and oxohalides that are stable in both melts, can be
accomplished by the theoretical analysis of the curves obtained by potentiometric titration [12,13].
The solubility of the oxides and oxyhalides can be determined theoretically from the analysis of the
experimental titration curves.
631

Figure 4. Potentiometric titrations of (a) 0.0797 mol kg -1 Pr(III) and
(b) 0.1040 Y(III) solutions by O 2- ions added as solid Na 2 CO 3 .in the eutectic LiCl-KCl at 450°C.
8
8
7
(a)
7
(b)
6
6
5
5
4
4
pO 2- 3
α=1
pO 2- α =1,5
3
2
2
1
1
0
0 0,5 1 1,5 2
α
0
0 0,5 1 1,5 2 2,5 3
α
The rare-earth ions were precipitated with oxide (added as Na 2 CO 3 or BaO), this reaction was
monitored with an ysze [7,14-16]. A e.m.f. jump occurrs at the point corresponding to the
stoichiometric precipitation of the corresponding oxide or oxohalide. Except for the YCl 3 all the
experimental curves obtained with lanthanide ions exhibited similar habits to the one of Figure 4.a.
The pO 2- values measured by the ysze (after calibration), show only one equivalent point for a
stoichiometric ratio:
α =
2
[ O ]
[ LnCl ]
− (7)
3
added
= 1.0
initial
This indicates that the reaction is:
Ln 3+ + O 2- + Cl - ↔ LnOCl(s) (8)
For the YCl 3 solutions titrated in the eutectic LiCl-KCl by O 2- (Figure 4 (b)), only one equivalence
point was observed for a stoichiometric ratio of α=1.5.
This value indicates that the reaction is in this case:
2Y 3+ + 3O 2- ⇔Y 2 O 3 (s) (9)
and it can be deduced that no oxychloride species were stable under the present experimental
conditions. The LnOCl and Y 2 O 3 formation was confirmed by XRD spectrometry analysis.
The theoretical equation corresponding to the titration curve can be elucidate from the mass
balance and the solubility product of the reactions, according to the procedure previously described
[7,14,15]. The solubility products of LnOCl and Y 2 O 3 , k s, were determined by applying the
Gauss-Newton non-linear least-squares method to the equation of the corresponding titration curves.
The values obtained are shown in Table 2.
632

Table 2. Solubility products, pk s of the different oxychlorides and oxides
Compounds LiCl-KCl (450°C) CaCl 2 -NaCl (550°C)
CeOCl 7.45 ± 0.05 5.62 ± 0.07
LaOCl 7.00 ± 0.09 5.19 ± 0.05
PrOCl 7.45 ± 0.25 5.27
Y 2 O 3 19.90 ± 0.22 –
3.4 Solubilization studies
With the solubility products of the Ln-O compounds and the equilibrium potentials of the
different red-ox couples involved, it is possible to establish the potential-acidity diagram for the Ln-O
species in both melts (Figure 5).
Figure 5. Comparison of the potential-acidity diagram for the Ce-O compounds with the E-pO 2-
diagram of the gaseous mixtures in the equimolar CaCl 2 -NaCl mixture at 550ºC. MIXTURES: I:
Cl 2 + O 2 , II: Cl 2 + C, III: Cl 2 + CO, IV: HCl + H 2 O, V: HCl + H 2 O + H 2 , VI: HCl + H 2 + CO
0,5
0
-0,5
O 2- (1atm)/O 2-
I
CeO 2
IV
Cl 2(g)
II
III
-1
-1,5
/V
-2
CeOCl
V
VI
CeCl 3(g)
-2,5
-3
-3,5
-4
Na(liq)
0 5 10 15 20 25 30
Ce
pO 2-
This type of diagrams gives the oxo-acidity and red-ox properties of the elements in the molten salt
mixtures, being possible to predict electrochemical properties and some chemical reactions. The
comparison of these diagrams to the those obtained for some gaseous mixtures in the same melt and
temperature [6,14,15], allows to predict optimal chlorinating conditions for rare earth oxides and
oxychlorides. It is observed in the Figure that all the LnOCl, CeO 2 and Y 2 O 3 can be chlorinated by HCl.
633

Experimental solubilization tests were carried out: i) with samples of LnOCl generated in situ, and
ii) samples of commercial La 2 O 3 , CeO 2 , Pr 6 O 11 and Y 2 O 3 . As chlorinating agent HCl was used, and the
reaction progress was followed by potentiometry with an ysze. After the dissolution an electrochemical
spectra was recorded.
During the chlorination, the oxide ions are transformed into water by HCl, according with the
following reactions:
2HCl + LnOCl ↔ LnCl 3 (disolved) + H 2 O (10)
6HCl + Y 2 O 3 ↔ 2YCl 3 (disolved) + 3 H 2 O (11)
For CeO 2 and Pr 6 O 11 , the chlorination could occur in two stages, in the first one it is formed the
oxidizing Ce(IV) which reacts with the melt evolving Cl 2 (g) and dissolving CeCl 3 , and the insoluble
PrOCl respectively.
The concentration of the dissolved rare earths was determined in situ by titration of the final
solution with sodium carbonate, showing efficiency values close to 100%.
3.5 Metal electrodeposition studies
The mechanism of electroreduction of rare-earth ions in the equimolar CaCl 2 -NaCl and eutectic
LiCl-KCl has been studied by electrochemical techniques. Previous experiments showed that
refractory metals such as tungsten or molybdenum are suitable materials to use as electrodes in both
melts. It is not possible to use glassy carbon due to the formation of Na-C or Li-C compounds [14].
The diffusion coefficient of the Me(III) ions could be calculated from the data obtained in the rare
earth electrodeposition studies (Table 3).
Table 3. Me(III) diffusion coefficients
LiCl-KCl (450ºC)
CaCl 2 -NaCl (550ºC)
W Mo W Mo
Ce(III) 1.0 × 10 -5 1.1 × 10 -5 9.2 × 10 -6 8.8 × 10 -6
La(III) 8.9 × 10 -6 8.4 × 10 -6 7.7 × 10 -6 7.8 × 10 -6
Pr(III) 9.4 × 10 -6 – 9.8 × 10 -6 –
Y(III) 1.5 × 10 -5 1.3 × 10 -5 1.3 × 10 -5 –
The diffusion coefficient values obtained in both media were similar. These results can be
explained considering the opposite effect of the temperature and viscosity of the melt: The higher
temperature in the case of the molten CaCl 2 -NaCl mixture should produce an increase in the diffusion
coefficient values, however, its higher viscosity leads to lower D values.
Chonoamperometric studies did not show evidence of nucleation phenomena in the equimolar
CaCl 2 -NaCl mixture under the experimental conditions tested. However, the I-t transients obtained in
the eutectic LiCl-KCl have proved that nucleation of metallic lanthanum and yttrium plays a
significant role in the overall electrodeposition process. Differences between the results obtained in
both molten chlorides could be related to the differences on surface tension of the melts, which affect
634

the interactions Me-substrate and Me-substrate-melt. Information about nucleation kinetics was
obtained applying a dimensionless method.
The efficiencies in the rare earth electrodeposition processes were calculated from double potential
step measurements, for several potential impossed. The results show that alkaline metal co-deposition
can interfere with the metal electrodeposition, complicating thus the process.
Acknowledgements
The authors would like to thank Francisco de la Rosa (Universidad de Valladolid) and
Luis Gutierrez (CIEMAT) for the technical assistance and ENRESA (Spain) for the financial support
(CIEMAT-ENRESA and CIEMAT-Universidad de Valladolid agreements).
REFERENCES
[1] H. Gruppelaar, J.L. Kloosterman and R.J.M. Konings, Advanced Technologies for the Reduction
on Nuclear Waste, ECN report, 1998.
[2] OECD Nuclear Energy Agency, Actinide and Fission Product Partitioning and Transmutation.
Status and Assessment Report, Paris, France, 1999.
[3] Y. Sakamura, T. Inoue, O. Shirai, T. Iwai, Y. Arai and Y. Suzuki, Proceedings of Global’99,
Jackson Hole, Wyoming, USA, 1999.
[4] T. Koyama, K. Kinoshita, T. Inoue, M. Ougier, J.P. Glatz and L. Koch, Workshop on
Pyrochemical Separation OECD/NEA, Work 3b-4, Avignon, France, March 2000.
[5] D. Ferry, Y. Castrillejo, G. Picard, Acidity and Purification of the Molten Zinc Chloride
(33.4 mol%)-sodium Chloride (66.6 mol%) Mixture, Electrochim. Acta, 1989, 34(3), 313-316.
[6] F. Seón, Tesis Doctoral de Estado, París, 1981
[7] Y. Castrillejo, A.M. Martínez,, G.M. Haarberg, B. B •rresen, K.S. Osen and R. Tunold,
Oxoacidity Reactions in Equimolar Molten CaCl 2 -NaCl Mixture at 550°C, Electrochim. Acta,
1997, 42, 1489-1494.
[8] G.C. Baker, Anal. Chim. Acta, 1958, 18, 118.
[9] J. Osteryoung, J.J. O’Dea, Determining Kinetic Parameters from Pulse Voltammetric Data,
Electroanal. Chem., 1986, 14, 209.
[10] A.J. Bard, L.R. Faulkner, Electrochemical Methods, J. Wiley New York, USA, 1980.
[11] J. Barin, O. Knacke, Thermochemical Properties of Inorganic Substances, Springer, Berlin,
Germany, 1973.
[12] H. Lux, Z. Elektrochemi., (1948), 52, 220, (1949), 53, 41.
635

[13] H. Flood T. Förland and K. Motzfeld, The Acidic and Basic Properties of Oxides, Acta
Chemica. Scandinavica., 1952, 6, 257.
[14] A.M. Martínez, Y. Castrillejo, E. Barrado, G.M. Haarberg and G Picard, A Chemical and
Electrochemical Study of Titanium Ions in the Molten Equimolar CaCl2-NaCl Mixtures at
550°C, J. Electroanal. Chem., 1998, 449, 67-80.
[15] Y. Castrillejo, M.R. Bermejo, A.M. Martínez, C. Abejón, S. Sánchez and G.S. Picard, Study of
the Electrochemical Behaviour of Indium Ions in the Molten Equimolar Cacl 2 -NaCl Mixture at
550°C, J. of Applied Electrochemistry, 29(1), 65-73, 1999.
636

CALIX[6]ARENES FUNCTIONALISED WITH MALONDIAMIDES AT THE UPPER RIM
AS POSSIBLE EXTRACTANTS FOR LANTHANIDE AND ACTINIDE CATIONS
Marta Almaraz, Sagrario Esperanza, Oriol Magrans, Javier de Mendoza, Pilar Prados
Departamento de Química Orgánica, Facultad de Ciencias,
Universidad Autónoma de Madrid, Cantoblanco, 28029-Madrid, Spain
Abstract
Lipophilic malondiamides have been recently employed successfully as extractants for lanthanide and
actinide cations from strongly acidic media. Many complexes between malondiamides and
lanthanide-actinides cations have been studied by different techniques. For many of these complexes it
has been observed that more than one malondiamide ligand participates in the complexation of each
metallic cation. Incorporation of two or three malondiamide moieties into a calixarene platform would
probably improve both extraction and selectivity with respect to the already tested malondiamides.
According to CPK examination, a calix[6]arene substituted at the upper rim with two or three
malondiamide moieties should constitute a promising ligand for lanthanide and actinide cations due to
co-operative complexation with the malondiamides. Based on these considerations, we synthesised
calix[6]arenes functionalised with malonic acid derivatives.
637

1. Introduction
The separation of lanthanides and specially actinides from the nuclear fuel and the transmutation
of long-lived isotopes to short lived ones are very important for the reprocessing of spent nuclear fuel.
The DIAMEX process in which malondiamides are used as extractants is one of the most
promising one because these kind of extractants are well-suited compounds to extract trivalent
actinides from nitric acid solutions. On the other hand, they are completely incinerable and have very
low water solubility.
Lipophilic malondiamides have been recently employed successfully as extractants for lanthanide
and actinide cations from strongly acid media. It has been observed that more than one malondiamide
ligands participates in the complexation of each metallic cation [1]. For that reason it was decided to
synthesize calix[6]arenes functionalised with malonic acids derivatives.
2. Malondiamide calix[6]arenes
As starting materials for this proposal, a variety of amines on the calixarenes were used with
different acyl chlorides.
Amines 1 and 2 were synthesised starting from p-tert-butyl calix[6]arene [2], as described in
Figure 1, by successive selective alkylation, nitration, total alkylation of the platform and finally
reduction to the desired amine [3].
Figure 1. Synthesis of amines 1 and 2
t-Bu
t-Bu t-Bu t-Bu
t-Bu t-Bu t-Bu
i) ii) iii)
t-Bu t-Bu t-Bu
OH 6
O OH OH
O OMe OMe
OH OMe OMe
2 2 2
iv)
NH 2 t-Bu t-Bu
vi)
NO 2 t-Bu t-Bu
v)
NO 2 t-Bu t-Bu
OMe OMe OMe
2
OMe OMe OMe
2
OH OMe OMe
2
1
t-Bu
t-Bu
t-Bu
NO 2
vii) viii) ix) x)
t-Bu
NO 2
t-Bu
NH 2
t-Bu
OH 6
OH
i) Me 3 SiOK, BrCH 2 C 6 H 4 CH 3 .
ii) NaH, Me 2 SO 4 .
iii) H 2 , Pd/C.
iv) HNO 3 /H 2 SO 4 .
v) NaH/Me 2 SO 4 .
vi) H 2 /PtO 2 .
vii) Na 2 CO 3 /IMe.
viii) HNO 3 /H 2 SO 4 .
ix) K 2 CO 3 /Me 2 SO 4 , x) H 2 /PtO 2 .
OMe
OH
OMe
OMe
3 3 3 3
OMe
OMe
2
OMe
638

These compounds were reacted with acyl chlorides 3, 4 and 5. Compound 3 is commercially
available and compounds 4 and 5 were obtained from commercially available di-esters by
monohydrolysis with MeOH/H 2 O/KOH followed by treatment of the resulting carboxylic acids with
Cl 2 SO (Figure 2).
Figure 2. Formula 3 and synthesis of compounds 4 and 5
O
O
Cl
OMe
Commercially available
3
O
O O O
MeO OR Cl OR
R 1
R 1 R 2
R 2
i) KOH,
ii) Cl 2 SO
i)
ii)
4: R=R 1 =Et, R 2 =H (88%)
5: R=Me, R 1 =R 2 =Et (86%)
The reaction of acyl chlorides 3, 4 and 5 with the amines 1 and 2 in CH 2 Cl 2 in the presence of
Et 3 N at room temperature gave the ester derivatives 6-10 (Figure 3).
Figure 3. Synthesis of esters 6-10
NH 2 t-Bu t-Bu
Cl
O O
OR
R 1 R 2
O
O
OR
R 2
HN R 1 t-Bu t-Bu
OMe OMe OMe
1
t-Bu NH 2
2
Cl
3: R=Me, R 1 =R 2 =H
4: R=R 1 =Et, R 2 =H
5: R=Me, R 1 =R 2 =Et
i)
O O
OR
R 1 R 2
t-Bu
OMe OMe OMe
6: R=Me, R 1 =R 2 =H (71%)
7: R=R 1 =Et, R 2 =H (51%)
8: R=Me, R 1 =R 2 =Et (36%)
O
HN
O
OR
R 1
R 2
2
OMe
2
OMe
3
3: R=Me, R 1 =R 2 =H
4: R=R 1 =Et, R 2 =H
i)
OMe
OMe
9: R=Me, R 1 =R 2 =H (23%)
10: R=R 1 =Et, R 2 =H (57%)
3
i) Et 3 N, CH 2 Cl 2 , r.t.
Aminolysis of these compounds with butylamine at reflux temperature gave malonamide
derivatives 11-15, as described in Figure 4.
639

Figure 4. Synthesis of amides 11-15
O
O
OR
HN
R 2
R 1 t-Bu t-Bu
O
O
NHBu
HN
R 2
R 1 t-Bu t-Bu
OMe OMe OMe
6: R=Me, R 1 =R 2 =H
7: R=R 1 =Et, R 2 =H
8: R=Me, R 1 =R 2 =Et
t-Bu
O
HN
O
OR
R 1
R 2
3
2
i)
i)
OMe OMe OMe
2
11: R 1 =R 2 =H (65%)
12: R 1 =Et, R 2 =H (82%)
13: R 1 =R 2 =Et (85%)
O
NHBu
O
R 2
t-Bu HN R 1
i) NH 2 Bu/∆
OMe
OMe
9: R=Me, R 1 =R 2 =H
10: R=R 1 =Et, R 2 =H
OMe
OMe
14: R 1 =R 2 =H (70%)
15: R 1 =Et, R 2 =H (32%)
3
These compounds are currently being tested in lanthanide-actinide extraction.
Acknowledgements
The work presented in this paper was supported by the Comunidad Autonoma de Madrid and by
the European Union (contract F14W6CT 960022).
REFERENCES
[1] a) G.Y.S. Chan, M.C.B. Drew, M.J. Hudson, P.B. Iveson, J.O. Liljenzin, M. Skalberg, L. Spjuth,
C. Madic, Solvent Extraction of Metal Ions From Nitric Acid Solution Using N,N´-substituted
Malonamides. Experimental and Crystallographic Evidence for Two Mechanism of Extraction,
Metal Complexation and Ion-pair Formation, J. Chem. Soc., Dalton Trans., 1997, 649-660.
b) P.B. Iveson, M.G.B. Drew, M.J. Hudson, C. Madic, Structural Studies of Lanthanide
Complexes With New Hydrophobic Malonamide Solvent Extraction Agents, J. Chem. Soc., Dalton
Trans., 1999, 3605-3610.
[2] C.D. Gutsche, B. Dhawan, M. Leonis, D. Steward, p-tert-butyl Calix[6]arene, Organic Synthesis,
1989, 238.
[3] A. Casnati, L. Domiano, A. Pochini, R. Ungaro, M. Carramolino, J.O. Magrans, P.M. Nieto,
J. Lopéz-Prados, P. Prados, J. de Mendoza, R.G. Janssen, W. Verboom, D.N. Reinhoudt,
Synthesis of Calix[6]arenes Partially Functionalised at the Upper Rim, Tetrahedron, 1995, 51,
12699-12720.
640

ACTINIDE(III)/LANTHANIDE(III) PARTITIONING USING n-Pr-BTP AS EXTRACTANT:
EXTRACTION KINETICS AND EXTRACTION TEST IN A HOLLOW FIBER MODULE
Andreas Geist, Michael Weigl, Udo Müllich, Klaus Gompper
Forschungszentrum Karlsruhe GmbH, Institut für Nukleare Entsorgung
POB 3640, 76021 Karlsruhe, Germany
E-mail: geist@ine.fzk.de
Abstract
2,6-di(5,6-dipropyl-1,2,4-triazin-3-yl)pyridine (n-Pr-BTP), first developed in our laboratory, is a very
promising extractant for the effective separation of actinides(III) from lanthanides(III). It is able to
extract actinides(III) with usable distribution coefficient from 1-2 molar nitric acid selectively over
lanthanides(III). The Am(III)/Eu(III) separation factor is approx. 135. The performance of this
extractant is further elucidated by kinetic investigations and a counter-current extraction experiment in
a hollow fiber module (HFM). The kinetic investigations were performed in a stirred cell. The fact that
extraction rate is independent of stirring speed reveals a slow chemical complexation reaction. The
HFM extraction test on americium(III)/fission lanthanides(III) separation showed good hydrodynamic
behavior. Depending on aqueous flow rate, which was varied, up to 94% americium could be removed
from the feed phase. Lanthanide co-extraction was in the range of 1%.
641

1. Introduction
Over the recent years, efforts have been made finding extractants capable of separating trivalent
actinides from fission lanthanides. This separation task is a key step in the separation of minor
actinides from high-level Purex effluents within the P&T strategy [1]. Ideally, such an extractant
would selectively extract trivalent actinides over lanthanides from moderately acidic media without
the need for pH adjustment or the use of salting-out agents. Furthermore, it would consists only of
C, H, O, and N atoms (“CHON principle”), making it fully combustible to gaseous products.
2,6-di(5,6-dipropyl-1,2,4-triazin-3-yl)pyridine (n-Pr-BTP), which was developed in our
laboratory [2,3], is an extractant capable of fulfilling this task. It is able to selectively extract
actinides(III) over lanthanides(III) from 1-2 molar nitric acid with usable distribution coefficient.
Am(III)/Eu(III) separation factor is approx. 135. The extractant complies to the CHON principle.
Figure 1. 2,6-di(5,6-dipropyl-1,2,4-triazin-3-yl)pyridine (n-Pr-BTP)
N N N
N
N
N
N
In shaking tubes, equilibrium is not attained very rapidly. We feel that this indicates a slow
chemical complexation reaction. To further elucidate the kinetics of extraction, we performed
experiments on Am(III) extraction in a stirred cell built specifically for our purposes [4].
A constant-interface, Lewis-type stirred cell is the best tool for extraction kinetics studies, if it is
calibrated [5,6]. It allows the discrimination of flow-dependent transport processes (diffusional
regime) from flow-independent interfacial reactions (chemical regime) as rate determining alternatives
by measuring extraction rates at varied stirring speeds. The cell we used in this work is tested on a
physical mass transfer system (toluene transfer into water), showing a linear dependency of mass
transfer rate on stirring speed over a wide range [4]. Therefore we can be sure that fluxes independent
of stirring speed indicate that an interfacial process (i.e. the chemical reaction) is rate determining.
With n-Pr-BTP, a hot mixer-settler test using a synthetic feed solution and a hot test in a 16-stage
centrifugal extractor using genuine DIAMEX raffinate for feed solution were performed in other
laboratories [7]. These tests were very successful. Americium(III) and curium(III) could quantitatively
be extracted (

2. Experimental
2.1 Synthesis and characterisation of n-Pr-BTP
We prepared two batches of 40-50 g n-Pr-BTP each as described in [3]. Melting points were in
the range of 105-107°C. NMR data confirmed the products’ identities. To further characterise the
products, we performed distribution measurements: Contacting a solution of 0.04 M n-Pr-BTP in TPH
(a kerosene-type diluent) modified with 1-octanol (70:30 vol.) and 1 M nitric acid labelled with 241 Am
or 152 Eu, we found an americium distribution coefficient of approx. 14 and a separation factor,
SF Am(III)/Eu(III) of 135. This is in good agreement with results published elsewhere [7].
2.2 Extraction kinetics
Kinetic measurements were performed in our special small stirred cell. Its half cell volume is only
60 mL [4]. The aqueous phase was a solution of americium(III) (2 500 Bq/mL) in 1 M nitric acid.
Organic phase was a solution of 0.04 M n-Pr-BTP in TPH/1-octanol (70:30 vol.).
The stirred cell was filled with both aqueous and organic phases, and stirrers were started with
appropriate stirring speeds (aqueous = organic). The activity in the aqueous phase was continuously
detected in a by-pass with a well-type NaI-detector. Measured activity was plotted vs. time and initial
metal fluxes were calculated.
2.3 HFM extraction test
2.3.1 Feed solutions
Aqueous feed contained 241 Am(III), fission lanthanides(III) and yttrium(III). Its composition
corresponds to a DIAMEX raffinate as used in other n-Pr-BTP tests [7], except that 241 Am was used in
trace amount, and Cm, Ru, Pd, Fe were not present. The composition is given in Table 1. The organic
phase was a solution of n-Pr-BTP (0.04 kmol/m 3 , 16.2 g/L) in TPH/1-octanol (70:30 vol.).
Table 1. Composition of aqueous feed solution
HNO 3 1.0 kmol/m 3
Y
89.0 mg/L
La
294 mg/L
Ce
566 mg/L
Pr
264 mg/L
Nd
998 mg/L
Sm
199 mg/L
Eu
35.7 mg/L
Gd
28.2 mg/L
241 Am 0.49 MBq/L
643

2.3.2 Set-up
We set up a single HFM extractor inside a glove box. This corresponds to the extraction stages in
commonly used extractor batteries, without scrubbing or stripping stages. The HFM set-up is shown
schematically in Figure 2. Both phases were passed through the module in counter-current, single-pass
mode, i.e. phases were not recycled. The organic phase flowed in the lumen of the hollow fibers (HF).
The module used was a Celgard LiquiCel Extra-Flow type module (10 000 HF, membrane material
polypropylene, average pore size 0.02 µm, porosity ε = 40%, tortuosity τ = 2.6 [9], HF inner
diameter = 0.24 mm, HF outer diameter = 0.30 mm, active length = 0.15 m). Static pressure in the
aqueous phase was kept approx. 0.5 bar higher than in the organic phase to maintain proper phase
separation. Aqueous flow rate was varied, 0.44 L/h•Q aq •1.75 L/h, organic flow rate was kept constant
at Q org = 0.50 L/h. Further details can be found in [10].
Figure 2. Single-pass, counter-current HFM set-up (schematic). ––– aqueous phase, - - - organic
phase, PI = pressure gauge, FI = flow meter (rotameter), CV = control valve, BP = bypass valve
PI
FI
PI
PI
PI
FI
BV CV
BV CV
2.3.3 Analytic
Samples of aqueous and organic effluents were drawn discontinuously. 241 Am γ activitiy was
determined on a γ counter (Packard Cobra Auto-Gamma). Lanthanide concentrations were measured
with ICP-AES after proper dilution with nitric acid (aqueous samples) or stripping into 0.01 nitric acid
(organic samples).
3. Results
3.2. Extraction kinetics
The plot of normalised initial americium fluxes, j t=0 /[Am 3+ ] t=0 , (Figure 3) characterises the kinetic
behaviour of americium extraction with n-Pr-BTP. The flow-independent fluxes (plateau rates)
indicate that the interfacial reaction is rate determining (i.e. non-equilibrium at the interface).
644

Figure 3. Americium(III) extraction into n-Pr-BTP. Stirring rate dependency of
normalised initial americium fluxes.
[Am 3+ ] = 2500 Bq/mL, [HNO 3 ] = 1.0 kmol/m 3 , [n-Pr-BTP] = 0.04 kmol/m 3 .
8
j t=0
/[Am 3+ ] x 10 6 [m/s]
6
4
2
0
0 100 200 300 400
stirring rate N [min -1 ]
This means that mass transfer is chemically controlled. This case is very interesting as it allows
studying the mechanism of the interfacial reaction. Therefore the dependency of the plateau rate from
the concentrations of all participating species (Am 3+ , H + , NO - 3
, n-Pr-BTP) must be studied. This leads
to the reaction orders of the interfacial reaction for each species. If the reaction orders are known a rate
equation can be expressed.
Some preliminary measurements indicate a first order dependency on the concentration of
n-Pr-BTP. The measurements for the other species are still in progress, therefore a mechanism for the
americium extraction with n-Pr-BTP cannot be postulated yet.
3.2. HFM extraction test
Throughout the experiment, which ran four hours, the hydraulic behaviour was highly
satisfactory, i.e. both aqueous and organic effluents were clear without any entrainment. This is a
benefit of the macroscopic phase separation by the membrane material.
Except with an aqueous flow rate of Q aq = 0.44 L/h, mass balances were (100 ± 2.5)% for
lanthanides, and 90-98% for Am(III). With an aqueous flow rate of 0.44 L/h, Am(III) mass balance
was 66%, lanthanides mass balances were approx. 88%. This is a sign that, at this flow rate, the
experiment was not run sufficiently long to reach steady state.
Aqueous effluent concentrations normalised to feed concentrations, [Me 3+ ] out /[Me 3+ ] in , over
aqueous flow rate, Q aq , are shown in Figure 4. With an aqueous flow rate of 0.44 L/h, 94% of Am(III)
could be removed from the aqueous phase. There is a strong dependency of Am(III) extraction
efficiency on aqueous flow rate, and hence residence time: With an aqueous flow rate of 1.75 L/h,
only 40% of Am(III) could be extracted. The steep flow rate dependency indicates that, at significantly
645

lower flow rates, Am(III) would be extracted almost quantitatively. However, the experimental set-up,
regarding the pumps, control valves and flow indicators installed, was not layed out for such low flow
rates.
Figure 4. Am(III)/Ln(III) separation in a HFM using n-Pr-BTP as extractant. Relative aqueous
effluent concentrations as a function of aqueous flow rate. Aqueous phase,
shell-side: Am(III) + Ln(III) in HNO 3 1.0 kmol/m 3 . Organic phase,
in HF: [n-Pr-BTP] = 0.04 kmol/m 3 in TPH/1-octanol (70:30 vol.). Q org = 0.50 L/h.
100
Ln
[Me 3+ ] out
/[Me 3+ ] in
[%]
10
Am
(non-steady state)
Y
La
Ce
Pr
Nd
Sm
Eu
Gd
Am
1
0.0 0.5 1.0 1.5 2.0
Q aq
[L/h]
Except for the non-steady state result with an aqueous flow rate of 0.44 L/h, no lanthanide
co-extraction was detectable in the aqueous phase. Lanthanide co-extraction as determined from
organic effluent samples was in the range of 1-2% for Sm, Eu, Gd, and well below 0.5% for other
lanthanides. We point out these good results concerning lanthanide co-extraction were realised without
lanthanide scrubbing, due to the high separation factor of n-Pr-BTP.
4. Conclusion
n-Pr-BTP is a very capable extractant for An(III)/Ln(III) separation from acidic media. Although
Am(III) extraction rate is not very fast, results from a HFM extraction experiment are promising.
Operation is stable, lanthanide co-extraction is small. Still better results can be expected from a
modified experimental set-up which can handle lower flow rates. Tests on lanthanide scrubbing and
back extraction will be performed to further evaluate the behaviour of the n-Pr-BTP extraction system
in a hollow fiber module.
Acknowledgements
The authors would like to thank the European Commission for their financial support.
646

REFERENCES
[1] OECD Nuclear Energy Agency, Status and Assessment Report on Actinide and Fission Product
Partitioning and Transmutation, (1999), (www.nea.fr/html/pt/pubdocs.htm).
[2] Z. Kolarik, U. Müllich and F. Gassner, Selective Extraction of Am(III) Over Eu(III) by
2,6-Ditriazolyl- and 2,6-Ditriazinylpyridines, Solvent Extr. Ion Exch. 17, 23 (1999).
[3] Z. Kolarik, U. Müllich and F. Gassner, Extraction of Am(III) and Eu(III) Nitrates by 2,6-Di(5,6-
dipropyl-1,2,4-triazin-3-yl)pyridine, Solvent Extr. Ion Exch. 17, 1155 (1999).
[4] M. Weigl, A. Geist, K. Gompper, J.I. Kim, Kinetics of Lanthanide/Actinide Co-extraction with
N,N’-dimethyl-N,N’-dibutyltetradecylmalonic diamide (DMDBTDMA) (submitted).
[5] G.J. Hanna, R.D. Noble, Measurement of Liquid-liquid Interfacial Kinetics, Chem. Rev. 85, 583 (1985).
[6] P.R. Danesi, Chapter 5 in: Principles and Practices of Solvent Extraction, J. Rydberg,
C. Musikas, G.R. Choppin (Eds), Marcel Dekker Inc., New York, Basel, Hong Kong (1992).
[7] C. Madic et al., New Partitioning Techniques for Minor Actinides, Final Report, EUR-19149 (2000).
[8] B.M. Kim, Membrane-based Solvent Extraction for Selective Removal and Recovery of Metals,
J. Membr. Sci. 21 (1984) 5.
[9] R. Prasad, K.K. Sirkar, Dispersion-free Solvent Extraction With Microporous Hollow-fiber
Modules, AIChE J. 34 (1988) 177.
[10] U. Daiminger, A. Geist, W. Nitsch, P. Plucinski, The Efficiency of Hollow Fiber Modules for
Non-dispersive Chemical Extraction, Ind. Eng. Chem. Res. 35, 184 (1996).
647

THE POTENTIAL OF NANO- AND MICROPARTICLES FOR THE SELECTIVE
COMPLEXATION AND SEPARATION OF METAL IONS/RADIONUCLIDES
C. Grüttner 1 , S. Rudershausen 1 , J. Teller 1 , W. Mickler 2 , H.-J. Holdt 2
1 Micromod Partikeltechnologie GmbH, Friedrich-Barnewitz-Str. 4,
18119 Rostock, Germany,
E-mail: www.micromod.de
2 Universität Potsdam, Institut für Anorganische Chemie,
Am Neuen Palais 10, Haus 9, 14469 Potsdam, Germany
Abstract
Nano- and microparticles for the selective complexation of metal ions and especially radionuclides on
their surface are presented. Beside several applications of such magnetic and non-magnetic particles in
the fields of biomedicine, diagnostics, molecular biology, bioinorganic chemistry and catalysis a high
potential exists for the complexation of radionuclides from nuclear wastewater on particle surfaces.
The magnetic properties of nano- and microparticles allow the fast magnetic separation of
radionuclides from the radioactive liquid waste stream, for example. The removal of radionuclides
from strongly acidic wastes requires a high stability of the particles in combination with the protection
of the incorporated iron oxide. The covalent binding of selective chelators allows the fractionation of
different types of radionuclides regarding the special needs of nuclear waste treatment.
649

1. Introduction
Nano- and microparticles are widely used for the immobilisation of metal ions [1] and
radionuclides [2,3] in the fields of biomedicine [4], molecular biology [5], medical diagnostics [6],
bioinorganic chemistry and catalysis [7]. In general there are two possibilities for the binding of metal
ions on particle surfaces. One method is based on the simple physical adsorption of chelators [8] or
metal ions on particle surfaces by inclusion into pores of the particles, adhesion processes or
electrostatic interactions The second more specific method consists of the complexation of metal ions
by selective chelators which are covalently attached to the particle surface. Nearly all applications of
the metal ion immobilisation on particle surfaces require an efficient complexation of the metal ions to
prevent traces of free metal ions in the special medium. Therefore the effective chemical binding of
metal ions on particle surfaces is our method of choice.
Here we report our results on the selective binding of metal ions and radionuclides on the surface
of magnetic and non-magnetic particles for the application in the magnetic field assisted radionuclide
therapy, for the selective binding of histidine-tagged proteins via the formation of a nickel(II) protein
complex, and for the complexation of palladium ions by sulfur-rich macrocyclic ligands on the surface
of silica particles.
These experiences resulting from the immobilisation of metal ions and especially radionuclides for
life sciences applications initiated our first attempts of the selective complexation of lanthanides and
actinides on the surface of magnetic silica beads. Current approaches for the recovery of lanthanides and
actinides from high level nuclear waste are based on the TRUEX process which utilises the highly
efficient, neutral, organophosphorous ligand octyl-phenyl-N,N-diisobutyl-carbamoylmethyl-phosphine
oxide (CMPO) [9]. Previously, calix[4]arene based extractants which incorporate CMPO moieties at
either the wide [10,11] or narrow rim have been reported [12]. Such pre-organisation of the chelating
ligands leads to a 100 fold (or even more) increase [10] in extraction efficiency combined with an
enhanced selectivity for actinides and lighter lanthanides [13]. Derivatives with single CMPO groups and
CMPO-substituted calixarenes were covalently attached to the surface of magnetic silica particles
allowing controlled ligand loading with defined orientation [14].
First solid-liquid extraction experiments were performed under conditions that simulate European
nuclear waste streams (4M NaNO 3 , 1M HNO 3 ). Separation of europium, cerium or americium, as
representatives of the early lanthanides and actinides, was evaluated. ICP-MS measurements of the
initial nuclide activity in the aqueous phase and the activity after shaking with the particles were used
to calculate the percentage extraction [14].
2. Selective complexation of the radionuclides 99m Tc/ 188 Re and 111 In/ 90 Y on the surface of
microparticles for therapeutical purposes
The magnetic field assisted radionuclide therapy aims the targeting of diagnostically and
therapeutically important radionuclides like 99m Tc or 111 In and 188 Re or 190 Y, respectively, to the
tumour. Therefore the radionuclides are efficiently complexed on the surface of biocompatible
magnetic nanoparticles. These nanoparticles are injected in the near of the tumor region and kept in
the target area by means of external magnetic fields. This leads to a high concentration of radioactivity
at the tumor and prevents side effects on the healthy tissue [15]. Another strategy for successful
tumour treatment is the intracavitary radionuclide therapy: The radioactive labelled microspheres are
immobilised in the tumour because of their size, and irradiate the tumour cells. After the radioactivity
is faded away the microspheres are biotransformed into harmless metabolites [16]. Both strategies
ideally require magnetic or non-magnetic biodegradable particles able to complex radionuclides
650

efficiently and stable. This can be reached by reacting one of the best known chelators for technetium
and rhenium, MAG 3 1, or for indium and yttrium, DOTA 2, with a functionality, e.g. an amino group,
on the surface of the microsphere (Schemes 1 and 2).
Scheme 1. Chemical structures of mercaptoacetyl-triglycine (MAG 3 ) [17] 1 and 1,4,7,10-
tetraazacyclododecane-N,N’,N’’,N’’’-tetraacetic acid (DOTA) [18] 2
O
COOH
COOH
O
NH
HN
N
N
SH
HN
O
N
N
COOH
COOH
COOH
1
2
Scheme 2. Covalent binding of chelators to particle surfaces followed by radiolabeling
Then these microspheres can either be labelled with the radioactive isotopes 99m Tc or 111 In, which
are gamma emitters, or with the beta emitters 188 Re or 90 Y for therapeutic applications. In the case of
99m Tc we obtained first results for non-magnetic microspheres with a labeling efficiency (particlebound
activity) of 39% and an in-vitro stability of the particle bound MAG 3 -technetium complex of
79%. A relatively high labelling efficiency of non-magnetic microspheres of 72% and an in-vitro
stability of more than 90% could be reached with the gamma emitter 111 In.
In future we want to optimise the labelling procedure, develop new chelators for technetium,
rhenium, indium and yttrium and investigate a combination of the intracavitary radionuclide therapy
with the strategy of the magnetic drug targeting.
651

3. Selective removal of histidine-tagged proteins from fermentation solutions by nickel(II)-
protein complex formation on the surface of magnetic silica particles (sicastar ® -M)
The covalent binding of backbone-modified nitrilotriacetic acid (NTA) on the surface of magnetic
silica particles is the basis for the formation of a nickel(II) complex with a high complex stability. The
backbone-modification allows the interaction of four chelating sites of the modified NTA with
nickel(II) ions, which results in a more tightly binding of nickel(II) ions in comparison to systems with
only three sites available for the nickel(II) complexation (Scheme 3).
Scheme 3. a) Interaction of four donor atoms of backbone modified NTA with nickel(II) ions,
b) Interaction of only three donor atoms of non-modified NTA with nickel(II) ions. One
carboxylic acid group of the NTA is necessary for the covalent binding on the particle surface.
The high selectivity of this Ni(II)-NTA complex for proteins with an affinity tag of six
consecutive histidine residues allows a one-step purification of almost any protein from any
expression system under native or denaturing conditions (Scheme 4).
Scheme 4. Principle of the selective binding of his-tagged proteins on the surface of particles
containing chelated metal ions (nickel(II) ions) on the surface.
Electrokinetic measurements of the surface potential of magnetic silica beads have been carried
out to determine the optimal density of NTA on the surface of the particles. Therefore the density of
NTA was increased stepwise until a saturation of the surface with NTA was achieved. This saturation
range was detected by the comparison of the Zetapotential values of the particles at a constant pHvalue
of 8.0 (Figure 1a). In addition streaming potential measurements were carried out to determine
the corresponding particle charge density values by polyelectrolyte titration against 0.001 N
poly(diallyldimethylammonium chloride) (Figure 1b). An optimal surface was realised by reacting
250-500 µmol NTA-chelator with one g of functionalized particles. The binding capacity of the
optimized NTA-modified particles lies in the range of 2.5-3.0 nmol nickel(II) ions per g of magnetic
silica particles. The magnetic silica-NTA beads (sicastar ® -M) can be used to purify 6xHis-tagged
proteins from any expression system including baculovirus, mammalian cells, yeast, and bacteria.
652

Figure 1a. Zetapotential of NTA-modified silica
particles (sicastar ® -M) with increasing densities
of NTA-groups on the surface (electrolyte:
10 -4 M KCl, pH = 8.0)
Figure 1b. Particle charge density of NTAmodified
silica particles (sicastar ® -M) with
increasing densities of NTA-groups on the
surface measured by polyelectrolyte titration
against 0.001 N poly(diallyldimethylammonium
chloride).
0
0
-10
-0,5
Zetapotential [mV]
-20
-30
-40
PCD [µmol/g]
-1
-1,5
-2
-50
-2,5
-60
0 500 1000 1500 2000 2500
c (NTA) [µmol/g]
-3
0 500 1000 1500 2000 2500
c (NTA) [µmol/g]
4. Complexation of palladium ions by sulfur-rich macrocycles on the surface of silica particles
1,2-Dithioethenes are weak chelate-forming ligands [19]. In the case of bis(methylthio)maleonitrile
[20] the donor power of the sulphur atoms is further decreased by the electron withdrawing effect of the
cyano groups. Crowned dithiomaleonitriles are macrocyclic chelate ligands which extract Pd(II) at
sufficient rate in a very good yield. The reason for that extraction behaviour is the fact that Pd(II)
favours the square planar coordination geometry in opposite to the 3d-metals and thiophilic d 10 ions
like Ag(I) and Hg(II).
For the synthesis of the immobilised ligands the 2-allyloxy-1,2-propanediol is transformed into
the dichloro compound, which is reacted with a dithiolate ((Z)1,2-disodium-1,2-dicyanethene-1,2-
dithiolate, 1,2-disodium-4-methylbenzene-1,2-dithiolate [21]) at high dilution conditions yielding the
macrocycle. Than the allylsubstituted crown ether is silylated and the resulting alkoxysilane is
immobilized on activated silica beads (Scheme 5).
Scheme 5. Derivatization of the macrocycle for the immobilization on silica beads
NC
NC
S
S
O
O
CH 2 O CH 2 CH 2 CH 2
Si
O
O
O
Silica
The substituent forms simultaneously a spacer which can be modified in the future. The
selectivity of the ligand should be applied by immobilisation at an inactive matrix for the
accumulation of Pd(II) from diluted solutions. The extraction of Pd(II) was performed from nitric acid
solution with a yield of 93% into a ligand solution (chloroform, kerosene). The extraction equilibrium
is reached after 10 min. By AAS the metal concentration in the aqueous phase was determined to
calculate the extraction rate.
653

Scheme 6. Formation of the macrocycle-Pd(II) complex
O
O
CH 2 O CH 2 CH CH 2
O
CH 2 O CH 2 CH CH 2
S
S
+ PdCl 2
S
O
PdCl 2
S
NC
CN
NC
CN
The extraction rate increases from the acyclic compound through maleonitrile-dithio-21-crown-7,
maleonitrile-dithio-15-crown-5 and maleonitrile-dithio-18-crown-6 by modification of the cavity of
the macrocycle. The best results can be observed for the maleonitrile-dithio-12-crown-4. The rise of
the function lg D = f(lgc L ) gives the composition of the extracted compounds as 1:1. Summarising, a
very good separation of Pd(II) can be specified from 3d-metals and other thiophilic metal ions. In
addition to the extraction experiments and the crystal structures the formation constants of selected
chelates were determined by UV spectroscopy. The observed order corresponds to that found by the
extraction of palladium in the system water/chloroform.
5. Extraction of lanthanides and actinides by magnetically assisted chemical separation
technique
The removal of radionuclides from strongly acidic wastes requires a high stability of the particles
in combination with the protection of the incorporated iron oxide. Therefore magnetic silica particles
were additionally coated with functionalized alkoxysilanes to encapsulate the iron oxide and to
introduce functional groups for the covalent binding of chelators on the particle surface.
Carboxylic acid modified sicastar ® -M (I) were used as the basis for the attachment of simple CMPO
ligands directly onto the particle surface (II). The CMPO-modified particles (II) enable extraction of
152 Eu, 241 Am and 139 Ce albeit at a very low level only slightly higher than with I (Scheme 7, Table 1).
However, the calix[4]arene based particles (III), with a roughly identical concentration of ligating
functions, show a significantly higher level of extraction (Scheme 7, Table 1). This demonstrates the
importance of pre-organization of the chelating ligands on a suitable macrocyclic scaffold, prior to their
attachment at the particle surface [14].
Table 1. Percentage extraction after 19 h shaking
Magnetic silica particles
152 Eu
241 Am
139 Ce
I 3.7%

Scheme 7. Magnetic silica particles (sicastar ® -M) with carboxylic acid groups (I),
CMPO-derivatives (II), and calix[4]arenes with
pre-organised CMPO-derivative units (III) on the particle surface
Partition coefficients comparable to those seen for the systems with adsorbed ligands [8] are
obtained for europium extraction. However, in contrast, larger K D values per mass of ligating function
are found for americium. Thus the covalent systems show considerably enhanced extraction of
americium over europium and offer the potential of selective separation. This reinforces the
importance of initial pre-organisation in imparting selectivity. CMPO extractants, such as octyl phenyl
N,N-diisobutyl carbamoylmethyl phosphine oxide are unable to discriminate greatly between actinides
and lanthanides showing only a slight preference for the heavier lanthanides. In contrast, it has
previously been shown, with non-particulate systems [10,11], that incorporation onto a calix[4]arene
allows differentiation between the actinides and lanthanides based on their cationic radii; the actinides
and lighter lanthanides with larger radii being extracted more efficiently. The ease of separation of
magnetic particles from the waste stream using magnetic fluidised bed techniques makes this system
more attractive for future industrial development.
6. Conclusion
Nano- and microparticles have a high potential for the selective binding of metal ions on their
surface. Thus the particles can be applied in different fields by variation of the particle matrix, the
particle size and chelators, which are covalently bound on the particle surface. Beside the established
particle use in the life sciences and chemistry the special application of magnetic nano- and
microparticles increases for the removal of heavy metal ions and radionuclides from wastewater.
655

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Optimization of Magnetite Carrier Precipitation Process for Plutonium Waste Reduction, Separation
Sci. Technol., 1997, 32, 127-147.
[4] U.O. Häfeli, S.M. Sweeney, B.A. Beresford, J.L. Humm, and R.M. Macklis, Effective Targeting
of Magnetic Radioactive 90 Y-microspheres to Tumour Cells by an Externally Applied Magnetic
Field. Preliminary in vitro and in vivo Results, Nucl. Med. Biol. Int. J. Rad. Appl. Instr. Part B,
1995, 22, 147-155.
[5] J. Gu, C.G. Stephenson, and M.J. Iadarola, Recombinant Proteins Attached to a Nickel-NTA
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Lokalisierung funktioneller Gruppen in Mesoporösen Materialien durch hochauflösende
Transmissionselektronenmikroskopie, Angew. Chem., 1998, 110, 2847-2851.
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657

NEW EXTRACTANTS FOR PARTITIONING OF FISSION PRODUCTS
J. Plešek, B. Grüner, J. %iþD
Institute of Inorganic Chemistry, 5Hå, Czech Republic
P. Selucký, J. Rais, N.V. Šistková
Nuclear Research Institute 5Hå, 5Hå, Czech Republic
B. ýasenský
Katchem, 5Hå, Czech Republic
Abstract
New progress made in the field of dicarbollide [closo-commo-(1,2-C 2
B 9
H 11
) 2
-3Co)] - (COSANs) based
extractants for partitioning of fission product from spent nuclear fuel, especially Sr 2+ and actinides,
made during past years in the Czech Republic, are described in the paper. The synthetic methods for
two classes of new extraction agents containing in the molecule either hydrophobic arylene bridge
substituents or metal selective groups with donor atoms able to co-ordinate polyvalent cations have
been developed. The structures of the recently prepared extraction reagents are presented, along with
ideas on which syntheses were based and the basic relations between structures and extraction
properties of the compounds.
659

1. Introduction
137
90
Extraction process for removal of Cs and Sr from radioactive waste, based on
cobaltadicarbollide anion [closo-commo-(1,2-C 2
B 9
H 11
) 2
-3Co)] -
(COSANs) (see Figure 1) derivatives
as extractants, was designed by IIC ASCR and NURI Re in 1972 and further developed during
subsequent decade [1-4]. The parent COSAN was later chlorinated in order to protect positions B(8)
and B(8’) of the cage toward oxidation. The hexachloroderivative, [(8,9,12-Cl 3
-C 2
B 9
H 8
) 2
-3-Co] (-) -
COSAN was found appreciably more stable towards HNO 3
, radiation, and even more hydrophobic
than the parent compound. Drawback of chlorinated COSANs based process lies, however in the use
of polar and environmentally dangerous solvents i.e. nitrobenzene, or halogenated hydrocarbons.
Presently, there is a co-operative research on this technology between USA and Russia, but details in
the open literature are scarce [5]. In Russia, a plant based on a modified process based on Russian-
Czech Patent [6] was launched in autumn 1996.
Figure 1. Schematic structure of the parent [closo-commo-(1,2-C 2
B 9
H 11
) 2
-3Co)] - (COSAN) anion
(for clarity, terminal hydrogen atoms are omitted)
10
6
11 12
5
9
2
1‘
1
2‘
7
Co
4‘
4
3
3‘
7‘
8
8‘
6‘
5‘
11‘
9‘
12‘
10‘
The targets of the recent and current investigations proceeding under framework of two INCO-
Copernicus EEC Projects [7] have been directed to find more powerful extractants effective also for
actinides and to minimise environmental risks, i.e. they should be able to meet with the EEC
ecological requirements. New extractants of closo-borate type have been tailored with the aim to find
selective reagents for individual fission products and to improve solubility of boron type extractants
in solvents ecologically more acceptable than nitrobenzene, originally used in the above-mentioned
dicarbollide process. Two groups of extractants were prepared starting both from sandwich skeleton
of COSAN incorporating hydrophobic and selective substituents into extractant molecule.
2. Results and discussion
During past years, attention has been paid on the increase of the hydrophobicity of the molecules
introducing arylene rings (phenyl (1), tolyl (2), ethylbenzyl (3), xylyl (4), biphenyl (5), tetraline (6),
etc.) bonded in positions 8,8’ of the COSAN molecule as a bridge substituents (see Figure 2, for
example). Extraction experiments proved that several promising extractants were successfully
prepared. The novel class of 4,8’, 8,4’- R 2
-Bis-arylene bridged COSANs (R = Ph (7), R = tolyl (8),
R = ethylphenyl (9)), and especially the basic member of the series bisphenylene-COSAN (7), exhibit
extreme complexation properties for caesium cation, overcoming significantly extraction ability of
dicarbollide itself and displaying enhanced solubility in aromatic solvents. Indeed, bis-bridged class
660

of COSAN derivatives allowed for use of aromatic solvents (toluene, xylene, etc.) in extraction,
provided that some aromatic sulfo compounds were added to the organic phase as so-called
“solubilizers” [7]. X-ray studies of the Cs + complex of the anion (7) revealed, that that the angle 72 o
between planes of phenylene substituents of this species is very favourable in order Cs + cation can be
strongly captured [8]. Distribution ratio of Cs +
using above anions has been found so high, that
imposes a consecutive problem of its back-extraction. This could be only accomplished using nitric
acid of high concentration. On the other hand, in comparison with chlorinated COSANs, these
compounds seem less stable towards oxidation effect of nitric acid. According to the preliminary
extraction studies it seems that addition of urea to the extraction system would probably suppress this
effect.
Figure 2. Schematic structures of the 8,8’ phenylene – COSAN (1)
and 4,8’, 8,4’ – bis-phenylene bridged COSAN (7)
Co
Co
The present study is devoted to COSAN and hydroborate based extractants containing selective
groups, including phenoxy groups, linear polyethyleneglycol chains, crown ethers and phosphorus
containing moieties, which should allow for selective transfer of strontium and especially M 3+ and M 4+
lanthanides and actinides into low polar organic phase without use of any additive.
The target syntheses are based on the idea originated from our previous experience in the
synthesis and testing of a large number of borate extractants. According to our knowledge, the
polyvalent cations M 3+ and M 4+ can be effectively extracted only provided that the anionic COSANbased
extractant amalgamates in one molecule both, hydrophobic anionic part, and a ligating selective
moiety containing electron donor atoms, i.e. oxygen, phosphorus, sulphur, etc. able of tight coordination
to the cation. Such functionalised anionic particles are able to build up, in solution, a
multi-component “supercomplex” with the cation. Formation of the complex in the extraction system
proceeds spontaneously via a self-assembly mechanism. Inner shell of these complexes contains
encapsulated metal particle bonded to polar donor atoms, outer shell of the “supercomplexes” is
composed by hydrophobic hydroborate core. The charge of the cation is then fully compensated by
the inherent negative charge of several particles of the extractant, and the hydrophobic electroneutral
supercomplex is pulled into organic phase. With the increasing number of hydrophobic anion
particles involved in the complex, the tendency for the transport to the organic phase would increase.
The validity of this principle (i.e. 3:1 complex formation for M 3+ and its transfer to organic phase,
schematically depicted on Figure 3), was confirmed by extraction results, and a recent
electrochemical study [9].
661

Figure 3. Schematic drawing of the complex particle formation
1-
HYDROPHOBIC
ANION
0
n
charge (1-) is
spread over the
surface
1-
R
n
M 3+
R
R
n
1-
A significant advantage of COSAN lies also in remarkable flexibility of its possible substitution
modifications by groups behaving as mono- to poly-donors.
Synthetic strategies to most of such compound have been based on 8-dioxane-COSAN (10) [10]
derivative ring opening procedure by a suitable organic base, deprotonized in situ using NaH (see
Figure 4). This method seems synthetically the most feasible, efficient, and high-yield way for
synthesis of series of COSAN based anionic species with solvent extraction properties (SER). The
series of organic terminal groups successfully attached on COSAN via compound 10 dioxane ring
opening include: 1.t-Octylphenoxide (11), 3-CF 3
-phenoxide (Trifluorocresol) (12), 2-Benzylphenoxide
(13), 2-phenylphenoxide (13), 2-MeO-phenoxide (Guaiacol) (14), [(BUO) 2
P = O] - (15) and
[(PhO) 2
P = O] - (16) (last two as an end-group with powerful sequestering properties). Recently, also
diphenylphosphine oxide moiety has been chosen for its well-known properties to act, even alone, as
efficient sequestering agents for lanthanides and actinides. The species containing the Ph 2
P(O) (17)
moiety as terminal group bonded on the diethyleneglycol chain was prepared via reaction of NaPPh 2
with COSDIOX and subsequent air oxidation of the zwitterionic intermediate by air. All SER of this
type are capable to transfer the target ions (M 3+ ) from aqueous solution into aromatic hydrocarbons
without any other additives.
Figure 4. Schematic drawing of the general route leading
to the synthesis of anionic species 11-21
6
11
10
12
9
2
1‘
1
2‘
5
7
4
3
Co 3‘
4‘
7‘
8
8‘
O
O
+
NaL
2
1‘
1
2‘
Co
O
O
L
6‘
5‘ 11‘
9‘
12‘
10‘
COSANdioxanate
662

On the other hand, all the above anions (11-17) have proved to be effective Eu 3+ extractants only
under neutral or slightly acidic conditions. No one of the investigated SER of this type seemed
promising for technological application in strongly acidic medium. The unfavourable dependence on pH
could be explained in terms of protonation of a strongly basic oxygen B (8)
-O-R leading to a [SER (-) . H (+) ]
“zwitterion”, no longer capable to sequester the target ion and especially to compensate its (+) charge.
To test this idea, the low efficient dibutyl ester 15 was converted via alkaline hydrolysis to the
PHOSDIOX with the monobutyl ester of phosphonic acid (18) as the terminal group and after complete
hydrolysis to PHOSDIOX Acid (19) with –P(O)(OH) 2
group on the spacer chain. Indeed, these
compounds were found reasonably more effective and amazingly specific for Eu 3+ . However, a decrease
of D Eu3+
values with increase of HNO 3
concentration could still be seen (see Tables 1 and 2).
From the study made on a series of model, purely organic phosphonic acid derivatives followed:
the oxygen in the spacer arm between COSAN and phosphorus containing moiety plays no role in
sequestration of the Eu 3+ -ion. Probably the acidity of the end group and its donating properties are
decisive for cation binding.
Table 1. Extraction of some fission products by PHOSDIOX extractant 18
C HNO3
D Cs
D Sr
D Eu
0.01 50.5 10.9 9.98
0.03 – – 36.5
0.05 – – 191
0.08 – – 847
0.11 6.18 0.22 12519
0.31 – – 160
1.01 0.368 0.004 1.94
2.01 0.113 – 0.252
3.01 0.047 – 0.111
0.05 M PHOSDIOX in toluene (obtained solution), up to 0.05 M HNO 3
some reagent losses to the aqueous phase.
Table 2. Europium extraction by PHOSDIOX acid 19
C HNO3
0.01 0.05 0.11 0.51 1.01 1.51 2.01 3.01
D Eu
111 2 376 4 959 63.1 3.99 1.32 0.45 0.08
0.05 M PHOSDIOX acid in toluene, in all cases reagent losses to the aqueous phase.
Two compounds of the above type with (CH 2
-15-crown-5) (20) and (CH 2
-21-crown-7) (21)
terminal groups bonded via diethyleneglycol chain were prepared and tested. Especially the last
compound exhibits good selectivity and enhanced extraction properties for Sr 2+ .
More recently, new synthetic methods for direct attachment of phosphorus containing substituents
on COSAN cage have been successfully developed starting from COSAN-OH (22) and COSAN-(OH) 2
(23). The ancient synthetic procedures [11] leading to hydroxyderivatives of COSAN were revised and
substantially improved. The species 22 and 23 were used as useful synthons for bonding a large variety
of metal selective phosphorus containing groups on the cage. Non bridged 8-(HO) 2
PO-O- COSAN acid
663

(24), 8-PhPO(OH)-O-COSAN (25), and bridged 8,8 , -µ-HO(O)P(O) 2
COSAN (26) and 8,8 , -µ-Ph(O)P(O) 2-
COSAN (27) anions containing phosphorus moiety were prepared in amounts sufficient for testing.
Further attention has been paid to improve their extraction properties and the solubility in less polar
solvents. Compound containing the bridging diethylphosporamide (28) moiety was synthesised and
characterised.
Figure 5. Examples of anionic compounds with non bridged 23 and
bridged 25 B-O-P bonded phosphorus containing selective group
O
Co
O
P OH
OH
Co
O
O
P
O
OH
24
From the standpoint of Eu 3+ extraction neither non-bridged phosphoric acid derivatives 24, 25
nor their bridged analogue 27 were exceptionally effective reagents. Best extraction results have been
observed with the species 26 with the -8,8’-O 2
>P(O)(OH) bridge substituent, which has been found
efficient in europium extraction (see data in Tables 3 and 4). This compound exhibit maximum on
nitric acid concentration dependence of Eu 3+ extraction, the maximum distribution ratio being over 10 3
at 0.2M HNO 3
then falling down but still sufficiently high at 1 M concentration.
Table 3. The Eu extraction by different bridge-type extractants 26-28
C HNO3
D Eu
0.1 0.3 0.5 1.0 2.0 3.0 5.0
26 158 61.6 16.0 4.79 0.910 0.462 0.247
27 32.2 – – 0.124 – – –
28 45.7 29.9 – 4.82 1.55 – –
0.01 M extractant in xylene.
Table 4. Acid dependence on europium extraction by phosphoric acid bridged COSAN 26
C HNO3
0.01 0.03 0.05 0.1 0.3 1.0 3.0
D Eu
157 157 363 158 61.1 4.79 0.46
0.01 M compound 26 in xylene.
664

All compounds presented above were adequately characterised by HPLC, FAB M.S., high field
multinuclear NMR and some of them by X-ray diffraction. The structures of all extraction reagents
were presented at the meeting, along with comprehensive extraction tests results.
General drawback of nearly all mentioned – otherwise successful – extractants still seems to be
their not sufficiently high solubility in low polar solvents. It is believed that further substitution of
their molecules can increase their hydrophobicity and solubility in solvents of interest. The
development of new possible extractants still continues within the framework of the EEC Project.
Therefore more efficient extractants could be found, which technology should be developed in the
future. According to the last results, a solution could be reached, COSAN extractants developed very
recently on the similar basis provided D Eu3+
extraction coefficient in the order of hundreds from
standard waste solution (1M HNO 3
+ 4M NaNO 3
) using 0 01 M extractant and either toluene or
xylene as the solvent.
Up to date, the samples of extractants were prepared in several gram quantities. If the process
based on their use is accepted for technological use, Katchem Prague, Ltd. is supposed to be their
main producer, and the technology of large scale production scale should be developed and optimised
in co-operation with IIC.
Acknowledgements
The partial support from EEC Project IC-CT-155-221, the Grant Agency of the Czech Republic
(Grant 104-99-1096) and the grant of Czech Ministry of Education OK 429(2000) was highly
appreciated.
665

REFERENCES
[1] Kyrš M., +H PiQHN S., Rais J., Plešek J., Czechoslovak Patent 182 913 (11.02.1972).
[2] Selucký P., Baše K., Plešek J., +H PiQHN S., Rais J., Czechoslovak Patent 215282 (01.08.1981).
[3] Rais J. Selucký P., Kyrš M., J. Inorg. Nucl. Chem. 1976, 38, 1376.
[4] Plešek J., +H PiQHN S., Czechoslovak Patent 188587 (04.11.1981).
[5] Romanovsky V.N., Proceedings of the 5th International Information Exchange Meeting on
Actinide and Fission Product Partitioning and Transmutation, Mol, Belgium, Nov.25-27, 1998,
EUR18898 EN, pp. 77-85, OECD/NEA (Nuclear Energy Agency), Paris, France, 1999.
[6] Lazarev L.N., Lyubtsev R.I., Galkin B.Ya., Romanovsky V.N., Shishkin D.N., Kyrš M.,
Selucký P., Rais J., +H PiQHN S., Plešek J., USSR Patent 1031088 (06.06.1981).
[7] Final Report, Project CIPA-CT93-0133, European Commission, February 1997.
[8] Plešek J., +H PiQHN S., Collect. Czech. Chem. Commun., 1995, 60, 1297-1302.
[9] 0DUHþHN V., Jäncherova J., Plešek J., Grüner B, Paper under preparation.
[10] J. Plešek, S. +H PiQHN, A. Franken, I. &tVD RYi and Ch. Nachtigal, Coll. Czech. Chem.
Commun., 1997, 62, 47.
[11] Francis J.N., Hawthorne M.F., Inorg. Chem. 1971, 10, 594.
666

INFLUENCE OF INTERMEDIATE CHEMICAL
REPROCESSING ON FUEL LIFETIME AND BURN-UP
A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
The influence of intermediate chemical processing of nuclear fuel with removal of fission products on
the fuel burn-up and lifetime for heavy water CANDU type reactors operating with fuel on base of
natural uranium is studied in this paper. Two types of nuclear fuel are considered: natural and slightly
enriched uranium (with enrichment up to 1.4%) and thorium fuel on basis of 232 Th- 233 U. Intermediate
chemical processing permits to prolong lifetime and to increase fuel burn-up. However, the effect is
not so high, the increase of burn-up is about 20%. More effect is gained by use of a fuel with
increased enrichment.
667

1. Introduction
A heavy-water CANDU-type reactor has good neutron-physical characteristics due to the use of
heavy water as moderator and coolant. In particular, it allows using natural uranium as nuclear fuel,
whereas in other types of thermal neutron reactors, it is necessary to use uranium with enrichment of
several percents. In natural uranium fuel, the relative role of plutonium produced from 238 U is great.
Nevertheless, fuel burn-up and lifetime are small because fuel multiplying properties at burning out
are quickly reduced. One of the opportunities to increase the lifetime is the transition to nuclear fuel
with slight enrichment. Another opportunity, intermediate chemical processing of fuel can be
considered where the fission products are removed and fuel nuclides are recycled for further burning.
For the future atomic power, a nuclear fuel cycle on base of 232 Th- 233 U can be rather perspective.
Thorium cycle has essential advantages over traditional uranium – plutonium cycle because of
considerably less amount of transuranium long-lived radioactive wastes (though considerably more
amount of rather harmful 232 U). In CANDU reactors, operation in thorium cycle and, the intermediate
fuel cleaning of the fission products could also prolong lifetime and lower the requirement of
specially obtained 233 U.
In this paper, results are given of a calculation study of the influence of intermediate chemical
processing of nuclear fuel with removal of fission products on fuel lifetime and burn-up in CANDUtype
reactor are given. Uranium fuel on basis of natural and slightly enriched uranium (with
enrichment up to 1.4%) and thorium fuel on base of 232 Th- 233 U are considered.
2. Calculation model
The reactor design is described in [1]. In an active core, 380 fuel assemblies are placed. Every
assembly contains 37 uranium pins with zirconium cladding in zirconium tube. Heavy water is used
as the coolant and moderator. Fuel assemblies are located in a square lattice with a pitch of 23.5 cm.
Height of an active core is 594 cm. Natural uranium loading in reactor is 114 tonnes.
It was accepted in calculations that reactor multiplying properties can be approximately
described by multiplication factor of an elementary cell as follows. Multiplication factor of an
elementary cell k eff
varies in function of fuel burn-up. The reactor operates in a mode of continuous
refuelling. Fuel assemblies with various burn-up from fresh fuel up to maximum burn-up are situated
in core at every moment. The on-load refuelling is carried out independently in different channels
after achieving the maximal burn-up. It allows accepting the value of multiplication factor in an
elementary cell average over fuel lifetime with correction on neutrons leakage from reactor as
approximate reactor multiplication factor. Such approximation is quite justified for comparative
calculations of the effect of intermediate nuclear fuel cleaning.
At calculations of lifetime, it was considered that the neutrons leakage makes 1% and the value
= 1.01 was accepted.
In reactors with continuous refuelling at constant power, the value of neutron flux varies in time
very slightly. It is necessary that the power of one fuel assembly varies in time because of changing of
fissile nuclide amount, and the fuel assemblies with different burn-up have appreciably different
power. Calculations of fuel burn-up and transformation of isotopes were carried out at constant
neutron flux.
668

3. Natural or slightly enriched uranium fuel
The calculations of reaction rates and multiplication factor in an elementary cell were performed
with the code [2]. A fuel assembly with pins was represented as a 4-ring coaxial assembly with the
same volumes of all-structural materials and fuel loading. The enrichment of uranium from 0.714%
up to 1.4% was considered. The amount of uranium in fresh fuel assembly was accepted the same for
all enrichments, and the 235 U amount corresponded to the enrichment. For natural uranium fuel,
thermal neutrons flux was equal to Φ = 5 10 13 n/cm 2 s. Neutron flux for the enriched fuel was
determined in a way to get the same power of fresh fuel assembly as in the variant with natural
uranium. At calculations of nuclide transformation, isotopes of uranium, neptunium, plutonium,
americium and curium up to 244 Cm were taken into account.
In uranium fuel, 239 Pu is produced. This isotope gives the essential contribution in reactivity of
CANDU-type reactor with natural or slightly enriched fuel already in the initial period of fuel
lifetime. The dependence of multiplication factor k eff
in an elementary cell on irradiation time T was
calculated for initial enrichment of uranium from 0.714% up to 1.4% without intermediate cleaning.
The neutron capture by fission products was taken into account by means of an “effective fission
fragment” [3]. A poisoning by 135 Xe, 105 Rh and absorption of neutrons by 149 Sm and 151 Sm were
additionally taken into account. The fuel lifetime T f
was determined by value = 1.01. Table 1,
fuel lifetime T f
and burn-up FP defined by amount of fission products in 1 tonne of fuel without
intermediate cleaning for different initial uranium enrichment C.
Table 1. Fuel lifetime T f
and burn-up FP without intermediate cleaning
C, % 0.714 1.0 1.2 1.4
T f
, years 2.0 2.8 5.38 6.67
FP, kg/tonne 11.4 17.4 20.8 23.4
Analogous time dependence of multiplication factor in an elementary cell k eff
with intermediate
processing with cleaning from accumulated fission products was calculated for natural uranium fuel and
fuel with enrichment 1%. The intermediate processing was carried out at time T p
= 0.8, 1.2, 1.6 years.
For enrichment 1%, the intermediate processing was carried out also at T p
= 2 and 2.4 years. These data
allow to estimate fuel lifetime corresponding to average over lifetime multiplication factor = 1.01.
They are presented in Table 2. T p
= 0 corresponds to the mode without intermediate cleaning.
Table 2. Fuel lifetime T f
and burn-up FP with intermediate processing
T p
, years
C = 0.714% C = 1.0%
T f
, years FP, kg/ton T f
, years FP, kg/ton
0.0 2.0 11.4 3.9 17.4
0.8 2.52 13.8 4.71 19.9
1.2 2.4 13.3 4.76 20.1
1.6 2.4 13.3 4.69 19.8
2.0 – – 4.57 19.5
2.4 – – 4.41 19.0
669

These data show that fuel lifetime and burn-up in modes without intermediate processing
essentially depend on fuel enrichment. At transition from natural uranium to enrichment 1%, the
lifetime is increased 1.95 times (and 1.4 times because of reduction of flux density necessary to
obtain the same power of fresh fuel assembly) and the burn-up grows 1.5 times. At transition from
natural uranium to enrichment 1.4%, the lifetime is increased 3.3 times and the burn-up grows
2 times. In variants with intermediate processing, the lifetime increase is not so high. The maximal
lifetime for natural uranium, 2.52 years, and burn-up, 13.8 kg/tonne, correspond to time point of
processing T p
= 0.8 years. The lifetime is longer by 26% and burn-up is greater by 21% than without
intermediate processing. For 1% enrichment, the maximal lifetime is 4.76 years and burn-up makes
20.1 kg/tonne. Processing will be done at T p
= 1.2 years. The increase in lifetime is 22% and that of in
burn-up is 16% in comparison with a mode without processing.
4. Thorium fuel 232 Th- 233 U
In thorium mode of operation, all fuel assemblies were considered alike, containing identical fuel
on basis of 232 Th and 233 U. The same elementary cell was studied as for uranium fuel. The thorium
amount in fuel zones was accepted the same as 238 U in uranium fuel. The share of 233 U in fresh fuel
was chosen 1.96% with respect to the amount of 232 Th. That has ensured necessary over-criticality of a
cell for appropriate fuel lifetime and burn-up.
During calculation of nuclide transformation, the production of isotopes of protactinium,
uranium, neptunium, plutonium up to 242 Pu was taken into account. The neutron flux is considered
constant over a lifetime and equal to 5⋅10 13 neutr/cm 2 s. In modes with intermediate processing, it was
considered that short-lived 233 Pa at an intermediate reactor shutdown completely decays into 233 U. The
intermediate processing is made at time points T p
= 0.4, 0.8, 1.2 years. Fuel lifetime and burn-up
corresponding to = 1.01 for different points of intermediate cleaning are shown in Table 3.
Table 3. Fuel lifetime T f
and burn-up FP in thorium modes
T p
, years T f
, years FP, kg/ton
0 1.45 9.5
0.4 1.71 11.1
0.8 1.75 11.3
1.2 1.72 11.1
The lifetime without intermediate cleaning makes 1.45 years, burnup is 9.5 kg/ton. The maximal
increase of fuel lifetime and burn-up at the expense of intermediate processing in comparison with a
usual mode is achieved at processing at T p
= 0.8 years and makes about 20%.
5. Conclusion
The research performed has allowed to establish how it is possible to extend lifetime and to
increase fuel burn-up at the expense of increase of uranium enrichment or at the expense of intermediate
processing of uranium and thorium fuel with fission products removal. If a power of fresh fuel assembly
remains constant with increase of uranium enrichment, it is necessary to reduce the neutron flux. At the
expense of this effect, the lifetime is extended even at the same burn-up of fuel. Increase of burn-up and
additional lifetime increase are caused by reactivity rise. At transition from natural uranium to
670

enrichment 1%, the burn-up grows 1.5 times, the lifetime is extended 1.95 times from 2 up to 3.9 years,
the burn-up corresponding to natural uranium is achieved after 2.24 years. At transition from natural
uranium to enrichment 1.4% the burn-up grows 2 times, the lifetime is extended 3.3 times. In modes
with intermediate cleaning of fission products, an increase of lifetime is not so high. The lifetime for
natural uranium raises by 26%, burn-up by 21%. For 1% uranium, an increase of lifetime is 22% and
that of burn-up is 16%. The optimum time point of processing is somewhat less than half of lifetime
without processing. In thorium mode, the maximal increase of fuel lifetime and burn-up at the expense
of intermediate cleaning in comparison with a usual mode makes about 20%. Thus, the increase of burnup
and lifetime are obtained much more effectively at the expense of fuel enrichment.
REFERENCES
[1] Karachi Nuclear Power Plant, In: Directory of Nuclear Reactors. Vol. IX, Power Reactors.
IAEA, Vienna, 1971, pp. 167-174.
[2] A.Ya. Burmistrov, B.P. Kochurov, Space-energy Neutron Distribution in Cylindrical Cell of a
Reactor (Code TRIFON), Moscow, Pre-print ITEP, 1978, #107.
[3] A.D. Galanin, Introduction in the Theory of Nuclear Reactors on Thermal Neutrons, Moscow,
Energoatomizdat, 1990, pp. 362-367.
671

RECENT PROGRESSES ON PARTITIONING STUDY IN TSINGHUA UNIVERSITY
Chongli Song, Jingming Xu
Institute of Nuclear Energy Technology, Tsinghua University
100084 Beijing, China
Abstract
Recent progresses on partitioning studies in Tsinghua University are reviewed. Declassification
of the commercial HLLW to a waste that is suitable to shallow land disposal is possible. An enhanced
TRPO process with optimal process parameters can meet the required DF of TRU elements. A Total
Partitioning process for commercial HLLW was developed by modification of the TP process for
Chinese HLLW. The Total Partitioning process for commercial HLLW consists of an enhanced TRPO
process to remove TRU elements and 99 Tc, a CESE process to separate strontium, a KTiFC ion
exchange process to segregate cesium and an An/Ln separation process with HBTMPDTP. The flow
sheet of the total partitioning process for commercial HLLW was given.
673

1. Introduction
The final disposal of radioactive waste is one of the key problems that effect the development of
nuclear energy industry. Partitioning and Transmutation (P&T) concept [1] involves chemical
separation of transuranium (TRU) elements as well as long-lived nuclides (for example, 99 Tc, 129 I, etc.)
from HLLW, and transmutation of them to either stable or short-lived nuclides. The P&T constitutes
an advanced nuclear fuel cycle. The implementation of the P&T could significantly reduce long-term
risk of the radioactive waste.
The partitioning of HLLW can also be used as a pre-treatment method of HLLW to reduce α-
waste and HAW volume. In recent years a clean use of nuclear energy (CURE) concept was proposed
for the back-end of nuclear fuel cycle [2]. In the CURE concept the partitioning requires not only to
remove the TRU, 99 Tc and 129 I, but also to segregate 90 Sr and 137 Cs from HLLW. After partitioning the
original HLLW would be de-classified to a non-α, low and intermediate lever radioactive waste that
could be suitable for shallow-land disposal. So for the CURE concept the required decontamination
factors (DF) for TRU elements will be much higher than that for the P&T concept. The required DF of
TRU, 99 Tc, 137 Cs and 90 Sr are given in Table 1 for a typical commercial HLLW. The spend nuclear
fuel had a burn-up of 33 000 MWd/tU, a cooling time of 10 years and 99.75% of U and Pu had been
removed in reprocessing [3]. In Table 1 the α- waste standard of 4 × 10 5 Bq/kg is chosen and
0.40 m 3 concrete/tU waste is supposed to be produced after solidification of the declassified liquid
waste. In order to get a higher waste volume reduction, the separation of lanthanides (Ln) and
actinides are necessary for commercial HLLW. The required DF for TRU elements in Ln fraction
should be higher than 2.4 × 10 5 [4].
Nuclides
Table 1. The required DF for treatment of typical commercial HLLW
to a waste suitable to shallow land disposal
Activity in HLLW
Bq/tU
Chinese standard
GB-9132-88
Bq/kg
Required DF (Solidify by
cementation,
0.4 m 3 /tU)
TRU 1.26 × 10 14 4 × 10 5 4.0 × 10 5
99 Tc 4.78 × 10 11 – –
90 Sr 2.07 × 10 15 4 × 10 10 71
137 Cs 3.01 × 10 15 4 × 10 10 104
In recent years, the study on the partitioning process was carried out in Tsinghua University in
order to meet the DF requirement for typical commercial HLLW. The aim is to declassify the HLLW
to a waste suitable to shallow land disposal. In this paper the recent progresses on partitioning studies
in Tsinghua University will be reviewed. The flow sheet of total partitioning process for commercial
HLLW was given.
2. The enhancement of TRPO process for commercial HLLW
A TRPO process was developed in Tsinghua University for removing TRU elements from HLLW
[5,6] in 1980s. Hot tests of the TRPO process were carried out with HLLW of WAK in Institute for
Transuranium Elements (ITU) at Karlsruhe, Germany in 1993 [7]. The hot test was completed with
24 stages of miniature centrifugal contactor in hot cell. The DF value of TRU elements obtained in the
674

hot tests (See Table2) was enough for the P&T requirement. However, it is not sufficient to meet the
required DF for CURE project because the TRPO process was designed for P&T project in the period.
In recent year, the TRPO process was improved in order to increase the DF values of TRU elements.
Table 2. The DF of TRU elements, 99 Tc and Nd in the TRPO hot tests
HNO 3
in feed
Extraction
stages 237
Np
238
U
Decontamination factors
239
Pu
Am/ 241 Pu
243
Am
99
Tc
144
Nd
Run 1 0.75 6 12.4 >5 400 >760 >2 800 >900 >1 400 >22 000
Run 2 1.36 10 >4 100 >7 000 >950 >3 200 >760 >1 700 >33 000
Tetra- and hexa-valent TRU elements are highly extracted by 30% TRPO-kerosene. The
controlling elements for the removal of TRU elements are trivalent americium and curium. The
extraction behavior for Am and Cm is very similar. So improving Am extraction is a key issue. A
simplified optimised objective function [8] was introduced into the TRPO mathematical model for
americium extraction [9]. The objective function was designed as that the second waste from the
TRPO process should have a minimal volume to improve the safety, cost and to decrease the
environmental impact. In addition, in the calculation, the Am decontamination factor should be above
4.0 í 10 5 and for a conservative consideration, the DF of 4.0 í 10 6 for Am was fixed. The acidity of
feed and scrub solution was chosen to improve neptunium extraction, and was 1.35 M and 0.5 M
respectively. The optimal parameters of the TRPO process for Am extraction were obtained [8] and
are listed in Table 3.
A multistage counter current cascade experiment with simulated HLLW skipped with 241 Am was
carried out to verify the calculated results [10]. A set of 20 stages miniature centrifugal extractor was
installed in a glove box. Optimal parameters were used in the process experiment (see Table 3). The
cascade included 12 stages for extraction, 2 stages for scrubbing and 6 stages for Am stripping. The
simulated feed solution had a specific volume of 1 850 L/tU and skipped americium with a specific
activity of 7.83 í 10 6 Bq/ml. The flow ratio of feed/organic/scrub/stripping was 1/1.21/0.265/1.21.
Very good results were obtained in the cascade experiments [10]. The obtained DF Am was
1.25 í 10 6 and the material balance for Am was 92.6% in the experiments. The americium profiles in
each stage are given in the Figure 1. The experiments show that the calculated results fit with the
experimental ones very well. The required DF for treating typical commercial HLLW to a waste, that
is suitable for sallow land disposal, can be reached with the TRPO process.
675

Table 3. Calculated and experimental parameter of the TRPO process
(For a typical HLLW of a burn-up of 33 000 MWd/tU, Calculated DF Am = 4.0 × 10 6 )
Parameter
Calculated
value[8]
Experimental
value [10]
Specific volume of Feed (F) 1 750 L/tU 1 875 L/tU
Volume of 30%TRPO-kerosene 1.18 F 1.21 F
Volume of scrubbing solution 0.267 F 0.265 F
Volume of stripping solution 1.18 F 1.21 F
Number of extraction stages 10 10
Number of scrubbing stages 2 2
Number of stripping stages – 6
HNO 3 concentration in feed solution 1.35 mol/L 1.35 mol/L
HNO 3 concentration in scrubbing solution 0.5 mol/L 0.5 mol/L
HNO 3 concentration in stripping solution – 5.0 mol/L
Figure 1. Americium activity profiles in TRPO process
1E+08
1E+07
1E+06
Aqu.(Bq/ml)
Org.(Bq/ml)
Am activity, Bq/ml
1E+05
1E+04
1E+03
1E+02
1E+01
1E+00
1E-01
0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21
Stage number
3. The separation of lanthanide and actinides
The separation of trivalent lanthanide (Ln) and actinides is a difficult subject in the separation
chemistry. However the separation of Ln and Actinides is necessary no matter how for the P&T
concept or for the CURE concept. The study on the separation chemistry of lanthanide and actinides is
one of research subjects in Tsinghua University.
An S-coordinated extractant bis (2,4,4 trimethylpentyl) dithiophospninic acid (HBTMPDTP) had
been proven to be an effective extractant for the separation of trivalent Am from Ln [11]. The
HBTMPDTP is prefers to extract Am rather than Ln and the separation factor reaches to 5 000 for
676

trace amount of Am and Eu. The HBTMPDTP (>99% purity) was obtained by purification of a
commercial extractant Cyanex 301 [12]. The extraction chemistry of Am and Ln was studied with
HBTMPDTP. An empirical model of distribution ratio for Am and Ln was derived and a computer
program for counter current separation of Am/Ln by HBTMPDTP extraction was compiled [13]. The
An/Ln separation process parameters were calculated and were verified by batch multistage counter
current extraction experiments.
A conceptual Am/Ln separation flow sheet by HBTMPDTP extraction was proposed for the
Am/Ln fraction from partition process of HLLW [14]. The feasibility of the separation flow sheet was
verified with a hot test of crossing flow extraction [15]. Am specific activity of was 2 × 10 5 Bq/ml and
the lanthanide concentration was 0.021M in the feed solution. After denitration to 0.3 M HNO 3 , the
feed solution was first extracted by Cyanex 301 to remove impurities. It was adjusted to pH 3.5 and
was then fed into extraction unit. More than 99.999% of Am was extracted into the organic phase with
4 stages of cross extraction. The Am concentration in the raffinate was 1 Bq/ml. Only ~3% Ln was
extracted by HBTMPDTP. The average separation factor between Am and Ln was 3 500 for first three
stages. The hot test results proved that the separation process was effective.
The synergic extraction and separation of Am and Ln by HBTMPDTP/TBP-Kerosene was also
studied. At pH about 2.8, quite high separation factor for Am/Ln could be obtained. A multistage
counter current cascade experiment was performed. It included 7 stages for extraction, 3 stages for
scrubbing and 2 stages for stripping. Americium was effectively separated from Ln. The separation
factor of Am from Ln was 5 × 10 4 and the separation factor of Ln from Am was 2500 [16].
4. The Total Partition process for commercial HLLW
A Total Partition (TP) process was developed in Tsinghua University during 1990s for Chinese
high saline (defence) waste [17,18]. The TP process consists of the TRPO process to remove TRU
elements, a Crown extraction process (CESE) to separate strontium and a potassium titanium
ferrocyanide (KTiFC) ion exchanger to segregate caesium. After the treatment by the TP process, the
high saline HLLW was declassified to a non-α low and intermediate waste, that could be cementation
and sallow land disposal. The hot test proved that the TP process worked very well for the waste.
In commercial HLLW, the salt content is much lower than that in the high saline HLLW. Low
salt content benefits the TRU extraction by TRPO extractant. However, it is detrimental to the
strontium extraction by crown ether DCH 18 C 6 . This problem can be solved by addition a concentration
and denitration unit between the TRPO and the CESE process. The unit was used for increasing the
salt content by evaporation and then to adjust the acidity of the solution. The crown ether extraction
and the KTiFC ion exchanger can meet the required DFs of strontium and cesium for commercial
HLLW. The hot test of the TP process for Chinese HLLW had proved the fact. So the TP process for
high saline HLLW can also be used for commercial HLLW after modification. The general flow sheet
of the TP process for commercial HLLW is given in Figure 2. In the flow sheet, the An/Ln separation
process is also included.
677

Figure 2. General flow-sheet of Total Partition process for commercial HLLW
HLLW
30% TRPO
Adjustment
TRU Extraction
Concentration
and denitration
0.5 M HNO 3
Scrub
Scrub
NH
Am+RE
Np+Pu Cyanex 301 U
4OH
Remove impurity by Adjust to pH 3.5
Denitration to Cyanex 301 extr.
0.3 M HNO 3
0.1M DCH 18C 6
1M HNO 3 Cyanex 301 to cleaning H 2O
0.5 M HBTMPDTP
Am Extraction
RE raffinate
Sr Extraction Scrub Stripping of Sr
Am Stripping
Recycling of
Sr (HAW)
Recycling Am
Cs Ion exchange
DCH 18 C 6 HBTMPDTP
H 2O
0.1M HNO 3
Ion exchanger
waste (HAW)
Cleaning of TRPO
Solidification by
cementation
Non α LLW/MAW
Shallow land disposal
5.5 M HNO 3 0.1 M HNO 3 0.6 M H 2C 2O 4 0.05 M HNO 3 5% (NH 4) 2CO 3
Stripping of Am
To MAW
Stripping of Np+Pu
Recycling
TRPO
Stripping of U
The auxiliary processes of the TP process are now being studied. They include the denitration and
calcination for Am (RE) stripping solution, Np/Pu separation in H 2 C 2 O 4 -HNO 3 solution, the
conversion process for uranium stripping solution and immobilization process for Cs-loaded KTiFC
ion exchanger. The advanced extraction equipment such as pulsed column and centrifugal contactor
for the TP process are also being studied.
5. The radiation stability of TRPO extractant
The radiation stability of TRPO extractant was studied in recent year. The physical properties of
30% TRPO do not have obvious change between a dose of 1 × 10 4 to 5 × 10 5 Gy [19]. Main gaseous
radiolytic products and acidic radiolytic products of 30% TRPO-kerosene extractant were analysed.
Their radiation yield (G value) was determined [20]. At a dose of 1 × 10 4 to 1 × 10 6 Gy, radiolytic
products do not have obvious effect on the extraction. When the radiation dose was above 2 × 10 6 ,
some retention of heave elements were observed [21]. Research indicates that the polymeric products
with high molecular weight cause the retention. The studies show that the TRPO extractant is much
more stable than TBP.
678

6. Conclusion
Declassification of the commercial HLLW to a waste that is suitable to shallow land disposal is
possible. An enhanced TRPO process can meet the required DF for TRU elements with optimal
process parameter. A Total Partition process for commercial HLLW was developed by modification of
the TP process for Chinese HLLW. It consists of an enhanced TRPO process to remove TRU elements
and 99 Tc, a CESE process to separate strontium, a KTiFC ion exchange process to segregate caesium
and an An/Ln separation process with HBTMPDTP.
REFERENCES
[1] A.C. Croff, J.O. Blomeke, Actinide Partitioning-Transmutation Program, Final Report, ORNL
5566 (1980).
[2] S.E. BINNEY, CURE: Clean Use of Reactor Energy, WHC-EP-0206 Westinghouse Hanford
Company, Richland, WA 99352, 1990.
[3] C. Song, Study on Partitioning of Long Lived Nuclides from HLLW in Tsinghua University, in
Energy Future in the Asia/Pacific Region, Proceeding of the International Symposium, Beijing,
China, 2000, pp. 89-99.
[4] Y. Zhu, J. Chen, R. Jiao, Hot Test and Process Parameter Calculation of Purified Cyanex 301
Extraction for Separating Am and Fission Product Lanthanide, Proceedings of the Global’97
Conference, Yokohama, Japan, 1997, Vol. 1, pp. 581-585.
[5] Y. Zhu and C. Song, Recovery of Neptunium, Plutonium and Americium from Highly Active
Waste, Tri-alkyl phosphine Oxide Extraction, in Transuranium Elements: A Half Century,
Edited by L.R. Morss and J. Fuger, 1992, ACS, Washington D.C. USA, pp. 318-330.
[6] C. Song, Y. Zhu, D. Yang, L. He, J. Xu, Chinese J. Nucl. Sci. Eng., 1992, 12 (3), 225 (in
Chinese).
[7] J-P. Glatz, C. Song, L. Koch, H. Bokelund, H. He, Hot Tests of the TRPO Process for the
Removal of TRU Elements From HLLW, Proceedings of the Global’95 Conference, Versailles,
France, 10-14 Sept.1995, Vol. 1, pp. 548.
[8] J. Chen, J. Wang, C. Song, Optimization of TRPO Process Parameters for Americium Extraction, to
be published in Tsinghua Science and Technology, 2001 (in English).
[9] C. Song, J.-P. Glatz, Mathematical Model for the Extraction of Americium from HLLW by 30%
TRPO and its Experimental Verification, in A Value Adding Through Solvent Extraction:
International Conference on Solvent Extraction, Vol. 2, The University of Melbourne, Australia,
1996.
679

[10] J. Wang, B. Liu, J. Chen, C. Song, R. Jiao, G. Tain, X. Liu, R. Jia, Test of Removing Americium
From Simulated Commercial High Level Liquid Waste, to be published in J. Tsinghua
University (Science and Technology) (in Chinese).
[11] Y. Zhu, J. Chen, R. Jiao, Extraction of Am(III) and Eu(III) from Nitrate Solution With Purified
Cyanex 301, Solv. Extr. & Ion Exch. 1996, 14, pp. 61.
[12] J. Chen, R. Jiao, Y. Zhu, Purification of Cyanex 301 and its property, Chinese J. Applied Chem.
1996, 13(2), 46 (in Chinese).
[13] J. Chen, Y. Zhu, R. Jiao, Separation of Am(III) from Fission Product Lanthanide by bis(2,4,4-
trimethyl pentyl) dithiophosphinic Acid Extraction – Process Parameters Calculation, Nuclear
Technology, 1998, 122, pp. 64.
[14] J. Chen, R. Jiao, Y. Zhu, A Conceptual Flow Sheet for Am/Ln Separation by HBTMPDTP
Extraction, to be published.
[15] J. Chen, R. Jiao, Y. Zhu, A Cross-flow Hot Test for Separating Am From Fission Product
Lanthanide by bis(2,4,4-trimethylpenthyl) dithiophosphinic acid, Radiochimica Acta, 1997, 76,
pp. 129.
[16] X. Wang, Y. Zhu, R. Jiao, Separation of Am from Lanthanides by a Synergistic Mixture of
Purified Cyanex 301 and TBP, to be published in J. Radioanal. Nucl. Chem.
[17] C. Song, The Concept Flow Sheet of Partitioning Process for the Chinese High-level Liquid
Waste, Atomic Energy Science and Technology, 1995, 29, 201-9 (in Chinese).
[18] C. Song, J. Wang, R. Jiao, Hot Test of Total Partitioning Process for the Treatment of High
Saline HLLW, in Global’99: International Conference on Future nuclear systems, Proceedings,
August 29-September 3, 1999, Jackson Hole, USA.
[19] R. Xin, P. Zhang, J. Liang, C. Song, Study on the Radiation Stability of Trialkyl Phosphine
Oxide, to be published.
[20] R. Xin, C. Song, J. Jiao, J. Liang, Investigation of Radiolytic Products of Trialkyl Phosphine
Oxide by Gas Chromatography, Chinese J. Spectroscopy Laboratory, 1999, 16, pp. 498-502.
[21] P. Zhang, C. Song, J. Liang, R. Xin, Extraction and Retention of Plutonium with -irradiated
30% Trialkylphosphine Oxide-Kerosene Solution, to be published in Solv. Extr. & Ion Exch.
680

POSTER SESSION
BASIC PHYSICS: NUCLEAR DATA AND EXPERIMENTS
&
MATERIALS, FUELS AND TARGETS
P. D’Hondt (SCK•CEN)
681

DESIGN AND CHARACTERISTICS OF THE n_TOF EXPERIMENT AT CERN
D. Cano-Ott
On behalf of the n_TOF collaboration
CIEMAT
Avda Complutense 22, Madrid Zip: 28040, Spain
Abstract
The n_TOF is a 180 m long neutron time-of-flight facility being built at CERN [1]. The aim of the
experiment is to measure with high accuracy neutron-induced reaction cross-sections in nuclei relevant
to ADS and Transmutation, as well as to Astrophysics. The neutrons are produced by spallation in a
massive lead target using the 20 GeV/c CERN-PS proton beam. It is foreseen that its operation will
start by fall 2000. An overview of the design of installation will be reported, putting special emphasis
on those aspects particular to the n_TOF: the excellent energy resolution and the high-energy spectrum
of the neutrons.
Most of the samples relevant to ADS and transmutation are radioactive and available only in reduced
amounts of high purity. This introduces some constraints to the beam optics that have to be
considered. It will be shown how the design of the neutron beam line has been adapted to the sample
dimensions, in order to preserve the features of the installation and at the same time allow clean
experiments.
Another challenge in the n_TOF design is coming from its geometry. The beam line is located inside a
closed tunnel, which has certainly a great impact in the reduction of the background. This becomes of
particular importance in the attenuation of high-energy neutrons (up to several GeV), since they can
traverse through large amounts of materials producing many secondary particles. It will be illustrated
how the design of the collimation system and shielding elements along the beam line guarantees
acceptable neutron and gamma backgrounds at the measuring station, providing in this way a clean
environment for the detectors.
683

1. Introduction
Among the large number of topics relevant to the design of the n_TOF facility at CERN, a
question of primary importance from the experimental point of view is the determination and
definition of the characteristics of the installation. The time-of-flight (TOF) measurements require a
geometrically well-defined neutron beam at the sample position and the absence of backgrounds.
Moreover, the neutron beam has to be adapted to the size of the samples, which is limited by the
available amounts of high purity materials, by the target construction procedure and also by their
intrinsic radioactivity (in case of unstable isotopes). The neutron beam has to be compatible with the
requirements coming from the various proposed experimental techniques. The experimental
programme of the n_TOF project covers a wide range of measurements summarised as follows:
• (n,γ) cross-section measurements with C 6 D 6 detectors (in a first phase) and with a total
absorption 4π calorimeter (in the second phase).
• (n,f) cross-section measurements with Parallel Plate Avalanche Chambers (PPAC).
• (n,xn) cross-section measurements with Ge or Si detectors.
Even though many sources of background can be highly suppressed through the time-of-flight
tagging, those background events having a time correlation similar to the neutrons may not be rejected
and may severely distort the measurements. Such background sources can be classified in two
categories: i) neutron reactions at the sample without the proper time-energy relation and ii) signals in
the detectors (produced by photons, neutron recoils or other particles) not originating from the reaction
under study. In the first case, the background can be highly reduced by designing the optical and
shielding elements (beam tube, collimators and walls) of the neutron beam line. The second
background species are also reduced in this way, but in addition the number of secondary reactions (of
neutrons and charged particles) has to be minimised also at the experimental area and its vicinity. The
n_TOF collaboration has dedicated a big effort to the studies needed for the definition of the neutron
beam design. Such studies covered a large spectrum of issues that can be summarised as follows:
• Study of the neutronic properties of the spallation target. In particular, production rates, the
energy, the time, the spatial and the angular distribution of the neutrons.
• Study and design of the beam optics, i.e. beam tube and collimators.
• Design of the necessary shielding elements, which guarantee clean experimental conditions.
It should be emphasised that all these investigations were made under the scope of providing
optimal conditions to the measurements, according to the following directives:
• The neutron beam size should have a radius as small as 2 cm, given the availability of the
samples, its intrinsic radioactivity, the overall detection efficiencies and the experimental
requirements arising from the capture (both the C 6 D 6 detectors and the 4π calorimeter), the
fission and the (n, xn) measurements. However, further developments are considered in order
to adapt the beam characteristics to particular measurements by studying and designing
variable size collimators.
• Design the corresponding beam optics (beam tube and collimators) such as to achieve neutron
and gamma backgrounds by order of magnitudes lower than the beam flux.
• Define the shielding elements such, as to improve the background attenuation and respect at
the same time the conditions imposed by the CERN safety rules, which impose the
accessibility and emergency escape paths of the installation.
684

The realisation of the present design implied complex and time consuming Monte Carlo
simulations, in order to achieve quantitative solutions. Several codes based on different models and
evaluated cross-section libraries, were used for this purpose (FLUKA [2], MCNPX [3], GEANT3 [4],
GEANT4 [5] CAMOT [6] and EAMC [7]) and the results have been cross-checked among them, in
order to asses their reliability. In this sense, our studies represent on themselves a benchmark between
the most advanced codes available in neutron physics today.
2. The spallation source
A detailed description of the properties of the lead spallation target can be found in [8]. This
section is devoted only to those characteristics of the target that have an impact to the definition of the
neutron beam. The study of the CERN neutron source had, among others, two major goals:
• To evaluate the most relevant properties of the spallation source, such as the flux and its
spectral function.
• To parameterise these properties in order to implement them in time efficient, but realistic,
Monte Carlo simulations, necessary for most of the studies.
The neutron energies at the n_TOF extend closely up to 20 GeV due to the 20 GeV/c momentum
of the PS proton beam. A FLUKA Monte Carlo simulated energy spectrum of the neutrons at the exit
of the water moderator of the spallation target, is shown in Figure 1. The 38% of the neutrons emerge
with energies below 0.3 eV. The range from 0.3 eV to 20 keV accounts for 23% of all neutrons and
evidences almost exact isolethargic behaviour, as a consequence of the moderation in the water.
However, a significant fraction 32% of the neutrons have energies between 20 keV and 20 MeV, and a
further 7% extends above 20 MeV. Such a hard component, the signature of the spallation reactions,
differentiates substantially the neutron energy distribution at the CERN facility from those of
alternative neutron production mechanisms used in other neutron TOF facilities.
Figure 1. Energy distribution (normalised to area unity) of the neutrons
at the exit of the Pb target after the water moderator
counts
10 -2
10 -3
10 -4
10 -5
10 -3 10 -2 10 -1 1 10 10 2 10 3 10 4 10 5 10 6 10 7 10 8 10 9 10 10 10 11
energy (eV)
685

The simulation of the spallation process is excessively time-consuming for high-energy protons.
An event generator appears to be necessary in order to perform effectively all calculations [9]. For this
purpose, the interactions of the 20 GeV/c protons with the Pb spallation target (80 ×80 × 40 cm 3 ) were
simulated by means of FLUKA [2] and EAMC [7]. A very detailed geometry of the target, the
surrounding water moderator, the mechanical parts and the shielding was included. The position,
energy, time and the direction cosines of the particles emanating entering into the TOF tube were
recorded on a data summary tape (DST). We report here on the results obtained for the neutrons, but
details about other emerging particles can be found in [8]. By taking the TOF tube as the z axis in a
co-ordinate system, the proton beam lies in the y/z plane and enters into the target at x = y = 0 cm and
z = -40 cm, forming an angle of θ = 10º with respect to the z-axis. Such an incident angle balances the
intensity losses in the neutron beam and its contamination due to high-energy charged particles, γ-rays
and others. However, it introduces an asymmetry in the spatial distribution of the neutron source along
the y-axis. The analysis of the DST provided the one-dimensional probability distributions of the
neutron x- and y-positions, p(x) and p(y), the angular distributions of their momenta, p(θ) and p(φ), the
time, p(t), and the energy, p(E), distributions. It was found that all single-variable distributions were
strongly correlated with the neutron energy, and thus, its parameterisation was made as a function of
the energy. Some illustrative examples can be found in Figure 2. In the upper left corner, the energytime
relation can be observed. Such relation is the working principle of the TOF measurements, since
it links the time of flight with the initial neutron energy.
Figure 2. Upper left: Time-energy relation of the spallation neutrons.
Upper right: x and y projections of the spatial distribution of neutrons
with energies between 1 MeV and 10 MeV. Lower left: cos(θ) distribution for
neutron energies between 10 keV and 100 keV. Lower right: φ distribution for
neutron energies between 100 MeV and 1 GeV.
log 10 (t) (t in s)
time vs energy
x-projection
y-projection
log 10 (E) (E in eV)
(m)
cos(q) distribution
f distribution
cos(q)
f (rad)
686

In the upper right corner of Figure 2, the x and y spatial distributions for neutron energies
between 1 MeV and 10 MeV are shown. Both distributions are clearly not uniform and present
pronounced peaks with similar r.m.s. values. It can also be observed that the centroid of the
y-distribution is displaced towards positive values, while the x-distribution remains centred in x = 0.
The displacement varies with the neutron energy ranging from 4.1 cm, for energies below 1 eV, to
6.6 cm, for energies between 10 MeV and 100 MeV, and 9.8 cm for energies above 1 GeV. The r.m.s
of both distributions presents also energy dependence. For neutron energies below 1 eV, a broad
distribution is obtained with a r.m.s of 15.9 cm. The width reduces to 12.9 cm for energies between
1 MeV and 10 MeV and gets its smallest value of 5 cm for energies above 1 GeV.
In the lower left corner, the cos(θ) distribution for energies between 10 keV and 100 keV is
shown. It can be observed that the distribution is not uniform and that there is a clear preference of the
neutrons to be emitted in the forward direction. The effect is enhanced for highest neutron energies.
However, if only small emission angles of the neutrons are considered (θ

Table 1. Parameters of the two collimators. The z co-ordinates are referred to
the exit face of the Pb target. The samples are assumed to be placed at
z = 185 m, where the beam has a spread of 2 cm radius or 4 cm diameter.
Material
Internal
radius
(cm)
Design of the first collimator
External
radius (cm)
Initial z
coordinate (m)
Final z coordinate
(m)
Length (m)
Part 1 Iron 5.5 25 135.54 136.54 1
Part 2 Concrete 5.5 25 136.54 137.54 1
Material
Internal
radius
(cm)
Design of the second collimator
External
radius (cm)
Initial z
coordinate (m)
Final z coordinate
(m)
Length (m)
Part 1
5% borated
Polyethylene
0.9 20 175.35 175.85 0.5
Part 2 Iron 0.9 20 175.85 177.1 1.25
Part 3
5% borated
Polyethylene
0.9 20 177.1 177.85 0.75
From pure geometrical considerations, it can be realised that a beam of 2-cm radius at 185 m can
be prepared if and only if the spallation target has a lateral size of almost 20 cm. This can be achieved
by the use of an additional collimator which partially screens the Pb target. It should be pointed out
that the spallation target should be considered an object consisting by two parts of the same material.
A central cylindrical part of 20 cm radius representing the main part of the Pb spallation source is
surrounded by an Pb reflector forming thus together an object of 40 cm radius, the actual TOF target.
The advantages of Lead as spallation source and efficient reflector are well established [10]. It has
been observed by Monte Carlo simulations that the neutron flux at energies below 1 MeV is increased
about 50% due to this Pb reflector. The source screening by the additional collimator reduces the
neutron flux, since the complete Pb target is no longer visible from the sample position through both
collimators. However, this effect is less significant as naively expected, because the spatial distribution
of the emanating neutrons is not uniform, but peaked at the centre of the Pb target. Moreover, the use
of two collimators turns out to be the most effective way of reducing the neutron and γ background in
the experimental area. The screening of the neutron source reduces by one order of magnitude the
number of neutrons hitting the second collimator, which represents the strongest source of background
observed therein.
The position and inner diameter of the two collimators were optimised [11] by minimising the
losses in flux and by leaving the possibility of future upgrades (already under study) open. In
particular, the study considered explicitly a first collimator that is also appropriate for producing
neutron beams of 8 cm diameter (or even broader). The optimal configuration is described in Table 1.
The first stage consists of a 2 metres long collimator (Source Screening Collimator, SSC) at 135.54 m
from the Pb target and with an inner aperture of 5.5-cm radius. The second stage is a 2.5-m long
collimator (Beam Shaping Collimator, BSC) at 175.35 m from the Pb target and with an inner aperture
of 0.9-cm radius. Such a configuration provides a neutron beam of 4-cm diameter at the sample
location, 185 m downstream from the Pb target.
688

Figure 3. Top view of the area where the BSC is located
y axis
chicane
neutrons
air
2 nd collimator
(BSC)
wall 3.2 m
sample position
(7m after BSC)
concrete
z axis
The optimal composition of the collimators was investigated as a separate issue by Monte Carlo
simulations [2,3,4]. The best results were achieved when a combined or “sandwich-like” collimator is
used [12]. In order to moderate the neutrons with energies below 20 MeV, a 40-50 cm long segment
made of an hydrogen-rich compound doped with a neutron absorber (borated polyethylene – CH 2 B –
in our case) was found to be very efficient. On the other hand, the neutrons above 20 MeV can not be
efficiently stopped in CH 2 B . A 1 m long segment of an intermediate Z material (natural iron in our
case) was found to be necessary and optimal for moderating, through elastic and inelastic scattering,
the high-energy part of the TOF spectrum. For this reason, the SSC consists of a 1 m of iron and a 1 m
of concrete cylindrical segments. A much more careful design was made for the BSC, only 5-7 m
away from the measuring station, where the lowest background rates are required. Thus, a design
based on three segments was adopted. The first part of the BSC, made of 50 cm of CH 2 B moderates
and captures most of the neutrons below 20 MeV. The second part, made of 1.25-m iron, moderates
and diffuses the high-energy neutrons below 20 MeV. The third part of the BSC, made of 75 cm of
CH 2 B moderates and captures the neutrons scattered in the preceding iron segment.
4. The shielding elements
Several shielding elements have to be placed outside the TOF tube in order to bring the
background at the experimental area to an optimal level. It has been observed by Monte Carlo
simulation that it is of crucial importance to avoid high-energy neutrons (above several tens of MeV)
reaching the vicinity of the experimental area and produce secondary particles. Thus, the best strategy
is to take profit of the geometrical factor, by placing the shielding elements as far away as possible
from the measuring station.
689

Figure 4. Radial distribution of the neutron beam profile and the neutron background at the
sample position. The values correspond to the neutron fluence through concentric cylinders:
the first 20 cylinders with radial increments of 1 cm and the last 7 as indicated in the figure.
neutron fluence (n/cm 2 /MCNPX n with θ=0.22341)
10 -6
10 -7
10 -8
10 -9
10 -10
10 -11
beam of 2 cm radius
10 -5 5 10 15 20 25
background
Beam pipe (19.844 - 20 cm)
Air (20 - 40 cm)
Air (40 - 60 cm)
Air (60 - 80 cm)
Air (80 - 100 cm)
Air (100 - 165 cm)
Air (165 cm - wall )
10 -12
region number
The first important shielding element is the pipe itself. Several tens of metres after the Pb target,
the incidence angle of the neutrons hitting the pipe is small and the effective length of material seen by
the neutrons is large. In this way, a large fraction of the neutrons are deviated from its initial trajectory
and produce background in the upstream half of the tunnel. Due to the telescopic structure of the pipe,
the places at the diameter reductions (80 cm to 60 cm and 60 cm to 40 cm) should be reinforced, since
the neutrons traverse the pipe almost perpendicularly. At these positions, the pipe was covered after
the reduction with 1-m thick external iron cylinders.
The second important shielding element is the first collimator (SSC) starting at 135.54 m. Such a
strong scattering centre had to be shielded by a 3 m concrete wall. After the SSC, the neutron beam
divergence is small and the neutrons do not hit the pipe before the second collimator (BSC). The 2.5 m
long BSC diffuses and moderates very efficiently the whole neutron spectrum. However, it is a strong
scattering centre that has to be efficiently shielded by a 3.2-m thick wall, placed 1 m
beam-downstream. Figure 3 shows the top view of the area in scale, where the BSC is located, as has
been designed for the MCNPX simulations. The 2.5-m long BSC is observed on the left-hand side of
the 3.2-m concrete wall. Also visible is the chicane, a necessary escape path in case of emergency. The
impact of this shielding opening on the background at the sample position was studied with
Monte Carlo simulations.
The neutron beam profile and background at the sample position is shown in Figure 4. All the
values correspond to the neutron fluence of concentric cylinders: the first 20 values with radial
increments of 1 cm and the last 7 as indicated in the figure. The first two bins correspond to the main
neutron beam, which has a radius of 2 cm. The remaining bins correspond to the background level at
the different places. It can be seen that the neutron background is on average at the level of
2·10 -12 n/cm 2 per neutron emitted from the Pb target with a θ angle smaller than 0.223º. Such a value,
when compared to the beam fluence of 2·10 -5 , leads to a background to signal ratio of 10 -7 . This
background level is of the same order than the background of scattered neutrons produced at 15 cm
from a 10 mg 235 U sample placed in-beam. A similar result was found for the γ background produced
690

y neutron reactions. The level reached at a separation of 7 metres from it is 10 -7 times smaller than
the neutron beam shown in level by another by factor of ten. Such a value is comparable to the
gammas emitted by a 10 mg 235 U sample. At 15 cm distance, the γ fluence is also seven orders of
magnitude below the neutron beam fluence. As a general remark, it should be noticed that applying the
necessary time of flight cuts would reduce the background levels shown in this section. In addition,
several possibilities of reducing the neutron and γ background in the experimental area were also
investigated. From the Monte Carlo simulations with GEANT and MCNPX, it was early observed that
covering the walls with some neutron moderator and γ shielding could diminish the background level
by another factor of ten.
Figure 5. Various projections of the beam profile at 5 and 7 metres after
the BSC´s for 1 keV neutrons
Beam profile at 5 metres after the 2 nd collimator
Beam profile at 7 metres after the 2 nd collimator
counts
1000
800
600
400
200
0
counts
1000
800
600
400
200
0
-2 0 2 x (cm)
-2 0 2 y (cm)
counts
900
800
700
600
500
400
300
200
100
0
-2 0 2 x (cm)
counts
900
800
700
600
500
400
300
200
100
0
-2 0 2 y (cm)
x projection 1 keV
y projection 1 keV
x projection 1 keV
y projection 1 keV
counts
1200
1000
800
600
400
200
0
0 0.5 1 1.5 2
radius (cm)
radial projection 1 keV
counts
10 3
10 2
10
1
0 0.5 1 1.5 2
radius (cm)
radial projection 1 keV
counts
900
800
700
600
500
400
300
200
100
0
0 0.5 1 1.5 2
radius (cm)
radial projection 1 keV
counts
10 3
10 2
10
1
0 0.5 1 1.5 2
radius (cm)
radial projection 1 keV
5. The beam at the experimental area
The neutron beam at the first phase of the n_TOF operation will have a maximal spread of 2-cm
radius. Although the maximal size of the beam does not depend on the neutron energy, due to the
energy dependence of the neutron spatial distributions at the exit of the Pb target, the beam profile at
the sample position does also depend on the energy. The beam profile has been calculated by the
geometric transport through the collimators for different neutron energies. Figure 5 shows various
projections of the beam profile for 1 keV neutrons at several positions after the BSC. It can be
observed from the radial projection in Figure 5 how the beam spread at 7 m after the BSC’s end is
inside the desired limit of 2 cm. However, it should be noticed that 5 m after the BSC, the beam spot is
smaller and has a diameter of 3.2 cm. Thus, it remains to the particular experiment to decide which
beam size has to be adopted.
It has been calculated by Monte Carlo simulations (FLUKA [2]) that for a PS bunch of 7·10 12
protons, the fluence in absence of collimators is of 7·10 5 n/cm 2 . With the described collimating system,
the fluence amounts to 1.3·10 5 n/cm 2 per proton bunch, which is smaller by a factor of 5. It should be
emphasised that this loss in fluence is sine qua non to the definition of a neutron beam and a TOF
facility fulfilling the requirements and providing the clean conditions necessary for experimentation.
691

6. The neutron escape line
The common experimental situation is that most of the neutrons in the beam do not interact with
the samples. They continue its path undisturbed until it is finally interrupted by any interposed
construction element. Such neutrons are a potentially dangerous source of background, since their
interactions with the surrounding materials in the vicinity of measuring station can interfere with the
experiments. It is necessary that they continue unperturbed its travelling for a long distance before
being scattered. A typical solution adopted at other TOF facilities is to have an experimental area of
large dimensions (several tens of metres). There the neutrons can travel (in air or vacuum) a long path
before being scattered and finally absorbed. In addition, the background introduced by them can be
highly suppressed by the TOF tagging, since the escape path is usually comparable (or larger) in
length than the TOF distance.
Figure 6. Geometry of the neutron escape line as included in the MCNPX simulations
11 m
concrete
neutrons
sample position
(185 m after Pb target)
wall
1.6 m
wall
2 m
7 m
80 cm
80 cm
80 cm
1.2 m
80 cm
11.5 m
concrete
neutrons
50 cm
BF 3
counters
50 cm
pipe diameter
φ = 20 cm
(could be smaller)
polyethylene
40 cm
196.5 m after Pb target
However, the particular situation at the n_TOF makes the mentioned strategies difficult to be
applied. First, the facility is located inside a small closed cave, which limits severely the size of the
experimental area. Second, the TOF tube hits the ground at 200 m from the Pb target, just 15 m after
the measuring station. This situation introduces an unavoidable source of background close to the
measuring station. Third, such a distance is much smaller than the 185 m TOF flight path, which
affects the suppression capabilities of the time tagging. Fourth, the high-energy component of the
neutron energy spectrum. All this considerations make preferably to design a neutron escape line and a
beam dump at its end (properly shielded) instead of having the uncontrolled collision of the neutron
beam with the floor at 200 m. Additional motivations for the neutron escape line from radio-protection
considerations have not been studied.
692

The geometry of the beam dump proposed and included in the simulations can be seen in
Figure 6. At the leftmost part of the figure, the sample position at 185 m from the Pb target is marked
by a dashed line. After it, the unperturbed fraction of the neutron beam continues its travelling in
vacuum for another 11 m, inside an aluminium pipe. It is worth to mention that the tube considered in
our study has a diameter of 20 cm, which is larger than really necessary. As it is shown in Figure 6, the
neutron beam has a maximal spread of 8 cm diameter at 200 m, which would allow the use of a
smaller pipe diameter. However, in order to make sure about the conclusions, it is preferable to know
that the number of neutrons backscattered at the beam dump is even acceptable for a broader tube
(larger geometric acceptance).
After the 11 m flight path, the neutrons cross a 1 mm thick tube end cup and enter into air.
Several materials for the tube end cup have been studied: 1 mm of Al, 1 mm of natural Fe, 1 mm of
carbon fibre and also the ideal case of no end cup (only vacuum). The MCNPX [3] Monte Carlo
simulations revealed no influence of the end cup election on the background distribution at the sample
position.
After the vacuum pipe and 50 cm of air, the neutrons first enter into a 40 × 50 × 50 cm 3
polyethylene block followed by a 2 m thick concrete wall. The polyethylene block has two main
purposes. First, it preferences the scattering of the neutrons in the forward direction, thus reducing the
background in the upstream area. Second, it is the frame for a monitoring detector array based on three
BF 3 detectors. The 2-m thick concrete wall behind the polyethylene block is in charge of further
moderation and absorption of the neutrons. Only the neutrons of highest energies, above several tens
of MeV are capable to cross it.
In order to minimise the number of scattered neutrons entering back into the experimental area, a
1.6-m thick concrete wall is interposed in their way. The wall is placed at 7 m from the sample
position, as far as possible from the point of view of the neutron and γ background. Also visible in
Figure 6 is the 80 cm broad opening in the wall, necessary from the point of view of safety.
Nevertheless, the opening is covered by an additional 80-cm thick concrete shielding forming a
chicane. By keeping the distance of this shielding to the centre of the pipe, also 80 cm, the neutrons
can go through the wall opening only after having suffered more than one interaction.
It should be said at this point that the geometry of the tunnel described in Figure 6 is only an
approximation to the real one. The curvature of the tunnel and its slope were not included. However, if
the distances marked in the figure are kept as they are, the results for the real geometry must be the
same or better, since the approximations made were always done by following the tendency of
increasing the background at the experimental area. Moreover, the geometry described in Figure 6
allows further improvements, because the dimensions of the present shielding elements are not at its
practical limits. For example, there is still space for enlarging them or moving all the shielding
elements 2 metres away from the sample position. However, the present configuration provides
already satisfactory results.
693

Figure 7. Neutron background at the sample position. The values correspond to the neutron
fluence through concentric cylinders: the first 20 cylinders with radial increments of 1 cm
and the last five as indicated in the figure.
neutron fluence (n/cm 2 /MCNPX n with θ=0.22341)
10 -11
10 -12
10 -13
Air (20-40 cm)
Air (40-60 cm)
Air (60-80 cm)
Air (80-100 cm)
Air (100 cm-wall)
10 -10 0 5 10 15 20 25
region number
The lower part of Figure 6 is an enlarged view of the beam dump at the end of the neutron escape
line. Three commercially available BF 3 detectors are embedded on the polyethylene block forming a
long counter [13]. The hole of 4-cm radius is only necessary for the BF 3 long counter working
principle. The main goal of this monitoring set-up is to control possible variations in the gross
properties of the neutron beam such as its intensity or its position. Figure 7 shows the radial
distribution of the neutron background at the sample position. All the values correspond to the neutron
fluence through concentric cylinders: the first 20 values with radial increments of 1 cm and the last
five outside the pipe as indicated in the figure. Such a background can be compared directly to the
previously shown in Figure 4. Two different components can be observed. The first one corresponds
to the initial 10 bins and accounts for the neutrons scattered at the beam dump and travelling back
inside the pipe. Such a background can be considered as a negligible contamination of the beam of
1 part in 10 6 (see Figure 4 to compare with the main beam). A suppression if this background can be
achieved by placing a sheet of Cd in front of the polyethylene block. The second component is the
neutrons travelling outside the pipe, which arrived at the experimental area either by crossing the tube
hole or the opening in the wall. This background is at the same level as the one in Figure 4, that is,
seven orders of magnitude below the main neutron beam. The energy distribution of the background at
the sample position shows that 95% of the neutrons have energies below 1eV. In order to investigate if
such neutrons are coming from the moderation of higher energy neutrons or had initially thermal
energies, a Monte Carlo simulation was performed with an energy spectrum starting at 1 eV. The
result is that 53% of the thermal neutrons at the sample position are produced by neutrons of energies
higher than 1 eV. The other 47% is already originated by neutrons with initial thermal energies. This
implies additional background suppression because the neutron energy spectrum at the sample position
is affected at energies below 0.1 eV because of gravitation. In fact, gravitation will make that the
thermal neutrons emitted from the Pb target fall down along the TOF path of 200 m and do not reach
the samples. The Monte Carlo simulations revealed also that at the sample position, the γ background
due to the beam dump is at least 7 orders of magnitude below the main neutron beam.
694

REFERENCES
[1] S. Abramovich et al., Proposal for a Neutron Time of Flight Facility, CERN/SPSC 99-8,
(March 1999).
[2] A. Ferrari and P.R. Sala, Intermediate and High Energy Models in FLUKA: Improvements,
Benchmarks and Applications, Proc. of Int. Conf. on Nuclear Data for Science and Technology,
NDST-97, ICTP, Miramare-Trieste, Italy, May 19-24 (1997).
[3] Laurie S. Waters, Editor, MCNPX Users Manual, TPO-E83-G-UG-X-00001, Los Alamos
National Laboratory (1999).
[4] GEANT3, Detector Description and Simulation Tool, CERN Program Library W5013, Geneva
(1994).
[5] GEANT4, http://wwwinfo.cern.ch/asd/geant4/geant4.html.
[6] C. Coceva, M. Magnani, M. Frisoni, A. Mengoni, CAMOT: a Monte Carlo Transport Code for
Simulations of Pulsed Neutron Sources, Unpublished (2000).
[7] H. Arnould et al., Neutron-driven Nuclear Transmutation by Adiabatic Resonance Crossing:
TARC. EUR 19117. Project Report of the NST of EURATOM of the EC. ISBN 92-828-7759-0
(1999).
[8] E. Radermacher, Editor, Neutron TOF Facility (PS 213) – Technical Design Report, CERN
(February 2000).
[9] D. Cano-Ott et al., First Parameterisation of the Neutron Source at the n_TOF and its Influence
on the Collimation System, DFN/TR-03/II-00.
[10] H. Arnould et al., Experimental Verification of Neutron Phenomenology in Lead and
Transmutation by Adiabatic Resonance Crossing in Accelerator Driven Systems. Phys. Lett. B
458 (1999) 167.
[11] D. Cano-Ott et al., Proposal for a Two-step Cylindrical Collimator System for the n_TOF
Facility, DFN/TR-04/II-00.
[12] D. Cano-Ott et al., Design of a Collimator for the Neutron Time of Flight (TOF) Facility at
CERN by Means of FLUKA/MCNP4b Monte Carlo simulation, DFN/TR-07/II-99.
[13] A.O. Hanson and M.L. McKibben, Phys. Rev. 72, 673 (1947).
695

RECENT CAPTURE CROSS-SECTIONS VALIDATION ON 232 TH
FROM 0.1 EV TO 40 KEV AND SELF-SHIELDING EFFECT EVALUATION
L. Perrot, A. Billebaud, R. Brissot, A. Giorni, D. Heuer, J.M. Loiseaux, O. Meplan, J.B. Viano
Institut des Sciences Nucléaires/CNRS-IN2P3/UJF
51 avenue des Martyrs, 38026 Grenoble, France
Abstract
Research on ADS, related new fuels and their ability for nuclear waste incineration leads to a revival
of interest in nuclear cross-sections of many nuclides in a large energy range. Discrepancies observed
between nuclear databases require new measurements in several cases. A complete measurement of
such cross-sections including resonance resolution consists in an extensive beam time experiment
associated to a long analysis. With a slowing down spectrometer associated to a pulsed neutron
source, it is possible to determine a good cross-section profile in an energy range from 0.1 eV to
40 keV by making use of a slowing-down time lead spectrometer associated to a pulsed neutron
source. These measurements performed at ISN (Grenoble) with the neutron source GENEPI requires
only small quantities of matter (as small as 0.1 g) and about one day of beam by target.
We present cross-section profile measurements and an experimental study of the self-shielding effect.
A CeF 3
scintillator coupled with a photomultiplier detects gamma rays from neutronic capture in the
studied target. The neutron flux is also measured with a 233 U fission detector and a 3 He detector at
symmetrical position to the PM in relation to the neutron source. Absolute flux values are given by
activation of Au and W foils. The cross-section profiles can then be deduced from the target capture
rate and are compared with very detailed MCNP simulations, which reproduce the experimental
set-up and provide also capture rates and flux. A good agreement between experimental and
simulated profiles for well-known cross-sections like Au for different thicknesses is found in our
energy range, and therefore validates the method and the taking into account of self-shielding effects.
The method is then applied to 232 Th, of main interest for new fuel cycle studies, and is complementary
to higher energy measurements made by D. Karamanis et al. [1] (CENBG). Results obtained for three
target thicknesses will be compared with simulations based on different data bases. Special attention
will be paid to the region of unresolved resonances (>100eV).
697

1. Introduction
At the dawn of XXI century, energy is a crucial problem to study. By 2050, it is predicted the
energy demand will double. In the same time, some greenhouse effect emission scenarios predict
approximately the same increase. The nuclear energy contribution in the energy production represents
4.5%. Nuclear utilisation and further development can be one of the possible responses to the increase
of energy demand and the greenhouse effect limitation.
However, it is necessary to minimise the radioactive waste production such as actinides and
long-lived fission products and this can be obtained by using the 232 Th/ 233 U fuel cycle, in an accelerator
driven system or critical reactors.
But a good prediction for this new way of producing energy is strongly dependent of the material
neutronic properties and more particularly the 232 Th capture cross-section for the fuel.
In the present work, an experimental method is proposed that carries out a validation of available
databases. A lead slowing down spectrometer coupled with a neutron pulsed generator allows to
measure reaction rates (n,γ) or (n,f) over a wide energy range from 0.1 eV to 40 keV for different
thicknesses of material. Experimental data are then compared with precise simulation calculation
using ENDF/BVI, JEF2.2 and JENDL3.2 databases.
The gold results for which the capture cross-section is well know, provides a validation of the
method for three different thicknesses. Tantalum, indium and thorium data for three thicknesses (for
self-shielding effect studies) are presented. The accuracy of the validation method is estimated to be
around 5%. From 0.1 eV to 300 eV, it is shown that predictions of reaction rates using the different
databases agree between themselves and with the experimental data. From 300 eV up to 40 keV
discrepancies between database predictions can be as large as ±20%. The experimental data allow to
either indicating the best databases to be used, or the need to measure again the cross-section in a
certain energy range. For the 232 Th, experimental capture rate data agrees with the prediction of
ENDF/BVI and JEF2.2 bases within 5%.
2. The experimental set-up
2.1 The neutron source
This accelerator has been specially designed for neutronic experiments taking place in the
nuclear reactor MASURCA located at CEA Cadarache Centre (France). In fact, for these experiments
the neutron pulse length must be of the order of the neutron lifetime in a fast reactor. The pulse
intensity must be as big as possible.
The GENEPI (GEnérateur de NEutrons Pulsés Intense) produced fast pulses. The pulse duration
is typically 1.0 µs. Deuterons are produced by a duoplasmatron source, especially studied for a pulsed
use. The frequency can vary from a few Hz up to 5 000 Hz. The deuterons are accelerated at the
maximal energy of 250 keV. The maximum peak intensity is 50 mA. The deuterons are focalised
through a long five meters tube (glove finger) on a deuterium or tritium target. The nuclear reactions
D(d,n) 3 He or T(d,n) 4 He product neutrons with an energy of respectively 2.67 MeV and 14.0 MeV.
This accelerator can produce in the case of a tritium target 5.0 10 6 neutrons/4π per pulse [8].
698

Figure 1. The GENEPI accelerator and the slowing down time spectrometer
Granite support
Deuteron 250 keV
1 µs 1kHz Beam guide 50 mA
Lead Block
Tritium target
1 meter
2.2 The slowing down time spectrometer
The slowing down time spectrometer is an assembly of 46.45 tons of lead with a cubic symmetry
in a relation to the beam axis GENEPI. The neutron production took place in the centre of the lead
block. This block is made up of 8 blocks having the following dimensions: 80 × 80 × 80 cm 3 . Each
block has two channels (10 × 10 cm 2 in section) parallel to the beam axis. They are used both for
handling of the block and for insertion of detectors. In the last case, the dimensions of the holes are
reduced to 5 × 5 cm 2 . Pure lead (99.99%) was chosen to ensure that impurities have negligible effect
on the neutron flux. Impurities of lead are less than 5 ppm, principally silver, bismuth, cadmium,
copper, antimony, tellurium. The lead block is shielded with a cadmium foil to capture the neutron
escaped from the lead block and backscattered by concrete walls, which can deteriorate the energytime
correlation.
In a slowing-down time lead spectrometer, there is a correlation between the neutron time of
flight in the block and its kinetic energy. The scattering mean free path of a neutron in the lead
medium, λ s
= 2.76 cm ,
is about constant over the energy range 0.1 eV to a few tens of keV. The
relation between the slowing down time and the neutron mean energy can be written in this form [2]:
Figure 2. The energy-time correlation from MCNP calculation
E
K
=
(t + t )
0
2
K = 166 ± 1keVµ
s
t
0
= 0.5µ
s
2
699

The K parameter value, function of the neutron masse m n
, the scattering mean free path λ s
and
the medium properties, has been experimentally determined and well understood by MCNP
calculation. The quantity t 0
can be considered as a time correction owing to the fact that the initial
neutron is not created at infinite energy but at energy E 0
= 14 MeV and at velocity v 0
, it suffers
inelastic or (n,2n) reactions before being slowed down only by elastic scattering [4].
2.3 Detectors
2.3.1 Neutrons source monitoring
The reaction on the target produces associated charged particles: α in the case of T(d,n)α
reaction and protons in the D(d,p)T reaction which occurs about as often as the D(d,n) 3 He reaction.
The charged particle is emitted with a 180° angle with respect to the neutron. For 0° neutrons, the
charged particle goes upstream the incident beam, being focused in the glove finger and is bent by the
GENEPI magnet. Two silicon detectors are placed side by side in the vacuum of the magnet chamber.
They detect both the α and the p associated to the d(d,p)n reaction, as fortunately they have the same
magnetic rigidity. One of them is covered with a 10 µm aluminium foil to stop the α particle. We can
measure then through α detection the relative source of 14 MeV neutrons and through p detection of
2.5 MeV neutrons due to D + implantation in the T target.
2.3.2 Neutron spectrum normalisation
The detector used for neutron spectrum normalisation is a classical proportional 3 He-gas counter
belonging to a series of counters developed at ISN Grenoble for fast neutrons spectroscopy in reactor.
We detect the p and/or t produced by the reaction 3 He(n,p)T, Q = 764 keV. It collects energy
deposited by the products of the exothermic reaction. The effective zone is a 6 cm long cylinder of
1 cm in radius. The counter is filled up with 70 mbar of 3 He, 3.3 bars of Argon and 2.5 mbar of CO 2
(quencher gas). The detector is described in Figure 3. It is placed at a symmetrical position to the (n,γ)
detector in relation to the beam axis.
Figure 3. Mechanical structure of the 3 He gas proportional counter detector
25 m stainless steel wire
argon welding
spring
insulator
3x 1.6mm stainless
steel shanks
copper tube argon welding
crimped on the wire
socket stand
gaz lock
gaz entry
Al crimped needles
Al
output socket
wire tension spring
shank stand
copper cap
stainless steel body
700

2.3.3 Neutron flux measurements
• Absolute calibration
The integral flux is measured by activation of nickel foils. These foils are put against the
GENEPI target. The dimensions of the foils are about 5 mm in radius and 0.5 mm in
thickness. Six hours of irradiation with an intensity of 66 µA are enough to reach saturation.
Following reactions are used: 58 Ni(n,2n) 57 Ni with a threshold 13 MeV, 58 Ni(n,np) 57 Co with a
threshold 13 MeV. These activated foils are then counted in the low radioactive laboratory at
ISN.
•
233 U fission detector
The GENEPI neutron pulse, generated at time zero by the reaction T(d,n)α in the lead block
centre region, gives at position a neutron flux φ(E,t, ) which is measured with a
detection system using the exothermic reaction 233 U(n,fission). The reaction rate versus time
is proportional to the quantity φ(E,t, )σ(E). Assuming that the cross-section σ(E) is
known, the measurement of the reaction rate gives an experimental access to the quantity
φ(E,t, ). The fission fragments produced in the reaction 233 U(n,fission) (Q = 180MeV) are
collected by a silicon detector. The 233 U target of 200 µg/cm2 is pure electro-deposited 233 U
on a 200 µm thick aluminium foil. This small detection device is enclosed in lead. Two small
charge-preamplifier are connected to the detectors inside the steel cylinder [4].
2.3.4 Capture rate reaction measurement
A scintillator coupled with a photomultiplier is used for sample (n,γ) reaction rate measurements.
The PM is a XP1911 type from Philips [5]. It was chosen for its reduced dimensions (φ = 19 mm).
Teflon has been chosen for the embase material, to avoid hydrogen and subsequent neutron energy
degradation. PM gain variation has been minimised by adequate decoupling capacitances. CeF 3
scintillator has been chosen for its quick time response time (30 ns) and for its low neutron captures
cross-section. The detection system and the sample embedded in a lead box in order to have a good
reproducibility of the detection geometry. Every two samples, the background has been
systematically measured in order to check the stability of the PM gain. The beam intensity was
adjusted to have a low dead time for each sample (0.1 evts/pulse during the first 10 µs). This detector
and its lead box are shown in Figure 4.
Figure 4. Picture of the PM in its lead box
701

3. Experimental results
The detector signals are recorded with a timing module referenced to the neutron source pulse
with 100 ns precision. For each detector, time spectra are built giving the number of events as a
function of the time of flight of the associated neutron.
3.1 Flux measurements
The 233 U fission rate σ f
φ(t) is obtained with the silicium detector as described above. The α
emission due to 233 U disintegration introduces a background in the fission rate time spectrum.
Fortunately, the energy deposition of fission products and α particles in the silicium can easily be
separated allowing building a pure fission rate time spectrum. Assuming the same efficiency for α
and fission products detection and knowing the neutron production (Ni foil activation, 1.7 10 6 neutron
per pulse), the fission rate σ (n,f)
φ(t) could be normalised per source neutron. In Figure 5 the time
spectrum exhibits at 300 µs a peak corresponding to the well-known 1.7 eV 233 U fission resonance.
Figure 5. Timing spectra of the 233 U detector: 200µg/cm 2
3.2 Neutron flux monitoring
For these measurements we used a 3 He gas detector. The Figure 6 presents a typical time
spectrum obtained with this counter. The (n,p) cross-sections is particularly smooth in the 10 -4 eV to
1.0 MeV energy range and elastic cross-section is negligible under 100 keV. Therefore we have a
good assurance that the flux is the same in these two holes, one of them being used afterwards for
capture rate measurements. The figure shows the good agreement between two measurements made
in symmetrical channels.
702

Figure 6. Time spectrum of 3 He gas detector in two measurements holes of lead block
3.3 (n,γ) experiments
3.3.1 Background study
The photomultiplier with the scintillator and the sample embedded in the lead housing are
inserted in a channel of the lead block. This detection system is very sensitive to gamma rays emitted
by neutron captures in surrounding materials. Therefore background measurements and a good
understanding of its structure are necessary.
The Figure 7 shows a background measurement. The general exponential dependence is due to
the decrease of the neutron flux associated with the scattering process in the lead block.
Super-imposed structures can be seen which are due to neutron capture resonances in various
nuclides. This background has been simulated taking into account all the elements contained in the
detection system itself (CeF 3
scintillator, PM) and in the lead impurities, with proportions as free
parameters. Measured and simulated spectra over a 300 µs time range are showing in Figure 7.
703

Figure 7: Experimental (full line) and simulated reconstructed (dotted line) background time
spectra with energy of the resonances and identification of associated nuclides
3.3.2 Results for gold and thorium targets
Figure 8 shows the normalised reaction rate for thorium and gold. Due to a larger cross-section
gold spectrum is less affected by the background. The large structure that appears at 180 µs
correspond to the 27 000 barns well-known 4.9 eV resonance. In the case of thorium a high
radioactivity level is observed above 140 µs. The peak at 85 µs is due to the 21.8 eV and 23.5 eV
unresolved resonances.
10 -2
Figure 8. Capture rate for gold (left) and thorium (right)
10 -3
10 -4
0 100 200 300 400 500
704

4. Analysis
4.1 Background subtraction method
All spectra are normalised by the counting rate of the 3 He reference monitor. Due to the
self-shielding effects in the target, the background subtraction cannot be directly made. Thus, we
proceed in three steps. Both activation and natural radioactivity induce a constant counting rate over
the whole time range. This constant level for background and target can be measured when the
neutron flux becomes negligible i.e. for time bigger than 2 ms. The first steps consist in subtracting
this level for each spectrum. In a second step, measured background must be corrected as it is higher
than its real contribution in the presence of the target, γ-rays due to captures in the surrounding
materials being slightly absorbed in the target. This correction is evaluated according to the density
and the thickness of the target. The last step consists in taking into account the neutron flux
perturbation induced by the target. This correction factor is obtained by the ratio of simulation
performed with and without target.
As the last this corrected background is subtracted from the spectrum obtained in the first step. It
must be noticed that the two last corrections are second order effects.
4.2 Monte Carlo simulation
We use the MCNP/4B code for simulations [7]. This code allows a very detailed description of
the experimental set-up: detectors, generator components, lead block and concrete walls. The reaction
rates σ(n,γ)φ(t) are calculated in order to be directly compared with experimental data. Three different
databases have been used: ENDF/B-VI, JENDL3.2 and JEF2.2.
4.3 Simulation and experimental results
Simulation provides capture rate per source neutron as a function of time. Both the simulated and
experimental time spectra are converted into energy spectra, by means of the time-energy correlation
described in a previous section. The simulations to experiment ratios are calculated for each energy
bin, and are presented in Figures 9 to 12. We present 4 targets results: Au in order to validate the
procedure, tantalum, indium and at last, we present the thorium target results of main interest for new
fuel cycle studies.
705

Figure 9. ENDF/B-VI and JEF2.2 simulation to experiment ratio for 1250 µm,
500 µm and 125 µm Gold samples in the energy range from 0.1 eV to 40 keV.
The large grey box around C/E = 1 correspond to an uncertainty of 5%
Gold:
First of all, in order to validate the
procedure, we used gold for which
capture cross-section is well known. The
Figure 9 shows results for 1 250 µm,
500 µm and 125 µm gold targets. A
good agreement is found from 0.2 keV
to 40 keV with ENDF/B-VI and JEF2.2
capture cross-section databases. However,
a noticeable discrepancy is observed
between 2 keV to 6 keV, which remains
unexplained.
1.4
1.3
1.2
1.1
1
0.9
0.8
0.7
0.6
10 -1 1 10 10 2 10 3 10 4 10 5
1.4
1.3
1.2
1.1
1
0.9
0.8
0.7
0.6
10 -1 1 10 10 2 10 3 10 4 10 5
1.4
1.3
1.2
1.1
1
0.9
0.8
0.7
0.6
10 -1 1 10 10 2 10 3 10 4 10 5
Figure 10. ENDF/B-VI, JEF2.2 and JENDL3.2 simulation to experiment ratio for 2000 µm,
200 µm and 100 µm Tantalum samples in the energy range from 0.1 eV to 40 keV
Tantalum:
In the case of 2 000 µm, 200 µm
and 100 µm tantalum targets, the resolved
resonance zone from 1 eV to 200 eV
is correctly described. However, a deficit
is observed for neutron energies lower
than 1 eV for the thickest targets.
A good agreement between simulation
and experimental results is found
with JENDL3.2 database for
300 eV

Figure 11. ENDF/B-VI, JEF2.2 and JENDL3.2 simulation to experiment ratio for 2 000 µm,
500 µm and 300 µm indium samples in the energy range from 0.1 eV to 40 keV
Indium:
In the case of 2 000 µm, 500 µm and
300 µm Indium targets, ENDF/B-VI,
JEF2.2 and JENDL3.2 simulation to
experiment ratio give a good agreement
in the range from 0,1 eV to 1keV (see
Figure 11). For 1 keV

5. Conclusion
The neutron capture cross-section profile of various targets (Gold, Tantalum, Indium and
Thorium) have been measured with a slowing down lead spectrometer in the neutron energy range
from 0.1 eV to 40 keV with a precision of 5%. The experimental results are compared to Monte Carlo
simulations (MCNP/4B) code using ENDF/B-VI, JEF2.2 and JENDL3.2 databases. Measurements on
the well-know gold nucleus are well reproduced by simulation. The agreement with different targets
thickness validates our method, and shows that the self-shielding effect is well taken into account by
MCNP. For tantalum and indium targets, a discrepancy between experiment and simulation is
observed for neutron energy greater than 300 eV, in the region of the unresolved resonances. For
thorium targets, the JENDL3.2 cross-section seems under evaluated by 10% in the energy range from
300 eV to 3 keV.
In conclusion, the lead spectrometer appears to be a very useful tool, allowing quick
cross-section validation and transmutation rates evaluation.
REFERENCES
[1] D.Karamanis et al., Neutron Radiative Cross-section of 232 Th in the Energy Range from 0.06 to
2 MeV, Proceedings of the 6th OECD/NEA Information Exchange Meeting on Actinide and
Fission Product Partitioning and Transmutation, Madrid, Spain, 11-13 Dec. 2000, OECD
Nuclear Energy Agency, Paris, France, (2001).
[2] R.E. Slovacek et al., 238 U(n,f) Measurements Blow 100 keV. Nuclear Science and Engineering,
62 (1997) 455.
[4] European Commission, Neutron Driven Nuclear Transmutation by Adiabatic Resonance
Crossing, TARC, Final Report, Euratom, EUR1911-EN, (1999).
[5] Philips, Photomultiplier Tubes, Technical report.
[6] Rene Brun and Fons Rademakers, ROOT – An Object Oriented Data Analysis Framework,
Proceedings AIHENP’96 Workshop, Lausanne, Sept. 1996, Nucl. Inst. & Meth. in Phys. Res. A
389 (1997) 81. See also: http://root.cern.ch/.
[7] MCNP, A General Monte Carlo Code for Neutron and Photon Transport, J.F. Briesmester Ed.,
LA-12625-M, (1993).
[8] J.L. Belmont, J.M. De Conto, L'accélérateur “GENEPI”, conception, technologie, caractéristiques,
Rapport Interne ISN00.77, July 2000, ISN-CNRS, France.
708

DOUBLE DIFFERENTIAL CROSS-SECTION FOR PROTONS EMITTED
IN REACTIONS OF 96.5 MeV NEUTRONS ON ENRICHED 208 Pb TARGETS.
F.R. Lecolley, C. Le Brun, J.F. Lecolley, M. Louvel, N. Marie
LPC, ISMRa et Université de Caen, CNRS/IN2P3, France
P. Eudes, F. Haddad, M. Kerveno, T. Kirchner, C. Lebrun
SUBATECH, Université de Nantes, France
A. Ataç, J. Blomgren, N. Olsson
INF, Uppsala University, Sweden
P.U. Renberg
TSL, Uppsala University, Sweden
X. Ledoux, Y. Patin, P. Pras
DPTA/SPN CEA, Bruyères-le-Châtel, France
F. Hanappe
ULB, Brussels, Belgium
L. Stuttgé
IreS Strasbourg, France
Abstract
Transmutation techniques involve high-energy neutrons created by the proton-induced spallation of a
heavy target nucleus. The existing nuclear data libraries developed for the present reactors go up to
about 20 MeV, which covers all available energies for that application; but with a spallator coupled to
a core, neutrons with energies up to 1-2 GeV will be present. Although a majority of the neutrons will
have energies below 20 MeV, a small fraction at higher energies has still to be characterised. Above
200 MeV, direct reaction models work reasonably well, while at lower energies nuclear distortion
plays a non-trivial role. This makes the 20-200 MeV region the most important for new experimental
cross-section data.
Very little high-quality neutron-induced data exist in this energy domain. For (n,xp) reactions, different
experimental programmes have been run at Los Alamos [7] and TRIUMF [1] facilities but with limited
coverage in particle energy and angle. Better coverage has been obtain by the Louvain-la-Neuve Group
up to 70 MeV [9].
Due to this particular lack of data above 70 MeV and in the framework of the European concerted
action “Lead for ATD” and the HINDAS project (see J.P. Meulders contribution in these
proceedings), in March’99 we performed an experiment in order to measure double differential crosssections
for protons and other light charged particles emitted in reactions of 96.5 MeV neutrons on
enriched 208 Pb targets, at the neutron facility of The Svedberg Laboratory (TSL), Uppsala, Sweden [2].
709

1. Experimental set-up
The charged particles (p, d, t, 3 He and alpha) were detected using the MEDLEY device [4] which
allows to measure continuous energy distributions in the forward direction (10°-80°). At larger angles,
due to the relatively low intensity of the neutron beam and due to the weak estimated cross-sections,
only the low-energy part of the spectra could be measured (E p < 40 MeV at θ = 160°). In order to
improve the counting rate at backward angles and to measure the high-energy part of the proton
spectra, we used a multi-target box together with the two arms of SCANDAL [5]. This set-up covered
the angular range 10°-140°.
The MEDLEY detector set-up is installed inside a cylindrical scattering chamber of 100 cm
diameter. It consists of eight detector telescopes which are mounted inside the vacuum chamber and
placed every 20 degrees. They cover scattering angles ranging from 20 up to 160 degrees. In order to
obtain a good separation between the different particles (p, d, t, 3 He and alpha) over a large dynamic
range, i.e. from a few MeV alpha particles up to 100 MeV protons, each telescope is composed of
three detectors: two silicon surface barrier detectors and one CsI(Tl) crystal. The front detectors (dE 1 )
are either 50 or 60 mm thick, while the second ones (dE 2 ) are 400 or 500 •m. The CsI(Tl) crystal, used
as E detectors, are long enough to stop 100 MeV protons. Using the well-known dE1-dE2-E method,
we are able to identify with no ambiguities light charged particles.
SCANDAL, SCAttered Nucleon Detection AssembLy, is a CsI hodoscope with auxiliary
detectors: drift chambers used to determine the proton trajectory and plastic scintillators used to trigger
the acquisition. SCANDAL is designed for protons and neutrons in the 30-130 MeV interval.
While the proton energy threshold is around 10 MeV for MEDLEY, for SCANDAL this
threshold is above 30 MeV because particles have to go through different materials before reaching the
CsI(Tl) detectors.
2. Preliminary results and comparison with theoretical prediction
The Double Differential Cross-Section (DDCS) for protons emitted in reactions of 96.5 MeV
neutrons on lead targets are presented in Figure 1.
We observe that in the energy region covered by both set-up there is a good agreement between
MEDLEY and SCANDAL measurements despite a small underestimation of high energy proton
production with MEDLEY at forward angles. This experimental effect is associated to the relative low
thickness of the second detector which induces detection efficiency lower than 100% for the high part
of the proton spectrum.
Comparisons with theoretical predictions are shown in Figures 1 and 2. A good agreement is
obtained either with the GNASH-CEA [8] or the MINGUS [6] or the CUGNON [3] calculations, but
there are still several problems (see Figure 3 for example with a linear y-axis) which are listed below:
• With the MINGUS calculation, the low energy part of the spectra is underestimated at
forward angles and overestimated above 60 degrees.
• The GNASH-CEA calculation always underestimates the DDCS at low energies.
• The CUGNON calculation which is based on intra-nuclear cascade and optimised for incident
energy above 200 MeV, works reasonably, except at low energies for the evaporative
component.
710

3. Conclusion
Double differential cross-section measurements for protons emitted in reactions of 96.5 MeV
neutrons on enriched lead targets were performed using the TSL facilities.
Preliminary results were compared with different theoretical calculations: they have reasonable
predictions nevertheless they have to be improved in order in particular to reproduce the low energy
part of the proton spectra.
This conclusion will be reinforced or cancelled
• By doing the same analysis on lead target for the other light charged particles (d, t, 3 He and
alpha) measured with the MEDLEY set-up.
• By studying DDCS with iron target (the dedicated experiment has been performed in May
2000 at TSL) and an uranium one (the experiment is planned in autumn 2001 at TSL).
Figure 1. Preliminary results of double differential cross-sections for protons
emitted in reactions of 96.5 MeV neutrons on enriched 208 Pb
• Experimental Data : SCANDAL (black circle) and MEDLEY (black square).
• Theoretical Prediction : MINGUS (open circle) and GNASH-CEA (open triangle).
• From left to right and top to bottom, angles are ranging from 20 up to 120 degrees with
20 degrees interval.
711

Figure 2. Preliminary results of double differential cross-sections for protons
emitted in reactions of 96.5 MeV neutrons on enriched 208 Pb
• Same as Figure 1.
• Theoretical Prediction: CUGNON (open star).
712

Figure 3. Preliminary results of double differential cross-sections for protons
emitted in reactions of 96.5 MeV neutrons on enriched 208 Pb
Same as Figure 1 with linear y-axis.
713

REFERENCES
[1] Alford W.P. and Spicer B.M., 1998, Nucleon Charge-exchange Reactions at Intermediate
Energy, Advances in Nuclear Physics 24, 1.
[2] Condé H. et al., 1990, A Facility for Studies Neutron Induced Reactions in the 50-200 MeV
Range, Nucl. Instr. Meth. A292, 121.
[3] Cugnon J. et al., 1997, Improved Intranuclear Cascade Model for Nucleon-nucleus Interactions,
Nucl. Phys. A, 620, 475-509.
[4] Dangtip S. et al., 2000, A Facility for Measurements of Nuclear Cross-sections for Fast Neutron
Cancer Therapy, Nucl. Instr. Meth. Phys. Res, A, 452, 484-504.
[5] Klug J. et al., 2000, SCANDAL – A Facility For Elastic Neutron Scattering Snakes in. the
50-130 MeV range, (to be published).
[6] Koning A. et al., 1997, Phys. Rev. C, V56.
[7] Rapaport J. and Sugarbaker E., 1994, Isovector, Excitations in Nuclei, Annu. Rev, Nucl. Part. Sci.
44, 109.
[8] Romain P. et al., CEA/DAM, Private Communication.
[9] Slypen I. et al., 1994, Charged Particles Produced in Fast Neutron Induced Reactions on 12C
in the 45-80 MeV Energy Range, Nucl. Instr. Meth., A337, 431.
714

MEASUREMENTS OF PARTICULE EMISSION SPECTRA IN PROTON INDUCED REACTIONS
OF INTEREST FOR THE DEVELOPMENT OF ACCELERATOR DRIVEN SYSTEMS
N. Marie, C. Le Brun, F.R. Lecolley, J.F. Lecolley, F. Lefèbres, M. Louvel, C. Varignon
LPC de Caen, Université de Caen, IN2P3-CNRS/ISMRA – France
Ph. Eudes, S. Auduc, F. Haddad, T. Kirchner, C. Lebrun
SUBATECH, Université de Nantes, IN2P3-CNRS, École des Mines de Nantes, France
Th. Delbar, A. Ninane
Institut de Physique Nucléaire, Belgium
F. Hanappe
FNRS et Université Libre de Bruxelles, Belgium
X. Ledoux, Y. Patin, Ph. Pras
DPTA/SPN, CEA, France
L. Stuttge
IreS, IN2P3-CNRS, Strasbourg, France
Abstract
In the framework of the concerted action “Lead for ADS” program, we have measured the double
differential cross-sections of neutrons produced in reactions induced by a proton beam on a lead
target at 62.5 MeV. The experiment was performed on the S-line of the CYCLONE facility in
Louvain-la-Neuve. The neutrons were detected using DEMON counters and their energy was derived
from the time-of-flight technique.
715

1. Introduction
For many accelerator driven system projects [1,2], lead has been chosen as a representative
spallation target material. Therefore, Pb (p, X n), Pb (p, X p), Pb (p, X lcp) double differential crosssections
(DDCS) are required with high priority for the development of simulation codes. These codes
are used for feasibility studies and optimisation of such hybrid systems in which complex combinations
of nuclear processes are involved. Combined with complementary Pb (n, X n), Pb (n, X p), Pb (n, X lcp)
DDCS, these data represent the best test for evaluating the global capabilities of the models. In addition,
such data provide important constraints which allow the predictive power of the codes to be improved
in the 20-150 MeV energy range.
In this context and in the framework of the Concerted Action “Lead for ADS” programme, we
measure the DDCS of neutrons and light charged particles (p, d, t, 3 He, 4 He) produced in reactions
induced by a proton beam, impinging on a lead target at 62.5 MeV. In this contribution we present
results concerning only the neutrons.
Figure 1. Simulation with GEANT of the experimental set-up geometry (see text)
Target
position
716

2. Experimental set-up
The experiment was performed on the S-line of the CYCLONE facility in Louvain-la-Neuve.
The lead target is 10.7 mg/cm 2 thick and the neutrons are detected using five DEMON large volume
NE213 liquid scintillator counters [3]. The following table gives, for each counter, its angle theta
relative to the beam and the distance between the target and its entrance window.
DEMON counter theta (°) d (mm)
1 120 2 960
2 80 2 507
3 55 3 039
4 35 3 887
5 24 5 347
Each detector is surrounded by a lead cylinder installed inside a “BOMBARDE” barrel filled
with paraffin and boron - materials that are efficient shields against background neutrons.
Figure 2. Compilation of the measured DEMON detector efficiencies (symbols)
compared to the predictions of the KSU code (black line)
Due to the necessity of shielding the DEMON counters from the very high radiation background
resulting from the proton beam dump, a wall made of concrete and paraffin is also built in the
experimental area. Taking into account the experiment configuration (various materials and
dimensions), the floor-space and the weight of concrete blocks, the wall dimensions is optimised
performing GEANT simulations. The final wall geometry divides by a factor of twenty the
background of the most exposed DEMON counter to the parasite neutron flux, resulting in a signalto-noise
ratio of 2.1. The Figure 1 illustrates the efficiency of the shielding wall. It presents part of the
geometry of the experimental set-up: the faraday cup placed at the end of the beam line and four
DEMON counters imbedded in BOMBARDE barrels, and the shielding wall built between the
forward DEMON counter and the beam dump. Dashed lines symbolise neutrons escaping from the
717

eam dump. We observe that the majority of the neutrons emitted in the direction of the counters are
stopped by the wall or deviated in their trajectories.
3. Data analysis
We discriminate neutrons from gammas by pulse shape analysis of the photomultiplier output.
For each neutron, the time-of-flight is derived from the start given by the DEMON counter, and the
stop given by the following beam high frequency signal (period = 54 ns). The same procedure is
employed for gammas so that the time-of-flight spectra are calibrated with the reference peaks
associated to gammas. In order to detect without ambiguity the lowest energy neutrons, nine beam
bursts out of ten are suppressed. Neutron energies are derived from their time-of-flight, taking into
account the depth at which the particle interacts inside the detector. This depth is estimated using an
iterative procedure since it depends on the detection efficiency, which is itself a function of the
neutron energy. The DEMON detector efficiency can be found in [4] and it is shown in Figure 2.
During the experiment, attention was paid to alternatively collect data with Pb targets and with
blank-targets in order to be able to subtract the background noise. The acquisition dead time is also
kept under twenty percent and a correction for this effect is applied to the data. For the cross-section
calculation, the number of incident protons is derived from the intensity of the beam measured using
the faraday cup. The detector efficiency and solid angle, as well as the orientation of the target are
also taken into account in deriving the absolute normalisation factor.
Figure 3. Double differential cross-section of neutrons produced in reactions
induced by a proton beam impinging on a lead target at 62.5 MeV
718

4. Results
Figure 3 presents neutron energy spectra obtained at five different angles. The energy uncertainty
is derived from the length uncertainty of the time-of-flight path (±1 mm), combined with the
uncertainty on the depth at which the neutron interacts inside the scintillator (±1 cm) and the
electronic chain resolution. The resulting energy uncertainty increases smoothly with the neutron
energy from 0.03 MeV to 4.2 MeV at 62.0 MeV. In order to calculate the cross-section uncertainty,
we take into account the detector efficiency uncertainty which is lower than 5.8% over the entire
energy range. The contribution of the statistical uncertainty to the relative total uncertainty is
estimated to be lower than 2% for energies smaller than 30 MeV, it increases up to 4.4% for an
energy value of 60 MeV at a detection angle of 80°, and it reaches a maximum value of 16% at the
most backward angle, for the larger energy. Those values result in a total relative uncertainty of the
cross-section lower than 5.6% for energies smaller than 30 MeV, it increases up to 6.5% for an energy
value of 60 MeV at a detection angle of 80°, and it reaches a maximum value of 17% at the most
backward angle and for the larger energy. Due to the logarithmic representation the associated error
bars are not visible on the figure.
5. Conclusion
We present neutron double differential cross-sections measured at five different angles for the
Pb(p, X n) reaction, at 62.5 MeV. These results will contribute to the extension up to 150 MeV of
evaluated nuclear data libraries, which are a combination of experimental and calculated data. Such a
database is planned to be implemented in different simulation codes which are used for the
conception of the future hybrid systems.
REFERENCES
[1] C.D. Bowman et al., Nucl. Instr. Meth. A320 (1992)336.
[2] C. Rubbia et al., preprint CERN/AT/95-44/ET (1995).
[3] I. Tilquin et al., Nucl. Instr. Meth. A365(1995)446.
[4] C. Varignon, Thèse de Docteur en Physique Nucléaire, soutenance déc. 99.
719

INTERMEDIATE ENERGY NEUTRON-INDUCED FISSION
CROSS-SECTIONS FOR PROSPECTIVE NEUTRON PRODUCTION TARGET IN ADS
V.P. Eismont, I.V. Ryzhov, A.N. Smirnov, G.A. Tutin
V.G. Khlopin Radium Institute, 2oi Murinskiy Prospect 28, Saint-Petersburg 194021, Russia
H. Condé, N. Olsson, A.V. Prokofiev*
Department of Neutron Research, Angström Laboratory, Uppsala University,
Box 525, S-751 20 Uppsala, Sweden
Abstract
Up-to-date status is considered of the experimental database on neutron-induced fission cross-sections
of tantalum, tungsten, lead, mercury, gold, and bismuth nuclei in the neutron energy range from the
fission threshold to 175 MeV. The perspective of creating a more complete database is discussed,
including (n,f) cross-sections for separated isotopes of lead and tungsten.
* On leave from V.G. Khlopin Radium Institute, St. Petersburg, Russia.
721

1. Introduction
Intermediate-energy fission data are of interest not only for fundamental physics, but also for
applied nuclear research. First of all it is connected with problems of accelerator driven systems
(ADS) for power production and transmutation of long-lived radioactive waste (see e.g. [1]). Such
elements as Ta, W, Hg, Pb and Bi are either already used as neutron producing target materials or
considered as potential candidates. Since fission and spallation reactions at intermediate energies are
the main reaction channels of neutron interactions with heavy nuclei, they have the most practical
significance, but are poorly investigated. Fission reaction contributes to the generation of the neutron
field in the target-blanket assembly, as well as to the production of radionuclides and chemically toxic
products in the target. For relatively light nuclei, such as Pb and Bi with fission cross-section of only
a few percent of the total reaction cross-section, the residual activity of the fission products with high
energy release and, often, long half-lives, is expected to be significant. It is estimated that the
contribution of the fission products to the overall residual activity of a lead target irradiated by
1.6-GeV protons may be as much as 10-15% for cooling times of about one year [2]. At present,
theoretical description of sub-actinide nuclei fission cannot match the practical needs. For example,
the nat W(p,f) cross-section predicted by the LAHET code was found to be about 20 times lower than
the experimental results [3]. If one bears in mind the insufficient predictive power of available
nuclear reaction models, especially with respect to fission, it is possible that the real residual activity
may be significantly different.
Intermediate-energy fission data are also of interest for nuclear standards, and particle beam
monitoring required for other applications as neutron cancer therapy, shielding of accelerators,
cosmic studies, thermonuclear synthesis etc. Furthermore, due to the insensitivity to low energy
neutrons, the 209 Bi(n,f) cross-section has been approved by IAEA as a secondary standard for neutron
flux determination at intermediate energies [4].
As a response to the outlined needs, V.G. Khlopin Radium Institute and Uppsala University
perform a joint program of (n,f) cross-section measurements for sub-actinides in the energy region
between 20 and 180 MeV. Earlier, results for the absolute and relative (n,f) cross-sections of 238 U, 209 Bi
and 208 Pb have been published [5-7]. In the framework of ISTC project #540 the measurements for
208
Pb and 209 Bi were continued with better experimental conditions, including new measurements on
181
Ta, nat W, nat Hg, 197 Au and nat Pb. The measurements on gold were included because of their methodical
and theoretical importance. Preliminary results for the above listed sub-actinides have been published
recently [8-11]. In the present work the results of further processing of the data are given and the
status of the 209 Bi (n,f) cross-section standard is discussed.
The prospects for creating of more complete database are considered, including (n,f) crosssections
for separated isotopes of lead and tungsten, specifically for 208 Pb and 184 W , which crosssections
are included in the High Priority Request List of nuclear data for the nucleon energy region
up to 200 MeV [12]. These data are needed for the development of adequate nuclear fission models,
as well as computer codes for ADS. The needs of fission cross-section measurements for the above
mentioned nuclides are stressed, not only with neutrons, but also with protons, in the same projectile
energy region. Comparison of proton- and neutron-induced fission cross-sections [13], carried out on
a common physical basis [14], gives added credence to the experimental database.
722

2. Up-to-date status of the (n,f) cross-sections of sub-actinides
The (n,f) cross-sections of 181 Ta, nat W, nat Hg, 197 Au, 208 Pb and nat Pb published in [9-11] have been
obtained with the assumption that the fraction of the total fission events induced by full energy
neutrons is equal for the studied nuclides and for the monitor reaction 209 Bi(n,f). In this work we have
calculated the fraction of the total fission events induced by full energy neutrons for each reaction
under study. For this purpose the TOF spectra of fission events have been simulated using the
experimental neutron spectra [15-19] and the given time parameters of the proton beam. The final
cross-sections and the fractions of the full energy fission events have been obtained as a result of an
iteration procedure with the initial cross-sections taken from [4,9]. The results of the calculations
carried out for the 209 Bi(n,f) and nat W(n,f) reactions at a peak neutron energy of 96 MeV are shown in
Figure 1 together with the neutron spectrum from the 7 Li(p,n) reaction measured by Nakao et al. [15]
at the similar incident proton energy.
Figure 1. (a) Neutron spectrum from 7 Li(p,n) reaction at 100 MeV proton energy [15];
(b) and (c) TOF spectra of 209 Bi(n,f) fission events at
12.8 m and 2 m flight distances, correspondingly;
(d) TOF spectrum of nat W(n,f) fission events at 2 m flight distance.
Open dots in (b), (c), and (d) are experimental TOF spectra. Solid curves are calculated TOF
spectra. Filled areas under dashed curves are calculated fractions
of fissions induced by low energy tail neutrons.
The results for (n,f) cross-section ratios are given in Table 1. The absolute (n,f) cross-sections
obtained using the 209 Bi(n,f) cross-section as a standard [4] are shown in Figure 2 together with fits
according to Equations 1) and 2) below and our previously reported data for 208 Pb [6].
723

Table 1. Relative neutron-induced cross-sections
E n
(n.f) cross-section ratio
MeV
nat Pb/ 209 Bi
208 Pb/ 209 Bi
nat Hg/ 209 Bi
197 Au/ 209 Bi
nat W/ 209 Bi
181 Ta/ 209 Bi
35 0.166 ± 0.079

Figure 2. The 181 Ta(n,f), nat W(n,f), nat Hg(n,f), 197 Au(n,f), 208 Pb(n,f) and nat Pb(n,f) cross-sections. Solid
and dashed curves are the fits made with the use the formulae 1, 2 (see the text below)
ission cross-section (mb)
10 2 nat
Pb(n,f)
this work
10 -2
10 -3
10 -4
197 Au(n,f)
10 1
10 0
[20] 1996
[21] 1955
[22] 1984
param. 1
param. 2
208
Pb(n,f)
this work
[5] 1996
param.1
param.2
nat
W(n,f)
this work
param. 1
param. 2
10 1
nat
Hg(n,f)
10 0
10 -1
10 -1
10 -2
10 -3
10 50 100
this work
[20] 1996
[21] 1955
param. 1
param. 2
10 50 100
this work
[21] 1955
param. 1
param. 2
181
Ta(n,f)
this work
param. 1
param. 2
10 50 100
Neutron energy (MeV)
The available experimental data on the nat Pb(n,f) and 197 Au(n,f) cross-sections [20] and the data of
the present work are in qualitative agreement. There are, however, systematic discrepancies between
these data sets in the neutron energy regions below about 50 MeV and above about 100 MeV for
197
Au(n,f) and practically in the whole energy region for nat Pb(n,f). This fact, together with the
comparison for 197 Au/ 209 Bi and nat Pb/ 209 Bi cross-section ratios performed in [10], leads to a suggestion
that some background may not have been taken fully into account in the LANSCE measurements of
the nat Pb(n,f) and 197 Au(n,f) cross-sections. The low-energy nat Pb(n,f) data of the present work are
compatible with the upper limits of the cross-section obtained by Vorotnikov [22]. The old data of
Reut et al., Goldanskiy et al., and Dzhelepov et al., [21] for nat Pb(n,f), 197 Au(n,f), and nat W(n,f) crosssections
at a neutron energy of 120 MeV are considerably (two-three times) higher than our data in
the same energy region.
The fits in Figure 2 were made using the same formulae that have been used for parameterization
of the 209 Bi(n,f) cross-section [4]:
σ nf
above about 70 MeV [23], and
= p ⋅( 1− exp( p ⋅( E − p ))) (1)
11 12 n 13
σ nf
= p21 + p22 ⋅ E n
+ p23
⋅ E n
2
exp( ln ln )
(2)
725

elow about 70 MeV [24] , where σ nf is the (n,f) fission cross-section (mb), E n is the neutron
energy (MeV) and p ij
are variable parameters. The parameters p ii
as functions of the parameter Z 2 /A of
the compound nuclei are available upon request.
Taking into account the uncertainties coming from the experimental technique and the crosssection
standard we can state that the accuracy of the presented data is about 20% for most nuclei in
the energy range above 75 MeV and 30-50% below 75 MeV. As the neutron energy and the target
atomic number are decreased, the uncertainty is increased.
Figure 3. Experimental data on the 209 Bi(n,f) cross-sections
Fission cross-section (mb)
80
70
60
50
40
30
20
10
TFBC (Uppsala, 1994, 1996)
IC (Los Alamos, 1995)
FGIC (Uppsala, 1996, 1999)
TFBC (Uppsala, 1999, 2000)
Recommended (IAEA, 1997)
209 Bi(n,f)
0
20 40 60 80 100 120 140 160 180 200
Neutron energy (MeV)
As it was mentioned above, the present status of the experimental data on the 209 Bi(n,f) crosssection
is of particular interest, because it is a new standard in the energy region above 50 MeV. All
experimental data available in the energy range of interest are given in Figure 3 along with the
209
Bi(n,f) cross-section parameterization from [4]. It is obvious that there is a discrepancy between the
recommended parameterization and some experimental results published more recently [8,9].
Specifically, our data obtained at the neutron energies 96 MeV and 133 MeV (solid circles) are
systematically lower than the recommended curve and do not fall into the confidence interval (10%)
stated in [4]. To understand whether (or not) this discrepancy will eventually result in a
recommendation to change the standard, a closer look at the quality of the experimental data must
first be undertaken. The data under consideration have been obtained with the use of three fission
fragment detectors: thin-film breakdown counters (TFBC), a conventional parallel plate ionization
chamber (IC), and a Frisch-gridded ionization chamber (FGIC).
The TFBCs are insensitive to the background radiation and offer excellent timing
characteristics [25]. Previous data obtained with the use of the TFBC [5,6] and the FGIC [7], together
with earlier data [21,22] have been used as a basis for the parameterization, recommended as a
standard [4]. However, recently it was found out that the decomposition procedures applied in
[5,6,11] to the TOF spectra of fission events lacked accuracy at some neutron energy points. We
suppose, that a more sophisticated background approximation (see e.g. [7]) has to be used to extract
from TOF spectra a number of fission events induced by “peak” neutrons. Solid triangles in Figure 3
show our data from [9,11] corrected due to a more accurate decomposition procedure as well as data
726

obtained more recently. It is seen that these data deviate systematically from the parameterization [4]
and become closer to our data obtained with the FGIC [8,9].
An ionization chamber (IC) is ideally suitable for (n,f) cross-section measurements at the
currently available high-energy neutron beams, because this device offers nearly 100% detection
efficiency with no limitations on the fissile target dimensions. It should be noted, however, that the
energy spectrum of fission fragments is contaminated by light charged particles arising in upstream
material from energetic neutrons. This background makes the determination of the fission fragment
yield difficult. As the incident neutron energy increases, the situation becomes more complicated
(particularly for sub-actinides) due to larger overlap between fission and background spectra. We
suppose that data obtained with the simple ionization chamber [20] may be subject, especially at high
neutron energies, to some systematic errors caused by the background problems.
The FGIC not only incorporates the main advantages of conventional parallel plate ionization
chambers, but also offers some extra ones for (n,f) cross-section measurements. The key advantage is
that FGIC allows discrimination against background charged particles [8]. The principle of so called
angular discrimination lies in the fact that fission fragments and light charged particles give different
combinations of anode and grid signals, and thus may be separated from each other by off-line
processing. Taking also into account that an accurate decomposition procedure has been applied to
the TOF spectra in [8,9], we consider the data obtained with the FGIC as the most reliable at present.
All aforesaid gives some grounds to expect changes of the standard in the future, but in the
present work all data on (n,f) cross-sections for sub-actinides are given relative to the old standard.
3. Prospects for advancement of the existing experimental data base
Further development of the experimental techniques is needed to improve the quality of the data.
This refers both to the characteristics of neutron beam and to the fission fragment detectors.
To increase the number of fissile targets to be irradiated simultaneously, the new FGIC has been
designed and manufactured at the KRI within the framework of ISTC project #1309. The chamber
consists of seven units. Each unit constitutes a twin Frisch-gridded ionization chamber with a
common cathode. By this means 14 different targets may be irradiated simultaneously.
Due to the insensitivity of TFBCs to light ionizing particles it is possible to carry out the (p,f)
and (n,f) cross-section measurements under comparable geometrical conditions. Experiments can be
done with a broad proton beam passing through the target and the TFBC [26]. This makes it possible
to reduce the uncertainty in the comparison analysis of data on the proton and neutron-induced fission
cross-sections.
Combined analysis of (n,f) and (p,f) cross-sections is of special interest for studies of the fission
process. The quantitative comparison of these cross-sections carried out in [13] revealed the
following empirical dependence of the cross-section ratios on the Z 2 /A parameter of the target
nucleus:
σ pf
/σ nf
= exp [k(37- Z 2 /A)] (3)
where k is a function of energy (k>0 for Z 2 /A≤37, and k = 0 for Z 2 /A>37). This dependence was
explained in terms of the fissility of the nucleus (P f
= σ pf
/σ in
) which is defined by the fission and
evaporation widths: P f
= Γ f
/Γ f
+ Γ n
+…
727

Since for sub-actinides Γ f
/Γ n

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p. 1223.
[5] V.P. Eismont, A.V. Prokofiev, A.N. Smirnov, K. Elmgren, J. Blomgren, H. Condé, J. Nilsson, N.
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of 208 Pb, 209 Bi and 238 U in the Intermediate Energy Region, Phys. Rev. C53, 2911 (1996).
[6] V.P. Eismont, A.V. Prokofiev, A.N. Smirnov, J. Blomgren, H. Condé, K. Elmgren, J. Nilsson,
N. Olsson, and E. Ramström, Measurements of Neutron-induced Fission Cross-sections of
Heavy Nuclei in the Intermediate Energy Region, Proc. 2nd Int. Conf. on ADTTA, Kalmar,
Sweden, June 3-7, 1996, ed. H. Condé: Uppsala University, 1997, Vol. 2, pp. 606-612.
[7] V.P. Eismont, A.V. Kireev, I.V. Ryzhov, G.A. Tutin, H. Condé, K. Elmgren and S. Hultqvist,
Measurement of the Neutron-induced Fission Cross-sections of 209 Bi at 45 and 73 MeV, Proc.
2nd Int. Conf. on ADTTA, June 3-7, 1996, Kalmar, Sweden, ed. Henri Condé: Uppsala
University, 1997, Vol. 2, pp. 618-623.
[8] V.P. Eismont, A.V. Kireev, I.V. Ryzhov, S.M. Soloviev, G.A. Tutin, H. Condé, K. Elmgren,
N. Olsson, and P.-U. Renberg, Neutron-induced Fission Cross-section Ratios of 209 Bi to 238 U at
75 and 96 MeV, The 3rd International Conference on Accelerator Driven Transmutation
Technologies and Applications, Praha, Czech Republic, 1999, (available as CD-ROM, paper
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[9] V.P. Eismont, Measurements of Neutron Induced Fission Cross-sections in Energy Region
15

[10] V.P. Eismont, A.V. Prokofiev, A.N. Smirnov, S.M. Soloviev, H. Condé, K. Elmgren, and
N. Olsson, Neutron-induced Fission Cross-sections of nat Pb and 197 Au in the 35-180 MeV Energy
Region, Proc. 3rd International Conference on Accelerator Driven Transmutation Technologies
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[11] V.P. Eismont, A.V. Prokofiev, A.N. Smirnov, H. Condé, K. Elmgren, N. Olsson, P.-U. Renberg,
Neutron-induced Fission Cross-sections of 181 Ta, nat W, 197 Au, nat Hg, 197 Au, 208 Pb, nat Pb, and 209 Bi Nuclei
in the Energy Range 35-175 MeV, TSL Progress Report 1998-1999, Ed. A. Ingemarsson, Uppsala
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M. Österlund, F.P. Brady, and Z. Szeflinski, A Facility for Studies of Neutron-induced
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T. Nakamura, Su. Tanaka, S. Meigo, H. Nakashima, Sh. Tanaka, and N. Nakao. Characterization
of a 40-90 MeV 7 Li(p,n) Neutron Source at TIARA Using a Proton Recoil Telescope and a TOF
Method, Nucl. Instr. and Meth. in Phys. Res. A428, 454 (1999).
730

[20] P. Staples, P.W. Lisowski, and N.W. Hill, Presented in APS/AAPT Conference, Washington,
April 18-21, 1995, Bull. Am. Phys. Soc. 40, 962 (1995), P. Staples, private communication
(1996) with an update of the data.
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Neutrons, ZETP (Sov. J. of Experimental and Theoretical Physics) 29, 778 (1955).
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Bismuth Nuclei, Sov. J. Nucl. Phys. 40, 552 (1984).
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Proceedings of the advisory group meeting organised by IAEA, Vienna, October 9-12, 1990,
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(2000), p. 43.
731

NUCLEON-INDUCED FISSION CROSS-SECTIONS CALCULATIONS
AND DEVELOPMENT OF TRANSMUTATION-ACTIVATION
DATA LIBRARY FOR TRANSITIVE ENERGY REGION 20-200 MEV
S. Yavshits, V. Ippolitov, S. Packhomov, G. Boykov
V.G. Khlopin Radium Institute
28, Second Murinski street, St. Petersburg, Zip: 194021, Russian Federation
O. Grudzevich
Institute of Nuclear Power Engineering
Studgorodok, 1, Obninsk, Kalaga reg., Zip: 2492, Russian Federation
Abstract
The results of a new approach for fission cross-sections at transitive energies are presented and it is
shown that the calculations describe experimental data well for both neutron and proton induced
fission. The development of the approach and corresponding code system for the calculations of
independent and cumulative yields of residual nuclei and fission fragments are discussed and
preliminary results are presented.
733

1. Introduction
Nuclear fission of heavy nuclei induced by nucleons at transitive energy region 20-200 MeV is
one of the main reaction channels (the fission cross-sections for actinide region can reach value 0.8-0.9
of the reaction cross-section). The accurate knowledge of fission probability for different fission
chances allows also defining yields of fission products (fission fragment yields and neutrons emitted
from fragment). These data added by particle and isotope yields from other reaction channels (direct
and pre-equilibrium emission of nucleons, evaporation of nucleons and light nuclei) give us the
possibility to develop the nuclear data necessary for the evaluation of activation of the Pb-Bi target of
the accelerator driven systems as well as the data on the fission of fuel and transmutation of actinide
radioactive waste.
The energy region 20-200 MeV is the transitive region from the well-investigated low energy
region to intermediate energy nucleon-induced reactions. It is well-known that for energies of
incoming particles up to 10-20 MeV the mechanism of reaction is defined by the competition between
fission and particle evaporation from compound nucleus and fission cross-sections and yields of
reaction products are well-reproduced by the statistical models. For higher energies the contributions
from the direct and pre-equilibrium reaction stages arise which is used to describe in the framework of
intranuclear cascade and exciton models, correspondingly.
In the given work, the new model approach and computer code is developed where the main
properties of nucleon-induced reactions on heavy nuclei at transitive energies are calculated in the
unified scheme on the base of reliable and detailed description of all main stages of the reaction.
2. The model approach
The scheme of new code is shown in the Figure 1. We use for entrance model simulation the
coupled channel method in Raynal’s version (code ECIS [1]). The deformation of target nucleus in the
ground state as well as the 3 lowest excited levels are taken into account in the calculations. The global
optical model potential for all nuclei from Pb up to Cf for transitive nucleon energies has been developed
by ours early [2]. For all these nuclei and beam energies the reaction cross-section data library has been
developed which is further used in the cross-section calculation of secondary reactions.
The intranuclear cascade model in Dubna version [3] has been included in the code for the
description of the direct stage of the reaction. The chains of primary and secondary cascades lead to
the emission of fast nucleons and the population of residual nuclei in the different nuclear states
characterised by mass and charge numbers, excitation energy and number of excitons (A,Z,E*,XpYh
distribution) which serve as the input data for the pre-equilibrium processes. The decay of excited
states on the pre-equilibrium reaction stage leads again to the particle emission and distribution of
excited nuclei (A,Z,E* distribution). The hybrid exciton model with Monte Carlo simulation [4] has
been used for the calculation of this reaction stage.
The last stage of statistical decay has been described in the framework of a detailed statistical
model based on the well-known code STAPRE [5] for each of the nuclei formed in the previous stages
of the reaction. The statistical part of reaction calculations contains a lot of parameters and special
efforts have been made in order to reduce the number of fitted parameters and to raise the predictive
ability of the code. The new data libraries for the level density parameters, fission barriers, nuclear
masses have been developed and included in the calculation scheme [2].
734

Figure 1. Scheme of the new code
Z A + p,n
ECIS-94
Reaction crosssection
library
for Z ≥ 82, E >
20 MeV
Global optical
model potential
Z A + p,n
Intranuclear cascade model
Distribution of (A,Z,E * ) and p-h states Z A, E * , XpYh
Hybrid exciton model
Distribution of (A,Z,E * )
Fission barriers library
Statistical fission/evaporation
model
Nuclear mass library
Cross-sections library
Level density parameters
library
735

3. Results and perspectives
On the base of the developed code system, the systematic calculations of proton and fission crosssections
for the nuclei from Pb to Pu have been carried out for transitive energies of incoming
particles. Some results of these calculations are presented in the Figures 2 and 3 in comparison with
the experimental data. It can be seen from the figures that our calculations describe the experimental
data rather well without any fitting of the model parameters. The preliminary results on the isotope
and neutron production are shown in the Figures 4 and 5. The calculations have been done within the
framework of the same approach. The agreement with the experimental data is quite satisfactory for
these data, too.
It is necessary for the fission product yields calculation to include in the code the model of fission
fragment production probability. Such a model should be used at each fission chance for each
fissioning nuclei formed at direct+pre-equilibrium stages of the reaction. At present we are working
out the statistical model of fission fragment yields in Fong’s approximation [6].
Figure 2. The proton-induced fission cross-section of 208 Pb
208 Pb+p
σ, mb
10 2 This work
10 1
Smirnov et al.
10 0
10 -1
100
E p
, MeV
736

Figure 3. Neutron-induced fission cross-sections of actinides
in comparison with experimental data
2.0
238 U(n,f)
σ f , b
1.5
2.5
235 U(n,f)
2.0
1.5
This work
Lisowski et al.
Scherbakov et al.
10 100
E n, MeV
2.5
σf, b
239 Pu(n,f)
2.0
1.5
Lisowski et al.
Scherbakov et al.
This work
2.5
237 Np(n,f)
2.0
1.5
Lisowski et al.
Scherbakov et al.
This work
10 100
E n
, MeV
737

Figure 4. Yields of U isotopes in the reaction 238 U (1GeV) + p
in comparison with the experimental data
σ, mb
10
238 U(1GeV) + p
U-isotopes
1
0.1
0.01
exp.(Schmidt et al.)
this work
0 2 4 6 8 10 12
Neutron loss
Figure 5. Neutron spectra from reaction Pb + p
100
208 Pb + p (120 MeV)
10
1
0.1
0 20 40 60 80 100 120 140 160
E, MeV
exp.
This work
Contribution of pre-equilibrium stage
Evaporation part
Contribution of cascade stage
738

REFERENCES
[1] J. Raynal, Proceedings of a Specialists Meeting, 13-15 November 1996, Bruyères-le-Chatel,
France, p. 159.
[2] S. Yavshits et al, Proceedings of IX Int. Conf. On Nuclear Reaction Mechanism, 5-9 June,
Varenna, Italy, p. 219.
[3] K.K. Gudima, S.G. Mashnik, V.D. Toneev, Nucl. Phys, 1983, Vol. A401, p. 329.
[4] M. Blann, Phys. Rev., 1996, Vol. C54(3), p. 1341.
[5] O.T. Grudzevich et al., Proc. of Int. Conf. on Nuclear Data for Science and Technology, 1988,
Mito, Japan, p. 1221.
[6] P. Fong, Phys. Rev., 1964, Vol. 135B, p. 1338.
739

NEUTRON RADIATIVE CAPTURE CROSS-SECTION OF 232 Th
IN THE ENERGY RANGE FROM 0.06 TO 2 MeV
D. Karamanis 1 , M. Petit 1 , S. Andriamonje 1 , G. Barreau 1 , M. Bercion 1 ,
A. Billebaud 2 ,B. Blank 1 , S. Czajkowski 1 , R. Del Moral 1 , J. Giovinazzo 1 ,
V. Lacoste 3 , C. Marchand 1 ,L. Perrot 2 , M. Pravikoff 1 , J.C. Thomas 1
1 CEN/Bordeaux-Gradignan, France
2 ISN, Grenoble, France
3 CERN, Geneva, Switzerland
Abstract
Neutron capture cross-section of 232 Th have been measured relative to σ(n,γ) for 197 Au and σ(n,f) for
235 U in the energy range from 60 keV to 2 MeV. Neutrons were produced by the 7 Li(p,n) and T(p,n)
reactions at the 4 MV Van de Graaff Accelerator of CEN/Bordeaux. The activation technique was
used and the cross-section was measured relative to the 197 Au(n,γ) standard cross-section up to 1 MeV.
Above this energy, the reaction 235 U(n,f) was also used as a second standard and the fission fragments
were detected with a photovoltaic cell. The results after applying the appropriate corrections indicate
that the cross-sections are close to the JENDL-3 database values up to 800 keV and over 1.4 MeV. For
energies in the intermediate range, values are slightly lower to the ones from all the libraries.
741

1. Introduction
During the past few decades a growing concern for the continuous accumulation of large amounts
of nuclear wastes and for the future of energy production systems has emerged. The green house
effect, the foreseeable limits in fossil fuel resources and the pollution of the environment with
combustion by-products, point to the need towards alternative, innovative and safer strategies. New
challenges for the different fuel cycle options and nuclear waste management, have produced an
impetus in the research for extension of the life span of presently operating reactors, the increase of the
fuel burn-up and plutonium recycling (in particular the incineration of actinides and long-lived fission
products). Furthermore, the 232 Th- 233 U fuel cycle is studied extensively for energy production and as a
waste management option in the next generation of systems like the Energy Amplifier and ATW [1],
thermal or fast reactors and the accelerator driven systems (ADS) [2]. Unfortunately, uncertainty in the
parameters of systems employing the Th-U cycle, caused by discrepancies in the nuclear data available
at present, appears to be higher than the uncertainty caused by different calculational schemes [3].
Hence, the need arises for bringing the quality of these nuclear data to the same level of accuracy as
that of the U-Pu cycle [3, 4].
The most crucial reaction channel in the Th-U fuel cycle is the neutron capture on 232 Th, which
leads to 233 U after two successive β - decays. Moreover, the above reaction cross-section is currently
required with an accuracy of 1-2% in order to be used safely in simulated techniques for predicting the
dynamical behaviour of complex arrangements in fast reactors or ADS [3]. As an example of its
importance in ADS, a 10% change in the 232 Th capture cross-section gives rise to a 30% change in the
needed proton current of the accelerator if the system has to be operated at a sub-critical level of
K eff ≈ 0.97 [5].
Since 1946, there have been a number of relative measurements of the 232 Th(n,γ) reaction
cross-section by employing prompt γ-ray detection or activation techniques. However, these are almost
the half when compared to the measurements of 238 U(n,γ) reaction cross-section that presents a number
of common features. Furthermore, experiments for the exclusive measurement of the above reaction
cross-section are even less. Since the measured values differ substantially in the energy range
0.05-1.5 MeV and due to the difficulty of re-normalisation, the current evaluations in the above energy
range present discrepancies of the order of 10-30% [4]. Therefore, additional measurements are
needed in order to satisfy the required accuracy.
In the present work, the neutron capture cross-section of 232 Th was measured in the neutron
energy range from 60 keV to 2.0 MeV. The activation technique was used and the cross-section was
measured relative to the 197 Au(n,γ) standard cross-section of ENDF/B-VI up to 1 MeV. The
characteristic γ lines of the product nuclei 233 Pa and 198 Au were measured with a 40% HPGe detector.
Above this energy, the reaction 235 U(n,f) with values from ENDF/B-VI, was used as a second standard
and the fission fragments were detected with a photovoltaic cell. Several experiments and simulations
were also performed in order to check all the factors influencing the cross-section values, with special
emphasis, to minimise the associated uncertainties and errors.
2. Experimental procedure
Neutrons were produced by either the Li(p,n) or the T(p,n) reaction on the 4 MV Van de Graaff
Accelerator of the Centre d’Études Nucléaires de Bordeaux-Gradignan (CENBG). A deep (1.7 m
depth) “neutron hole” was installed in the neutron beam line of the accelerator in order to eliminate
scattered neutrons from the environment (floors, walls and ceiling). The beam was focused and
collimated to a spot roughly 0.5 × 0.3 cm 2 at the target while beam currents were of the order of
742

15 µA. At the end of the proton beam line, the target holder was cooled with a continuous flow of a
very thin film of water. This was used instead of compressed air cooling since it was experimentally
checked that the mono-energetic neutron beam was better stabilised when the water flow was applied.
The threshold of the 7 Li(p,n) 7 Be reaction was used for the energy calibration of the VdG
accelerator. The selected proton energies with the corresponding neutron energies are presented in
Table 1. In this table, the minimum (maximum) neutron energy corresponds to maximum (minimum)
proton energy degradation in the LiF target and maximum (minimum) angular spread in the neutron
beam at the position of the Th target. With Li as a neutron source and Ep ≥ 2.4 MeV, a second group
of mono-energetic neutrons was produced due to the population of the first excited state of 7 Be. The
intensity of the second group was calculated with a Monte Carlo simulation.
Table 1. Selected proton energies and corresponding reactions in the activation measurements
1
Neutron source
(mg cm -2 )
Irradiation
Ep (Min)
(keV)
Ep (Max)
(keV)
En (Min)
(keV)
En (Max)
(keV)
LiF:0.260 Au-Th-Au 1 1 881 1 907 0.875 100
0.515 – 1 928 1 991 130 220
0.540 – 1 928 1 995 135 225
0.515 – 2 008 2 070 236 320
0.500 – 2 017 2 077 247 324
– – 2 108 2 166 354 425
0.540 – 2 195 2 256 452 525
– – 2 253 2 313 516 588
0.515 Au-Th-U 2 2268 2 325 522 600
0.500 Au-Th-Au 1 2 270 2 325 534 600
– – 2 294 2 349 560 626
– – 2 346 2 340 616 680
– – 2 422 2 475 697 760
– – 2 499 2 551 778 840
– – 2 577 2 627 860 920
– – 2 654 2 704 940 1 000
TiT: 1.0 Au-Th-U 2 1 700 1 803 870 1 000
– Au-Th-Au 1 1 710 1 803 880 1 000
– Au-Th-U 2 1 953 2 047 1 126 1 250
– – 2 205 2 292 1 379 1 500
– – 2 458 2 539 1 632 1 750
– – 2 458 2 539 1 632 1 750
– – 2 710 2 786 1 882 2 000
Al box.
Fission chamber.
2
Thorium metal targets (Goodfellow SARL) were of high purity (99.5%), 1 mm thick and a
surface of 1 × 1 cm 2 . Two experimental devices were used for samples irradiation, a Cd box and a
fission chamber. In the first one, the Th foils were packed together with two Au foils on each side
743

(same surface and 0.5 mm thickness) and were placed in a Cd box (5 × 5 cm 2 and 1 mm of Cd
thickness) in order to eliminate any contribution from the thermal neutron background. The box was
tied with tiny nylon wires at the centre of a thin Fe ring and aligned with the proton beam; the
assembly was placed at 5 cm from the Cu backing of the neutron source.
Figure 1. Schematic representation of the neutron fission chamber
In the fission chamber, a 235 U thin foil (0.04 mg cm -2 ) on an Al backing (0.047 cm) was used as a
second reference foil. The configuration used with the fission chamber is schematically depicted in
Figure 1. Fission fragments were detected online with a photovoltaic (solar) cell. This device was used
because of its minimal mass inside the neutron fission chamber. The cell was a monocrystalline n + p
junction with 360.36 mm 2 surface. The dead zone of the photovoltaic cell was determined by direct
micro-measurement of the thickness of the Al grids. The detection efficiency for fission fragments
from 235 U was considered twice that for alpha particles that were observed with a Si surface barrier.
The later was collimated in order to suspend exactly the same solid angle as the cell during the
irradiations.
The Cd box or the fission chamber were irradiated around 20 hours at 0° with respect to the
proton beam. During the irradiation, the flux was monitored with a He 3 detector that was placed
2 meters from the neutron source. After each irradiation, the intensity of the γ ray lines emitted by the
de-excitation of the produced nuclei 233 Pa (Eγ = 312 keV (38.6%)) and 198 Au (Eγ = 412 keV (96%)),
was measured with γ spectroscopy with a 40% HPGe and for 5 cm source to detector distance. The
acquisition time varied between 1-2 hours for Au and 1-2 days for Th. The pulser method was used for
dead time correction. The photopeak areas were determined with the use of PAW program [6]. The
efficiency of the Ge detector was determined with a 152 Eu punctual source and the use of GEANT3.21
[7] and MCNP4B [8] simulation codes.
Two more experiments were performed for the estimation of the thermal and epithermal neutron
background in the experimental hall. Prior to irradiations, the coincidence technique of the alpha and
triton produced in the 6 Li(n,α)T reaction and detected with two Si junctions (J1-J2) in a “sandwich”
arrangement, was used. Several combinations with different target support materials, with and without
cadmium shielding and distances from the neutron production target were investigated. The optimal
744

conditions with the minimum thermal background were produced by placing the cadmium-shielded
device in around 5 cm from the neutron production LiF target that was supported on copper. At the
end of irradiations, the background was measured by irradiating simultaneously three Cd boxes with
Au foils inside and at different distances from the neutron source.
3. Analysis and results
The neutron capture cross-section of 232 Th relative to the cross-section σ(n,γ) for 197 Au and σ(n,f)
of 235 U is given by the relation:
Th Rf
Th
σ Th(
< En > ) ε(
Rf) I N f(T ,t)
(1)
= ⋅ ⋅ ⋅
Rf Th
Rf
σ ( < En > ) ε (312 ) I N f(T ,t)
Rf
where is the mean neutron energy in the sample, I the photopeak or fission fragments area, ε the
efficiency for the 233 Pa or 198 Au γ lines or the fission fragments, N the number of atoms in the samples
and f a time factor relating the measured peak intensities to the end of irradiation. In the case that the
235 U foil is used as a second reference, the time factor is simply the time of irradiation. In the present
experiment, the neutron flux in the Th target was assumed to be the same as the mean value of the flux
calculated from the two foils (Au-Au or Au-U) in each side of the Th target. Therefore, the above
formula was modified to account for the time difference in measuring the two Au reference targets or
the different cross-sections in the case of Au-U foils.
The efficiency of the HPGe detector for the activity measurements of thorium or gold samples
was determined with the following way. The dependence of recording decay events by the HPGe as a
function of photon energy was determined with a punctual 152 Eu source that was placed at 5 cm or
10 cm. The detector’s geometry was then simulated with MCNP4B and GEANT3.21 in order to
reproduce the observed experimental values. Finally, the detector's efficiency was determined by
replacing the punctual source in the simulation with the extended gold or thorium box. In this way, the
values of (1.19 ± 0.03)% for the recorded 412 keV line of 198 Au and (1.27 ± 0.03)% for the 312 keV
line of 233 Pa, were obtained. The above values were also confirmed experimentally.
Since the photovoltaic cell was calibrated by replacing it with a collimated and suspending the
same solid angle Si surface barrier, the combined product ε U • N U was measured instead of each
separate term in Equation 1 and a value of 2 × (3.27 ± 0.10) × 10 16 was determined. Furthermore, the
dead zone of the photovoltaic cell was determined by direct measurement of the grids thickness and
was found (4.45 ± 0.02)%.
3.1 Corrections
One of the factors with primary importance in fast capture measurements with the activation
method is the thermal-epithermal background present in the experimental hall. Its contribution and
spoiling of experimental results can be non-negligible and have to be corrected. For this reason, two
experiments were performed. In the first, the “room background” was evaluated with the coincidence
technique of the reaction 6 Li(n,α)T. Since the alpha and triton particles from thermal or fast neutrons
are emitted in opposite directions with different production rates (300:1) and with different energies,
they can be separated. In this way the contribution of the “room background” was estimated to be
insignificant. However and due to the increased apparatus mass seen by the neutron flux, a
quantitative analysis of the experiment could be of less value.
745

Figure 2. Dependence of the total activation on the target–sample distance. Solid curve
represents linear fit of Equation 2 to the data after corrected the 2nd and 3rd Au foils
for flux attenuation (A = 7 205, B = 0.7). Dashed curve represents linear fit to
the data after corrected the foils with an MCNP simulation (A = 7 185, B = 2.3)
Therefore, a simple irradiation experiment was performed with an assembly of three Au foils at
different distances from the neutron source. Since the room-scattered neutrons are uniformly
distributed in the vicinity of the target, the produced 198 Au activity should be constant with targetsample
distance. The activity due to the primary neutron source should vary as 1/r 2 whereas the inscattering
contribution from the structural materials near the target or the sample should also exhibit a
distance dependence close to 1/r 2 . Hence, the total activation of any foil will be given by:
I(r) = A / r 2 + B (2)
According to the results obtained and shown in Figure 2, the “room” neutron background
contribution can be considered negligible.
However, the effect of other factors is not negligible and the observed activation of any foil had
to be corrected due to:
• Neutron energy spread because of proton energy degradation in the source.
• Effects of finite dimensions of neutron source and targets on neutron energy spread.
• Second neutron energy group from the 7 Li(p,n) 7 Be* reaction for energies higher than
600 keV.
• Inelastic and elastic neutron scattering within the intermediate experimental environment
• Multiple elastic and inelastic scattering in the target foils
Therefore, a Monte Carlo code with all the above effects was developed for an IBM AIX 4.3.2
operating system. The code includes source angular and energy distribution, energy-dependent
cross-sections and considers multiple scattering. All the cross-sections are entered as tables with linear
interpolation. Neutrons are randomly produced inside the target and their angular distribution and
different emission probability in the energy range of the proton degradation in the target is chosen
according to evaluated cross-sections [9,10]. Their path is followed and their energy is monitored. The
746

only assumption of the code is that the angular distribution of the produced neutrons is constant in
their energy spread interval.
Diagram 1. Flow diagram of the analog Monte Carlo code for the determination of the correction
factors in the neutron capture cross-section of 232 Th
Load materials
Load cross-sections
Load geometry
Load random generator
Calculate proton energy degradation
Neutron production according
to the total cross-section
Neutron emission according to
the differential cross-section
Find neutron location
Find interaction length (IL)
Find boundary length (BL)
If capture Au or Th or fission U
ÅSave and kill
otherwise kill
No
Perform reaction and
change energy
Yes
Elastic
or inelastic ?
Yes
BL

Figure 3. Neutron radiative capture cross-section of 232 Th of the present work
in comparison with a) the existing evaluated data from the four major neutron
data reference libraries and b) with experimental and normalised data
748

4. Conclusions
The values of the neutron capture cross-section of 232 Th that were measured in the present work
are in agreement with the JENDL database up to 800 keV and over 1.4 MeV. For energies in the
intermediate range, values are slightly lower to all the databases but in agreement with the most recent
values of Davletshin et al., [11]. Moreover, a re-normalisation of 232 Th and 238 U neutron capture data
of Lindner et al., [12] (the most reliable considered data and the basis of many evaluations) with 238 U
from the current ENDF and JENDL databases was undertaken (Figure 3b). The later reveals that a
very good agreement is reached with the data of the present work, data from a time of flight
experiment that were the basis for the JENDL evaluation [13] and data that could be the basis of a new
BROND or ENDF evaluation [11,12].
Acknowledgements
This work was partially supported by GEDEON-PACE Programme, Région Aquitaine and
European Commission Nuclear Fission Safety Programme. The support in codes and references from
the Nuclear Data Centres of NEA and IAEA, is kindly acknowledged. One of the authors (D.K.)
would like to thank the European Commission Nuclear Fission Safety Programme for providing the
Marie Curie Postdoctoral Research Fellowship (Contract No. FI4W-CT98-5004).
REFERENCES
[1] C.D. Bowman et al., Nuclear Energy Generation and Waste Transmutation Using an
Accelerator-driven Intense Thermal Neutron Source, NIMA 320 (1992) 336-367; C. Rubbia
et al., Conceptual Design of a Fast Neutron Operated High Power Energy Amplifier,
CERN/AT/95-44(ET) (1995).
[2] AIP Conference Proceedings No. 346 of the Int. Conf. on Accelerator-driven Transmutation
Technologies and Applications, Las Vegas, NV, USA, July 1994.
[3] V.G. Pronyaev, Summary Report of the Consultants’ Meeting on Assessment of Nuclear Data
Needs for Thorium and other Advanced Cycles, INDC(NDS)-408, IAEA, August 1999.
[4] B.D. Kuzminov and V.N. Manokhin, Status of Nuclear Data for the Thorium Fuel Cycle,
Nuclear Constants, No. 3-4 (1997), p. 41.
[5] M. Salvatores, Nuclear Data for Science and Technology, Conf. Proceedings Vol. 59, G. Reffo,
A. Ventura and C. Grandi (Eds.), Bologna 1997.
[6] PAW, Physics Analysis Workstation, CERN Program Library Long Write-up Q121, CERN
Geneva (1995).
[7] GEANT, Detector Description and Simulation Tool, CERN Program Library Long Write-up
W5013, CERN, Geneva (1993).
749

[8] MCNP, A General Monte Carlo Code for Neutron and Photon Transport, J.F. Briesmester (Ed),
(LA-12625-M, 1993).
[9] H. Liskien and A. Paulsen, Neutron Production Cross-sections and Energies for the Reactions
7
Li(p,n) 7 Be and 7 Li(p,n) 7 Be*, Atomic Data and Nuclear Data Tables 15 (1975) 57-84.
[10] H. Liskien and A. Paulsen, Neutron Production Cross-sections and Energies for the Reactions
T(p,n) 3 He, D(d,n) 3 He and T(d,n) 4 He, Atomic Data and Nuclear Data Tables 11 (1973) 569-615.
[11] A.N. Davletshin et al., Neutron Radiative Capture Cross-sections for the 232 Th and 197 Au Nuclei
in the 0.37-2.5 MeV Region, INDC(CCP)-375 and INDC(CCP)-389 (1994).
[12] M. Lindner, R.J. Nagle and J.H. Landrum, Neutron Capture Cross-sections from 0.1 to 3 MeV
by Activation Measurements, Nucl. Sci. & Engin. 59 (1976) 381-394.
[13] K. Kobayashi, Y. Fujita and N. Yamamuro, Measurement of Neutron Capture Cross-section of
Thorium-232 from 1 keV to 408 keV, J. Nucl. Sci. Technol. 18 (1981) 823-834.
750

DETERMINATION OF THE NEUTRON FISSION CROSS-SECTION
FOR 233 Pa FROM 0.5 TO 10 MeV USING THE TRANSFER REACTION 232 Th( 3 He,pf) 234 Pa
M. Petit 1) , M. Aiche 1) , S. Andriamonje 1) , G. Barreau 1) , S. Boyer 1) , O. Busch 1) ,S. Czajkowski 1) ,
D. Karamanis 1) , F. Saintamon 1) , E. Bouchez 2) , F. Becker 2) , F. Gunsing 2) , A. Hurstel 2) ,
Y. Le Coz 2) , R. Lucas 2) , M. Rejmund 2) , C. Theisen 2) , A. Billebaud 3) , L. Perrot 3)
1 CEN Bordeaux, BP 120, 33175 France
2 CEN Saclay, DAPNIA, 91191 Gif-sur-Yvette, France
3 ISN Grenoble, 53, Avenue des Martyrs, 38026 Grenoble, France
Abstract
Neutron induced fission cross-section of 233 Pa in the fast neutron energy range from 0.5 to 10 MeV
was determined for the first time as a two term product of the fission probability of 234 Pa nucleus and
the same compound nucleus formation cross-section. The first term was measured with the transfer
reaction 232 Th( 3 He,p) 234 Pa while the second one was calculated. The tendency of the resulting data to
agree with the existing evaluated one, is a proof for the validity of the utilised method.
751

1. Introduction
New reactors using the uranium-thorium fuel cycle are under studies in order to provide safer and
cleaner nuclear energy as highly radiotoxic actinide waste (Pu, Am and Cm isotopes) will be produced
in lower quantities than the currently used uranium fuelled reactors. Moreover, further developments
of this thorium based cycle rely on nuclear data libraries of the quality achieved for the uranium ones.
The primary reaction of importance using the thorium cycle is the one producing 233 U from
neutron capture on 232 Th, the net production of 233 U is controlled by the 27 days half-life of 233 Pa: a
fertile nucleus ( 232 Th) is transformed into a fissile nucleus ( 233 U) after neutron capture and 2 successive
β decays.
+n β β
232 Th
233 Th
233 Pa
(22 min) (27 days)
233 Pa as a precursor of 233 U may capture neutrons and could lead to a reactivity decrease of the
reactor, conversely and after a shut down, the build up of 233 U increases this reactivity; this so called
Protactinium effect should add severe requests on the 233 Pa and 233 U inventories for reactivity control.
There is no equivalent effect in the well studied 238 U- 239 Pu fuel cycle as the equivalent intermediate
isotope 239 Np has a relatively shorter half-life (2.35 days).
Furthermore, neither the neutron capture cross-section nor the neutron induced fission
cross-section have been measured for 233 Pa up to now. The reason of this absence is the short decay
half life of the 233 Pa nucleus (27 days) that leads to an extreme activity of 7.10 8 bq µg -1 s-1. Due to this
high radioactivity, there is no technique presently available to measure directly the 233 Pa(n,f) reaction
cross-section.
The particular aim of this work is to provide data for the neutron induced fission of 233 Pa in the
fast neutron energy range from 0.5 to 10 MeV. To overcome the problem of the induced radioactive
233 Pa, we have used the transfer reaction 232 Th( 3 He,p) 234 Pa that leads to the desired 234 Pa nucleus as it
should be observed in the 233 Pa(n,f) reaction. Several years ago, this method has been used
successfully to estimate the neutron-induced fission of short-lived targets like 231 Th (25.6 h), 233 Th
(22.1 min) etc.
233 U
2. Theory
Transfer reaction measurements give access to the fission probability Pf(E * ) as a function of
excitation energy. The equation relating the two quantities can be written as:
Pf(E
exc
where α(E exc
,J,π) is the relative population of spin states (J, π).
⎛
⎞
⎜
Γ (E
* π ⎟
⎜
⎟
∑ α π F
,J, )
)= ( E
⎜ exc
, J, )
⎟
(1)
Jπ
⎜
∑Γ
i
(E
*
,J, π)
⎟
⎝
i
⎠
Neutron induced fission measurements give access to the fission cross-section σ F (En) as a
function of the incident neutron energy that is relating to the excitation energy as:
E exc
≈ Sn + En (2)
752

Therefore, the neutron induced fission cross-section can be determined from the following
equation:
π
T
J , π N(
J , En)
σ F ( En ) = σ NC ( En)
∑ pn,
f
π
J , π ∑ N(
J , En)
(3)
where N(J π ,En) is the relative final spin states (J π ) population and Pn,f is the fission probability for the
spin states (J π ) given from the relation:
Γ f ( En,
J,
π )
pn,
f =
∑ Γ ( En,
J,
π )
(4)
i
i
Concluding the neutron induced fission cross-section can be found as a two term product or:
J
σ f
(E) ≈ σ NC
* P f
(E exc
≈ Sn + En) (5)
The last relation is well verified if we assume that we have enough fission channels so that the
total spin population differences between the two processes do not affect the output channel. This
assumption should hold for the odd-odd fissionning system 234 Pa.
3. Experimental procedure and data reduction
The 3 He beam was provided by the IPN Orsay Tandem facility at three different energies (24, 27
and 30 MeV). The 232 Th targets (100 µg/cm 2 ) were deposited onto 50 µg/cm 2 carbon backings. The
( 3 He,p) channel has been discriminated among the other competing channels (d, t and alpha outgoing
light charged particles) with a ∆E-E counter telescope placed at 5 cm from the target and at 90°
relative to the 3 He beam axis. The ∆E detector was 300-µm thick fully depleted Si detector. The
E counter was 5-mm thick Si detector.
The telescope energy calibration has been obtained with the reaction 208 Pb( 3 He,d) 209 Bi using
ground state Q value and excited states in 209 Bi, as it is indicated in Figure 1.
753

Figure 1. Excited states of 209 Bi that were used for the telescope energy calibration
Fission fragments were detected with two multi-solar cell arrangements placed at 5 cm from the
target and at 0° and 90° relative to the recoil direction of the 234 Pa nucleus (∼27°). The Saclay VXI
acquisition system was used in this experiment, as it was designed and used for the Euroball and
Saphir multi-detector arrays [1]. The master trigger was generated either by a triple coincidence of
signals from the ∆E, E and fission counters (coinc.) or by the double coincidence from the ∆E and E
detectors (singles). The singles spectra have been corrected for the contribution from the carbon
backing by subtracting the spectrum from a separate carbon irradiation run. The two spectrums are
shown in the Figure 2. The coinc. spectra were corrected for random coincidence events by using the
appropriated scaled singles spectra.
The fission probability of 234 Pa was calculated from the number of fission events detected in
coincidence with the outgoing light charged particles according to the relation:
P f
2 Ncoinc
= π (6)
Ω Nsingles
f
754

Figure 2. Singles protons and deuterons spectra from the transfer reactions
a) 232 Th( 3 He,p) 234 Pa (solid line) and b) 12 C( 3 He,p) 14 N (dashed line)
In the last relation Ω f is the solid angle efficiency for fission fragment detection and it was
determined from a calibrated 252 Cf source, placed in the same geometry.
The consistency of the method and the reliability of our measurements were checked with the
reaction 232 Th( 3 He,df) 233 Pa. Since this reaction had been studied in the past by Back et al. [2], their
fission probability of 233 Pa was compared with the one obtained in the present work and an excellent
agreement was found (Figure 3).
755

Figure 3. The fission probability of 233 Pa as obtained in the present work (open cycles)
and in the Los Alamos experiment by Back et al [2] (full cycles).
Figure 4. Neutron induced fission cross-section of 233 Pa and the existing evaluated data
756

Following the procedure proposed by J.D. Cramer and H.C. Britt [3], the neutron induced fission
cross-section 233 Pa(n,f) was determined as the product of the 234 Pa measured fission probability and the
computed compound nucleus formation cross-section of 234 Pa. In the later calculation, the transmission
coefficients T(l,s = ±1/2) of Perey and Buck [4] were used. The results obtained are shown in Figure 4
in comparison with the existing evaluated data from the ENDF/B-VI and JENDL-3 reference libraries.
4. Conclusion
Although the results should be viewed as preliminary, a tendency for following the JENDL
evaluation can be observed, at least for neutron energies greater than 4 MeV. The high errors for lower
neutron energies, do not permit a safe conclusion. Due to this, a new experiment is planned in the very
near future which combined with updated transmission coefficients will provide a clear determination
of the neutron induced fission cross-section of 233 Pa. It is also planned that the present work will
extend to the measurement of the 233 Pa(n,γ) reaction in order to complete our knowledge on the 233 Pa
effect.
REFERENCES
[1] C. Thiesen, C. Gautherin, M. Houry, W. Korten, Y. Lecoz, R. Lucas, G. Barreau, C. Badimon,
T.P. Doan, The SAPHIR Detector, in Ancillary Detectors and Devices for Euroball, H. Grawe
(Ed.) GSI 1997 p.47.
[2] B.B. Back, H.C. Britt, O. Hansen, B. Leroux, J.D. Garrett, Fission of Odd-A and Doubly Odd
Actinide Nuclei Induced by Direct Reactions, Phys. Rev. C10 (1974) 1948-1965.
[3] J.D. Cramer and H.C. Britt, Neutron Fission Cross-sections for 231 Th, 232 Th, 235 U, 237 U, 239 U, 241 Pu,
and 243 Pu from 0.5 to 2.25 MeV Using (t,pf) Reactions, Nuclear Science and Engineering 41
(1970) 177-187.
[4] E.H. Auerbach and F.G.J. Perey, Optical Model Neutron Transmission Coefficients, 0.1 to
5.0 MeV, Brookhaven National Laboratory, BNL 765 (T-286) (1962).
757

MEASUREMENT OF DOUBLE DIFFERENTIAL CROSS-SECTIONS
FOR LIGHT CHARGED PARTICLES PRODUCTION IN NEUTRON
INDUCED REACTIONS AT 62.7 MeV ON LEAD TARGET
M. Kerveno, F. Haddad, P. Eudes, T. Kirchner, C. Lebrun
SUBATECH, Nantes, France
I. Slypen, J.P. Meulders
Institut de Physique Nucléaire, Louvain-la-Neuve, Belgique
V. Corcalciuc
Institute of Atomic Physics, Bucharest, Roumania
C. Le Brun, F.R. Lecolley, J.F. Lecolley, M. Louvel, F. Lefèbvres
Laboratoire de Physique Corpusculaire de Caen, Caen, France
Abstract
In the framework of nuclear waste transmutation, we have measured d 2 σ / dΩdE
for protons,
deuterons, tritons and alpha production in neutron induced reactions on a lead target. Due to the
structure of the neutron beam, incident neutron energies between 30 and 62.7 MeV have been
obtained at once. The analysis of 62.7 MeV neutron is now complete for hydrogen isotopes and a first
set of comparisons has been done with calculations. On one hand, it is found that the GNASH-ICRU
data do not give the correct cross-sections (neither absolute value nor shape). In the other hand, a
comparison for protons using FLUKA is working reasonably well except an underestimation of the
pre-equilibrium emission around 30 MeV at forward angles and an overestimation of thermal
emission at backward angles. Further data on protons induced reactions at the same energy, obtained
within an European concerted action, will be available soon allowing a stronger constraint on
theoretical calculations.
759

1. Introduction
The renewal interests on intense neutron source have put forward the necessity of new sets of
nuclear data. This is particularly true, in the intermediate energy range between 20 and 200 MeV, for
the development of new options for nuclear waste management based on the concept of hybrid system
which combines an intense high energy proton beam with a sub-critical fission reactor. One important
point of these studies is to know precisely the characteristics of the nuclear reactions taking place in
the spallation target that is intended to be in Pb-Bi or Hg. In particular, it’s necessary to estimate, in
reactions induced by neutrons, the production of light charged particles (lcp) which may have critical
effects on materials.
At present, code calculations are used to simulate these phenomena. Below 20 MeV, the upper
limit of the databases, codes provide results with a good level of confidence. Above 150 MeV, Intra
Nuclear Cascade calculations provide also good results. On the contrary, in the intermediate energy
region where the pre-equilibrium emission is important, new theoretical approaches seem to be
necessary to ensure a good link between low and high-energy processes.
These new approaches based on pre-equilibrium models will allow increasing the upper limit
energy value (from 20 to 150 MeV) of data bases providing that theoretical codes could have
sufficient predictive power in this energy range. Thus it’s necessary to measure new cross-sections to
constrain these codes in order to improve their predictive power and to evaluate the quantity of
hydrogen and helium isotopes that will be emitted from the lead target and eventually estimate their
interactions with structure materials. A large concerted program of nuclear data measurements is now
carrying out by several French and European laboratories to measure double differential crosssections
production for light charged particles in neutron induced reactions on different targets.
We report hereby double differential cross-sections for protons, deuterons and tritons production
from a lead target at 62.7 MeV incident neutron energy.
2. Experimental set-up
The experiment has been done at the fast neutron facility existing at the cyclotron CYCLONE at
Louvain-la-Neuve [1]. The neutron beam is obtained using the 7 Li(p,n) 7 Be gs
(Q = -1.644 MeV) and
7
Li(p,n) 7 Be* (Q = 0.431MeV) reactions. The neutron facility is presented in Figure 1. The important
features of this line are the presence of a beam peak off, BPO, upstream the lithium target to get the
time at which the neutrons are created and a faraday cup which collect the non-interacting deflected
protons. The scattering chamber is located 3.28 m after the neutron production point and is followed
by a second chamber which contains a second beam monitor system.
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Figure 1. Global view of the experimental set-up
About 10 6 n/s are available in the reaction chamber when a 10µ A proton beam interacts on a
3 mm thick natural lithium target [2]. The neutron spectrum is presented in Figure 2. It consists of a
well-defined peak located at 62.7 MeV containing about 50% of the neutrons and a flat continuum at
low neutron energy, which is 8 times lower than the peak maximum. The full width at half maximum
of the peak is 4 MeV. The neutron beam spot is quite large at the reaction point as shown in the inset
of Figure 2 that presents the radial neutron distribution normalised to the intensity on the centre.
Figure 2. Neutron spectrum produced by a 65 MeV protons beam on a lithium target. The inset
shows the beam profile, normalised to the centre intensity, 3 m downstream the lithium target.
The experimental set-up is based on the one used by the group of J.P. Meulders [3,4]. The
reaction chamber allows to use simultaneously six telescopes. Each telescope is composed of a ∆E
detector (100 µm thick and 4 cm in diameter NE102 plastic scintillator) and an E detector (22 mm
thick and 38.1 mm in diameter CsI crystal). A set of two collimators is inserted in the telescope as
shown in Figure 3 to precisely define the detection solid angle. The ∆E detector gives a fast time
signal, which allows time of flight measurement and ensures a good reconstruction of the incident
neutron energy. The CsI thickness has been optimised to stop the light charged particles produced in
our experiment and a pulse shape analysis of the signal is performed.
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Figure 3. Schematic view of a telescope
During the experiment, a quite complete angular distribution has been obtained from 20 o to 70 o
by step of 10 o in the forward hemisphere and at 110 o and 160 o in the backward hemisphere. Two
different configurations have been used to allow two times longer recording data time at backward
angles.
3. Data analysis
The particle identification is obtained by performing a pulse shape discrimination of the CsI
detector signal. Plotting the slow (CsI_s) versus the fast (CsI_f) component of the CsI light output
allows to separate the different hydrogen isotopes as well as the helium. As shown on Figure 4, the
good quality of the discrimination added to the possibility to suppress most of the background
(neutron and γ) using the ∆E-E correlation facilitates the particle identification.
Figure 4. CsI slow versus fast component of the light output
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To get the proton (deuteron) energy calibration of our detectors we have used recoil protons
(respectively deuterons) from elastic neutron scattering on CH 2
(DH 2
) target at 6 different angles from
20 o to 70 o . The time calibration is also extracted from these data.
Figure 5. Total time of flight as a function of
the measured energy for alpha particles in a lead run
The triton and the alpha calibration have been performed using the time of flight information
available for these particles, following the relation T cp
=T tot
- T n
where T tot
, corresponds to the
measured time between the BPO and the ∆E signals, T n
to the time made by the neutron to go from
the BPO to the target and T cp
to the time of flight of the particle from the target to the ∆E detector.
The special neutron beam energy distribution (see Figure 2) allows selecting only the neutrons from
the peak. The bi-parametric plot shown on Figure 5 presents T tot
versus the measured energy in
channel for alpha particles in a lead run. A clear band appears corresponding to the 62.7 MeV
incident neutrons for which T n
is known. For several points on this band, it is then possible to extract
the alpha particle energy from T cp
and to determine the energy calibration curve. The same method
applies also for the other kind of particles. In particular, the calibration obtained by this method for
protons and deuterons gives similar results as the first method based on elastic scattering. In Figure 6,
calibration curves used for telescope 1 are summarised for isotopes under study.
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Figure 6. Calibration curves used for telescope 1. Proton corresponds to the full black line,
deuteron to the dashed line, triton to the dotted line and alpha to the dashed-dotted line.
Using the calibration curves, it is possible, event by event, to determine the lcp energy and then
T cp
to deduce T n
and the neutron energy. As an illustration, in Figure 7, the total deuteron spectrum is
plotted as a function of energy. The inset of Figure 7 shows the reconstituted neutron incident energy
distribution. By selecting a slice in the neutron spectrum, the deuteron spectrum can be obtained for
the corresponding neutron incident energies. As examples, deuterons created by neutrons of
62.7 MeV (respectively 43 MeV) are represented as hashed histogram (squared histogram). It is then
possible in one experiment using 65 MeV protons to measure cross-sections at neutron incident
energy ranging from 30 MeV to 62.7 MeV.
Figure 7. Deuteron spectrum obtained by selecting 62.7 MeV (respectively 43 MeV)
incident neutron energy is presented as hashed (squared) histogram
The absolute normalisation of the lead double differential cross-section is obtained by using the
n-p scattering cross-section extracted from the CH 2
calibration runs [3].
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3.1 Corrections
Several corrections have to be made on our data. One concerns the particle scattering on the
telescope collimators, the others are coming from the target thickness (0.3 mm). In order to quantify
these corrections, we used the GEANT code [5] to simulate as closely as possible the experimental
setup and the beam structure.
3.1.1 Diffusion on the collimator set
In Figure 8, the effect of the collimators on a well define energy beam, as shown on the inset of
Figure 8, have been plotted. The diffusion leads to a long tail at low energy. The broadening of the
peak is due to the energy losses in the target. Using the simulation, it is possible to estimate the
pollution of the tail, normalised to the peak population, in each energy bin. Doing these calculations
from 5 to 70 MeV allows us to have an estimation of the full diffusion contribution. The iterative
correction procedure consists on removing the tail contribution from the spectrum starting from the
highest bin: the population of the highest energy bin does not contain any pollution and the
corresponding tail contribution can be estimated from the simulation and discarded for each bin of the
spectrum.
Figure 8. Simulation showing the effect of the proton scattering on the collimator set.
The inset shows the particle energy before entering our telescope.
The result of such a procedure is shown on Figure 9 for protons at 20 o
in the laboratory. The
highest spectrum corresponds to the non-corrected one and the lowest one to the corrected one. The
effect of the correction is increasing with decreasing energy bin due to the accumulation of
corrections coming from higher bins. Such correction corresponds to an overall effect of 13% for
protons, 10% for deuterons, 5% for tritons and is negligible for alpha.
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Figure 9. Proton spectrum before and after scattering corrections at 20 o .
3.1.2 Thick target corrections
Another correction consists of taking into account the energy lost in the target in order to get the
emitted energy from the measured energy. In Figure 10, the correlation between emitted energy and
mean measured energy is presented for triton and alpha. In both pictures, the dashed line characterises
the equality between both energies whereas dots show the effect of our thick lead target. For triton,
and the other hydrogen isotopes, the difference is small and of the order of our energy binning. It
implies that the correction is a simple shift in energy. On the contrary, the effect for alpha is
important and a special method is being developed.
Figure 10. Emitted energy versus measured energy for triton (left)
and alpha (right) for the used target. The dashed line corresponds to no thickness effect.
The last correction affects only the low energy particles, which are created without enough
energy to cross the entire target and to be detected. This indicates that only the particles created in a
fraction of the target, the part close to the output side, can be detected. It is then possible to determine
a so-called active target fraction (ATF) which can varied between 0 (nothing can escape) and 1
(everything can escape). The correction depends on the emitted energy and on the type of the particle.
Figure 11 shows the evolution of ATF as a function of the emitted energy for tritons and alpha. For
hydrogen isotopes, the correction starts below the maximum of the coulomb barrier down to 0 and its
effect is small due to the low population. For alpha, this effect goes up to 43 MeV implying a special
treatment, which is still under study.
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Figure 11. Active target fraction (see text) as a function of emitted energy.
4. Results
Dealing with all these corrections, cross-sections can be extracted for proton, deuteron and triton.
Using the Kalbach [6] systematic, it is possible to determine the differential cross-section in energy.
Figure 12 presents the dσ/dE for the proton (dots), deuteron (triangles) and triton (square). Energy
bins of 2 MeV have been used for proton and deuteron and of 3 MeV for triton. The proton spectrum
shows a smooth behaviour with a maximum around 18 MeV. For the deuteron spectrum, the
maximum is less pronounced and a small rise appears above 57 MeV due to direct processes. Since
our most forward angle is 20 o , we do not have enough information to fit properly this part of the
spectrum and we decide not to determine the cross-sections above 57 MeV for deuterons. For tritons,
the low statistic does not allow us to determine the cross-section above 47 MeV. The integration of
these spectra gives a production cross-sections of 290 ± 22 mb for protons, 70 ± 5 mb for deuterons
and 24 ± 2 mb for tritons.
Figure 12. dσ/dE for proton (dots), deuteron (triangles)
and triton (squares) in n + Pb at 62.7 MeV.
Before starting any comparison with theoretical calculations, it is interesting to compare our
experimental results with those found in the literature. No data exists on neutron induced reactions at
767

this energy and the deuteron data of [7] obtained in proton induced reactions are the only data
available to compare with. Since we are looking to deuteron production and that Bi and Pb are
neighbours, the production cross-sections in proton and neutron have to be similar in the preequilibrium
region. On Figure 13, double differential cross-sections are plotted for 20 o , 60 o and 160 o .
The black dots correspond to our data and the triangles to the Bertrand and Peelle one [7]. A good
agreement is found on the overall angular distribution.
d 2 σ
Figure 13. Deuteron at 20 o , 60 o and 160 o . The dots correspond
dEdΩ
to neutron induced reaction (this work) and triangles to proton induced reactions [7].
5. Comparison with theoretical calculations
As a first set of comparisons, we used two well-known code FLUKA and GNASH. The GNASH
data are coming from a publication of ICRU [8] whereas FLUKA [9] results have been obtained
locally. In Figure 14, the double differential cross-sections for protons are reported at 3 different
angles. The left column shows ICRU data as dashed line whereas FLUKA data are plotted on the right
column as dotted line. In all pictures, the black dots correspond to our data. The ICRU data
overestimates, in all spectra, our experimental results. In addition, it presents, at forward angles, a
double humped structure localised at low and high energy which is not present in our data that are
maximum at medium energy. FLUKA is giving a good total cross-section (270 mb) thanks to the
compensation of the underestimation of the medium energy part of the spectrum at forward angles
and the overestimation of the low energy part at backward angles. Nevertheless, the shapes of the
spectra are in close agreement with the data.
768

d 2 σ
Figure 14. Proton for n + Pb at 62.7 MeV. Dots are the experimental
dEdΩ
data whereas curves in the left (right) column correspond to ICRU [8] (FLUKA [9]) results.
For composite particles such as deuterons, the discrepancy is greater as is shown on Figure 15
where the dashed line corresponds to ICRU data and black dots to our experimental data.
d 2 σ
Figure 15. Deuteron for n + Pb at 62.7 MeV. Dots are the experimental
dEdΩ
data whereas curves correspond to ICRU [8] results.
6. Conclusion
Proton, deuteron and triton double differential cross-sections have been measured in 62.7 MeV
neutron-induced reactions on natural lead target. A special attention has been devoted to the
correction procedures coming from our use of thick target and collimators. Measurements were done
with a good statistic and are in good agreement with data of [7]. The comparison with some wellknown
theoretical data from GNASH-ICRU and FLUKA shows some disagreements. The largest
differences are found for GNASH-ICRU that neither reproduces the shape of the spectra nor the
769

absolute values of proton spectra. The composite particles are also not correctly reproduced. FLUKA
is giving a good total cross-section value thanks to differences cancelling each other at forward and
backward angles. Further comparisons with theoretical approach are underway especially with model
including pre-equilibrium emission such as MINGUS [10]. Other data on lead using proton-induced
reactions at the same energy beam are under analysis and will be delivered soon to enrich the data
tables.
This work has been supported by the European Commission under the concerted action N o FI4I-
CT98-0017 and by the GDR GEDEON (research group CEA – CNRS – EdF – FRAMATOME).
REFERENCES
[1] A. Bol, P. Leleux, P. Lipnik, P. Macq and A. Ninane, A Novel Design for a Fast Neutron Beam,
NIM, 1983, 214, 169-173.
[2] I. Slypen, V. Corcalciuc and J.P. Meulders, Geometry and Energy Loss Features of Charged
Particle Production in Fast-neutron Induced Reactions, NIMB, 1994, 88, 275-281.
[3] I. Slypen, V. Corcalciuc and J.P. Meulders, Proton and Deuteron Production in Neutron-induced
Reactions on Carbon at E n
= 42.5, 62.7 and 72.8 MeV, Phys. Rev. C, 1985, 51, 1303-1311.
[4] S. Benck, I. Slypen, V. Corcalciuc and J.P. Meulders, Light Charged Particle Emission Induced
by Neutrons with Energies Between 25 and 65 Mev on Oxygen I. Protons and Deuterons, Eur.
Phys. J. A., 1998, 3,149-157.
[5] GEANT: Detector Description and Simulation Tool, CERN program library long write-up W5013.
[6] C. Kalbach, Systematic of Continuum Angular Distributions: Extensions to Higher Energies,
Phys. Rev. C, 1988, 37, 2350-2370.
[7] F.E. Bertrand and R.W. Peele, Complete Hydrogen and Helium Particle Spectra from 30 to
60 Mev Proton Bombardment of Nuclei With A = 12 to 209 and Comparison with the
Intranuclear Cascade Model, Phys. Rev. C, 1973, 8, 1045-1064.
[8] International Commission on Radiation Units and Measurements, Nuclear Data for Neutron
and Proton Radiotherapy and for Radiation Protection, reports 63, 1999.
[9] A. Ferrari and P.R. Sala, A New Model for Hadronic Interaction at Intermediate Energies for
the FLUKA Code, The MC93 international conference on Monte Carlo Simulation in High
Energy and Nuclear Physics, Tallahassee, Florida, 1993, 277-288, World scientific, Singapore.
[10] A.J. Koning, M.B. Chadwick, Microscopic Two-component Multistep Direct Theory for
Continuum Nuclear Reactions, Phys. Rev. C, 1997, 56, 970-99.
770

HIGH AND INTERMEDIATE ENERGY NUCLEAR DATA FOR
ACCELERATOR DRIVEN SYSTEMS – THE HINDAS PROJECT
J.P. Meulders 1 , H. Beijers 2 , J. Benlliure 3 , O. Bersillon 4 , J. Cugnon 5 , Ph. Eudes 6 ,
D. Filges 7 , A. Koning 8 , J.F. Lecolley 9 , S. Leray 10 , R. Michel 11 , N. Olsson 12 ,
K.H. Schmidt 13 , H. Schuhmacher 14 , I. Slypen 1 , H. Synal 15 , R. Weinreich 16
1 Institut de Physique Nucléaire, Université Catholique de Louvain, 1348 Louvain-la-Neuve, Belgium
2 Kernfysich Versneller Instituut, Rijksuniversiteit Groningen, 9747 Groningen, The Netherlands
3 University of Santiago de Compostela, 15706 Santiago de Compostela, Spain
4 CEA-Bruyères-le-Châtel, Service de Physique Nucléaire, 91680 Bruyères-le-Châtel, France
5 Institut de Physique B5, Université de Liège, 4000 Sart Tilman Liège 1, Belgium
6 Subatech, Université de Nantes, Ecole des Mines de Nantes, IN2P3-CNRS, 44307 Nantes, France
7 Forschungszentrum Jülich GmbH, Institut für Kernphysik, 52425 Jülich, Germany
8 Nuclear Research and Consultancy Group, 1755 Petten, The Netherlands
9 Laboratoire de Physique Corpusculaire de Caen,
IN2P3-CNRS/ISMRa et Université de Caen, 14050 Caen, France
10 DAPNIA/SPhN CEA/Saclay, 91191 Gif-sur-Yvette, France
11 Zentrum für Strahlenschutz und Radioökologie, Universität Hanover, 30167 Hanover, Germany
12 Department of Neutron Research, Uppsala University, 75121 Uppsala, Sweden
13 Gesellschaft für Schwerionenforschung mbH, 64291 Darmstadt, Germany
14 Physikalisch Technische Bundensanstalt, 38116 Braunschweig, Germany
15 Institute of Particle Physics, ETHZ, 8093 Zurich, Switzerland
16 Paul Scherrer Institute, 5232 Villingen-PSI, Switzerland
Abstract
The HINDAS project (High and Intermediate energy Nuclear Data for Accelerator driven Systems) is
a three years project supported by the European Commission under the Fifth Framework Program. The
gathering of 16 partners, both experimentalists as theoreticians, allows to measure the wealth of new
nuclear reaction cross-sections in the energy range between 20 MeV and 2 GeV on 3 elements of
crucial importance for ADS systems: Pb as a target element, U as an actinide and Fe as a shielding
element. The new experimental data will help to benchmark the existing theoretical models or to
improve them. The assembly of nuclear data tables on those elements will allow interpolating to other
elements appearing to be important in the design and the construction of an European ADS
demonstrator.
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1. Introduction
The HINDAS project (High and Intermediate energy Nuclear Data for Accelerator driven
Systems) is supported by the European Commission under the Fifth Framework Program (September
2000-August 2003) and involves 16 European laboratories. Its general objective is to obtain a
complete understanding and modelling of nuclear reactions in the 20-2 000 MeV region, in order to
build reliable and validated computational tools for the detailed design of the spallation module of an
accelerator driven system. This essential goal can only be accomplished by means of a well-balanced
combination of basic cross-section measurements, nuclear model simulations and data evaluations.
Therefore, three nuclides, Fe, Pb and U have been chosen which provide a sufficiently broad
coverage of the periodic table and are representative of the target, structure and core materials of the
ADS. Hence, not only a few of the top-priority materials are chosen but, more importantly, with
detailed theoretical and experimental knowledge of these particular elements, the nuclear models
present in the foreseen simulation codes of this project will be fine-tuned. These will be employed to
generate nuclear codes and data libraries for the materials that are requested by the ADS community.
The measurements will be performed at six nuclear physics laboratories in Europe, where beams
of proton, neutron and heavy ion (in conjunction with inverse kinematics) as well as relevant
measurement are available.
There appear to be a transition region around 200 MeV for the theoretical models. In the
20-200 MeV region, the theoretical calculations include direct interaction, pre-equilibrium, fission and
statistical models, all with many uncertainties. Above 200 MeV, the theoretical analysis includes the
intra-nuclear cascade model together with fission and evaporation models. A similar transition appears
at about the same energy in the experimental facilities and in the measuring techniques.
The HINDAS project is therefore divided in experimental as well as theoretical work packages,
according to this energy limit. The detection techniques differ also substantially for neutrons, protons
and residual nuclide production, which has motivated a further division of the work packages
according to the detected nuclides. This paper presents these work packages with the different results
that would be available at the end of the project.
2. Experimental work between 20 and 200 MeV
The experimental work in the intermediate energy range from 20 MeV to 200 MeV is condensed
in 3 parts. The measurements of cross-sections of nuclear reactions induced by protons and neutrons
on Fe, Pb and U targets cover 2 work packages and the measurement of residual nuclide production is
the object of the third part. Those experiments will be performed at 4 European accelerators which
allows to cover the entire energy range from 20 to 200 MeV for the proton beams and two of those
facilities can produce monoenergetic neutron beams with time of flight facilities.
2.1 Light charged-particle production induced by neutrons or protons between 20 and 200 MeV (WP1)
The measurement of double differential cross-sections of light charged-particles p, d, t, 3 He and
induced by protons or neutrons on the different chosen targets will be obtained by measuring the
energy spectra of each charged-particle over a large angular range from 15° to 160°. By integration
over the angle, energy differential cross-sections are obtained at each angle, and by integration over
the energy, angular differential cross-sections of the produced particle are deduced. The information
contained in the double differential cross-sections is very stringent for theoretical models of nuclear
772

eactions, since the pre-equilibrium reactions have to be taken into account, in addition to the direct
interaction contribution at the high energy part of the spectrum and the statistical evaporation
component at the low energy side of the spectra.
The (p,xlcp) reactions on Pb and U will be measured by the partners of UCL, Subatech and LPC-
Caen at 65 MeV at the CYCLONE cyclotron (UCL, Louvain-la-Neuve), and the same reactions on Fe,
Pb and U at 135 MeV by the partners of Subatech, LPC-Caen and RuG at the KVI cyclotron
(Groningen, The Netherlands) (see also N. Marie, this meeting). For these measurements, 8 triple
telescopes (Si-Si-CsI) allow to measure the light charged-particles over their entire energy range (with
low energy thresholds). Figure 1 shows a picture of the reaction chamber with the triple telescopes.
Figure 1. The reaction chamber and the triple telescopes used in the (p,xlcp) reactions
The (n,xlcp) reactions on Fe, Pb and U will be measured at 65 MeV at the CYCLONE cyclotron
(UCL, Louvain-la-Neuve) (see also M. Kerveno, this meeting). Six (-E telescopes (NE 102 plastic
scintillator – CsI(Tl) detector) detect the charged particles produced by the neutrons on the target. The
information from the telescopes coupled to the time of flight method with excellent time resolution
(less than 1 ns) allows reconstructing, event by event, the energy spectra for each ejectile. Double
differential cross-sections are obtained for the neutron mono-energetic peak (~63 MeV) and also for
energies from the continuum of the neutron energy spectrum [1] (from 30 to 57 MeV).
The (n,xlcp) reactions on Fe and Pb at 100 MeV will be measured at The Svedberg cyclotron
(UU, Uppsala) with a similar detection setup as for the proton induced-reactions. Partners of Subatech
(Nantes), LPC-Caen and UU (Uppsala) are involved in these measurements ([2] and F.R. Lecolley,
this meeting).
Finally, charged-particles multiplicities will be measured in proton-induced reactions on Fe, Pb
and U in the energy range between 130 and 200 MeV by partners from RuG (Groningen).
2.2 Neutron production induced by neutrons and protons (WP2)
Partners of UU and LPC-Caen study elastic neutron scattering (n,n), at 100 MeV for Fe and Pb [2].
Such data are important to determine the nuclear optical potential to high precision in an energy range
where data are essentially lacking. With this model at hand, cross-sections for elastic scattering, which is
the most important reaction channel in the moderation and transport of the source neutrons, can be
773

calculated. Moreover, the optical potential is a necessary component in the description of many other
reaction channels, since it accounts for the behaviour of a neutron entering or emerging from a nucleus.
The measurements will be performed using a recently developed detector set-up, consisting of
two identical detector sets, which can be arranged to cover, e.g., the 10-50 and 30-70 degree ranges.
Each detector set consists of a front veto scintillator, a 1 cm thick plastic scintillator for conversion
into recoil protons, two drift chambers with x-y position sensitivity for proton tracking, and an array of
12 large CsI detectors for proton energy measurement. Absolute cross-sections will be determined by
comparison with the reasonably well-known neutron-proton scattering cross-section.
Furthermore it is proposed to study the feasibility of (n,xn) reactions on Pb at 100 MeV. Such
experiments are difficult to perform, but information is of great importance to understand and improve
quantum-mechanical multi-step direct and classical pre-equilibrium models, as well as statistical
models built on multiple Hauser-Feshbach emission. The measurements will make use of part of the
previously described set-up, together with an active target for conversion of the emitted neutrons into
recoil protons. The active conversion target will be positioned outside the neutron beam, but close to
the Pb target, to obtain a large solid angle for neutrons. The recoil protons will be traced by a couple of
drift chambers, and finally the energy will be determined in the CsI detector array.
Finally, the measurement of double-differential spectra from (p,xn) reactions in Pb and U, using a
65 MeV proton beam will be performed by partners of LPC-Caen, Subatech and UCL. In these
experiments the emitted neutrons will be detected by well-shielded NE213 neutron detectors, placed
around the scattering centre to measure angular distributions. The neutron energy distribution will be
determined using time-of-flight techniques. Neutrons will be distinguished from gamma-rays using the
pulse shape discrimination properties of this kind of detectors.
2.3 Residual nuclide production induced by neutrons and protons and production of long-lived
radionuclides (WP3)
Reliable cross-sections for the production of residual nuclides by medium-energy proton- and
neutron-induced reactions are essential for ADS to calculate the radioactive inventories of the
spallation target, of structural materials and of ambient matter. Production of residual nuclides by GeV
protons in thick or massive targets are a complex phenomenon the modelling of which needs to follow
in detail the inter- and intra-nuclear cascades, the production and transport of primary and secondary
particles. The spectra of primary and secondary particles strongly depend on the material irradiated as
well as on geometry and depth inside the target. To calculate activation rates and radioactive
inventories such calculated spectra have to be folded with the energy-dependent cross-sections of the
underlying nuclear reactions for energies from thresholds up to the initial energy of the primary
particles. Presently, there is no model or code available to predict the required cross-sections with an
accuracy of better than a factor of two on the average. Therefore, one has to rely for the important
nuclides on experimental cross-sections. Such experimental cross-sections are also needed if one tries
to improve models and codes as a basis for validation.
Due to the importance of nuclear reactions of secondary particles, neutron-induced reactions will
dominate the radionuclide inventory of the spallation target though the high-energy primary protons
will significantly contribute. As a consequence, one needs cross-sections for both proton- and neutroninduced
reactions for a reliable modelling of residual nuclide production over the entire energy range.
The data to be determined in this section will provide an experimental basis to calculate such
inventories of the spallation target, of shielding and structural materials for an accelerator driven
system a few minutes after shut-down as well as to validate theoretical work which is needed to
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calculate the very short-lived radionuclides which make up an essential part of the spallation target
during operation of a facility. With respect to the long-term behaviour and the final disposal of
spallation targets and structural materials the precise modelling of long-lived radionuclides will be
essential. Up to now, there are no inventory calculations which take into account long-lived
radionuclides, mainly due to the lack of respective cross-sections.
For the modelling of radionuclide inventories it will be sufficient as a first approximation to have
neutron-cross-sections up to 200 MeV. Measurements of residual nuclide production induced by
neutrons between 30 and 180 MeV are foreseen. For proton-induced reactions one needs the complete
excitation functions up to the energy of the primary beam. The latter do exist from recent work of our
collaboration for most relevant target elements [3]. Measurement of production cross-sections of longlived
radionuclides via accelerator mass spectrometry (AMS) after chemical separation (partners ZSR
and ETHZ) will be performed between 40 and 75 MeV.
3. Experimental work between 200 and 2 000 MeV
The aim of this work will be to collect high quality data and compare them with the state-of-theart
nuclear models. Data will be either measured in the framework of the project or have already been
measured by the partners but not yet fully interpreted. In any case, they will be delivered as a ready-touse
file to be included in international data banks.
Particular attention will be paid to the impact of the new data for applications. Calculations of
several quantities important in the design of ADS target or window will be performed using standard
High Energy Transport Codes. In these codes the elementary cross-sections generated by the old nuclear
models will be replaced either by the most recent version of models from J. Cugnon validated on data
from our collaboration or, when possible and if models are not yet reliable enough, directly by the
measured cross-section. Errors or uncertainties due to the use of the standard codes will be assessed.
3.1 Light charged-particle production (WP4)
This part will be devoted to the collection of data concerning the production of light chargedparticles.
These data are important to probe the high-energy nuclear models in which the competition
between neutrons and charged-particles, and the emission of composite nuclei (deuterons, alphas) are
not yet treated satisfactorily. Moreover, the production yields of hydrogen and helium are essential for
estimation of gas production in the window or structure materials of an ADS.
Production cross-sections for hydrogen and helium are being measured using a 4π silicon ball
detector. So far, experiments have been performed at 0.8, 1.2, 1.8 and 2.5 GeV on several targets [4]
and further experiments are foreseen. These measurements are also performed in coincidence with
neutron multiplicity distributions. This allows studying the production rates of protons and alphas as a
function of the excitation energy in the nucleus remaining at the end of the Intra-Nuclear Cascade
stage. All these data will be analysed and compared to high-energy nuclear models.
Implications of the results from this experiment for gas production in some of the components of
an ADS will be assessed, for instance on the lifetime of the window or on structure materials.
Moreover, a new magnetic spectrometer able to measure with a high resolution doubledifferential
cross-sections for the production of light charged-particles (induced by protons) in
coincidence with low energy neutrons will be designed.
775

Figure 2. The Berlin ball detector system
3.2 Neutron production induced by protons in thin and thick targets (WP5)
In this work-package, different types of neutron production data measured recently by the
partners in both thin and thick targets will be collected, cross compared and compared with models.
Up to recently, very little high-quality data concerning double-differential cross-sections of
neutron production were existing above 800 MeV and below there were significant discrepancies
between different sets of data. Partners of CEA-Saclay, CEA-Bruyères and LPC-Caen have measured
neutron energy spectra and complete angular distributions using two complementary experimental
techniques: time-of-flight for the low energy part of the neutron spectrum and neutron-proton
scattering on a liquid hydrogen converter with a magnetic spectrometer measuring the momentum of
the recoiling proton for high energy neutrons. This has allowed to obtain energy spectra of (p,xn)
reactions with a high resolution from 2 MeV to the incident energy, on several targets at 800, 1 200
and 1 600 MeV [5]. The same apparatus was used to measure neutron energy spectra from thick
targets with different length and diameters.
Partner from FZJ has participated to a collaboration using a 4π liquid scintillator detector able to
measure event-wise the multiplicity of neutrons up to 150 MeV on both thin and thick targets of
different length and diameter for incident proton energies of 0.4, 0.8, 1.2, 1.8 and 2.5 GeV over a wide
range of structural and target materials for ADS applications [4]. The neutron multiplicity distribution
in thin targets reflects the excitation energy distribution of the nucleus remaining at the end of the
Intra-Nuclear Cascade stage and is therefore important to understand the reaction mechanism. The
average value of the neutron multiplicity distribution in thick targets is directly interesting for
applications.
The measurements performed by FZJ and CEA-Saclay, CEA-Bruyères, LPC-Caen are
complementary both for technical (energy range of the measurements) and physics reasons (highenergy
neutrons test the intra-nuclear cascade stage while low energy neutrons probe the evaporation
stage). So far, no coherent simultaneous analysis of both experiments has been done. This will be the
goal of this work-package in which, for example, the average multiplicity distributions measured with
the neutron ball will be compared to those inferred from the integration of the double-differential
776

cross-sections; the secondary reactions induced in the neutron scintillator detector will be assessed
using results of the high energy neutron spectrum measured by the double-differential cross-section
experiment, etc.... Comparisons with the same high energy nuclear models for thin targets, the same
high-energy transport codes for thick targets, taking into account the rather complex experimental
acceptance of both experiments, will be performed. Results will be used to assess the remaining
deficiencies in the codes to be improved in the theoretical section of the HINDAS project. Simulation
of thick target results will also be realised. Direct applications of the thick target experiments such as
average neutron multiplicities or high-energy neutron leakage for shielding estimation will be
discussed.
3.3 Residual nuclide production in inverse kinematics (WP6)
In spallation reactions of heavy nuclei induced by protons of about 1 GeV, mostly short-lived
radioactive nuclei are produced. The spallation residues are stopped inside the target. They decay
towards stable isobars predominantly by beta decay. After irradiation, long-lived radioactive residues
are identified in mass and atomic number by gamma spectroscopy and by accelerator mass
spectrometry. These experiments provide reliable and comprehensive data on cumulative yields, from
which long-lived activities and final element yields can be deduced. In addition, these techniques
allow for measurements over a large range of bombarding energies. A previous inter-comparison with
available data has revealed that the calculations with nuclear-reaction models are not realistic enough,
but it is difficult to pin down the deficiencies of the models on the basis of cumulative yields. For this
purpose, a complete systematic of isotopic production cross-sections emerging from the nuclear
reaction is urgently needed.
In particular for proton energies above 200 MeV, a substantially different technique, based on the
use of inverse kinematics, has been developed recently which allows identifying all short-lived
radioactive nuclides produced as spallation residues prior to beta decay. Heavy nuclei are provided as
projectiles, impinging on a liquid-hydrogen target. The spallation residues are identified in-flight in a
high-resolution magnetic spectrometer. These experiments allow a much more direct insight into the
reaction mechanism than experiments in normal kinematics and therefore are best suited to improve
nuclear-reaction models which are known to be unable to reproduce available data. In addition, this
technique allows to determine the kinetic energies of the spallation residues [6], an information of
highest importance for estimating radiation damages in structure material of an ADS. That means that
these experiments provide unique and valuable information which complements the results obtained in
normal kinematics. Due to electronic interactions in the spallation target, the primary protons loose
energy and induce nuclear reactions in a wide energy range. However, the higher energies are
particularly important for residual-nuclide productions, since more than 75% of the primary protons of
1 GeV undergo nuclear reactions in the spallation source in an energy range above 700 MeV.
Additional measurements with a liquid deuterium target are aimed to provide information on
spallation reactions induced by neutrons.
The experiments in inverse kinematics and the data analysis being rather complex, only few
projectile species and energies can be investigated. Therefore, the measurements are restricted to 208 Pb,
238 U and 56 Fe at 1 A GeV and partly at 500 A MeV. During the 3-year period of the project, final data
on 208 Pb and 238 U will be available. It is expected that the full isotopic distributions and kinetic
energies obtained in inverse kinematics in combination with detailed excitation functions of specific
reaction products obtained in normal kinematics provide sufficient information to develop
substantially improved nuclear-reaction models which can then be used in transport codes to predict
realistic energy-integrated production yields in thick targets.
777

Finally, a new experimental technique will be developed to also measure neutrons and light
charged-particles in inverse kinematics. This will allow establishing coincidences between these
particles and the heavy residues, an information still more relevant for modelling the nuclear reaction
correctly.
Calculations of the activities, radiotoxicities and element distributions in a realistic lead spallation
target will be performed using transport and evolution codes. The elementary cross-sections generated
by the old nuclear models will be replaced either by the most recent version of models from
J. Cugnon, or directly by the measured production yields on Pb at 1 000 MeV, extrapolated at nonmeasured
energies using the energy dependence of the excitation functions measured in WP3.
4. Theory and evaluation
For research on accelerator-driven systems, cross-sections for the important materials need to be
known for ALL possible outgoing channels, outgoing channels, outgoing energies and angles. This
total amount of required information is so large that experiments alone can never cover the nuclear
data needs for ADS. To fill this gap, the data are simulated computationally, with the help of
theoretical reaction models. The development of this simulation is done in close correspondence with
the experiments: adjustable parameters of the theoretical models are adjusted in such a way that the
latter reproduce the measurements as closely as possible. The critical assumption is then that the
models can also be used in areas where no measurements exist. Hence, the actual provision of nuclear
data in a form usable for ADS design will be done in two work-packages
• Nuclear data libraries, improved and extended up to 200 MeV, based on nuclear models.
• Intra-nuclear cascade models and codes for the higher energies.
4.1 Nuclear data libraries and related theory (WP7)
This part concerns nuclear model calculations for a theoretical analysis of the between 20 and
200 MeV and predictions for the unmeasured channels for energies up to 200 MeV. In combination,
this will be used to construct complete nuclear data libraries for 56 Fe, 208 Pb and 238 U up to 200 MeV,
which will show a clear improvement over all other existing nuclear data files and methods [7].
Theoretical calculations will be performed with a variety of nuclear models at NRG-Petten and at
CEA-Bruyères-le-Châtel. The new model code system will be extended to include a proper treatment
of all channels precisely. Coupled-channels optical models will be constructed for the simulation of
the elastic and inelastic channels, not only for the total (angle-integrated) cross-sections, but also for
the angular distributions. For the continuum reactions, complete outgoing energy and angular spectra
will be included for all light particles. These will be predicted, and compared with the new
measurements, using quantum-mechanical multi-step direct (MSD) and classical pre-equilibrium
models that include novel models for microscopical particle-hole level densities and the optical
models. Multiple pre-equilibrium emission beyond the second step will be included for the highest
incident energies. Complete evaporation of the residual nuclides is accounted for by means of multiple
Hauser-Feshbach emission that includes competition of all possible outgoing particle channels and
fission, while conserving energy, angular momentum and parity. Simultaneously with the doubledifferential
spectra, the calculated residual production cross-sections will be compared with the
experiments as described in WP3. Both types of observables must be described by one and the same
calculation. High-energy fission will also be included by means of an extension of the Brosa model.
778

All possible nuclear reactions will be evaluated simultaneously, in order to ensure flux conservation
and energy balance. The results will be compared with the American GNASH code.
The calculated results will be processed automatically into the ENDF6-format. The results will be
combined with the data below 20 MeV to come to one consistent final library. If the existing data file
below 20 MeV turns out to be inadequate, the cross-sections will be improved in the low energy
regime as well to ensure a smooth transition from low to high energies. All the nuclear data will be
stored in the common ENDF-6 format and will be checked according to a standard QA-system. As
basis for the new high-energy evaluations, the European JEFF library will be used.
4.2 High-energy models and codes (WP8)
The high energy codes, although globally rather successful, suffer from some deficiencies. Both
their inter-comparison and the comparison with existing experimental data reveal, in some identified
regime, discrepancies which are beyond the accuracy required by the engineers working on projects of
ADS or spallation sources. These observations call for improvements of the physics already included
in the cascade codes (in-medium corrections, Pauli principle, mean field dynamics,...), of the
evaporation codes (level densities) and of fission codes (viscosity, evaporation-fission competition at
high excitation energy) [8]. These improvements are part of the specific theoretical task in this project.
They will be realised in successive steps at Ulg-Liège.
The first step will consist of improving the existing codes by including physics aspects not included
so far and by refining some of the physics which is already implemented. For the most recent intranuclear
cascade (INC) code, this concerns a proper description of the nuclear surface, an improvement of
Pauli blocking, which present too much fluctuations, and refinements of the in-medium corrections. For
the evaporation codes, the first step will involve a careful examination of the input data and an advanced
development of the fission model at high excitation energy, taking advantage of the forthcoming
measurement of the fission component in reverse kinematics (see WP6). The improvements will be
inspired by the most recent theoretical progress in nuclear dynamics far from equilibrium.
The second step aims at a validation of the improved codes (and other standard codes). An
extensive comparison with the neutron differential cross-sections measured at SATURNE (800 to
1 600 MeV) and with the neutron multiplicities and light charged-particle spectra measured at Jülich
(WP5) will be performed, for both thin and thick target data. In addition, an extensive comparison
with the experimental residue production data to be provided by WP6 will be realised.
As a third step, a new improvement of the codes will be undertaken, if necessary. This work will
involve an adjustment of the introduced parameters to describe less well known physics aspects, like
the parameters regulating the coupling between the INC and evaporation codes and some parameters
of the fission model, especially viscosity.
The final goal will consist in the elaboration of a version of a high-energy transport code
including these new simulation tools. This version could be tested on the thick target data generated by
this project.
Acknowledgements
All partners thank the European Commission for supporting the project HINDAS under contract
number FIKW-CT-2000-00031.
779

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780

A STUDY ON BURNABLE ABSORBER FOR A FAST SUB-CRITICAL REACTOR HYPER
Yong Hee Kim, Won Seok Park
Korea Atomic Energy Research Institute
P.O. Box 105, Yusong, Taejon, 305-600, Korea
Jong Seong Jeong
Department of Nuclear Engineering, Seoul National University
Shinrim-dong Kwanak-ku Seoul 151-742, Korea
Abstract
This paper is concerned with development of burnable absorber technologies for an accelerator driven
system (ADS) with fast neutron spectrum, HYPER (Hybrid Power Extraction Reactor). Concerning
the ADS loaded with TRUs (Transuranic Elements), one of the major problems is a large burn-up
reactivity swing and the consequent unfavourable slanting of the radial power distribution over a
depletion period. In order to reduce the reactivity drop during core burn-up, B 4 C is introduced as a
burnable absorber and its efficacy is evaluated for the HYPER system. Taking into account the
radioactive TRU fuel, the inner surface of the fuel clad is coated with a thin B 4 C layer. Two concepts
of the burnable absorber application are considered, homogeneous and heterogeneous loading of B 4 C.
In the homogeneous application, B 4 C is used in all fuel rods, while the burnable absorber is utilised
only in the outer zone of the core in the heterogeneous loading. The burn-up characteristics of the
HYPER cores with and without the B 4 C burnable absorber are analysed with a Monte Carlo code,
called MCNAP.
781

1. Introduction
In Korea, an accelerator driven system (ADS), which is called HYPER (Hybrid Power Extraction
Reactor) is currently under development for transmutation of TRUs (Transuranic Elements) [1].
Concerning the uranium-free fast reactors like HYPER, one of the big problems is a very large
reactivity swing, regardless of the sub-criticality of the core. In an ADS, a large burn-up reactivity
swing means a large reservation of the proton beam current. This large reserved proton current may
result in several unfavourable safety features as well as adverse impacts on the economics of the
system. Also, another concern associated with the large reactivity change is a one-way change of the
radial power distribution during depletion of the core. To resolve this problem, an on-power refuelling
concept, as in CANDU, was studied previously for HYPER [1]. However, the on-power refuelling
makes the system fairly complex and may cause serious engineering concerns.
Several types of burnable absorbers such as boron, gadolinium, and erbium are successfully used
to suppress the initial excess reactivity and to control the power distribution in thermal reactors like
PWRs [2]. Regarding the critical fast reactors such as LMR (Liquid Metal Reactor), poisoning the
core with a burnable absorber is not used to control the reactivity. This is mainly because the excess
reactivity in conventional LMRs is fairly small due to self-generation of fissile elements and thus the
burn-up reactivity swing can be easily controlled by control rods. Of course, it is well recognised that
there is no effective burnable absorber for fast neutron systems due to small neutron capture crosssections.
When it comes to the fast-neutron ADS loaded with TRUs, however, the situation is quite
different from those of the conventional critical thermal and fast reactors. Basically, any excess
reactivity, which should be suppressed by external control mechanisms, should not be allowed in ADS
in order to guarantee its surmised advantages. Consequently, control rods or absorber-containing
coolant cannot be used to control the reactivity of an ADS. Thus, fixed burnable absorbers, if any,
could only be utilised as the reactivity control mechanism for ADS.
Previously, Stone et al. [3] studied a dual spectrum core to reduce the reactivity swing of the
ATW core, where a thermal spectrum zone is placed in the periphery of the core and 237 Np and 241 Am
are loaded in the thermal region. They showed that the reactivity swing could be reduced by a factor of
2 in the dual spectrum ATW. However, the smaller reactivity change in the modified ATW is mainly
due to the reduced power density. In addition, the dual spectrum core may lead to a large power
peaking in the interface region between hard and soft spectrum regions.
The simplest way to reduce the reactivity swing is to adopt a low power density core. However, a
low power density needs a large core volume, thus it is not favourable from the economics point of
view. Recently, Hejzlar et al. [4] studied burnable absorbers for a critical Pb-Bi-cooled transmutation
reactor. They evaluated various candidate materials such as B, Re, Hf, Gd, Er, etc. They showed that
10 B has the largest neutron capture cross-sections and can reduce the reactivity swing a little. Finally,
they discarded the burnable poison option, in favour of excess reactivity compensation through control
rods.
In this paper, we have re-evaluated the potential of 10 B as a burnable absorber for the sub-critical
HYPER core to reduce the burn-up reactivity swing and also to control the radial power distribution.
All calculations are performed with a Monte Carlo code, MCNAP [5], which was developed at Seoul
National University, Korea. It is worthwhile to note that MCNAP has its own built-in depletion
routine
782

2. Burnable absorber for HYPER
2.1 The HYPER core
HYPER is a Pb-Bi-cooled ADS under development at KAERI with the aim of transmuting both
TRUs and LLFPs such as 99 Tc and 129 I. The HYPER system is rated at 1 000 MW th thermal power and
the minimum required sub-criticality is k eff = 0.97. Figure 1 shows a schematic configuration of the
evolving HYPER core. In HYPER, a linear accelerator produces the proton beam of 1 GeV and the
proton impinges on the Pb-Bi target in the core central region, generating about 28.84 spallation
neutrons a proton. The proton beam is delivered to inside of the core through a beam tube to maximise
the source neutron importance and also to obtain favourable axial power distribution. For emergency,
3 locations are reserved for safety zones. The fuel blanket region is divided into 3 TRU enrichment
zones (low, medium, high) to obtain acceptable radial power distribution. The low and high TRU fuels
are loaded in the innermost and outermost zones, respectively.
A unique feature of the HYPER core is transmutation of 99 Tc and 129 I in a localised thermal
neutron zone. Inner region of the FP (Fission Product) assembly is composed of I and moderator
(CaH 2 ) rods to produce thermal neutron and 99 Tc-is placed in the peripheral region to block the
thermal neutron leakage into the neighbouring fuel assemblies [6]. Currently, two fuel types are
considered for HYPER, one is the TRU-Zr metal and the other one is the TRU-Zr dispersion fuel,
where TRU-Zr particles are dispersed in Zr matrix. In this work, the dispersion fuel is assumed. Spent
fuels from PWRs of 33 GWD/MTU burn-up, after 30-year cooling time, are reprocessed with a
pyrochemical processing and then recycled into the HYPER core. In the present work, a uranium
removal rate of 99.9% is assumed. Consequently, the HYPER core is not completely free from
uranium elements, instead, uranium occupies about 9 w/o in the fuel as shown in Table 1.
Table 1. Feed fuel composition in weight percent
(33 GWD/MTU, 30-year cooling)
Isotopes
234 U
235 U
236 U
238 U
237 Np
238 Pu
239 Pu
240 Pu
241 Pu
242 Pu
241 Am
242m Am
243 Am
243 Cm
244 Cm
245 Cm
246 Cm
Weight percent (w/o)
0.2000E-2
0.7894E-1
0.3840E-1
0.8920E+1
0.4449E+1
0.9909E+0
0.4756E+2
0.2168E+2
0.2689E+1
0.4101E+1
0.8649E+1
0.3868E-2
0.7591E+0
0.1207E-2
0.6604E-1
0.7321E-1
0.8515E-3
783

Figure 1. Configuration of the Pb-Bi-cooled HYPER core (183 fuel assemblies)
2.2 B 4
C-coated cladding
In order for a material to be an effective burnable absorber, its neutron capture cross-section
should be much larger than those of fuel elements. Also a neutron capture of a burnable absorber
should not generate nuclides with large capture cross-sections and at the same time daughter nuclides
should be naive in terms of radiotoxicity. Taking into account the above constrains on burnable
absorbers, 10 B seems to be the best candidate for the burnable absorbing material of the HYPER core.
10 B absorbs a neutron through ( , γ )
reaction, is an exothermic process:
n or ( , α)
n reaction. The ( n , α)
reaction, i.e. helium production
B +
10 1 7 4
5
+ n0
→ Li3
+ He2
Q
where Q is about 2.79 Mev for thermal neutrons and is a little larger in fast neutron systems. Table II
compares one-group effective cross-sections of boron and plutonium isotopes in the HYPER fuel
assembly. As shown in Table 2, the capture cross-section of 10 B is a little larger than the fission crosssection
of 239 Pu, the major fissile isotopes of the TRU fuel. Neutron absorptions of 11 B and Li-7 are
negligibly small. Table 2 shows that the depletion rate of 10 B is a little faster than that of 239 Pu even in
very hard neutron spectrum.
,
784

Table 2. One-group effective cross-sections of boron and plutonium isotopes in the HYPER core
σ ( n , γ ) , barn ( , α)
σ n , barn σ f , barn
10 B
11 B
7 Li
238 Pu
239 Pu
240 Pu
241 Pu-
2.978E-4
4.468E-5
3.097E-5
0.631
0.397
0.427
0.375
2.307
–
–
–
–
–
–
–
–
–
1.089
1.693
0.358
2.296
The ( n , α)
reaction of 10 B is exothermic and produces helium gas as well as Li-7. Therefore, care
must be taken, when using 10 B as a burnable absorber in nuclear reactors. It is well known that direct
mixing 10 B with fuel impairs the fuel integrity since the fuel swelling is enhanced due to helium gas
and liquid-phase Li-7. To overcome this limitation, Westinghouse has developed a burnable absorber
technology for PWRs, where the fuel rod is coated with ZrB 2 [2], and achieved successful
performance. Thickness of the ZrB 2 layer is about 0.002 cm.
Unfortunately, it is not easy to use the Westinghouse approach directly for HYPER, since the
TRU fuel of HYPER is very radioactive. Therefore, we have used a slightly different option, i.e.
B 4 C-coated cladding, where the inner surface of cladding is coated with B 4 C. In the present work, B 4 C
is used, instead of ZrB 2 , since it is easily available and has more boron elements than ZrB 2 . Thickness
of the B 4 C layer is 0.0009 cm or 0.0012 cm. In the natural boron, abundances of 10 B and 11 B are 19.8%
and 80.2%, respectively. In order to maximise the 10 B loading, it is assumed that 10 B is enriched up to
90% atomic percent in this paper. Figure 2 shows the burnable absorber rod for the HYPER core. It
should be noted that two cutback regions, where the absorber is not applied, are adopted to flatten the
axial power distribution.
Figure 2. Fuel rod with B4C-coated cladding in HYPER
785

3. Numerical results
The performance of the burnable absorber in Section 2.2 is evaluated in terms of spallation neutron
multiplication and radial power distribution. Two types of burnable absorber applications are compared
with the unpoisoned reference HYPER core, one is a homogeneous loading of B 4 C (HYPER-HBA) and
the other one is a heterogeneous loading (HYPER-OBA). All fuel rods have the B 4 C coating in the
homogeneous application, while the burnable absorber is used only in middle and outer zones of the core
in the heterogeneous loading of B 4 C. In HYPER-HBA, a B 4 C layer of 0.0012 cm thickness is used and a
little thinner layer, 0.0009 cm, is utilised in HYPER-OBA to enhance the 10 B depletion rate. Numerical
tests are conducted for the initial core, which has the largest burn-up reactivity swing.
All calculations were done with the MCNAP code for a 3-dimensional model of the HYPER core,
where each assembly was homogenised by using the volume-weighting method and treated as an
independent cell and the active core was divided into 6 segments in the axial direction. The zone-wise
TRU enrichments were adjusted such that the initial k eff should equals 0.97 at the beginning of cycle
(BOC) and also the radial power distribution should be acceptable. The radial power distribution is not
optimised since the objective of this work is to evaluate the potential of the B 4 C burnable absorber.
Table 3 shows TRU enrichments of the three cores, k eff values, and corresponding multiplication
factors at BOC. TRU inventory of HYPER-HBA is about 24% larger than that of the reference core
and it is increased by about 13.4% in HYPER-OBA. Initial sub-criticality was determined by a critical
mode calculation and the depletion calculations were based on the fixed-source mode with a 30-day
time step. In the fixed-source calculations, a generic source distribution was assumed.
Table 3. Zoning of TRU enrichment and initial k eff
(L = Low, M = Medium, H = High)
Core type
TRU enrichment (w/o)
Reference L (19.18), M (24.70), H (30.60)
TRU Loading: 2774.58 kg
HYPER- L (22.63), M (29.07), H (36.18)
HBA TRU Loading: 3436.46 kg
HYPER- L (19.22), M (28.15), H (34.05)
OBA TRU Loading: 3146.11 kg
a) Standard deviation.
10 B (kg) k eff Multiplication
(M s
)
25.281
– 0.96975
(0.0010) a) (0.043) a)
21.858 0.96940 21.547
(0.0010) (0.050)
13.176 0.97021 21.946
(0.0011) (0.043)
The burn-up reactivity drop was evaluated for each core. Currently, the MCNAP code cannot do
both critical and fixed-source calculations for a burn-up point. Therefore, the reactivity change over a
180-day depletion was indirectly evaluated in terms of a spallation neutron multiplication factor (M s
). In
this paper, M s
is defined as 1 plus the number of fission neutrons produced by a spallation neutron in the
fuel blanket. In a critical reactor, the multiplication of a fission source neutron can be represented by
1/(1-k eff ). However, the spallation neutron multiplication, in sub-critical reactors, is quite different from
the multiplication of a fission source in a critical reactor. Readers who are interested in the multiplication
of a spallation neutron in ADS are referred to our work [7]. Although the source neutron multiplication
cannot exactly represent the reactivity, it can be generally said that the larger k-eff value, the larger M s
.
As far as the accelerator power is concerend, M s
has more practical meaning than k-eff for a sub-critical
reactor, since a large k eff does not always mean a high multiplication of the spallation neutron. Table 3
confirms this argument; k eff is almost 0.97 in all the three cores, even though the multiplication factor
varies fairly significantly. It should be noted that the proton beam current, required for a constant power
786

of the core, is directly determined by the M s
value, not by k eff .
Figure 3 shows the evolution of the spallation neutron multiplication during a 180-day burn-up
period for the reference core, HYPER-HBA, and HYPER-OBA. Figure 4 compares the proton beam
currents required for 1 000 MW th fission power. It is observed that multiplication of source neutrons at
BOC is quite different from each other, despite that the k eff values are almost the same. Specifically,
M s of HYPER-HBA is significantly smaller that that of the reference core, while HYPER-OBA has a
little larger M s than HYPER-HBA. The small multiplication factor of HYPER-HBA is due to the fact
that a significant fraction of the spallation neutrons, which are generated in the central target zone, is
absorbed by B 4 C before they give birth to their descendants. Meanwhile, the relatively high
multiplication factor of HYPER-OBA is because the inner zone is poisoned with burnable absorbers,
thus the probability for a source neutron to be parasitically absorbed is lower than in HYPER-HBA.
One can see a rapid decrease of the M s
values, in Figure 3, at the early period of depletion. This is
due to the fact that M s
is proportional to 1/(1-k eff ) and partly because fission products are accumulated in
the core. In Figure 3, it is clearly observed that HYPER-HBA has the smallest burn-up reactivity swing
among the three cases. However, if the cycle length is short, e.g. 120 days, this smaller reactivity swing
has little advantage since larger proton beam current is required, as shown in Figure 4. On the contrary,
for a 180-day operation, it is worthwhile to note that total accelerator power is almost comparable to the
reference case and the peak beam current is smaller than that of the unpoisoned core. This advantage is
attributed to the smaller reactivity swing of the HYPER-HBA core.
For HYPER-OBA, one can note that change in M s
is a little smaller than that of the reference due
to reduced reactivity swing. In addition, the M s
value of HYPER-OBA is a little larger except in the
vicinity of BOC, compared with the reference case. From Figure 4, it is clear that HYPER-OBA needs
smaller integrated accelerator power and also smaller peak beam current than the reference HYPER
core, if the depletion period is greater than 60 days. Figures 3 and 4 indicate that HYPER-OBA has
slightly larger reactivity swing than HYPER-HBA. This is mainly because the amount of B 4 C in
HYPER-OBA is about a half of that of HYPER-HBA. If thickness of the B 4 C layer is increased, the
reactivity swing of HYPER-OBA would be reduced further.
Figures 5 to 7 show the normalised radial power distributions at three burn-up points, 0-day,
120-day, and 180-day. In the reference core (see Figure 5), it is seen that slanting of the radial power
distribution is very significant; the inner zone power increased considerably during the burn-up
periods, while the outer zone power decreased. Especially, power density in the innermost fuel
assembly increased by a factor of 1.256 (120-day operation) or 1.394 (180-day operation). Currently,
the maximum allowable radial peaking factor is set to 1.50 for the HYPER core. Consequently, for a
relatively long cycle length, e.g. 180-day, the initial powers of the innermost fuel assemblies should be
much lower than the current values. Of course, the inner zone powers should also be lowered even for
a 120-day operation, since the peaking factor of the inner zone fuel assemblies might be fairly large.
This result confirms that radial power distribution control is a big concern in a sub-critical core with
large reactivity swing. In addition, too large change in the radial power distribution is not favourable
from the discharge burn-up distribution.
For HYPER-HBA in Figure 6, a similar behaviour can be observed as in Figure 5. On the other
hand, one can see quite different trend in the HYPER-OBA core. As shown in Figure 7, the HYPER-
OBA core has also a one-way change in the radial power distribution, i.e. monotonic increase in the
inner zone and decrease in the outer zone. However, the power increasing rate of the innermost
assembly is significantly suppressed, compared with the reference core. In HYPER-OBA, the power
of the innermost assembly is increased by a factor of 1.229 (120-day operation), or 1.300 (180-day
787

operation), respectively. This is because the 10 B burnable absorbers burn in the middle and outer
zones. This advantage of HYPER-OBA can be used for a longer fuel cycle. If the radial power
distributions are the same at BOC for HYPER-OBA and the reference core, the peak power of the
reference core would reach the criterion earlier than in HYPER-OBA. Based on the current results, it
is conjectured that the fuel cycle length of HYPER-OBA would be at least a month longer than that of
the reference design. Previously, we have seen that the initial source neutron multiplication of
HYPER-OBA is smaller that that of the reference. This is partly because of the lower power density of
the inner zone in HYPER-OBA, as shown in Figures 5 and 7. If the inner zone power were increased
in HYPER-OBA, the initial multiplication of source neutrons would also increase.
In Figure 8, the depletion behaviours of 10 B in HYPER-HBA and HYPER-OBA are given. In
HYPER-HBA, 10 B burns on the average at a rate of 2.22%/month and 10 B depletion rate of HYPER-
OBA is 2.11%/month. HYPER-OBA has a slower depletion rate of 10 B since the burnable absorbers
are loaded in relatively low-flux region, i.e. middle and outer zones. Assuming a 3-batch fuel
management in an equilibrium cycle, it is expected that only about 50% of 10 B would burn out in the
HYPER-OBA design. Therefore, the residual negative reactivity of 10 B is large in HYPER-OBA.
However, as stated previously, the advantages of the HYPER-OBA design such as smaller proton
beam current and longer cycle length would compensate for the negative impacts of 10 B.
Figure 3. Multiplication of spallation neutrons over a 180-day depletion in the HYPER cores
26
24
Neutron Multiplication (Ms)
22
20
18
16
14
12
10
8
Reference
HYPER-SBA
H YPE R- DB A
6
0 30 60 90 120 150 180
Time (day)
788

Figure 4. Required proton beam currents for the reference and poisoned cores
9
Proton Current (mA)
8
7
6
5
4
Reference
HYPER-SBA
HYPE R- DB A
3
2
0 30 60 90 120 1 50 180
Time (day)
Figure 5. Normalised radial power distributions of the reference HYPER core
ÃÃÃÃ'D\
'D\
'D\
789

790
Figure 6. Normalised radial power distributions in HYPER-HBA
ÃÃÃÃÃ'D\
'D\
'D\
Figure 7. Normalised radial power distributions in HYPER-OBA
ÃÃÃ'D\
'D\
'D\
%&
ORDGHG

Figure 8. 10 B depletion in HYPER-HBA and HYPER-OBA
24
HYPER-SBA
HYPE R- DB A
22
20
B-10 Mass (kg)
18
16
14
12
10
0 30 60 90 120 150 180
Time (day)
4. Conclusions
A homogeneous application of the B 4 C burnable absorber can be effectively used in reducing the
burn-up reactivity swing. However, it is not favourable in terms of source neutron multiplication, since
a significant fraction of spallation neutrons are parasitically absorbed by B 4 C before they are
multiplied. In sub-critical reactor, absorbing materials should not be placed in the neighbourhood of
the target zone.
Loading of 10 B burnable absorbers in the outer zones is required in order to minimise the parasitic
neutron absorption by 10 B. In this application of 10 B burnable absorber, the integrated and peak proton
beam powers are lower than those of the reference design. In addition, this kind use of 10 B can
considerably mitigate the slanting phenomenon of the radial power distribution, which is a critical
problem in TRU-loaded sub-critical reactors. Consequently, outer zone loading of B 4 C can lead to a
longer cycle length, compared with the unpoisoned reference core.
Finally, it is concluded that 10 B has a relatively high potential as a burnable absorbing material for
fast sub-critical reactors and introduction of a burnable absorber would open a new research field to
optimise the core design of ADS.
791

REFERENCES
[1] W.S. Part et al., HYPER (Hybrid Power Extraction Reactor): A System for Clean Nuclear
Energy, Nuclear Engineering and Design, 199, p. 155, 2000.
[2] A. Jonsson, Comparison of Economic Performance of Burnable Absorbers for 17 × 17 Fuel,
Proceedings of ANS Topical Meeting, March 1999.
[3] N. Stone et al., A Dual Spectrum Core for the ATW-Preliminary Feasibility Study, PHYSOR 2000.
[4] P. Hejzlar et al., Conceptual Reactor Physics Design of a Lead-bismuth-cooled Critical Actinide
Burner, MIT-ANP-TR-069, 2000.
[5] H.J. Shim et al., Monte Carlo Depletion Analysis of a PWR with the MCNAP, M&C 99 Int.
Conf. on Mathematics and Computation, reactor physics and Environmental Analysis in
Nuclear Applications, Sep. 1999, Madrid, Spain, Vol. 2, 1026-1035, 1999.
[6] W.S. Park et al., Fission Product Target Design for HYPER System, to be presented at the
6th Information Exchange Meeting on Actinide and Fission Product Partitioning and
Transmutation, Madrid, Spain, 11-13 Dec. 2000, OECD/NEA (Nuclear Energy Agency), Paris,
France, 2001.
[7] Y.H. Kim, W.S. Park, and T.Y. Song, Source Neutron Multiplication in Sub-critical Reactors, to be
presented at the IAEA Technical Committee Meeting on ADS to be held at Argonne, 2000.
792

POSTER SESSION
TRANSMUTATION SYSTEM
W.S. Park (KAERI)
793

MA AND LLFP TRANSMUTATION IN MTRs AND ADSs:
THE TYPICAL SCK•CEN CASE OF TRANSMUTATIONS
IN BR2 AND MYRRHA. POSITION WITH RESPECT TO THE GLOBAL NEEDS
Ch. De Raedt, B. Verboomen, Th. Aoust, E. Malambu, A. Beeckmans de West-Meerbeeck,
P. Kupschus, Ph. Benoit, H. Aït Abderrahim, L.H. Baetslé
SCK•CEN, Boeretang 200, 2400 – Mol, Belgium
E-mail: cdraedt@sckcen.be
Abstract
The proposed paper indicates the performances, in the domain of the transmutation of MAs and
LLFPs, of the high flux materials testing reactor BR2 located at SCK•CEN, and compares them with
those of the multipurpose ADS MYRRHA, the pre-design of which is at the present time being
finalized at SCK•CEN. With thermal neutron fluxes reaching 9 10 14 n/cm 2 s in thermal positions and
4 10 14 n/cm 2 s in the reactor core and, in the latter position, a fast flux (E>0.1MeV) of 7 10 14 n/cm 2 s,
BR2 has a transmutation throughput of the order of 1.5 kg Np+Am per 200 EFPD. This capacity can
be used for investigating at the technological scale the transmutation of MAs and LLFPs in a thermal
neutron spectrum with a high contribution of epithermal and fast neutrons. The metallurgical
behaviour of the targets can hence be studied. In MYRRHA, higher fast fluxes are expected to be
attained in irradiation positions near the spallation source, viz fast fluxes (E>0.75 MeV) up to
10 15 n/cm 2 s. One of the purposes of MYRRHA is therefore its utilisation for the investigation of
actinide transmutation feasibility with ADSs.
795

1. Introduction
In the framework of partitioning and transmutation (P&T), transmutation is the only technology
which is capable of accelerating the natural decay sequence, of influencing the decay schemes of
actinides and of reducing the radiotoxic inventory of some nuclides. For some radionuclides the
natural decay reactions take hundred thousands of years to reach the initial uranium ore toxicity level.
If partitioning of minor actinides (MAs) and long-lived fission products (LLFPs), such as e.g. 99 Tc
and 129 I, is successful, the way to transmutation is open. Transmutation in thermal or fast neutron
spectra has been thoroughly studied and both have their merits. Thermal neutrons are very effective
for fissile trans-uranium (TRU) nuclides ( 239 Pu, 241 Pu, 242 Am, 245 Cm) whereas fast neutron spectra
(in FRs and ADS) are indispensable for fissioning the fertile ( 237 Np, 241 Am, 243 Am) and even
mass-number nuclides ( 238 Pu, 240 Pu, 242 Pu, 244 Cm...).
The present paper makes an analysis of the possibilities of both approaches in relation with
technological demonstration experiments which could be performed in BR2, and later on in the
planned MYRRHA ADS-facility, to investigate the issues related to target optimisation, cladding
selection and structural material behaviour under intense irradiation.
2. BR2 and MYRRHA
2.1 The high flux materials testing reactor BR2
BR2 is a heterogeneous thermal high flux materials testing reactor [1-4]. Routine operation
started in 1963, and, to this very day, BR2 continues to contribute to the development of nuclear
projects within the European Community and for nuclear partners throughout the world. Figure 1 (left
hand side) shows a horizontal cross-section of the reactor core at the reactor midplane with a typical
loading.
Figure 1. Horizontal cross-section (left) of the BR2 reactor at the reactor midplane
with a typical loading and (right) of a type 6n-G fuel element with central aluminium plug
BR2 is cooled and moderated with pressurised light water (12 bar) in a compact core of highly
enriched uranium fuel elements, positioned in, and reflected by, a beryllium matrix. The ultimate
cooling capacity, initially foreseen for 50 MW, has been increased in 1971 to 125 MW. The reactor
nominal full power depends on the core configuration used; at the present time it ranges from 50 to
796

70 MW. The beryllium matrix has 79 cylindrical holes in a hexagonal lattice of 96.44 mm pitch at the
reactor midplane: there are 64 standard channels (84.2 mm diameter), 10 small peripheral channels
(50 mm diameter) and 5 large channels (200 mm diameter). All channels can receive fuel elements,
control rods, beryllium plugs or experiments, which allows a great flexibility of operation.
Each fuel element has a 762 mm active fuel length. The presently used 6n-G fuel elements
(Figure 1, right hand side) contain, when fresh, 400 g 235 U in the form of UAl x (1.3 g U/cm 3 ) + 3.8 g
boron (B 4 C) + 1.4 g samarium (Sm 2 O 3 ). The reactor core is loaded with 10 to 13 kg 235 U (30 to 40 fuel
elements, not all fresh). The concentration at discharge of the fuel elements is about 50% of the initial
fissile content value. The present nominal heat flux at the hot spot is 470 W/cm 2 , the maximum value
allowed for nominal cooling conditions (probable onset of nucleate boiling) is 600 W/cm 2 . Typical
neutron fluxes are (in the reactor hot spot plane):
• Thermal conventional neutron flux : v 0 n = v 0 ∫ 0
0.5 eV
n (E) dE:
2 to 4 10 14 n/cm 2 s in the reactor core
2 to 9 10 14 n/cm 2 s in the reflector and core flux trap (H1).
• Fast neutron flux : Φ >0.1 MeV = ∫ 0.1
∞MeV (G(
4 to 7 10 14 n/cm 2 s in the reactor core.
2.2 MYRRHA, a multipurpose ADS for R&D
MYRRHA, in its present development stage, is described in another paper of this conference [5].
It is based on the coupling of a proton cyclotron with a liquid Pb-Bi windowless spallation target,
surrounded by a sub-critical neutron multiplying medium in a pool type configuration [6]. Ion Beam
Applications (IBA), a world leader in accelerator technology, is in charge of the design of the
accelerator. The accelerator parameters presently considered are 5 mA continuous current at 350 MeV
energy. The proton beam will impinge on the spallation target from the top. The spallation target
circuit is separated from the core coolant as a result of the windowless design presently favoured to
avoid window overheating and embrittlement and loss of energy. To meet the goals of material
studies, fuel behaviour studies, radioisotope production, transmutation of MAs and LLFPs, the
MYRRHA facility should include two spectral zones: a fast neutron spectrum zone and a thermal
spectrum one.
The core pool contains the fast spectrum core zone, cooled with liquid Pb-Bi, and several islands
housing thermal spectrum regions located in in-pile sections (IPSs) at the periphery of the fast core. In
its present design phase, the fast core is fuelled with typical fast reactor fuel pins (triangular pitch:
10.2 mm) with an active length of 600 mm arranged in hexagonal assemblies with 122 mm pitch. The
central hexagon position is left free for housing the spallation module. The fast sub-critical core of
MYRRHA is made of 18 MOX fuel assemblies of which 12 have a Pu content of 30% and 6 a Pu
content of 20%: a horizontal cross-section is given in Figure 2 (the circumscribing circle has a
diameter of 610 mm).
The design of MYRRHA needs to satisfy a number of specifications such as:
• Achievement of the neutron flux levels required by the different applications considered in
MYRRHA:
Φ >0.75 MeV = 1.0 10 15 n / cm²s at the locations for MA transmutation,
797

Φ >1 MeV = 1.0 10 13 to 1.0 x 10 14 n / cm²s at the locations for structural material and fuel
irradiation,
Φ th = 2 to 3 10 15 n / cm²s at locations for LLFP transmutation or radioisotope production.
• Sub-critical core total power: ranging between 25 and 35 MW.
• Safety: k eff ≤ 0.95 in all conditions, as in a fuel storage, to guarantee inherent safety.
• Operation of the fuel under safe conditions: average fuel pin linear power

Table 1. Neutronic design parameters of the MYRRHA facility [9]
Zone Parameter Value
Spallation source
Fast sub-critical core
E p
I p
350 MeV
5 mA
k eff 0.948
k s 0.959
MF = 1/(1 - k s ) 24.51
thermal power
~32 MW
peak linear power
475 W/cm
max Φ > 1 MeV: around the spall. target 0.83 10 15 n/cm 2 s
first fuel ring
0.73 10 15 n/cm 2 s
max Φ > 0.75 MeV: around the spall. target 1.14 10 15 n/cm 2 s
first fuel ring
1.03 10 15 n/cm 2 s
2.3 Typical spectra in BR2 and MYRRHA
Typical neutron flux spectra in BR2 and in MYRRHA are shown (with the same scales) in
Figure 3.
Figure 3. Typical spectra in BR2 and MYRRHA (all normalised to 1.00)
1.E+00
1.E+00
1.E-01
1.E-01
1.E-02
1.E-02
flux per unit lethargy
1.E-03
1.E-04
BR2 : axis of a 6n-G fuel element
flux per unit lethargy
1.E-03
1.E-04
1.E-05
1.E-05
BR2 : reflector channel
1.E-06
1.E-06
MYRRHA : R = 5.75 cm
MYRRHA : R = 33.4 cm
1.E-07
1.E-07
1.E-03
1.E-02
1.E-01
1.E+00
1.E+01
1.E+02
1.E+03
1.E+04
1.E+05
1.E+06
1.E+07
1.E+08
1.E-03
1.E-02
1.E-01
1.E+00
1.E+01
1.E+02
1.E+03
1.E+04
1.E+05
1.E+06
1.E+07
1.E+08
energy (eV)
energy (eV)
799

3. Irradiation targets and irradiation positions in BR2 and MYRRHA
All irradiation targets considered (for the irradiations both in BR2 and in MYRRHA) were
assumed to have an outer diameter of 8.36 mm and to be clad with HT-9 steel with outer diameter
9.5 mm. The target compositions are indicated in Table 2. The targets were given an active length of
200 mm, positioned symmetrically with respect to the maximum axial flux level of BR2 and
MYRRHA.
Table 2. Composition of the various targets
MA or LLFP Chemical form of the target Density
(%TD)
237 Np
77.9 wt% 241 Am
+22.1 wt% 243 Am
99 Tc
129 I
20 vol% NpO 2 + 40 vol% MgAl 2 O 4 + 40 vol% Al
20 vol% Am 2 O 3 + 40 vol% MgAl 2 O 4 + 40 vol% Al
Tc-metal
NaI
90
90
75
70
The BR2 irradiation targets were assumed to be introduced into a loop consisting of concentric
aluminium tubes with in between cooling water circulation. The outer diameter of the outer tube was
25.4 mm, allowing the loop to be substituted for the central aluminium plug (also with diameter
25.4 mm) of a standard BR2 fuel element such as shown in Figure 1, right hand side. As irradiation
position in BR2 the high-flux channel B180 (see Figure 1, left hand side) was selected.
For the irradiations in MYRRHA, the targets with their cladding were assumed to be introduced
into the irradiation space between the pressure tube surrounding the spallation source and the six
inner hexagonal assemblies of the fast sub-critical core (see Figure 2), at a radial distance of 59 mm
from the MYRRHA main axis.
4. Calculated transmutation yields
The transmutation rates of the MAs and of the LLFPs, both in BR2 and in MYRRHA, were
calculated with the aid of the Monte Carlo code MCNP-4B [10]. The way BR2 was modelled in great
detail is explained in [11,12,13] (e.g. the meat and the cladding of each of the 6 × 3 fuel plates of each
of the fuel elements of the BR2 loading is considered as a separate zone). Also for MYRRHA the
calculations were performed in great detail: each of the 2286 fuel pins (with for each: fuel, gap and
cladding described separately) was modelled. The neutron cross-section library used was ENDF/B-VI,
except for the nuclides not present in ENDF/B-VI; for these ENDF/B-V was used.
In Table 3 the fission and “disappearance” reaction rates are indicated for the various nuclides
considered in the present study. The values are averaged over the target volumes. By “disappearance”
is meant the sum of all nuclear reactions leading to the removal of the nuclide considered from its
(A,Z) position in the table of isotopes. The “disappearance” reaction is hence, practically, the sum of
the processes 16 (= all (n,2n)), 17 (= (n,3n)), 18 (= total fission) and 101 (= neutron disappearance,
viz mainly (Q n,p), (n,d), (n,t), (n,He-3) and (Q ZKHUHWKHQXPEHUVUHIHUWRWKH07QXPEHUVLQ
the ENDF/B format. It should be noted that in the present calculations performed for MYRRHA the
very high energy part of the spallation neutron spectrum (above 20 MeV) is not taken into account.
800

Table 3. Target-volume-averaged direct fission and
disappearance reaction rates in the various targets irradiated in BR2 and MYRRHA
Target Nuclide Reaction rates in BR2 Reaction rates in MYRRHA
NpO 2 -MgAl 2 O 4 -Al
Am 2 O 3 -MgAl 2 O 4 -Al
Tc-metal
NaI
237 Np
241 Am
243 Am
99 Tc
129 I
direct fiss.
(s -1 )
disapp.
(s -1 )
direct fiss.
/disapp.
direct fiss.
(s -1 )
disapp.
(s -1 )
direct fiss.
/disapp.
5.50 10 -10
9.09 10 -10
4.75 10 -10 3.99 10 -8
7.05 10 -8
2.87 10 -8 0.014
0.013
0.017
1.29 10 -9
1.09 10 -9
8.67 10 -10 3.88 10 -9
3.89 10 -9
3.27 10 -9 0.33
0.28
0.27
3.93 10 -9
8.95 10 -10
5.86 10 -9 4.90 10 -10
Table 4 indicates the resulting percent MA and LLFP disappearance, by fissions and total, after
an irradiation period of 200 effective full power days (EFPD). The isotopes 241 Am and 243 Am are
considered separately. The fissions indicated are the “direct” fissions of 237 Np, 241 Am and 243 Am and
do not include the “secondary” fissions, viz those of the actinides formed by (Q UHDFWLRQVGXULQJWKH
irradiation, possibly followed by natural decay. In the case of irradiations in BR2, these “secondary”
fissions are very important: see further, Table 6.
Table 4. Percentage disappearance (through direct fissions and total)
of the MAs and the LLFPs irradiated in BR2 and MYRRHA during 200 EFPD
Target Nuclide Disappearance in BR2 Disappearance in MYRRHA
NpO 2 -MgAl 2 O 4 -Al
237 Np
% direct
fissions
0.69
%
total
49.79
direct
fiss./total
0.014
% direct
fissions
2.16
%
total
6.49
direct
fiss./total
0.33
Am 2 O 3 -MgAl 2 O 4 -Al
Tc-metal
NaI
241 Am
243 Am
99 Tc
129 I
0.91
0.65
70.42
39.10
8.99
9.63
0.013
0.017
1.82
1.46
6.50
5.49
1.53
0.84
0.28
0.27
5. Discussion
5.1 Comparison with data published previously for BR2 and MYRRHA
In [14], presented in 1992 at the Second Information Exchange Meeting on Actinide and Fission
Product Separation and Transmutation, the MA and LLFP transmutation rates in BR2 were calculated
for a large variety of targets as to their dimensions and material concentrations. Compared to the
calculations discussed in the present paper, entirely different codes and different cross-section
libraries were used. Nevertheless, for the 1992 target dimensions and MA and LLFP concentrations
corresponding to those studied in the present paper, the comparison of the calculated transmutation
rates indicates, for the MAs, an excellent agreement (within 5%), and for 99 Tc and 129 I, an agreement
within about 10%. The deductions made in [14] as to the total amounts of MAs and LLFPs that could
801

e transmuted in BR2 hence remain valid: the transmutation capacity of BR2, if dedicated to R&D
programmes on P&T, is about 1.5 kg Am or Np per 200 EFPD.
The comparison of the transmutation rates in MYRRHA calculated in the present paper with those
mentioned in [15], presented at Global’99, indicates a good agreement if one takes into account the fact that
in the present paper the total neutron flux in the targets amounts to 2.48 ... 2.56 10 15 n/cm 2 s, while in [15],
total fluxes of 1.0 10 15 n/cm 2 s were considered.
5.2 Transmutation of LLFPs
The transmutations of 99 Tc and 129 I considered in this study were assumed to occur in the hard
spectrum zone of BR2 and in the fast sub-critical core part of MYRRHA. The transmutation yields are
small. Irradiations in the thermal neutron spectrum zones of both devices would lead to higher yields.
The transmutation of the LLFPs is not further considered in the present paper.
5.3 Comparison of the performances of BR2 and MYRRHA
One observes – see Table 4 – that the total disappearance of MAs, as defined above, is much
larger when the targets are irradiated in BR2 than when they are irradiated in MYRRHA. This is
mainly due to the fact that the “disappearance” cross-sections are much larger in the thermal and
epithermal energy regions than in the fast energy region. As known, see Table 5, the (Q UHDFWLRQVRQ
the MAs considered in the present study lead to the formation of other actinides, mainly (for 237 Np
and for Am) to 238 Pu-and the further Pu family, and (for Am) to 244 Cm and the further Cm family.
Table 5. Main transmutation reactions occurring in the targets considered
237 Np (Q
(n,f)
241 Am (Q
§
238 Np
FP
242m Am
141 y
2.1 d 238 Pu (Q Pu-family
(n,f) FP
(Q
243 Am
(Q
§
(n,f)
242g Am
FP
16 h
242 Cm 163 d 238 Pu (n,f) Pu-family
(Q )3
243 Am (Q
244 Am
16 m...10 h
244 Cm (Q Cm-family
(n,f)
FP
99 Tc (Q
100 Tc
16 s
100 Ru
129 I (Q
130 I
12 h
130 Xe
802

Only the fission process allows complete removal of the MAs out of the actinide family. For this
process, ADS-type reactors such as MYRRHA, with a fast neutron spectrum, seem, at first sight, the
most attractive. Table 4 indicates that about 30% of the disappearance of MAs in MYRRHA are due
to “direct” fissions, while for BR2 the figure is less than 2%. Nevertheless, if one includes the
“secondary” fissions, viz, as mentioned in the previous section, those occurring in the fissile actinides
formed by (Q UHDFWLRQVGXULQJWKHLUUDGLDWLRQSRVVLEO\IROORZHGE\GHFD\RQHREWDLQVFRPSOHWHO\
different figures. This can be seen in Table 6, where the concentrations (in atom%) of the main
nuclides present in the 237 Np and the 241+243 Am targets are indicated, as calculated for irradiations of
200, 400 and 800 EFPD in BR2 and MYRRHA, followed by a cooling period of 5 years. “Fissium”
represents the nuclides that have disappeared from the actinide family and is hence equal to half the
total number of FP atoms. One observes the important percentage of fissions in BR2,
“direct” + (mainly) “indirect”. This contribution increases strongly with irradiation time, due to the
quadratic (or higher) order of the build-up curve of nuclides in long formation chains (the values
given in Table 6 are only approximate as both the neutron flux levels and the microscopic
cross-sections were assumed to remain constant during the irradiations. Control calculations
nevertheless show that the trends indicated remain valid, in particular that the percentage fissions
remains important in BR2). The important fraction of secondary fissions contributes to the high
depletion rate observed in a thermal neutron spectrum. In a fast spectrum however, the generation of
toxic long-lived actinides is practically non-existent for the irradiation of 237 Np.
From the neutron economics point of view, which is an essential topic when applying
transmutation on an industrial scale, ADSs (and FRs) have the undeniable advantage over thermal
reactors of needing shorter chains (i.e. less neutrons) to achieve fission. Larger quantities of fissile
material will be needed in thermal critical systems to compensate the reactivity decrease (while in a
thermal ADS this could be obtained by increasing the proton current). Concerning fast neutron
systems, and in particular ADS devices such as MYRRHA, the irradiation period needed to fully
deplete a MA target is much longer for the neutron flux level considered in this paper. Higher flux
levels would obviously shorten the irradiation time needed to achieve transmutation, probably with a
much lower fissile material inventory. A quantitative analysis of these issues would be worth while.
803

Table 6. Atom percent concentration of the various nuclides in MA targets irradiated
in BR2 and MYRRHA during 200, 400 and 800 EFPD, followed by 5 years cooling
Target
0 EFPD
BR2
MYRRHA
200 EFPD 400 EFPD 800 EFPD 200 EFPD 400 EFPD 800 EFPD
237 Np
Am
234 U
237 Np
238 Pu
239 Pu
240 Pu
241 Pu
242 Pu
241 Am
243 Am
244 Cm
Fissium
Sum
234 U
237 Np
238 Pu
239 Pu
240 Pu
241 Pu
242 Pu
241 Am
242m Am
243 Am
243 Cm
244 Cm
245 Cm
246 Cm
Fissium
Sum
0
100.0
0
0
0
0
0
0
0
0
0
100.0
0
0
0
0
0
0
0
78.08
0
21.92
0
0
0
0
0
100.0
1.0
50.2
25.4
6.4
1.6
1.1
0.4
0.6
0.1
~ 0
13.2
100.0
1.5
0.2
40.9
2.3
1.6
0.2
~ 0
22.9
0.2
14.0
0.9
6.2
0.5
0.1
8.3
99.8
0.8
25.2
20.3
6.1
1.9
1.8
1.6
0.6
0.7
0.3
40.5
99.8
1.6
0.1
39.3
5.7
2.7
0.9
0.3
7.0
0.1
8.8
1.0
8.7
0.8
0.3
22.4
99.7
0.3
6.4
6.9
2.2
1.1
0.8
1.6
0.2
1.4
1.6
75.9
98.4
0.9
~ 0
17.2
4.8
2.6
1.4
1.3
1.0
~ 0
3.3
0.3
8.4
0.8
0.9
55.4
98.3
0.1
93.5
4.0
~ 0
~ 0
~ 0
~ 0
0.1
~ 0
~ 0
2.3
100.0
0.1
0.7
3.1
~ 0
0.2
~ 0
~ 0
72.4
0.3
20.7
~ 0
0.7
~ 0
~ 0
1.8
100.0
0.3
87.4
7.4
~ 0
~ 0
~ 0
~ 0
0.1
~ 0
~ 0
4.8
100.0
0.2
0.7
5.8
0.1
0.3
~ 0
~ 0
67.6
0.6
19.6
~ 0
1.3
~ 0
~ 0
3.7
99.9
0.5
76.5
12.8
~ 0
~ 0
~ 0
~ 0
0.1
~ 0
~ 0
10.0
99.9
0.4
0.7
10.2
0.3
0.6
~ 0
~ 0
59.0
1.0
17.5
0.1
2.3
0.2
~ 0
7.7
100.0
When comparing the performances of MYRRHA and BR2, one should also take into account the
operation regime. Currently, BR2 only operates about 105 days per year while MYRRHA is to
operate about 9 months per year: the MYRRHA utilisation factor would hence be 2.6 times that of
BR2.
5.4 Comparison of the performances of Accelerator Driven Systems and Fast Reactors
In Reference [15], the performances of FRs and ADS systems were compared: the conclusion
was that the neutron-flux-averaged cross-sections governing the transmutation of MAs and LLFPs do
not lead to very important differences in the performances of ADS devices compared to FRs. The
absolute neutron flux levels, on the other hand, which are proper to each individual device, do
strongly influence the transmutation capacity. In the case of fast reactors the neutron flux levels can
only vary within certain limits, while in the case of ADS the neutron flux levels are directly
proportional to the proton beam current delivered by the accelerator and to the multiplication factor of
the subcritical system, and also depend on the energy of the proton beam (the higher the energy, the
higher the neutron/proton ratio in the spallation reaction). From the core reactivity control point of
view, the ADS devices present undeniable advantages in the case of variable core loadings with large
804

amounts of MAs. In addition, one should keep in mind that, as mentioned above, the very high energy
part of the spallation neutron spectrum was not taken into account, neither in Reference [15] nor in
the present study.
6. Required transmutation capacity and associated fuel cycle facilities [16]
A 100 GWe nuclear LWR UO 2 reactor park produces annually 2 200 t spent fuel. This inventory
contains approximately 25 t Pu, 1.6 t Np, and 1.6 t Am+Cm. If nuclear electricity production
continues during an indefinite time period, the spent fuel inventory will continuously grow and
become a large nuclear legacy which will have to be properly managed during a very long period,
exceeding human civilisation. To decrease the growth rate of this spent fuel inventory, particularly its
incorporated Pu mass, reprocessing of LWR-UO 2 is a first step to reduce the accumulation rate.
However the HLW resulting from reprocessing still contains the 3.2 t of MAs. These nuclides
constitute the long-term radiotoxic inventory of the high level waste (HLW) if no partitioning is
performed.
Partitioning of the MAs, followed by transmutation-incineration, could achieve a 10 to 100-fold
reduction of the residual radiotoxicity. The recycling of separated MAs can be envisaged by mixing
Np with the LWR-MOX fuel or, more effectively, by mixing all the MAs with FR-MOX fuel. A fuel
fabrication capacity of 60 t FR-MOX-(2.5%Np) and 60 t FR-MOX-(2.5%Am) would have to be
installed near the reprocessing plants. A composite reactor park containing 70 GWe-LWR-UO 2 ,
10 GWe-LWR-MOX and 20 GWe-FR+ADS is capable of stabilising the TRU inventory. Reduction
of the MA fuel mass to be handled in the FR-ADS systems is still possible if higher concentrations of
MAs can be introduced into the sub-critical cores of ADS systems. An 820 MW th ADS core (1.5 GeV,
40 mA) containing 60% MA and 40% Pu could transmute 250 kg TRU per year. With a thermal to
electric yield of 30% it would produce 246 MWe which would in part be recycled to the accelerator
(146 MWe) and in part delivered to the grid. Gradually the ADS capacity should increase from
8 GWe initially to 20 GWe at the end of the nuclear energy production to cope with the entire residual
TRU inventory.
7. Conclusions
With a routine thermal output of 60 MW th , BR2 has a limited irradiation potential for 12 targets
of 500 g Np+Am each and a transmutation throughput of the order of 1.5 kg Np+Am per 200 EFPD.
This transmutation capacity can be used for investigating, at the technological scale, the formation of
transmutation products ( 238 Pu, 239 Pu, FPs...) in a thermal neutron spectrum with large contribution of
epithermal and fast neutrons as well as the metallurgical behaviour of the targets. In particular, if the
irradiations are carried out during a long period, the calculated high fission-over-total-disappearance
rate in the 237 Np-and Am targets could be checked. It is indeed essential to take into account the total
length of the transmutation chains when performing calculations for high flux reactors.
One of the purposes of MYRRHA is its utilisation for the investigation of actinide transmutation
feasibility with ADSs. With a total power not to exceed 30 to 35 MW, fast fluxes (E>0.75 MeV) up to
10 15 n/cm 2 s are to be attained in irradiation positions near the spallation source. Calculations indicate
lower transmutation rates in MYRRHA than in BR2, but fast spectrum systems, and in particular ADS
devices, are characterised by better neutron economics in the transmutation process, i.e. by a higher
“direct” fission-over-total-disappearance rate. MYRRHA as multipurpose ADS for R&D is hence an
interesting tool to investigate transmutation of MAs in a fast neutron environment, with the
805

advantage, with respect to FRs, of a larger versatility and an improved core reactivity control. In
addition, MYRRHA is expected to operate with a high utilisation factor.
Compared to what BR2 and MYRRHA can offer from the materials investigation point of view,
the quantities generated in nuclear fuel from a 100 GWe park that could potentially be separated in an
advanced reprocessing plant (3.2 tonnes Np+Am per year) would of course require very large
irradiation facilities (from 8 to 20 GWe) to achieve a quantitative transmutation yield.
REFERENCES
[1] Brochure BR2, Multipurpose Materials Testing Reactor. Reactor Performance and Irradiation
Experience, SCK•CEN, November 1992.
[2] J.M. Baugnet, Ch. De Raedt, P. Gubel, E. Koonen, The BR2 Materials Testing Reactor. Past,
Ongoing and Under-study Up-gradings, First Meeting of the International Group on Research
Reactors, Knoxville, Tennessee, USA, Febr. 28-March 2, 1990.
[3] Ch. De Raedt, H. Aït Abderrahim, A. Beeckmans de West-Meerbeeck, A. Fabry, E. Koonen,
L. Sannen, P. Vanmechelen, S. Van Winckel, M. Verwerft, Neutron Dosimetry of the BR2
Aluminium Vessel, Ninth International Symposium on Reactor Dosimetry, Prague,
Sept. 2-6, 1996.
[4] E. Malambu, Ch. De Raedt, M. Wéber, Assessment of the Linear Power Level in Fuel Rods
Irradiated in the CALLISTO Loop in the High Flux Materials Testing Reactor BR2, Third
International Topical Meeting “Research Reactor Fuel Management (RRFM)”, Bruges,
March 28-30, 1999.
[5] H. Aït Abderrahim, P. Kupschus, E. Malambu, K. Van Tichelen, Ph. Benoit, B. Arien,
F. Vermeersch, S. Bodart, Th. Aoust, Ch. De Raedt, MYRRHA, a Multipurpose ADS for R&D
as First Step towards Waste Transmutation, 6th Information Exchange Meeting on Actinide
and Fission Product Separation and Transmutation, Madrid, Spain, Dec. 11-13, 2000,
EUR 19783 EN, OECD/NEA (Nuclear Energy Agency), Paris, France, 2001.
[6] H. Aït Abderrahim, P. Kupschus, E. Malambu, Ph. Benoit, K. Van Tichelen, B. Arien,
F. Vermeersch, P. D'hondt, Y. Jongen, S. Ternier, D. Vandeplassche, MYRRHA: a Multipurpose
Accelerator Driven System For Research and Development, in Special Issue on Accelerator
Driven Systems, H.S. Plendl (Ed.), Nucl. Instr. and Meth. A, in press.
[7] P. Cloth, D. Filges, R.D. Neef, G. Sterzenbach, HERMES, a Monte Carlo Program System For
Beam-materials Interaction Studies, User's Guide, Jül-2203 (1988).
[8] W.A. Rhoades, R.L. Childs, The DORT Two-dimensional Discrete Ordinates Transport Code,
Nuclear Science & Engineering, 99, 1, pp.88-89, May 1998.
806

[9] E. Malambu, Progress Report on Neutronic Assessment of the MYRRHA ADS Facility, Internal
Report SCK•CEN R-3474, Oct. 26, 2000.
[10] J.F. Briesmeister, Ed., MCNP, a General Monte Carlo N-Particle Transport Code, LA-12625-M,
Version 4B, March 1997.
[11] B. Verboomen, Th. Aoust, A. Beeckmans de West-Meerbeeck, Ch. De Raedt, Irradiation of
New MTR Fuel Plates in BR2, Fourth International Topical Meeting “Research Reactor Fuel
Management (RRFM)”, Colmar, March 19-21, 2000.
[12] Ch. De Raedt, E. Malambu, B. Verboomen, Increasing Complexity in the Modelling of BR2
Irradiations, PHYSOR 2000 International Topical Meeting “Advances in Reactor Physics and
Mathematics and Computation into the Next Millenium”, Pittsburg, USA; PA, May 7-11, 2000.
[13] B. Verboomen, A. Beeckmans de West-Meerbeeck, Th. Aoust, Ch. De Raedt, Monte Carlo
Modelling of the Belgian Materials Testing Reactor BR2: Present Status, Monte Carlo 2000,
International Conference on Advanced Monte Carlo for Radiation Physics, Particle Transport
Simulation and Applications, Lisbon, Portugal, Oct. 23-26, 2000.
[14] L.H. Baetslé, Ch. De Raedt, A. Delbrassine, A. Beeckmans de West-Meerbeeck, Transmutation
Capability of the High Flux Reactor BR2, Second Information Exchange Meeting on “Actinide
and Fission Product Separation and Transmutation”, ANL, Chicago, USA, Nov. 11-13, 1992.
[15] Ch. De Raedt, L.H. Baetslé, E. Malambu, H. Aït Abderrahim, Comparative Calculation of
FR-MOX and ADS-MOX Irradiations, International Conference on Future Nuclear Systems,
Global’99, Jackson Hole (Wyoming), Aug. 30-Sept. 2, 1999.
[16] OECD Nuclear Energy Agency, Actinide and Fission Product Partitioning and Transmutation.
Status and Assessment Report, 1999, Paris, France.
807

ENHANCEMENT OF ACTINIDE INCINERATION AND TRANSMUTATION
RATES IN ADS EAP-80 REACTOR CORE WITH MOX PUO 2 &UO 2 FUEL
S. Kaltcheva-Kouzminova, V. Kuzminov
Petersburg Nuclear Physics Institute,
Gatchina, St.Petersburg, Zip 188300, Russian Federation
M. Vecchi
ENEA,
Via Martiri di Monte Sole 4, 40129 Bologna, Italy
Abstract
Neutronics calculations of the accelerator driven reactor core EAP-80 with UO 2 &PuO 2 MOX fuel
elements and Pb-Bi coolant are presented in this paper. Monte Carlo optimisation computations of
several schemes of the EAP-80 core with different types of fuel assemblies containing burnable absorber
B 4 C or H 2 Zr zirconium hydride moderator were performed with the purpose to enhance the plutonium
and actinide incineration rate. In the first scheme the reactor core contains burnable absorber B 4 C
arranged in the cladding of fuel elements with high enrichment of plutonium (up to 45%). In the second
scheme H 2
Zr zirconium hydride moderated zones were located in fuel elements with low enrichment
(∼20%). In both schemes the incineration rate of plutonium is about two times higher than in the
reference EAP-80 core and at the same time the power density distribution remains significantly
unchanged compared to the reference core. A hybrid core containing two fuel zones one of which is the
inner fuel region with UO 2 &PuO 2 high enrichment plutonium fuel and the second one is the outer region
with fuel elements containing zyrconium hydride layer was also considered. Evolution of neutronics
parameters and actinide transmutation rates during the fuel burn-up is presented. Calculations were
performed using the MCNP-4B code and the SCALE 4.3 computational system.
809

1. Introduction
In the last years there is an increased interest to the problem of burning-up of plutonium and
minor actinides generated from waste and dismantled weapons. A project of an accelerator driven
sub-critical reactor prototype EAP-80 is considered in Italy. In the ANSALDO technical reports a
detailed description of the reference core configuration for the accelerator driven sub-critical
demonstration facility is given [1]. Preliminary neutronics analysis of an accelerator driven
sub-critical energy amplifier prototype (EAP-80) was presented in the Atzeni report [2]. The main
purpose of project [2] was to demonstrate the soundness of the basic principles of the energy
amplifier [3] with U-Pu MOX fuel which is similar to the fuel used in the super Phenix fast reactor. A
lead-bismuth eutectic is used as a neutron production target and also as a coolant. Neutron producing
target located in the centre of the reactor core is irradiated by 600 MeV proton beam. Due to the low
value of neutron absorption cross-sections on lead-bismuth eutectic a low gradient of neutron flux
may be obtained in the reactor core that leads to a small value of the power peaking factor.
Optimisation calculations of neutronics parameters of the reference core of the EAP-80 were
performed in ANSALDO and ENEA. The fuel composition used in the reference reactor core is
presented in Table 1.
Table 1. Fuel composition (10 24 cm -3 )
Atomic density Atomic density Atomic density
234
U 7.673302 × 10 -7 237 Np 1.056298 × 10 -6 241 Pu 1.390316 × 10 -4
235
U 9.345065 × 10 -5 238 Pu 1.468511 × 10 -5 242 Pu 6.823877 × 10 -5
236U 8.919691 × 10 -7 239 Pu 3.715382 × 10 -3 241 Am 1.235840 × 10 -4
238
U 1.774578 × 10 -2 240 Pu 1.307362 × 10 -3 16 O 4.642046 × 10 -2
The reference core contains fuel assemblies with plutonium enrichment of 23.2%. The arrangement
of fuel assemblies in the reference reactor core is shown in Figure 1.
2. Calculation method
The effective multiplication factor, power density distribution and the reaction rates for the reactor
core were computed by the MCNP-4b [4] code with DLC-181, DLC-189 (ENDF/B-6) cross-section
libraries at the temperature of 300 K. Calculations of fuel burn-up were performed using the ORIGENS
module and the SCALE 4.3 [5] computational system. Problem dependent cross-sections for the
burn-up computations for the considered reactor core were prepared using the NITAWL and
XSDRNPM modules in the SCALE system. The effective multiplication factor k eff for the reference
core at a temperature of T = 300K is equal to 0.9839 ± 0.0005. In the present computations the
temperature of the reactor core was taken equal to 300 K in order to compare the results of
optimisation calculations with the parameters of the reference core. The fuel in the reference core has
a high concentration of 238 U that leads to plutonium breeding reaction competitive with the burning-up
reaction on plutonium. We will define a plutonium burn-up rate as a sum of fission and capture rates
on 239 Pu and a subtraction of breeding rate of 239 Pu in the reaction (n,γ) of neutron capture on 238 U:
%
3X
= P ×
∫
9
∫
0H9
[ Σ (() + Σ (() − Σ (()] Φ( U
() G( G9
I
&
&
810

Here Σ fC , ( E)
are the fission and capture cross-sections, Φ ( U ()
is the neutron flux, and B Pu
is a
plutonium burning rate in units of kg per full power year (kg/FPY), and m is a normalisation
parameter used to take into account the reactor power. Supposing that a spatial distribution of fission
events in the reactor core with fission neutron source is the same as in the core with a spallation
neutron source (i.e. in the core with k eff
near criticality), we can estimate the B Pu
from the criticality
calculations. Calculations with a spallation neutron source located in the centre of the reactor core are
also presented. The spectrum and the mean energy E s
of evaporated neutrons in spallation reaction
was calculated by us using the medium energy code SITHA [6]. For the considered target it is equal
to E s
= 3.3 MeV. The total factor of neutron multiplication in the core with the spallation neutron
source was calculated using the formula
i i
∫∫ Φ(
r,
E)
∑ ρ
iν
σ ( E)
dE dV + ∫∫Φ
fis fis
i i
M = +
( r,
E)
ρ x σ ( E)
dE dV
tot
1
i n,
xn
i
i
For the reference core M tot
=50.9, the neutron source multiplication factor is equal to
k s
= (M tot
- 1)/M tot
= 0.98. The current of proton beam I p
= 1.67 mA is obtained for the reactor power
P r
= 80 MW th
. Plutonium burn-up rate in the reference core is equal to B Pu
= 13.4 kg/FPY and
B Pu
= 13.7 kg/FPY for fission and external neutron sources respectively at the reactor power
P r
= 80 MW. A spatial distribution of power density in the fuel elements was calculated using the
space distribution of fission events in fuel assemblies (heating by photons in the present calculations
was not taken in account due to the small values compared with the fission reactions). Power peaking
factor, k v
= k r
k z
, in the reference core with fission source is equal to 1.37 in the assembly near the
spallation neutron target. It should be noted that the spatial distribution of heating energy in the reactor
core calculated using the fission neutron source differs from the energy distribution in the core
calculated with the spallation neutron source. The radial peaking factor k r
is increased to k r
= 1.27
compared to the value k r
= 1.20 obtained for the fission neutron source. This difference depends on the
sub-critical level of the core and for the core with lower value of k eff
this difference will be greater.
Above mentioned phenomenon plays an important role in the dependence of the reactor power
distribution and power peaking factors versus the changing of the k eff
parameter during the reactor fuel
cycle.
In order to enhance the plutonium incineration rate we propose to shift the neutron spectrum in
the fuel into the energy region where the ratio of fission cross section on 239 Pu to capture cross-section
on 238 U is greater than in the reference core. At the same time the effective multiplication factor k eff
should be equal to or less than the value 0.984. Below we consider several ways of how to enhance
the burning rate of plutonium in the reactor core.
∑
3. Core with burnable absorber B 4
C in fuel elements
Application of thin axial burnable absorbers B 4
C in the gap between the fuel pellet and the stainless
steel cladding allows to increase the plutonium enrichment and to obtain hard neutron spectrum in the
core. Due to the neutron absorption on 10 B in the resonance energy region the rate of capture reaction on
238
U reduces and the ratio of fission to capture cross-sections for transuranium nuclides increases. The
fuel pellet with plutonium enrichment 43.7%, has a central hole r fuel
= 0.100 cm and the outer radius of
fuel pellet is equal to R fuel
= 0.347 cm. The thickness of 10 B 4
C layer is equal to 0.013 cm. The average
power density in fuel pellets in this scheme at the beginning of fuel cycle (BOC) is equal to q f
= 24 W/g
and is comparable with the reference value of power density (q f
= 22 W/g in the reference fuel pellets).
The plutonium 239 Pu burning rate is equal to B Pu
= 25 kg/FPY, i.e. about two times higher than the
plutonium burn-up in the reference core (B Pu
= 13.7 kg/FPY).
811

4. Core with zirconium hydride in fuel elements
For enhancement of plutonium burn-up rate a moderated zone containing zirconium hydride in
the fuel element may be used. Zirconium hydride is a good moderator and has good inert properties.
Application of zirconium hydride moderated zones in fast reactors were considered in papers [8-9].
We have considered the reactor core with fuel elements containing zirconium hydride layer
located in the gap between the fuel pellet and steel cladding or around the steel cladding. All
dimensions of the fuel pin were taken as in the reference core. Due to the shifting of the neutron
spectrum to the thermal energy region more than 70% of the plutonium burn-up events occurs in the
thermal energy region. The use of the fuel elements with zirconium hydride and plutonium
enrichment of 20%-23% permits to burn up about B Pu
= 25 kg/FPY (at the P r = 80 MW) of plutonium
at the average power density in the fuel of q f
= 227 W/cc. There are two possible arrangements of
zirconium hydride layer in fuel element: (1) HZr arranged abutting to the outer surface of the steel
cladding, and (2) HZr layer may be arranged in the gap between the fuel pellet and the steel cladding
of the fuel element.
In Table 2 results of computations of the reactor core with different arrangement of HZr layer in fuel
element are presented. Burn-up rate of plutonium in fuel elements with HZr layer
(∆d 2
(HZr) ≈ 0.045 ÷ 0.072 cm) arranged abutting to the outer surface of fuel pin is equal to
B Pu
= 24.8 kg/FPY, while B Pu
= 21.6 kg/FPY in the core with HZr layer (∆d 1
(HZr) ≈ 0.035 cm) arranged
inside the fuel element. The average power density is equal to q f
= 228 W/cc (21.8 W/g) in fuel elements
with HZr layer arranged abutting to the outer diameter of fuel cladding, and q f
= 258W/cc (24.7 W/g) in
fuel elements containing HZr layer inside fuel element.
Table 2. Fuel elements containing H 2
Zr cladding arranged in different positions:
(1) Abutting to the outer diameter of the steel cladding.
(2) In the gap between the fuel pellet and the steel cladding.
Parameter 23.225%
Reference core
Plutonium enrichment, %
20.0% 23.2% 55%
Position of HZr – (1) (2) (1) (2)
r fuel
, cm 0.09 0.09 0.08 0.09
R fuel
, cm 0.357 0.357 0.33 0.301
r clad
, cm 0.3685 0.3685 0.3685 0.3685
R clad
, cm 0.425 0.425 0.425 0.425
k eff
0.9839(5) 0.9831(5) 0.9811(6) 0.9828(6)
k s
0.980 0.982 0.981 0.984
I p
(mA) 1.67 1.51 1.61 1.41
q f
(W/g) 21.98 21.82 24.73 31.00
k v
= q max
/q f
1.35 1.37 1.32 –
B Pu
, (kg/FPY) 13.7 24.8 21.6 33.2
Fuel element with plutonium enrichment 55% and two moderated zones of zirconium hydride is
also considered. The first HZr zone is arranged abutting to the outer diameter of fuel element (the
thickness of the layer is equal to 0.07 cm) and the second zone is arranged in the gap between the fuel
812

pellet and the steel cladding. In order for the effective multiplication factor to be equal to k eff
= 0.984,
a small fraction (3%) of natural Eu absorber was added to internal zirconium hydride layer. In this
scheme the plutonium incineration rate is equal to B Pu
= 33.2 kg/PFY (P r
= 80 MW) and the average
power density in the fuel is equal 31.0 W/g.
5. Production of 210 Po from 209 Bi in the cores with B 4
C and zirconium hydride
Production of 210 Po in the lead-bismuth eutectic was calculated for the reference core in the
ENEA report [8]. It was indicated that the main contribution to the production of 210 Po give neutrons
with energy below 20 MeV in the reaction of neutron capture:
209
Bi(n,γ)→ 210g Bi(T 1/2
= 5.013 days)→ 210 Po
The fraction α 210g
of the cross-section value for 210g Bi in the total (n,γ) cross-section of 209 Bi according to
the report [8] is equal to α 210g
= 0.72. Here we present the results of depression of 210 Po production in
the ADS reactor with fuel elements containing burnable absorber 10 B 4
C or zirconium hydride.
Calculations were performed using the MCNP-4b code for heterogeneous model of the reactor with
detailed description of fuel assemblies and the spallation neutron production target. The cross-section
of radiative capture (n,γ) on 209 Bi used in MCNP-4b is shown in Figure 3. In order to estimate the
amount of produced 210 Po we use the value α 210g
= 0.72 from the report [8]. The reaction rate (n,γ) on
209
Bi was calculated for different parts of the reactor and is 1.5 times lower for the core with B 4
C, and
3.3 times lower for the core with H 2
Zr compared to the reference core. The amount of 210 Po produced
in Pb-Bi eutectic in the reference scheme is equal to 3.6 kg/FPY.
6. Hybrid core with reference FA and FA containing zirconium hydride zones
A hybrid core contains two zones with reference fuel assemblies and a small fraction of fuel
assemblies with zirconium hydride. We will consider three schemes of arrangement of fuel
assemblies in the reactor core:
• (C1) in the first scheme the inner fuel zone (3-5 rings of hexagons in the core) contains
reference FA and the outer zone (6-7 rings of hexagons) contains fuel elements with HZr
cladding (see Figure 2).
• (C2) fuel assemblies with HZr are arranged in 7-8 rings of hexagons so that a thick layer of
Pb-Bi eutectics is located between the inner fuel zone and the outer fuel zone.
• (C3) fuel assemblies with HZr layer are arranged in a scattered order in the core containing
reference fuel assemblies.
In all above schemes the number of fuel assemblies containing HZr is equal to 48 and their
fraction is equal to 40% from the total number of fuel assemblies in the reactor core. In Table 3 main
parameters for the hybrid cores are presented. For comparison the parameters of the reference core
and of the core containing FA with HZr are also shown.
813

Figure1. Reference reactor core
Figure 2. Hybrid core with FA containing HZr
2
2
4
2 2
4
2 2
4
4
TARGET
2
3
4
4
TARGET
2
3
4
4
4
4
5
3
5
3
FA
Pb&Bi
Control Rods
FA
FA with HZr
Pb&Bi
Control Rods
C1 CORE. In the core C1 the total amount of 239 Pu burn-up is equal to B Pu
= 21 kg/PFY
(12.7 kg/PFY in the reference, and B Pu
= 8.4 kg/PFY in the FA with HZr). The radial peaking factor is
equal to k r
= 1.20 and is comparable with the value for the reference core. The neutron flux averaged
over the reference fuel assemblies in the core C1 is equal Φ n = 6.5 × cm
fuel 1014 -2 s -1 , and in fuel
assemblies with HZr cladding the flux is equal to Φ n = 2.9 × 10 14
cm -2 s -1 . The flux averaged over all
fuel
fuel assemblies in the core C1 is equal to Φ n = 5.0 × 1014 cm -2 s -1 . The production of 210 Po in (n,γ)
fuel
reaction on 209 Bi in region of dummy assemblies and in heat transfer regions is equal to 0.0088 (n,γ)
reactions per one fission neutron which is about 3 times less than in the corresponding zones of the
reference core.
C2 CORE. The main parameters for the core C2 is presented in Table 3. The amount of 239 Pu
burn-up is equal to B Pu
= 20.6 kg/PFY. However the radial peaking factor k r
= 1.28. In order to
decrease the k r
and to flatten the power density distribution, the enrichment in the inner fuel zone
should be increased while in the outer zone containing FA with HZr cladding should be decreased
below 10%. The amount of produced 210 Po in dummy assemblies and in heat transfer regions is equal
to 0.013 (n,γ) reactions per fission neutron, which is 2.02 times less than the production of 210 Po in the
same zones of the reference core.
C3 CORE. In this core 48 fuel assemblies with HZr layer have plutonium enrichment of 10% and
arranged in a scatter order in the reactor core. In Table 3 the parameters for this core are presented.
814

Table 3.Neutronics parameters of hybrid two zones core
containing reference FA and FA with HZr layer calculated at BOC using the MCNP-4b code.
Type of core Reference FA
with HZr
(C1) (C1) (C2) (C3)
(mean enrich.) 23.2% 20.0% 25.0% 23.2% 20.0% 20.0%
Reference FA
Number of FA
Enrichment ,%
FA with HZr
Number of FA
Enrichment ,%
Hex rings of
120
23.2
0 72
29.6
0 48
20
48
18
48
28
48
10
48
10
3-7 6-7 6-7 7-8
3-7
FA with HZr –
6-dummy
Burn-up 239 Pu,
12.66 (Ref.) 3.30 (Ref.) 9.73 (Ref.)
8.40 (HZr) 22.2 (HZr) 10.83 (HZr)
B Pu9
(kg/FPY) 14.3 24.8 21.1
25.5
20.6
q f
, (W×cm -3 ) 216 217 216 216 215 209
72
20
72
26.6
72
26.6
13.51 (Ref.)
7.07 (HZr)
20.6
k r
1.20 1.19 1.20 1.80 1.28 1.17
Φ n , fuel
(n cm -2 s -1 ) 8.5 × 10 14 3.2 × 10 14 6.5 × 10 14 (Ref.)
2.9 × 10 14 (HZr)
5.0 × 10 14 (Av.)
4.1 × 10 14 (Ref.)
3.4 × 10 14 (HZr)
3.8 × 10 14 (Av.)
6.3 × 10 14 (Ref.)
2.8 × 10 14 (HZr)
4.9 × 10 14 (Av.) −
E n 0.518 (Ref.)
0.466 (Ref.)
fuel
0.572 (HZr)
0.566 (HZr)
(MeV) 0.430 0.650 0.530 (Av.)
0.489 (Av.)
210
Po,
per 1 fiss. n 0.0263 0.0078 0.0088 – 0.0130 0.0110
k eff
0.9839(5) 0.9831(5) 0.9852(6) 0.9824(5) 0.9802(6) 0.9812(9)
7. Comparison of incineration parameters for the reference core, for the core with fuel
elements containing B 4
C or zirconium hydride
The main parameters of the reference core and of the core containing 10 B 4
C absorbing zone, or
zirconium hydride moderator zone are presented in Table 4. The effective multiplication factor
calculated using the MCNP-4b code with cross-section library DLC-181 (T = 300 K) for the core with
B 4
C is equal to k eff
= 0.9833 ± 0.0005, and for the core with zirconium hydride k eff
= 0.9831 ± 0.0006.
The average power density in the fuel pellets with B 4
C is equal to q f
= 22 W/g (235 W/cc), and in the
core with H 2
Zr - q f
= 21.8 W/g (227 W/cc). The power peaking factors for the cores with 10 B 4
C, H 2
Zr
and for the reference core are equal to 1.33, 1.37 and 1.35 respectively. So, these parameters are very
close to the values in the reference core, but the plutonium burn-up rate at the BOC in fuel elements
with B 4
C, H 2
Zr is about two times higher compared to the reference core: 25 kg/FPY compared to
13.7 kg/FPY.
815

Table 4. Comparison of burn-up parameters of the reference EAP-80 core and
the core containing FA with burnable absorbers or moderated zones
Reference core FA with B 4
C clad FA with ZrH 2
matrix
Enrichment 23.2% 43.7% 20.0%
r fuel
, cm 0.09 0.1 0.09
R fuel
, cm 0.357 0.347 0.357
r clad
, cm 0.3685 0.3685 0.3685
R clad
, cm 0.425 0.425 0.425
k eff
0.9839(5) 0.9833(6) 0.9831(6)
k s
0.980 0.982 0.982
I p
(mA) 1.67 1.51 1.51
q f
(W/g) 21.98 22.60 21.82
k v
= q max
/q l
1.35 1.37 1.37
Φ n fuel (cm-2 s -1 ) 8.5 × 10 14 5.8 × 10 14 3.2 × 10 14
E n (MeV)
fuel
0.43 0.60 0.65
B Pu9
(kg/FPY) 13.7 25.0 24.8
The comparison of neutron spectra Φ n
averaged over all fuel pellets in the reference core, in the
core with B 4
C and in the core with zirconium hydride is presented in Figure 3. It is seen that B 4
C
absorber and zirconium hydride reduce the fraction of neutrons absorbed in (n,γ) reaction on 238 U in
resonance region, and zirconium hydride moreover shifts the neutron spectrum to the thermal energy
region.
n
The average energy E fuel of neutron spectrum (calculated using fission neutron source) in fuel
n
n
pellets with B 4
C is equal to E fuel = 0.60 MeV, while in the reference core E fuel = 0.43 MeV. Neutron
fluxes
nfuel in fuel pellets at the reactor power P = 80 MW are equal respectively to 8.5 × , 5.8 × 10
r 1014 14
and 3.2 × 10 14
(cm -2 s -1 ) for the reference fuel pellets, for fuel elements with B 4
C and with zirconium
hydride. The energy dependence of integrated plutonium burn-up rate (i.e. integrated over thin energy
interval and over all fuel elements) for the thermal energy region is shown in Figure 4 for the reference
core, for the core with B 4
C and with zirconium hydride. In Table 5 we present the fractions of plutonium
incineration rate in 6 energy intervals: 0 – 10 -4 MeV, 10 -4 – 10 -3 MeV, 10 -3 – 10 -2 MeV, 10 -2 – 10 -1 MeV,
10 -1 – 1 MeV and 1 – 10 MeV. In fuel elements with B 4
C absorber the plutonium is burning in the fast
energy region, while the plutonium is burning mainly in the thermal energy region in FA with zirconium
hydride. In the fast and thermal energy regions the ratio of fission cross section of 239 Pu to the capture
cross-section for 238 U is high.
816

Figure 3. Neutron spectra averaged over fuel
pins in the reference core, in FA with B 4
C, and
in Fa with H 2
Zr. Neutron cross-section
in the reaction 209 Bi(n,γ) is also shown
Figure 4. Energy distribution of plutonium
burn-up in the reference core,
in FA with B 4
C, and in FA with HZr
Table 5. Energy dependence of B Pu,
plutonium burn-up rate (%) versus the neutron energy in the
ADS EAP-80 in different cores containing: reference fuel elements,
fuel elements with thin B 4
C cladding, Fuel elements with H 2
Zr matrix.
Ei
Here B Pu
(E i-1,
E i
) = m
∫ [ Σ 239 ( E) + Σ 239 ( E) − Σ 238
( E)
] Φ( E)
dE.
f C C
Ei−1
Neutron energy
(MeV)
Energy dependence of plutonium B Pu
(E i-1,
E i
) burn-up,
%
(E i-1,
E i
) Reference core.
Enrichment 23.2%
Core with B 4
C cladding.
Enrichment 45.0%
Core with ZrH 2
matrix.
Enrichment 20.0%
0 – 10 -4 3.78% 0.284% 71.10%
10 -4 – 10 -3 9.78% 0.73% 14.66%
10 -3 – 10 -2 8.89% 4.06% 1.08%
10 -2 – 10 -1 4.74% 19.47% 0.00%
10 -1 – 1 54.81% 56.39% 6.87%
1 – 10 18.0% 19.07% 6.29%
B Pu
, (kg/FPY) 13.7 25.2 24.8
817

8. Power distribution in the reference core, in the core with B 4
C, H 2
Zr
8.1 Reference core
Power density distribution in the fuel elements was calculated using the spatial distribution of
fission events in fuel assemblies (photon heating in the present calculations was not taken in account
due to the small value compared to the fission reactions). The axial power peaking factor was
calculated as a ratio of fission energy in the axial layer to the average fission energy in the fuel
assembly where the axial layer is located. The radial power peaking factor was calculated as the ratio
of the fission energy in each fuel assembly to the energy averaged over all fuel elements. The spatial
distribution of energy in the reactor core with the fission neutron source differs from the energy
distribution in the core with spallation neutron source. This difference depends on the level of the
core sub-criticality and for the core with k eff
close to 1.0 this difference will be very small. However,
the changing of the parameter k eff
to the higher sub-criticality during the burning of fissile 239 Pu will
increase the difference in power peaking factors calculated for the cores with the fission source and with
the spallation neutron source. For the reference core with the fission source k v
= 1.37 (k eff
= 0.984) and
k v
= 1.45 for the spallation neutron source.
8.2 Fuel elements with B 4
C cladding
The radial peaking factor calculated for the core with the spallation neutron source is equal to k v
= 1.47
while in the core with fission source the maximal value of power peaking factor is equal to the value of
k v
= 1.37. The average power density in the fuel is equal to q f
= 235 ± W/cc (q f
= 244 ± 12 W/cc for the
spallation source), that is comparable with the reference power density q f
= 229 ± 12 W/cc.
8.3 Fuel elements with zirconium hydride cladding
The average power density in fuel elements with H 2
Zr cladding arranged abutting to the outer
surface of fuel element and with plutonium enrichment of 20% is equal q f
= 227 ± 12 W/cc, and the
maximal value of the peaking factor is equal to k v
= 1.35. The power density distribution in fuel
elements with the internal zirconium hydride cladding and plutonium enrichment of 23.2% has
increased slightly because of lower volume of fuel and is equal to q f
= 258 ± 12 W/cc and the
maximal value of the peaking factor is equal to k v
= 1.32.
9. Fuel depletion and actinide transmutation
Time evolution of fuel composition and calculation of actinide transmutation rate in fuel cycle
were performed using the ORIGENS module of the SCALE 4.3 computational system. Problem
dependent effective cross-section for all nuclides in the fuel were prepared using the COUPLE,
NITAWL and XSDRNPM modules in the SCALE system. The reactor power was taken equal to
80 MW. Time dependent nuclear concentrations of actinides and of fission products were used in
MCNP-4b Monte Carlo computation of changing the effective multiplication factor, neutron
multiplicity factor and of the reaction rates in the fuel during the fuel irradiation.
The transmutation rate of actinides is defined as the fractional difference in the final mass of the
actinides (Pu, Am, Cu) and the initial mass of actinides. Transmutation rate for all actinides
calculated after 1 FPY of the fuel irradiation in the Reference Core is equal to -1.4%, in the core
818

containing FA with B 4
C cladding is equal to -2.0%, and in the core with FA with zirconium hydride
cladding -2.3%. In the hybrid two-zones reactor core the transmutation rate is equal to -2.2%.
In Table 6 the mass of actinides at the BOC and after 1 FPY of fuel irradiation, as well as the
burn-up rate B Pu
(kg/FPY) for 239 Pu, are presented for different compositions of the reactor core:
Table 6. Mass of actinides in the BOC and after 1 FPY of fuel irradiation
in the reference core, in the core containing FA with B 4
C, and FA with HZr.
Time T = 0 (BOC) T = 360 days
Core
(Enr., %)
Reference
(23.2%)
B 4
C
(43.7%)
HZr
(24%)
Hybrid
(23.2%)
Reference
(23.2%)
B 4
C
(43.7%)
HZr
(24%)
Hybrid
(23.2)
239
Pu, kg 519.7 908.2 539.1 560.7 506.7 884.3 513.8 539.4
B Pu
, kg 0 0 0 0 -13.0 -23.9 -25.3 -21.3
241
Pu, kg 19.6 34.2 20.3 21.2 19.5 32.8 25.3 23.5
241
Am, kg 17.4 30.4 18.1 18.8 17.6 31.6 16.5 18.5
Pu, Am, Cu, kg 750.0 1311.3 778.0 809.3 739.8 1284.8 759.9 791.3
The burn-up rate of 239 Pu in FA with 10 B 4
C or HZr is about two times higher than in the reference
core. It should be noted that the burn-up rates of 239 Pu calculated using the MCNP-4b reaction rates
(see Table 4) are slightly different from results in the above table, because we supposed that reaction
rates remain constant during the time of irradiation. In the core with 10 B 4
C the burn-up rate of 241 Pu is
about -1.4 kg/FPY, while in the core with HZr the mass of 241 Pu is increased by 5 kg/FPY. The amount
of 241 Am in the reference core is increased by 0.2 kg/FPY, in the core with 10 B 4
C is increased by
1.2 kg/FPY, while in the core with HZr the amount is reduced by -1.6 kg/FPY, or by -0.3 kg/FPY in
the hybrid core.
It should be noted that the effective cross-section of 10 B in FA with 10 B 4
C is about 1.6 barn, and
the maximum of the neutron spectrum is located in the energy region higher then 0.1 MeV. So, the
burning of 10 B 4
C cladding in the fuel element during fuel cycle is small. Nevertheless, the burning of
10
B 4
C cladding was taken into account in the MCNP calculations. Production of fission fragments is
comparable in all considered schemes of the reactor core: 0.002 kg of 135 Xe, 0.1 kg of 149 Sm, 1.0 kg of
137
Cs, 0.7 kg of 99 Tc, 0.2 kg of 129 I per 1 FPY. The effective absorption cross-section of 135 Xe in fuel
elements with HZr is three orders of magnitude lower than in the thermal power reactors and is about
1 800 barns, and samarium 149 Sm has effective cross-section of absorption equal to 80 barns. In
calculations of reactivity loss during the fuel cycle the presence of all fission fragments were taken
into account. Reactivity calculations were performed using the MCNP-4b code. In the reference core
the reactivity loss ∆ρ during 1 FPY is equal to ∆ρ = -1.6%/FPY, in the core with HZr zone
∆ρ = -2.0%/FPY, in the hybrid core with HZr ∆ρ = -1.8%/FPY, while in the core with
10
B 4
C ∆ρ = –1.1 %/FPY.
The dependence of proton beam current on k eff
may be estimated using the simple formula:
P
r
= I
p
k
eff
M
( 1 − k )
eff
n , xn
ν
f
× N
n / p
ε
f
,
819

where P r
is the reactor power, I p
is the proton beam current, k eff
is the effective multiplication factor,
N n/p
is the multiplicity of spallation neutrons in Pb-Bi target irradiated by protons, ε f
= 190 MeV, ν f is
the average number of neutrons per fission event, M n,xn
multiplication factor taking into account nonfission
multiplication reactions (n,xn) for primary spallation neutron. If we assume that the reactor
power is constant during the reactor company, and that N n/p
also does not depend on time then the
dependence of proton beam current I p
(T) on k eff
may be estimated as
,
S
( 7 ) = , ( )
S
N
N
HII
HII
( )
( )
( ) − N
HII
( 7 )
( 7 ) − N ( )
where I p
(0) is the proton beam current and k eff
(0) is the effective multiplication factor at the beginning
of the reactor company. Using the values of reactivity loss ∆ρ and T = 1FPY we may estimate the
ratio I p
(T)/I p
(0). For the reference core I p
(T)/I p
(0) = 2.0, and for the core containing FA with 10 B 4
C
I p
(T)/I p
(0) = 1.6.
HII
10. Conclusion
Application of a thin cladding of burnable absorber B 4
C or a thin zone containing hydride
zirconium moderator in the PuO 2
&UO 2
MOX fuel element of the EAP-80 accelerator driven reactor
prototype permits considerably to increase the plutonium burn-up rate. Incineration of 239 Pu in FA
with B 4
C or with HZr is about 25 kg/FPY at the reactor core power of P r
= 80 MW th
which is about
2 times greater compared to the reference core. Plutonium enrichment in fuel elements with B 4
C
cladding was chosen to be 43.7%. In the core containing fuel elements with zirconium hydride a fuel
with plutonium enrichment of 20-23% is used. The geometry of the reactor core and of fuel
assemblies in the proposed schemes was unchanged and remains the same as in the reference core. A
hybrid core containing reference fuel elements in the inner region of the core and fuel elements with
zirconium hydride cladding in the outer region of the core was considered. A reactor core with
scattered arrangement of fuel assemblies with HZr is also considered. The fraction of fuel elements
with zirconium hydride cladding in the hybrid cores was equal to 40% from the total number of fuel
elements loaded into the core. The average power density in fuel with B 4
C or with H 2
Zr and in hybrid
cores at the beginning of fuel cycle is comparable with the power density in the Reference Core and is
equal to q f
= 224 W/cc, q f
= 227 W/cc and q f
= 215 ÷ 217 W/cc, respectively. The power peaking
factor in these cores remains the same as in the reference core and is equal to k v
= 1.37 (for exception
of one hybrid scheme). The dependence of main neutronic, actinide and FP transmutation
characteristics for different core models during time of irradiation has been calculated. The
transmutation rates for 239 Pu, 241 Am and for total actinides is maximum for the core with fuel elements
containing zirconium hydride cladding. The transmutation rate of 241 Pu is maximal in the core with
fuel elements having B 4
C cladding. The reactivity loss during time of irradiation is minimal in the
core containing fuel elements with the cladding of B 4
C burnable absorber.
Acknowledgements
S. Kaltcheva-Kouzminova and V. Kuzminov are grateful to ENEA at Bologna (Italy) where this
work was performed. We would like to thank G. Glinatsis for supplying us with the detailed
information about the EAP-80 reference core and for discussion, and also L. Cinotti, R. Tinti, K. Burn
and A. Konobeev, Yu. Petrov, and V. von Schlippe for their helpful discussion. We are grateful also
to N. Voukelatou for the help during the preparation of this work.
820

REFERENCES
[1] ANSALDO Technical Reports for the Reactor Core of the ADS Demonstration Facility,
1999-2000.
[2] S. Atzeni, Preliminary Neutronics Analysis of the Energy Amplifier Prototype, 1998, Report EA
D2.02 4 407.
[3] C. Rubbia et al., Conceptual Design of a Fast Neutron Operated High Power Energy Amplifier,
CERN/AT/95-44 (ET), 1995.
[4] T. Wakabayashi, Status of Transmutation Studies in a Fast Reactor at JNC. Actinide and
Fission Product Partitioning and Transmutation. Proc. of the 5 th Int. Inform. Exch. Meeting on
Actinide and Fission Product Partitioning and Transmutation, Mol,Belgium, 1998, EUR 18898
EN, OECD/NEA (Nuclear Energy Agency), Paris, France, 1999.
[5] Judith F. Breismeister, MCNP – A General Monte Carlo N-Particle Transport Code,
LA-12625-M.
[6] The SCALE 4.3 Computational System.
[7] A.V. Daniel, V.Yu. Petrov, E.A. Sokolov, Program SITHA, Communication of the Joint
Institute for Nuclear Research, 1991, Dubna.
[8] W. Hongchun, T. Takeda. Minor Actinides Incineration By Loading Moderated Targets in Fast
Reactor, Proc. of Int. Conf. Mathematics and Computation, Reactor Physics and Environmental
Analysis in Nuclear Applications, Madrid, 1999, p. 580-586.
[9] Hongchun Wu, Diasuke Sato, Toshikazu Takeda. Minor Actinides Incineration By Loading
Moderated Targets in Fast Reactor, Journal of Nuclear Science and Technology, 2000, Vol. 37,
No. 4, p. 380-386.
[10] T. Sanda, K. Fujumura, K. Kobayashi, K. Kawashima, M. Yamawaki, K. Konashi, Fast Reactor
Core Concepts for Minor Actinides Transmutation Using Hybrid Fuel Targets, Journal of
Nuclear Science and Technology, 2000, Vol. 37, No. 4, p. 335-343.
[11] A.Yu. Konobeev, M. Vecchi, Nuclide Composition of Pb-Bi Heat Transfer Irradiated in
80 MW Sub-critical Reactor, ENEA Activity Report, 1999, Workshop on Spallation Module.
821

REMARKS ON KINETICS PARAMETERS OF A
SUB-CRITICAL REACTOR FOR NUCLEAR WASTE INCINERATION
Juan Blázquez
Department of Nuclear Fission, CIEMAT,
22, Complutense Av., 28040 Madrid, Spain
Abstract
Accelerator driven systems (ADS) are being designed as nuclear waste incinerator in order to be
complementary with the proposed geological disposal. When the ADS is based on a highly
sub-critical reactor, some elementary concepts designed for critical reactors need to be reconsidered:
• How much sub-critical means highly sub-critical?
• Is the spallation source really external?
• Are the kinetics parameters depending on the neutron source?
• Can standard neutron noise techniques be used when the neutron source is not Poisson-like?
The article remarks the subtleties behind those questions aiming to clarify future experiments with
spallation source focused to nuclear waste incineration.
823

1. Introduction
At present, there are 423 nuclear power plants (NPP) operating around the world. Most of them
are light water reactors, being the PWR type the more abundant, 247 NPP, followed by the BWR
type, 92 NPP. As a consequence, the 3 000 MW th
PWR type is used as a reference for roughly
estimation of the actinide amounts the humankind has to deal with shortly.
The total nuclear power around the world amounts 351.7 GWe, and it seems to follow the Pareto’s
law for the Economy, in the sense that the 20% of the 31 countries where nuclear power is installed have
the 75% of the power [1]. Therefore, the nuclear waste problem has different weight for each country
and no common solutions are expected to be taken. Based on the global power, operating during
40 years – 330 days/year – the forecast waste for the year 2010 is approximately [2]:
Table 1. The nuclear waste for the year 2010
Total spent fuel
Plutonium isotopes
Neptunium
Minor actinides
Long lived fission fragments
99
135
Tc
Cs
I
129
281 600 t
2 816 t
131 t
113 t
235 t
84 t
56 t
The case of Spain can be meaningful [3,4]. The nuclear power is 7.74 GWe, about the 2.1% of
the world nuclear power. According to Table 1, the expected amount for plutonium is 60 tonnes
roughly. This material may be regarded either as a waste to take care of, either a fissile material to
extract energy from.
After fuel reprocessing, accelerator driven systems (ADS), based on a proton accelerator coupled
to a highly sub-critical reactor, can be designed for the actinide and long lived fission fragment
incineration, acting as a complementary solution for the nuclear waste problem [5]. If so, it has
several advantages:
• The geological disposal will be cheaper.
• Help for public acceptance of a geological disposal.
• Nuclear waste is converted into fuel.
The last item comes from the energy content of the actinides, about 940 MWD/Kg. The
economical income coming from the incineration of the actinides produced in Spain will cover the
cost of the incinerator, hence, a priori is not a bad option for the nuclear waste problem. Nevertheless
there are still many uncertainties around the ADS designs requiring detailed research. Highly
sub-critical reactor driven by a spallation neutron source shows novel quests, in particular how to
measure, control and even understand the k eff
, when the reactor is far from critical and what does it
mean to be far from critical.
824

2. Sub-critical multiplication
The incineration rate should be high. This is achieved with a neutron flux higher than
5.10 13 n cm -2 s -1 , corresponding to a typical PWR. According to the flux, the neutron population N can
be calculated from φ = vN/V R
, where v is the neutron mean velocity and V R
the reactor volume. In its
turn, the global neutron population is: N = Sl/(1-k), where l is the prompt neutron mean life and S the
external neutron source intensity in n/s units [6]. Preliminary designs contemplate a magnitude about
5.10 15 n cm -2 s -1 for φ; a flux high enough for fast incineration and breeding purposes – 233 U from 232 Th.
Such a flux needs a sub-critical reactor with a spallation neutron source. The closer to critical the
higher the flux, but the restriction of non increasing the actual actinide inventory limits the fissile fuel
chosen for the incinerator; as a consequence, preliminary design deals with k eff
about 0.95, which
yields a multiplication of 20. Hence, the magnitude of S will be around 4.10 17
n/s. Such a high
intensity external neutron source is one of the key points of the incinerator. Preliminary designs are
focused on the spallation of neutrons caused by 1.0 GeV proton beam coming from an accelerator;
protons collide with a metallic target of high mass number -normally lead- causing neutrons to be
expelled out from the nuclei until protons are stopped. The neutron source features are defined by the
target and the accelerator current; some figures can define the order of magnitude: the operating
accelerator current between 29 mA, the energetic neutron cost -in molten lead- about 36 MeV, the
neutron spectrum is evaporation-type, similar to fission spectrum but with a higher energy tail.
Figure 1. Evolution of k eff
Along the fuel burn-up the fuel composition will not remain constant, k eff
and the other kinetics
parameters will drift slowly. That is a difference with critical reactors where the neutron flux
increases in order to compensate the decreasing fuel density, keeping the power constant; the
capability of keeping k eff
= 1 defines the burn-up period. In a sub-critical reactor the burn-up period is
much longer because it is not limited by k eff
, but a drift of 1% in k eff
yields a drift of 20% in the flux,
so in order to operate with constant power, it is important to measure k eff
accurately, and it is not easy
for such a low multiplication with a non stationary source.
The relationship between this source and the k is even deeper. Being low the multiplication, the
neutron shape function is affected by the source position; so, a change in the position of the spallation
825

target means a change in k and the rest of kinetics parameters. And not only in the space domain, but
in the time domain, because the spallation source can have a fast pulsed nature due to the accelerator,
which in its turn affects to the flux shape normalisation, causing that static and dynamic reactivity
measurements might differ. The above ideas are to be explored in the next sections.
3. External source and sub-criticity
In the case of a highly sub-critical reactor the k depends on the external neutron source. When
close to critical state, this dependence is too weak and can be neglected. To argue that, the qualitative
definitions of k is:
k
fission rate
=
( absortion + leakage)
rate
according to the diffusion equation the leakage rate can be written as:
− D ∆φ
= νΣ
φ − Σ φ + S
f
where S V
stands for the stationary external source per volume unit and the other symbols have the
habitual meaning. In terms of the buckling : B 2 = (νΣ f
-Σ a
)/D = (k ∞ -1)/L 2 :
k
=
( Σ
a
νΣ
φ
f
2
a
+ DB ) φ + S
In critical reactors the term DB 2 φ stands for the neutron leakage rate and DB 2 is regarded as a
correction for Σ a
, but in sub-critical reactors the leakage rate is a function of the source which can be
ignored when the source intensity is negligible compared with the fission rate, i.e. for large
multiplication. Considering that φ is proportional to S V
it is clear that k does not depend on the power
level.
In the Figure 2 two radial neutron flux distribution are plotted. They correspond to a cylindrical
sub-critical reactor energy amplifier type with spallation source placed at the centre [5]. It is well
known that the radial flux distribution is:
• Cosinus like k ∞ >1.
• Straight line k ∞ = 1.
• Exponential like k ∞

Figure 2. Radial flux distribution for a energy amplifier type
Let be the flux shape ψ be defined by: φ(r,t) = N(t)ψ(r,t) with the normalisation condition:
=1
where the bracket denotes the scalar product and W is the adjoint flux. Even when the k eff
is calculated
by the usual procedure:
k
eff
< W ; νΣ
fψ
>
=
< W;(
Σ − D∆)
ψ >
due to the laplacian operator, a change of ψ affects to k eff
. As a conclusion, for ADS the spallation
source is not external and has to be considered for the calculation of the sub-critical multiplication
factor.
The same argument applies for the rest of the kinetics parameter β and Λ. It must be
distinguished between β as a quantity and β eff
. as a parameter. The quantity is a property of a given
fissionable nucleus, so it does not depend on the spallation source; but the parameter does not
correspond to any nucleus and it is calculated using the shape function [6,7], so it depends on the
spallation source.
a
4. The stationarity of the spallation source
It is to be remarked that static and dynamic reactivity measurements might not coincide in a ADS
with a spallation source. In the case of fast pulsing accelerator, the neutron flux shape function may
be time dependent. If so, the Gyftopoulos term [8] should be added to the point kinetics equation
causing that both type of measurement yield different results. Explicitly:
827

Following the normal procedure for point kinetics derivation [7] the shape function is defined
with the condition:
so, for the global neutron population N(t):
< W; ψ > = 1
v
∂ ψ ∂N
[ N < W;
> ] =
∂t
v ∂t
but when the transients are so fast that < W; ψ / v >≠ 1, because the shape function is time dependent,
then:
∂ ψ ψ ∂N
1 ∂Ψ
[ N < W;
> ] =< W;
> + N < W;
>
∂t
v v ∂t
v ∂t
therefore a new term is carried into the point kinetics equations, now appearing as [8]:
dN
dt
ρ − β
= ( − λS ) N + λC
+ S
Λ
dC
dt
β
= N − ( λ + λ ) S
C
Λ
where all the symbols have its habitual meaning except:
1 ∂ψ
< W;
>
λ ≡
v ∂t
S
ψ
< W;
>
v
This new term vanished when the shape function does not depend on time, the static case. For
fast transients, if one defines ρ
g
≡ ρ + λSΛ
and λ
g
≡ λ + λS
the point kinetics normal form is
restored; but in that case, the kinetic and static measurements of reactivity might differ.
5. The neutron noise procedure
Neutron noise measurements are proposed for reactivity control in large sub-critical reactor.
Because reactivity feedback will hardly change the multiplication factor, some procedures useful in
zero power reactors could be used. Particularly the Rossi-alpha and the Feynman-alpha techniques
seem the most promising procedures for estimating the sub-critical reactivity and the kinetics
parameters. For instance, in the case of Feynman-alpha [9]:
2
σ ( t)
Dν
= 1+
ε
Z ( β − ρ)
2
1−
e
(1 −
αt
where Z is the average neutron detector count during the time t, ε is the detector efficiency, σ the
standard deviation when the experiment is repeated several times, D ν the Diven factor and
α = (ρ-β)/Λ the parameter to be determined in order to measure the reactivity.
−αt
)
828

Using such a traditional expression, the reactor should be sub-critical and stationary. The
formula is based on the assumption that the neutron source follows the Poisson statistics, i.e. the
probability of emitting one neutron in dt is dt/S -1 = Sdt. Clearly the pulsed spallation source is not a
Poisson like source because they are at least two well defined times: the pulse duration and the
inverse pulse rate, so some research is still needed determining an equivalent Feynman-alpha
expression.
A correction is proposed recently, multiplying the Diven’s factor by:1+SD S
(-ρ)/(νD ν ), where D S
is the Diven’s factor for the source.
D
S
< S(
S −1)
>
=
2
S
aver
but the derivation is based on a continuous spallation source rather than pulsated. Besides, if S were
known, the simpler multiplication factor expression is much more convenient determining the k. In
spite of being of little practical interest, the expression above alert us about the correction of D ν , not
only because of the spallation source, but also because of ν, which is drifting continuously as a
consequence of the fuel burn-up.
As a way out, the Feynman-alpha technique suggest that:
σ
=
Z
a +
b
P
where a and b are two constants to be fitted, having in mind that the left hand term does not depend
on the detector efficiency and the power P can be changed by changing the accelerator pulse rate
without affecting to the reactivity.
6. Summary
Some remarks for sub-critical reactors with spallation neutron source are made:
• After fuel reprocessing option, the actinide content will cover the incinerator costs.
• The k eff
is going to drift along the reactor operation so new methods for reactivity control are
to be found.
• The neutron source is of capital importance for highly sub-critical reactor and cannot be
considered as “external”.
• The k eff
and the others kinetics parameters depend on the source position.
• The shape function of the neutron flux might not be constant along a given fast transient due
to the fast pulse rate of the spallation source. If so, static and dynamic measurements of
reactivity might differ.
• Neutron noise analysis techniques for controlling reactivity can be used, but the traditional
way of doing it should be corrected.
As a conclusion, the ADS option deserves still a lot of research, even for those elemental and
well-established concepts.
829

Acknowledgements
This study has been performed within the collaboration expert of CIEMAT-ENRESA for the
Transmutation of Long Lived Isotopes.
REFERENCES
[1] Energía 2000, Foro Nuclear, Dep. Legal M-28233-2000, Spain, 2000, pp. 116-117.
[2] C. Rubbia, S. Buono, E. González, Y. Kadi and J.A. Rubio, A Realistic Plutonium Elimination
Scheme With Fast Energy Amplifiers and Thorium-plutonium Fuels, CERN/AT/95-53(ET),
1995, p. 3.
[3] R. Cortés, Los Residuos en las Centrales Nucleares, Gestión de Residuos Radiactivos, Serie
Ponencias, CIEMAT, Spain, 1990, Vol. I, pp. 6.6-6.7.
[4] F. Alvarez Mir, Los Residuos Radiactivos en las Centrales Nucleares, Ponencias Generación de
Iberdrola 1999, Dep. Legal M-32495-2000, Spain, Vol. I, pp. 197-222.
[5] C. Rubbia, J.A. Rubio, S. Buono et al., Conceptual Design of Fast Neutron Operated High Power
Energy Amplifier, CERN/AT/95-44(ET), 1995, pp. 23-58.
[6] D.L. Hetrick, Dynamics of Nuclear Reactors, The University of Chicago Press, 1971, p. 19.
[7] A.F. Henry, Nuclear Reactor Analysis, The MIT Press, 1975, pp. 296-333.
[8] E.P. Gyftopoulos, General Reactor Dynamics, The Technology of Nuclear Reactors Safety,
The MIT Press, Vol. I, pp. 175-204.
[9] M.H.R. Williams, Random Processes in Nuclear Reactors, Pergamon Press, 1974, pp. 26-29.
830

NOISE METHOD FOR MONITORING
THE SUB-CRITICALITY IN ACCELERATOR DRIVEN SYSTEMS
Y. Rugama 1 , J.L. Muñoz-Cobo 1 , T.E. Valentine 2 , J.T. Mihalczo 3 , R.B. Perez 3 , A. Perez-Navarro 4
1 Universidad Politécnica de Valencia, Chemical and Nuclear Engineering Department
P.O.Box 22012, 46071 Valencia.Spain.
E-mail: yrugama@iqn.upv.es
2
Computation Physics and Engineering Division, Oak Ridge National Laboratory
Oak Ridge, TN, 37831-6362, USA
3
Instrumentation and Controls Division, Oak Ridge National Laboratory
Oak Ridge, TN, 37831-6004, USA
4
LAESA,
Plaza Roma, F-1, 1a planta, 50010 Zaragoza, Spain
Abstract
In this paper, an absolute measurements technique for the sub-criticality determination is presented.
The development of ADS, requires of methods to monitor and control the sub-criticality of this kind of
systems, without interfering it’s normal operation mode. This method is based on the Stochastic
Neutron and Photon Transport Theory developed by Muñoz-Cobo et al. [1], and which can be
implemented in presently available neutron transport codes. As a by-product of the methodology a
monitoring measurement technique has been developed and verified using two coupled Monte Carlo
programs. The spallation collisions and the high-energy transport are simulated with LAHET. The
neutrons transports with energies less than 20 MeV and the estimation of the count statistics for
neutron and/or gamma ray counters in fissile systems, is simulated with MCNP-DSP.
It is possible to get the kinetics parameters and the k eff
value of the sub-critical system through the
analysis of the counter detectors.
831

1. ADS description
A conceptual design of the fast energy amplifier has been proposed by Rubia et al. (1995) [2] at
CERN. The prototype design used in this paper consists of a fast neutron sub-criticality facility fuelled
with U 233
and Th 232
and which is cooled with liquid lead-bismuth under natural convection. The driving
neutron source are spallation neutrons produced by an intense beam (about 11 mA and 380 MeV) proton
pulse from a cyclotron. The proton beam is injected in the lead-bismuth coolant slightly above the core
centre, which is used as a spallation target for the neutron generation that will be used as source for the
ADS (Accelerator Driven System).
The neutron lifetime of this sub-critical system is about 1 µ s and this value depends on the
cooling fluid and the composition and geometry of the assembly.
The envisaged k eff
values are ranging between 0.945 and 0.985.
2. The cross-power spectral density
The behaviour of neutrons and gamma rays in a nuclear reactor or configuration of fissile materials
can be represented as a stochastic process. Measurements of the fluctuations of the neutrons and also of
gamma rays in a system are typically used to characterise the fissile material. The representation of the
statistical descriptor by Muñoz-Cobo et al. [1] for neutrons and Muñoz-Cobo et al. [3] for coupled
neutron and photon, is applicable for many neutron noise analysis measurements.
In this paper, we have used fission detectors sensitive to neutrons and the formalism used, to
determine the noise methodology for monitoring the sub-criticality of the ADS, was developed by
Verdu et al. [4] and Muñoz-Cobo et al. [1].
The cross-power spectral density CPSD 12
(w) between one source detector and system detector
will give us the possibility to know the k eff
value from the position of the first pole in the phase and
Bode diagrams.
From the source probability-generating function defined as:
∞
∑ ∑
∞
N = 0 M =
NC
M C
GS
( z1,
z2,
d1,
d2
) = z1
z2
PN
M
( d1,
d
2
)
(1)
Where the probability is defined as:
P ( d , d
N 1 2
) → Probability to have N
CM C
C
counts in detector 1, and
start up of the proton beam source at time t o
.
C
C
0
C
C
M
C
counts in detector 2, upon the
The source for the ADS is a proton pulsed beam which has been defined, on the theory and
simulation, as a periodic array of Dirac Deltas with period T.
N
∑ − 1
o
n=
0
S(
t)
= S δ ( t − nT )
(2)
Defining the covariance between detectors D 1
and D 2
by:
Cov( d1,
d
2
) = NC ( d1)
M
C
( d
2
) − NC
( d1)
× M
C
( d
2
)
(3)
832

And the descriptor used in noise analysis, the cross-correlation function, as the following limit of
the covariance function:
Cov(
d1,
d
2
)
Φ12
( tf1,
tf
2
) = lim∆ tc1→0
(4)
∆tc
→0
∆tc
∆tc
2
From this definitions and using the adjoint transport equation, it is possible to obtain the
CPSD 12
(w) expression from the Fourier transform of (4), given by:
1
D2
CPSD12 ( w)
= dw2S(
w w2
) g
n / p
exp( i(
w w2
) tf1)
h(
w2
) I ( w)
2
∫ ∞ +
+
(5)
π −∞
Where the average importance for source neutrons is given by:
I
D
∫ d r
S
( r´)
∫ dv´
∫ dΩ´
f
sp
( v´,
Ω´)
nC
( r´,
v´,
Ω
D 2 2
( 3
1
2
w)
= ρ ´, w)
(6)
n D2 ( r´,
v´,
´, w
C
Ω ) is the average number of counts produced in detector D 2
, per injected neutron at
the phase point and frequency ( r ´, v´,
Ω ´, w)
and S ( w + w2
) is the Fourier Transform of the proton
pulse source, h(w 2
) is the Fourier Transform of the response function for detector D 1
, ρ
S
is the spatial
distribution of source neutrons and f sp
is the energetic and directional distribution for the source
neutron from spallation, g
n
is the neutron proton gain expressed as the number of neutrons obtained
p
per each source proton.
If Equation (5) is the general equation for an arbitrary source then, the cross-power spectral
density for the proton pulsed beam, as described by equation (2), is:
CPSD
N −1
2
12
(
o n ∑ ∫ 2
2
2 1 2
w
p
n=
0
D
w)
= S g exp( −iwnT
) dw exp( −iw
nT )exp( i(
w + w ) tf ) h(
w ) I ( ) (7)
Fourier Transforming the adjoint Boltzman equation, it is obtained:
⎛ iw + ⎞ D
⎜−
+ L ⎟n
2 (1, w)
( r,
v
C
= Σ
D
)
⎝ v ⎠
Where we can expand 2 (1, w ) , in α modes as follows:
n D C
∑
D2
D2
n ( 1, w)
= T ( w)
Φ (1)
+
Where (1)
is solution of the eigenvalue equation:
Φ m
+ + α
L Φ
m
(1) = − Φ
v
And satisfying the biorthogonality condition:
C
⎛
⎜
⎝
m
m
+
m
+
m
(1)
(8)
(9)
(10)
Φ + p
, Φ ⎟
m
= δ
mp
N
m
(11)
v
⎞
⎟
⎠
833

Substitution of (9) into (8), and use of equation (10) yields:
∑
(2)
( Σ Φ )
D2
D m +
n (1, w)
= −
Φ
m
(1)
m ( iw + α
m
) N
m
From the fundamental mode approximation, we obtain:
Considering (13) the CPSD 12
(w) will be:
∑
(2)
( Σ Φ )
D
D 0 +
n
0 2 ( 1, w)
= −
Φ
0
(1)
m ( iw + α
0
) N
0
N
∑ − 1
D1
( CPSD12
( w))
0
= Sog
n
( exp( −iw(
tf1
− nT )) Wn
1
⎢
p
n= 0 +
Φ Φ ⎣(
(
0
0
,
0
)
v
1
(12)
(13)
⎡ C(
D ⎤
2
)
⎥
α + iw)
(14)
⎦
D1
Where W
n
, is a weight factor that depends on the transfer function of the detector D 1
(source
detector).
∫
D 1 D1
D1
W = dw2
exp( −iw2(
tf1
− nT )) h ( w2
) = h ( tf1
− nT
n
)
(15)
We note that h D1 ( tf nT 1
− ) becomes zero by the causality condition when nT > tf1.
And C(D 2
) depends of the system detector D 2
:
3
∫ d r∫ S
( r´)
∫ dv´
∫
C ρ (16)
(2)
+
( D2 ) = dΩ´
f
sp
( v´,
Ω´)
Σ
D
Φ
0
(1) Φ
0
(1)
Being detector D 2
one of the system detectors and C(D 2
) is a function of the location of this
detector, the source neutron distribution ρ , inside the target, and the direct and adjoint fluxes Φ 0
(1)
S
+
and Φ
0
(1)
. From the phase and amplitude diagram of CPSD 12
(w), we can get the α
0
value and from
this value and the previous calculated Λ value the k eff
. From this methodology we can know the pole
( α
0
) location independently of the source, it means that can be used for all sub-critical systems,
always with k eff
values ranging from 0.94 to 1.
The graphical representation of the Equation (14) is showed on the Figure 1 and with data from
the simulation as we will show later.
834

Figure 1. Amplitude versus frequency and phase versus frequency computed
using Equation (14) for ( CPSD 12
) 0
abs(g12) for 1 pulse from the algorithme
3.50E+01
3.00E+01
2.50E+01
2.00E+01
20*logG12
1.50E+01
abs(g12)
1.00E+01
5.00E+00
0.00E+00
1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
frequency
phaseg12 for 1 pulse from the algorithme
0.00E+00
1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
-2.00E-01
-4.00E-01
-6.00E-01
20*logG12
-8.00E-01
-1.00E+00
faseg12
-1.20E+00
-1.40E+00
-1.60E+00
-1.80E+00
frequency
3. LAHET + MCNP-DSP
The coupling of both Monte Carlo codes provide an estimator of the time-dependent response,
that can be used to design a sub-criticality monitoring system. The LAHET code provides a neutron
or/and gamma source to the MCNP-DSP and the final output contains the correlated detector
responses in the time or frequency domains.
As described in Valentine et al. [5], the LAHET file is read by MCNP-DSP as the source for the
subsequent calculation, provided the spallation source given at the interaction of the proton beam on
the target. MCNP-DSP uses this information to determine the location of the particles in the
MCNP-DSP model. MCNP-DSP simulates the interaction for neutrons with energies less than
20 MeV and also the neutron detection.
835

With the coupling LAHET + MCNP-DSP, we have the capability to simulate a pulsed source
which correspond to the conceptual design of the ADS. The proton pulse frequency will be limited by
the measurements set-up frequency given by the maximum frequency and the number of bins per
block.
4. LAHET + MCNP-DSP simulator of ADS
Using LAHET + MCNP-DSP we can obtain the CPSD 12
(w) function between two detectors. The
detectors location will be given by the users and the CPSD 12
(w) will be dependent on the detectors
location. This dependence is related with the higher sub-critical modes and will be more important
when the system becomes more sub-critical.
Figure 2 was obtained using one proton source per data block as was done on Figure 1, to
compare between both figures. The fundamental mode approximation done on Equation (14) is in
good agreement with the calculated CPSD 12
(w), where all the higher modes are included.
Figure 2. Amplitude versus frequency and phase versus frequency
computed with the LAHET&MCNP-DSP
40
Amplitude G12 for keff = 0.94346 and 1 pulse
30
20
20*logG12
10
0
1.00E+0 1.00E+0 1.00E+0 1.00E+0 1.00E+0 1.00E+0 1.00E+0
-10
-20
-30
frequency
2
Phase G12 for Keff = 0.94346 and 1
1.5
1
0.5
phase (rad)
0
1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
-0.5
-1
-1.5
-2
frequency
836

As suggested by Uhrig [6], a method to obtain the k eff
, values was from the plot of the crosspower
spectral density versus the frequency. The simpler way is to find out the value of the eigenvalue
α
0
, from the break frequency value. The break frequency value for a single pole, can be obtained
from the plot of the phase ( Φ
12
( w)
) versus frequency, looking for the frequency which has a phase
equal to ( − π ). This method can be applied to the simulations performed with LAHET/MCNP-DSP
4
by Y. Rugama et al. [7] see Figure 2.
The simulations performed with LAHET + MCNP-DSP confirm that for the case of one
accelerator proton pulse per data block, one obtain the correct value of the eigenvalue from the
CPSD 12
(w) module and phase vs. frequency plots and then from the break frequency f
b
in Hz. The
eigenvalue is given by:
α0 = 2πf b
(17)
The multiplication constant of the system can be obtained from the expression:
α 0
Keff −1
keff
=
Λ
(18)
Where Λ is obtained from some previous calculation or determination. The mean generation
time for the 233 U/ 232 Th FEA cooled by liquid metal, from a previous calculation, was equal
to: Λ = 8.25846 × 10 -7 s.
We have checked that this method gives the correct value of the multiplication constant k eff
with
an acceptable error. For instance from the phase diagram obtained with LAHET + MCNP-DSP we get
a value of the Keff = 0. 985983 while with MCNP-4A, the K eff
from the sub-critical systems was
equal to 0 .96627 ± 0. 00067 .
The α
0
eigenvalue of the system for the fundamental mode, so it will be independent of the
numbers of proton pulses per block. In this study the data will be given using one pulse per data block
because as we can observe in Figure 3. The pole location using 1 pulse per data block or 5 pulses is at
the same point, but only for 1 pulse, the frequency at the phase equal to − π will give us the exactly
4
value for the break frequency.
837

Figure 3. Comparison for 1 and 5 pulses per block on amplitude and phase
comparison CPSD12 for 1 pulse and 5 pulses
40
30
20
10
20*logCPSD12
0
1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
-10
1 pulse
5 pulses
-20
-30
-40
-50
(
comparison CPSD12 phase for one and five pulses
2
1.5
1
0.5
phase (rad)
0
1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
1 pulse
5 pulses
-0.5
-1
-1.5
-2
frequency
5. Conclusions
We have analysed in this paper a method to measure the multiplication constant of the FEA
system. The method can be used with the accelerator proton beam turned on and is based on the plots
of Φ ( ) versus frequency. We have showed that this CPSD function has a single pole at the first
12
w
eigenvalue of the system which is related to the sub-criticality value of the fast energy amplifier. The
simulations performed with the LAHET and MCNP-DSP codes showed that we can obtain the subcriticality
value directly from the plot of the phase of Φ
12
( w)
versus frequency, looking for the
frequency value which has a phase equal to − π . The calculations that we have performed give the
4
correct value of the system sub-criticality with very good precision when we have one proton pulse per
block (each block is formed by 512 or 1 024 bins depending on the data acquisition systems). We have
838

observed that the precision of the method becomes poor when the number of protons pulses per block
increased.
Also we have also showed in this paper that both, theoretical and simulated results for the
amplitude of the CPSD i.e. Φ
12
( w)
in decibels versus frequency agree pretty well. This prediction is
very good in spite of the complexity of the simulation results. This fact tells us that the physical bases
of the theoretical results are well established.
REFERENCES
[1] J.L. Muñoz-Cobos, R. Perez, T. Valentine, Y. Rugama, J. Mihalczo, A Stochastic Transport
Theory of Neutron Photon Coupled Fields: Neutron and Photon Counting Statistics in Nuclear
Assemblies, (2000), Annals of Nuclear Energy 27 1087-1114, (2000).
[2] C. Rubia, J.A. Rubio, S. Bruno, F. Carminati, N. Fietier, J. Galvez, C. Geles, Y. Kadi,
R. Klapisch, P. Mandrillon, J.P. Reval, Ch. Roche, Conceptual Design of a Fast Neutron
Operated High Power Energy Amplifier, CERN/AT/95-44(ET), (1995).
[3] J.L. Muñoz-Cobos, R. Perez, G. Verdu, Neutron Stochastic Transport Theory, (1987), Nuclear
Science and Engineering, Vol. 95, p. 83-105, (1987).
[4] G. Verdu, J.L. Muñoz-Cobos, J.T. Mihalczo, W.T. King, Trans. Am. Nucl., Soc.,42, 452 (1984).
[5] T. Valentine, Y. Rugama, J.L. Muñoz-Cobos, R. Perez, Coupling MCNP-DSP and LAHET
Monte Carlo Codes for Design Sub-critical Monitors for Accelerator Driven System (2000).
Proceedings of MC2000 (Monte Carlo 2000, Lisbon), in press.
[6] R.E. Uhrig, Random Noise Techniques in Nuclear Systems, Ronald Press, New York, (1970).
[7] Y. Rugama, J.L. Muñoz-Cobos, T. Valentine, Noise Method for Monitoring the Sub-criticality
in Accelerator Driven Systems, Proceedings MC 2000 (Monte Carlo 2000, Lisbon), in press.
839

MOLTEN SALTS AS POSSIBLE FUEL FLUIDS
FOR TRU FUELLED SYSTEMS: ISTC #1606 APPROACH
Victor Ignatiev 1 , Raul Zakirov 1 , Konstantin Grebenkine 2
RRC-Kurchatov Institute, 123182, Moscow, Russian Federation
2
VNIITF, 456770, Snezinsk, Russian Federation
1
Abstract
The principle attraction of the molten salt reactor (MSR) technology is the use of fuel/fertile material
flexibility (easy of fuel preparation and processing) for gaining additional profits as compared with
solid materials. This approach presents important departures from traditional philosophy, applied in
current nuclear power plants, and to some extent contradicts the straightforward interpretation of the
defence-in-depth principal. Nevertheless we understand there may be potential to use MSR
technology to support back end fuel cycle technologies in future commercial environment.
The paper aims at reviewing results of the work performed in Russia, relevant to the problems of
MSR technology development. Also this contribution aims at evaluation of remaining uncertainties
for molten salt burner concept implementation. Fuel properties & behaviour, container materials, and
clean-up of fuels with emphasis on experiments will be of priority. Recommendations are made
regarding the types of experimental studies needed on a way to implement molten salt technology to
the back-end of the fuel cycle.
To better understand the potential and limitations of the molten salts as a fuel for reactor of
incinerator type, Russian Institutes have submitted to the ISTC the Task #1606 Experimental Study of
Molten Salt Technology for Safe and Low Waste Treatment of Plutonium and Minor Actinides in
Accelerator Driven and Critical Systems. The project goals, technical approach and expected specific
results are discussed.
841

1. Introduction
Last years important R&D efforts were placed worldwide to find the ways to reduce the long
term radionuclide inventory resulting from the nuclear power generation. This approach calls for the
introduction of some innovative technologies to overcome some of technical hurdles presented by
traditional ones. In our understanding, introduction of the innovative reactor concept to back end of
fuel cycle pursues the following goals:
• Reduced actinides total losses to waste.
• Low plutonium and minor actinides total inventory in the nuclear fuel cycle.
• Minimal 235 U support.
• Minimal neutron captures outside actinides.
Within this context the general matrix describing major innovative reactor & fuel cycle options
could be written as follows:
⏐Dedicated TRU burners; Once-through fuel cycle ⏐
⏐TRU-free fuel cycle system; Recycling of actinides⏐
The use of the molten salts as the fuel material has been proposed for many different reactor
types and applications. Molten salt fuelled reactor (MSR) concepts have been prepared for fast
breeders and thermal reactors more particularly in the USA, Russian Federation, France and Japan.
Though, molten salt nuclear fuel concept has been proven by the successful operation experimented
in MSRE experimental reactor at ORNL [1,2], this approach has not been implemented in industry.
The fuel chosen for the operation of MSRE and for subsequent reactors of this type was a mixture of
7
LiF–BeF 2
–ZrF 4
(ThF 4
)–UF 4.
In Russia, the MSR programme started in the mid-seventies. RRC-Kurchatov Institute (KI) was a
basic organisation which supervised a collaboration (an expert group composed by) of specialised
institutions. A reduction of activity appeared after 1986 due to Chernobyl accident as well as a
general stagnation of nuclear power and nuclear industry. Then at the end of the eighties there was an
increase of conceptual studies as a result of the interest to the inherent safe reactors of a new
generation. An extensive review of MSR technology developments at RRC-KI through 1989 is given
in the publication in reference [3].
Today’s interest in MSR technology stems mainly from an increased fuel residence time in the
system, reduced actinides mass flow rate and relatively low waste stream when purifying and
reconstituting the fuel by pyrochemical methods. We could then expect that in the future the MSR
technology could find a role in symbiosis with standard reactors in the management of TRUs and
thorium utilisation. New MSB concept (a reactor of incinerator type) requires a reconsideration of
prior MSR concept, including optimisation of the neutron spectra and power density in the core,
selection of the salt composition and new approaches towards its behaviour control and clean-up.
Below, we try to give our understanding of key issues, remaining uncertainties, methods
available and improvements to integrate P&T recycle in molten salts. Also, planned tests for three
years within the ISTC Task #1606 Experimental Study of Molten Salt Technology for Safe and Low
Waste Treatment of Plutonium and Minor Actinides in Accelerator Driven and Critical Systems will
be discussed.
842

2. The fuel salt for MSB concept
Many chemical compounds can be prepared from several “major constituents”. Most of these,
however, can be eliminated after elementary consideration of the fuel requirements. Consideration of
nuclear properties leads one to prefer as diluents the fluorides of Be, Bi, 7 Li, Pb, Zr, Na, and Ca, in
that order. Simple consideration of the stability of these fluorides towards reduction by structural
metals eliminates, however, the bismuth fluorides (see Table 1).
Several fluorides salts satisfy the characteristic properties of the thermal stability, radiation
resistance, low vapour pressure and manageable melting point. To achieve lower melting
temperatures, two or more salts are combined to produce still lower melting mixtures.
Table 1. Thermodynamic properties of fluorides
Compound
(solid state)
-∆G f,1000 ,
kcal/mole
-∆G f,298 ,
kcal/mole
E 0 298, V
(Me|F 2
)
LiF 125 140 6.06
CaF 2
253 278 6.03
NaF 112 130 5.60
BeF 2
208 231 5.00
ZrF 4
376 432 4.70
PbF 2
124 148 3.20
BiF 3
159 200 2.85
NiF 2
123 147 3.20
UF 3
(UF 4
) 300 (380) 330 (430) 4.75
PuF 3
(PuF 4
) 320 360 (400) 5.20
ThF 4
428 465 5.05
AmF 3
325 365 5.30
CeF 3
345 386 5.58
LaF 3
348 389 5.63
Note, that ZrF 4 , as a part of basic solvent, was found to distil from melt and condense on cooler
surfaces in the containment system [1]. Control of the ZrF 4
mass transport was considered too
difficult to ensure, so the 2LiF-BeF 2
solvent system was chosen as basic option at ORNL and later at
RRC-KI. Also, in order to minimise problems associated with chemical treatment of the fuel salt and
associated reduction of the basic components, the priorities should be given to the system with lowest
possible ZrF 4
and PbF 2
content. Note, that use of Zr and Pb, instead e.g. the sodium in the basic
solvent will lead to the increased generation of the long-lived activation products in the system [4].
Trivalent plutonium and minor actinides are stable species in the various molten fluoride salts [5,6].
Tetravalent plutonium could transiently exist if the salt redox potential was high enough. But for
practical purposes (stability of potential container material) salt redox potential should be low enough
and corresponds to the stability area of Pu(III). PuF 3
solubility is maximum in pure LiF or NaF and
decreases with addition of BeF 2
and ThF 4
. Decrease is more for BeF 2
addition, because the PuF 3
is not
soluble in pure BeF 2
.The solubility of PuF 3
in LiF-BeF 2
and NaF-BeF 2
solvents is temperature and
composition dependent and PuF 3
solubility seems to be minimal in the “neutral” melts. For last ones, it
843

eaches about 0.5% mole at a temperature of 600°C and increases to about 2.0% mole at 800°C. For
some excess free fluoride solvent system Li, Be, Th/F (72-16-12% mole), studies indicated that
solubility of PuF 3
increased from 0.7-0.8% mole at 510°C to 2.7-2.9% mole at 700°C.
The other TRU species are known to dissolve in Li, Be, Th/F solvent, but no quantitative
definition of their solubility behaviour exists. Such definition must of course be obtained, but the
generally close similarity in behaviour of the AnF 3 makes it most unlikely that solubility of this
individual species could be a problem. As expected substitution of a small quantity of AnF 3 scarcely
changes the phase behaviour of the solvent system.
The trifluoride species of AnF 3 and the rare earth’s are known to form solid solutions so, that in
effect, all the LnF 3 and AnF 3 act essentially as a single element. It is possible, but highly unlikely, that
the combination of all trifluorides, might exceed this combined solubility at the temperature below
the reactor inlet temperature. A few experiments must be performed to check this slight possibility.
The solubilities of the AnO 2 in Li, Be, Th/F are low and well understood. Plutonium as PuF 3 shows
little tendency to precipitate as oxide even in the presence of excess BeO and ThO 2 . The solubility of
the oxides of Np, Am, Cm has not been examined. Some attention to this problem will be required.
The molten fluoride chemistry (solubility, redox chemistry, chemical activity, etc.) for the
2LiF-BeF 2
system is well established and can be applied with great confidence, if PuF 3
fuels are to be
used in the 2LiF-BeF 2 solvent. The properties of the most developed 2LiF-BeF 2
solvent however, are
not all near the optimum for MSB application (very limited PuF 3 solubility, high enough melting
point, etc.). Alternative solvent composition which will meet the lower liquidus temperature and
increased PuF 3
solubility may be chosen e.g. from Na,Li,Be/F system. However, new less understood
solvents system must be considered carefully before application in order to avoid severe problems
with process operation. For MSB’s needs next more important is consideration of PuF 3
chemical
behaviour in these solvent systems: PuF 3 solubility in Li, Be/F, Na, Be/F and Li, Na, Be/F solvent,
oxide tolerances of such mixtures and redox effects of the fission products.
The specific physical properties which were measured within the Russian MSR program include
density, heat capacity, heat of fusion, viscosity, thermal conductivity and electrical conductivity.
Particular emphasis has been placed for U/Th fuelled reactor cores. Properties are in each case
adequate to the proposal service. Many of properties required for the MSB concept development are
estimates rather than measured values. In some cases, especially for alternative solvents, careful remeasurement
of some properties (e.g. thermal conductivity) is reasonable and desirable.
3. Fission products clean-up
For molten salt fuels, fission products could be grouped in the three broad classes: 1) the soluble
at salt redox potential fission products, 2) the noble metals and 3) the noble gases. The MSR would
manage the noble gas removal by sparging with helium. As it was mentioned before, the problem here
is to prevent the xenon from entering the porous graphite moderator. For the noble metals the
situation is not so good and more experimental efforts is required in order to control their
agglomeration, adhesion to surfaces and transport in purge gas.
In MSRs, from which xenon and krypton are effectively removed, the most important fission
products poisons are among lanthanides, which are soluble in the fuel. Also, in combination of all
trifluorides, AnF 3
solubility in the melt is decreased by lanthanides accumulation. Since plutonium
844

and minor actinides must be removed from the fuel solvent before rare earth’s fission products the
MSR must contain a system that provides for removal of TRUs from the fuel salt and their
reintroduction to the fresh or purified solvent.
The available thermodynamic data (calculated or measured) for An/Ln trifluorides include
considerable uncertainties and dispersion. For example, estimations on An-Ln separation ability have
shown the most favourable situation for fluoride melts in comparison with chloride ones [7]. The
similar conclusion was obtained in paper [8]. From our point of view, the comparison of
thermodynamic potentials given in paper [9], for actinides and lanthanides both in chloride and
fluoride systems for the benefit of chlorides is not correct enough, as the values of Gibbs energy for
fluorides of lanthanides are underestimated. All mentioned above specify a necessity for further
measurements of thermodynamic potentials of Ln and An fluorides by an uniform technique (e.g.
EMF method with solid fluorine ion conducting diaphragms).
A number of pyrochemical processes (reductive extraction, electrochemical deposition,
precipitation by oxidation and their combinations) for removing the soluble fission products from the
fluoride based salt have been explored during the last years. Studies of the full scale MSB fuel salt
chemical processing system are not as far advanced, but small scale experiments lead to optimism,
that a practicable system can be developed.
3.1 Reductive extraction
Selective extraction from molten fluoride mixtures into liquid metals have been studied in details
for essentially all pertinent elements [2]. This method of processing is optimal from technological
perspective. It allows to realise on-line processing of fuel composition using simple design of
extractors. The process capacity is rather high, and can be easily enlarged by intermixing. Obviously,
use of the metal transfer system essentially simplify and accelerates the process, but several stages are
required to reach desired recovery.
Rare earth removal unit based on Bi-Li reductive extraction flow-sheet developed in ORNL for
LiF-BeF 2
solvent system could provide negligible losses of TRU (about 10 -4 ) by use of several counter
current stages. The separation factors Θ of AnF n and LnF n are approximately equal to 10 3 for the Li,
Be/F and Li, Na,K/F solvents. These values are very convenient for the lanthanide’s separation by the
reductive extraction. However, when the melt is complicated by the addition of large quantities of
ThF 4 the situation becomes considerably less favourable. Under the conditions used U and Zr are the
most easily reduced of the species shown above. U and Pa should be easily separated under the proper
conditions, Pu-Pa separation is possible.
Note the following drawbacks of reductive extraction as applied to MSB:
• Less favourable scheme of An and Ln separation due to decreased difference in
thermodynamic potentials of An and Ln (alloys with liquid metal).
• Materials compatibility pose substantial problems.
• Poor separation of thorium from rare earths for fluoride system; it can be made by use of LiCl.
• Changes in fuel composition because of the significant amount of lithium required; rare
earths are removed in separate contactors in order to minimise the amount of Li required.
845

3.2 Precipitation of oxides
Although the metal-transfer process appears to give the best fuel salt purification in case of
processing system with relatively short cycle (10-30 days), there are other possibilities for rare earth
removal that are perhaps worth keeping in mind. If a bismuth containing system proves expensive or
if unseen engineering difficulties (e.g. material required) develop, the other methods may be
applicable, especially at longer processing cycle times. If treatment of a MSB fuel on a cycle-time of
100 days or more is practicable, such an oxide precipitation might be used for periodic removal of
rare-earths. In experiments [10] a successful attempt was made to precipitate mixed uranium,
plutonium, minor actinides and rare earths from LiF-NaF molten salt solution by fluor-oxide
exchange with other oxides (for example CaO, Al 2 O 3 , etc.) at temperatures 700-800°C. The rare earths
concentration in the molten salt solution was about 5-10 mole%. It was found the following order of
precipitation in the system: U-Pu-Am-Ln-Ca. Essentially all U and TRU were recovered from the
molten salt till to rest concentration 5 × 10 -4 %, when rare earths still concentrated in solution.
Main advantage of a method of processing the fuel composition by a sequential sedimentation of
its components by oxides, non-soluble in fluoride melts, is the simplicity of the equipment for
processing unit and more acceptable corrosion of structural materials in comparison with reductive
extraction. Note, that recovery of oxide precipitates from a molten salt need further development.
3.3 Electrochemistry
Main advantage of electrochemical methods is a possibility for the fuel clean up without
introduction to the melt of additional reagents, which could change salt composition and influence its
chemical behaviour and properties. Some processing flow-sheets with electrochemical deposition on
solid electrodes, at first of all TRU elements, and after that the fission products, by the same way, or
as alternative, for example, by oxides precipitation are possible. The reintroduction TRU in molten
salt could be carried out electrochemically, or by simple dissolution, for example, of HF use.
In principle, An/Ln separation on solid electrodes could be more attractive in comparison with
liquid electrodes. For the first case the overall reaction is the following:
and the latter:
Ln + AnF 3
= An + LnF 3
Bi(Ln) + AnF 3
= Bi(An) + LnF 3
In first case equilibrium in the separating process is carried out at the maximum value of a
difference of thermodynamic potentials for actinides and lanthanides.
Regarding technological aspects, the electrochemical method has the important advantage
consisting in the possibility of a continuous quantitative control of the process. During the process its
rate and also depth of fuel processing are set and controlled by the value of a potential on electrodes
of an electrolytic bath.
Note the following drawbacks of this method:
• Space limitation on processes area (bath electrodes), decreased capacity of units as
contrasted to chemical processes in volume, especially for the end phase of the process.
• Necessity of dendrites control when use solid electrodes.
846

4. Container material studies
4.1 Fuel and coolant circuits
An important part of Russian MSR program dealt with the investigation of the container
materials [3,11,12]. The development of domestic structural material for MSR was substantiated by
available experience accumulated in ORNL MSR programme on nickel based alloys for UF 4
containing salts [1,2]. In addition, compatibility tests were conducted to re-examine the possibility of
using iron based alloys as container materials for MSRs. These alloys are more resistant to tellurium
penetration and generate less helium under irradiation than Ni based alloys. However, their use would
limit the redox potential of the salt and operating temperature in MSR.
Corrosion resistance of materials was studied in RRC KI by two methods. The first is the method
of capsule static isothermal test of reference specimens in various molten salt mixtures. Also, flibe,
flinak and sodium fluoroborate eutectic salts have been circulated for thousands of hours in natural
and forced convection loops constructed of iron and nickel based alloys to obtain data on corrosion,
mass transfer, and material compatibility. Not only normal, but also lowered and high oxidation
conditions were present in the loops.
The alloy HN80MT was chosen as base. Its composition (in wt.%) is Ni(base), Cr(6.9), C(0.02),
Ti(1.6), Mo(12.2), Nb(2.6). The development and optimisation of HN80MT alloy was envisaged to be
performed in two directions:
• Improvement of the alloy resistance to a selective chromium corrosion.
• Increase of the alloy resistance to high temperature helium embrittlement and to tellurium
intergranular cracking.
The results of combined investigation of mechanical, corrosion and radiation properties various alloys
of HN80MT permitted to suggest the Ti and Al-modified alloy as an optimum container material for the
MSR. This alloy named HN80MTY (or EK–50) has the following composition (in wt.%): Ni(base),
Fe(1.5), Al(0.8–1.2), Ti(0.5–l), Mo(11–12), Cr(5–7), P(0.015), Mn(0.5), Si(

potentials that must be maintained to avoid intergranular cracking for nickel-based alloys. Techniques
developed under other reactor programs to improve the resistance of stainless steels to helium
embrittlement should be extended to include nickel-base alloys.
4.2 Fission product clean up unit
The materials required for fission product clean up unit depend of course, upon the nature of the
chosen process and upon the design of the equipment to implement the process. For MSB the key
operation in the fuel treatment is removal of rare earth, alkali-metal, alkaline-earth fission products
from the fuel solvent before its return, along with the TRUs, to the reactor. The crucial process in
most of the processing vessels is that liquids be conducted to transfer selected materials from one
stream to the another.
Such a fission product cleanup unit based on metal transfer process at least will present a variety
of corrosive environments, including:
• Molten salts and molten alloys containing e.g. bismuth, lithium or other metals at 650°C.
• HF-H 2
mixtures and molten fluorides, along with bismuth in some cases, at 550-650°C.
• Interstitial impurities on the outside of the system at temperature to 650°C, particularly if
graphite and refractory metals are used.
Certainly, the R&D on the materials for the fission product clean up unit for MSB is at very early
stage. A layer of frozen salt will probably serve to protect surfaces that are worked under oxidising
conditions, if the layer can be maintained in the complex equipment. RRC-KI preliminary tests at
molten salt loops [11,12] showed that the thickness of the frozen film on the wall was predictable and
adhered to the wall.
The only materials that are truly resistant to bismuth and molten salts are refractory metals (W,
Mo, Ta) [1] and graphite (e.g. graphite with isotropic pyloric coating tested in RRC-KI [12] for both
working fluids), neither of which is attractive for fabricating a large and complicated system.
Development work to determine if iron base alloys can be protected with refractory metal coatings
should probably be considered for higher priority. The approach taken to materials development could
be to initially emphasise definition of the basic material capabilities with working fluids and
interstitial impurities, and then to develop a knowledge of fabrication capabilities.
5. ISTC supported R&D
To solve some of the mentioned in previous sections essential issues, Russian Institutes (RFNC-
All-Russian Institute of Technical Physics, (Chelyabinsk-70), RRC-Kurchatov Institute (Moscow),
Institute of Chemical Technology (Moscow) and Institute of High Temperature Electrochemistry
(Ekaterinburg)) have submitted to the ISTC the Task #1606 Experimental Study of Molten Salt
Technology for Safe and Low Waste Treatment of Plutonium and Minor ActinidesiIn Accelerator
Driven and Critical Systems.
The general purpose of the project is to perform an integral evaluation of potential of the selected
molten salt fuel as applied to safe, low-waste and proliferation resistant treatment of radwaste and Pu
848

management. The major developments that will be pursued in the framework of the project are the
following:
• Experimental study of behaviour and fundamental properties of prospective molten salt
compositions.
• Experimental verification of candidate structural materials for fuel circuit.
• Reactor physics and fuel cycle consideration and recommendations on the key points of
MSB concept development.
First two objectives are considered as the most crucial to the MSB further development. There
are no doubts that the candidate constituents of the solvent system for MSB concept must be LiF,
NaF, BeF 2 , maybe with minor additions of some other fluorides in order to provide required fuel
properties. In contrast with well-established 2LiF-BeF 2 solvent system, for the other prospective
compositions there is essential uncertainty in fundamental data, necessary to estimate their potential
for MSB fuelled by plutonium and minor actinides.
Some of the salt components required for the experiments are commercially available and will be
purchased, but some other salts have to be prepared basing on the technology developed by the
participating institutes. Particularly, Pu, Np and Am trifluorides will be prepared at VNIITF site by
pyrochemical method of HF gas treatment. Although starting materials of very high purity will be
used in production of the fluorides, a careful analysis and, in some cases, a further purification will be
needed before the salts usage in cell and loop systems. Two steps of purification would be required:
one for the removal of oxides and sulfides and one for the removal of structural metal fluorides (NiF 2 ,
FeF 2
, CrF 2
, etc.)
The measurements of basic properties of prospective molten salt fuels will be carried out by means
of techniques, which are mastered or will be developed in the participating institutions. First, it is
planned to evaluate a phase diagram for the selected solvent and perform its experimental verification on
several points. Also, it is planned to measure molten salt properties, such as actinides/lanthanides
solubility, viscosity, thermal conductivity, standard potentials of actinides and rare-earth fluorides and
equilibrium distribution coefficients of lanthanides in the two-phase system. Heat capacity, vapour
pressure and density of the selected mixture can be calculated with a reasonable accuracy.
Measurements of the properties will be carried out in a range of temperatures from 350°C to 800°C.
Two main types of candidate structural materials for the MSB primary circuit will be tested. First
ones are samples of the Ni-based alloy Hastelloy-N, modified according to the ORNL and RRC-KI
recommendations, the second will be prospective stainless steel samples. The corrosion tests will be
carried out with selected fuel salt under conditions simulating the design ones with a working
temperature up to 700-750°C, fuel salt heat up about 100°C, as well as with additions simulating the
main fission products. The tests will be carried out at thermal convection loops with the samples
exposure time in a flow till to 1 000 hours. It is planned to develop a technique for redox potential
control and monitoring and apply it at the corrosion test loops. After the samples exposition, their
detailed examinations will be carried out and corrosion rate will be measured. On the basis of these
studies, the conclusion about the candidate structural materials compatibility with the selected salt
will be made.
The experimental data will be fed into the conceptual design efforts. The objectives of the
conceptual studies are to consider a candidate flow-sheet for MSB concept, that would be feasible.
849

Based on experimental data received for the project, the recommendation on choice of the fuel
composition, the core configuration and the fuel cycle parameters, will be done.
The specific expected results of the Task #1606 will be:
1. Identification of the place of the molten salt technology in future fuel nuclear power system
and suggestions on the strategy of its implementation for the fuel cycle harmonisation.
2. Recommendations on the choice of fuel compositions, core configuration and fuel cycle of
the MSB fuelled by Pu and MA.
3. Measurements of the key properties of the selected molten salt fuel composition. A whole set
of experimental facilities will be created. It allows continuing examinations of other
candidate fuel compositions according to requests of the foreign collaborators.
4. Tests of candidate structural materials in corrosion loops and verification the materials
compatibility with the selected fuel composition.
5. Fission product clean up feasibility study, including experimental studies of basic data for
the pyrochemical processes as applied to MSB needs.
In March 2000 the Governing Board of the ISTC has approved Task #1606 for financial support.
Funding source is EU. In the meantime the work plan for the Task #1606 has been developed.
Numerous comments and suggestions of foreign collaborators have been taken into account.
6. Summary
It is obvious from the discussion above that the use of molten fluorides as fuel and coolant for a
reactor system of energy production and incinerator type faces a large number of formidable
problems. Several of these have been solved, and some seem to be well on the way to be solved. But
it is also clear that some still remain unsolved. The molten salts have many desirable properties for
such applications, and it seems likely that – given sufficient development time and money – a
successful TRU free or burner system could be developed. The properties of the MSR basic option
salts however, are not all near the optimum for MSB applications. Only performing some additional
experimental work give us the possibility to understand the practicability of operating an MSB.
It is still too early to guarantee that a MSR could truly operate as a reliable long term incinerator
of TRUs and producer of energy in U-Pu or U-Th fuel cycle. It may even be uncertain whether such a
system would serve a useful purpose if its successful development were assured. It is certain that
effort to date has thrown light on e. g. much elegant high temperature non-aqueous chemistry and has
shown how molten salts can operate under hard and strong conditions. Finally, it opens perspectives
significantly different to the present reactor and fuel cycle technology.
Acknowledgements
The authors acknowledge the ISTC for it’s financial support. They also thank the staff members
for their friendly assistance, which was very much appreciated.
850

REFERENCES
[1] H.J. MacPherson, Development of Materials and Systems for the Molten-salt Reactor Concept,
Reactor Technology, Vol. 15, No. 2, (1972).
[2] J.R. Engel et al., Development Status and Potential Program for Development of Proliferationresistant
Molten Salt Reactors, ORNL/TM-6415, March, (1979).
[3] V.M. Novikov, V.V. Ignatiev, V.I. Fedulov, V.N. Cherednikov, Molten Salt Nuclear Energy
Systems – Perspectives and Problems, Energoatomizdat, Moscow, (1990), (in Russian).
[4] V.I. Oussanov et al., Long-lived Residual Activity Characteristics of Some Liquid Metal
Coolants for Advanced Nuclear Energy Systems, In Proc. of the Global’99 international
conference, Jackson Hole, USA, (1999).
[5] C.J. Barton, J.D. Redman, R.A. Strelow, J. Inorg. Chem., 1961, Vol. 20, No. 1-2, p. 45.
[6] V.F. Afonichkin et al., Interaction of Actinide and Rare-earth Element Fluorides with Molten
Fluoride Salts and Possibility of their Separation for ADTT Fuel Reprocessing, In Proc. of the
second international conference on ADTTA, Kalmar, Sweden, 1996, June 3, pp. 1144-1155.
[7] R.Y. Zakirov, V.N. Prusakov, Role of Electrochemistry for Fuel Processing in Molten Salt
Reactors, Pre-print IAE-6061/13, RRC-KI, Moscow, (1998).
[8] F. Lemort, X. Deschanels, R. Boen, Application of Pyrochemistry to Nuclear Waste Processing,
In Proc. of the Global’99, Jackson Hole, USA, (1999).
[9] C. Pernel et al., Partitioning of Americium Metal from Rare Earth Fission Products by
Electrorefining, In Proc. of the Global’99, Jackson Hole, USA, (1999).
[10] V.F. Gorbunov et al., Experimental Studies on Interaction of Plutonium, Uranium and Rare
Earth Fluorides with Some Metal Oxides in Molten Fluoride Mixtures, Radiochimija, 1976,
Vol. 17, pp. 109-114, (in Russian).
[11] V.V. Ignatiev, V.M. Novikov, A.I. Surenkov, V.I. Fedulov, The State of the Problem on
Materials as Applied to Molten-salt Reactor: Problems and Ways of Solution, Pre-print IAE-
5678/11, Moscow, (1993).
[12] V.V. Ignatiev, V.M. Novikov, A.I. Surenkov, Molten Salt Test Loops (In and Out Reactor
Experimental Studies), Pre-print IAE-5307/4, Moscow, (1991).
[13] V.V. Ignatiev, K.F. Grebenkine, R.Y. Zakirov, Experimental Study of Molten Salt Reactor
Technology for Safe, Low-waste and Proliferation Resistant Treatment of Radioactive Waste and
Plutonium in Accelerator-driven and Critical Systems, In Proc. of the Global’99 international
conference, Jackson Hole, USA, (1999).
851

COMPARATIVE ASSESSMENT OF THE TRANSMUTATION EFFICIENCY
OF PLUTONIUM AND MINOR ACTINIDES IN FUSION/FISSION HYBRIDS AND ADS
Marcus Dahlfors
Department of Radiation Sciences, Uppsala University
Uppsala, Sweden
Yacine Kadi
Emerging Energy Technologies, CERN, European Organisation for Nuclear Research
Geneva, Switzerland
Abstract
A preliminary comparative assessment relevant to the transmutation efficiency of plutonium and
minor actinides has been performed in the case of ANSALDO’s Energy Amplifier Demonstration
Facility based on molten lead-bismuth eutectic cooling, classical MOX-fuel technology and operating
at 80 MW th
. The neutronic calculations presented in this paper are a result of a state-of-the-art
computer code package, EA-MC, developed by C. Rubbia and his group at CERN. Both high-energy
particle interactions and low-energy neutron transport are treated with a sophisticated method based on
a full Monte Carlo simulation, together with modern nuclear data libraries. Detailed Monte Carlo
transport calculations were performed for different types of external neutron sources: D-D and D-T
fusion sources and proton induced spallation neutron sources. The fuel core was described on a pinby-pin
basis allowing for detailed scans of the main neutronic properties, e.g. neutron flux spectra and
power density distributions.
853

1. Introduction
The Energy Amplifier Demonstration Facility (EADF) [1] is a hybrid system designed to be
driven by an external source. The present design utilises a proton induced spallation neutron source for
providing the external neutrons, but as such, the system is not limited to any particular choice of
source as long as neutrons of suitable energy are provided. In this study, the neutronic properties of the
reference spallation source driven EADF system has been compared to those of the system driven by
two different alternative neutron sources: D-D and D-T fusion sources.
2. The EA-MC code package
EA-MC is a general geometry, “point-energy”, Monte Carlo code which stochastically calculates
the distribution of neutrons in three-dimensional space as a function of energy and time. The neutron
data are derived from the latest nuclear data libraries [2]: ENDF/B-VI 5 (USA), JENDL-3.2 (Japan),
JEF-2.2 (Europe), EAF-4.2 (Europe), CENDL-2.1 (China), EFF-2.4 (Europe) and BROND-2 (Russia).
The general architecture of the EA-MC code is shown in Figure 1. The geometrical description is
first automatically translated into FLUKA’s combinatorial geometry, and the high-energy particle
transport is carried out [3,4]. Neutrons are transported down to 20 MeV and then handed over to EA-
MC which continues the transport.
Figure 1. General architecture of the EA-MC simulation of
neutron transport and element evolution
The EA-MC code is designed to run both on parallel and scalar computer hardware. Having used
standard language elements, the code can be implemented on different systems. A common
initialisation section is followed by a parallel phase where every CPU runs an independent simulation
with the same initialisation data. A parallel analysis program collects the results and calculates the
standard deviation among the different CPUs. This gives an estimate of the statistical fluctuations [5].
3. The Energy Amplifier Demonstration Facility
The key objective of the Energy Amplifier Demonstration Facility [1], in a first approximation,
aims at demonstrating the technical feasibility of a fast neutron operated accelerator driven system
cooled by molten lead-bismuth eutectic (LBE) and in a second phase that of incinerating TRUs and
long-lived fission fragments while producing energy.
854

3.1 Reference configuration
As in the case of the Energy Amplifier’s conceptual design [6], the EADF core consists of an
annular structure immersed in molten lead-bismuth eutectic which serves as primary coolant and
spallation target (Figure 2). The central annulus contains the spallation target unit which couples the
proton accelerator to the sub-critical core. The core is arranged in a honeycomb-like array forming an
annulus with four coaxial hexagonal rings of fuel sub-assemblies. The fuel core is itself surrounded by
an annular honeycomb-like array of four rings of dummy sub-assemblies, which are essentially empty
ducts. The detailed description of the EADF reference configuration can be found in [1].
Figure 2. Schematic view of the reactor system assembly [1]
The coupling of the accelerator system to the sub-critical core is realised via the target unit. The
design approach chosen for the EADF [1], has been to keep the spallation products confined where
they are generated. The lead-bismuth eutectic spallation target is therefore kept separated from the
primary coolant and confined within the structure of the target unit. The target unit structure is located
in the central opening of the sub-critical core, which has an equivalent diameter of 630 mm. The beam
pipe penetration takes place from the top of the reactor vessel.
3.2 Global neutronic parameters at beginning-of-life
The present version of the EA Monte Carlo code package enables a rather complete and detailed
model of the EADF reference configuration at the level of individual fuel pins or heat exchanger tubes
(presently arranged in square lattices). All the major core components have been taken into
consideration. The main global results for the beginning-of-life performance of the EADF reference
configuration are summarised in Tables 1 and 2.
855

Table 1. Main parameters of the EADF reference configuration
Global parameters Symbol EAP80 Units
Initial fuel mixture MOX (U-Pu)O 2
Initial fuel mass m fuel
3.793 ton
Initial Pu concentration m Pu
/m fuel
18.1 wt%
Initial fissile enrichment Pu 39,41 /U 38 18.6 wt%
Thermal power output P th
80 MWatt
Proton beam energy E p
600 MeV
Spallation neutron yield N (n/p)
14.51 ± 0.10
Neutron multiplication M 27.80 ± 0.56
Multiplication coefficient k = (M-1)/M 0.9640 ± 0.0007
Energetic gain G 42.73 ± 0.88
Gain coefficient G 0
1.54
Accelerator current I p
3.20 ± 0.07
Table 2. Core power distributions of the EADF reference configuration
Av. fuel specific power P th
/m fuel
24.5 W/g
Av. fuel power density P th
/V fuel
255 W/cm 3
Av. core power density P th
/V core
55 W/cm 3
Radial peaking factor P max
/P ave
1.25
Axial peaking factor P max
/P ave
1.18
3.3 Neutronic properties of the different source cases
The effective neutron multiplication factor, k eff , is an intrinsic property of the system. If the flux
distribution is not an eigenstate of the operator, the net neutron multiplication factor, k, will be
different, but this will not change the value of k eff . We can still formally define a value of k as,
ksrc
= 1−1 Msrc
but it will depend on the neutron flux as well as on the system. In particular, in the
presence of an external source, this value will depend on the position and energy spectrum of the
source neutrons. Hereinafter, k src will indicate the value of k calculated from the net multiplication
factor M src
in the presence of an external source.
By definition, a constant power operation requires ν/k eff
neutrons per fission, which means that an
external source has to provide a number of neutrons per fission that is:
⎛
µ eff
= ν ⎜
1
⎝
k eff
⎞
−1
⎟
⎠
=
ν
M eff
−1
856

if they are distributed exactly as the eigenfunction of the stationary problem. In the case of an arbitrary
external source, this number becomes:
⎛
µ src
= ν 1 ⎞
⎜ −1
⎟
⎝ k src ⎠
= ν
M src
−1
The ratio:
ϕ * = µ eff
= 1− k eff
µ src
1− k src
( )( k eff
ν)
( )( k src
ν)
is known as the importance of source neutrons. ν * is an effective number of neutrons per fission and
thus contains a correction for non-fission multiplicative processes such as (n,Xn) reactions, which are
of great importance for lead-bismuth or lead cooled fast reactors.
The neutronic properties of three different source cases has been examined: a spallation source
driven reference configuration (in short: reference or ADS case), a deuteron-deuteron fusion source
case (D-D or DD) and a deuteron-triton fusion source case (D-T or DT). As compared with the
reference case, the D-D configuration stayed within a range of 260 pcm (1 pcm = 1·10 -5 ) in terms of
the neutron multiplication factor (k src ), cf. Table 3. The D-T source configuration exhibits a
distinctively higher k src than the reference case, the difference in k src being 1 500 pcm (0.015). Other
central neutronic parameters are shown in Table 4. It should be noted that the required fusion source
intensities (“External n/s” in Table 4) are remarkably high. For comparison, it can be mentioned that
the source intensity of large-scale inertial confinement fusion experiments reaches a level of ~10 18 n/s
for the higher yielding D-T fusion.
Table 3. Neutron multiplication factors for the different source configurations
k eff
k src
ADS Fusion-DD Fusion-DT
0.9634 0.9640 0.9614 0.9790
Table 4. Main neutronic parameters for the different source configurations
ADS Fusion-DD Fusion-DT
ϕ* 1.0196 0.9478 1.7727
M = 1/(1-k src
) 27.8 25.9 47.6
Total n/s 8.01 10 18 7.98 10 18 8.13 10 18
External n/s 2.88 10 17 3.08 10 17 1.71 10 17
fiss/s 2.50 10 18 2.50 10 18 2.50 10 18
ν* 3.09 3.07 3.19
σ capt
(U 238 )/σ abs
(Pu 239 ) 0.68 0.68 0.68
Since the D-T source configuration case was found to show the most notable differences in
comparison with the reference ADS case, the onus will be on comparing the reference ADS and D-T
source configuration cases. In Table 5, the neutron balance of the whole EADF device is presented.
The fuel core neutron balance is presented in Table 6.
857

Table 5. Neutron balance in the whole device
Neutron absorption inventory ADS Fusion-DD Fusion-DT
Reactor containment 0.32% 0.32% 0.32%
LBE target 1.90% 1.76% 3.49%
Flow guides 0.15% 0.15% 0.15%
Heat exchangers 0.90% 0.89% 0.86%
Purification units 0.03% 0.03% 0.03%
Conv. enh. units 0.17% 0.17% 0.17%
Neutron shield 2.53% 2.50% 2.47%
Core upper reflector 5.33% 5.31% 5.21%
Core radial reflector 2.05% 2.03% 2.00%
Core lower reflector 6.69% 6.66% 6.55%
Fuel core 72.81% 73.15% 71.83%
Primary coolant 6.69% 6.66% 6.55%
Outs 0.20% 0.15% 0.15%
Total 100% 100% 100%
Main nuclear reactions ADS Fusion-DD Fusion-DT
Capture 66.09% 66.20% 65.01%
Fission 31.16% 31.29% 30.72%
n,Xn 2.13% 1.95% 3.69%
Others 0.42% 0.41% 0.43%
Outs 0.20% 0.15% 0.15%
Total 100% 100% 100%
Table 6. Neutron balance in the fuel core
Neutron absorption ADS Fusion-DD Fusion-DT
Fuel 89.80% 89.79% 89.80%
Cladding 3.77% 3.78% 3.77%
Sub-assembly wrapper 1.93% 1.93% 1.91%
Coolant 4.50% 4.50% 4.52%
Main nuclear reactions ADS Fusion-DD Fusion-DT
Capture 54.40% 54.44% 54.42%
Fission 42.81% 42.77% 42.76%
n,Xn 2.27% 2.26% 2.29%
Others 0.53% 0.53% 0.53%
858

From the perspective of the whole device, it is seen that the D-T source produces significantly more
reactions in the LBE target than the ADS source does, and leakage is smaller 1 . This increase is linked
with a relative decrease of reactions in the fuel core. The composition of reaction types is, in
consequence 2 , altered by an increase of the relative occurrence of (n,Xn) reactions at the expense of
capture and fission reactions. Regarding the neutron absorption inventory and the composition of the
main reaction types of the core, no particular or distinguishable trends can be reported (cf. Table 6). An
explanation for the abundancy of (n,Xn) reactions in the LBE target of the D-T source driven
configuration can be sought by studying Figure 3, which shows the LBE target neutron spectra for the
different source configurations. Just around 14 MeV, the spectrum of the D-T driven target shows clearly
distinguishable peaks, which arise due to the typical energy distribution of a D-T source. In this energy
region, cross sections for both (n,2n) and (n,3n) reactions come close to their maximum, and thus the
overall (n,Xn) reaction ratio will increase 3 . Table 7 compares the neutron flux distributions of the cases.
Figure 3. LBE target neutron spectra
1 The slightly higher leakage from the ADS geometry stems from the spatially asymmetrical neutron flux
distribution of an ADS source.
2
The LBE target is inclined for (n,Xn) reactions; capture and other reactions occur relatively infrequently.
3
The neutrons from D-D fusion appear at significantly lower energies, around 2.5 MeV. Thus, they do not
produce any significant changes in the overall (n,Xn) reaction occurrence.
859

Table 7. Neutron flux distributions throughout the device
Reactor region ADS Fusion-DD Fusion-DT
Reactor vessel 1.2 × 10 11 1.1 × 10 11 1.2 × 10 11
Safety vessel 3.6 × 10 10 3.5 × 10 10 3.6 × 10 10
LBE target 5.9 × 10 14 6.9 × 10 14 6.6 × 10 14
Target vessel 8.2 × 10 13 8.3 × 10 13 8.3 × 10 13
Heat exchangers 5.8 × 10 11 5.7 × 10 11 5.7 × 10 11
HX secondary coolant 7.2 × 10 11 7.2 × 10 11 7.2 × 10 11
Core neutronic protection 1.5 × 10 13 1.5 × 10 13 1.5 × 10 13
Av. fuel 1.3 × 10 14 1.3 × 10 14 1.3 × 10 14
Av. fuel cladding 3.4 × 10 13 3.4 × 10 13 3.4 × 10 13
Core radial reflector 7.1 × 10 13 7.0 × 10 13 7.0 × 10 13
As is seen from Table 7 (and Figure 3), the neutron flux distributions do not differ to any larger
extent except for in the LBE target. The characteristical peaks at energies of ~2.5 MeV and ~14 MeV
are found for the D-D and D-T fusion sources (respectively). Upon noting that the scale of Figure 3 is
logarithmic and comparing the high-energy part of the spectra, it is readily conceived that the integral
flux of the fusion cases indeed is higher than that of the reference ADS configuration
The axial and radial neutron flux distributions of all the configurations were examined for
possible differences. As can be perceived from Figures 4, 5 and 6 presenting the results graphically,
the only differences are found near the radial centre, where the LBE target is located.
Figure 4. Normalised neutron flux distribution over the height of the fuel core
1.4
1.2
ADS
dd
dt
1
0.8
0.6
0.4
0.2
0
-50 -40 -30 -20 -10 0 10 20 30 40 50
Axial distance [cm]
860

Figure 5. Radial distribution of the neutron flux
1x10 15 0 50 100 150 200 250 300 350
ADS
1x10 14
dd
dt
1x10 13
1x10 12
1x10 11
1x10 10
Radial Distance (cm)
Figure 6. Close-up near the LBE target of the radial distribution of the neutron flux
8x10 14 0 5 10 15 20 25 30
7x10 14
6x10 14
ADS
dd
dt
5x10 14
4x10 14
3x10 14
2x10 14
1x10 14
0x10 0
Radial Distance (cm)
A direct consequence of an increased high-energy neutron flux load on the LBE target will be a
higher displacement rate 4 . The EA-MC code allows for estimation of the damage to structural material
from the generated neutron flux spectra, accounting for damage induced by both high-energy particles
and low-energy neutrons. Table 8 presents the displacement rate in some of the main structural
components. The LBE target serves as an important attenuator of neutron flux and energy, which is
clearly recognised from the table: the DPA/year values are similar for the other structural components
than the target itself. Both fusion sources produce roughly the double amount of damage than the ADS
configuration does, but the damage is effectively absorbed in the target.
4 The displacement rate is measured in units of displacement per atom/year (DPA/year).
861

Table 8. Displacement rates
DPA/year ADS Fusion-DD Fusion-DT
Reactor vessel 8.3 × 10 -6 7.5 × 10 -6 8.1 × 10 -6
Safety vessel 2.5 × 10 -6 2.4 × 10 -6 2.6 × 10 -6
LBE target 1.566 2.988 2.643
Target vessel 0.230 0.235 0.235
Heat exchangers 1.5 × 10 -4 1.5 × 10 -4 1.5 × 10 -4
HX secondary coolant 1.1 × 10 -4 1. × 10 -4 1.1 × 10 -4
Core neutronic protection 5.7 × 10 -3 5.6 × 10 -3 5.7 × 10 -3
Av. fuel 0.807 0.808 0.809
Av. fuel cladding 0.152 0.152 0.152
Core radial reflector 0.146 0.145 0.145
Table 9 reports the transmutation efficiencies for the nuclides to be destroyed. In the case of 238 Pu
and 241 Pu, the D-T configuration proves to have the most efficient transmuter capabilities, although the
difference is slight. This benefit arises due to the harder spectrum of the D-T source (cf. Figure 3).
Table 9. Transmutation efficiencies
Fission/Capture ADS Fusion-DD Fusion-DT
Pu 238 1.83 1.85 1.94
Pu 239 3.33 3.32 3.33
Pu 240 0.66 0.66 0.66
Pu 241 6.07 6.05 6.11
Pu 242 0.41 0.41 0.41
Np 237 0.28 0.52 0.19
Am 241 0.13 0.13 0.14
Since the neutron flux distributions were essentially the same throughout the core 5 independent of
source configuration, also the transmutation rates remained unaffected by the choice of neutron
source. Table 10 presents a comparison between the transmutation rates for the EADF fuelled with
(Th-Pu)O 2
, the EADF fuelled with (U-Pu)O 2
and a standard PWR fuelled with UO 2
.
5 As earlier was seen, there is a difference between different source configuration for the flux over the LBE target. However,
since this region carries no fissile materials, no effect on transmutation rates is seen.
862

Table 10. Transmutation rates (kg/TW th
h) of plutonium and minor actinides
Nuclides
EADF
(Th-Pu)O 2
ENDF/B-VI
EADF
(U-Pu)O 2
ENDF/B-VI
PWR
UO 2
233
U +31.0 – –
Pu -42.8 -7.39 +11.0
Np +0.03 +0.25 +0.57
Am +0.24 +0.17 +0.54
Cm +0.007 +0.017 +0.044
99
Tc prod +1.08 +1.07 +0.99
99
Tc trans -3.77 -3.77 –
4. Conclusion
The neutronic properties of four different neutron source configuration of the EADF have been
examined by means of 3-D Monte Carlo techniques. The results indicate that the accelerator driven
and the fusion-DD source systems would exhibit similar neutronic properties. The fusion-DT source
driven configuration differs from the others, giving rise to a neutron multiplication factor ~1 500 pcm
higher than the others. The effect was seen to be mainly due to the DT-fusion characteristical neutron
emission energy spectrum, which produces a significantly larger share of (n,Xn) produced neutrons.
Finally, it was established that the choice of source configuration has negligible impact on
transmutation efficiencies and transmutation rates.
REFERENCES
[1] Ansaldo Nucleare, Energy Amplifier Demonstration Facility Reference Configuration:
Summary Report (ANSALDO Nucleare, EA-B0.00-1-200 – Rev. 0, January 1999).
[2] OECD Nuclear Energy Agency, Data Bank (Paris, France, 1994), http://www.nea.fr.
[3] A. Fassó, A. Ferrari, J. Ranft, P.R. Sala, G.R. Stevenson, J.M. Zazula, Nuclear Instruments and
Methods A, 1993, 332, 459, also, CERN report CERN/TIS-RP/93-2/PP (Geneva, 1993).
[4] A. Fassó et al., Intermediate Energy Nuclear Data: Models and Codes, Proceedings of a
Specialists meeting (Paris, France, May 30-June 1, 1994), p. 271 (OECD, 1994) and references
therein.
[5] S. Atzeni, Y. Kadi, C. Rubbia, Statistical Fluctuations in Monte Carlo Simulations of the
Energy Amplifier, CERN report CERN/LHC/EET 98-004, Geneva, April 20, 1998.
[6] C. Rubbia et al., Conceptual Design of a Fast Neutron Operated High Power Energy Amplifier,
CERN report CERN/AT/95-44 (EET), Geneva, September 29, 1995.
863

DEEP UNDERGROUND TRANSMUTOR (PASSIVE HEAT
REMOVAL OF LWR WITH HARD NEUTRON ENERGY SPECTRUM)
Hiroshi Takahashi
Brookhaven National Laboratory
Upton, New York, 11973, USA
E-mail: takahash@bnl.gov
Abstract
To run a high conversion reactor with Pu-Th fuelled tight fuelled assembly, which has a long burn-up
of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat
can be removed passively not only during a steady-state run and also in an emergency case of loss of
coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the
evacuation area around the reactor can be minimised, and the reactor placed near the consumer area.
This approach reduces the cost of generating electricity by eliminating the container building and
shortening transmission lines.
865

1. Introduction
The concept of a high conversion light water reactor using a high concentration of Pu-fuel tightlattice
has been proposed [1]. This reactor has a hard neutron energy spectrum close to that of an Nacooled
fast reactor, and high burn-up of fuel can be obtained. A reactor with uranium fertile material
has a positive water-coolant void coefficient, so, to get a negative void-coefficient, it is required a
pancake-type flat core configuration or a fuel assembly with a neutron-streaming void section which
reduces neutron economy. The use of thorium fertile material [2], however, provides a negative voidcoefficient
without having neutron-leaky core configuration; the neutron economy accordingly is
improved and a higher burn-up of fuel can be obtained compared with a reactor with uranium fertile
materials. However, the pumping power of water coolant has to be substantially increased to remove
the high-density heat from a tight latticed-fuelled core. During steady operation, coolant flow can be
maintained by increasing pumping power several times above that of the regular LWR. But during
emergencies, such as an outage of on-site power or loss of coolant, heat removal becomes serious
problem. This accident scenario has been studied analysed in detail and an experimental study for heat
removal from a tight lattice has been planned in the Japanese research programme.
Due to the hard neutron energy spectrum from the high concentration of the Pu-fuel tight lattice
and the good neutron economy, this reactor can be used for transmuting minor actinide (MA) and long
lived fission product (LLFP). MA has a similar neutronics properties as the fertile material of thorium.
By capturing the neutrons, MA will be converted into fissile material, and thereby contribute to long
burn-up of fuel. To transmute the LLFP of Tc, which has low neutron-capture cross-section, a
moderator such as zirconium hydride is used in the transmutor cooled by Na or Pb-Bi; however, the
neutrons can be effectively moderated by light water, so that the configuration of the light-watercooled
transmutor becomes simple.
2. Passive heat removal
To withstand an emergency case of loss of pumping power, a passive cooling system, such as
heat removal using the natural circulation of the coolant, has been studied. I propose here using a
tight-latticed water reactor embedded in a deep underground location, so that it is cooled by the natural
circulation of the water. The high pressure difference between the inlet and outlet in the narrow water
channel of the tight lattice is generated by the difference in gravity force between the low density of
boiled water and the high density water condensed after the steam passes through steam turbine. To
obtain such a high-pressure difference, the vacuum condenser must be located far above the boiling
point of the water. The pumping-pressure difference needed to circulate water in a regular BWR and
PWR are, respectively, 2 atm and 1.5 atm which is equivalent to a 20-15 meter difference in water
height. But for our high conversion (HC) LWR with a tight lattice, the difference in pumping power is
increase several times; a water height of more than 80-60 meters is needed to naturally circulate
coolant water. By putting the reactor deep underground, we can provide enough space to get such a
high pressure difference between the inlet and outlet using the density difference between the steam
section and the water which is condensed after passing through the steam turbine and steam condenser
which in our configuration are located far above the reactor vessel.
By locating the reactor even deeper, the pressure imposed on the pressure vessel is increased by
the gravitational force of the surrounding earth. A water pressure of 100 atm and 150 atm for a BWR
and a PWR can be provided, respectively, by the earth’s pressure at a depth of 400- and 600-meters.
The passive cooling system using natural circulation which is conventionally proposed, is
operated in an environment wherein there is not enough pressure, so that the steam -water state is not
well defined and some instability is created; hence this is not necessarily a safe operation, even in the
866

passive state. By operating at a high enough pressure, these non-linear effects can be eliminated, and
we can obtain safe operation of the reactor deeper underground. From this point of view, there are
many advantages to the concept of a deep underground reactor. Due to the pressure of earth’s gravity,
the pressure vessel can be quite thin, thus the reactor would be much lighter than that of a regular
LWR operated on the earth surface. A huge heavy crane would not be required to move this reactor
and it could be constructed with a modular-type design.
It has been proposed to use super-critical steam for gaining high efficiency of electric generation [3],
this reactor requires 250 atm water pressure, this can be achieved by the earth pressure in the 1 000 meter
deep under ground: 260 atm. The more high water pressure can be obtained by providing thick pressure
vessel.
3. A deep underground facility
Deep underground geological storage of high level waste has been studied. The depth of the
Yucca Mountain Repository is about 300 meters depth, so that a tunnel more than tens of kilometers
long is planned.
To measure neutrino oscillation, a super-kamiokande detector with a 50-meter high and 40-meter
diameter water-tank is installed 1 000 meters deep in a mountain in Kamioka mine in Japan. Many
other high-energy facilities such as the Grand Sasso (Italy) have been used for such high-energy
experiments.
The cost of the digging a large hall underground is not as expensive as digging above ground. I
was informed that the cost of a 10 × 20 meter tunnel is about 10 000 dollars per 1 meter depth in
Japan, although this depends on geological features. Nevertheless, the cost of the digging in hard rock
deep underground is cheaper than excavating in shallow but fractured rock [4].
Figures 1 and 2 respectively show the conceptual layout of the installation of a BWR and a PWR
in deep underground. The emergency cooling can be installed high above the reactor, to provide the
high pressure needed to cool the decay heat. In the PWR version, the heat exchanger with primary
boiling water is used to obtain the high-pressure difference between the inlet and outlet of narrow
water channel.
To get a high gravitational force, deep tunnelling such as in the case in Yucca Mountain
programme can be utilised instead of making a deep vertical hole, the reactor can be embedded in the
tunneling’s wall. By putting the steam turbine and vacuum condenser far above level of the reactor we
can get high pressure difference between inlet and outlet in the narrow cooling water channel so that
heat can be removed by the natural circulation of water. In a light reactor without a thick pressure
vessel, when the fuel must be exchanged (which is infrequent due to long burn-up of fuel in Pu-fuel
with thorium fertile with a tight latticed reactor), the reactor can be taken out of the wall of the tunnel,
and fuel can be exchanged in the space adjacent the tunnelling wall. Although the tubes of steam and
condensed water must be disconnected and connected again to do this, it can be done without
difficulty.
4. Economy of the deep underground reactor, transmission lines, container building’s
emergency cooling system and evacuation
To protect the public from radiation hazards in the fall out from radioactive releases from regular
nuclear power plants, a container building for the reactor is provided. By putting the reactor deep
867

underground, the radiation field generated from any release would be very small, and we could
minimise the number of people who would have to be evacuated.
Thus, we could build a NP near a consumer area and thereby shorted the transmission lines. This
would entail substantial reduction in the cost of electricity generation. Generally the cost of
transmission lines is very substantial, especially establishing lines in densely populated areas where
the cost of land is high; it was estimated that the construction of a transmission line of more than
400 km is greater than the cost of building the power plant itself.
Although installing many facilities, such as the steam turbine and vacuum condenser is more
expensive than building on the earth’s surface, by not having to construct long transmission lines and a
double-walled container building and other facilities associated with having a seismically strong
building structure, lowers the overall cost of the constructing the reactor deep underground. However,
a detailed cost evaluation is still required.
To remove decay heat after an accident, emergency-cooling borated water is stored in the
container building; in the case of the tight-lattice reactor, the height of the cooling water must be
considerable to get the needed high pressure; also it is very vulnerable for seismic protection.
When the reactor is deep underground the emergency borated water also can be stored
underground where the movements of an earthquake is smaller than at shallower depths. Also there is
enough room to store a huge amount of water high above the reactor and so water under high pressure
can be provided to cool the reactor.
Also to protect against re-criticality of the core due to its melting, a large pool of water can be
provided under an underground reactor without difficulty.
Installing a nuclear facility in under deep underground ensures that public are well protected. We
can eliminate containment building and reduce the seismic hazards avoiding the strong earth motions
at the surface. The area of emergency evacuation, which closed down the Shoreham NP near BNL in
the last decade, also be minimised, thus there is possibility of constructing the NP near a consumer
area, and the expensive construction of high voltage electricity transmission lines can be lessened.
Therefore, building a deep underground reactor might be more economical than building the nuclear
power plant on the above ground. Although it might be very difficult to obtain public acceptance in the
present political climate, it might be wise to built future reactor deep underground.
5. The accelerator driven reactor
I have discussed a power-generating reactor so far, but this reactor with its hard-neutron energy
spectrum, can also be used for transmuting the minor actinides and LLFP. As the neutron energy
spectrum becomes harder, neutron economy increased, but the negative void-coefficient might be
small and so the delayed neutron portion also becomes small. To run this type of reactor in a safer
mode, it is desirable to run it in a sub-critical condition.
For transmuting minor actinides (MAs) and long-lived fission products (LLFPs), several kinds of
nuclear reactor concepts were proposed, such as the Na, Pb, Pb-Bi cooled fast reactor in critical
operation, and in sub-critical operation driven by spallation neutrons. For a transmutor of Tc, which
has a low neutron-capture cross-section at intermediate energy, to a get high transmutation rate, the
high-energy neutron is thermalized by moderator such as Zr hydride. In this water-cooled reactor the
excellent moderator of water is in the core and so it is not necessary to install another moderator.
868

I have proposed using an accelerator driven reactor [5], run sub-critical condition by providing
spallation neutrons created by medium energy proton. When the spallation target is equipped with a
beam window, it should be protected from radiation damage by expanding the proton beam. By
locating the reactor deep underground, there is enough space to install the beam expansion section.
6. Conclusion
The construction of a nuclear power plant underground was proposed by A. Sakharov and
E. Teller [6] for protection against radiation hazard, and Russian Pu and Electric Generation nuclear
power plant is operated in Enisei river [7]. However, my proposal for a deep underground reactor uses
earth’s gravity force to provide passive heat removal using natural circulation of the reactor coolant,
such a nuclear power plant can be operated more safely. The high pressure required for heat removal
from the coolant can be supplemented by use of earth’s gravity. Also it provides the pressure
difference required removing fission heat from the nuclear fuel with natural circulation of gravity
force. This natural-water-coolant circulation can eliminate concerns about an on-site electricity
blackout. The storage facility for the emergency coolant system can be built far above the reactor
because there is enough space available in a deep underground installation.
Also for defence purposes in protecting people from nuclear hazards created by nuclear plants
smart bombing, future reactor should be built in deep underground.
Here, I have discussed mainly the light water reactor, but this concept equally can be applied to
gas-cooled reactors, which require high pressure, and it will apply many other types of rectors.
The capability of removing heat, not only during steady-state operation but also in accidents
involving a loss of coolant or an outage in on-site power is essential especially for the HC reactor with
tight latticed fuel assembly. By putting the reactor deep underground and removing heat passively,
public safety is ensured.
Acknowledgements
The author would like to express his thanks to Dr. B.D. Chung, Dr. J. Herczog, Dr. U. Rohatgi,
Dr. S. Mtsuura, Dr. Yamazaki, Dr. K. Takayama, Dr. K. Higuchi, Dr. Ikuta, Prof. B.W. Lee for their
valuable discussion.
869

REFERENCES
[1] T. Iwamura et al., Resarch on Reduced-moderation Water Reactor (RMWR), JAERI_research
Report, 99-058 (In Japanese).
[2] Hiroshi Takahashi, Upendra Rohatgi, A Proliferation Resistant Hexagonal Tight Lattice LWR
Fuelled Core for Increased Burn-up and Reduced Fuel Storage Requirements, Progress report
of NERI: Aug. 1999 to May 2000.
[3] Y. Oka, Physics of Supercritical Pressure Light Water Cooled Reactors, 1998 Frederic Joliot
Summer school, Aug. 17-Aug. 26 1998, Cadarache, France.
[4] K. Higuchi, Ikuta, Private Communication, Nov. 2000.
[5] H. Takahashi, Safe, Economical Operation of a Slightly Sub-critical Fast Reactor and
Transmutor with a Small Proton Accelerator, Proc. Int. Conf. on Reactor Physics and Reactor
Computation, p. 79, Y. Ronen and E. Elias (Eds.), Tel-Aviv, Jan. 23-26, 1994, Ben-Gurion
Univ. of the Negev Press (1994).
[6] Edward Teller, Muriel Ishikawa, Lowell Wood, Roderick Hyde and John Nukolls, Completely
Automated Nuclear Reactors for Long Term Operation II, ICENES 96.
[7] Informal meeting between US and Russia for Pu reduction at BNL, Sept. 2000.
870

Figure 1. Layout of deep underground BWR
Earth surface or mountain
Earth
500m~1000m
Emergency borate
Cooling water
6WHDP7XUELQH
Generator
A
Condenser
Steam
B
Core
Plenum
Condensed Water
871

Figure 2. Layout of deep underground PWR
Earth surface or mountain
Earth
800m~1000m
6WHDP7XUELQH
Generator
Condenser
Emergency borate
Cooling water
Steam
A
B
Heat Exchanger
Condensed Water
High Temp
Water
Height Difference
300~400m
Core
Plenum
Low Temp
Water
872

RADIATION CHARACTERISTICS OF PWR MOX SPENT FUEL AFTER LONG-TERM
STORAGE BEFORE TRANSMUTATION IN ACCELERATOR DRIVEN SYSTEMS
B.R. Bergelson, A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova, T.S. Zaritskaya
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP),
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
Changes in a radiotoxicity and decay heat power of actinides from spent uranium and uraniumplutonium
nuclear fuel for PWR-type reactors at long-term storage are investigated. The extraction of
the most important nuclides for transmutation permits to reduce radiotoxic content of wastes remaining
in storage. The decay heat power of actinides is determined by the same nuclides as the radiotoxicity is.
The radiotoxicity and decay heat power of actinides of uranium-plutonium fuel is 2.5 times higher, than
that of uranium fuel because of the greater accumulation of plutonium, americium, and curium isotopes.
873

1. Introduction
The problem of the management of long-lived radioactive waste from spent nuclear fuel is
closely connected to the prospects for development of nuclear power. Some directions for decisions to
this problem are now studied. One of the opportunities is the construction of a long-term controllable
storage facility. Other ways are connected with the realisation of nuclear transmutation of long-lived
waste. Apparently, the resolution of the problem will combine several approaches. To determine their
correct combination, it is necessary to know how the major characteristics of radioactive waste vary
during long-term storage.
One of the important radiation characteristics of radwaste is the decay heat power. It influences
the design of a storage facility and the type of heat removal system. Other important characteristic of
radwaste is the radiotoxicity. It is a more representative characteristic than activity because
radiotoxicity includes the influence of radiation of separate nuclides on the human body.
Radiotoxicity is a base to evaluate ecological danger of stored radwaste.
Time dependence of radwaste radiation characteristics during storage allows identifying most
important nuclides in different stages of storage. Their separation and extraction from storage with
subsequent transmutation permits to reduce the radiologic danger of wastes staying in storage.
Removal of nuclides with a decrease of the remaining decay heat power in storage permits to ease
requirements to the heat removal systems at long-term storage of wastes.
Changes in radiotoxicity and decay heat power of actinides from spent uranium – plutonium
nuclear fuel of VVER-1000 type reactors at storage during 100 000 years are investigated in this
paper.
The radiotoxicity RT i
of each nuclide i by air or by water is determined by the ratio:
RT i
= A i
/MPA i
where A i
– activity of considered amount of a nuclide i, MPA i
– represents the maximum permissible
activity of this nuclide by air or by water according to radiation safety standards. Total radiotoxicity
is equal to a sum of radiotoxicities of all nuclides taken in those amounts in which they are contained
in the considered mix of nuclides. For the calculations, data of MPA accepted in the Russian
Federation [1] were used. For the calculations of a decay heat power, the contributions from alpha-,
beta- and gamma – radiations [2] were taken into account.
2. Calculation results
The total radiotoxicity of actinides in air and in water and the contributions of the most important
actinides in total radiotoxicity at storage of spent uranium-plutonium fuel of a VVER-1000 type
reactor during 100 000 years are presented in Tables 1 and 2. The amount of actinides corresponded
to their contents in 1 tonne of spent fuel with burn-up of 40 kg of fission products per 1 tonne and
subsequent cooling during 3 years. The fresh fuel was a mix of depleted uranium with addition of
3.5% 239 Pu. Only isotopes of neptunium, plutonium, americium, and curium were considered as
actinides.
The total decay heat power of actinides and the contributions of the most important actinides at
storage of spent uranium-plutonium fuel during 100 000 years is given in Table 3. Decay heat power,
as well as the radiotoxicity, corresponds to the content of actinides in 1 tonne of unloaded fuel.
874

Table 1. Radiotoxicity of actinides from uranium-plutonium spent fuel in air, m 3 air
T, year 1 10 100 1 000 10 000 100 000
238
Pu 4.39 + 16 4.09 + 16 2.02 + 16 2.35 + 13 – –
239
Pu 9.57 + 15 9.57 + 15 9.55 + 15 9.34 + 15 7.38 + 15 5.68 + 14
240
Pu 2.62 + 16 2.63 + 16 2.63 + 16 2.40 + 16 9.23 + 15 –
241
Pu 8.78 + 16 5.69 + 16 7.49 + 14 1.19 + 12 – –
242
Pu 1.01 + 14 1.01 + 14 1.01 + 14 1.01 + 14 9.92 + 13 8.41 + 13
241
Am 3.67 + 16 8.54 + 16 1.55 + 17 3.69 + 16 2.90 + 13 –
242m
Am 2.93 + 14 2.81 + 14 1.87 + 14 3.08 + 12 – –
243
Am 1.03 + 15 1.03 + 15 1.03 + 15 9.42 + 14 4.05 + 14 –
243
Cm 8.21 + 14 6.59 + 14 7.39 + 13 – – –
244
Cm 9.15 + 16 6.48 + 16 2.07 + 15 – – –
Total 2.99 + 17 2.86 + 17 2.15 + 17 7.13 + 16 1.72 + 16 6.80 + 14
Table 2. Radiotoxicity of actinides from uranium-plutonium spent fuel in water, kg water
T, year 1 10 100 1 000 10 000 100 000
238
Pu 1.98 + 14 1.84 + 14 9.10 + 13 1.06 + 11 – –
239
Pu 4.27 + 13 4.27 + 13 4.26 + 13 4.17 + 13 3.29 + 13 2.54 + 12
240
Pu 1.17 + 14 1.17 + 14 1.18 + 14 1.07 + 14 4.12 + 13 –
241
Pu 4.24 + 14 2.75 + 14 3.62 + 12 – – –
242
Pu 4.52 + 11 4.52 + 11 4.52 + 11 4.51 + 11 4.45 + 11 3.77 + 11
241
Am 1.54 + 14 3.59 + 14 6.50 + 14 1.55 + 14 1.22 + 11 –
242m
Am 1.32 + 12 1.27 + 12 8.44 + 11 – – –
243
Am 4.50 + 12 4.50 + 12 4.46 + 12 4.10 + 12 1.76 + 12 –
243
Cm 3.53 + 12 2.84 + 12 3.18 + 11 – – –
244
Cm 3.51 + 14 2.49 + 14 7.93 + 12 – – –
Total 1.30 + 15 1.24 + 15 9.19 + 14 3.09 + 14 7.67 + 13 3.11 + 12
Table 3. Decay heat power of actinides from uranium-plutonium spent fuel, W
T, year 1 10 100 1000 10 000 100 000
226
Ra – – – – 0.015 0.114
238
Pu 106 99.0 48.9 0.057 – –
239
Pu 20.1 20.1 20.1 19.6 15.5 1.19
240
Pu 55.0 55.3 55.4 50.4 19.4 0.0014
241
Pu 10.6 6.85 0.090 – – –
242
Pu 0.209 0.209 0.209 0.209 0.206 0.174
241
Am 96.5 225 407 96.9 0.076 –
243
Am 2.70 2.70 2.68 2.46 1.06 –
242
Cm 15.5 0.763 0.506 – – –
243
Cm 3.24 2.60 0.292 – – –
244
Cm 398 282 9.00 – – –
245
Cm 0.162 0.162 0.160 0.149 0.071 –
Total 708 694 544 170 36.5 1.66
875

The data presented show that the radiotoxicity of actinides of spent uranium-plutonium fuel in air in an
initial period of storage is determined by nuclides 244 Cm, 241 Pu and 238 Pu. Their contribution in beginning of
the storage is about 75%. All isotopes of plutonium give 56%, 244 Cm – 30%. In addition, 241 Am creates 12%
of radiotoxicity. At storage there is the conversion 241 Pu into 241 Am. After 100 years of storage, total
radiotoxicity of actinides decreases 1.4 times. The main contribution (72%) comes from 241 Am. The
contribution of plutonium isotopes makes 26%. The amount of 244 Cm decreases essentially because of the
decay. Its radiotoxicity falls 44 times and makes 1% of total radiotoxicity at the end of 100-year storage.
After 1 000 years of storage, radiotoxicity in air falls 4.2 times, after 10 000 years – 17 times, after
100 000 years it falls 440 times.
The radiotoxicity in water is determined by the same nuclides. In initial period of storage, all
plutonium isotopes give the contribution in total radiotoxicity 60%, 244 Cm – 27%, 241 Am – 12%. After
100 years of storage, total radiotoxicity of actinides decreases 1.4 times. The main contribution 71%
gives 241 Am. The contribution of plutonium isotopes makes 28%, 244 Cm – 0.9%. The 241 Am gives
maximal respective contribution 72% after 300 years. After 1 000 years its share decreases quickly.
After 1 000 years, total radiotoxicity in water reduces 4.2 times, after 10 000 years – 17 years, after
100 000 years – 420 times.
The decay heat power of actinides of spent uranium-plutonium fuel in initial period of storage is
determined by a nuclide 244 Cm, which creates 56% of power. The contribution of plutonium isotopes
makes 27%, 241 Am – 13%. After 100 years of storage, total power of actinides decreases 1.3 times.
The main contribution 75% gives 241 Am, plutonium isotopes – 23%, 244 Cm – 1.6%. After 10 000 years,
power reduces 20 times, after 100 000 years – 460 times.
The radiotoxicity of actinides of uranium-plutonium fuel appears 2.5 times more and the decay heat
power appears 2.7 times more than that of usual uranium fuel because of greater (by 2-3 times)
accumulation of 239 Pu, 240 Pu, 241 Pu, and 244 Cm.
3. Conclusion
Recommendations could be done to perform chemical separation of plutonium, americium,
curium before long-term storage. Americium should be separated after 50-70 year of storage
sufficient for conversion 241 Pu in 241 Am. Curium can be separated in the beginning of storage. This will
allow reducing radiotoxicity of the remaining actinides by 20-30%. If we abandon a separation of
curium then it decays in 100 years almost fully. Extracted americium (possibly, with long-lived
curium isotopes) should be directed to transmutation and plutonium – to repeated use. The separation
of actinides is expedient also to reduce decay heat power. So, extraction of americium after
241
Pu decay and decay of greater part of 238 Pu-permits to reduce essentially decay heat power of the
plutonium fraction.
REFERENCES
[1] Radiation Safety Standards (NRB-99), Minzdrav of Russia, Moscow, 1999.
[2] Schemes of Decay of Radionuclides. Energy and Intensity of Irradiation, Publication 38 ICRS,
Moscow, Energoatomizdat, 1987.
876

RADIATION CHARACTERISTICS OF URANIUM-THORIUM SPENT FUEL IN LONG-TERM
STORAGE FOR FOLLOWING TRANSMUTATION IN ACCELERATOR DRIVEN SYSTEMS
B.R. Bergelson, A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova, T.S. Zaritskaya
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
Changes in radiotoxicity and decay heat power of actinides from spent thorium-uranium nuclear fuel
for PWR-type reactors at long-term storage are investigated. The extraction of the most important
nuclides for transmutation permits to reduce radiotoxic content of wastes remaining in storage.
Actinide accumulation in the thorium-uranium fuel cycle is much less than in the common-type
uranium fuel cycle. The radiotoxicity of actinides of thorium-uranium fuel in air is 5.5 times less and
that of in water is 3.5 times less than that of uranium fuel.
877

1. Introduction
The problem of the radiotoxicity of long-lived radioactive wastes produced in various nuclear
fuel cycles is important from the viewpoint of ecological impact of these cycles. Separation of the
most important nuclides and their extraction from storage with subsequent transmutation permits to
reduce radiologic danger of remaining wastes in storage. The removal of nuclides with increased
decay heat power from storage permits to ease requirements to heat removal systems of long-term
storages. Quantitative comparison of the radiologic characteristics of minor actinides produced in
various fuel cycles is also of interest.
Changes in radiotoxicity and decay heat power of actinides from spent thorium-uranium nuclear
fuel of VVER-1000 type reactors at storage during 100 000 years are investigated in this paper.
The radiotoxicity RT i
of each nuclide i by air or by water is determined by the ratio:
RT i
= A i
/MPA i
where A i – activity of considered amount of a nuclide i, MPA i – represents the maximum permissible
activity of this nuclide by air or by water according to radiation safety standards. Total radiotoxicity
is equal to a sum of radiotoxicities of all nuclides taken in those amounts in which they are contained
in the considered mix of nuclides. For the calculations, data of MPA accepted in Russia [1] were
used. For the calculations of the decay heat power, the contributions from alpha-, beta- and gamma –
radiations [2] were taken into account.
2. Calculation results
Total radiotoxicity of actinides in air and in water and the contributions of most important
actinides in total radiotoxicity at storage of spent thorium-uranium fuel of a VVER-1000 type reactor
during 100 000 years are presented in Tables 1 and 2. The content of actinides in spent thoriumuranium
fuel was calculated for the neutron spectrum created by basic uranium fuel in a VVER type
reactor. The data correspond to burn-up of basic uranium fuel, 44 kg of fission products per 1 tonne
and subsequent cooling during 3 years. The fresh fuel was a mix of thorium with an addition of 3.3%
233
U. Isotopes of thorium, uranium, and more heavier nuclides were taken into account.
Total decay heat power of actinides and the contributions of the most important actinides at
storage of spent thorium-uranium fuel during 100 000 years is given in Table 3. Decay heat power, as
well as the radiotoxicity, corresponds to the content of actinides in 1 tonne of unloaded fuel.
The data presented show that radiotoxicity of actinides of spent thorium-uranium fuel during the
first 100-300 years is determined by the nuclide 232 U and its daughter nuclides, first of which is 228 Th.
The half-life of 232 U is 68.9 years, of 228 Th, 1.9 years. The subsequent daughter nuclides in the decay
chain of 232 U after 228 Th are short-lived. Among other actinides, the most important are 238 Pu and 234 U.
Their contribution is 1-2 order lower. The appreciable contribution in radiotoxicity in air introduces
also 233 U. The contributions from 239 Pu, 240 Pu, 241 Pu, 241 Am, 232 Th are 4 order lower, that of 232 U. At
100 years storage, the total radiotoxicity in air decreases 2.4 times, than that of in water – 2.6 times, at
1 000 years – 410 times and 280 times. After 1 000 years storage, the 232 U with daughter nuclides
gives no contribution to radiotoxicity. It is determined by 234 U, 230 Th. After 3 000 years there is an
increase of radiotoxicity because of accumulation of 226 Ra, 229 Th, 230 Th.
878

Table 1. Radiotoxicity of actinides from thorium-uranium spent fuel in air, m 3 air
T, year 1 10 100 1 000 10 000 100 000
226
Ra – – – – 2.87 + 12 2.16 + 13
228
Th 1.63 + 16 1.66 + 16 6.74 + 15 7.70 + 11 – –
229
Th – – 4.20 + 11 3.92 + 12 2.65 + 13 4.18 + 13
230
Th – 1.02 + 11 2.21 + 11 1.40 + 12 1.25 + 13 7.43 + 13
232
Th 7.66 + 11 7.66 + 11 7.66 + 11 7.66 + 11 7.66 + 11 7.66 + 11
232
U 3.68 + 15 3.36 + 15 1.36 + 15 1.56 + 11 – –
233
U 2.31 + 12 2.31 + 12 2.31 + 12 2.31 + 12 2.30 + 12 2.21 + 12
234
U 4.03 + 13 4.03 + 13 4.04 + 13 4.03 + 13 3.93 + 13 3.04 + 13
237
Np 1.36 + 11 1.36 + 11 1.36 + 11 1.37 + 11 1.36 + 11 1.32 + 11
238
Pu 1.56 + 15 1.45 + 15 7.12 + 14 5.81 + 11 – –
239
Pu 7.95 + 11 7.95 + 11 7.93 + 11 7.73 + 11 5.97 + 11 –
240
Pu 5.87 + 11 5.87 + 11 5.82 + 11 5.29 + 11 2.04 + 11 –
241
Pu 1.65 + 12 1.07 + 12 – – – –
241
Am 5.99 + 11 1.52 + 12 2.84 + 12 6.75 + 11 – –
244
Cm 1.47 + 11 1.04 + 11 – – – –
Total 2.16 + 16 2.15 + 16 8.86 + 15 5.24 + 13 8.52 + 13 1.71 + 14
Table 2. Radiotoxicity of actinides from thorium-uranium spent fuel in water, kg water
T, year 1 10 100 1 000 10 000 100 000
226
Ra – – – 4.91 + 09 1.72 + 11 1.29 + 12
228
Th 2.50 + 13 2.53 + 13 1.03 + 13 1.49 + 09 – –
229
Th 9.97 + 07 3.24 + 08 2.55 + 09 2.38 + 10 1.61 + 11 2.54 + 11
230
Th 1.21 + 09 1.36 + 09 2.94 + 09 1.87 + 10 1.67 + 11 9.91 + 11
232
Th 6.26 + 09 6.26 + 09 6.26 + 09 6.26 + 09 6.26 + 09 6.26 + 09
232
U 1.23 + 14 1.12 + 14 4.53 + 13 6.54 + 09 – –
233
U 2.73 + 10 2.73 + 10 2.73 + 10 2.73 + 10 2.72 + 10 2.62 + 10
234
U 4.45 + 11 4.45 + 11 4.45 + 11 4.44 + 11 4.33 + 11 3.36 + 11
238
Pu 7.00 + 12 6.52 + 12 3.20 + 12 2.62 + 09 – –
239
Pu 3.55 + 09 3.55 + 09 3.54 + 09 3.45 + 09 2.66 + 09 –
240
Pu 2.62 + 09 2.62 + 09 2.60 + 09 2.36 + 09 – –
241
Pu 7.98 + 09 5.18 + 09 – – – –
241
Am 2.52 + 09 6.38 + 09 1.19 + 10 2.84 + 09 – –
Total 1.55 + 14 1.44 + 14 5.93 + 13 5.44 + 11 9.71 + 11 2.91 + 12
The decay heat power of actinides of spent thorium-uranium fuel is determined by the same
nuclides as the radiotoxicity. After 100 years it decreases 2.4 times, after 1 000 years – 280 times,
after 3 000 years it increases because of accumulation of 226 Ra, 229 Th, 230 Th.
The radiotoxicity of actinides of thorium-uranium fuel in air in begin of storage appears 5.5 times
less and that of in water is 3.5 times less and decay heat power appears 1.2 times more than those of
usual uranium fuel.
879

Table 3. Decay heat power of actinides from thorium-uranium spent fuel, W
T, year 1 10 100 1 000 10 000 100 000
226
Ra – – – 0.013 0.454 3.41
229
Th – – 0.004 0.037 0.250 0.395
230
Th – – 0.001 0.009 0.084 0.499
233
U 0.058 0.058 0.058 0.058 0.058 0.056
234
U 1.00 1.00 1.00 1.00 0.977 0.757
238
Pu 3.76 3.51 1.72 0.001 – –
239
Pu 0.002 0.002 0.002 0.002 0.001 –
240
Pu 0.001 0.001 0.001 0.001 – –
241
Am 0.002 0.004 0.007 0.002 – –
232
U 44.9 41.0 16.6 0.002 – –
228
Th 266 270 110 0.014 – –
232
Th 0.002 0.002 0.002 0.002 0.002 0.002
Total 315 315 129 1.14 1.83 5.12
3. Conclusion
The overwhelming share of radiotoxicity and decay heat power is determined by 232 U which is of
the same chemical element as the main fuel isotope 233 U. It is obvious that the repeated use of
thorium-uranium fuel connected with various variants of new 233 U addition will be accompanied by
accumulation of radiotoxicity.
At unitary use of thorium-uranium fuel with deep 233 U burn-up, it is necessary to perform additional
deep burn-out (transmutation) of uranium fraction containing both 233 U and 232 U. The further reduction of
radiotoxicity by several orders can be achieved through extraction and transmutation of plutonium
faction ( 238 Pu). The transmutation of 228 Th – daughter nuclide of 232 U – is not necessary because 228 Th
decays practically completely after 10 years together with its short-lived daughter nuclides.
REFERENCES
[1] Radiation Safety Standards (NRB-99), Minzdrav of Russia, Moscow, 1999.
[2] Schemes of Decay of Radionuclides. Energy and Intensity of Irradiation, Publication 38 ICRS.
Moscow, Energoatomizdat, 1987.
880

INTERNATIONAL CO-OPERATION ON CREATION OF A
DEMONSTRATION TRANSMUTATION ACCELERATOR DRIVEN SYSTEM
A.S. Gerasimov, G.V. Kiselev
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
The opportunity to construct a Demonstration ADS in Russia with a high flux of thermal neutrons for
the effective incineration of long-lived fission products and minor actinides is considered. In this
connection, it is necessary to carry out international comparison of different versions of blankets for a
Demo ADS to define the main requirements for the ADS as a first stage for an international cooperation.
881

1. Scientific activity on ADS in world
Several research centres in different countries are carrying out scientific activities on ADS:
Belgium, Belarus, Czech Republic, France, Germany, Netherlands, Italy, Japan, Republic of Korea,
Russian Federation, Switzerland, Spain, Sweden, USA, and international scientific centres: CERN
(Geneva) and Joint Institute of Nuclear Research (JINR, Dubna). The following organisations in
Russia actively carry out investigations on different versions of ADS:
• State Scientific Centre of the Russian Federation Institute of Theoretical and Experimental
Physics (SSC RF ITEP, Moscow) – scientific leader on ADS in Russia.
• State Scientific Centre of the Russian Federation Institute of Physics and Power Engineering
named after A.I. Leipunski (SSC RF IPPE, Obninsk of Kalugskoi region) – scientific leader
on transmutation of radioactive waste (RW) in fast reactors in Russia, activity on ADS
target.
• Federal Nuclear Centre of the Russian Federation “Kurchatov Institute” (FNC RF KI,
Moscow) – investigation of ADS version with molten salt.
• Federal Nuclear Centre of the Russian Federation Institute of Experimental Physics (FNC RF
IEP, Arzamas-16) – nuclear data, study of safety of ADS.
• Federal Nuclear Centre of the Russian Federation Institute of Technical Physics (FNC RF
ITP, Chelaybinsk-70) – study of ADS version with active target.
• Research Design Institute of Power Engineering (RDIPE, Moscow) – conceptual study of ADS.
• State Scientific Centre of the Russian Federation Institute of Non-organic Materials named
after A.A. Bochvar (SSC RF INOM, Moscow) – development of fuel, structural materials,
reprocessing, vitrification, storage of RW.
• Radium Institute (RI, St-Petersburgh) – reprocessing, management of RW.
• Moscow Institute of Physics and Engineering (MIPE, Moscow) – conceptual study of ADS.
• Institute of Atomic Power (IAP, Obninsk) – conceptual study of ADS.
• Research Institute of Nuclear Research (INR, Troitzk of Moscow region) – experimental
study of ADS.
• Institute of Physics of High Energy (IPHE, Protvino of Moscow region) – experimental
study of ADS.
• Moscow Radiotechnical Institute (MRTI, Moscow) – study of ADS accelerator.
• Design Bureau of Building Machinery (DBBM, N. Novgorod) – design of ADS target-blanket.
• Design Bureau of Hydropress ( DBHP, Podolsk of Moscow region) – design of ADS target-blanket.
• Research and Design Institute of Power Technology (RDIPT, St-Petersburgh) – design of ADS.
Joint Institute of Nuclear Research (JINR, Dubna) is actively participating in Russian
programme on ADS too.
The current state of scientific activity and obtained results of conceptual investigations of ADS
allows to take conclusions on the principal opportunity for the development of a project for a
Demonstration Transmutation ADS (DTADS). The development of this project can be organised on
882

an international base to combine efforts of research centres and specialists of different countries. The
OECD/NEA, which has a long-term experience on organising such international projects, could be
one of these organisations of the international team on DTADS. Most obvious examples are the
activity of the OECD/NEA on the international study of physical features of ADS blanket and the
study on underground storage of RW in Europe and other projects.
2. On the opportunity of a DTADS construction in the Russian Federation
From our point of view, the site for DTADS construction must be near a reprocessing plant to
receipt transmuted RW immediately to manufacture the target. In connection with that, it is expedient
to consider 2 sites for construction of DTADS in Russia. The first site is the enterprise “Mayak”
located in South Ural, where there is the reprocessing plant RT-1, second – Krasnoyarsk Mine-
Chemical enterprise (KMCE) – where the construction of the reprocessing plant RT-2 is foreseen.
The SSC RF ITEP considered the opportunity of DTADS construction at these sites with the directors
of those enterprises. The directors of “Mayak” and KMCE agreed with our proposal with one
important requirement – providing financial support for the construction of DTADS.
The Scientific Council of Minatom was consulted and approved the large programme of activity
on ADS prepared by the SSC RF ITEP and IPPE. Professor E.O. Adamov as Chairman of Scientific
Council and Minister of Minatom approved this programme. He agreed the construction of DTADS in
the framework of an international co-operation including financial support.
It is necessary to note that according to the existing Russian Law on protection of environment it
is actually forbidden to import radioactive materials in Russia besides transportation from abroad and
reprocessing of spent fuel. But the Committee of Parliament is considering an amendment to this law,
which could allow the importation of radioactive materials in the Russian Federation for reprocessing.
The Law on protection of the environment does not prevent joint development of international project
and joint investigations on different kind of nuclear facilities.
3. Possible measures on organisation of joint co-operation for DTADS
In the case of a positive response to the proposal of the constitution of an international team on
the project of DTADS with the collaboration of OECD/NEA it would be convenient that OECD/NEA
organises a working group. This working group would set-up a preliminary study of all aspects and
formulation of technical tasks and organisational topics on the development of an international
project of DTADS for further consideration in participating countries, including the following
aspects:
• Technical: to choose the concept and principal scheme of DTADS.
• R&D grounding: to define a list, volume and schedule of time for R&D activity.
• Economics: to consider possible place for construction of DTADS in the Russian Federation
and possible consumption of money for development of projects.
• Political (legal): to define requirements and conditions for transportation of radioactive
materials in Russia and reprocessing.
• Public relation: to develop a corresponding programme.
On the base of this conceptual study it would be convenient to prepare a programme of activities
on DTADS creation for further consideration and approval.
883

4. Possible technical tasks on creation of DTADS
The general (principal/main ?) aim of DTADS is the transmutation of long-lived RW (LLRW) to
decrease the amount of LLRW before long-term storage or underground burial. Possible technical
tasks on creation of DTADS are presented below.
4.1 Possible principal schema of DTADS
The principal schema includes an accelerator, neutron-producing target, a blanket, and a trap of
protons in emergency. Following versions could be considered. Horizontal design – accelerator,
target-blanket, and trap of neutrons are located horizontally in one line. Vertical design – accelerator
is located horizontally, then proton beam is turned by magnet system into vertical direction, and it is
transported to blanket which is located below an accelerator (upper beam) or above an accelerator
(under beam). In both last versions there are two proton traps, the first one is after horizontal part of
proton beam, and the second one is after the vertical part of the beam. The capacity of these traps
must be approximately 10-100 MW.
4.2 Possible design of DTADS blanket
• First topic: level and neutron’s spectra. Thermal neutron spectrum is more preferable than
fast neutron spectrum. High flux more than 10 15 neutr/(cm 2 s) are preferable to middle-range
neutron flux about 10 14 neutr/(cm 2 s).
• Second topic: choice of coolant and moderator for thermal neutron spectrum. Light or heavy
water should be used rather than liquid metal coolant Na, Pb, Pb-Bi typical for fast spectrum.
• Third topic: heterogeneous or homogeneous blanket. It seems, as a base for consideration on
mastering in nuclear engineering that heterogeneous blanket is preferable.
• Fourth topic: choice of structure schema for blanket. It would be convenient to study the
following schema of blanket. First version – vessel type with H 2
O as coolant in central part
and D 2 O as moderator in reflector. Second version: channel-vessel type with H 2
O as coolant
in channels and D 2
O outside channels.
• Fifth topic: sectional blanket with compound neutron valve or without sections? It needs
additional investigation.
• Sixth topic: fuel: it seems that MOX-fuel is preferable.
• Seventh topic: target with transmuted radionuclides. As possible technical decision Tc,
dissolved into coolant (moderator) and cermets of MA.
There are numerous other topics: control for level of sub-criticality, distribution of neutron flux
on volume of blanket, water-chemical regime, schema of refuelling, etc.
4.3 Possible design of DTADS neutron-producing target
• First topic: choice of material for target. Ta, W and U in neutron-producing target should be
used. Alternative variant with Pb, Pb-Bi, molten salt could be rejected.
884

• Second topic: interaction of proton’s beam with target. There are two problems: choice of
form for target and division of proton’s beam at several beams or scanning on target.
• Third topic: choice of type for window between volumes of accelerator and target. Problems:
solid window, super-sonic window, without window.
• Fourth topic: target cooling. Water for target cooling is more preferable for thermal spectra in
blanket and water cooling of blanket on base of consideration on provision of safety of
DTADS.
There are number of other technical tasks on target.
4.4 Possible design of DTADS accelerator
There are following tasks for DTADS accelerator:
• First topic: warm or superconductivity of accelerating structure.
• Second topic: development of operation continuous mode.
• Third topic: development of frequency generator with high life.
• Fourth topic: provision of extremely low losses of protons during acceleration.
There are other topics too.
5. Possible stages of joint co-operation
The following main stages of joint co-operation on DTADS could be proposed:
• Comparison of existing versions of ADS for transmutation of LLRW and choice of main
version for further development in the following directions: blanket, neutron-producing
target, accelerator.
• Preparation of technical requirements for development of conceptual design.
• Preparation of draft of governmental agreement on joint co-operation.
• Development of conceptual design of DTADS.
• Preparation and execution of R&D Programme.
6. Conclusion
The proposed ideas concerning a joint co-operation on DTADS requires the consideration of the
interested participants and, in the first place of the international organisations: the OECD/NEA, IAEA
and others. After that it would be convenient to organise meetings on consideration of organisational
questions if positive attitude to this idea will be between specialists.
885

ON NECESSITY OF CREATION OF ACCELERATOR DRIVEN SYSTEM
WITH HIGH DENSITY OF THERMAL NEUTRON FLUX
FOR EFFECTIVE TRANSMUTATION OF MINOR ACTINIDES
A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
The opportunity for using a high thermal neutron flux for the effective incineration of fission products
and minor actinides is studied in this paper. Incineration rates of main long-lived fission products in
various fluxes are presented. An efficiency of actinide transmutation is studied for three types of
installations: thermal power reactor with neutron flux 5⋅10 13 neutr/(cm 2 s), fast neutron power reactor
with neutron flux 5⋅10 15
(neutr/cm 2 s), and homogenous heavy-water blanket of ADS with thermal
neutron flux 5⋅10 15 neutr/(cm 2 s).
887

1. Introduction
The role of accelerator driven systems (ADS) in promising atomic power engineering can be
very important. It is mainly explained by the fact that they decrease the amount of long-lived
radioactive waste. Specific properties of ADS permit to use different kinds of nuclear fuel such as
uranium, plutonium, thorium, minor actinides (and their combinations), and then to use various
nuclear fuel cycles. In ADS, long-lived fission products and actinides can be effectively transmuted.
Characteristics of transmutation process depend on parameters of the transmutation installation.
The main characteristics are neutron flux and spectrum and excess of neutrons, which can be used for
transmutation. High thermal neutron fluxes make possible to obtain high transmutation rates and
provide low loads of incinerated nuclides in the installation. ADS can provide also rather high excess
neutrons for transmutation. Unfortunately, there is no possibility to increase significantly the neutron
flux in fast neutron spectrum installations. However, fast neutron spectra could be preferable from the
view point of neutron balance.
The application of high-flux ADS for transmutation of long-lived fission products and actinides
is studied in this paper.
2. Fission product transmutation
Experts of different institutions came to an agreement about a list of long-lived fission products
that should be transmuted rather than ultimately stored. The main nuclides are 99 Tc and 129 I. Their halflife
is too long, and accumulation in spent nuclear fuel is rather high. Other radioactive fission
products either could not be transmuted effectively or have a low importance.
In Table 1, masses (in grams) of nuclides for transmutation of 99 Tc are presented for irradiation in
constant neutron flux Φ = 10 14 neutr/(cm 2 s) and spectrum hardness γ = 0.4 typical for a light water
blanket of an ADS (spectrum hardness is the share of epithermal neutrons in neutron flux). Data are
normalised by 1 gram of initial 99 Tc. In Table 2, the same data are presented for Φ = 10 15 neutr/(cm 2 s)
and spectrum hardness γ = 0.1 typical for heavy water blanket of ADS.
In Tables 3 and 4, analogous data for transmutation of 129 I are given. They are normalised by 6.35 grams
of initial 129 I and 1.40 gram of initial 127 I. Iodine isotopes are accumulated in spent fuel of light water power
reactors in those amounts on account of power produced 1 MW-year.
Table 1. Transmutation of 99 Tc with Φ = 10 14 neutr/(cm 2 s) and γ = 0.4
T, year
99
Tc
100
Ru
101
Ru
102
Ru
103
Rh
107
Pd
0 1 0 0 0 0 0
1 6.11-1 3.83-1 5.94-3 2.85-4 4.89-7 2.14-14
2 3.74-1 6.05-1 1.93-2 1.95-3 5.23-6 4.15-12
3 2.28-1 7.30-1 3.57-2 5.68-3 1.85-5 8.25-11
4 1.40-1 7.96-1 5.26-2 1.16-2 4.25-5 6.49-10
5 8.53-2 8.26-1 6.86-2 1.98-2 7.77-5 3.09-9
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Table 2. Transmutation of 99 Tc with Φ = 10 15 neutr/(cm 2 s) and γ = 0.1
T, year
99 Tc
100 Ru
101 Ru
102 Ru
103 Rh
107 Pd
0 1 0 0 0 0 0
0.1 8.43-1 1.55-1 1.52-3 2.19-5 2.56-8 2.37-17
0.2 7.11-1 2.83-1 5.65-3 1.65-4 3.39-7 5.37-15
0.5 4.26-1 5.43-1 2.83-2 2.18-3 8.08-6 5.83-12
1 1.82-1 7.25-1 8.00-2 1.33-2 6.63-5 9.08-10
2 3.31-2 7.29-1 1.70-1 6.56-2 3.86-4 1.02-7
Table 3. Transmutation of 129 I with Φ = 10 14 neutr/(cm 2 s) and γ = 0.4
T, year
129 I
130 Xe
131 Xe
132 Xe
133 Cs
134 Cs
135 Cs Ba
0 6.35 0.0 0.0 0.0 0.0 0.0 0.0 0.0
1 5.57 7.40-1 2.51-2 1.32-2 2.26-5 2.31-6 2.12-7 6.08-7
2 4.89 1.32 6.55-2 7.65-2 2.44-4 4.05-5 7.97-6 2.33-5
3 4.29 1.77 1.02-1 1.94-1 8.60-4 1.79-4 5.57-5 1.68-4
4 3.77 2.11 1.32-1 3.57-1 1.95-3 4.61-4 2.00-4 6.19-4
5 3.30 2.36 1.54-1 5.55-1 3.50-3 8.98-4 5.05-4 1.61-3
Table 4. Transmutation of 129 I with Φ = 10 15 neutr/(cm 2 s) and γ = 0.1
T, year
129
I
130
Xe
131
Xe
132
Xe
133
Cs
134
Cs
135
Cs
0 6.35 0.0 0.0 0.0 0.0 0.0 0.0 0.0
0.1 5.76 5.60-1 2.09-2 4.08-3 2.90-6 1.03-7 8.31-9 2.61-8
0.2 5.23 1.02 6.64-2 2.75-2 3.85-5 2.21-6 3.73-7 1.18-6
0.5 3.92 1.94 2.22-1 2.69-1 8.94-4 7.92-5 3.71-5 1.22-4
1 2.42 2.48 3.64-1 1.11 6.57-3 1.66-3 1.63-3 6.91-4
2 9.21-1 2.05 3.40-1 3.09 2.85-2 9.37-3 2.12-2 1.24-2
These data show that in a middle-range neutron flux, the rate of 99 Tc and 129 I transmutation is not
high. In transmutation of 99 Tc, isotopes of Ru, Rh and Pd are produced, while in transmutating 129 I,
isotopes of Xe, Cs and Ba are produced. High flux provides much higher rate of transmutation.
Ba
3. Actinide transmutation
As for MA incineration, there are various viewpoints concerning the most preferable neutron
spectrum. The choice of MA transmutation conditions is important to define a type of reactor or ADS
installation. From this point of view, it is convenient to compare an efficiency of actinide transmutation
in different facilities. In Table 5, calculated characteristics of transmutation modes in 3 types of
transmutation facilities are presented: thermal power reactor PWR with neutron flux 5⋅10 13 neutr/(cm 2 s),
fast neutron power reactor with neutron flux 5⋅10 15 neutr/(cm 2 s), and homogeneous heavy-water blanket
of ADS with thermal neutron flux 5⋅10 15 neutr/(cm 2 s). These data were obtained by experts of ITEP and
889

MEPI within the framework of Project of ICST #17 [1]. Two types of feed by actinides were considered:
total 39 kg/year with isotopic composition 45% 237 Np, 44% 241 Am, 8.5% 243 Am, 2.2% 244 Cm, and total
83 kg/year with isotopic composition 57% 241 Am, 30% 243 Am, 11.5% 244 Cm. Note that second feed is
typical for actinides from spent uranium-plutonium fuel.
Table 5. Transmutation characteristics
Type of installation Thermal neutrons Fast neutrons High flux ADS
Neutron flux, neutr/(cm 2 s) 5⋅10 13 5⋅10 13 5⋅10 15 5⋅10 15 5⋅10 15 5⋅10 15
Feed by actinide, kg/year 39 83 39 83 39 83
Time of equilibrium, year 50 50 40 40 0.5 0.5
Equilibrium actinide mass, kg 700 2 300 880 2 100 2.7 2.3
Equilibrium respective radiotoxicity 1 3.1 1.6 3.4 0.03 0.1
Reference time, year 250 170 400 200 4 2
The main feature of MA transmutation is that the radiotoxicity of incinerated actinides is
increased in the initial period of irradiation because of higher radiotoxic nuclide production. After
some period, equilibrium is established. In equilibrium, the rate of actinide incineration is equal to
actinide feed. In the installation, there is an equilibrium actinide mass and radiotoxicity. The time of
equilibrium achievement is almost the same both for thermal and fast neutron reactors. The high
thermal flux installation has an obvious advantage, because the time of equilibrium achievement is
significantly (up to 100 times) shorter than for other types of installation. Radiotoxicity is presented
in respective units, it is normalised by initial value for thermal neutron installation for a first type of
feed. Equilibrium actinide mass and radiotoxicity is also much more preferable for high flux ADS
installation, it is 40-50 times less than for common-type thermal or fast neutron installation.
An interesting parameter is the reference time. It is a parameter for comparison of processes of
transmutation and storage. If actinides are located into transmutation installations with constant rate
(feed) then in some time equilibrium radiotoxicity in the installation is established. If actinides are
located in storage with the same rate then radiotoxicity of actinides in storage uniformly increases.
Reference time in Table 5 is a time when a radiotoxicity would be equal to that of in transmutation
installation. So, for common-type installation, reference time makes 200-400 years. It means that
during 200-400 years, simple actinide storage with uniform addition in storage facility gives us less
radiotoxicity than that of established in transmutation installation. However, transmutation in high
flux ADS installation will be preferable after 2-4 years of transmutation.
Table 6. Respective radiotoxicity in irradiation of plutonium
T, days 10 14 neutr/(cm 2 s) 10 15 neutr/(cm 2 s)
0 1 1
100 1.9 5.4
200 2.0 14
300 1.8 18
500 2.1 13
700 2.7 7.1
1 000 5.4 2.2
890

Another important problem concerning to radiotoxic actinide accumulation is weapon-grade
plutonium utilisation. In Table 6, respective radiotoxicity of actinides in process of almost pure 239 Pu
irradiation is presented. Neutron flux 10 14 and 10 15 neutr/(cm 2 s) and neutron spectrum of heavy-water
ADS blanket are considered. There is an increase of radiotoxicity during a long period of irradiation.
It is caused in an initial period by accumulation of 241 Pu and then by accumulation of 244 Cm. A major
part of radiotoxicity is produced by 244 Cm. Maximal radiotoxicity is 18 times higher than that of initial
plutonium.
4. Conclusion
The data presented show that both long-lived fission products and actinides can be successively
transmuted in ADS installations. For transmutation of fission products, a most important feature of
ADS is high amount of excess neutrons, which can be captured in transmuting nuclides. Low
equilibrium amount nuclides does not play so important role as in transmutation of actinides.
For actinide transmutation, a high flux ADS installation is needed. The low equilibrium
radiotoxicity of incinerated actinides and high level of nuclear safety preventing reactivity accident
are the main features of ADS important for transmutation facility.
REFERENCES
[1] B. Bergelson et al., Transmutation of Minor Actinides in Different Nuclear Facilities. Proceedings
of the International Workshop “Nuclear Methods for Transmutation of Nuclear Waste”, Dubna,
Russian Federation, 29-31 May 1996. World Scientific Publishing Co., 1977, pp. 67-76.
891

NEW ORIGINAL IDEAS ON ACCELERATOR DRIVEN SYSTEMS IN RUSSIA AS BASE
FOR EFFECTIVE INCINERATION OF FISSION PRODUCTS AND MINOR ACTINIDES
G.V. Kiselev
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
The analysis of original ideas on ADS proposed by Russian specialists in the last time is given for a
joint consideration by specialists dealing with ADS. Ideas are concerned by the following problems of
ADS design: accelerator, target, sub-critical blanket, sectioned blanket, necessity of high neutron flux
for transmutation of long-lived fission products and minor actinides, two-blanket installation with
fluid fuel, denaturation of weapon-grade plutonium by joint irradiation with neptunium, use of
secondary nuclear fuel in ADS with reduced requirements to level of purification of spent fuel.
893

1. Introduction
During the last years leading Russian nuclear centres conducted conceptual researches of various
variants of electronuclear installations (ADS). This study stimulated occurrence of a number of
original technological ideas, directed on improvement of the characteristics ADS with the purpose,
first of all, of the effective destruction of long-lived radioactive waste (LLRW). The review of these
offers is not only a question of priority but is also of interest for the development of ADS future
projects. In the beginning of the review the author presents a brief description of ADS technological
schema with an indication of the current problems related to the realisation of ADS.
2. General aim of ADS
From the author’s point of view, the main purpose of ADS consists of the effective destruction
(transmutation) of long-lived products of fission (FP) and minor actinides (MA) up to level at which
probably radiation – equivalent burial of LLRW – as it is offered by the experts of the RDIPE [1]. If
to proceed from indicated purpose, it is necessary to develop the concept and, accordingly, design
blanket with high flux of thermal neutrons, as it is shown in many papers of experts of the SSC RF
ITEP and MEPI, see, for example, [2]. The creation of such high flux blanket by thermal capacity not
less than 1 000 MWt with density of thermal neutron flux 10 15 cm -2 c -1 and higher with small campaign
for fuel is already enough complex technological problem, even at modern level of nuclear
engineering. However it does not reach the list of technological problems, claiming their decision and
experimental study.
After these brief remarks, we shall pass to a compendious description of new, original ideas,
concerning to ADS and formulated by the Russian experts in different time. We shall originally stay
on exposition of the offers, relating to ADS blanket, as its design in many respects determines the
basic schema and parameters of ADS.
3. Sub-critical blanket of ADS
3.1 Sectional blanket
It is possible to note without exaggeration, that really technically the revolutionary idea concerns
to blanket of ADS, i.e. division of blanket on multiplied sections with one-sided neutron connection.
The idea of the neutron multiplier, stated for the first time more than 40 years ago by L.B. Borst [3],
was originally supposed to be used in systems with “initiating” reactor for reception of higher burn-up
for fuel in sub-critical sections and for reception of extremely high flux of thermal neutrons. In USSR
Dr. B.G. Dubovsky from the SSC RF Institute of Physics Power Engineering (IPPE) in Obninsk has
briefly considered this topic for the first time of activity on atomic power [4]. The significant theoretical
contribution to development of idea of sectioning or connected reactors or reactors with one-sided
neutron connection for pulsed systems and ADS was proposed by the experts of the Federal Centre
Institute of Experimental Physics (FNC VNIEP, Arzamas-16) under management of V.P. Kolesov. In
papers of experts of the FNC VNIEP a basic opportunity of essential increase of a level of multiplication
of neutrons for an external source and decrease of requirements to proton’s current in 5-10 times for
sectional blanket with by loading of neptunium in first fast section was shown [5].
894

During conceptual investigations of ADS the experts of SSC RF ITEP N.M. Danilov,
Yu.D. Kàtargnov, G.V. Kiselev, V.V. Kushin, V.G. Nedopekin, S.V. Plotnikov, S.V. Rogov and
I.V. Chuvilo and the expert of the FNC Institute of Technical Physics (VNIITP), K.F. Grebenkin have
offered a schema of ADS sectional blanket with compound neutron valve (NV) beginning 1993. The
authors of this schema received the patent of Russian Federation with priority from 27.04.1993 [6].
According to the offered schema for the first multiplying section (2) of blanket of cylindrical form with
fast spectrum of neutrons contains in centre of neutron-producing target, and on periphery – NV,
separating from second of multiplying section with thermal spectrum of neutrons. Compound neutron
valve NV consists of 2 parts: an absorber of thermal neutrons, for example, boron by thickness 2-5 mm,
bounded with first multiplying section, and moderator of thermal neutrons, for example, graphite by
thickness 30-50 cm, contiguous to second section, and is executed as continuous ring cylinder. The role
of compound NV consists of maintenance unilateral neutron connection between sections of blanket. In
result of investigations conducted by the various Russian experts it is possible to indicate following
physical properties of sectional fast- thermal blanket:
• Multiplying property of thermal section poorly influence number of neutrons, born from
fissions in fast section.
• The change of physical properties of thermal section poorly influences factor of
multiplication for complete system.
• There is the limiting ratio between capacities of fast and thermal sections.
• There is the basic opportunity of decrease of a current of protons in case of realisation
sectional blanket, by various estimations, from 2 up to 5 times.
However allocation of compound NV with layer of an absorber of neutrons, for example, from
carbide of boron, results, first, in deterioration of the neutron balance in system, that, certainly, it is
undesirable. Secondly, there is the problem of nuclear safety in case of probable emergency
destruction of compound NV. Therefore expediently to consider a problem about optimum method of
suppression of a feedback between sections, without essential deterioration of the neutron balance and
maintenance of significance k eff
always it is less than unit in case of occurrence of an emergency and
destruction compound NV.
In this connection the specialists of the SSC RF ITEP, G.V. Kiselev and MIPE V.A. Apse,
G.G. Kulikov and A.N. Shmelev have conducted calculations of a compound NV, which does not
worsen the neutron balance and at the same time provides sub-criticality of blanket in emergency.
With this purpose was considered compound NV with use of depleted uranium oxide (first variant)
and oxide of neptunium (second variant) in an absorber part NV [7]. It was shown (not staying on
interesting and, at the same time, transparent physics), that the application of depleted uranium oxide
does not provide necessary nuclear safety. Use compound NV, in structure of which was present
oxide of neptunium-237, does sectional blankets of ADS completely safe at emergencies, as
according to calculations destruction of a absorber layer of compound NV from oxide of neptunium
results in reduction k eff
.
In paper of the specialists of SSC of the RF “Kurchatov Institute” P.N. Alekseev, V.V. Ignatjev,
O.E. Kolayskin and other [8] is indicated opportunity of use of lanthanide’s GdF 4
, SmF 4
as means for
suppression of a feedback between sections of ADS blanket, which follows to include into structure
first (fast) section, that completely, in their opinion, provides conditions of nuclear safety at
emergency.
895

Other original proposal of the indicated above experts of the SSC RF ITEP and FNC VNIITP on
sectional blanket of ADS concerns the mutual location of neutron-producing targets in sections of
ADS blanket and use dipole triplet for management of a flux of protons [9]. In traditional
configuration of ADS the neutron-producing target is placed in centre of vertical or horizontal
blanket. In offered by the indicated authors [9] to the constructive schema of ADS some neutronproducing
target in regular intervals place on volume external ring of multiplying section of the
cylindrical form blanket. In each section are available CNV, passed through neutrons on direction
from peripheral section to centre blanket, that permits to reach high density of a neutron flux in
irradiated volume, were in centre of blanket, down to 10 16 cm -2 s -1 . For realisation of this idea between
horizontally located accelerator of protons, working in pulsing mode, and vertical blanket is entered
so-called dipole triplet- system for distribution of a protons beam on targets. The dipole triplet
consists from first dipole, executed in kind rotary on 90° magnets for change of a direction of a
proton’s beam from horizontal to vertically located target. The second and third dipoles of triplet
serve for circular distribution of a proton’s beam on targets and are executed in kind of a pair of highcycling
rejecting magnets, the phases of currents of which are moved on 90°, and the frequency F of
sending of current pulses for start-up of the accelerator and frequency f of a current in windings of
magnets are connected by a ratio F = nf, where number of targets n = 3-16. Availability of shift on
phase 90° between currents in windings of magnets with frequency f, synchronised with frequency of
repetition of accelerator pulses F, results in that a pulsing current protons turned on circle and the
pulses of a proton’s current consistently interact with neutron-producing target.
The authors of the patent [9] specify following advantages of the offered schema sectional
blanket with dipole triplet:
• Basic opportunity of achievement of high density of a neutron’s flux in irradiated volume,
were in centre of blanket, that will allow to execute effective transmutation of LLRW.
• Opportunity of reduction of a accelerator current about in 10 times for three- sectional
blanket with k eff = 0,97 in comparison with traditional configuration blanket without sections,
that will allow to reduce a proton’s current and heat deposition of neutron-producing targets.
• To reduce consumption of power for supply of the accelerator at least in 5 times and more.
The main drawback of the indicated offer is the necessity of the introduction in ADS schema of
appropriate blocking and additional trap of a proton’s beam in case of loss of power supply or failures
in dipole triplet operation. One trap of a proton’s beam should be stipulated in any case at horizontal
configuration of the accelerator protons and vertical blanket.
The important aspect of ADS nuclear safety provision during normal operation and transitive
modes is the control of a level of blanket sub-criticality. This control acquires a special significance
for ADS sectional blankets. For this purpose it is possible to use the offer developed by the experts of
the SSC of RF ITEP N.M. Danilov, Yu.D. Katargnov, G.V. Kiselev, V.V. Kushin, V.G. Nedopekin,
S.V. Plotnikov, S.V. Rogov and I.V. Chuvilo, on the use of special reactivity meter for sectional
blanket and gauge of measurement of a current protons of the accelerator, the description of which is
indicated in patent [10]. For illustration of the offered idea we shall consider three-sectional blanket
of ADS with two CNV, passed through neutrons on direction from neutron-producing target to
periphery of blanket. The reactivity meter for control of a sub-criticality level of each section operates
together with gauge of a current protons and synchroniser mode of operations of the accelerator and
consists from integrator-meter, connected through blocks of the co-ordination with neutron detectors,
available a minimum on one in each multiplying section, block of formation of signals, block of
896

delay, interface and computer. The feature of the offered schema of measurement of a sub-criticality
level consists that the pulse of a current in detector of neutrons in first multiplying section arises
practically simultaneously with pulse from gauge of a proton’s current. The pulses of a current in
detector of neutrons in second and third multiplying sections are late on time of slowing down of
neutrons in CNV. By estimation this time of delay makes about 100 ms for second section and can
reach 200 ms for third section. According to it temporary characteristics of appropriate pulses of
synchronisation in block of delay are established. At each pulse of a accelerator current the computer
processes the indications of all integrator-meters and calculates current significances of effective
multiplication factor k ef.i (t) for each multiplying section, which are compared with given benchmark
size k ef.i . By results of comparison a managing command on rods of system for monitoring and control
is issued.
Expediently to develop the specific schema offered reactivity meter and to conduct its
experimental examination in conditions of critical assembly. The proposed schema can be base for
development of system of monitoring and control for ADS in future.
The idea on sectional blankets of ADS was investigated in various Russian nuclear centres
during last years.
The experts of the SSC of the RF “Kurchatov Institute”, P.N. Alekseev, V.V. Ignatjev,
O.E. Kolayskin and others have offered fast-thermal cascade molten-salt blanket with two multiplying
sections and internal properties of safety: in first section salt is used NaF53-ZrF 4
41- MF 4
(4-6), in second
section – LiF69-BeF 2 28- MF 3 (3-5) (are indicated vol. shares) [8]. Using indicated sectional blanket
permits to receive high multiplication of neutrons (before 1500), that enables to consider opportunity for
application of electronic accelerators as external source of neutrons, though use proton’s cyclotron, as
means of reduction of the investments and operational costs is possible, on that specify the authors [8].
The experts of the FNC VNIITP under management of K.F. Grebenkin have offered the schema of
ADS with unilateral connection in kind a multiplying target with k eff = 0.95 and blanket with k eff = by
0.92 and thermal capacities 500 MWt [11]. In one of variants loading in a multiplying target of nuclear
fuel from oxide or nitride of uranium for type fuel assemblies of fast reactor with 20% by enrichment on
235
U is stipulated, liquid lead as coolant is considered. By evaluations, by use of the offered schema
probably to reduce requests to current protons in 3-5 times, that increases realisation of ADS.
The experts of the SSC RF ITEP B.P. Kochurov, O.V. Shvedov and V.N. Konev have offered twosectional
blanket of ADS, consisting from internal section as pool type, contiguous to target and having fast
spectrum of neutrons, with Pb-Bi eutectic as material of a target and coolant for the first section, and outside
section with thermal spectrum of neutrons and heavy water as coolant in channels under pressure [12]. As fuel
in both sections is used weapon-grade plutonium or mix weapon-grade and power plutonium. Technetium-99
is entered into coolant of thermal section for purposes transmutation and simultaneously the compensation of
reactivity. On calculations the thermal and electrical capacity of fast section is equal 300 and 120 MWt,
thermal section – 3 000 and 580 MWt accordingly. The steel wall of fast section and internal wall of thermal
section serve NV, providing of function unilateral neutron connection. The significance k eff
of this system
makes 0.99 at k eff
in thermal section 0.95, that permits to use a proton’s current 10 mA at energy 1 GeV. Deep
burn-up is about 70 GWtd/t in fast section and 35 GWtd/t in thermal section is reached by operation on
nominal capacity during about 6 years. For this period more than 90% initial plutonium in thermal blanket
transforms to fission products or minor actinides. During 30-years of period are transmuted about 25 t weapongrade
plutonium and 3 000 kg 99 Tc.
897

Other original offer about sectional molt-salt blanket of ADS by thermal capacity 2 000 ÌWt and
k eff
0.98 was investigated by the experts of the SSC of the RF VNIINM named after A. Bochvar
V.I. Volk, A.Yu. Vahrushin and experts of the SSC RF ITEP, B.P. Kochurov, O.V. Shvedov,
A.Yu. Kwarazheli and V.N. Konev [13]. Liquid salt 66PbF 2 -NaF with dissolution up to 10% of heavy
particles (Pu, MA) as material of a target and carrier blanket, with temperature melting 498°C was
considered.
Taking into account perspective of sectional sub-critical blanket of ADS expediently to provide
continuation of their study and organisations of experimental researches of physics of this systems,
including topic of safety.
3.2 Various variants sub-critical blankets of ADS
Except sectional blankets the Russian experts have considered a number of constructive
schema’s non-sectional blankets of ADS, which have novelty and differ from schema’s, investigated
by the foreign experts. Solid-fuel and liquid-fuel (or in kind of water solutions of salts or molten-salt)
blankets with various coolant (helium, sodium, lead, Pb-Bi eutectic) and moderators (heavy and light
water) were investigated.
3.2.1 Variants blankets, investigated in the SSC RF ITEP
The experts the SSC RF ITEP have conducted conceptual investigations of ADS for the
following variants of blankets with solid and liquid fuel, for which common is first, the vertical
configuration, secondly, location of a neutron-producing target in central part of blanket.
1. One of original offers, developed by the experts of the SSC RF ITEP in 1985, concerns the
combination (or integrated) schema target-blanket of ADS as one vessel [14]. The combined
design of target-blanket represents vertical vessel without pressure, inside which circulates
from below – upwards Pb-Bi eutectic. As nuclear fuel and simultaneously a target material
(alongside with Pb-Bi) is used depleted uranium in kind fuel of the spherical form as reactor
HTGR fuel, weighted in flux of the coolant and interacting with proton’s beam. The offered
schema permits to exclude a steel wall between volumes of a target and blanket and, thus, to
exclude a problem of its radiating damage and replacement.
Late, during activity on project ICST #17 following integrated schema’s as target-blanket of
ADS as vessel were considered:
−
−
Combined target-blanket with Pb-Bi-eutectic as material of a target and molten-salt NaF-
ZrF 4
-PuF 3
as carrier in blanket [15].
Homogeneous blanket with solution of fuel in heavy water for ADS UTA with high density
of thermal neutrons about 5⋅10 15 cm -2 s -1 , which the experts of Reactor Department of the
SSC RF ITEP, B.R. Bergelson and others have justified [16]. On accounts of the authors
one installation UTA can destroy MA from 40-45 reactors of a type VVER-1000 (PWR).
898

2. Channel-vessel design of ADS blanket with heavy water as coolant and (or) moderator is
developed in following updating:
−
−
As designs of domestic industrial heavy-water reactors by thermal capacity near 1 000 ÌWt
and density of a flux of thermal neutrons 5⋅10 14 cm -2 s -1 and as core of heavy-water reactor
CANDU by thermal capacity 2 064 MWt with various kinds of fuel (enriched uranium,
weapon-grade and power Pu, Th) and targets in solid phase [17].
Heavy-water blanket with channels-modules, inside which compulsory circulate of liquid
fuel in kind of a solution or slurry in heavy water with help of special independent
pumps, located in the bottom part of each module is carried out. Each module has
individual lines of extraction of a fuel mix for purification from fission products [18].
− As core of heavy-water reactor CANDU with molten salt 7 LiF-BeF 2 -PuF 4 (77%-22%-1%)
and heavy-water moderator, in which is dissolved 99 Tc-for maintenance of a given subcriticality
level [19].
The number of the constructive schema’s of blankets, except indicated above sectional blankets,
was offered by the experts of other leading Russian nuclear centres.
Indicated here short description of the various conceptual original offers of the Russian experts
under constructive schema sub-critical blanket of ADS testifies to availability enough justified base
for realisation of comparison of these variants and subsequent choice of variant for development of
the project for demonstration ADS.
4. Neutron-producing targets of ADS
Except constructive schema of neutron-producing targets, indicated in the previous section
(integrated configuration of target-blanket and the active target with fuel assemblies of fast reactors,
offered by FNC VNIITP), in the Russian nuclear centres were studied the following variants of targets.
4.1 Variants of targets, investigated in the SSC RF ITEP
The experts of the SSC RF ITEP independently and with the participation of other organisations
have offered a number of various variants of ADS neutron-producing targets:
Tungsten solid target of the cone form, available in centre of heavy-water blanket, with cooling
by heavy water and proton’s current up to 30 mA [20]. The development of the constructive schema
of this target is executed by the experts Design Bureau of Building Machinery (DBBM,
N. Novgorod).
Leaden solid target in kind of a small diameter balls, hydraulic weighted in a interaction zone
with beam of protons by pressure of heavy-water coolant [21]. Around target a buffer zone in kind of
a layer of heavy water by thickness 50-70 mm with the purpose of break of connection on fast
neutrons between target and blanket is provided. An absorber of neutrons (boron or gadolinium) for
interruption of chain reaction in emergency case of blanket can be entered into buffer zone. This
variant of a target permits to remove 62.7 MWt of thermal capacity at energy of protons 1 GeV and
current 100 mA. The target has a diameter of 500 mm and height of a layer of 2 000 mm,
899

concentration of lead in heavy water about 0.3 at diameter of leaden spheres of 5 mm.
Lead-bismuth eutectic (Pb 44.5% + Bi 55.5%) target by capacity from 77 up to 116 MWt were
offered by the experts of the ITEP and MEPI [22], a feature of which is availability of a window of
the cone form from beryllium, sharing volumes of accelerator and target. As the authors [24] specify,
the cones form of a window permits, on the one hand, to receive distributed on axis of blanket a
source of converting neutrons, with other, to reduce about in 5 times fluence of fast neutrons by
constructive elements of a target and blanket.
The vertical tantalum target by capacity 25 MWt executed a kind of set constituted by 15 flat
horizontal cylindrical disks of a thickness of 30 mm each, with internal channels for pass of the
coolant [23]. The target was developed for sub-critical light water blanket as core reactor PWR.
4.2 Design of the IPPE and the Design Bureau Hydropress (DBHP) target
The experts of the IPPE named after A.I. Leipunsky with participation of the experts of the
DBHP were developed 2 conceptual projects of liquid targets: with lead on capacity 10 MWt and with
Pb-Bi eutectic on capacity 25 MWt at energy of protons 1 GeV and current 25 mA. Advantage of a
liquid target with Pb-Bi eutectic is availability of technology, developed for reactors of nuclear
submarines. Defect of Pb-Bi coolant for a target is formation of polonium-210, defect of lead is high
temperature melting (about 327 ° C) and high thermal expansion. A serious problem of liquid targets at
high significance of heat deposition is regeneration (purification) of a target material from formed
products of nuclear reactions. Now the experts of the IPPE with the participation of the DBHP
execute experimental check of a Pb-Bi target on basis of ICST grant.
5. Accelerator of protons
5.1 Linear accelerator of protons
The modern variant of the block diagram for the linear accelerator of protons (LPA) for ADS with
current and energy of protons 30 mA and 1 GeV accordingly, developed by the SSC RF ITEP and
Moscow Radiotechnical Institute (MRTI), is constructed under single-channel circuit [24]. The
accelerator consists from injector, initial, intermediate and main parts. A basic feature of the offered
circuit is use of superconductivity (SP) resonators with low gap in intermediate part for acceleration of
particles up to 100 MeV instead of traditional long multigap resonators, which were provided in initial
variants LPA. It permits to choose distances between centers of accelerating gaps outside of dependence
on speed of particles, that permits to reduced length of section to 30-40%, but also to continue process of
acceleration even at failure of small amount of accelerating resonators or path’s of their independent
power supply, having postponed their repair up to scheduled stop. Basically, cylindrical resonators with
drift tubes in intermediate part can be used. In main part, for acceleration of particles up to 1 GeV is
provided to use multigap SP resonators. The project on accommodation in each cryoinstallation not less
than 2 resonators is developed in the MRTI, each of which, in turn, consists of 9 accelerating cells of the
ellipsoid form. Structurally SP resonators it is supposed to execute from niobium. Length and diameter
of one resonator 0.41-1.12 m and 29-26 ñm accordingly. A rate of acceleration on length of the resonator
from 5 up to 15 MeV/m can be achieved. The excitation of SP resonators is supposed to execute from
clistron amplifiers by capacity 1.2 MWt each. The main calculation characteristics full-scale LPA
(1 GeV, 30 mA) with SP accelerating resonators in main part [24] are indicated in Table 1.
900

Table 1. Characteristics of the full-scale accelerator with SP resonators
Parameter Warm SP,
5 MeV/m
SP,
15 MeV/m
Approximate length of LPA, m 1000 400 135
Efficiency of resonators 0.4 1 1
Efficiency of HF-generators with power supply 0.65 0.65 0.65
HF power, MWt 75 33 33
Efficiency of LPA 0.2 0.6 0.55
Cost of accelerating system, mln. dol. 50 69 23
Cost of HF-generators with power supply, mln. dol. 125 49.5 49.5
Cost of the non-standard equipment, mln. dol. 275 120 72
Total cost of the equipment, mln. dol. 313 179 109
Total cost of LPA, mln. dol. 437 233 142
From Table 1 it is obvious, that using a SP accelerating system and high rate of acceleration
results in significant reduction of LPA cost up to 233-142 mln. dol. in comparison with cost of the
“warm” accelerator 437 mln. dol. The offered technical decision on constructive circuit of LPA can
render essential influence to technological characteristics of ADS.
6. Conclusion
The information on new original offers of Russian experts on ADS different units shows, at first,
on high potential of Russian atomic science and engineering, at second, on existence of scientific base
for development of project for Demonstration Transmutation ADS. But it is necessary to continue
conceptual comparison study of ADS different versions and carry out large volume of R&D activity.
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901

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902

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903

CONDITIONS OF PLUTONIUM, AMERICIUM AND
CURIUM TRANSMUTATION IN NUCLEAR FACILITIES
A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova, T.S. Zaritskaya
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
Weapon-grade plutonium burning and transmutation of the americium and curium isotopes from
spent nuclear fuel in reactor or accelerator-driven installations with various neutron fluxes and spectra
are analysed. The concentration of the nuclides up to 248 Cm and the radiotoxicity are calculated. The
problem of plutonium burning and minor actinides transmutation is that the radiotoxicity is increased
in the beginning of irradiation, and only after a period it decreases to the initial value.
905

1. Introduction
The problem of nuclear transmutation of minor actinides accumulated in spent nuclear fuel is
part of the problem of radioactive waste management. Weapon-grade plutonium released as result of
conversion of weapons programmes can effectively be used as nuclear fuel. However at burn-up of a
plutonium, significant amounts of transplutonium nuclides will be formed with essentially more high
levels of radioactivity and radiotoxicity than initial plutonium. The transmutation of neptunium
results in high-toxic 238 Pu. The main difficulties can arise with americium and curium. At irradiation
by neutrons, they are transformed to heavier nuclides and are partially fissioned.
In connection with the importance of minor actinide transmutation problem, it is necessary to
determine the main features of the process of transmutation of americium and curium and burn-up of
plutonium in various conditions.
2. Long-term plutonium burning
A mode of long-term continuous burning of almost pure 239 Pu in a spectrum of heavy-water
reactor, with constant neutron flux density Φ = 10 14 n/cm 2 s, is studied in order to clarify the role of the
produced minor actinides. Irradiation time is 1 000 days. In order to look after burn-up for curium
isotopes, irradiation in the high flux – 10 15 n/cm 2 s with same spectrum is also considered.
The relative values of total radiotoxicity in water (normalised by initial radiotoxicity of
plutonium) of isotopes of a plutonium, americium and curium (up to 248 Cm) are summarised in
Table 1. T is the time of irradiation. The radiotoxicity RT i
of each nuclide i in water is determined by
the ratio:
RT i
= A/MPA i
where A i
– activity of considered amount of a nuclide i, MPA i
– maximum permissible activity of this
nuclide in water according to radiation safety standards accepted in Russia [1]. Total radiotoxicity is
equal to a sum of radiotoxicities of all nuclides taken in those amounts in which they are contained in
the considered mix of nuclides.
Table 1. Total radiotoxicity in plutonium irradiation
modes with neutron flux 10 14 and 10 15 n/cm 2 s
Irradiation time,
days
10 14 n/cm 2 s 10 15 n/cm 2 s
0 1 1
100 1.9 5.4
200 2.0 14
300 1.8 18
500 2.1 13
700 2.7 7.1
1 000 5.4 2.2
The time dependence of the radiotoxicity as shown in Table 1 has two maxima. The first rather
weak maximum at Φ = 10 14
n/cm 2 s and T = 200 days is determined by a nuclide 241 Pu which is
accumulated at irradiation of plutonium. Its radiotoxicity per gram is 32 times more than that of 239 Pu.
906

The further increase after 500 days is determined by accumulation of 244 Cm whose radiotoxicity per
gram is 600 times more than that of 239 Pu. The contribution of 244 Cm in total radiotoxicity after
700 days of an irradiation exceeds 90%, after 1 000 days more than 99%. The second maximum of
the radiotoxicity is determined by 244 Cm (at Φ = 10 15 n/cm 2 s and T = 300 days). In the calculation with
Φ = 10 14 n/cm 2 s corresponds to T = 8.2 years. The maximum radiotoxicity exceeds the radiotoxicity of
initial plutonium by 18 times. For longer irradiation, the radiotoxicity decreases slowly. For
irradiation of 27 years (and Φ = 10 14 n/cm 2 s), radiotoxicity exceeds 2.2 times initial one. The total
amount of plutonium after 1 000 days results in 6.2%, americium, 1.5%, curium, 1% of the initial
plutonium. At the time of the 244 Cm maximum, plutonium makes 1.7% (basically 242 Pu), americium
1% ( 243 Am), 244 Cm 3.3%, 246 Cm 0.3% of initial plutonium.
3. Single burn-up of americium and curium
Two modes of irradiation of americium and curium with constant neutron flux density are
considered: an irradiation in a thermal spectrum, typical for light water reactor, at several values of
neutron flux density; and in a fast spectrum typical for BN-800 reactor [2].
The initial masses of nuclides corresponded to the contents of isotopes in an annual unloading of
spent nuclear fuel of VVER-1000 reactor: 241 Am – 17.4 kg, 242m Am – 7.0 g, 243 Am – 3.38 kg, 243 Cm –
5.5 g, 244 Cm – 894 g, 245 Cm – 63.5 g, 246 Cm – 9.1 g. In the calculation of the radiotoxicity, the nuclides
238
Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, 241 Am, 242m Am, 243 Am, 243 Cm, 244 Cm, 245 Cm, 246 Cm, 247 Cm, 248 Cm were taken
into account. It was considered that rather short-lived 242 Cm (T 1/2 = 163 days) after end of an
irradiation, completely decays in 238 Pu.
Table 2 indicates the number of nuclei of the most important nuclides 238 Pu, 241 Am, 242m Am, 243 Am,
242
Cm, 243 Cm, 244 Cm, fission products FP, for irradiation in a thermal spectrum of light water reactor
with Φ = 10 14 n/cm 2 s. The number of nuclei corresponds to above-listed above initial masses of
nuclides.
Table 2. Nuclei number in americium and curium irradiation in thermal spectrum, 10 24 nuclei
Nuclide
T, year
0 0.5 2 5 10
238
Pu 0 8.4 5.8 0.40 8.6-3
241
Am 43.5 8.1 0.82 0.12 2.4-3
242m
Am 0.017 0.13 0.012 1.7-3 3.6-5
243
Am 8.36 2.7 0.093 4.4-3 9.3-5
242
Cm 0 20 4.0 0.32 6.8-3
243
Cm 0.014 0.68 0.22 0.016 3.3-4
244
Cm 2.21 1.5 0.61 0.064 1.4-3
FP 0 6.2 36 52 54
Table 3 summarises the total radiotoxicity in water for three flux values Φ = 10 13 , 10 14 and 5⋅10 14 n/cm 2 s
and the contributions of most significant nuclides 238 Pu and 244 Cm. As for 238 Pu, the contribution of 238 Pu formed
at irradiation and 238 Pu by decay of 242 Cm after irradiation are presented separately. The time of an irradiation at
different Φ corresponds to identical neutron fluence.
907

Table 3. Radiotoxicity in americium and curium irradiation in thermal spectrum, 10 14 kg water
10 13 n/cm 2 s
T, years
0 5 20 50 100
Total 55 106 18 1.1 1.4-2
238
Pu 77 14 0.89 1.2-2
238
Pu from 242 Cm decay 9 0.8 0.07 9-4
244
Cm 12 1.9 0.05 2-4
10 14 n/cm 2 s
T, years
0 0.5 2 5 10
Total 55 147 50 3.9 0.084
238
Pu 35 24 1.7 0.036
238
Pu from 242 Cm decay 85 17 1.3 0.029
244
Cm 15 6 0.6 0.014
T, years
5⋅10 14 n/cm 2 s
0 0.1 0.4 1 2
Total 55 160 121 34 3.7
238
Pu 9 16 4.7 0.5
238
Pu from 242 Cm decay 122 87 23 2.6
244
Cm 15 10 3.5 0.4
In Table 4, the number of nuclei and total radiotoxicity in water for transmutation in fast
spectrum are given.
Table 4. Number of nuclei, 10 24 nuclei, and total radiotoxicity
for transmutation in fast spectrum
Nuclide
T, years
0 0.5 2 5 10
238
Pu 0 2.9 11 5.9 0.8
241
Am 43.5 30 10 1.5 0.25
242m
Am 0.02 0.84 0.72 0.12 0.01
243
Am 8.36 6.7 3.5 0.98 0.11
242
Cm 0 6.6 5.1 0.93 0.11
243
Cm 0.01 0.19 0.54 0.17 0.02
244
Cm 2.21 1.8 1.0 0.35 0.05
FP 0 3.4 17 39 51
RT, 10 14 kg water 55 84 91 35 4.7
The transmutation of americium and curium is characterised by the following major factors. In
the initial actinide mix, an overwhelming share is 241 Am. 242 Cm is produced by irradiation of 241 Am
through intermediate short-lived 242g Am. It decays into 238 Pu. The 238 Pu concentration reaches a
maximum after some years of irradiation, and then decreases. The 242 Cm concentration quickly
reaches a maximum, then decreases being in balance with the decreasing concentration of 241 Am.
Decay of accumulated 242 Cm after irradiation results in additional accumulation of 238 Pu. Further, the
concentration of 244 Cm decreases monotonously, its additional formation from 243 Am slows this
908

decrease. Total actinide amount much decreases and the amount of fission products FP accordingly
grows after 5-10 years of an irradiation.
The radiotoxicity, as it is visible from Table 3, is determined by the nuclides 244 Cm and 238 Pu and,
only at initial stage, also by nuclide 241 Am. The radiotoxicity of 244 Cm per gram is 14 times higher than
that of 241 Am and 2.4 times higher than that of 238 Pu. The initial radiotoxicity is determined in equal about
shares by the nuclides 241 Am and 244 Cm. In the first years of an irradiation, the radiotoxicity, grows
because of accumulation of 238 Pu. Thus the share of 244 Cm in total radiotoxicity makes 10-20% both in
thermal and in fast spectrum, and the share of 241 Am is rather small. A maximum radiotoxicity in thermal
spectrum is 1.2-1.6 times greater than in a fast spectrum. It exceeds initial radiotoxicity by 2-3 times.
The maximum long-lived radiotoxicity in fast spectrum exceeds the initial radiotoxicity by 1.7 times.
The rate of decrease of radiotoxicity by an essential way depends on neutron flux density. It is
explained by the double role of 242 Cm in formation of 238 Pu – at the expense of decay occurring during
irradiation and decay after end of an irradiation. It is expedient to compare different neutron fluxes at
times corresponding to an identical fluence.
At small neutron fluxes and accordingly long time of irradiation, the equilibrium concentration of
242
Cm is small, initial 241 Am at first stage of an irradiation is transformed into 238 Pu through 242 Cm. It is
well illustrated by the first part of the Table 3 corresponding to Φ = 10 14
n/cm 2 s. The share of a
radiotoxicity given by 238 Pu formed from 242 Cm after irradiation is small. Total radiotoxicity at the end of
irradiation decreases much. At high flux and accordingly smaller times (third part of the Table 3,
appropriate Φ = 5⋅10 14 n/cm 2 s), the equilibrium concentration of 242 Cm essentially grows, and its burn-up
at high flux is even small in comparison with its decay. The share of 238 Pu formed during irradiation
decreases. The significant part of radiotoxicity is determined by 238 Pu formed by 242 Cm decay after
irradiation. This share does not transform by irradiation. Therefore at high flux, radiotoxicity as function
of a fluence decreases essentially slower than at low flux. However the maximum radiotoxicity depends
rather poorly on flux density.
4. Stationary mode of a minor actinides transmutation
The stationary mode in which there is a continuous feed of new actinides and continuous
removal of fission products is convenient model for comparison of transmutation modes at various
neutron flux and spectra. The stationary concentration produced in this mode is determined by
neutron flux density, effective cross-sections, feed intensity and isotopic composition of feeding
nuclides.
The characteristics of stationary transmutation modes in thermal neutron power reactor with
neutron flux density 10 14 n/cm 2 s and fast breeder reactor are presented in Table 5. It was considered
that burn-out and feed of minor actinides occurs separately from nuclear fuel in the nuclear reactor
and the nuclides from nuclear fuel do not influence the isotopic composition in stationary mode.
Isotopic composition of feed was the same as for single transmutation of americium and curium with
addition of 237 Np. The feed intensity was defined so that the annual feed is equal to the annual
unloading of actinides from one VVER-1000 reactor. The data on americium and curium are listed
above. The annual unloading of 237 Np is 17.3 kg. In Table 5, M represents total mass of actinides
taking part in stationary transitions, Q is their activity, RT is their radiotoxicity.
909

Table 5. Characteristics of stationary transmutation modes
Characteristics Thermal neutron power reactor Fast breeder reactor
M, kg 170 150
Q( 242 Cm-), 10 6 Curi 26 21
Q(long-term), 10 6 Curi 3.2 2.0
RT( 242 Cm), 10 16 kg water 8.0 6.5
RT(long-term), 10 16 kg water 11.7 9.4
Part in long-term radiotoxicity
238
Pu 56% 82%
244
Cm 41% 13%
The data presented show that, in general, transmutation in regular thermal neutron reactor is
close to the transmutation on base of a fast breeder reactor. Long-lived radiotoxicity is defined
basically by two nuclides: 238 Pu and 244 Cm.
5. Conclusion
Submitted data permit to make conclusions on the main regularities of conversion of nuclides at
burn-up of plutonium and transmutation of minor actinides. At long-term burn-up of plutonium, there is
the increase of radiotoxicity determined by accumulation of 244 Cm. The maximum amount of 244 Cm
makes about 3% of initial plutonium, the maximum radiotoxicity is 18 times the radiotoxicity of initial
plutonium. The maximum is reached after 8 years irradiation in flux 10 14 n/cm 2 s. In real, burn-up of a
plutonium should be done by much shorter lifetimes with intermediate processing and addition of new
plutonium. So, it is expedient at processing to extract curium isotopes. Since 244 Cm half-life makes
18.1 years, one can organise a controllable storage of extracted curium.
For transmutation of americium and curium, there is the increase of radiotoxicity on initial stage
of irradiation. The maximum is reached in 0.5-2 years of irradiation in thermal or fast reactors. The
radiotoxicity is reduced up to initial level in 2-3 years of irradiation. It forms basically by two
nuclides 238 Pu and 244 Cm both in thermal and in fast spectrum. The rate of radiotoxicity decrease
depends significantly on neutron flux density. At high flux, radiotoxicity as function of a fluence falls
down slower at the expense of additional formation of 238 Pu from 242 Cm after irradiation.
At stationary transmutation of minor actinides with continuous feed, the characteristics of the
transmutation process are identical as for facilities with thermal and fast neutron spectrum. The
stationary radiotoxicity in thermal spectrum is determined in identical measure by nuclides 238 Pu and
244
Cm. In fast spectrum the share of 238 Pu is essentially higher.
REFERENCES
[1] Radiation Safety Standards (NRB-99), Minzdrav of Russia, Moscow, 1999.
[2] V.M. Murogov et al., Stimuli of the Development of Fast Reactors with Sodium Coolant.
Atomic Energy, 1993, Vol. 74, Issue 4, pp. 285-290.
910

DEMONSTRATION ACCELERATOR DRIVEN COMPLEX FOR
EFFECTIVE INCINERATION OF 99 Tc AND 129 I
A.S. Gerasimov, G.V. Kiselev, L.A. Myrtsymova
State Scientific Centre of the Russian Federation
Institute of Theoretical and Experimental Physics (RF SSC ITEP)
25, B. Cheremushkinskaya, 117259 Moscow, Russian Federation
Abstract
Design of an ADS complex for transmutation of 99 Tc and 129 I is discussed. The complex contains
following units: a linear accelerator or proton cyclotron with small current, a neutron producing target
and a sub-critical blanket. A schema of a linac with superconducting systems is proposed. Versions of
Pb-Bi and tantalum targets are considered. The special design of a channel-vessel type of blanket is
outlined.
911

1. Introduction
The atomic power engineering exists for more than 40 years. One of the main problems of
atomic power engineering is radiation safety of power plants and long-lived radwaste management.
Nuclear transmutation is one of the alternative technologies of radwaste management. The nuclear
transmutation consists in the incineration of radioactive nuclides by neutron irradiation. For
production of neutrons, a sub-critical reactor (which in this case is called blanket) with neutron source
can be used. A neutron source is a neutron producing target and proton accelerator. A neutron source
permits to have excess neutrons, which can be used for transmutation. This electronuclear system is
called accelerator driven system (ADS). The conceptual researches conducted recently by Russian
and foreign nuclear centres and institutes show a prospectivity and technical feasibility of an
electronuclear method of neutron production for radwaste transmutation.
In the State Scientific Centre of Russian Federation Institute of Theoretical and Experimental
Physics (SSC RF ITEP) electronuclear technology of radwaste incineration is studied for a long time.
It is directed to decrease radiation risk in connection with the opportunity to reduce the amount of
radwaste at the expense of ADS. The development of ADS represents an example of high
technologies used for deciding on one of the complex problems of modern engineering.
2. Choice of nuclides for transmutation
What are reasons to choose nuclides for transmutation? The main danger in radioactive wastes is
determined by two types of long-lived nuclides: fission products and actinides accumulated in spent
nuclear fuel. Their half-life is from hundreds up to millions years. Among basic fission products, the
main contribution in radiation danger during the first 300-1 000 years is given by 90 Sr and 137 Cs with
half-life 30 years. A further storage, it is given by 99 Tc (2.1⋅10 5 years) and 129 I (1.6⋅10 7 years). 129 I is
dangerous because it can be accumulated in the human body.
For 90 Sr and 137 Cs destruction, a too high neutron flux is necessary. These nuclides can not be
transmuted and they can be put in surface storage facility. Among the other fission products, 99 Tc and
129
I should be transmuted. Processes of 99 Tc and 129 I transmutation have specific physical features. At
99
Tc transmutation, the chain of conversions at step-by-step neutron capture by secondary nuclides is:
99
Tc(20/340) 100 Ru(5/11) 101 Ru(3/100) 102 Ru(1/4) 103 Rh(145/1100) 104 Pd(0.6/16) 105 Pd(20/98) 106 Pd(0.3/6) 107 Pd(2/90)
Values of thermal cross-sections and resonance integrals of reactions (n,γ) are given in brackets.
Short-lived intermediate nuclides are not shown. Initially, only 99 Tc is presented in a target. During
irradiation, a 8-step neutron capture, a nuclide 107 Pd-with half-life 6.5⋅10 6
years is produced. This
nuclide is radiotoxic, however its amount is low. The other intermediate nuclides are stable. At 129 I
transmutation, the analogous chain is:
129
I(27/36) 130 Xe(26/14) 131 Xe(85/900) 132 Xe(0.5/46) 133 Cs(29/440) 134 Cs(140/54) 135 Cs(9/62) Ba
Two nuclides are presented in initial target: 129 I and 127 I. An amount of 127 I in spent fuel of PWR is
about 22% of 129 I. A main product of a transmutation is 132 Xe. At 7-step neutron capture, long-lived
135
Cs with half-life 2.3⋅10 6 years is produced. Its amount is important.
912

Thus, stable nuclides will be formed as result of transmutation with rather low content of new
radioactive nuclides. Thermal neutron flux of about 5⋅10 14 cm -2 s -1 is desirable for transmutation and
that is technically feasible at modern level of nuclear engineering.
3. ADS structural scheme
ADS differs from common-type reactor by the sub-criticality of a blanket. This fact excludes an
opportunity of blanket runaway on prompt neutrons and increases a level of nuclear safety.
The different basic schemes of ADS are possible: with horizontal and vertical disposition of a
target and blanket; with top and bottom supply of target by proton beam; with one accelerator and
several blanket and on the contrary, one blanket and several accelerators. The target can be solid (for
example, tungsten, lead, uranium) or liquid (for example, lead, lead-bismuth, melted salts).
4. Design of a blanket
The blanket is the most complex unit in an ADS. Its design can vary by the type of fuel, coolant,
construction scheme, etc. For transmutation of 99 Tc and 129 I, facility with high neutron flux of the order
of 5⋅10 14 cm -2 s -1
is necessary. It is developed in [1]. Parameters of such facility are submitted in
Table 1. The blanket is of channel-vessel type with heavy water as coolant and moderator. The choice
of heavy water is explained by an opportunity to have great volume for irradiation of target materials
and by a wide experience of heavy-water reactor development and operation.
Table 1. Parameters of heavy-water blanket of ADS
Thermal power, MW 1 100
Electrical power, MW 350
Proton beam current, mA 15
Energy of protons, MeV 1 000
Multiplication factor 0.97
Neutron flux density, cm -2 s -1 up to 5⋅10 14
Consumption of 235 U or 239 Pu, kg/year 305
Temperature of a coolant on input of fuel assembly, 0 C
on output of fuel assembly, 0 C
270
300
Rate of 99 Tc transmutation, kg/year 120
Rate of 129 I transmutation, kg/year 35
Number of consumed reactors VVER-1000 4
Such a design can be realised, as the technical parameters are not intense. Neutronic
characteristics can be improved in case of use of a sectioned blanket with neutron valve which are
now studied in various Russian nuclear centres.
5. Neutron-producing target
An important element of a design is neutron-producing target. Targets with liquid lead-bismuth
alloy [2] or solid tantalum target [3] can be used for offered ADS. Total feature of these targets is
their location in central part of a blanket and vertical configuration. Main parameters of two variants
913

of lead-bismuth target are given in Table 2. The offered target is designed for high proton current and
power.
Table 2. Main parameters of lead-bismuth target
Variant 1 Variant 2
Beam current of 1-GeV protons, mA 111 166
Proton to neutron ratio 29 29
Thermal power of a target, MW 77 116
Flow for cooling of a target, m 3 per hour 200 200
Maximal temperature of shell, ° C 600 680
Tantalum target with power of 25 MW is more preferable for ADS designed for 99 Tc and 129 I. It is
made as a set of 15 flat horizontal cylindrical disks with thickness of 30 mm each, with internal
channels for a coolant. Thermal parameters of this target are presented in Table 3.
Table 3. Thermal parameters of tantalum target
Target power, MW 25
Flow of cooling water, kg/s 212
Maximum temperature of a surface, ° C 193
Pressure of water, MPa 1.63
Temperature of cooling water, ° C 30
Maximum temperature increase in a channel, ° C 48
Maximum heat density, kW/m 3 9.5⋅10 5
These thermal parameters of a target are rather moderate, that permits to hope for its reliable
operation. It is necessary to note that use of a solid target permits to refuse special membranes which
separate the vacuum volume of accelerator and target.
6. Conceptual project of the linear accelerator
The linear accelerator of protons for the ADS is chosen in connection with the necessity of high
current of protons with intermediate energy of 0.8-1.5 GeV. For 99 Tc and 129 I transmutation, currents
of the order of 10 mA is required.
A modern variant of linac for the ADS with current of 30 mA and proton energy 1 GeV [4] is
constructed under single-channel scheme. Mode of operations is continuous, on the contrary to
existing proton accelerators which operate in pulsing mode. The accelerator consists of injector,
initial part, intermediate part, and main parts. In table 4, the main calculated characteristics of a linac
with superconducting accelerating resonators in main part giving a rate of acceleration 15 MeV/m are
presented. Use of cryogenic accelerating system and high rate of acceleration results in significant
decrease of linac cost in comparison with cost of the “warm” accelerator.
914

Table 4. Characteristics of linac with cryogenic resonators
Length of linac 135
Efficiency of resonators 1
Efficiency of HF-generators with supply 0.65
HF-power, MW 33
Electric power supply, MW 52
Efficiency of linac 0.55
7. Technical problems to take into consideration
• Blanket: it is necessary to determine experimentally the greatest possible value of effective
multiplication factor which could completely exclude probability to increase it above unit in
any mode of operation. This will allow increasing considerably a level of nuclear safety in
comparison with power reactors. It is necessary to choose a type of fuel and target. One of
possible decisions consists in solution of 99 Tc in coolant, that claims the substantiation of
water-chemical mode.
• Neutron producing target: important question is maintenance of uniformity of proton beam
interaction with target surface.
• Accelerator: it is necessary to check experimentally cryogenic resonators which are a basis of
accelerating structures. Resource tests of klystrons for high-frequency system should be made.
There are a number of other technical questions.
REFERENCES
[1] B.R.Bergelson et al., On Necessity of the Development of Heavy-water ADS Blanket with High
Neutron Flux Density, Proceedings of the Conference on Improved Heavy-Water Reactors
(Moscow, November 18-20, 1997), Moscow, ITEP, 1998, pp. 139-151.
[2] B.R. Bergelson et al., Sub-critical Facility (Target-blanket) for Transmutation of Actinides.
Atomic Energy, 1996, Vol. 82, Issue 5, pp. 341-347.
[3] A.G. Chukhlov, Solid Target for Electronuclear Facility of High Power, Proceedings of the
International Conference on Sub-critical Accelerator Driven Systems (Moscow, October 11-15,
1999), Moscow, ITEP, 1999, pp. 151-154.
[4] B.Yu. Sharkov et al., The Problem of Design of High Current Proton Linear Accelerator for
Electronuclear Systems, Proceedings of the International Conference on Sub-critical
Accelerator Driven Systems (Moscow, October 11-15, 1999), Moscow, ITEP, 1999, pp. 12-16.
915

CRITICAL AND SUB-CRITICAL GT-MHRs FOR
WASTE DISPOSAL AND PROLIFERATION-RESISTANT FUEL CYCLES
D. Ridikas, G. Fioni, P. Goberis, O. D’eruelle, M. Fadil, F. Marie
CEA Saclay
91191 Gif-sur-Yvette Cedex, France
Abstract
The gas turbine modular helium-cooled reactor (GT-MHR), using an annular graphite core and
graphite inner and outer reflectors, is known probably as the best option for the maximum plutonium
destruction in once-through cycle. Combination of the critical GT-MHR with an accelerator driven
(AD) sub-critical GT-MHR core (AD-GT-MHR) would allow to merge the best features of both
systems to achieve near total destruction of fissionable plutonium. We perform detailed simulations
along these lines and compare different scenarios in order not only to reduce considerably the mass of
plutonium isotopes but also to diminish the waste radiotoxicity in the long term.
917

1. Introduction
Nuclear waste radiotoxicity in the long term, say, more than 1 000 years, is strongly dominated by
plutonium isotopes if the spent fuel originating from the once-through cycle is considered as waste.
The disposal of this waste has become an environmental and political issue. The main concerns are
related to the potential for radiation release and exposure from the waste, and also the possible
diversion of fissionable material.
Plutonium is the unique waste component. It is fissionable, and therefore capable of releasing
significant amount of energy. It is also a hazardous material, particularly if inhaled in particulate form.
Because of its potential use in nuclear weapons, there is great sensitivity about isolating plutonium
from other components of the nuclear waste stream.
As it was shown in a number of studies, gas-cooled reactor technologies offer significant
advantages in accomplishing the transmutation of plutonium isotopes and nearly total destruction of
239
Pu in particular (see [1-7] and also references therein). GT-MHR uses well-thermalised neutron
spectrum, operates at high temperature without the need for fertile material and employs
ceramic-coated fuel. It utilises natural erbium as a burnable poison with the capture cross-section
having a resonance at a neutron energy such that ensures a strong negative temperature coefficient of
reactivity. The lack of interaction of neutrons with coolant (helium gas) means that temperature
feedback is the only significant contributor to the power coefficient. As a matter of fact, no additional
plutonium is produced during the fuel cycle since no 238 U is used.
This paper describes the application of critical as well as sub-critical (accelerator driven)
GT-MHR for transmutation of plutonium originating from the spent nuclear fuel in the once-through
cycle. We examine a few different scenarios in order to obtain the maximum destruction of plutonium
as well as to minimise the irradiated fuel radiotoxicity in the long term. MCNPX [8], MCNP4B [9],
MONTEBURNS [10] and CINDER’90 [11] codes are employed at different stages of our simulations.
The performances of the codes have been reasonably benchmarked in part in [12] by simulating the
fuel cycle of the high flux reactor at ILL Grenoble.
This work is organised as follows. Major GT-MHR characteristics and system modelling details
are summarised in the following two sections. After we present different scenarios considered, which
are followed by simulation results and discussion. The paper ends with conclusions and outlook.
2. Major GT-MHR characteristics
The Gas Turbine – Modular Helium Reactor (GT-MHR) [1] is an electric generation power plant
that couples the passively safe critical reactor with a highly efficient energy conversion system.
Conceptual design of GT-MHR was developed in a joint project of the Russian Federation, USA,
France and Japan with the major interest in plutonium based fuel cycles in general and weapons grade
plutonium (WGPu) destruction in particular. We refer the reader to Reference [1] for further details on
technical characteristics of GT-MHR. Table 1 gives only major reactor parameters essential for the
modelling of the system.
A simplified model of GT-MHR reactor which is shown in Figure 1 has been created using
MCNP4B geometry set-up [9]. It consists of the reactor core (C), inner reflector (Ri), side reflector
(Rs), top reflector (Rt), bottom reflector (Rb), and reactor vessel (V) as shown explicitly in Figure 1.
Further details on modelling can be found in [13], while some of the results in part have already been
reported in [14].
918

Table 1. Basic GT-MHR reactor parameters
Power, MW(th) 600
Active core size:
- height, cm
800
- outer diameter, cm
484
- inner diameter, cm
296
Active core volume, m 3 92.1
Average power density, W/cm 3 6.5
Outer diameter of the side reflector, cm 684
Total height of the core, cm
1 090
(with top and bottom reflectors)
Vessel size:
- height, cm
- outer diameter, cm
- thickness, cm
Averaged temperature, °C:
- active core
- inner reflector
- side reflector
- top and bottom reflector
- vessel
- helium at the core inlet/outlet
2 022
750
20
1 230
1 500
1 100
1 000
440
487.8/850.0
3. Modelling tools and procedure
The major reactor parameters used for modelling of the GT-MHR reactor are given in Table 1. As
we have already mentioned above, a three-dimensional geometry set-up was constructed and is
presented in Figure 1. MCNP4B was also used to obtain k-eigenvalues and neutron fluxes. In the case
when a sub-critical GT-MHR (AD-GT-MHR) is considered, a single difference was the prevision for
an accelerator target (for neutron production by spallation) and surrounded pressure vessel, both
located in the central graphite reflector. We have considered a fluidised bed of tungsten particles
cooled by helium as a spallation target. In this particular case the MCNPX code in proton source mode
was applied to all calculations of neutron fluxes.
Burn-up calculations were made with MONTEBURNS [10]. MCNPX [8] was used to write a low
energy (E n

Figure 1. Lengthwise section view of GT-MHR reactor model.
The following notation is employed: C – core, Rt – top reflector,
Ri – inner reflector, Rb – bottom reflector, Rs – side reflector, and V – reactor vessel.
The change of neutron flux spectra because of fuel burn-up also has been determined. A typical
neutron spectrum evolution for plutonium fuel poisoned with natural erbium is shown in Figure 2. The
observed increase of the thermal flux can be explained by the loss of 239 Pu and 240 Pu in addition to
burnable poison 167 Er during the fuel burn-up. In other words, the mass of the elements with the highest
thermal neutron capture cross-sections is considerably decreased with time.
We note separately at this point that the change of the energy spectra of neutrons will change the
average cross-sections to be used in the burn-up calculations, in some cases by a factor of two or more
[13,14]. Therefore for these types of spectra fuel evolution calculations have to be performed with
corresponding variable neutron fluxes as it is done with MONTEBURNS [10].
Another important result of our calculations is related to the estimation of neutron fluxes in their
absolute value, again as a function of time. For example, we found that typical average GT-MHR
neutron fluxes in the active core may increase by 50-100%, i.e. from ~1⋅10 14 n/(cm 2 s) to ~2⋅10 14 n/(cm 2 s)
at the beginning and at the end of the fuel cycle respectively. This shows that irradiation/burn-up
conditions may change as a function of time what has to be taken into account for the follow up burn-up
calculations.
920

Figure 2. Typical change of the averaged energy spectra of neutrons in the active core of
GT-MHR for different fuel burn-up expressed in full power days (fpd).
4. Scenarios examined
First of all, in Table 2 we fix the transuranic waste composition to be transmuted. This table
represents typical form of LWR discharge (40 GWd/tonne burn-up) when uranium and fission
products are separated. 1 500 kg of the waste is taken arbitrarily.
Table 2. Initial composition of the TRU considered in this study.
Nuclide (%) (kg)
237
Np 4.9 73.5
238
Pu 1.7 25.5
239
Pu 54.5 817.5
240
Pu 22.8 342.0
241
Pu 5.4 81.0
242
Pu 3.7 55.5
241
Am 5.7 85.5
243
Am 0.9 13.5
242
Cm 0.1 1.5
244
Cm 0.3 4.5
Total Pu 88.1 1321.5
Total 100 10500
921

Scenario S0. This is our reference point when TRU represented in Table 2 is left to decay
naturally, i.e. no irradiation takes place.
Scenario S1. In this case only plutonium isotopes are placed in the critical GT-MHR for destruction.
The length of the fuel cycle is ~1 550 days. Table 3 lists in detail initial load and discharge isotopic
composition (see column Discharge S1) correspondingly. Total neutron fluence was ~1.7⋅10 22 n/cm 2 .
Table 3. Isotopic composition of the TRU considered for transmutation: cases S1 and S2.
19.2 kg of 167 Er was put at the beginning of the fuel cycle for Scenario S1 and
additional 13.5 kg of 167 Er was put at the beginning of the fuel cycle for Scenario S2
(at the end of the fuel cycle of Scenario S1).
Nuclide Load (kg) Discharge S1 (kg) Discharge S2 (kg)
234
U 0.0 0.6 0.6
235
U 0.0 0.2 0.2
236
U 0.0 0.1 0.1
237
Np 0.0 0.0 0.0
238
Pu 25.5 26.5 20.7
239
Pu 817.5 24.8 4.4
240
Pu 342.0 47.3 17.6
241
Pu 81.0 87.1 14.0
242
Pu 55.5 141.0 149.0
241
Am 0.0 9.1 2.1
243
Am 0.0 35.0 38.8
242
Cm 0.0 6.1 4.7
243
Cm 0.0 0.2 0.2
244
Cm 0.0 32.7 48.7
245
Cm 0.0 2.1 1.5
246
Cm 0.0 0.5 1.3
Total Pu 1321.5 326.5 205.7
Total 1321.5

Scenario S4. This case is actually a continuation of Scenario 3. GT-MHR continues running in its
sub-critical mode for another 250 days with decreasing k eff
(k eff
∼0.88→0.60) and consequently
decreasing reactor power P th
(P th
= 1.0P 0
→0.30P 0
). The accelerator power should be increased from
∼42 MW to ∼62 MW. A detailed composition of TRU is presented in Table 4 (column Discharge S4).
Total neutron fluence (including Scenario S3) was ∼2.5⋅10 22 n/cm 2 .
Table 4. Isotopic composition of the TRU considered for transmutation:
cases S3 and S4. 29.2 kg of 167 Er was put at the beginning of the fuel cycle.
Nuclide Load (kg) Discharge S3 (kg) Discharge S4 (kg)
234
U 0.0 1.7 1.7
235
U 0.0 0.4 0.4
236
U 0.0 0.1 0.2
237
Np 73.5 29.5 20.7
238
Pu 25.5 90.6 69.1
239
Pu 817.5 27.5 13.0
240
Pu 342.0 42.4 20.9
241
Pu 81.0 72.0 18.1
242
Pu 55.5 151.0 153.0
241
Am 85.5 11.9 2.9
243
Am 13.5 40.4 43.0
242
Cm 1.5 11.4 9.7
243
Cm 0.0 0.6 0.4
244
Cm 4.5 50.6 67.5
245
Cm 0.0 3.8 2.7
246
Cm 0.0 1.3 2.5
Total 10321.5 383.5 274.1
Pu
Total 10500

Scenario 3 shows that mixing of waste plutonium with the rest of actinides (Np, Am, Cm) can
influence considerably GT-MHR performances. First of all, the system cannot start its cycle in the
critical mode, i.e. an accelerator is needed already at the very beginning of the operation. Secondly, as
soon as the sub-critical GT-MHR starts its fuel cycle, k eff
is constantly increasing as a function of time;
the system becomes critical and can run in its critical mode for a certain period of time without
accelerator. At this point there is a saturation of the 241 Pu build up from 240 Pu, i.e. until now 241 Pu mass
was constantly increasing and starts decreasing from now on. Finally, due to considerable burn-up of
239
Pu (compared to its initial value) and also 241 Pu (compared to its saturation value) k eff
falls below
unity and there is again need for an accelerator to continue the fuel cycle.
Again like in the Scenario S2, Scenario S4 shows that in order to obtain deeper burn-up of fissile
materials the system should be run as long as possible even with decreasing core power. The initial
reactor power cannot be kept constant due to large decrease in k eff
and also due to limitations on
available accelerator power. Although, Scenario S2 ends up with slightly better burn-up rates both for
239
Pu and 240 Pu, at the very beginning it requires full separation of Np, Am and Cm isotopes. In the case
of Scenario S4, the mass of 238 Pu is actually increased due to the presence of 237 Np in the fuel, what is
not the case for Scenario S2. However, if the total mass of actinides is compared at the end of the fuel
cycles, i.e. (Discharge S2 + 178.5 kg of Np, Am and Cm isotopes not irradiated) with (Discharge S4),
the numbers are in favour of Scenario S4 by -57.5 kg.
The question which remains to be examined is the actual behaviour of fuel radiotoxicity for all
different cases considered in this study. Figure 3 presents a change in radiotoxicity for Scenarios S0,
S2 and S4 as described in the previous Section. It is important to note that the curve TRU: S2 contains
the contribution to the radiotoxicity due to 178.5 kg of Np, Am and Cm isotopes (see Table 2) not
irradiated in this particular scenario. In brief, there is no considerable gain in decreasing TRU
radiotoxicity in a long term. As it is clearly seen from Figure 3, the total radiotoxicity of the irradiated
fuel is higher by a factor of 3-4 during first 100 years of cooling. Later curves cross and irradiated fuel
becomes less radiotoxic only by a factor of 4-5 after 1 000 years of cooling. After 100 000 years the
radiotoxicity curves of irradiated fuel again approach the one corresponding to non irradiated TRU.
Even more, Scenario S4 for a while gives even higher radiotoxicity compared to Scenario S0 as shown
in the same Figure.
6. Conclusions and outlook
The problem of elimination of Pu isotopes has been addressed in terms of once-through fuel
cycle. We confirm that the GT-MHR technology offers the potential to eliminate essentially all
weapons-useful material present in nuclear waste. In addition, wide spectrum of plutonium isotopic
compositions with or without isotopes of Np, Am and Cm prove GT-MHR potentials to use the
plutonium as fuel without generating large amounts of minor actinides. In brief, we emphasise that
significant levels of plutonium destruction (~99.5% for 239 Pu and ~84.4% for all Pu isotopes) can be
achieved using critical and accelerator-driven subcritical thermal assemblies. The use of accelerator is
essential to provide the needed neutrons for deep burn-up of 239 Pu and an important further destruction
of 240 Pu and 241 Pu in the thermal regime without reprocessing.
924

Figure 3. A change in radiotoxicity of TRU for different scenarios considered:
TRU:S0 – no irradiation, TRU:S2 – Scenario S2 and TRU:S4 –
Scenario S4. See Tables 2, 3 and 4 for detailed discharge compositions.
On the other hand, none of the scenarios considered could reduce significantly the total waste
radiotoxicity in the long term. Although only a small amount of higher actinides (mainly Am and Cm)
is created during the fuel cycle, these elements are much more toxic than initial plutonium, what
actually “compensates” the decreased total mass of actinides. Consequently, the question why other
actinides but plutonium cannot be eliminated via fission in GT-MHR should be addressed. There are at
least a few interlinked answers explaining this situation:
1. Np, Am and Cm are not efficiently fissioned in thermalised neutron flux simply due to very
small σ f
/σ c
ratio.
2. Neutron fluxes available in GT-MHR in absolute value are rather low, say, of the order of
1-2⋅10 14 n/s/cm 2 ⋅
3. Typical fuel cycles of GT-MHR are rather short, i.e. 4-6 years.
Since other actinides with an exception of plutonium are more inclined to fission in a fast neutron
energy spectrum, one could consider an additional fast stage, following a thermal stage, in order to
eliminate the remaining actinides. Consequently, much bigger gain in reduction of the total
radiotoxicity of the waste could be achieved. This is a subject of our future study, and the work along
these lines is already in progress elsewhere [4,6,7].
925

REFERENCES
[1] International GT-MHR project, Conceptual Design Report, OKBM, Russia, report GT-MHR
100025 (1997).
[2] C. Rodriguez, J. Zgliczinski, and D. Pfremmer, GT-MHR Operation and Control, IAEA meeting
on Gas Reactors, IAEA report (1994).
[3] A.J. Neylan, F.A. Silady, and A.M. Baxter, Gas Turbine Modular Reactor (GT-MHR): A
Multipurpose Passively Safe Next Generation Reactor, General Atomics, LA-12625-M (1995).
[4] C. Rodriguez, A. Baxter, Gas-cooled Transmutation of Nuclear Waste Using Thermal and Fast
Neutron Energy Spectra, Presentation at the Massachusetts Institute of Technology, USA (2000).
[5] X. Raepsaet, F. Damian, R. Lenain, and M. Lecomte, Fuel Cycle Performances in High
Temperature Reactor, CEA and Framatome case study (1998).
[6] A.M. Baxter, R.K. Lane, and R. Sherman, Combining a Gas Turbine Modular Helium Reactor
and an Accelerator and near Total Destruction of Weapons Grade Plutonium, AIP report,
General Atomics, USA (1995) 347.
[7] C. Rodriguez et al., The Once-through Helium Cycle for Economical Transmutation in Secure,
Clean and Safe Manner, Paper 73 in the Proceedings of the 6th Information Exchange Meeting
on Actinide and Fission Product Partitioning and Transmutation, 11-13 December 2000,
Madrid, Spain, OECD/NEA (Nuclear Energy Agency), Paris, France, 2001.
[8] L.S. Waters, MCNPX TM USER’s MANUAL, Los Alamos National Laboratory, pre-print
TPO-E83-G-UG-X-00001 (November 1999). See also: http://mcnpx.lanl.gov/.
[9] J. Briesmeister for Group X-6, MCNP-A, A General Monte Carlo Code for Neutron and Photon
Transport, Version 4A, Los Alamos National Laboratory, pre-print LA-12625- M (1993).
[10] H.R. Trellue and D.I. Poston, User’s Manual, Version 2.0 for Monteburns, Version 5B, Los
Alamos National Laboratory, pre-print LA-UR-99-4999 (1999); H.R. Trellue, Private
communication (April 2000).
[11] W.B. Wilson, T.R. England and K.A. Van Riper, Status of CINDER’90 Codes and Data, Los
Alamos National Laboratory, pre-print LA-UR-99-361 (1999), submitted to Proc. of 4th
Workshop on Simulating Accelerator Radiation Environments, September 13-16, 1998,
Knoxville, Tennessee, USA; W.B. Wilson, private communication (May 2000).
[12] D. Ridikas, G. Fioni, P. Goberis, O. D’eruelle, M. Fadil, F. Marie, S. Röttger, On the Fuel Cycle and
Neutron Fluxes of the High Flux Reactor at ILL Grenoble, CEA-report, DAPNIA/SPHN-00-52
(2000); available at http://www-dapnia.cea.fr/Doc/Publications/Sphn/sphn00.html.
926

[13] P. Goberis, Modelling of Innovative Critical Reactors with Thermalized Neutron Fluxes in the
Frame of the Mini-Inca Project, rapport du stage at CEA Saclay, DAPNIA/SPHN-00-68
(2000); available at the request by e-mail: D.Ridikas (ridikas@cea.fr).
[14] G. Fioni, O. D’eruelle, M. Fadil, Ph. Leconte, F. Marie, D. Ridikas, Experimental Studies of the
Transmutaion of Actinides in High Intensity Neutron Fluxes, Proc. of the 10th International
Conference on Emerging Nuclear Energy Systems, ICENES 2000, Petten, the Netherlands,
September 25-28, 2000, p.374.
927

THE USE OF PB-BI EUTECTIC AS THE
COOLANT OF AN ACCELERATOR DRIVEN SYSTEM
Alberto Peña 1 , Fernando Legarda 1 , Harmut Wider 2 , Johan Karlsson 2
1 University of the Basque Country
Nuclear Engineering and Fluid Mechanics Department
Escuela Técnica Superior de Ingenieros, Alameda de Urquijo s/n, Bilbao, 48013, Spain
2 Institute for Systems, Informatics and Safety
Joint research Centre of the European Commission
Ispra, Italy
Abstract
The use of the Pb-Bi eutectic appears necessary for designs of spallation targets for ADSs. Even in
ADS facilities cooled by gas, the target unit for the system contains lead-bismuth. Including this liquid
metal as the primary coolant of the sub-critical reactor has important advantages in the safety field.
Natural circulation, which can be enhanced by inert gas injection, avoids mechanical pumps or
electrical induction pumps. Calculations made with CFD codes show that Pb-Bi coolant circulation by
buoyancy forces is an important safety aspect. Even in the case of a loss of the heat sink, the core is
still coolable with passive devices such as a Reactor Vessel Auxiliary Coolant System. Results from
two different codes demonstrate similar conclusions about this passive emergency cooling system.
929

1. Introduction
The Nuclear Engineering and Fluid Mechanics Department in the Engineering School of Bilbao,
is working on computational capacity of CFD codes, concentrating the efforts on liquid metals
thermohydraulics. There is also a support of calculations of radiological protection, dosimetry, and
shielding.
At the Joint Research Centre of the EC at Ispra (Italy), the ISIS group is doing calculations of an
ADS prototype designed by ANSALDO [1]. A 2-D representation of this design has been set-up using
the Computational Fluid Dynamic (CFD) STAR-CD code. Including the core, a riser cylinder and a
heat exchanger in the upper part of the downcomer, as well as a gas plenum on top of the fluid. The
vessel, the safety vessel around this one, and the reactor vessel air coolant system (RVACS) are also
modelled.
The collaboration of these two groups led to the setting up of the FLUENT code for a similar
representation of the ANSALDO’s facility as the STAR-CD one. And the main objective of the
calculations was to compare both codes, and see if they led to the same conclusions.
Using normal operation parameters, the model was driven to a steady state situation. And then, it
was supposed to come in a situation of a station blackout to see the behaviour of the Pb-Bi eutectic
flowing in natural convection.
This benchmark work, not yet finished, will lead to a better comprehension of the behaviour of
the codes, and to propose possible modifications of some of the models. On the other hand the detailed
study of the comparison results will help to achieve a better design for the demonstration facility.
In this paper some first FLUENT calculations are shown together with the STAR-CD ones, and it
can be said that results are not that different, taken into account that two different codes were used and
two different persons were doing the calculations.
2. Description of the accident
The description of the ANSALDO’s prototype design is fully explained in [1], but apart from the
steady state calculation based on this normal operation design, it is important to demonstrate that the
ADS design is safe. And a simulation of an accident by a CFD code, could be a good source of
demonstrating it.
The accident considered for this report is the following. After a certain working period of normal
operation, there is a station blackout. So, the proton beam is switched off, there is a secondary coolant
loss (it is assumed that once the secondary coolant is lost, it takes ten seconds, after the black out, to
switch the beam off), and the bubble injection, used to help the primary coolant in its circulation, is
also stopped. From that point, the coolant must circulate by natural convection, as the bubble injection
has disappeared, and the reactor must remain coolable during the accident sequence.
The power law used for the decay heat power from the core has been taken from [2], and it is
considered that the prototype has been working for almost two years.
930

P d
T = 6.0E+07, time at reactor full power.
t = cooling time.
P 0 = reactor power.
P d = decay heat power.
This transient will be running for 40 hours.
0
−0.2
−0.2
( t − ( T + ) )
( t,
T ) = 0.0622P
t
3. Fluent simulation inputs
Modelling of the ANSALDO’s design includes the core, a riser cylinder and a heat exchanger
(Number 3, in Figure 1) in the upper part of the downcomer, as well as a gas plenum on top of the
fluid (Number 6). The vessel, the safety vessel around this one (the gap between the two vessels is
depicted with Number 5), and the reactor vessel air coolant system (RVACS) (Number 4) are also
modelled.
The simulation was done in a 2 D axysimmetric geometry, as shown in Figure 1. The total
number of cells is 1 458, and they are all quadrilateral cells. This grid is the finest one of the two used
in the paper, and the more similar to the STAR-CD one.
Figure 1. Grid
931

Some simplifications have been made in the geometry to get the best 2 D simulation. This can be
seen in Figure 1.
For an axisymmetric geometry as this one, a cylindrical heat exchanger has been used, although
this is not the real foreseen design. However, the dimensions have been adjusted to the ones of the
ANSALDO’s heat exchangers.
The core (Number 1) has 120 fuel assemblies, but to simplify the model, the tubes have been
substituted by the pressure drop they produce. And the same has been done with the heat exchanger
and the dummy assemblies (Number 2). These pressure drops are the ones indicated by [1], that is:
20 kPa inside the core, and 7 kPa for the heat exchangers. This last pressure drop has been calculated
from the difference between the core and the total pressure drop, which is 29 kPa. 2kPa were assumed
for the rest of the system.
Most of the fluid should pass through the core for cooling the fuel assemblies. So, a very small
flow area has been assumed for the dummy assemblies, in order to get the Pb-Bi passing mainly
through the core. The flow rates are: 5 345 kg/s through the core, and 491 kg/s through the dummy
assemblies.
The proton beam is not represented, so, the power inserted in the core is a constant energy source.
Then, the secondary circuit (the heat exchanger), is represented as a power sink. All the power
generated in the core is taken out by the heat exchanger, so it is represented as a negative energy
source.
The boundary conditions and the general data have been taken from [1].
There are several physical models involved in this simulation, so it is worth to mention them, and
to give a brief explanation of the assumptions made. For more details see [4]:
• Basic models: continuity, momentum, energy.
• k-eps RNG model for the turbulence equations. This model was chosen because it is more
accurate than the standard k-eps model, and it is not as elaborated as other models that solve
the Navier-Stokes equations, taking into account the Reynolds stresses. A term that accounts
for low-Reynolds numbers is included.
• The model used for radiation calculations is based on the expansion of the radiation intensity,
I, into an orthogonal series of spherical harmonics, treating all walls as gray and diffuse.
• Density is treated as a constant, except in the buoyancy term in the momentum equation, that
a Boussinesq approximation is used.
• Discrete phase model. This model has been included in order to simulate the injection of
Argon to pump the primary coolant. The diameter of the bubbles is 1 mm, and they are
injected 0.5 meters above the core outlet.
Another important point of the simulation is the material properties input. The Pb-Bi eutectic
properties have been taken from [2] and [3].
Radiation is important in this simulation due to the high temperatures, and for the walls an
emission of 0.7 has been input.
932

The partial differential equations describing the flow are transformed into discretized analytical
equations. And for these calculations the first order UPWIND scheme is used for momentum and
energy equations, and the second order UPWIND scheme for the turbulence equations.
4. Calculational results
4.1 The steady state
The whole problem, with all the above-mentioned models, is driven into a steady state. And to get
to this situation, the injection of Argon and the use of the porous media model, to get the pressure
drops in the display, are the main key. Figure 2 shows the evolution of temperature inside the vessels,
as the flow is heated when going through the core, and getting colder in the heat exchanger. The
maximum temperature is reached in the core outlet, 676 K. And once the coolant has gone through the
heat-exchanger, the temperature goes down to 570 K, approximately.
Figure 2. Contours of temperature
There is some undesirable re-circulation above the dummy and below and above the heat
exchanger (see Figure 3). Some of this re-circulation could be avoid with buffers above the core. And
it is an option seriously taken for the final design.
933

Figure 3. Velocity vectors
Some of the lift force needed from the Argon is missed in the re-circulation, as some lines of
bubbles are re-circulating in the main stream.
With this injection of bubbles, the correct velocities were reached, as well as the correct pressure
drop and the proper temperature difference between the core inlet and the core outlet. Previous
calculations without bubbles, could not reach the predicted velocities. And due to these lower
velocities, temperatures were higher, up to 700 K (25 K increase).
In the next pictures (Figures 4 and 5), pressure drop, temperature difference and velocities are
depicted. The average temperatures indicate that there is a difference of approximately 100 K between
the core inlet and the core outlet. In the calculations, temperatures vary from 580 K at the core inlet, to
676 K at the core outlet, while in the ANSALDO’s design 573.15 and 673.15 are foreseen.
Figure 4. Temperature differences in the core
934

Figure 5. Core and heat exchanger pressure drops
The pressure drops were adjusted through the porous media model, as a momentum source
inserted in the momentum equation.
In this Figure 5 it is shown that pressure drops of approximately 20 kPa in the core and 7 kPa in
the heat exchangers are predicted by the calculations. And those are the pressure drops foreseen by
ANSALDO. Once these pressure drops were the predicted ones, and the temperature was the correct
one, the accident could be calculated.
4.2 The accident
The calculations have been done with the Argon injection switched off and a loss of the heat sink
(the heat exchangers do not extract heat anymore), due to a station blackout accident.
On the other hand, to make a conservative approach, it has been considered that there are ten
seconds after the blackout, in which the proton beam is still working. So, the full power is on and the
temperatures increase during this period by approximately 20 K (see Figure 6).
Once the beam power is off, the lead-bismuth coolant must be capable of managing the core
cooling. At this point, the core is heated due to the decay heat power, and therefore, the energy source
for the core heating is now determined by the function mentioned in the description of the simulation
inputs.
The accident is assumed to continue for forty hours, and several conclusions can be drawn from
these first results. The profile of the core outlet temperature is used to study the most important
consequences of the accident considered.
The first calculation with FLUENT, with a very coarse mesh, shows that from the beginning of
the blackout till the complete shut down of the proton beam (ten seconds), the temperature increases
by 20 K, to 693 K. And once the power of the beam is off, and during the next 200 s, the temperature
decreases 100 K, down to 590 K. Then, there is another escalation of temperatures at five hour after
the accident.
935

Figure 6. Temperature evolution at the core outlet
800
750
Temperature (K)
700
650
600
550
0 50 000 100 000 150 000
time (s)
FLUENT coarser mesh FLUENT finer mesh STARCD
The behaviour of the temperatures with a finer mesh, and the ones calculated by STAR-CD are
qualitatively similar, but the actual values are somewhat different. FLUENT calculates a first peak of
772 K (100 K above the steady state temperature of 673 K), and STAR-CD predicts a temperature of
675 K from 659 K, in the steady state).
During this slope, a cloud of high temperature is formed above the heat exchanger that remains
there for 7 000 seconds. The peak of 620 K is reached in the coarser mesh calculation by FLUENT;
the STAR-CD calculation reaches 613 K, and the finer mesh calculation with FLUENT, 625 K. And
from this time on the coolant begins to cool down, just by the natural convection of the Pb/Bi and the
air. This means that the Pb-Bi boiling temperature is never reached (1 670 K). And what is more
important, temperatures are always far below 1 273 K, a temperature above which vessel creep
becomes a problem. At the end of these forty hours, the temperature has decreased to 573 K at the core
outlet.
Another important result is the velocity decrease. From 0.42 m/s, the coolant circulation reduces
to 8 mm/s in the period in which the mentioned temperature cloud is active (7 000 s). Then, it
increases to 2.5 cm/s when the normal re-circulation is re-established.
The RVACS also assures the vessel cooling. This cooling is possible and useful because of the
good conduction characteristics of lead-bismuth.
Recent FLUENT calculations have been made using an unsteady formulation to reach the steady
state, instead of a direct steady state formulation. The core outlet temperature is 659 K, similar to the
STAR-CD one. The maximum temperature peak in the STAR-CD calculations is twelve hours after
the accident, while the FLUENT calculations that appear in this paper predict the peak five hours after
the accident, and the new calculations, after eight hours of transient.
5. Conclusions
Results in this paper have been taken from the first running of a basic simulation of the
ANSALDO’s ADS design. No grid independence studies have been done, and the detailed comparison
936

with the STAR-CD results is still to come. But some important conclusions can be extracted from this
preliminary work:
• The good safety characteristics of lead-bismuth as a coolant for an ADS core. Its great heat
capacity plus its high thermal conductivity are a good warranty for having time for an
operator response, if necessary. A comparison between helium, water and lead-bismuth,
shows the following properties at 700 K:
– Pb-Bi: Thermal conductivity (k): 12.616 W/m.K; Specific heat (cp): 146.56 J/kg.K.
– Helium: k: 0.152 W/m.K; cp: 5 193 J/kg.K.
– Water: k: 0.6 W/m.K; cp: 4 182 J/kg.K.
The passive way of functioning is an important safety item since fewer things can go wrong
and human errors cannot play a role.
• One of the main problems to get the correct temperatures, has been the calculation of the
porous media model parameters. The problem is that FLUENT considers all the area as
totally open, so it is not possible to get the correct velocities in these geometries, because it
considers velocities as if the cross section area is much larger than in reality. But once the
correct momentum sources are calculated through the porous media model the core interior
velocities should not lead to problems if the main flow is the correct one.
• The addition of bubbles has been the key to get the predicted velocities and temperatures.
Since one is enhancing circulation, velocities are higher and temperatures are lower because
of a better cooling. But it has also been interesting to do calculations without these bubbles,
because it has been proven that natural circulation alone guarantees reactor cooling even
without gas bubbles enhancing the flow. In this last case maximum temperatures get to
700 K, still far from the eutectic boiling temperature of 1 670°C.
• The vessel cooling by air is another important item of this ADS design. In this calculation the
RVACS has been modelled as a cavity of air surrounding the safety vessel. But the original
idea is to use several tubes, up to 84 U-tubes, with atmospheric air circulating through them.
For this more complicated geometry a 3D simulation will be needed.
The thermal condition of the external wall is isothermal, with a temperature of 293.15 K. This
could lead to non-physical results, because it is not very realistic to have the same
temperature during normal operation, and during the accident.
• Before some serious benchmark studies are done between FLUENT and STAR-CD
calculations, some comparisons have been done, and results are similar except for one result.
Temperatures are similar during the transient. For example, the maximum differences are
only between 7 K to 10 K. This is not very much if we take into account that the simulations
are not completely the same. In the steady state temperatures there are also differences (15 K
approximately). But there is a very odd thing in the time when the maximum temperature is
reached. While FLUENT predicts this maximum temperature at about five hours after the
beginning of the accident, STAR-CD calculations predicts this time to be twelve hours. But
as it has been mentioned, new FLUENT calculations using the unsteady formulation of the
flux governing equations, seem to get more similar results compared with the STAR-CD
ones. Tendencies, anyway, are similar, although FLUENT profiles show a quicker increase of
temperature in the first seconds.
937

• Two main reasons can explain this large difference:
– The injection of bubbles: while FLUENT models this injection of ARGON as a discrete
phase interacting with the main flow, STAR-CD varies the density of the coolant above
the core with a user defined function.
– The convective heat transfer correlations: FLUENT is using default ones, and STAR-CD
inputs some new ones taken from reference [2]:
On the air side: Nu = 1.22 Re
On the Pb-Bi side: Nu = 0.5Re
• Calculations have been carried out with a first order discretization scheme, which gives more
diffusive results. The stability of the solution is rather good, but not so the accuracy of the
results.
Calculations with a second order scheme must be done to assure better results.
• The grid used for the preliminary results is still very coarse, and a finer one is a necessity for
grid independence results.
0.456
0.5
Pr
Pr
0.4
0.5
REFERENCES
[1] ANSALDO. Energy Amplifier Demonstration Facility Reference Configuration, Summary
Report. EA B0.00 1 200 – Rev.0, January 1999.
[2] Frank P. Incropera, David P. DeWitt, Fundamentals of Heat Mass Transfer, John Wiley &
Sons, 4th edition.
[3] The Reactor Handbook. Volume 2 Engineering, Technical Information Service, US Atomic
Energy Commission, May 1955.
[4] FLUENT Manuals.
938

ONE WAY TO CREATE PROLIFERATION-PROTECTION OF MOX FUEL
V.B. Glebov, V.A. Apse, A.E. Sintsov, A.N. Shmelev
Moscow State Engineering Physics Institute (MEPhI)
Moscow Zip: 115409, Russian Federation
E-mail: vglebov@nr.mephi.ru
M.V. Khorochev
Scientific and Engineering Center for Nuclear and Radiation Safety (SEC NRC)
Moscow, Russian Federation
E-mail: khom@grs.de
Abstract
The potential of the accelerator driven systems (ADS) is evaluated for enhancing the proliferation
resistance of LWR-MOX fuel. It is considered as a concept to create an inherent radiation barrier in
MOX fuel by the admixture of 232 U and followed by a short-term irradiation in a blanket of the ADS
with a solid-state target. ADS with extremely moderate physical parameters of pool-type sub-critical
light-water cooled blanket can be applied to create the inherent radiation barrier in fresh MOX fuel
assemblies at the level of Spent Fuel Standard. It would promote development of proliferationresistant
nuclear fuel cycles.
939

1. Introduction
Actually, we noticed the extension of the fabrication of mixed uranium-plutonium oxide fuel
(MOX-fuel) as a promising direction for utilisation of the reactor-grade plutonium disposed in spent
nuclear fuel. In addition, the Russian Federation and the USA are developing programmes on the
utilisation of the surplus of weapons-grade plutonium in power nuclear reactors as a real way to
nuclear disarmament [1]. In this connection, the protection of the plutonium containing MOX-fuel
against non-peaceful applications becomes a more and more urgent problem.
The solution of this problem might be found in creating protection barrier systems of different
type. Inherent radiation protection barriers are considered among them.
An important aspect of the radiation protection barriers is the duration of their action. In [2] a
possibility has been analysed to create an inherent radiation protection barrier by short-term
irradiation of manufactured fuel assemblies (FAs) in an accelerator driven facility (ADF). The
disadvantage of such an approach is the short action time of the protection barrier because of the rate
of exposure dose (RED) for γ-radiation emitted by fission products is relatively quickly reduced. So,
other combinations are required to be used for creating a long-term protection barrier. The approach
of a pre-supposed admixture of long-lived radionuclides to fuel in process of fuel fabrication may be
considered as a possible option. In this case, it is obvious that the procedures of MOX-fuel
fabrication, fuel rods and FAs manufacturing have to be performed under increased RED of γ-
radiation. Therefore, amounts of radionuclides to be admixed should be chosen in such a way to keep
the conditions of radiation safety protection of the staff.
For modern technologies of MOX-fuel fabrication the radiation safety is ensured by remote
handling with nuclear materials. As it follows from [3] the requirement to permissible RED of
γ-radiation in manufacturing of FAs for MOX-fuelled light-water reactors is about 0.6 rem/h at 1-m
distance from FA.
This extremely stringent requirement really excludes from analysis the radionuclides which are
intense sources of γ-radiation under their decay. In this case, it is interesting to address the use of the
radionuclide 232 U as an admixture to MOX-fuel.
The radionuclide 232 U is the first isotope of a long chain of decays, and decays of some daughter
radionuclides in final part of the chain are accompanied by high-energy γ-radiation. Therefore,
γ-activity of 232 U increases with a certain time delay due to gradual accumulating of emitters of highenergy
γ-radiation. Then, γ-activity of 232 U mounts to its peak value, approximately after 10 years, and
exponentially reduces with rather long half-time (about 69 years). Such a time behaviour of 232 U
activity (really, activity of 232 U decay products) is convenient for creating the inherent radiation
protection barrier which is negligible at the stage of fuel fabrication and FAs manufacturing, then,
increased and slowly reduced after attaining the peak value at stage of FAs storage, transportation and
preparation to use at nuclear power plants. Thus, admixture of 232 U to MOX-fuel provides a “window”
of low activity at stage of FAs manufacturing, and this “window” gradually self-closes afterwards.
2. Evaluation of FAs protection with MOX-fuel containing 232 U
Modern technologies of reactor fuel fabrication are characterised by the use of high-productivity
equipment, intense flows of nuclear materials and, as a consequence, relative short time of FAs
manufacturing [4].
940

By supposing that FA manufacturing takes about 10 days, then, by limiting RED of FA during
manufacturing (0.6 rem/h), it is possible to evaluate permissible amount of 232 U admixed to fuel,
which will be about 0.01%.
The functional sequence SAS2H of the computer code SCALE-4.3 [5] was used for determining
the radiation parameters of irradiated fuel. This functional sequence is able to perform numerical
analysis of a fuel composition at every time moment of its irradiation and one-dimensional transport
calculation of the system with use of a two-step procedure. At the first step, the lattice of fuel rods is
analysed while the lattice of fuel assemblies is considered at the second step. Results of transport
calculations for each time step of fuel burn-up are applied for the determination of changes in fuel
nuclide composition during irradiation. Upon completion of fuel burn-up analysis, dose parameters of
irradiated fuel are evaluated.
By using the functional sequence SAS2H, the calculations were carried out for a time behaviour
of RED at 1 m distance from FA with MOX-fuel containing 0.01% 232 U (see Figure 1). The
computations were performed with the application of 44-group neutron cross-section library
generated from the evaluated nuclear data file ENDF/B-V. It was found that RED attained 100 rem/h
in 5 years of cooling time and did not reduce below this value for the next 20 years. This level is
considered by the US Nuclear Regulatory Commission and by the IAEA as a sufficient level to
provide self-protection for spent FAs of nuclear reactors [1]. However, RED is lower than this selfprotection
level during the initial 5 years after FAs manufacturing and does not exceed 30 rem/h
during the first year.
Figure 1. RED at 1 m distance from non-irradiated FA with MOX-fuel containing 0.01% 232 U
100000
10000
ate of exposure dose, rem/h
1000
100
10
Total
RED of fission products
1
0 10 30 50
Time, yr
So, admixture of 232 U in amounts permissible for modern technologies of nuclear fuel fabrication
allows to keep protection of MOX-fuel at aforementioned level for rather long period but during the
first several years there is a temporary “window” when fuel protection is substantially less than the
required level.
941

This problem can be solved by using a combined approach to protect MOX-fuel, namely:
combination of 232 U admixture to fuel with short-term irradiation of manufactured FAs in ADF. Timedependent
recession of RED from the FA irradiated for 90 days in ADF blanket [2] is shown in
Figure 2. It can be seen that during the first two years the RED of fission products accumulated in
irradiated FA is substantially larger than 100 rem/h while during the next two years the RED remains
at only 50% of this protection level. But, in combination with the RED from 232 U decay products, total
RED becomes sufficient to ensure radiation protection of MOX-fuel at a level no less than 100 rem/h
for a 25-year period.
Figure 2. RED at 1 m distance from irradiated FA with MOX-fuel containing 0.01% 232 U
100000
10000
Rate of exposure dose, rem/h
1000
100
10
Total
RED of fission products
1
0 10 30 50
Time, yr
As mentioned above, the level of radiation protection barrier (100 rem/h) is defined by permissible
232
U concentration in fabricated fuel and, finally, by capabilities of the technology used for MOX-fuel
fabrication. For example, at MOX-fuel fabrication plant in Sellafield (United Kingdom) a semi-distant
technology is applied. Such a technology allows to fabricate MOX-fuel and to manufacture FAs with
RED at 1-m distance from ready-made FA about 0.6 rem/h.
Of course, if it is necessary, this level of protection barrier may be heightened up to the required
value. However, it leads to proportional upgrade in radiation level at stage of MOX-fuel fabrication
and requires proper application of sophisticated technology. For instance, if the value of 1 000 rem/h
for RED at 1-m distance from ready-made FA was chosen as criterion for reliable radiation
protection, then the 232 U content in fuel may be increased up to 0.1%, and RED at the end of FA
manufacturing will be about 6 rem/h. So, the value of RED adopted as protection criterion and RED
from ready-made FA differ approximately by a factor of 170.
Here, it is appropriate to underline once more an advantage of 232 U as compared with γ-active
radionuclides ( 137 Cs, for instance). This advantage is related with the delayed nature of γ-radiation
emitted by 232 U. Intensity of this radiation, negligible at initial stage, increases along with gradual
accumulation of γ-active isotopes in the decay chain of 232 U, attains a peak value and only afterwards
942

exponentially reduces with rather long half-time. Quite the contrary, intensity of γ-radiation emitted
by other radionuclides exponentially reduces just from the very beginning, i.e. from the moment of
their introduction into composition of the fuel to be protected. To have RED from the FA containing
137
Cs at level 100 rem/h after 25 years, the value of RED at fabrication stage should be about
180 rem/h, i.e. higher by a factor of 300 than that in case of 232 U admixture.
However, admixture of 232 U requires irradiation of ready-made FAs in ADF blanket to close the
low-radioactivity “window” while application of γ-active radionuclides does not require such an
irradiation.
3. About cooling time of irradiated FAs before transportation to nuclear power plants
After irradiation in ADF blanket, protected MOX FAs must be cooled for some time period until
residual heat generation reduces down to the level acceptable for their transportation in standard
containers. As known [6], the standard containers used for transportation of spent FAs discharged
from power reactors of VVER-1000 type have the following values of acceptable specific heat
generation:
• If cooling-time is half a year – 23 W per kg of fuel.
• If cooling-time is one year – 12 W per kg of fuel.
• If cooling-time is two years – 5.6 W per kg of fuel.
Time-dependent behaviour of specific heat generation in MOX FA after irradiation in ADF
blanket was evaluated by using the empirical formula presented in [7]:
W(t) = 0.07×W(0)×[t -0.2 - (t + t irrad
) -0.2 ] (1)
where t: time after irradiation, s.
W(t): specific heat generation at time moment t.
W(0): specific heat generation at the end of irradiation.
t irrad
: duration of irradiation, s.
Comparative calculations with application of the computer code SCALE-4.3 confirmed that
formula (1) satisfactorily describes residual heat generation in spent FAs of light-water reactors till
t < 10 7 s (about 4 months).
The evaluations performed with the application of formula (1) demonstrated that MOX FAs
irradiated in ADF blanket [2] for 90 days have specific heat generation at the level of 24 W/kg just in
a day after the end of irradiation, in 5 days, at the level of 13 W/kg. It means that the needed cooling
of FAs irradiated in ADF blanket does not cause real time delay for their transportation to nuclear
power plants in standard containers.
4. Evaluation of ADF productivity and needs in 232 U
In steady-state refueling regime the ADF productivity in the creation of inherent radiation
protection barrier in MOX FAs is essentially defined by the number of FAs in ADF blanket and
duration of their irradiation.
943

The evaluations obtained in [2] have demonstrated that ADF is able to ensure a sufficient level of
radiation protection for 150 FAs loaded into blanket and irradiated for 90 days. Since a conventional
light-water reactor of 1 GWe power requires about 50 FAs for annual refueling, such ADF is able to
supply with protected FAs the park of light-water reactors with a total capacity of 12 GWe. It means
that about 30 such ADFs are required to support the total nuclear power of the world.
To upgrade the productivity, it is desirable to reduce exposure time of FAs in ADF blanket as
short as possible.
It is obvious that the shortening of irradiation would lead to a proportional extension of the
serviced reactor park with the same reduction in the number of ADFs needed. However, this
shortening should not be accompanied by a decrease in the level of FA radiation protection. One
possible way to solve this problem is the disposition of FAs in ADF blanket with increased gap
between them. There are some reasons to expect that additional softening of neutron spectrum in
peripheral layers of fuel rods will intensify the build-up processes of radionuclides in these fuel rods
which are the major contributors to total radioactivity of irradiated FA.
For 232 U content in MOX-fuel at a level of 0.01%, one fresh FA contains about 50 g of 232 U.
Evaluations of 232 U burn-up in VVERs demonstrated that to the end of a 3-year fuel lifetime, the
content of 232 U in discharged FA approximately halved. Since one third part of FAs loaded in the
reactor core (~50 FAs) is replaced during the annual refueling, under gradual transition to usage of
protected MOX-fuel, for the first several years a need in 232 U will be about 2.5 kg/(GWe*year). Then,
when 232 U recycle will be closed and taking 50% burn-up of 232 U into account, only 1.25 kg
232
U/(GWe*year) is needed. Extrapolating this value to total capacity of the worldwide nuclear power
(360 GWe at the end of 1998), annual need in 232 U will be about 900 kg during the transition period
and then, in established recycle regime, about 450 kg for compensation of 232 U burn-up.
5. Some ways for 232 U production
The isotope 232 U is produced under neutron irradiation of thorium together with accumulation of
well-fissionable nuclide 233 U and considered as an undesirable by-product that impedes in high degree
the power utilisation of 233 U. The main channel for 232 U generation is the following chain of neutron
reactions:
232
Th (n,2n) 231 Pa (n,γ) 232 U (2)
Reaction 232 Th(n,2n) is a threshold reaction with high enough energy threshold (~6.5 MeV).
Since the fraction of so high-energy neutrons is relatively low in operating thermal and fast reactors, a
small amount of 232 U is generated in these reactors (about 0.1% in ratio to amount of 233 U).
Generation of 232 U in fusion facilities may be considered as a promising option because in their
neutron spectra the fraction of high-energy component is larger than in power nuclear reactors. The
evaluations presented in [8] have demonstrated that about 0.3 nuclei of 232 U per one (D,T)-reaction in
plasma may be produced in thorium-containing blanket of the ITER-type fusion reactor. It means that
about two tonnes of 232 U a year may be produced in the ITER fusion reactor under nominal neutron
loading on the first wall (1 MW/m 2 ).
944

Thus, 232 U generation by only one ITER-type fusion facility is large enough to service all nuclear
power of the world both for transition period and for established regime of 232 U feeding. Moreover,
the available reserve of the ITER productivity may be used for 2-fold enhancement of MOX-fuel
protection.
It should be noted that dose parameters of recycled MOX-fuel containing 232 U depend on fuel
purification from 228 Th.
6. Conclusions
The calculational studies have been performed, and the results obtained allow making the
following conclusions:
• The ADS with extremely moderate physical parameters of pool-type sub-critical light-water
cooled blanket can be applied to create the inherent radiation barrier in fresh MOX fuel
assemblies at the level of Spent Fuel Standard. It would promote the development of the
proliferation-resistant nuclear fuel cycles.
• Evaluation of the ADS productivity in creating the protective radiation barrier of fresh MOX
fuel assemblies demonstrates that a fraction of such ADS in park of serviced power reactors
may be rather small. In this case, such ADS operating, for example, at international nuclear
technology centres might be considered as a specialised technological facility with no
electricity production at all. The latter factor allows providing maximal simplicity of the
ADS design and operation safety.
• A combined approach to create the inherent radiation barrier was analysed. Such an approach
includes admixing of various radionuclides to MOX fuel composition under fabrication stage
followed by short-term irradiation of fabricated fuel assemblies. This combined approach
allows to have a controlled “window” of reduced radiation level at the stage of fuel
fabrication and fuel assembly manufacturing, and, afterwards, the “window” is closed due to
short-term irradiation of these fuel assemblies in the ADS blanket. It was shown in Figure 2
that admixture of 232 U to MOX-fuel in amounts permissible for modern technology of fuel
fabrication followed by short-term irradiation in ADS blanket keeps radiation protection at
the level of 100 rem/h during 25 years. It can be noted that the protection level can be
changed depending on permissible γ-background during MOX-fuel fabrication stage.
945

REFERENCES
[1] Management and Disposition of Excess Weapons Plutonium, Committee on International
Security and Arms Control, National Academy of Sciences, National Academy Press,
Washington, D.C., 1994.
[2] Shmelev A.N., Sintsov A.E., Glebov V.B., Apse V.A., 1999, Communications of Higher Schools,
Nuclear Power Engineering, Obninsk, 4, 77-82.
[3] MOX – Mixed Oxide Fuel, Document of Sellafield Visitors Centre: Briefing Notes, 1999.
BNFL WebSite, http://www.bnfl.com/.
[4] Sharov M.Yu., 1997, Proc. of the Russian International Conference on Nuclear Material
Physical Protection, Control and Accounting, Obninsk, Russia, March 9-14, Vol. 3, 617-624.
[5] Bowman S.M., 1997, NUREG/CR-0200, ORNL/NUREG/CSD-2/R5, Oak Ridge.
[6] Bagdasarov Yu.E., 1969, Technical Problems of Fast reactors, Atomizdat, Moscow.
[7] Sinev N.M., Baturov B.B., 1984, Economics of Nuclear Power. Fundamentals of Technology
and Economics of Nuclear Fuel, Energoatomizdat, Moscow.
[8] Shmelev A.N., Tikhomirov G.V. et al., 1998, Communications of Higher Schools. Nuclear
Power Engineering, Obninsk, 4, 81-92.
946

TRANSMUTATION OF LONG-LIVED NUCLIDES IN
THE FUEL CYCLE OF BREST-TYPE REACTORS
A.V. Lopatkin, V.V. Orlov, A.I. Filin
(RDIPE, Moscow, Russia)
947

1. Background
Radiation background is an integral part of nature around us. Radioactive nuclides, such as
uranium isotopes, thorium, potassium-40, carbon-14, etc. occur widely in nature. Much attention has
been paid to the consequences of high radiation exposure caused by human activities. However, lower
natural radiation background may also turn negative for the mankind. Today, natural radioactive
sources (uranium and thorium) are extracted from earth for the needs of nuclear power industry,
which in turn is producing a large amount of new radioactive materials, namely, fission products,
actinides, and irradiated structural materials. Some of these – actinides and certain fission products –
are classified as high-level long-lived waste (HLW) for the purpose of assessing long-term
environmental impact of nuclear generation. At present, there is a wide discussion in Russia and
elsewhere on various aspects of HLW management, in particular, transmutation. Usually the term
“transmutation” implies the following:
• If used in respect of actinides: incineration (fission) in neutron flux, i.e. conversion into
fission products whose biological equivalent activity is 3 to 5 orders of magnitude lower than
that of the source actinide mixture (Pu, Am, Cm, Np) if cooled for more than 5 000 years.
• If referred to long-lived nuclides from fission products: irradiation in neutron flux, to be
converted in a nuclear reaction (mostly, (n,γ)) into a stable or short-lived stable nuclide.
The suggestions concerning transmutation of long-lived nuclides in a flux of charged particles
(photons and other) have not been seriously looked into because of the low efficiency of these
processes.
To be able to assess the required transmutation scope and choose proper strategy, it is necessary
to have quantitative criteria. In most of the foreign and in many domestic studies, HLW transmutation
has been treated for long as a task of doing away with the wastes, as completely as possible. If so, the
task is hard and extremely costly to accomplish.
A group of authors’ [1] proposed to judge the demanded scope of HLW transmutation against
such quantitative criterion as radioactive properties of natural uranium consumed for the purpose of
nuclear generation. In this case, HLW should be transmuted until the biological equivalent activity of
HLW sent to disposal diminishes to that of feed natural uranium. This means actually that it is
suggested that the principle of radiation equivalence between the feed radioactive material and
radioactive waste sent to disposal should be taken as a basis for working out the policy of radioactive
materials management in the fuel cycle of nuclear power. Such equivalence can be established both at
the time of waste disposal and in a historically short reliably predicted time interval (e.g. 200-1 000 years).
Radiation hazard from natural uranium is an obvious criterion for comparing the risks from radioactive
waste put in disposal. Uranium is a natural source of radioactivity, which remains with biosphere
throughout the whole time of its existence. Taking uranium from nature, nuclear power upsets the
existing natural balance by lowering it. To return activity to nature, in the amount biologically
equivalent to that of natural uranium extracted from earth is to restore the natural radiation balance
and hence ensure that the conditions of biosphere existence will not be impaired. In other words, with
this concept, nuclear power will be waste-free in terms of radioactivity. Radiation equivalence
approach allows reasonable minimisation of HLW weight and activity. Specific conditions of HLW
disposal should correspond to local health and other standards and regulations. The most consistent
implementation of the radiation equivalence principle will be if HLW are buried in uranium-mining
areas worked out using new techniques that would meet environmental requirements.
948

In the existing open fuel cycle, in which the irradiated nuclear fuel is seen solely as waste,
radiation equivalence can be struck only after 100 000 to 500 000 years of spent fuel cooling before
disposal, i.e. is practically impossible.
To implement the radiation equivalence principle and transmute HLW, it is necessary to
reprocess all spent fuel and separate it into appropriate fractions.
2. Transmutation demand for thermal reactors
Current nuclear power is based on thermal reactors. Let us discuss the long-term variation of
potential biological hazard (PBH) 1 of irradiated fuel using the example of VVER-1000 fuel. Similar
discussion in respect of fast reactor BREST is given below. After a long storage, radiation
characteristics of spent fuel from VVER-1000 show PBH variation in the bulk of the spent fuel
accumulated worldwide (PWRs and BWRs). Following a 10-year cooling, the key PBH contributors
are strontium-90 and caesium-137 with associated decay products, and plutonium (See Figure 1).
Figure 1. Potential biological hazard (ingestion) of irradiated
fuel from VVER-1000 rated for 1 kg of irradiated actinides
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After 100 years of storage, the key contributors are Am and Pu, and in the period from 1 000 to
100 000 years the main species responsible for PBH, is Pu. Therefore, from the viewpoint of reducing
1 Potential biological hazard (PBH) is the dose of inner human exposure due to long-term consumption of
radioactive materials. Individual biological risk (IBR) of a radioactive nucleus is defined as total exposure dose
due to 1 Bq of activity of this nucleus, consumed with food or inhaled continuously in the course of 50 years,
with due regard for biological specifics of removal and impact. PBH from a mixture of radioactive nuclei is a
sum of individual PBH defined as IBH multiplied by decay rate of a particular nucleus.
949

the long-term hazard of spent fuel, it is necessary to extract Pu, Am, Cm, Cs and Sr from HLW subject
to long-term burial. The extraction and transmutation of Pu and Am are of crucial importance. Other
nuclides, considering their relatively short half-lives, may or may not be extracted, depending on the
demand from the side of fuel recycling and RW management. Curium whose main isotopes (in terms
of mass and activity) – 243 Cm and 244 Cm have T 1/2
= 29 and 18 years, respectively, should preferably be
cooled in monitored storage facilities until they decay into 239 Pu and 240 Pu which should be sent back
to the fuel cycle for incineration. It is practically impossible to transmute 137 Cs and 90 Sr on account of
low neutron cross-sections, so they can be put in monitored storage for 150 to 200 years till total
decay (T 1/2
~ 30 years). The efficiency of these steps is illustrated in Figure 2. Two cases have been
considered:
1) 0.1% of U, Pu, Am, Cm and Np go in waste subject to disposal (with the bulk of these
actinides sent to transmutation or cooling), as well as 100% of other actinides and 2% of Sr,
Cs, Tc and I present in spent fuel;
2) The same as in 1), but with 100% of all fission products sent to waste.
Figure 2. Potential biological hazard (ingestion) of waste
from 1 kg irradiated fuel of VVER-1000
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Calculations were performed assuming that SF is stored for 30 years prior to reprocessing. It was
found that in Case 1 the radiation equivalence between RW and source uranium could be struck in
80 years after SF removal from reactor. In Case 2 the equivalence would be attained in 220 years,
following the decay of 137 Cs and 90 Sr. Radiation characteristics of 9.76 kg of natural uranium used for
fabrication of 1 kg of uranium fuel with 4.4% enrichment ( 235 U), have been taken as a reference
criterion in Figures 1 and 2.
Spent fuel taken annually from VVER-1000, contains approximately 230 kg of “civil”-grade Pu
and 23 kg of minor actinides (~68% Np, 22% Am and 10% Cm). If SF is first cooled for a long time
(e.g. 30 years) before reprocessing, short-lived FPs will decay completely, 137 Cs and 90 Sr
concentrations will be twice lower, 241 Pu will decay into 241 Am, and the concentration of 243 Cm and
244
Cm will diminish by a factor of two and more. Plutonium mass in SF will decrease by ~12%, and
that of minor actinides (MA) will increase 2.2-fold with the changes in the composition which will
950

then include 31% of Np, 66% of Am and 3% of Cm. With the likely 30-year average cooling time of
VVER-1000 spent fuel before the beginning of its mass reprocessing at the RT-2 plant (planned plant
for reprocessing of all VVER-1000 spent fuel which start operating before the beginning start of fast
reactor deployment), the values of 202 kg of Pu and 51 kg of MA (or 49.5 kg without Cm) should be
taken as more correctly depicting the levels of their annual production in VVER-1000 in the analysis
of the demanded capacity of transmutation facilities.
Figure 2 shows the possibility of striking the radiation balance in a thermal reactor system
including a transmutation facility. The decision on the type of the transmutation facility is dictated by
long-term nuclear policy. At least two approaches are possible:
1) Nuclear power sector is based on thermal reactors and operates until economically
acceptable natural reserves of uranium come to an end. After that, nuclear power is phased
out. Fast reactors (FR) are not used at a practically meaningful level.
2) Nuclear power is developed based on fast reactors using Pu present in spent fuel of thermal
reactors which also remain in operation for a long time, together with FRs.
With the first approach, nuclear energy sources become but a short episode in the history of
mankind, with the need to take special measures to get rid of accumulated HLW. Plutonium can be
transmuted in thermal reactors operating in a closed fuel cycle, despite the poor present economics of
this step. This option will require cardinal changes in “fresh” fuel management strategy and will
provide ~20% increase in effective fuel resources of the nuclear power sector based on thermal
reactors. Minor actinides can be transmuted at the same time, which will somewhat lower fuel burnup.
Plutonium and MA can be transmuted in dedicated transmutation facilities widely discussed in
literature: accelerator-driven and fusion sub-critical installations, thermal and fast reactors with
molten fuel, etc. If implemented, this approach will in fact shut the door to rapid development of
large-scale nuclear power sector based on fast reactors.
The second approach virtually resolves the issue of resource limitations in nuclear power
development because of an ~100-fold increase in the efficiency of U incineration, which affords the
use of much more expensive uranium. All Pu accumulated in spent fuel of thermal reactors, will be
involved in the fuel cycle of fast reactors, becoming its integral part. Minor actinides can also be
efficiently transmuted in fast reactor cores. With lower (or nil) demand for Pu breeding, surplus
neutrons can be used to transmute long-lived fission products. Fast reactors with small reproduction
of surplus Pu (i.e. more then they need for their own purposes) allow implementing transmutation fuel
cycle in large-scale and continuing to grow nuclear power.
3. Transmutation fuel cycle
In reality, radiation equivalence can be attained if a large-scale nuclear power sector is developed
based on safe cost-effective fast reactors of a new generation and a closed fuel cycle, along with
thermal reactors (advanced VVERs, etc.) fueled with uranium. Fast reactors of a new generation with
full Pu reproduction in the core and without breeding blanket will serve not only for energy
production but also to incinerate transuranic species and long-lived FPs produced by thermal and fast
reactors.
951

In a transmutation fuel cycle including SF reprocessing with the possibility of waste fractioning,
radiation equivalence can be attained provided certain requirements are met. As follows from the
earlier studies performed by the authors’ [2], radiation equivalence is attained if:
• All SF is put in reprocessing.
• U, Pu and Am extracted during reprocessing, are totally incinerated (transmuted) as a result
of repeated in-pile irradiation, first of all, in fast reactors.
• Extracted Cm is stored for about 100 years to decay into Pu which is then incinerated in
reactors.
• No more than 0.1% of reprocessed U, Am and Cm, 0.01-0.1% of Pu, 1-10% of Np and 1-5%
of 137 Cs, 90 Sr, 99 Tc, and 129 I go in waste put in disposal.
• Np is extracted and stored for some time to be used later as source material for production of
radioisotope sources on the basis of 238 Pu, or is immediately transmuted as part of FR fuel.
• Cs and Sr are extracted during reprocessing and can be either used as heat/radiation sources
or stored in a dedicated storage facility for about 200 years until total decay into stable
elements.
99
• Tc and 129 I extracted during reprocessing, are converted in a stable state (i.e. transmuted) in
neutron flux as a result of irradiation in FR blanket.
The key prerequisites for attaining radiation equivalence, are reprocessing of all irradiated fuel
with appropriate fractioning, fast reactors used to incinerate the bulk of actinides, and intermediate
storage of high-level waste prior to final disposal.
4. Transmutation in fast reactors of BREST type
Transmutation fuel cycle can be built around lead-cooled fast reactors of the BREST type in a
closed uranium-plutonium cycle (See Table 1), which have been developed by RDIPE together with
other institutes for the last 10 years. BREST reactor development has been carried out proceeding
from the following considerations:
• Total Pu reproduction in the core without uranium blankets, with breeding ratio (BR) ~1 and
moderate power density.
• Natural (inherent) safety of the reactor, with the most dangerous accidents, such as fast
runaway, loss of coolant, fires, steam and hydrogen explosions resulting in fuel failure and
catastrophic radioactive releases, excluded deterministically.
• Minimisation of the radiation hazard of radwaste due to in-pile transmutation of the most
hazardous long-lived actinides and fission products and thorough radwaste cleaning from the
above in order to attain radiation balance between the radwaste put in disposal and uranium
extracted from earth (waste-free energy production in terms of radioactivity).
• Impossibility of using closed fuel cycle facilities to extract Pu from irradiated fuel
(nonproliferation of nuclear materials).
• Provision of the economic competitiveness of fast reactor plants with modern thermal reactor
NPPs (VVER, RBMK, BWR, PWR, CANDU).
952

Let us discuss the radiation balance of a BREST-1200 reactor. The reactor is assumed to be
running in an equilibrium state with U, Pu, Am and Np recycling in the fuel cycle and reactor madeup
with natural uranium and Np and Am from three VVER-1000. Plutonium produced in VVERs,
goes into the first cores of newly commissioned BREST reactors. Curium from BREST and VVER is
stored for about 100 years until 243 Cm and 244 Cm decay into Pu isotopes. The resultant Pu is returned to
the fuel cycle of BREST reactors. The contributions of individual species into PBH of irradiated fuel
of BREST reactors are shown in Figure 3. In the time period from 40 to 10 5 years, PBH of BREST
spent fuel is defined by Pu and Am. With the above recycling, when the bulk of U, Pu, Am, Np and
Cm are recycled and transmuted and only 0.1% of them goes to waste, as well as 100% of other
actinides and 5% of SR, Cs, Tc and I of those present in spent fuel, radiation equivalence between
RW and source natural uranium can be reached after 180 years of storage (See Figure 4). If 100% of
Cs and Sr are sent to RW, radiation equivalence can be struck in 350 years. The time needed to reach
the balance depends on Cs and Sr decay. PBH of the waste is compared with PBH from 13.7 kg of
natural uranium, which is actually the mass of U nat
needed to produce in a thermal reactor 1 kg of fuel
for the first core of BREST reactor. This value was defined taking into account that the first load
would be recycled 12 times in the course of reactor lifetime, i.e. 60 years.
Figure 3. Potential biological hazard (ingestion) of irradiated fuel from
BREST-1200 rated for 1 kg of irradiated actinides (1.06 kg of nitride fuel)
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953

Figure 4. Potential biological hazard (ingestion) of high-level waste from BREST-1200 rated for
1 kg of irradiated actinides (1.06 kg of nitride fuel and 0.132 kg of steel)
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Waste composition:
1: 1 kg of irradiated fuel + 132 g of steel.
2: 5% (Sr, Cs, Tc, I) + 100% other FP + 0.1% (U, Pu, Am, Cm) +100% (Th, Pa, Np, Bk, Cf) + 132 g of steel EP823.
3: 100% FP + 0.1% (U, Pu, Am, Cm) + 100% (Th, Pa, Np, Bk, Cf) + 132 g of steel EP823.
Steel: 123 g of stainless steel.
Natural uranium: 13.7 kg of natural uranium.
The data given in Sections 2 and 4 only illustrate the transmutation fuel cycle and the possibility
of reaching radiation equivalence. The full radioactivity balance in a well-developed nuclear power
system can be found in [2].
4.1 Some remarks concerning natural uranium consumption
• For the purpose of comparison, Sections 2 and 4 discuss natural uranium consumption
typical for the current nuclear power and that of the nearest future (within the 21st century).
Thermal reactors operate in an open fuel cycle. Critical loads of BREST reactors of the first
generation will be fabricated from reprocessed spent fuel of thermal reactors. However, with
the growth and long-term functioning of the fast reactor system, natural uranium
consumption will be going down tending to 1 t of U nat
/1 t of fission products, i.e. will
decrease ~100-fold as compared to the level assumed in Figures 2 and 4. Obviously, this will
complicate radiation balance attainment. It will be necessary to provide deeper cleaning of
the waste put in disposal from actinides, to resolve carbon-14 issue, by all means, to
transmute Np, Tc and I. (There is no need to do this within the next century. It will suffice
just to extract and store these materials, which can hardly cause serious difficulties,
considering their low specific activity [3]).
• Potential biological hazard from natural uranium has been calculated considering all
products of U decay. If long-lived decay products – 231 Pa, 230 Th and 236 Ra are separated from
U in the course of mining and returned to the earth or sent to waste, then the value of natural
uranium PBH, used in the comparison, should be reduced approximately 10-fold. This also
imposes higher demands on RW cleaning from long-lived nuclides.
954

Higher requirements for the cleaning of RW sent to waste do not undermine the effort to develop
an on-site reprocessing cycle for BREST fuel. This technique is suitable for the bulk of the fuel of
future reactors. The RW generated by the on-site cycle – and their mass will be less than 10% of that
of the recycling fuel - will be in need of additional reprocessing, probably, at some central facility.
Table 1. Main characteristics of BREST-1200 [1]
Characteristic
Characteristic
Power, MWe Fuel charge (U+Pu)N, t 60
- thermal 2 800 Pu/( 239 Pu+ 241 Pu) charge, t 7.34/4.93
- electric 1 200 Fuel life, years 5-6
Core Time between refuellings, years 1
- diameter, mm 4 750 Inlet/outlet lead temperature, °C 420-540
- height, mm 1 100 Maximum temperature of
650
fuel cladding, °C
Fuel rod diameter, 9.1, 9.6, 10.4 Maximum lead flow rate, m/s 1.6
mm
Fuel pitch in a
13.6 Steam temperature at SG outlet, MPa 24.5
square lattice, mm
Fuel UN+PuN(+MAN) Net efficiency of a generating unit, % ~43
4.2 Physics of BREST transmutation
• Minor actinides should be transmuted in the core. Transmutation may be homogeneous (i.e. as
part of fuel) or heterogeneous (in dedicated fuel rods or fuel assemblies) (See Section 5 for
further details). All MA isotopes have their own critical mass without reflector, i.e. they surpass
238
U as regards their reproduction properties. Their addition to the core will not increase the
critical charge of BREST reactor. All isotopes are efficiently incinerated in the core spectrum.
MA transmutation requires neither extra neutron expenditure nor significant extra costs, while
contributing to the main reactor objective, i.e. electricity generation.
• Tc and I (probably, other long-lived fission products as well) should preferably be
transmuted some distance from the core, in a blanket, with appropriate softening of neutron
spectrum. It is quite realistic to spend ~0.5 neutron per one fission in such transmutation. In
this, some 250 kg of Tc and I will be transmuted during effective year of irradiation. The
realistic rate of transmutation is ~7-10 % annually, i.e. I and Tc loading in the blanket will be
2 500-3 500 kg.
5. Homogeneous and heterogeneous transmutation of Am, Cm, Np in BREST core
This issue is discussed in detail in [4]. Below are summarized the main points of the discussion.
MA addition alters core performance. The level of 5% h.a. has been chosen as acceptable safe
MA inventory in the core. Given the critical core charge of 56 t h.m., this will actually mean
2 800 kg, which is equivalent to 126 annual MA yields (Np and Am) in VVER-1000, with 3-year
955

cooling prior to SF reprocessing and with extraction of MA and Pu, or to 57 annual yields (without
Cm) if SF is first cooled for 30 years (See Section 2). In effective operation year, ~100 kg of MA
from VVER-1000 will be burned while the remaining amount will be safely “kept” in the closed fuel
cycle. It would suffice to take Pu for the critical charge from 37 annual outputs of VVER-1000 spent
fuel with 30-year cooling. With such proportion between Pu and MA, there will be no problem with
MA disposition at the RT-2 plant. MA-Pu balance shows that the transmutation of MA from thermal
reactors will be carried out with their lower concentration in fuel, or only in some of BREST-1200
reactors.
There might be homogeneous (as part of fuel) or heterogeneous (in dedicated fuel rods or fuel
assemblies) MA transmutation in the core. In both cases, the transmutation rate in the core centre may
run to 8.5% for Np and 10% for Am in effective irradiation year.
Equilibrium MA content in BREST fuel is ~0.7% h.a. With Cm extraction during reprocessing
(the presence of Cm raises decay heat release 4-fold and neutron activity in fuel 300-fold) radiation
performance of fuel does not differ much from that of equilibrium uranium-plutonium fuel (in the
absence of MA recycling), except for gamma dose rate which increases an order of magnitude. MA
effect in case of in-pile irradiation is also insignificant. If Am and Cm extracted from VVER fuel, are
added to the fuel (keeping to the authorized 5% h.a.), decay power in fuel increases ~5-fold, neutron
activity – 1.5 times, gamma dose rate ~100-fold. It appears acceptable in case of remote production,
but requires greater attention to fuel cooling at all production steps.
In case of heterogeneous transmutation of the mixture of Am and Cm extracted from
VVER-1000 fuel, decay power is about 30 times higher, gamma dose rate is ~50 times greater than
those of equilibrium fuel of BREST using its own Np and Am. The initial multiplication properties of
(Np+Am)N fuel and of equilibrium BREST fuel are approximately the same, however, burn-up
performance differs dramatically. On in-pile irradiation, decay heat in (Np+Am)N fuel rods may be
some 1.5-2.5 times greater (depending on fuel composition and density) than that in BREST fuel
placed nearby, which is intolerable and calls for addition of inert material. This, however, impairs
transmutation efficiency and changes core performance. Loading intervals for Np- and Amcontaining
FAs are at variance with the standard loading of BREST fuel. “Pure” critical mass of
Np+Am mixture is less than 100 kg. Moreover, this circumstance makes heterogeneous transmutation
undesirable from nonproliferation point of view as well. Though it is very difficult to make a charge
from this material, the very possibility of doing so cannot be excluded.
6. Local radiation equivalence
The developed groundwork for a transmutation fuel cycle allows making fundamental strategic
decisions concerning the lines of nuclear power development in Russia. The cornerstones of future
nuclear power are reprocessing of all fuel from thermal and fast reactors, transmutation of long-lived
nuclides, predominant deployment of fast reactors of the new generation, high-active fuel and
relatively “clean” wastes. However, this is but a first step in the development of the scientific basics
for creation of environmentally acceptable nuclear power. The next step is substantiation of the areas
of final disposal of radioactive waste based on the preservation of local radiation balance and nature.
Most consistently, this condition can be met in uranium mining areas. Pioneered in Russia by
B.V. Nikipelov, work has been started on developing reference principles for such disposal. Needless
to say, this does not mean rejection of existing developments in the area of RW disposal; they just
should be reappraise. The key modern principle in the substantiation of the environmental strategy of
956

RW disposal is multibarrier configuration in case of high-level waste, which is hard to prove for the
time of RW storage of 10 4 -10 6 years.
The problem of preserving local radiation equivalence is closely connected with uranium mining
technique and management of associated long-lived Pa, Th and Ra actinides. Once separated from
uranium, they are sent to surface tails, which seriously disturbs local radiation balance, so that the
problem of rehabilitation becomes insoluble. It is hardly possible to change what has already been
done, but in future, in a sensibly organized nuclear power sector, the approach to U mining should be
different. From the viewpoint of contamination and subsequent rehabilitation of the area, uranium
mining is no less important than the RW minimisation and disposal technique. Definition and
substantiation of requirements for environmentally safe uranium mining needs further investigations
as well.
Apart from radiation-equivalent activity, it is necessary to take into consideration heat balance,
radionuclide migration as a result of matrix failure, etc. Thus, if migration is taken into account, it is
necessary to pay attention not only to Pu, MA, 99 Tc, 129 I, but also to 126 Sn, 135 Cs, 79 Se, 231 Pa, which play
insignificant role in the analysis of the total radiation balance of nuclear power.
The line mentioned in the title of the Section, requires comprehensive systematic studies.
7. Conclusions
Transmutation of long-lived nuclides produced as a result of nuclear generation, should be set up
proceeding from the principle of reasonable sufficiency, expressed as radiation equivalence between
the radwaste sent to disposal and source natural uranium. In this case, introduction of fast reactors of
new generation (such as BREST or other reactors based on similar philosophy) will resolve
transmutation problems even with the thermal-to-fast reactor capacity ratio of 2:1. The authors of the
“Strategy of nuclear power development in Russia” [5] foresee, and substantiate their prediction, that
fast reactors of the new generation will account for no less than 2/3 of nuclear capacity in future
large-scale nuclear power sector. Fast reactors will be the basis of a transmutation fuel cycle, which
will remove the need of creating additional transmutation facilities.
957

REFERENCES
[1] White Book of Nuclear Power, Ed. E. Adamov, 1st edition, RDIPE, Moscow, 1998.
[2] E. Adamov, I. Ganev, A. Lopatkin, V. Muratov, V. Orlov. Transmutation Fuel Cycle in Largescale
Nuclear Power of Russia, Monograph, RDIPE, Moscow, 1999.
[3] B. Gabaraev, I. Ganev, A. Lopatkin, V. Orlov, V. Reshetov, Regional Storage Facility for
Long-term Monitored Storage of Long-lived High-level Waste, RDIPE Preprint, Moscow, ET-
2000/51.2000.
[4] I. Ganev, A. Lopatkin, V. Orlov, Homogeneous and Heterogeneous Transmutation of Am, Cm,
Np in BREST Cores, (To be published soon in Atomnaya Energiya).
[5] Strategy of Nuclear Power Development in Russia in the First Half of 21st Century. Summary.
Minatom of Russia. Moscow, 2000.
958

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