PROGRESS REPORT - ENEA - Fusione

PROGRESS REPORT
2006
Nuclear Fusion and Fission, and Related Technologies Department
ITALIAN NATIONAL AGENCY FOR NEW TECHNOLOGIES ENERGY
AND THE ENVIRONMENT

This report was prepared by the Scientific Publications Office from contributions provided by the scientific and
technical staff of ENEA’s Nuclear Fusion and Fission, and Related Technologies Department.
Scientific editors: Paola Batistoni, Adriana Romagnoli, Gregorio Vlad
Design and composition: Marisa Cecchini, Lucilla Crescentini, Lucilla Ghezzi
Artwork: Flavio Miglietta
English revision: Carolyn Kent
See http://www.fusione.enea.it for copy of this report
Cover picture: The six BNC cables and
Piccolo–Micromegas assemby inside the TRIGA
reactor
Published by:
ENEA - Nucleo di Agenzia
Edizioni Scientifiche,
Centro Ricerche Frascati,
C.P. 65
00044 Frascati, Rome (Italy)
Tel: +39(06)9400 5016
Fax: +39(06)9400 5015
e-mail: cecchini@frascati.enea.it

A FUSION PROGRAMME 6
Contents
A1 MAGNETIC CONFINEMENT 6
Introduction 6
A1.2 FTU Facility 7
A1.3 Experimental Results 8
Lower hybrid current drive studies in ITER-density-relevant plasmas 8
Liquid lithium limiter experiment 10
MHD real-time control experiment 12
Electron cyclotron current drive experiment 12
Disruption studies 13
Dusty plasmas 14
A1.4 Plasma Theory 16
Theory of beta-induced Alfvén-eigenmodes 17
Electron fishbones: theory and experimental evidence 17
Analysis and modelling of LHW propagation in toroidal plasmas
by asymptotic methods 18
Modelling of the ICRH experiment on JET 19
Simulation of burning plasma dynamics by ICRH accelerated minority ions 20
Particle simulation of bursting Alfvén modes in JT–60U 21
Theory of Alfvén waves and energetic particle physics in burning plasmas 23
Nonlinear equilibria, stability and generation of zonal structures in toroidal plasmas 23
A1.5 JET Collaboration 24
Participation in the JET EP/EP2 24
Participation in experimental campaigns C15-C17 26
A1.6 Proto–Sphera 29
A2 PRELIMINARY DESIGN OF FT3 32
Introduction 32
A2.2 Scientific Motivation of the Proposal 33
A2.3 Preliminary Design Description 36
A3 TECHNOLOGY PROGRAMME 40
Introduction 40
A3.2 Divertor, First Wall, Vacuum Vessel and Shield 40
Manufacturing of small-scale W monoblock mockups 40
Engineering Design Activities: V and VI test campaigns 42
Hydraulic characterisation of full-scale divertor components 42
H permeation through EUROFER and heat exchanger material (Incoloy, Inconel) 43
Formal trials for the new ITER divertor cassette refurbishment 43
A3.3 Breeder Blanket and Fuel Cycle 44
DEMO breeding blanket 44
European Breeding Blanket Test Facility 44
Thermo-mechanical characterisation of HCPB mockup 44
TRIEX loop for studying technologies for extracting tritium from Pb-17Li 46
Conceptual design of auxiliary systems for HCPB-TBM 46
Structural analyses during em loading 46
VDS catalyst tests 47
Permeator tubes 48

Contents
A3.4 Magnet and Power Supply 48
ITER magnet casing welds 48
ITER pre-compression ring fibreglass composite material 48
High-frequency/high-voltage solid-state modulator for ITER gyrotrons 48
A3.5 Remote Handling and Metrology 48
A3.6 Neutronics 50
Quality assurance for neutronics analysis for ITER 50
ITER systems: nuclear design 51
TBM HCPB and HCLL neutronics experiments 51
Experimental validation of neutron cross sections for fusion-relevant materials 52
A3.7 Materials 53
Flat-top indenter for mechanical characterisation 53
A3.8 IFMIF 54
Remote handling of the back-plate bayonet concept – bolted solution 54
Lithium corrosion and chemistry: LIFUS III facility 54
Preliminary remote handling handbook for IFMIF facilities 55
Inventories and dose rates induced by deuterons and neutrons in
the accelerator system 56
Inventories and dose rates induced by deuterons and neutrons in
the cooling system 56
A3.9 Safety and Environment, Power Plant Studies and Socioeconomics 56
Failure mode and effect analysis for the European test blanket modules 56
Failure mode and effect analysis for remote handling transfer systems of ITER 57
Validation of computer codes and models 57
Dust removal experiments in STARDUST 58
Feasibility study of a torus-shaped facility for dust mobilisation studies 58
Post-accident occupational exposure and radioprotection 58
Integration of design modifications (in Rapport Préliminaire de Sûreté)
to tritium building and detritiation system 59
Collection and assessment of data related to JET occupational
radiation exposure 60
JET data collection on malfunctions and failures of ICRH system components 60
JET dust in-vitro experiment: result assessment and in-vivo experiment
literature review 60
Study on recycling of fusion activated material 61
A4 SUPERCONDUCTIVITY 62
Introduction 62
A4.2 ITER and ITER-Related Activities 62
ITER toroidal field cable conductor 62
Current redistribution study on ITER conductors 64
EFDA dipole 64
Barrel bending experiments 65
Optimisation of NbTi strand for PF1/PF6 performance 65
A4.3 JT-60SA 66
A4.4 High–Temperature Superconducting Materials 66
Evolution and control of cube texture in Ni-W substrates for YBCO-coated
conductors 66
Nickel-copper alloys as textured substrates for YBCO–coated conductors 68
MOD-TFA YBCO films 69

Introduction of artificial pinning sites in YBCO films 70
Magnetic characterisation of superconducting wires for fast ramped
superconducting dipoles 71
MARIMBO experiment: application of MgB 2 72
Transport and thermal stability characterisation of HTS wires and tapes:
analysis of quench propagation on YBCO-coated conductors 73
A5 INERTIAL FUSION 74
A6 PUBLICATIONS, PATENTS AND EVENTS 78
A6.1 Publications 78
Articles 78
Articles in course of publication 81
Contributions to conferences 82
Reports 86
A6.2 Patents 86
A6.3 Conferences and Events 87
A6.4 Seminars 87
B FISSION TECHNOLOGY 88
B1 R&D ON NUCLEAR FISSION 88
B1.1 Innovative Fuel Cycles Including Partitioning and
Transmutation 88
Partitioning technology 88
Transmutation systems and related technology 90
VELLA - Virtual European Lead Laboratory 102
B1.2 Evolutionary and Innovative Reactors 102
International Reactor Innovative and Secure 103
European Lead-Cooled Fast System 104
Very high temperature reactor 106
B1.3 Nuclear Safety 107
Code validation and accident analysis 107
Severe accident analysis 109
Reliability and risk analysis 111
B1.4 Nuclear Data 112
General quantum mechanics 112
Nuclear reaction theory and experiments 113
Nuclear data processing and validation 113
Computer code development 115
Radioactive ion-beam production for nuclear-structure studies 116
B1.5 TRIGA RC-1 and RSV TAPIRO Plant-Operation for Application
Development 117
B2 MEDICAL, ENERGETIC AND ENVIRONMENTAL APPLICATIONS 118
B2.1 Boron Neutron Capture Therapy 118
The epithermal column EPIMED at TAPIRO 118
Employment of the thermal column HYTHOR at TAPIRO 121

Study of BNCT applied to lung tumours 121
Design of a facility at TRIGA to treat explanted livers 122
B2.2 Solar Thermal Energy 123
B2.3 Development Activities for Antarctic Drilling 124
B3
PARTICIPATION IN INTERNATIONAL WORKING GROUPS
AND ASSOCIATIONS 128
B4 PUBLICATIONS 130
B4.1 Publications 130
Articles 130
Reports 133
Contributions to conferences 134
C NUCLEAR PROTECTION 140
C1 RADIOACTIVE WASTE MANAGEMENT AND ADVANCED
NUCLEAR FUEL CYCLE TECHNOLOGIES 140
Introduction 140
C1.2 Entrustment of ENEA’s Fuel Cycle Facilities and
Personnel to Sogin 140
C1.3 Characterisation, Treatment and Conditioning of Nuclear
Materials and Radioactive Waste 140
C1.4 Radioprotection and Human Health 142
Methodological proposal for the evaluation of a physiological comfort
index in indoor environments 142
LCA of strippable coating and the principal competing technology
used for nuclear decontamination 143
C1.5 Integrated Service for Non-Energy Radwaste 143
C1.6 Transport of Nuclear Material 144
Packaging for transport of radioactive material 144
C1.7 Disposal of Radioactive Waste 145
Artificial barriers for disposal units 145
D MISCELLANEOUS 146
D1 Advances in the IGNITOR Programme 146
D2 Ultra-Pure Hydrogen Production 147
D3 Non-ITER Activities 148
D4 Condensed Matter Nuclear Science 149
ORGANISATION CHART 152
ABBREVIATIONS AND ACRONYMS 154

Preface
This report describes the research activity carried out
during 2006 by the laboratories belonging to the ENEA
Nuclear Fusion and Fission, and Related Technologies
Department (Dipartimento Fusione, Teconologie e
Presidio Nucleari (FPN)).
An important point to note is that during 2006 ENEA
implemented a new organisation that combines fusion
and fission activities in the same department FPN. This
choice is clearly advantageous for both fields.
In the fusion field, a historical event took place in 2006 -
the signature of the agreement for the construction of
ITER in Europe (Cadarache). ITER concentrates the efforts
of the most advanced countries in the world in utilizing fusion as a safe, environmentally sustainable
and inexhaustible energy source. The participants in this challenging enterprise are Europe, China,
Korea, India, Japan, the Russian Federation and the United States.
ENEA, in the framework of the Euratom-ENEA Association for fusion, continues to contribute to
broadening plasma physics knowledge as well as to developing the relevant technologies. This
report describes the 2006 research activities carried out by the ENEA Fusion Research Group of the
FPN with the contributions of other ENEA research groups. The following fields were addressed:
magnetically confined nuclear fusion (physics and technology), superconductivity and inertial
fusion.
During 2006 the scientific activity at the Frascati Tokamak Upgrade (FTU) was focussed on ITERrelevant
aspects of plasma scenarios. In the meantime the conceptual design activity and a report
discussing the scientific motivation of the FT3 device in the context of the European Accompanying
Programme were completed. This new proposal also involves the other participants in the ENEA-
Euratom Association, namely, the Reversed Field Pinch Experiment (RFX) Consortium and the
National Research Council (CNR) Milan. The technological R&D programme was performed in the
ITER framework and under the Broader Approach Agreement with Japan. Collaboration with
industry in view of the participation in construction of ITER was further strengthened.
In the fission field one of the most important events at international level was the Global Nuclear
Energy Partnership (GNEP), launched by the USA Government. The GNEP is a comprehensive
strategy aimed at making it possible to use economical, environmentally responsible nuclear energy
to meet growing electricity demand worldwide, while virtually eliminating the risk of nuclear material
misuse. This initiative, together with the Generation-IV projects and the 6th European Union
Framework Plan, is the reference frame in which the ENEA FPN Department operated during 2006
in the nuclear fission field.
ENEA benefits from a wide range of collaborations with other international research centres and
with industry. Several patents were granted in 2006 and spin-off activities are in progress. High-tech
services are supplied to Italian industry.
The work summarised in this report is amply documented in published articles and conference
communications (most of which invited).
Frascati, December 2006
Alberto Renieri

A1 Magnetic Confinement
A Fusion Programme
Scientific activity at the Frascati Tokamak Upgrade (FTU) continued to be focussed on ITER-relevant
aspects of plasma scenarios. Reliable plasma operations with lithizated walls were achieved thanks to the
newly installed liquid lithium limiter (LLL). A new experimental activity on dust creation/mobilisation was also
started. Good results have been obtained although the 2006 experimental activity was somewhat limited.
The spring campaign was first delayed by lightening hitting the electrical substation and then shortened
because of an optical window cracking due to focus deterioration of the laser beam of the Thomson
scattering diagnostic. Including a short autumn campaign, the whole 2006 experimental activity fully
dedicated only 27 days to scientific programmes, out of a total of 50 operational days.
The objective to push the performance of the wide (r/a ≥ 0.6, with r the radial coordinate and a the minor
radius of the torus) internal transport barriers (ITBs) obtained in 2005 was not pursued in 2006 because of
limited availability of lower hybrid (LH) power, necessary for controlling the current profile at higher plasma
current and density. Experiments and studies were concentrated on ion transport in the presence of electron
ITBs with ion heating determined by electron-ion collisional energy transfer.
The LLL installed in 2005 allowed operations with extremely clean plasmas where the content of heavy Z
impurities, typical of FTU metallic operations, was close to zero. In these conditions, discharges exhibit
better confinement (~20% above ITER-97L), comparable with results obtained with freshly boronized walls.
With the LLL acting as the main limiter new regimes exhibiting peaked density profile, up to density limit
values, were found.
Experiments dedicated to magnetohydrodynamic (MHD) control were aimed at improving m=2 mode
stabilisation by modulating the electron cyclotron (EC) power in phase with the island rotation and at
enhancing the signal-to-noise ratio to better identify the electron cyclotron heating (ECH) absorption
position. Disruptions (induced by impurity injection and the density limit) were mitigated by electron
cyclotron resonance heating (ECRH) power and were completely avoided when the power deposition
coincided with the location of the modes responsible for the disruptions.
Theoretical and experimental activities concerning dust in the plasma scrape-off layer (SOL) were
successfully started in collaboration with the universities of Naples and Molise and with the Max Planck
Institute for Extraterrestrial Physics. In the experimental work, in particular, evidence of dust particles
collected in the SOL of FTU discharges was found on Langmuir probes.
Mutual and positive feedbacks between theory and experiments led to i) clear identification of highfrequency
MHD activity in FTU; ii) modelling of ion cyclotron resonance heating (ICRH) experiments on the
Joint European Torus (JET); iii) numerical simulation of energetic ion transport and nonlinear Alfvénic
fluctuations in situations of experimental relevance in present-day experiments, in the framework of a
collaboration with the Japan Atomic Energy Research Institute (JAERI) JT-60U team.
More basic activities were focused on electron-fishbone mode excitations by LH additional power, the
propagation and absorption of radiofrequency (rf) waves in toroidal plasmas, the investigation of energetic
ion dynamics in burning plasmas, and activities on plasma turbulence and turbulent transport.
In 2006 the JET experimental campaigns were put off to July, and the participation of Frascati scientists was
then limited to restart and high-level commissioning of the JET Enhancement Programme (EP) and the
organisation of new enhancements for JET-EP2. The main experimental activity was resumed in autumn
Progress Report 2006
6

2006, with ENEA having direct responsibility for experiments both in Task Force S1, regarding hybrid scenarios with
dominant electron heating, and in Task Force S2, on high-performance ITBs.
The FT3 conceptual design activity and a report discussing the scientific motivation of the device in the context of
the European programme were completed. Various plasma scenarios can be investigated and it is shown that the
various heating systems are capable of producing the plasma conditions needed for the ITER physics investigation.
The report includes a preliminary design of the machine, auxiliary heating systems and diagnostics, and a
preliminary assessment of different sites and construction and operation costs.
Finally, the construction of the poloidal field shaping coils of MULTI-PINCH (initial set-up of PROTO-SPHERA)
continued during 2006 and will be completed by the beginning of 2007. The collaboration with the United Kingdom
Atomic Energy Authority (UKAEA) continued, mainly on modelling and on experiments aimed at plasma start-up and
plasma current ramp-up in the absence of the central solenoid in MAST.
A1.2 FTU Facility
During 2006 the FTU machine achieved 91% of successful pulses, continuing the high level of reliability of
the previous years.
Experimental work started at the end of March and continued up to the first week of July without
suspensions. The second experimental session ran from mid-September to mid-October. In 2006, 1144
shots were successfully completed out of a total of 1257 performed over 50 experimental days. The
average number of successful daily pulses was 23.11. Table A1.I reports the main parameters for
evaluating the efficiency of the experimental sessions. Figure A1.1 reporting the indicator trend from 1999
up to 2006 shows that experimental time and successful pulses are stable, while experimental days are
lower due to power supply problems and to a vacuum loss caused by a hole in the scattering window.
For the control and data acquisition system:
a) Work was started on developing a software framework to obtain a user-friendly environment for carrying
out all the phases (i.e., control law design, simulation, automatic source code generation, debug and
software release) related to the FTU real-time control system. A software simulation tool was also
implemented and released. The whole work should be finished by 2007. A 10 PC Cluster has been
installed to allow FTU data analysis in a Linux environment. In the initial phase the cluster is employed
as a test-bed to characterise real-time network protocols suitable for ITER.
b) A set of computing resources was released on the EGEE-GRID (i.e., Enabling Grids for E-sciencE) site
of ENEA for the FUSION Virtual Organisation: in particular a 1-TB storage area is available for use by the
Integrated Tokamak Modelling Task Force.
c) A web tool was developed to handle the configuration of a data acquisition system (DAS) similar to the
FTU control and data acquisition system (CODAS) and with the same data and parameter configuration.
d) Work was started for a European Fusion Development Agreement (EFDA) task aimed at achieving a
fully revised version of the ITER control data access and communication (CODAC) specifications ready
for fusion internal review. In particular, ENEA has to revise the CODAC documentation, bringing it up to
date for April 2007; prepare and organise internal and external reviews (including experts outside fusion)
and a peer review of the CODAC design in agreement with the ITER International Team and ITER
Participant Teams; incorporate into the CODAC design common proposals that will have to be
discussed in the review process.
To model the CODAC structure, the capabilities of UML language were studied. Preliminary results
indicated that Matlab/Simulink could be suitable for the final design work, but a hybrid solution (UML code
into Matlab/Simulink diagram) is being investigated.
7
Progress Report 2006

A1 Magnetic Confinement
Table A1.I – Summary of FTU operations in 2006
Jan. Feb. March April May June July Aug. Sept. Oct. Nov. Dec. Total
A Fusion Programme
Total pulses 0 0 103 130 268 385 0 0 120 251 0 0 1257
Successful pulses (sp) 0 0 97 117 251 343 0 0 113 223 0 0 1144
I(sp) 0.94 0.90 0.94 0.89 0.94 0.89 0.91
Potential experimental days 0.0 0.0 8.0 11.0 10.5 17.0 4.0 0.0 12.0 8.5 0.0 0.0 71.0
Real experimental days 0.0 0.0 4.0 6.5 10.5 15.5 0.0 0.0 4.5 8.5 0.0 0.0 49.5
I(ed) 0.50 0.59 1.00 0.91 0.00 0.38 1.00 0.70
Experimental minutes 0 0 1680 2136 4555 6294 0 0 2137 3913 0 0 20715
Delay minutes 0 0 743 1888 1943 3098 0 0 815 1388 0 0 9875
I(et) 0.69 0.53 0.70 0.67 0.72 0.74 0.68
A(sp/d) 24.25 18.00 23.90 22.13 25.11 26.24 23.11
A(p/d) 25.75 20.00 25.52 24.84 26.67 29.53 25.39
Delay per system (minutes)
Jan. Feb. March April May June July Aug. Sept. Oct. Nov. Dec. Total %
Machine 0 0 24 89 104 266 0 0 98 148 0 0 729 7.4
Power supplies 0 0 366 471 669 867 0 0 322 282 0 0 2977 30.1
Radiofrequency 0 0 0 20 87 286 0 0 0 100 0 0 493 5.0
Control system 0 0 16 17 158 352 0 0 89 120 0 0 752 7.6
DAS 0 0 77 30 156 116 0 0 6 42 0 0 427 4.3
Feedback 0 0 0 8 52 46 0 0 0 81 0 0 187 1.9
Network 0 0 0 0 0 0 0 0 0 0 0 0 0 0.0
Diagnostic systems 0 0 27 174 121 200 0 0 15 104 0 0 641 6.5
Analysis 0 0 149 179 173 492 0 0 108 393 0 0 1494 15.1
Others 0 0 84 900 423 473 0 0 177 118 0 0 2175 22.0
TOTAL 0 0 743 1888 1943 3098 0 0 815 1388 0 0 9875 100
Fig. A1.1 – Indicator trend from 1999 up to
2006. I(sp): successful/total pulses. I(et):
real/total experimental time. I(ed): real/total
experimental days
A1.3 Experimental Results
Lower hybrid current drive studies in ITER-density-relevant plasmas
The LH radiofrequency (f LH =8 GHz) heating system in FTU is used mainly to create and maintain
radial profiles of the toroidal current j(r) that are suitable for sustaining plasma regimes with an ITB
that improves core plasma confinement. The ECH radiofrequency system (f EC =140 GHz) facilitates
this task, and the rf waves of both interact only with electrons. Ions are, instead, heated only via
collisional damping of the hotter electrons. These regimes are currently the most valuable option for
steady-state operation in ITER and future tokamak reactors. Indeed, the better confinement
0.9
0.7
0.5
I(ed)
I(et)
1999 2001 2003 2005
Years
I(sp)
Progress Report 2006
8

Fig. A1.2 – Ion thermal conductivity vs normalised minor radius, for the
highest density a) and the widest b) steady ITB discharges. Experimental
(χ i,exp ) and neoclassical (χ i,neo ) ion thermal conductivities are shown in
full lines, while dotted segments limit the variability range during the
whole ITB phase. Also shown are the ITB radial location and the ion
thermal conductivity range during the Ohmic phase
m 2 /s m 2 /s
1
0.5
χ i,exp
0
χ
1 i,neo
0.5
OH-exp
OH-exp
χ i,neo
r ITB
# 26671 highest n e
# 27928
b)
widest radius
r ITB
a)
obtained in ITB regimes would allow operation at a lower 0
plasma current I p in order to obtain the same τ E with respect
0 0.2 0.4 0.6
r/a
to the standard scenario. In addition, demands on the
external current drive (CD) sources would be greatly reduced because the self-generated bootstrap current
I bs would increase due both to the lower I p , and to the steeper pressure radial gradients arising in reduced
transport conditions, I bs /I p ∝I p
2·∇p.
The peculiarity of FTU is that ITBs can be formed by using electron heating and current drive with no direct
ion heating or external momentum injection, which is a similar condition to that foreseen for ITER. The
additional capability to establish an ITB starting from a fully relaxed current profile at high density makes
FTU unique. Unfortunately, in 2006 the LH and ECH performances were not at the level required for
significant experimental progress, so activity in the ITB field was focussed mainly on exploiting at best the
data previously obtained and on preparing the 2007 experiments.
Crucial questions to be answered for ITER concern the effect collisional energy transfer between electron
and ions has on ITBs and how electron ITBs, with little or no induced rotation, affect ion transport. Here
the FTU contribution may be important and indeed the 2005 report illustrates the encouraging results on
the first point, while ion transport has been treated recently in an overview of FTU results [A1.1] and in a
more dedicated paper [A1.2]. Figures A1.2a) and A1.2b) plot the ion thermal conductivity χ i during an ITB
as a function of radius for the two most representative steady discharges, one obtained at the highest
density (fig. A1.2a)), and the other at the widest radius (fig. A1.2b)). The vertical bars limit the variability
range of χ i during the ITB evolution phase. Irrespective of the ITB radius (r ITB ) χ i appears to drop below
neoclassical at r≤r ITB . Although the magnitude of χ i,neo might be overestimated due to the uncertainty on
the safety factor q(r), χ i,neo ∝q 2 , for two very different ITB discharges χ i drops just at the barrier footprint
and falls even below the value it has in the Ohmic phase. This is of particular relevance since energy
transport is usually faster if the temperature increases, while Ohmic temperatures are lower. Therefore a
q(r) profile with low shear (which is typical of ITB regimes) appears suitable for reducing not only the
electron but also the ion transport, without the support of induced plasma rotation. Although limited so far
to low ion central temperature values T i0 this result is promising for ITER. Consistently, the drop in the
turbulence level close to the barrier foot derives from decorrelation of the modes that could affect both
electron and ion transport [A1.3, A1.4].
The possible application of lower hybrid current drive (LHCD) to ITER, however, still has to satisfy the
requirement of high efficiency η CD . Previous FTU results [A1.5] show that η CD does not degrade up to and
χ i,exp
[A1.1] V. Pericoli–Ridolfini et al., Proc. 21 st IAEA Fusion Energy Conference (Chengdu 2006), on line at: http://wwwnaweb.iaea.org/napc/physics/FEC/FEC2006/papers/OV_3-4.pdf,
and submitted to Nuclear Fusion
[A1.2] V. Pericoli–Ridolfini et al., Proc. 21 st IAEA Fusion Energy Conference (Chengdu 2006), on line at: http://wwwnaweb.iaea.org/napc/physics/FEC/FEC2006/papers/EX_P1-15.pdf
[A1.3] V. Pericoli–Ridolfini et al., Plasma Phys. Control. Fusion 47, B285–B301 (2005)
[A1.4] M. De Benedetti et al., Proc. 32 th EPS Conference on Plasma Physics (Tarragona 2005), on line at:
http://epsppd.epfl.ch/Tarragona/pdf/P4_035.pdf
[A1.5] V. Pericoli–Ridolfini et al., Nucl. Fusion 45, 1386-1395 (2005)
References
9
Progress Report 2006

A1 Magnetic Confinement
A Fusion Programme
J LH (r) (arb. units)
q(r)
1
0.8
0.6
0.4
FRTC + scattering
#27928, t=0.7s
Hard-X ray
(FEB)
0.2
FRTC - NO
0
scattering
0 0.05 0.1 0.15 0.2 0.25 0.3
r(m)
10
8
6
4
2
LHstar
Conventional
MSE
0
0 0.2 0.4 0.6 0.8 1
r/a
Fig. A1.4 – Safety factor profile q(r) vs normalised
minor radius for a current-hole discharge of JET:
comparison between LHCD radial profiles from
experiment (MSE diagnostic, dashed line), from the
recently developed LHstar code (full line) and from
conventional calculation (dotted-dashed line)
beyond the ITER density (line average
n _ e =1×1020 m -3 ), while the favourable scaling of
η CD with electron temperature T e leaves hope for
the desired value >0.35×10 20 Am -2 /W. However,
the lower ITER LH frequency (f LH,ITER =5 GHz)
may induce some concern on the basis of the
past results on ASDEX [A1.6] and JET [A1.7].
The ratio of plasma to wave frequency (f pe /f LH ),
which could be an important parameter, had to
be f pe /f LH ≤15, while in ITER it will be ~18.
Although f pe /f LH ≤15 in FTU, some precursors of
efficiency loss started to appear: the frequency
spectral broadening of the LH pump was no
longer negligible, the main cause being the
interaction of the LH waves with the edge
plasma. Here, the low temperatures, even more
than 100 times below the core, and the relatively
high densities, larger than 0.1 times the core,
can either exalt the linear scattering on density
fluctuations or trigger nonlinear phenomena,
such as parametric decay instability (PDI). Both
effects, which however may also coexist, can
cause noticeable degradation of the N ||
spectrum and of the trajectories of the launched
LH waves (N || is the parallel index of refraction
and governs the LHCD efficiency).
In this context both effects were modelled by considering the available data. For turbulent scattering
in the SOL, the model follows the one proposed in [A1.8]. More details can be found in [A1.9]. For
the scattering case, figure A1.3 reports a comparison of the LH power radial deposition derived from
the fast electron bremsstrahlung (FEB) camera and the deposition according to the newly
developed fast ray tracing code (FRTC) and to the conventional calculation [A1.10]: only when
scattering is taken into account is there good agreement with the experiment. The case considered
is the ITB discharge in figure A1.2b). For the PDI case figure A1.4 shows the fairly good agreement
for JET between the q(r) profiles derived from the motional Stark effect (MSE) diagnostic and those
calculated with the newly developed code LHstar [A1.11], which takes into account the nonlinear
interaction LH waves-edge plasma. Conversely the agreement is a good deal poorer for the profile
calculated conventionally.
Liquid lithium limiter experiment
Fig. A1.3 – Comparison between LHCD radial profiles in
FTU computed by FRTC: LH wave edge scattering by
density fluctuations (full line, FRTC+scattering), no
scattering (dash-dotted line, FRTC-NO scattering), and the
experiment (dashed line, same discharge as in fig. A1.2,
hard x ray [FEB])
During 2006, experiments to test a liquid lithium limiter with a capillary porous system (CPS)
configuration on FTU [A1.12, A1.13] were continued and a full analysis of the first results obtained
at the end of 2005 was performed. The programme in collaboration with TRINITI & Red Star
(Russian Federation) was also begun: the aim is to investigate the behaviour of liquid lithium in view
of its possible application as plasma-facing material and in the framework of a more general study
on liquid metals. Lithium was chosen because of its low atomic number, good thermal properties
Progress Report 2006
10

Fig. A1.5 – Plasma viewed by a visible CCD camera. At the bottom a bright ring
separated from the main toroidal limiter by a darker zone is clearly visible. LLL is
located on the far bottom right, where the glow is most intense
and strong capability to pump deuterium and impurity
particles. The LLL, composed of three similar units with
dimensions respectively of 100 mm and 34 mm in poloidal and
toroidal directions has been inserted 1.0-2.0 cm within the
SOL, from the bottom vertical port 1. It has been used for
depositing a thin Li film on the FTU metallic walls during
plasma discharge (lithization) and as a liquid material facing the
plasma.
Infrared and visible detectors viewing the LLL surface, plus Langmuir probes placed 5 mm from the LLL
leading edge, have been used to determine surface temperature [A1.14], Li release, electron density and
temperature in the SOL plasma at the LLL position. In 2006, experiments in Ohmic conditions confirmed
the previous results [A1.15, A1.16]. Plasma discharges with heating power up to 0.85 MW are
characterised by the lowest Z eff , P rad and D α signals (as monitor of particle recycling) ever observed on
FTU, and whether the LLL is inserted or not inside the vacuum chamber makes no substantial difference.
Strong modifications occur in the SOL [A1.15, A1.17] with respect to the standard metallic wall conditions.
Electron temperature increases by more than ΔT e ,SOL~10 eV, due to the strong reduction in deuterium
and impurity recycling together with the low radiation from Li atoms/ions eroded by the walls. When the
LLL is inserted inside the vessel, instead, the liquid surface represents a strong localised source of Li
atoms/ions, which increases radiation losses in a region that is close to the LLL poloidal location and
toroidally uniform. Figure A1.5 shows a plasma image recorded by a visible CCD camera. The radiation in
front of the LLL surface reduces the power flux onto the limiter surface which, in turn, is able to sustain
thermal loads exceeding 5 MW/m 2 with no damage and no lithium bloom occurring. Thermal analysis with
the ANSYS code together with the interpretation given in the framework of the 2D edge physics code
TECXY [A1.18] support this view. Associated with the low particle recycling, enhanced performance
operations, near or beyond the Greenwald limit, are easily obtained after lithization in the explored plasma
current ranges (I p =0.5-0.9 MA), with no MHD activity. For I p =0.5 MA, B T =6T, the density limit
(n _ e =2.7×1020 m -3 ) is 1.7 times higher than after a fresh boronization and a factor of 1.4 higher than the
[A1.6] V. Pericoli–Ridolfini et al., Nucl. Fusion 34, 469-481 (1994)
[A1.7] V. Pericoli–Ridolfini et al., Plasma Phys. Control. Fusion 39, 1115-1128 (1997)
[A1.8] P.L. Andrews and F. Perkins, Phys. Fluids 26, 2537-2545 (1983)
[A1.9] V. Pericoli–Ridolfini et al., Nucl. Fusion 38, 12, 1745-1755 (1998)
[A1.10] G. Calabrò et al., Proc. 33 rd EPS Conference on Plasma Physics (Rome 2006), on line at: http://epsppd.epfl.ch/Roma/pdf/P5_077.pdf
[A1.11] R. Cesario et al., Nucl. Fusion 46, 462-476 (2006)
[A1.12] V.A. Evtikhin et al., Fusion Eng. Des. 56-57, 363-367 (2001)
[A1.13] A. Vertkov et al., Technological aspects of liquid lithium limiter experiment on FTU tokamak, presented at the 24 th Symp. on Fusion
Technology - SOFT, (Warsaw 2006)
[A1.14] A.G. Alekseyev et al., Proc. 33 rd EPS Conference on Plasma Physics (Rome 2006), on line at:
http://epsppd.epfl.ch/Roma/pdf/P1_162.pdf
[A1.15] M.L. Apicella et al., First experiments with lithium limiter on FTU, presented at the 17 th Inter. Conference on Plasma Surface Interactions
- PSI, (Hefei 2006), to appear in J. Nucl. Mater.
[A1.16] G. Mazzitelli et al., Proc. 21 st Fusion Energy Conference (Chengdu 2006), on line at: http://wwwnaweb.iaea.org/napc/physics/FEC/FEC2006/papers/ex_p4-16.pdf
[A1.17] V. Pericoli–Ridolfini et al., Modification of the SOL properties with the liquid lithium limiter in FTU - experiment and transport modeling,
presented at the IEA Large Tokamak IA Workshop on Edge Transport in Fusion Plasmas - ETFP (Kraków 2006), to be published in
Plasma Phys. Control. Fusion
[A1.18] R. Zagórski and H. Gerhauser, Physica Scripta 70, 2/3, 173 (2004)
References
11
Progress Report 2006

A1 Magnetic Confinement
corresponding Greenwald limit. In this case the electron density profile reaches a very high peaking
factor n e0 /=2.2 [A1.19] ( indicates the volume average).
A Fusion Programme
With the LLL well inserted in the SOL, peculiar new regimes are observed at high density
(n _ e ≥1×1020 m -3 ) where, without particle fuelling, a spontaneous transition at higher n _ e occurs close
to the Greenwald limit, characterised by peaked density profiles n e0 /2. This phenomenology,
well described in [A1.17], is related to the high Li pumping rate that strongly depresses deuterium
and impurity recycling, thus reducing to a great extent the instabilities due to multifaceted
asymmetric radiation from the edge (Marfe).
Transport and energy balance analysis was performed with the JETTO code for plasma discharges
at I p =0.5 MA, n e =0.7×10 20 m -3 , B t =6 T after lithization, fresh boronization and with very clean
metallic walls (e.g., oxygen-free) [A1.20]. An improvement in energy confinement time τ E by a factor
of 1.3 was found for lithizated and boronized discharges compared to the metallic case, mainly due
to the strong reduction in Ohmic power produced by the lower Z eff . For boronized and lithizated
discharges τ E /τ ITER97L =1.25 was found, which is sensibly larger than the values observed for
standard metallic FTU Ohmic discharges, which range between an average value of
τ E /τ ITER97L =0.92 [A1.21] up to τ E /τ ITER97L =1.1 in the case of very good clean plasma.
During 2006, preliminary operations with the LLL in plasma-heated discharges with LH and ECRH
at power levels in the MW range were obtained without any particular problem, but careful analysis
is required to gain a full physical and technological understanding of the experimental results.
MHD real-time control experiment
An active automatic system for MHD mode location and feedback control via ECRH power is
installed on FTU. The system [A1.22, A1.23] is able to identify, in real time, mode presence/location
and the position of ECRH absorption, and to proceed to suppress the mode.
The aims of the 2006 campaign were i) to look for a more efficient m=2 mode stabilisation obtained
by modulating the ECRH source in phase with the island rotation and ii) to optimise the technique
for identifying the ECRH absorption position, by enhancing the signal-to-noise ratio.
An overall experimental time of three days was allocated to the experiment: two during the spring
campaign and one during the autumn campaign. Only a preliminary result [A1.24, A1.25] was
obtained for target i) because of the difficulties found in plasma target production (an m=2 mode
with rotation frequency less than 3 kHz): an MHD mode together with availability of the active ECRH
system occurred only in a few shots, which were used to optimise the experimental setup. Various
induced MHD production schemes and a Mo laser blow-off technique were tested. The
achievement of a reliable target is still an open question in the experimental programme, and
dedicated experiments should be planned. Further investigation is needed in order to close the
control loop and complete the experiment.
Regarding target ii), new modulation schemes were developed, using non-periodic ECRH
modulation, to get a more enhanced signal-to-noise ratio and a better picking factor than with the
usual fixed frequency modulation scheme. A partial power scan was performed, but a lower
detectable power limit has still to be found [A1.26].
Electron cyclotron current drive experiment
The aim of the experiment was to explore at full EC power (1.5 MW) the capability of electron
cyclotron current drive (ECCD) to modify the plasma current profile. Modifications would allow
control of plasma core confinement and MHD instabilities (e.g., sawteeth) at ITER-relevant plasma
density n (n=0.6–0.7×10 20 m -3 ) and magnetic field (B T =4.6–5.1 T).
Progress Report 2006
12

The immediate goal was to fix the minor radius range where local re-shaping of the plasma current density
and safety factor q profiles can be modified by driving, with oblique injection of EC waves, well localised
ECCD in co-/counter- directions (co, to reduce q, counter, to increase it). In the previous 2005 campaigns
the range of minor radius was explored up to r/a=0.3, using 75% of the available EC power (1.1 MW).
Significant non-inductive current for plasma current density re-shaping was obtained (6-7% I p ), and
sawtooth stabilisation effects by local tailoring were observed by driving counter-current on-axis (±10°) in
target plasmas with I p =360 kA, =0.75×10 20 m -3 and T e =5 keV [A1.27].
In June 2006 one day of ECCD experiments (eight successful shots), with the same plasma conditions as
in 2005, allowed better investigation of the radial range, stabilisation of sawteeth also at ±20° oblique EC
injection, still using 75% of EC power. The goal to control the plasma current density in the plasma core,
using two oblique injection angles (±10°–±20°), was achieved even though the ECCD was low (

A1 Magnetic Confinement
Fig. A1.7 – P ECRH deposition scan: safety factor (q) values obtained from island viewed
through soft-x-ray tomography (except for q=1 determined from sawtooth inversion radius)
A Fusion Programme
t dis -t MHD (ms)
l/l p (%)
80
40
# 29984
q=1
q=3/2 q=2
40
Another important issue is the effect of LH
power on disruptions. The formation of large
0
runaway electron currents (fig. A1.8) has been
0 40 80 120 160 found to occur more often in FTU in discharges
(dl that disrupt during LH injection. Contrary to the
p /dt) max (MA/s)
theoretical expectations for electron thermal
Fig. A1.9 – Runaway current fraction vs maximum runaway generation (based on the usual
plasma current derivative during current quench for LH Dreicer and avalanche mechanisms), the
runaway plateau disruptions. Ohmic runaway plateau largest runaway currents correspond to the
disruptions included for comparison
slowest plasma current decay rates (fig. A1.9).
This trend is opposite to what is observed in
most tokamaks. Such anomalous behaviour is
attributed to pre-existent wave-resonant suprathermal electrons being accelerated during the
disruption decay phase [A1.29]. These results could be relevant for the operation of ITER whenever
a sizeable amount of LH power is used.
Dusty plasmas
# 29979 & # 29963
q=3
disruption
avoidance
0
0 10 20
r dep (cm)
80
350 kA;T e
=80 eV
LH -500 kA
OH-500 kA
350 kA;T e
=44 eV
LH - 350 kA
OH- 300 - 400 kA
500 kA;T e
=42 eV
Research on the problem of dust in tokamak plasmas is carried out in the framework of a
collaboration with the universities of Naples and Molise and the Max Planck Institute for
Extraterrestrial Studies. Interest in this subject is increasing due its relevance for fusion reactors in
terms of safety and operation [A1.30].
Preliminary theoretical studies were dedicated to analysing, in un-magnetised dusty plasmas,
fundamental dust interactions and fluctuations. In the framework of linear, fluid theory, it was shown
that over-screening and attraction between negatively charged dust particles can occur if cations
are released by the dust surface [A1.31]. Problems associated with a full kinetic model of such dust
interaction were discussed and solved in principle, although analytical calculations still have to be
completed [A1.32]. The kinetic theory of fluctuations was used to describe changes in the spectral
densities of plasma fluctuations in un-magnetised plasmas in the presence of dust [A1.33].
n/s I p (MA)
80
0.4
I p
19989
10 11 BF 3
a)
0.2
0
V I
40
0
10 13
b)
NE213
1.00 1.02 1.04 1.06 1.08
Time (s)
V I (V)
Fig. A1.8 – Plasma disruption showing the formation of a
0.3–MA runaway current: a) plasma current I p (solid) and
loop voltage V loop (dashed); b) neutron rate: BF 3 (solid)
and NE213 (dashed) signals. The NE213 line is absent
during the plateau phase because of saturation
Progress Report 2006
14

Fig. A1.10 – Conditionally averaged signals with
threshold of 4 a) and 8 rms b). Number of elementary
charges collected during typical large event (inset in b)
Experimental analysis of scattered laser light signals in FTU
discharges during disruption events confirmed the presence of
dust particles [A1.34], formerly observed by this kind of
diagnostic in JIPPT-IIU [A1.35]. Laser scattering signals were
observed by the Thomson scattering (TS) system installed in
FTU. The spectral transmission of the filter of the spectral channel
used for alignment has been centred at the laser wavelength so
that it can reveal elastic light scattering, which might be due to
the presence of dust particles [A1.35]. Elastic scattering
observed in several discharges after a disruption can last more
than 1 s after the end of the discharges. Preliminary analysis of
the laser light scattering data suggests the presence, after a
disruption, of sub-micron size (

A1 Magnetic Confinement
A Fusion Programme
probes, due to the impact of μm-sized
dust at a velocity of the order of ten km/s.
This interpretation is supported directly
by electron microscope analysis of the
probe surface, which revealed the
presence of 10 to 100–μm–sized craters,
a typical footprint of the impact ionization
processes (fig. A1.11). A number of
spherically shaped, iron-rich μm-sized
particles was also observed to be
embedded in the probe surface. Neither
craters nor embedded particles were
detected on the surface of the “virgin”
probe. The size, number and distribution
of the observed craters are consistent
with impact ionization processes
Fig. A1.11 – Electron microscope analysis of the probe occurring at an average rate of a few
surface
hundred Hz, which corresponds to
10 4 m -3 density of fast μm–sized dust
particles accelerated at velocities of the order of 10 km/s by ion drag forces associated with plasma
flows in the SOL of FTU.
A1.4 Plasma Theory
Mutual and positive feedbacks between theory and experiments have led to a clear identification of
high-frequency MHD activity (high frequency with respect to that typical of MHD fluctuations) in FTU
as evidence of nonlinear Alfvén mode excitations by a large magnetic island.
Electron-fishbone mode excitations by LH additional power only are explained within a general
theoretical framework, which fully accounts for the various experimental evidence of such modes
and also provides a simple yet relevant model for interpreting the rich nonlinear dynamic behaviour,
observed experimentally.
The theory of propagation and absorption of rf waves in toroidal plasmas has been explored in both
its more basic aspects as well as with detailed applications of practical relevance, such as the
modelling of ICRH experiments in JET and the investigation of burning plasma dynamics issues by
ICRH accelerated minority ion supra-thermal tails.
The investigation of energetic ion dynamics in burning plasmas has been articulated along three
main lines: i) identification of the relevant plasma parameters that make it possible to experimentally
study burning plasma physics issues in sub-ignited regimes; ii) numerical simulation of energetic ion
transport and nonlinear Alfvénic fluctuations in situations of experimental relevance in present-day
experiments; iii) first-principle-based analysis of fundamental processes involved in the collective
excitation of Alfvénic modes and in the fluctuation enhanced energetic ion transport. Item ii) has
been explored with the interpretation of nonlinear Alfvén wave dynamics and energetic ion transport
observed in JT-60U by means of hybrid MHD-gyrokinetic numerical simulations, carried out within
an ongoing collaboration with the Japan Atomic Energy Agency. For item iii) an overview of the
theory of Alfvén waves and energetic particle physics in burning plasmas has been given as a result
of work done within the continuing collaboration with University of California at Irvine (UCI) USA.
Within the same framework of UCI collaboration, recent theoretical work on plasma turbulence and
turbulent transport, or more specifically, on nonlinear equilibria, stability and generation of zonal
structures in toroidal plasmas has been summarised.
Progress Report 2006
16

Theory of beta-induced Alfvén-eigenmodes
Beta-induced Alfvén eigenmodes (BAEs) have frequency located in the low-frequency beta-induced gap in
the shear-Alfvén continuous spectrum, which is caused by finite plasma compressibility [A1.38-A1.40].
Their excitation can be due to the presence of fast ions and/or sharp thermal ion temperature gradients.
However recent observations in FTU [A1.41, A1.42], TEXTOR [A1.43] and JET have revealed the presence
of modes whose frequency is consistent with that of BAEs, coexisting with a large magnetic island, in the
absence of fast ions and without direct thermal ion heating. In this framework, the magnetic island appears
to play a causal role in the excitation of modes at BAE frequencies.
A kinetic stability analysis [A1.40, A1.44] of BAEs, including the effects of finite Larmor radius, finite orbit
width and toroidicity, led to a dispersion relation for BAE modes, obtained by asymptotically matching the
kinetic layer solution with that of the ideal MHD region. The resulting frequencies compare very well with
those seen experimentally in FTU [A1.45], so it can concluded that the modes observed are BAEs.
Moreover, their calculated growth rates (the experimental ones cannot be measured) are negative but small
in absolute value compared to their frequencies, so it can be inferred that such modes are marginally stable
and become nonlinearly excited above a critical amplitude threshold of the magnetic island. Analysis of
BAE destabilisation by a finite amplitude magnetic island is in progress.
Electron fishbones: theory and experimental evidence
The work described here was done in collaboration with UCI and the South-western Institute of Physics,
Chengdu P.R.C. Fishbone-like internal kink instabilities driven by electrons in conjunction with ECRH on the
high-field side were observed for the first time on DIII-D [A1.46]. The excitation was attributed to barely
trapped supra-thermal electrons, which are characterised by drift-reversal and can destabilise a mode
propagating in the ion diamagnetic direction in the presence of an inverted spatial gradient of the suprathermal
tail. Similar but higher frequency modes were observed in Compass-D [A1.47] during ECRH and
LH power injection, with chirping frequency comparable to that of the toroidal Alfvén eigenmode (TAE),
[A1.48] ω≤ω TAE . Observations of electron fishbones with ECRH only [A1.49, A1.50] and LH only [A1.51,
A1.52] have also been reported in HL-1M and FTU, respectively.
[A1.38] W.W. Heidbrink et al., Phys. Rev. Lett. 71, 855 (1993)
[A1.39] A.D. Turnbull et al., Phys. Fluids B5, 2546 (1993)
[A1.40] F. Zonca, L. Chen and R.A. Santoro, Plasma Phys. Control. Fusion 38, 2011 (1996)
[A1.41] P. Buratti et al., Nucl. Fusion 45, 1446 (2005)
[A1.42] P. Buratti et al., Proc. 32 nd EPS Conference on Plasma Physics (Tarragona 2005), on line at:http: //epsppd.epfl.ch/Tarragona/pdf/
P5_055.pdf
[A1.43] O. Zimmermann et al., Proc. 32 nd EPS Conference on Plasma Physics (Tarragona 2005), on line at:http: //epsppd.epfl.ch/Tarragona/pdf/
P4_059.pdf
[A1.44] F. Zonca et al., Plasma Phys. Control. Fusion 40, 2009 (1998)
[A1.45] S.V. Annibaldi, F. Zonca and P. Buratti, Proc. 33 rd EPS Conference on Plasma Physics (Rome 2006), on line at:http:
//epsppd.epfl.ch/Roma/pdf/ O2_016.pdf, and to appear on Plasma Phys. Control. Fusion
[A1.46] K.L. Wong et al., Phys. Rev. Lett. 85, 996 (2000)
[A1.47] M. Valovic et al., Nucl. Fusion 40, 1569 (2000)
[A1.48] C.Z. Cheng, L. Chen and M.S. Chance, Ann. Phys. 161, 21 (1985)
[A1.49] X.T. Ding et al., Nucl. Fusion 42, 491 (2002)
[A1.50] J. Li et al., Proc. 19 th IAEA Fusion Energy Conference (Lyon 2002), on line at: http://wwwpub.iaea.org/MTCD/publications/PDF/csp_019c/pdf/OV_5-1.pdf
[A1.51] P. Smeulders et al., Proc. of the 29 th EPS Conference on Plasma Physics and Controlled Fusion (Montreaux 2002), on line at:
http://epsppd.epfl.ch/Montreux/pdf/D5_016.pdf
[A1.52] F. Romanelli et al., Proc. 19 th IAEA Fusion Energy Conf. (Lyon 2002), on line at: http://wwwpub.iaea.org/MTCD/publications/PDF/csp_019c/pdf/OV_4-5.pdf
References
17
Progress Report 2006

A1 Magnetic Confinement
A Fusion Programme
keV MW 10 20 m -3 keV
8
1)
6
4
T
2 e0
n 2)
0.6 e,line
0.5
0.4
1.5 P LH
3)
1.0
0.5
0
0
-0.5
1
-1
st branch 2 nd branch ECE Ch 9 4)
0.20 0.25 0.30 0.35 0.40
Time (s)
Fig. A1.12 – Time evolution of thermal electron temperature 1),
electron density 2), LH power input 3) and (fast) electron
temperature fluctuation 4) in FTU shot #20865. It is clear that
the nonlinear behaviour of electron temperature fluctuations
(electron fishbone) reflects the level of LH power input
The peculiar features of electron fishbones were
analysed vs those of the well-known ion fishbone
[A1.53-A1.55]. Due to the frequency gap in the lowfrequency
shear Alfvén continuum for modes
propagating in the ion diamagnetic direction
[A1.55], effective electron fishbone excitation
favours conditions characterised by supra-thermal electron drift reversal, which is consistent with
experimental observations. For the same reason, the spatial gradient inversion of the supra-thermal
electron tail is necessary, explaining why ECRH excitation is observed with high-field side deposition
only [A1.46, A1.49, A1.50, A1.56]. Circulating supra-thermal electrons play a peculiar role in electron
fishbone excitations with LH only: the barely circulating population directly provides the mode drive
and the well circulating particles controls the drift-reversal condition. As in the case of ion fishbones,
two branches of the electron fishbone have been shown to exist: a discrete gap mode [A1.55] and
a continuum resonant mode [A1.54]. Contrary to the gap mode, the continuum resonant mode can
propagate in the electron diamagnetic direction as well. Thus, it does not require either drift-reversal
or inverted spatial gradient of the supra-thermal electron tail. However, its threshold condition is
higher and it requires high power densities to be excited. So, even the case of the continuum
resonant fishbone mode tends to favour the branch propagating in the ion diamagnetic direction,
which minimises continuum damping. If the effective temperature of the supra-thermal electron tail
is sufficiently high, the present theory predicts that fishbone oscillations can be excited at
frequencies comparable with those typical of the geodesic acoustic mode (GAM) [A1.57] or the BAE
[A1.38, A1.39]. Unlike the case of fishbone gap modes in the ion diamagnetic gap [A1.55] of the
low-frequency shear Alfvén continuum, fishbone gap modes in the BAE gap [A1.58] do not favour
propagation in the ion diamagnetic direction, since the gap structure is nearly symmetric in
frequency [A1.40]. One single general fishbone-like dispersion relation [A1.59] has been discussed,
describing mode excitation by trapped as well as circulating supra-thermal electrons in both
monotonic and reversed magnetic shear equilibria [A1.60].
The most interesting feature of electron fishbones is their relevance to burning plasmas. In fact,
unlike fast ions in present-day experiments, fast electrons are characterised by small orbits that do
not introduce additional complications in the physics due to nonlocal behaviour, similarly to alpha
particles in reactor-relevant conditions. Meanwhile, the bounce averaged dynamics of both trapped
as well as barely circulating electrons depends on energy (not mass); hence their effect on lowfrequency
MHD modes can be used to simulate/analyse the analogous effect of charged fusion
products. Furthermore, the combined use of ECRH and LH provides extremely flexible tools to
investigate diverse nonlinear behaviour, for which FTU experimental results provide a nice and clear
example (fig. A1.12). During high-power LH injection, an evident transition in the electron fishbone
signature takes place from almost steady-state nonlinear oscillations (fixed point) to regular bursty
behaviour (limit cycle). A simple yet relevant nonlinear dynamic model has been derived for
predicting and interpreting these observations [A1.61].
Analysis and modelling of LHW propagation in toroidal plasmas by asymptotic
methods
The LH full wave equation in the electrostatic approximation and in general magnetic field equilibria
has beenen critically analysed by applying asymptotic techniques when looking for the solution
(Wenzel, Kramer, Brillouin [WKB] approximation). The phase and the amplitude were obtained
numerically and analytically, and then compared [A1.62].
Progress Report 2006
18

Lower hybrid wave (LHW) propagation in a tokamak plasma 2D geometry can be correctly described only
with a full wave approach based on full numerical techniques or on a semi-analytical approach, by reducing
the wave equation into two nested equations of the first order, as shown in [A1.63]. To test and compare
the full numerical solution with that obtained by applying the WKB asymptotic expansion, a rigorous WKB
solution of the wave equation for the first two orders of the expansion parameter was presented, obtaining,
at the first order, the equation for phase and, at the next order, the equation for the field amplitude. The
nonlinear partial differential equation (PDE) for the phase was solved in a pseudo-toroidal geometry (circular
and concentric magnetic surfaces) by the method of characteristics. The associated system of ordinary
differential equations (ODEs) for the position and the wave-number was obtained and analytically solved
by choosing an appropriate expansion parameter. The quasi-linear PDE for the WKB amplitude was also
analytically solved, allowing reconstruction of the wave electric field inside the plasma. The solution was
also obtained numerically and compared with the analytical solution. Further developments, consisting in
generalising the solution to a Solov’ev analytical equilibrium geometry, are in progress. The validity of the
WKB approximation was analysed on the basis of the results obtained.
Modelling of the ICRH experiment on JET
The aims of this modelling study are first to evaluate the main features of the proposed ICRH heating
experiment and second to perform a detailed analysis of the experimental discharges. The proposed
experiment concerns essentially the possibility of obtaining internal transport barriers (ITBs) on both the ion
and the electrons species with only the use of the ICRH system in an ion-heating scheme, without neutral
beam injection (NBI) as an external momentum input. In this context an ITB regime on JET was obtained
by using 6 MW of ICRH in the minority heating scheme [A1.64].
The minority species involved is 3 He. The scheme should act at the fundamental cyclotron harmonic of the
minority species (ω=Ω cm ) located near the plasma centre, while the fundamental or the first harmonic of
the majority (ω=Ω cM or ω=2Ω cM ) is out of the plasma. This is the so-called isolated case. Cyclotron
resonance heating of the minority is very efficient because fast wave polarization is essentially determined
by the majority species alone, while damping is due essentially to the resonant minority ions (minority
heating regime). If the minority concentration increases too much, the screening due to the rotating electric
field is no longer negligible, and cyclotron damping decreases drastically, entering the “mode conversion
regime”.
When programming an ICRH heating experiment, it is important to establish the plasma and antenna
parameters that fit the goals well. In the ICRH minority heating experiment on JET, antenna and plasma
parameters are chosen by maximising the power coupled to the plasma, without dealing with edge
[A1.53] K. McGuire et al., Phys. Rev. Lett. 50, 891 (1983)
[A1.54] L. Chen, R.B. White and M.N. Rosenbluth, Phys. Rev. Lett. 52, 1122 (1984)
[A1.55] B. Coppi and F. Porcelli, Phys. Rev. Lett. 57, 2272 (1986)
[A1.56] Z.-T. Wang et al., Chin. Phys. Lett. 23, 158 (2006)
[A1.57] N. Winsor, J.L. Johnson and J.M. Dawson, Phys. Fluids 11, 2448 (1968)
[A1.58] M.S. Chu et al., Phys. Fluids B4, 3713 (1992)
[A1.59] F. Zonca and L. Chen, Plasma Phys. Control. Fusion 48, 537 (2006)
[A1.60] R.J. Hastie et al., Phys. Fluids 30, 1756 (1987)
[A1.61] F. Zonca et al., Electron fishbones: theory and experimental evidence, submitted to Nucl. Fusion
[A1.62] A. Cardinali, L. Morini and F. Zonca, Proc. of the Joint Varenna-Lausanne International Workshop on Theory of Fusion Plasmas, ed. by
J. Connor, O. Sauter, E. Sindoni (American Institute of Physics, Varenna), Vol. 871, 292 (2006)
[A1.63] A. Cardinali and F. Zonca, Phys. Plasmas 10, 4199 (2003)
[A1.64] F. Crisanti et al., Experimental evidence of ion internal transport barrier without injection of external momentum input, presented at the
Transport Task Force Meeting (Varenna 2004)
References
19
Progress Report 2006

A1 Magnetic Confinement
cut–offs in the low field side, and by choosing the right minority concentration in order to avoid the
mode conversion regime [A1.65].
A Fusion Programme
The following codes have been used to plan and to model the ICRH experiment:
1) A code that solves the cold plasma electromagnetic (em) dispersion relation in slab geometry in
order to clarify the dispersion characteristics of the experiment. Thus, the range of variation in
the main parameters can be established, e.g., power spectrum vs plasma density profiles to
assess the accessibility conditions; minority concentration to assess, in a plasma with two ion
species (or more), the localisation of the ion-ion resonance (and the associated cut-off), which
turns out to be very close to the ion cyclotron resonance of the minority species, etc.
2) A code that solves the warm plasma em dispersion relation in the complex space of the wavenumber
in order to clarify the effect of wave damping on the minority species, the effects of the
plasma parameters (minority concentration, ion and electron temperature, parallel wave-number)
on the transition to the mode conversion regime. The use of this code should also provide the
power deposition profiles and the power damping rate for the entire launched spectrum.
3) A 1D ray-tracing code in cylindrical geometry to take into account, at the lowest order, the
geometry of the tokamak plasma. The code uses the warm plasma em dispersion relation of
point 2).
4) When needed, a complex 2D ray-tracing code in tokamak geometry to take into account the
realistic geometry of the tokamak. This code is based on a complex full em dispersion relation
and complex integration of the trajectories.
5) A 1D full wave code (FELICE), which gives the linear distribution of the wave power on ion and
electron species. It accounts correctly for the electron Landau damping (ELD) in the fast wave
branch and in the ion Bernstein wave (IBW) branch; it also accounts for the realistic antennaplasma
coupling and calculates the whole effect of the power spectrum on the various species.
6) A 2D full wave code (TORIC) [A1.66], which has the same characteristics as the 1D FELICE
code, but includes the real geometry of the plasma (in the flux surface coordinate system).
7) A 2D full wave code (steady-state quasi-linear Fokker-Planck [SSQLFP] code), which does the
same as before but includes the evolution of the 2D distribution function for the ions and
electrons.
Simulation of burning plasma dynamics by ICRH accelerated minority ions
The main difference between present experiments and ITER will be the presence, as the main
heating source, of alpha-particles produced in DT reactions. Alpha particles will mainly heat
electrons, contrary to present experiments dominated by low-energy neutral beam injection that
mainly heats the ions. Moreover, alpha-particles can drive stronger collective modes.
As proposed in [A1.67], alpha-particle dynamics can be simulated in pure deuterium plasmas by
ions accelerated by rf waves. The use of ICRH in the minority scheme (H or 3 He) can indeed
produce fast particles (although with a different distribution function to that of fusion-generated
alpha-particles) which, with an appropriate choice of the minority concentration, rf power and
plasma density and temperature, can reproduce the dimensionless parameters ρ * fast and β fast
characterising the alpha-particles in ITER. Here, ρ * fast is the normalised fast-particle radius and β fast
the fast-particle beta. Thus, a device operating with deuterium plasmas in a dimensionless
parameter range as close as possible to that of ITER and equipped with ICRH as the main heating
scheme would allow investigation of some of the most important features of alpha-particle heated
plasmas and, therefore, it would be possible to assess these issues in relevant scenarios before their
implementation on ITER itself.
As an example, the following reference antenna and plasma parameters were considered: 24 MW
of ICRH power coupled to the plasma at a frequency f=81 MHz; toroidal magnetic field B T =8T;
volume average density =4×10 20 m -3 with generalised parabolic profile (1-(r 2 /a 2 )) α and 3 He
Progress Report 2006
20

Fig. A1.13 – Power deposition profiles vs plasma radius for
the various species
minority-heating scheme. Two scenarios were taken into account:
one characterised by the enhancement factor H=1.3, consistent with
an ITB and peaked profiles (α n =1 α T =1), which corresponds to
having β N =1.8% and on-axis values of density, temperature and beta
given by, respectively, n e0 =6×10 20 m -3 , T e0 =T i0 =12 keV; the other
characterised by an enhancement factor H=1, in H-mode, and flat
profiles α n =0 α T =1, with β N =1.4%, and n e0 =4×10 20 m -3 ,
T e0 =T i0 =10 keV. A parametric study of ICRH absorption was
performed, varying the resonant layer, coupled wave spectrum,
minority concentration, density and temperature, with the aim of
increasing the power coupled to the minority ions and obtaining the
maximum effective temperature of the tail. As an example, the results
obtained in the “enhanced H-mode scenario” are reported here. In
figure A1.13, the power density coupled to the various species in
(W/cm 3 ) is plotted vs the plasma radius, when an optimum 3 He
minority concentration of 2% (which maximises the power absorbed
by the minority) is considered. From the figure it is possible to get the
localisation of the deposition, r/a=0.1 the width of the deposition
layer, Δr/a=0.2 for the minority and broader for the electrons, and the
peak of the power density (45 Watt/cm 3 ).
The quasi-linear analysis, based on the linear results shown above,
allows calculation of the effective temperature of the minority ions as
well as the fraction of the minority at those energies. The effective
temperature was calculated to be ≈150 keV (on the peak of the
absorption layer), with a fast ion fraction of about 30% leading to a
fast ion β fast of about 0.8%. Figure A1.14 shows the effective
temperature of the ion minority in parallel and perpendicular
directions as a function of the plasma radius.
Particle simulation of bursting Alfvén
modes in JT–60U
A numerical investigation, based on particle-in-cell
simulations, of the bursting-mode phenomenology
observed in negative neutral beam (NNB)-heated
JT–60U discharges was performed [A1.68– A1.70]. It
was shown that the experimental observations can be
interpreted as the effect of nonlinear interaction
between Alfvén modes and the energetic ions
produced by NNB injection. In particular, the
b 3
keV
60
40
20
0
150
100
50
0
Total ICRH power density
Power to ion
Power to electrons
0 0.2 0.4 0.6 0.8 1
r/a
T eff, par
T eff, perp
0 0.2 0.4 0.6 0.8 1
r/a
Fig. A1.14 – Effective temperature of the
minority vs plasma radius
n H /n H0
1.0
0.8
0.6
0.4
0.2
Relaxed
Initial
After ALE
0.0
0.0 0.2 0.4 0.6 0.8 1.0
r/a
Fig. A1.15 – Nonlinear modifications of the energetic ion density
profile produced by EPM saturation. Blue curve: initial (simulation
and experimental) profile. Red curve: relaxed profile obtained in the
simulation. Black curve: experimentally inferred profile just after the
ALE occurrence, plotted here for comparison
[A1.65] A. Cardinali et al., Proc. 33 rd EPS Conference on Plasma Physics (Rome 2006), on line at:http://epsppd.epfl.ch/Roma/pdf/P1_065.pdf
[A1.66] M. Brambilla, Plasma Phys. Control. Fusion 41, 1 (1999)
[A1.67] F. Romanelli et al., Fusion Sci. Technol. 45, 483 (2004)
[A1.68] G. Vlad et al., Proc. of the Joint Varenna-Lausanne International Workshop on Theory of Fusion Plasmas, ed. by J. Connor, O. Sauter,
E. Sindoni (American Institute of Physics, Varenna), Vol. 871, 250-263 (2006)
[A1.69] G. Vlad et al., Proc. 21 st IAEA Fusion Energy Conference (Chengdu 2006), on line at: http://wwwnaweb.iaea.org/napc/physics/FEC/FEC2006/papers/TH_P6-4.pdf
[A1.70] S. Briguglio et al., Particle simulation of bursting Alfvén modes in JT-60U, accepted for publication in Phys. Plasmas
References
21
Progress Report 2006

A1 Magnetic Confinement
3
a)
3
b)
3
c)
2
2
2
A Fusion Programme
α
1
0
0 0.25 0.5 0.75 1
Ê
α
1
0
0 0.25 0.5 0.75 1
Ê
0
0 0.25 0.5 0.75 1
Ê
Fig. A1.16 – a) Variation in the energetic-ion distribution function in the (Ê,α) plane (Ê being the energy, α the pitchangle
of the energetic ions), after saturation of the EPM, averaged on the outer plasma region, where the TAE-like
mode grows. b) Volume average of the power transfer from particles to the wave (i.e., the resonance pattern), in the
same region, during the linear simulation stage. c) Power transfer after EPM saturation. Red corresponds to positive
values; violet to negative. The EPM saturation causes an increase in the energetic-ion distribution function at low
energy and large pitch-angle. It can be shown that the increase is due to an outward radial displacement from the
central (EPM) region. The displaced ions resonate with the outer mode, modifying the outer resonance
^ ^
F-F SD
0.004
0.003
0.002
0.001
0
-0.001
-0.002
-0.003
-0.004
3
2.5
2
1.5
α
1
0.8
0.6
0.4
0.2
E^
0.5 0
Fig. A1.17 – Distortion with respect to the slowing
down distribution function (F^-F^SD) of the energetic-ion
distribution function in the (Ê,α) plane produced by
EPM saturation
1
investigation, related to modes with toroidal
number n=1, showed that an energetic particle
mode (EPM) localised around the maximum of
the energetic-ion pressure gradient is driven
unstable by resonant interaction with such ions.
Its saturation produces radial displacement of
energetic ions, in fair agreement with the
experimental findings related to the so-called
abrupt-large-amplitude events (ALEs)
(fig. A1.15).
Simulation results demonstrate that displaced
ions resonate with Alfvén modes in the outer
region (fig. A1.16), causing a TAE-like mode to
become dominant as the saturation of the EPM
proceeds. It has also been observed that the
scattering due to the EPM is more effective on
resonant ions than non-resonant. Besides the
relaxation of the density profile, a distortion in
the velocity-space distribution function is then
produced (fig. A1.17). This fact can explain why
a quieter phase, characterised by weaker bursting modes (the fast frequency sweeping), is
observed after an ALE, allowing the system to restore the free energy needed for a new ALE. In the
absence of velocity-space distortion, any reconstruction of the density profile would indeed generate
relatively large amplitude modes: energetic ions would be further scattered by these modes and
their density profile would be essentially clamped to the relaxed profile produced by the ALE.
Once the phase-space distortion is fully taken into account, the free-energy reconstruction rate is
instead set by the need to rebuild both the density profile and the resonant part of the distribution
function. The slow time scale evolution of energetic ion equilibria in intermediate configurations
between two successive ALEs is then characterised by a lower drive than that corresponding to the
unperturbed velocity-space distribution function, and the weak modes excited are less effective in
contrasting the density profile reconstruction. Only when the combined restoration of the
configuration and velocity space distributions provides enough drive for a fast growing Alfvén mode,
does a new ALE occur.
α
1
Progress Report 2006
22

Theory of Alfvén waves and energetic particle physics in burning plasmas
An overview of the work presented here has been given in [A1.71]. A unique characteristic of burning
plasmas is that the energy density of fast ions (MeV energies) and charged fusion products is a significant
fraction of the total plasma energy density. Consequently, one can address two major issues of practical
concern in such plasmas: i) whether fast ions and charged fusion products are sufficiently well confined to
transfer their energy and/or momentum to the thermal plasma without appreciable degradation due to
collective modes; and ii) whether, on longer time scales, mutual interactions between collective modes and
energetic ion dynamics on the one hand and drift wave turbulence and turbulent transport on the other
may decrease the overall thermonuclear efficiency of the considered system.
The first issue was addressed by analysing theoretically the dynamics of shear Alfvén waves collectively
excited by energetic particles in tokamak plasmas. Both linear physics, such as spectral and stability
properties, as well as key nonlinear wave and particle dynamics have been identified and considered. The
investigations of such processes via computer simulations have also been discussed along with the
importance of benchmarking with existing or future experimental observations.
In terms of consequences, the two issues have different practical implications: the first has a direct impact
on the operation scenarios and boundaries, since energy and momentum fluxes due to collective losses
may lead to significant wall loading and damage to plasma-facing materials; the second poses soft limits
in the operation space.
In the framework of plasma theory, the first issue is connected with identification of burning plasma stability
boundaries with respect to collective mode excitations by fast ions and charged fusion products as well
as with nonlinear dynamics above the stability thresholds; the second is associated with long time-scale
nonlinear behaviour typical of self-organised complex systems.
Nonlinear equilibria, stability and generation of zonal structures in toroidal plasmas
The crucial role played by zonal flows [A1.72] in regulating the saturation level of drift wave turbulence and
ultimately of turbulent transport [A1.73] has led to significant attention being paid to determining the
quantity of zonal flows (ZFs) that can be spontaneously generated by the turbulence itself before the flows
become unstable, also due to Kelvin Helmholtz (KH)–like mode excitations [A1.74-A1.76]. In this
framework, drift waves (DWs) are the “primary” instability and spontaneously generate ZFs, the
“secondary” instability, which can be limited in amplitude by the onset of “tertiary” KH–like modes [A1.74-
A1.76]. The “tertiary” instability has been proposed to explain the nonlinear up-shift of the critical ion
temperature gradient (ITG) driven turbulence threshold [A1.77].
It has been proposed that long-lived saturated ZF structures, spontaneously generated by DW turbulence,
can be considered as generators of neighbouring nonlinear equilibria [A1.78]. In the present theoretical
framework, the general form of these neighbouring nonlinear equilibria has been computed in terms of
[A1.71] L. Chen and F. Zonca, Proc. 21 st IAEA Fusion Energy Conference (Chengdu 2006), on line at: http://wwwnaweb.iaea.org/napc/physics/FEC/FEC2006/papers/OV_5-3.pdf,
and submitted to Nuclear Fusion
[A1.72] A. Hasegawa et al., Phys. Fluids 22, 2122 (1979)
[A1.73] Z. Lin et al., Science 281, 1835 (1998)
[A1.74] B.N. Rogers et al., Phys. Rev. Lett. 85, 5336 (2000)
[A1.75] F.L. Hinton and M.N. Rosenbluth, Bull. Am. Phys. Soc. 45,7, 195 (2000)
[A1.76] E.-J. Kim and P.H. Diamond, Phys. Plasmas 9, 4530 (2002)
[A1.77] A.M. Dimits et al., Phys. Plasmas 7, 969 (2000)
[A1.78] L. Chen and F. Zonca, Proc. 21 st IAEA Fusion Energy Conference (Chengdu 2006), on line at: http://wwwnaweb.iaea.org/napc/physics/FEC/FEC2006/papers/TH_P2-1.pdf,
and submitted to Nucl. Fusion
References
23
Progress Report 2006

A1 Magnetic Confinement
A Fusion Programme
zonal structures as well as of the characteristics of the primary DW turbulence. The derived nonlinear
evolution equation for the zonal response consistently describes the temporal evolution of the zonal
structures, whose time-asymptotic behaviour corresponds to nonlinear equilibria. The stability of the
nonlinear equilibria determines the nature of the “tertiary” instability regime, the nonlinear up-shift of
critical thresholds, and the collisionless dissipation of the zonal structures. On a shorter time-scale,
the temporal evolution of the zonal response describes the DW-ZF generation and the regulation of
the DW intensity by the ZFs.
While the stability properties of the nonlinear zonal equilibria have been given in terms of integral
eigenmode equations [A1.75], simple estimates for the threshold condition for tertiary instability can
be derived in the local limit. It has also been discussed how this instability condition can be
translated into an estimate of the nonlinear up-shift of the critical threshold for the ITG turbulence
driven transport, known as the “Dimits-shift” [A1.77]. In fact, employing the time asymptotic
response of the zonal structures as the nonlinear equilibria allows one to directly connect the starting
reference equilibrium quantities to the nonlinear equilibrium features due to finite ZF amplitude as,
e.g., radial modulations in the temperature profile [A1.79]. It has been shown that tertiary instability
consists of trapped ion ITG modes (TITG), generated by these radial modulations of the ion
temperature profile. Employing the quasi-linear description, it is has been further demonstrated that
tertiary TITG turbulence [A1.78] can lead to collisionless dissipations of the zonal structures, i.e.,
their quasi-linear relaxations. In this respect, the existence of TITG turbulence can then lead to the
resurgence of the primary DW turbulence and of turbulent transport, which is strongly suppressed
by ZFs for reference plasma equilibrium gradients below the Dimits-shift.
A1.5 JET Collaboration
The JET machine operated during 2006 after the long shut-down for the new diagnostic systems,
new divertor, and NBI upgrading. The new ICRH antenna was not ready for integration in the
machine. Three campaigns (C15-C16-C17) were carried out. The organisation of the experimental
programme was led by two main task forces (S1 and S2). The experiments proposed by the other
task forces (Diagnostics, Heating, Magnetics, Exhaust, DT) were incorporated in the S1 and S2
experimental programme.
With regard to the European Fusion Development Agreement (EFDA) JET 2006 work programme,
ENEA participated in the commissioning of systems included in the JET Enhancement Programme
(EP), the organisation of new enhancements for JET-EP2 (2006-2008) and the realisation of
experiments in campaigns C15-C17. ENEA has provided EFDA JET with the EFDA associated
leader for JET, two task force leaders (Transport and Diagnostics), one deputy task force leader
(Advanced Tokamak Scenario, S2), one responsible officer (RO) in the EFDA JET close support unit
(CSU) (RO for H-mode Scenario (S1) and Transport task forces) and one RO for enhancements in
the EFDA JET CSU.
Participation in the JET EP/EP2
JET neutron profile monitor: fast data acquisition system for neutron/gamma
discrimination. Digital techniques for neutron detection/spectrometry are important in view of ITER
application as they allow the realisation of systems that provide high count rates (MHz),
simultaneous counting and spectroscopic measurements of both neutrons and gamma-ray
emission (at present not technically feasible with conventional analog pulse shape discrimination)
and the possibility of post-experiment (re)processing of data with different analysis techniques and
analysis of pile-up events.
Under an EFDA task the 14-bit 200 MS/s digitizer system for fast sampling of pulses and
Progress Report 2006
24

neutron/gamma digital pulse shape discrimination (DPSD)
[A1.80] to be used with scintillators was installed on the
central channel of the KN3 neutron camera at JET and
commissioning took place in November 2006. Data from
a large number of plasma discharges were acquired. The
results demonstrate the capability of the DPSD system to
provide simultaneously 2.5- and 14-MeV neutron count
rates as well as pulse height spectra: note the detection of
ion tails in a discharge with NBI and ICRH (fig. A1.18).
Comparison of analog and digital count rates indicates
good agreement between the two systems (fig. A1.19).
Optimisation of the system hardware and software is also
in progress under an EFDA Underlying Technology task. In
particular, pile-up is currently being investigated with the
use of a set of data acquired with the DPSD system on
JET. The plan is to prepare a software module to deal with
pile-ups, and also to perform neutron/gamma separation
and pulse height analysis on such events, which will
eventually be included in the existing software package.
Along the same line of research, two further tasks have
been started in collaboration with TRINITI on neutron
detector digital electronics and radiation hardness testing.
3×10 4
CVD diamond detectors for neutron measurement.
Shot #68569
Under the Small Enhancement agreement, two diamond
detectors produced by the chemical vapour deposition
1×10 4
(CVD) technique were installed at JET and worked
continuously during the C15-C17 experimental
50 54 58 62
campaigns. One detector is polycrystalline diamond
Time (s)
(p–CVD) covered with a thin layer (2 μm) of lithium fluoride
Fig. A1.19 – Comparison between analog and digital count
(LiF) 95% enriched in Li-6. The latter converts low-energy
rates in the 2.5-MeV neutron energy range (JET discharge
neutrons into alphas and tritons of about 2 MeV and
#68569)
2.7 MeV respectively, which are easily detected by the
diamond film and hence the total neutron emission can be
measured. To further enhance its response the detector is embedded in polyethylene. The other detector
is a single crystal diamond (SCD) film with a special heavily doped boron contact covered with a layer of
enriched 6 LiF (2 μm thick) for simultaneous detection of total neutron emission and 14-MeV neutron
emission from triton burn-up. This detector was also used in a first attempt to perform neutron
spectrometry. Both detectors also measured the time-dependent neutron emission during each pulse.
The goal was a) to demonstrate the capability and reliability of diamond detectors as neutron monitors
during long–lasting experiments under ITER-like working conditions; b) to demonstrate the capability of a
single SCD detector covered with LiF to simultaneously detect and discriminate between total and 14-MeV
neutrons produced by triton burn-up. The job output will be a comparison between CVD data and those
obtained from the official JET neutron detectors (fission chambers and silicon diodes).
Figure A1.20 shows the correlation between the total neutron emission measured by the p-CVD (as well
as the SCD) and that (average value) recorded by the fission chambers (FCs) available at JET. The degree
Counts (arb. units)
10 -2
10 -4
Shot #68445
t = 46.0 s
t = 47.4 s
t = 48.8 s
t = 50.2 s
t = 51.6 s
10 -6 0 4 8 12 16
Proton energy (MeV)
Fig. A1.18 – Pulse height spectra in JET discharge
#68445: NBI and ICRH are injected at t>46 s
Counts/s
7×10 4
5×10 4
neutron analog (10 ms)
gamma analog (10 ms)
neutron digital (10 ms)
gamma digital (10 ms)
[A1.79] S.E. Parker et al., Phys. Plasmas 6, 1709 (1999)
[A1.80] M. Riva, B. Esposito and D. Marocco, Proc. 10 th Inter. Conference on Accelerator & Large Expt. Physics Control Systems - ICALEPCS
(Geneva 2005), paper P-O2.041-4, http://epaper.kek.jp/ica05/proceedings/pdf/P3_041.pdf
References
25
Progress Report 2006

A1 Magnetic Confinement
A Fusion Programme
Diamonds (counts)
8×10 4
6×10 4
4×10 4
2×10 4 0
CVD02
SCD03
5×10 15
y=2.042×10 -12 x +9.463×10 2
R 2 =9.978×10 -1
of agreement between the two detectors is excellent in the whole range of interest for JET, that is
from 5.0×10 14 n/shot up to the highest neutron yields (>4×10 16 n/shot) produced during C17.
Figure A1.21 reports the behaviour of the p-SCD detector vs JET pulses and thus as a function of
time. Also in this case the stability is proven.
ENEA’s participation in the JET-EP2 concerns further implementation of the enhancements: fast
data acquisition for the neutron camera; new calibration, with new fast electronics for the NE213
compact neutron spectrometer; a new monochrystal CVD to be tested for the neutron
spectrometry; new CVD detectors to be tested for ultraviolet radiation detection.
Participation in experimental campaigns C15-C17
ENEA has presented JET with about 20% of the proposals considered for campaigns C15-C17. The
proposals are mainly dedicated to the work carried out by TFS2 (Advanced Tokamak Scenario), TFM
(Magnetics), TFD (Diagnostics). The following is a short presentation of some preliminary results of
the 2006 experiments.
Advanced Tokamak Scenario
SCD
y=9.953×10 -13 x +1.043×10 2
R 2 =9.978×10 -1
1.5×10 16 2.5×10 16 3.5×10 16
FC (counts)
Fig. A1.20 – Correlation between p-CVD and SCD
detectors and FC
• Optimisation of hybrid advanced regime with electron heating. The activity consisted in planning
and co-leading the sessions and coordinating the diagnostics in the control room. The hybrid
regime with T e >T i was established, and subsequently i) density scan, ii) current profile scan, iii)
power scan were carried out. This experiment was done to complete the database with more
refined diagnostic coverage, in particular, the charge exchange and motional Stark effect (MSE).
Ion temperature, rotation profiles and impurity density were carefully measured for transport
analysis. The plasma parameters of the reference pulse (#62779, in C13) were magnetic field
B T =3.2 T, plasma current I p =2.3 MA, neutral beam heating power P NBI =9 MW, ion cyclotron rf
power P ICRH =9 MW, lower hybrid current drive power P LHCD =1 MW. The hybrid current profile
was obtained with LHCD in preheating. The main heating was performed with equal neutral beam
and ICRH power, resulting in peak temperatures T e =9-11 keV, and T i =7-8 keV at a density of
3×10 19 m -3 . The confinement regime was H-mode with H89 ~2, and small edge localised modes
(ELMs). The scenario is characterised by a sawtooth-free period and by frequent and very small
ELMs, after LH preheating (fig. A1.22), where the H89 ~2, and small n=2 NTMs (neoclassical
tearing modes) are detected. The spatial q profile is typical of hybrid regimes (fig. A1.23) where a
large region of flat shear (with q min ≥1) is created at the plasma centre, when the maximum β N is
reached.
• ITER-relevant ITB scenario at high β N and bootstrap fraction. The ITER non-inductive scenario
has to reach a high H-factor (H98(y,2)~1.5) and normalised beta (β N ≥3), in the presence of a
fraction of bootstrap current J bs , close to 50% of the total current. The scenario is characterised
by a non-monotonic q profile and the formation of an ITB located close to the point of minimum
value of safety factor q min . Experiments were done on JET at plasma parameters: magnetic field
Ratio
1.5
1.3
1.1
0.9
0.7
y= -2.869×10 -6 × +1.252
R 2 =2.780×10 -4
0.5
68500 68700 68900 69100
Pulses
Fig. A1.21 – Ratio between p-CVD and FC counts vs
JET pulses
Progress Report 2006
26

Fig. A1.22 – From the top. Shot #68383 First plot: heating
waveforms LHCD total power (red), ICRH (magenta), neutral
beams (blue). Second plot: β N (red) and H89 factor (blue); Third
plot: n=1 (red), n=2 (blue) MHD modes. Fourth plot:
measurements of T e in two positions. Fifth plot: D α outer divertor
B T =2.3 T, plasma current I p =1.5 MA, neutral beam
heating power P NBI =22 MW, ion cyclotron heating
power P ICRH =6.2 MW, LHCD power
P LHCD =2.2 MW, density close to the Greenwald
limit n G =I p /(π a 2 )=0.5×10 20 m -3 , T e =3-5 keV. The
safety factor at the edge was q 95 ~5, and the
triangularity δ~0.4. A q profile with negative
magnetic shear is formed using LHCD early in the
discharge, and high NBI and ICRH power are
added when the minimum q is just above 2, to
trigger an ITB as q min crosses 2. Two configurations
were used for this experiment: ITER_AT (Advanced
Tokamak) and high beta poloidal, in two separate
sessions. To maintain low ELMs, neon or D 2 gas
puffing is used, and this tool helps also to make the
scenario suitable for operation with a Be wall.
Figure A1.24 shows the main results in terms of
normalised beta: the normalised beta is plotted vs
the no-wall beta limit (defined as β N no-wall =4l i ). The
best results β N,max ~3 are obtained when both
electron and ion ITBs occur at the same radius and
strength. The value of the fusion gain vs the density
normalised to the Greenwald density is consistent
with that useful for ITER (fig. A1.25). This scenario
so far is limited by strong ELMs appearing at peak
β N (fig. A1.26). Despite the large ELMs, the ITB
persists for several confinement times, but then it is
lost due to detrimental MHD activity linked to the
presence of low magnetic shear, as the q profile
evolves from negative to positive magnetic shear.
V 10 3 eV V
10 7 W
0.8
0.4
0
1.5
0.5
0.20
0.10
0
8
6
4
2
0.4
0.2
q
0
5
4
3
2
1
46 48 50 52 54 56
Time (s)
Time = 49.94 s
Time = 51.95 s
Time = 53.96 s
#68383
2.0 2.5 3.0 3.5
R(m)
Fig. A1.23 – Evolution of the q(r) profile as measured by MSE
β N
4.0
3.5
3.0
2.5
2.0
1.5
1.0
0.5
0
0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0
4I i
Fig. A1.24 – Normalised beta β N vs the no-wall
beta limit defined as β N no-wall =4l i
H 89 β N /q 95
2
0.35
0.30
0.25
0.20
0.15
0.10
0.05
0
0 0.2 0.4 0.6 0.8 1
n e /n G
Fig. A1.25 – Figure of merit of fusion gain
H 89 β N /q 2 95 vs normalised electron density
27
Progress Report 2006

A1 Magnetic Confinement
A Fusion Programme
Fig. A1.26 – Time evolution of β N , Z max
plasma position, elongation κ, H α and
MHD (n=1) monitors
Diagnostics
• High-resolution Thomson
scattering (KE11). ENEA
Frascati has been involved in
the commissioning of the highresolution
Thomson scattering
(HRTS) system required for the
exploitation of many
experiments. The analysis
code of the HRTS data was
developed and used during
β N
Zmax (m)
κ
H α
MHD n=1
3.0
2.8
2.6
2.4
1.85
1.75
1.76
1.72
1.68
2.0
1.5
1.0
0.5
commissioning on the plasma. A preliminary project
of a system for laser alignment control using
cameras has been outlined. The aim is to monitor
the laser beam incident on the input window and on
the beam dump internal to the vacuum vessel.
2
1
0
#68927
44.5 45.0 45.5 46.0 46.5
Time (s)
#70068
Time = 47 s
• Motional Stark effect. The tools presently available
to determine the spatial q profile were studied: i)
EFIT + MSE ; ii) EFIT+polarimetry; iii) point-to-point 4
analysis. The aim of the study was to determine the
Time = 46 s
sensitivity of the q profile to the data sources and to
2
the method, and the accuracy of the subsequent
determination, in particular for hybrid discharges
2.0 2.5 3.0 3.5
(exp S2-4.3 - Optimisation of a hybrid scenario with
R(m)
electron heating) where the q(r) spatial profile in the
Fig. A1.27 – Spatial profiles of the q safety factor
plasma central region is critical. For these
discharges substantial agreement has been
detected between the results of the various methods. The MSE data analysis was critically
reviewed by comparing different constraints on the EFIT data, e.g., choice of MSE channel
weights, data from polarimetry, variation in pressure constraints and polynomial degree. Where
possible, the results were checked against proof of the existence of rational surfaces for q given
by mode analysis of fast diagnostics or the occurrence of sawteeth. As an example, figure A1.27
shows the q profiles in the proximity of a 3/2 mode reconnection, with the final value of q in the
region interested by the mode close to 1.5. An alternative method for processing MSE data,
previously developed at JET [A1.81], has been retrieved and tested on several discharges against
the EFIT equilibrium. According to some hypotheses on the shape of internal flux surfaces, once
the last closed surface is known, it is possible to process individual MSE data points, obtaining
independent determinations of the radial q points. This minimises the effect on the overall profile
of the errors on individual channels (some of which are occasionally affected by spurious radiation
which makes them completely unreliable). More work is being done to extend the comparison to
several different experimental scenarios. The activity in support of S2 experiments was
concentrated on hybrid heating studies and on high β N . Inter-shot analysis of the q profiles has
been a useful tool to achieve the desired configuration, mainly monitoring the central q value.
Broadening of the q profile observed by MSE data in particular conditions of high beta discharges
is under detailed investigation.
• Polarimetry. Polarimeter data were analysed to find the consistency of measurements with various
theoretical models developed recently. A dataset including 300 discharges (2003-2006) was
created, where the validation of data of the interferometer and the light detection and ranging -
Thomson scattering (LIDAR-TS) system was accurately checked. The dataset contained
measurements of channel 3 of the JET polarimeter [A1.82]. A parasitic experiment (Polarimetry at
q
10
8
6
Progress Report 2006
28

0.08
Fig. A1.28 – Line integrated plasma density (n _ e ) deduced from Cotton-Mouton is
plotted vs neL measured by the interferometer
n e,Cotton-Mouton
0.06
0.04
0.02
0
-0.01
0 10 20 30
n e,interf.
s3/s2
0.22
0.18
0.14
0.10
KG4 data PHAS
shot#66002 ch#3
numerical solution
numerical solution
including T e
corrections
0.06
0.02
high n e and T e ) was partially executed (during the
40 45 50 55 60 65 70
TOFOR commissioning sessions, 19-21 April 2006) to
Time (s)
detect the effect of the (high, >6 keV) electron Fig. A1.29 – Cotton-Mouton phase shift measured by
temperature on model predictions (shots #66016, channel 3 of the polarimeter (blue trace) shot #66002,
#66068). The main conclusions of the analysis are that together with the calculation of the signal made using the
the Cotton-Mouton phase shift can be used for Stokes equations (black stars) and including the effect of
evaluation of the line integral of electron density. the electron temperature (red crosses)
Figure A1.28 shows the line integral of electron density
n _ e obtained from the polarimeter Cotton-Mouton measurement vs the n_ e measured by the
interferometer. Substantial agreement emerges from the two independent measurements of plasma
density. The model of Stokes equations (where the inputs are taken from LIDAR TS and EFIT equilibrium)
is in good agreement with measurements even when corrections for the electron temperature are
included in the analysis (fig. A1.29).
• Neutron emission profiles and fuel ratio measurements. ELMy-H mode plasma scenarios with tritium
puff of the Trace Tritium Experiment have been analysed by using simultaneous DD 2.5–MeV and DT
14-MeV neutron emission profile measurements. Two-dimension spatial profiles of the tritium
concentration were obtained, which provided useful information for transport analysis and tritium
diffusion [A1.80].
• NE213 liquid scintillator. A new neutron detection system has been built and installed at JET. It is based
on a liquid scintillator cell (NE213-BC501 A), a light emitting diode (LED) connected to stabilisation
hardware for PMT high voltage gain control, and a light guide as interface/coupler with the
photomultiplier XP2020. The LED system provides calibrated and stable light pulses (at 1 kHz) used as
reference for gain control purposes. The new system makes use of DPSD electronic hardware that
provides separate neutron and gamma signals for spectra acquisition. Pulse height spectra of various
plasma scenarios were acquired during the JET restart and 2006 experimental campaigns. The present
activity together with the project Prototype Digital Pulse Shape Discrimination Module, a new
neutron/gamma DPSD, is aimed at improving neutron spectroscopy at high count-rate operation for
future JET applications, and at assessing its potential for ITER.
A1.6 PROTO–SPHERA
The PROTO-SPHERA [A1.83] system proposed at the ENEA Frascati research centre is a simply
connected magnetoplasma configuration composed of a spherical torus (ST, with external diameter
[A1.81] R. Giannella et al., Rev. Sci. Instrum. 75, 4247 (2004)
[A1.82] F. Orsitto et al., Proc. 33 rd EPS Conference on Plasma Physics (Rome 2006), on line at: http://epsppd.epfl.ch/Roma/pdf/P1_073.pdf
[A1.83] F. Alladio et al, Nucl. Fusion 46, S613 (2006)
References
29
Progress Report 2006

A1 Magnetic Confinement
A Fusion Programme
Anode
Spherical torus
Divertor PF coils
Screw pinch
Cathode
R=R EL
a)
Water-cooled anode ring
b)
SP
SP
I e
Gas flux
ST
ST
I ST
R EL =0.4 m
Directly heated cathode ring
Gas exhaust
2R sph =0.7 m, with closed flux surfaces and toroidal
plasma current I ST ≤240 kA) and a hydrogen plasma
arc, taking the form of an electrode-fed screw pinch
(SP, with length L Pinch ~2 m and midplane diameter
2×ρ Pinch ~0.08 m, with open flux surfaces and plasma
electrode current I e =60 kA), see figure A1.30a). Such a
combined plasma configuration has been devised
theoretically under the name "flux-core-spheromak"
(FCS). The central metallic conductor of a spherical
tokamak will be replaced in PROTO-SPHERA by the
screw pinch acting as a plasma central column. The SP
and the ST will have a common embedded magnetic
separatrix: resistive instabilities will drive magnetic
reconnections, injecting magnetic helicity, poloidal flux
and plasma current from the electrode-driven SP into
the ST and converting into plasma kinetic energy a
fraction of the injected magnetic energy. The SP will be
magnetically given a disk-shape near each electrode
(fig. A1.30b)), with singular magnetic X-points on the
2 m
symmetry axis.
Fig. A1.31 – MULTI-PINCH
The MULTI-PINCH experiment (fig. A1.31) is being built
as an initial partial setup of PROTO-SPHERA, devoted
to assessing and clarifying the most critical points of the PROTO-SPHERA experiment from the SP
point of view: to explore the breakdown conditions and the pinch stability in the starting phase of
the PROTO-SPHERA discharge, in the presence of the poloidal field (PF) shaping coils alone and
therefore in the absence of the spherical torus.
2.5 m
Fig. A1.30 – a) Sketch of the PROTO-SPHERA system; b) Cut-out sketch of plasma and electrodes
As a sign of international support for this project, a collaboration in the field of spherical tokamaks
has been established with the UKAEA-Culham Association. ENEA Frascati has been contributing
with personnel to the MAST experiment in Culham since 2004. In 2004 UKAEA Culham donated to
Frascati the available START equipment (in particular the vacuum vessel), and further contributions
from UKAEA-Culham can be expected during the final construction phases, commissioning and
operation of MULTI-PINCH, with respect to diagnostics and manpower.
SP
Progress Report 2006
30

Fig. A1.32 – The four pairs of MULTI-
PINCH PF coils and an enlargement
of a few details
Fig. A1.33 – Isometric view of the MULTI-PINCH anode
Fig. A1.34 – Isometric view of the MULTI-PINCH cathode
MULTI-PINCH will produce a stable screw pinch with current I e ≤8.5 kA, namely, the current expected in
PROTO-SPHERA before the ST formation. The four pairs of PF shaping coils will be fully recovered for
PROTO-SPHERA. In 2005 the constructive design of the PF shaping coils was completed with ASG
Superconductors (Genoa, Italy) and their construction will be completed by the beginning of 2007
(fig. A1.32).
A European call for tender for the construction of the remaining parts of the MULTI-PINCH load assembly
will be sent out in spring 2007. Examples of the detailed drawings for the MULTI-PINCH load assembly are
given in figures A1.33-A1.34.
The power supplies have been defined and their procurement should be such as to have the machine
ready for operation in 2009.
If the MULTI-PINCH experiment gives positive results, the PROTO-SPHERA setup can be completed by
adding the ST compression coils, along with an improved power supply, capable of raising the pinch
electrode current from I e ≤8.5 kA to I e =60 kA in ∼1 ms.
31
Progress Report 2006

A2 Preliminary Design of FT3
A Fusion Programme
Fusion is the most promising energy source as it can satisfy the energy needs in a safe and environmentally
responsible way. For fusion to play a major role by the second half of the 21st century, rapid exploitation of
ITER and an adequate parallel programme of material development (IFMIF) are mandatory, as proposed in
the so–called “fast track” approach to fusion energy. Within this approach, the construction of a
demonstration/ prototype reactor (DEMO) could start after ten years of ITER operation. Such an ambitious
time schedule specifically requires rapid progress in the exploitation of ITER during the first ten years of
operation, so that the DEMO regimes of operation can be demonstrated by the start of DEMO construction.
This requires parallel R&D activities on devices that are able to simulate burning plasma conditions but are
more flexible than a nuclear device such as ITER.
The European Power Plant Conceptual Study shows that the DEMO regimes must go beyond the regimes
developed for ITER. Although the extrapolation in plasma parameters (with respect to ITER) is limited, their
demonstration will require a significant exploratory effort. Indeed, DEMO will operate with a fraction of selfgenerated
(bootstrap) current close to 70%, use sophisticated methods for plasma control and require
techniques to reduce the heat flux on the plasma–facing components. All these requirements push the
plasma close to the operational limits where the risk of plasma disruptions is high. Furthermore, different
technological solutions for plasma–facing components and control methods have to undergo testing, which
would clearly be difficult and expensive to perform directly on a nuclear device such as ITER. Thus, the
successful development of the DEMO scenarios, prior to testing them on ITER, requires a preparatory
activity on smaller (than ITER) devices with sufficient flexibility and capable of simulating burning plasma
conditions.
Although many of the existing devices can provide important contributions to the preparation of ITER
operation, the requirement that the plasma behaviour be sufficiently close to that of ITER sets stringent
constraints on the plasma conditions that must be achieved in order to investigate ITER-relevant scenarios
in a meaningful way. The aim of the present proposal is to show how the preparation of ITER scenarios can
be effectively implemented on a new facility that will: i) work with deuterium plasmas, hence avoiding the
problems associated with the use of tritium, and will simulate alpha–particle dynamics by using fast ions
accelerated by heating and current drive systems; ii) work in a dimensionless parameter range close to that
of ITER; iii) be capable of long pulse operation at high plasma performance; iv) test technical solutions (e.g.,
the use of full tungsten) for the first wall/divertor that are directly relevant to ITER and DEMO.
Such a facility (FT3) could be ready in advance of the ITER operation phase and would require, taking into
account the infrastructures available in Italy, limited investment and operation costs. FT3 would be
designed, constructed and operated in the framework of a collaboration with other associations. In
particular, FT3 would make use of the competence available at ENEA, the National Research Council (CNR)
Milan and at the Reversed Field Pinch Experiment (RFX) consortium and would be the focus of Italian
activities in fusion after completion of the FTU and RFX scientific programmes.
Progress Report 2006
32

A2.2 Scientific Motivation of the Proposal
Rationale for the choice of FT3 parameters. The conditions to be satisfied in order to reproduce
ITER–relevant plasmas can be summarised as follows:
• ITER–relevant geometry (same shape of magnetic surfaces and divertor configuration);
• a ratio between energy confinement time and electron-ion equipartition time similar to that of ITER;
• production and confinement of energetic ions in the half-MeV range;
• a large ratio between heating power and device dimensions to simulate the large heat loads on the
divertor plasmas;
• pulse duration (normalised to the plasma current diffusion time) similar to that of ITER to study plasma
scenarios in steady–state conditions.
It is possible to show that these conditions imply the following set of parameters:
• plasma current I above 4.6 MA;
• auxiliary heating systems able to accelerate the plasma ions to energies in the range of 400 keV;
• device dimension of 1.8 m;
• pulse duration up to 100 s.
To accelerate plasma ions up to 400 keV, it is impossible to use neutral beams produced by accelerating
positive ions, which is the most diffuse heating scheme, as neutralisation efficiency rapidly drops above
140 keV. Other methods such as ion acceleration by ICRH or neutral beams produced by accelerating
negative ions have to be employed.
The FT3 parameters are shown in table A2.I and compared with those of JET, JT60 SA (the proposed upgrade
of the JT60-U device, under the Broader Approach Agreement) and ITER. Comparison of FT3 with JET and
JT60-SA shows that the dominant heating scheme in FT3 is ICRH, whereas in JET and JT60-SA, positive
neutral beam injector is mostly employed; thus only in FT3 can fast ions in the correct energy range be
produced; also the pulse duration is much longer in FT3 than in JET.
The initial configuration will be equipped with ICRH (20 MW), ECRH (4 MW) and LHCD (6 MW) power.
Although such a configuration is adequate to investigate the physics issues relevant to the FT3 mission,
the machine is designed so that, if necessary, further upgrades in auxiliary power (in particular a neutral
beam injector) could be accomodated.
Plasma parameters and equilibrium configurations. The ITER design currently foresees the
investigation of three main equilibrium configurations: a) a standard H-mode at I=15 MA with a broad
pressure profile (p o /=2); b) a hybrid mode at I=11 MA with a narrower pressure profile (p o /=3); c) a
steady-state scenario at I=9 MA with a peaked pressure profile (p o /=4). The FT3 equilibrium
configurations have been designed so as to reproduce the ITER equilibrium configurations with the plasma
current being scaled by a factor of 3. The corresponding
Table A2.I – FT3, JET, JT60-SA and ITER parameters
plasma parameters are shown in table A2.II which reports
the parameters achievable with an auxiliary power of
FT3 JET JT60-SA ITER
20–30 MW for each scenario.
All the plasma equilibria satisfy the following constraints:
a) a minimum distance of 3 λ E between plasma and first
wall to avoid interaction between plasma and main
chamber (here, λ E is the energy e-folding length,
assumed to be 1 cm on the equatorial plane); b) current
density in the poloidal field coils not exceeding 30 MA/m 2 .
Within these constraints enough flexibility is maintained to
allow different plasma shapes, efficient pumping and
strike point sweeping. The location of the poloidal field
R(m)/a(m) 1.8/0.6 3.0/1.0 3.0/1.0 6.2/2.1
B(T) 6.7 3.9 2.7 5.3
I(MA) 5.0 3.9 5.0 15
P ICRH (MW) 20 12 0 20
P NNBI (MW) 0 0 10 40
P PNBI (MW) 0 25 24 0
P ECRH (MW) 4 0 7 20
P LH (MW) 6 3 0 20 (*)
t flat-top (s) ∞ 10 ∞ ∞
(*) to be decided
33
Progress Report 2006

A2 Preliminary Design of FT3
A Fusion Programme
Table A2.II – FT3 plasma parameters. The magnetic field is
6.7 T in all the cases
H-mode H-mode Hybrid Steady state
I(MA)/q 95 5/3 5/3 3.6/4 2.8/4.9
H 98 1.0 1.0 1.3 1.5
n 20 3.7 2.6 1.95 1.35
n/n GW 85% 60% 60% 60%
P(MW)/P th (MW) 30/13-23 20/11-17 20/9-13 30/8-10
β N 1.85 1.42 1.8 2.1
t flat-top (s) 6 6 30 100
τ E (s) 0.42 0.48 0.47 0.25
T o (keV) 7.9 8.6 8 13
T plate (eV) 22 26 67 76
f rad 32% 27% 30% 53%
coils has been optimised to minimise the magnetic energy, produce enough magnetic flux (up to
25 Wb) for the formation and sustainment of each scenario and produce a fairly good field null at
plasma breakdown (B p /B T

W/cm 3
60
40
20
Hf power absorption by species Minority ion distribution function
0
a) b)
Total ICRH power density
Power to ion
-1
Power to electrons
-2
Maxwellian
-3
0
0 0.2 0.4 0.6 0.8 1
r/a
log(F)
-4
-5
0 100 200 300 400
E(keV)
Fig. A2.2 – ICRH power deposition
profile for various absorption
mechanisms a) and minority-ion
distribution function b). The dominant
mechanism is minority absorption (red
curve) which produces localised
heating in the plasma centre, similar to
alpha-particle heating in ITER. The fast
ion energy is in the range of 400 keV
an energetic ion population in the direction parallel to the equilibrium magnetic field, complementing in this
way the physics that can be studied with ICRH–produced fast ions with velocity mostly perpendicular to
the equilibrium magnetic field.
Ion cyclotron absorption was estimated using the FELICE and TORIC codes. Both codes solve the integrodifferential
equation for wave propagation and absorption: FELICE solves the equation in slab geometry
wirh the use of the self-consistent electric field radiated by the antenna; TORIC solves the equation in
toroidal geometry by employing a spectral method. Three absorption regimes can possibly play a role:
minority absorption (which is the absorption channel to be maximised), electron Landau damping and
mode conversion to the ion Bernstein wave. The analysis by the FELICE code tends to give a lower
minority absorption than those from TORIC. A minority concentration of 2% yields a minority absorption of
50% (FELICE) or 70% (TORIC) with a two–strap antenna with relative phase 180°, the remaining power
being directly absorbed by the electrons. The power deposition profiles are shown in figure A2.2a) for the
various absorption mechanisms. The parameters of the ion tails produced at a power level of 24 MW are
in agreement with the Stix theory, with the local absorbed power density obtained by the deposition code.
For the case shown in the figure, the effective temperature of the ion tail predicted by the Stix theory is 188
keV for a power density of 45 MW/m 3 . The minority ion distribution
Steady state
function is shown in figure A2.2b). The fast–particle concentration is
β N =2.1 T 0 =13 keV n 0 =21×10 19 m -3
0.3%. Note that a slowing down distribution function with a maximum 0.12
P
energy of 400 keV has an average energy between 189 keV and
EC = 3 MW
q = 2
W sat
149 keV for a critical energy between E c =150 keV and E c =50 keV. 0.10
Therefore, the fast–particle population is in the correct range of
0.08
parameters to simulate the ITER fast–particle dynamics.
The MARS code was used to perform a preliminary analysis of the
global MHD stability for the steady–state scenarios in order to
investigate the possibility of stabilising resistive wall modes. The nowall
beta limit corresponds to β Nc =2.8, whereas an ideal wall at
r/a=1.3 has a beta limit corresponding to β Nc =3.24. Feedback
control analysis shows that the use of internal poloidal field sensors
can allow full stabilisation of the mode, using both internal and
external feedback coils, whereas radial field sensors do not allow
stabilisation.
(green) and with (red) EC power applied
The ECRH system on FT3 is mainly dedicated to stabilisation of
neoclassical tearing modes in hybrid and steady–state scenarios, at densities below 3.6×10 20 m -3 , which
is the cut-off for the ordinary mode for the chosen frequency of 170 GHz (the same as ITER).
As an example, figure A2.3 shows the m/n=2/1 island with (W) evolution for a 2.8 MA scenario and
β N =2.1. The wave is launched from an upper port at an angle of 10 ° . Wave propagation is evaluated with
the electron cyclotron wave Gaussian beam (EC GB) ray-tracing code. The m/n=2/1 island evolution as
determined by the modified Rutherford equation is also shown in figure A2.3. The island width (6 cm) is
maintained below the ECRH deposition width for an injected power of 3 MW and corresponds to about
50% of the value at saturation without ECRH applied.
Lower hybrid current drive can be used on FT3 to control the current density profile in advanced scenarios.
W(m)
0.06
Δ ' p =-2 j_peak EC /j bs ~ 1.08
w/δ j = 1
0.04 EC on
B t =6.7 T r s /a=0.68
0.02 β p =0.6(β cr =0.05) ω=1.3 kHz
0
0 0.5 1 0.5 2
t(s)
Fig. A2.3 – Evolution of the 2/1 island without
35
Progress Report 2006

A2 Preliminary Design of FT3
1.2
0.8
Fig. A2.4 – Lower hybrid wave trajectories for three values of the
parallel refractive index
A Fusion Programme
Z(m)
0.4
0
-0.4
-0.8
-1.20
-1.20 -0.8 -0.4 0
R(m)
0.4
0.8
1.20
A2.3 Preliminary Design Description
A study of LH penetration and absorption was
performed in a parameter range typical of FT3 scenarios.
Figure A2.4 shows the ray trajectories for a plasma
equilibrium representative of FT3 scenarios with an
average density 10 20 m -3 and a central temperature of
13 keV. The wave frequency is 3.7 GHz and three
values of the parallel refractive index are considered. In
all the cases the wave is absorbed around mid–radius as
requested for this kind of scenario. Current drive
efficiencies in the range 0.3×10 20 Am 2 /W are predicted.
Load assembly. The FT3 load assembly (fig. A2.5) consists of the vacuum vessel and its internal
components (first wall, divertor, passive stabiliser structure), the magnet system and the poloidal
field coils. Since the maximum flat-top duration is 100 s, an actively cooled oxygen–free copper coil
system has been chosen for the magnet. The cooling is guaranteed by pressurised sub-cooled
(ΔT sat =26 nitrogen [LN 2 ]) flowing through suitable channels carved in the coil turns. Each turn is fed
by LN 2 independently (the LN 2 flows in parallel in each turn) to limit
the pressure drop to an allowable value and therefore avoid
LN 2 vaporisation. The Ohmic power dissipated in the
magnet is about 50 MW, which implies a LN 2 mass
flow of about 1600 kg/s. The magnet consists of 18
coils, each made up of 14 copper plates suitably
worked in order to have 6 turns in the radial
direction. Ripple correction is made by ferritic
inserts. The 14 plates are welded corresponding to
the most external region in order to obtain a
continuous helix. The maximum turn thickness is
30 mm. The plates are tapered at the innermost region
to get the needed wedged shape; the minimum turn
thickness is about 15 mm. The magnet insulation is made of
Fig. A2.5 – FT3 Load assembly glass-fabric epoxy, both for ground and inter-turn.
The coils are fixed together by the surrounding steel structure. Two pre-compressed rings situated
in the upper-lower zone keep the whole toroidal magnet structure in a wedged configuration. The
structure is also used to position the poloidal coils, which surround the toroidal magnet, and to fix
the vacuum vessel supports. Cooling of the toroidal–magnet structure is obtained by the contact
with the actively cooled components. To enable operation at 80 K, the whole machine is kept under
vacuum by a metallic cryostat.
The magnet dimensions were determined by the cooling requirements. It was necessary to limit the
current density to 30 MA/m 2 . It turns out that from the structural standpoint the magnet section is
adequate to sustain the forces. The first rough evaluation of the stresses indicates the maximum Von
Mises to be below 200 MPa. The fabrication process is based on well–assessed technology utilised
for FTU and other prototypical components, so no further R&D is required for the construction of
the toroidal magnet.
The FT3 free-standing central solenoid (CS) is segmented in six coils to allow plasma–shaping
flexibility, to facilitate manufacture and to allow cooling. The poloidal field coils and busbars are made
of hollow copper conductors. They have to withstand both vertical and radial electromagnetic loads,
Progress Report 2006
36

and are free to expand radially. The power to be removed from the poloidal field coils, keeping the coils at
cryogenic temperature, is about 12 MW. The CS coils are layer wound and have an even number of layers
for the electrical leads to be located on the same side of the coil. The conductors are wrapped with glass
fabric and kapton tapes and vacuum impregnated with epoxy resin. Radial grooved plates at the interfaces
between coil segments maintain concentricity. To limit the pressure drop, thus avoiding LN 2 evaporation,
cooling is achieved by feeding each layer independently. Due to the large dimensions of the most external
poloidal coils, a pancake configuration is adopted in order to allow cooling of the single turn. As for the
toroidal magnet, the poloidal–coil section and conductor size were determined by the cooling
requirements.
The total FT3 LN 2 daily consumption, determined on the basis of four 100–s pulses a day or 16×10 s
pulses, for the toroidal magnet and the poloidal coils is 150 t. A further consumption of 15 t for the losses
through the ports and other feedthroughs as well as to keep the vacuum vessel at room temperature must
be considered.
The vacuum vessel is segmented by 20–degree modules. To minimise the vacuum vessel time constant,
the shell is made of Inconel and the port in stainless steel. The maximum thickness of the shell is 30 mm,
while the ports are 20 mm thick. The shell is manufactured by hot forming and welding. Following the
previous experience with FTU, the vacuum vessel will be supported by the toroidal field magnet system by
means of vertical brackets attached to the TF coil case through the vessel equatorial port. According to
this constraint scheme, thermal expansion/contraction of the vessel is allowed, while nonsymmetric
displacements that might appear during disruptions or plasma vertical displacement events are restrained.
Twelve vacuum vessel sectors are equipped with five access ports. The maximum force during plasma
disruption is about 300 t for a 5–MA operating scenario. The thickness of the wall is adequate to sustain
such a load. The vessel time constant is about 30 ms. The operating temperature of the vessel ranges
from room temperature to 100°C. A suitable water loop is dedicated to maintaining the vessel
temperatures.
The first wall and the divertor are actively cooled by pressurised water with velocity respectively 5 and
10 m/s. These components have been designed to exhaust up to a maximum heat power of 50 MW
during long pulse operation. The first wall surrounds most of the vessel wall. It consists of a bundle of tubes
armoured with 3–mm plasma–spray tungsten. The heat load impinging on the first wall is, on average,
1 MW/m 2 with a peak of about 3 MW/m 2 . The solution adopted is well suited to resisting these loads,
having been tested up to 7 MW/m 2 . The first wall is also able to work as a limiter during plasma startup.
Its temperature will be maintained around 100°C to avoid impurity adsorption. The design has to be
remote-handling compatible. Maintenance will be carried out from equatorial and upper ports. The divertor
has to withstand a heat flux in excess of 20 MW/m 2 . The only suitable technology in this case is
monoblock, which has been tested extensively in the relevant heat flux range. The armour consists of
hollow tungsten tiles inserted in a copper tube heat sink. The heat flux component will be supported by a
steel frame which acts also as a cooling circuit. To enhance the critical heat flux, swirl tapes are provided
in the most loaded zone. The configuration has to allow easy maintenance operation as the possibility of
having to substitute some components is likely. The scheme is similar to that of ITER, with the frame acting
as a carousel all around the machine. Maintenance will be carried out from the lower port.
A remote handling system, similar to the JET FARM, has been conceived for unplanned (emergency)
operations. Standard maintenance tasks will instead be accomplished by a plug-in design of the
diagnostics and the antennae and casked solutions. The ITER divertor maintenance procedure will be used
wherever possible. The procedure is based on the development of an ad hoc cassette–mover tractor
capable of grasping and moving the divertor cassette. Some of the maintenance tasks of the first wall are
similar to those foreseen for the divertor, so pipe sizes could be standardised to be able to share cut and
weld devices. For the first wall assembly and disassembly a classical articulated boom plus a front end
manipulator have been considered.
Power supply. The FT3 power supply system includes three main subsystems: the 400–kV main
switchyards, the poloidal field coil (PFC) power supplies and the toroidal field coil power (TFC) supplies.
Figure A2.6 shows the total power for a 5–MA scenario with a total heating power of 30 MW
37
Progress Report 2006

A2 Preliminary Design of FT3
Fig. A2.6 – Active, reactive and total power for the reference FT3 pulse
A Fusion Programme
P(MW) Q(MWAr) S(MVA)
500
400
300
200
100
P: active power
Q: reactive power
S: total power
0
-21 -15 -9 -3 0 6 12 18 24 30 36 42
-100
Time (s)
(corresponding to about 80 MW requested at the
grid) and a stationary load of 25 MW. Due to the
amount of requested power, connecting to a
powerful node of the 400–kV Grid would be
desirable. Nevertheless, an accurate check by the
National Grid Regulator (GRTN), including both
active and reactive power effects on the specific
grid, might show that a 220–kV line could be
adequate. Lacking such an evaluation, the
400–kV line is taken as the reference solution.
Within the assumed 400–kV reference solution,
FT3 needs a dedicated switchyard to supply the
PFC, TFC, additional heating systems and
auxiliaries. All the loads are fed by one main step–down transformer (400/36 kV) with three
secondary windings: two (225 MVA each) star connected and grounded through a resistor, to supply
FT3, and one (80 MVA) delta connected to allow free circulation of third harmonic currents. On the
request of the GRTN, active power shedding resistors could be connected to the tertiary winding.
Sharing the total power between two secondary windings has the aim of making it possible to use
the 36–kV level on the secondary sides (instead of the more expensive 75–kV level), limiting the rated
current within the present breaker capability at this voltage. Each circuit for the supply of the TFC and
the various PFCs is generally made up of a converter transformer, a thyristor converter unit, a protective
crow-bar and high-speed, solid–state switches for the additional resistance units. No specific study for
the breakdown phase has been made so far.
Table A2.III – ICRH system parameters
Operating frequency range ( MHz) 60±90
Peak power (MW) 20
Bandwidth (MHz)
±2MHz (-1db)
Pulse width (s) ≥ 100
Time interval between two
100–s pulses (s) 1800
Type of antenna
3 rows of 2 straps
Power per strap (MW)
1 (at generator)
Power coupled per antenna (MW) 5
Max radiated power density (MW/ m 2 ) 10
N. of antennae 4
Power per generator (MW) 2
N. of rf generators 12
Heating systems. The FT3 auxiliary heating
systems are consistent with the present state
of the art and do not require additional R&D
activity. FT3 is equipped with three systems:
ICRH, ECRH and LHCD. A description of the
ICRH system is given in table A2.III. At a
magnetic field of 6.7 T, the use of 3 He minority
requires a frequency of 68 MHz. In its initial
configuration the system will couple 20 MW to
the plasma. A possible design of the ICRH
antennae could be based on an array of six
(two toroidal by three poloidal) current straps
protected by a Faraday shield made of a set of
16 non–tilted elements, with a smoothed
rectangular cross section. The Faraday shield
has to suppress the components of the emitted radiation parallel to the local B-field, and shield the
electrically active components from direct contact with the plasma. All the antenna components
(straps and Faraday shield rods) are water-cooled. Each antenna is fed by three high–power
tetrodes “TH 526”, with a maximum rf power output of 2 MW in the frequency range 35-80 MHz.
Three of the generators are supplied by a 33–kV/380 A solid–state unit. The antenna, together with
the respective vacuum transmission lines and vacuum windows, is integrated in a plug inserted in
an equatorial port and removable as a single unit.
The performance of the antenna was studied with the TOPICA code on the reference FT3 H–mode
plasma scenario at 68 MHz with 2% 3 He minority. Electric current and magnetic current/electric field
distribution were obtained in vacuum and with the plasma. The analysis in vacuum of the optimised
antenna showed very good (low) inter-strap coupling. The analysis with plasma demonstrated the
good performance of the antenna array in terms of power coupled to the plasma: for the standard
configuration and for a maximum voltage of 30 kV, a power of 5 MW can be coupled to the plasma
by each array. The launched power spectrum has a maximum for n || =±6. Figure A2.7 shows the
current distribution on the straps, demonstrating the good efficiency obtained with this geometry:
Progress Report 2006
38

Fig. A2.7 – Distribution of current on the straps
the current on the straps has a very high absolute value and is almost
constant along the entire length of the straps.
Four identical units compose the ECRH system, each with a gyrotron, a
transmission line and a launcher. The four launchers are located in the same
port. Each gyrotron is fed by an independent power supply, capable of
high–frequency modulation (up to 10 kHz) and designed to be used as the actuator
in the feedback loop for mode suppression. The power is delivered from the source to
the launcher by means of an evacuated corrugated waveguide. The reference design for the launcher is
based on front steerable mirrors, with real–time control of toroidal and poloidal injection angles. No barrier
window is considered in the transmission line. The whole system is designed to operate in feedback mode
with real-time control of the main parameters (polarization, beam current, mirror steering, power, fault
management), addressing in this way technological issues relevant for a system working in a
thermonuclear plant. The considered gyrotron is a 170–GHz/1–MW source, with depressed collector and
a pulse length larger than 100 s, based on the results of the R&D activity for ITER. Each gyrotron is fed by
a 55–kV/50–A high–voltage power supply. The transmission line is an evacuated aluminium corrugated
waveguide (i.d. 63.5 mm) matched to the gyrotron output beam with an elliptical mirror. Since the power
dissipated on the waveguide is small and the pulse length does not exceed 100 s, no direct cooling of the
waveguide is considered, while all the other components (mitre-bends, polarizer, dc-break) must be
cooled. The launcher under study, based on the front steering concept for major flexibility in terms of beam
shaping and injection angles, is located in the upper vertical port. In this way, the intersection of the EC
resonance with the q=2 surface, for the relative neoclassical teaning model (NTM) stabilisation, is reached
with limited diffraction effects. The front mirror is real–time controlled at a speed compatible with all the
issues assigned to the ECRH system (NTM stabilisation, sawtooth control, disruption mitigation). The beam
spot radius (waist) in the plasma resonant region can be less than 3 cm, below the expected width of the
NTM m/n=2/1 island at saturation. The overall losses of the design of ECRH system (waveguide, mitrebends,
microwave components and launcher) are below 8%, which can be reduced further with a HE 11 to
Gaussian beam converter at the beginning and at the end of the waveguide.
The LHCD system is designed to routinely couple a rf power of 6 MW to
FT3 plasmas. The preliminary design is based on a frequency of 3.7 GHz in
pulsed regime, with pulse length up to 100 s. At this working frequency
high–power CW sources are available, i.e., the TH 2103 klystron, rated at
500 kW/CW and 650 kW/10 s; these klystrons (table A2.IV) are the
sources of the Tore–Supra and JET LHCD systems. The FT3 LHCD system
will be equipped with two passive–active multijunction (PAM) launchers,
which will simultaneously allow coupling LH waves in the plasma with severe
edge conditions and effectively water cool the antenna in long operations
and with heavy thermal loads. The dimensioning of the launcher, given the
frequency, is based on the requirement of launching a peak n || N ||peak =1.9
and by the cross section of the FT3 ports at the narrower point. The
Table A2.IV – TH 2103 main
parameters
resulting power density in the active waveguides is limited to P S =33 MW/m 2 , which is comparable with
the values normally achieved in JET and Tore Supra.
Taking into account ~20% rf losses in the transmission lines and in the launcher, a minimum rf power at
the generator of P Inst =7.5 MW has to be installed, i.e., 15 klystrons to be used.
Diagnostics. A specific activity has been dedicated to studying the capability of performing detailed
measurements of the FT3 plasma parameters. The basic diagnostics include the magnetic diagnostic
(diamagnetic loops, saddle coils, Hall sensors and pick–up coils), the CO 2 interferometer for measuring the
electron density profile, the ECE (Michelson, polychromator and radiometer) and Thomson scattering
systems for measuring the electron temperature, the bolometric measurements for plasma radiation,
various spectroscopic measurements (visible, UV and x rays) of the impurity content, the neutron camera
and neutron spectrometer for 2.45 neutron emission, the activation measurement and the gamma-ray
scintillator. These diagnostics will be taken from FTU with minor adaptations. Further diagnostics could be
provided as a contribution in kind from other associations. Discussions are under way to assess this
possibility.
ıJı (dB,interp)
-11.540
-14.704
-17.869
-21.033
-24.198
-27.362
-30.527
-33.691
-36.856
-40.020
Frequency
3.7 GHz
Bandwidth @ - 1 dB 10 MHz
Output power (CW) 500 kW
Gain
47 dB
Cathode voltage
60 kV
Beam current
20 A
Efficiency 42%
Modulating anode voltage 45 kV
Modulating anode current 50 mA
39
Progress Report 2006

A3 Technology Programme
A Fusion Programme
The technology activities carried out by the Euratom-ENEA Association in the framework of the European
Fusion Development Agreement (EFDA) concern the continuation of the ITER, DEMO and IFMIF R&D
programmes. ENEA has also started design and preliminary R&D activities under the Broader Approach
agreement between the EU and Japan.
In 2006 the most important results of the technology programme were achieved in the fields of plasmafacing-component
development and testing, neutron data, remote handling. However, it should be noted
that all the activities contributed substantially to the progress of the fusion programme as a whole (safety,
the Power Plant Conceptual Study, engineering activities).
The divertor CFC/W monoblock mockup fabricated using ENEA’s patented processes was tested under
fatigue heat loads and achieved results that surpassed those of the other technologies. The next step is to
industrialise the technology for application in the construction of the ITER divertor heat flux components.
Significant work was done to define quality assurance for neutronics analyses. Mockups of the ITER precompression
ring made in glass fibre epoxy were fabricated.
ENEA is also equipped to contribute to the ITER construction, not only through the continuing R&D
activities, but also through participation in the development of the ITER neutron radial camera and laser invessel
viewing system and, as an associate of the Consortium of Associations, in the construction of the
first nuclear fusion components - the Test Blanket Modules.
The activities and results documented in the following illustrate ENEA’s readiness to enter the new era
opened with the decision to build ITER.
A3.2 Divertor, First Wall, Vacuum Vessel and Shield
Manufacturing of small-scale W monoblock mockups
Manufacturing of the prototypical component by means of the two ENEA patented technologies
(pre-brazed casting [PBC] and hot radial pressing [HRP]) was successfully concluded (Underlying
Technology and European Fusion Development Agreement contract [EFDA] 03/1054) [A3.1-A3.5].
Thermograph non-destructive (ultrasonic, lock-in thermography) testing performed with the SATIR
(Commissariat à l’Energie Atomique ([CEA]) equipment showed there was no evidence of defective
zones. After the testing the mockup was sent to the FE200 e-beam facility at Le Creusot France for
thermal fatigue tests (fig. A3.1).
The testing plan started with a screening at 5 MW/m 2 of absorbed power. The mockup was
successfully tested at ITER-relevant heat fluxes: 10 MW/m 2 for 3000 cycles (all), 20 MW/m 2 for
2000 cycles on the carbon fibre composite (CFC) part, 15 MW/m 2 for 2000 cycles on the tungsten.
Progress Report 2006
40

Fig. A3.1 – Monoblock mockup installed in the FE200
facility before high heat flux testing
Figure A3.2 shows the infrared images taken during
the high heat flux testing and during the 3176 th
cycle performed at 20 MW/m 2 of absorbed power
on the CFC tiles and 15 MW/m 2 on the tungsten.
The images highlight the absence of surface
overheating. The mockup was also subjected to
critical heat flux (CHF) (fig. A3.3) to verify its
behaviour under ITER-relevant thermal-hydraulic
conditions. A CHF of 35 MW/m 2 was obtained and
it can be said that this value is well above that
estimated and gives a margin of 1.75 with regards
to ITER nominal loading. For the first time it was
possible to measure the CHF of a monoblock
component with armour tiles still joined on the tube.
Figure A3.4 shows the mockup after the CHF
testing, while still connected to the FE200 facility.
Fig. A3.2 – a) CFC part, b) W part
10/19/06 INFRAMETRICS 17:59:19
The complete manufacturing and successful testing
of this vertical target medium-scale mockup
(fig. A3.5) can be considered a success for both the
PBC and the HRP processes. A survey of the
manufacturing technologies for the ITER divertor
has shown that they are valid alternatives to the
current techniques.
2353
1997
2195 2098 1981
Fig. A3.3 – Infrared image during CHF testing at 34.2 MW/m 2
Fig. A3.4 – CFC surface after CHF testing
[A3.1] M. Merola et al., Fusion Eng. Des. 56-57, 173 (2001)
[A3.2] M. Rödig et al., Fusion Eng. Des. 56-57, 417 (2001)
[A3.3] M. Rödig et al., Investigation of tungsten alloys as plasma facing materials for the ITER divertor, presented at the 6 th Int. Symposium on
Fusion Technology - ISFNT-6 (San Diego 2002)
[A3.4] E. Visca et al., Fusion Eng. Des. 56-57, 343 (2001)
[A3.5] M. Rödig et al., Post irradiation testing of samples from the irradiation experiments PARIDE 3 and PARIDE 4, presented at the 11 th Inter.
Conference on Fusion Reactor Materials - ICFRM-11 (Kyoto 2003)
References
41
Progress Report 2006

A3 Technology Programme
Engineering Design Activities: V
and VI test campaigns
A Fusion Programme
Fig. A3.5 – Mockup after thermal fatigue testing (high heat
flux testing [HHFT] and CHF)
Fig. A3.6 – Mockup PH-S-39 B being tested in EDA-
BETA apparatus
The objective of the test campaigns in the
framework of Engineering Design Activities
(EDA) is to characterise the primary first-wall
(PFW) panels in terms of their behaviour
under thermal fatigue in ITER-relevant
conditions (temperature and thermal flux).
Thermal fatigue tests on the PFW mockups
under ITER-relevant operative conditions
were carried out to qualify the Be-metallic
heat sink (Cu alloys)-austenitic 316L steel
panel joints made by hot isostatic pressing.
In 2006 the EDA V campaign was
successfully concluded, with neither melting
nor erosion/failure of the Be tiles. Under the
new EDA VI campaign, started in late
autumn, the plan is to accomplish 30000
thermal cycles. As in the previous
campaigns, the two mockups (one shown in
fig. A3.6) delivered to ENEA by EFDA, are
being tested under thermal fatigue cycling
with an emitted thermal flux up to a
maximum of 0.65 MW/m 2 and a cycle
period of 300 s. The EDA-BETA
experimental setup (fig. A3.7), consisting of
a glove box operated under vacuum and
suitably instrumented (with a high specific
power CFC resistor), is connected to the
CEF 2 water loop at ENEA Brasimone for the
mockup cooling. The main features of the
whole experimental apparatus (EDA-BETA +
CEF 2) are summarised in table A3.I. The
experimental activity should be concluded in
late summer 2007.
Fig. A3.7 – The two mockups assembled in EDA-
BETA apparatus
Hydraulic characterisation of
full–scale divertor components
Table A3.I – EDA-BETA + CEF 2
The main aim of the activity, started in 2006,
EDA-BETA dimensions (Φ×l)
700×1200 mm is to perform an exhaustive thermal
Max. power delivered by the resistor 41 kW
hydraulic experimental campaign on the
ITER divertor plasma-facing components
Max. thermal flux emitted by the resistor 0.65 MW/m 2
(PFCs), i.e., outer vertical target (OVT),
Loop design temperature 140 °C
dome liner (DL) and inner vertical target (IVT)
Max. CEF1 pump flow-rate
2×70 kg/s
in stationary state and transient conditions.
Max. pump head
2×1.2 MPa
Both types of tests are carried out in the
CEF 1 water loop at ENEA Brasimone.
Hydraulic tests in stationary state are aimed at determining the pressure drop of each component,
verifying the balance of the parallel water flows and assessing possible conditions for the insurgence
of cavitation. The tests on the OVT and DL were successfully carried out in the last part of 2006.
Figure A3.8 shows the OVT connected to the CEF 1 loop. Figure A3.9 reports the pressure drops
across the OVT, experimentally determined at three different temperatures (20-50-100°C). Tests in
transient conditions are aimed at evaluating the efficiency of discharging the activated water, not
Progress Report 2006
42

Pressure drop (bar)
5.0
4.0
3.0
2.0
1.0
0.0
20°C
80°C
50°C
100°C
y=0.0228x1.8223
R 2 =1
y=0.0188x1.8809
R 2 =0.9998
y=0.0171x1.9083
R 2 =0.9997
y=0.0174x1.881
R 2 =0.9996
5 8 11 14 17 20
Water flow (kg/s)
Fig. A3.8 – OVT connected to CEF 1 loop
Fig. A3.9 – Pressure drops across OVT
drainable by gravity, into the divertor modules. The procedure 30
to efficiently accomplish this task has been identified as 20
consisting of a first phase of draining by high-pressure gas,
10
followed by drying of the residual water by low-pressure dry
gas. This part of the experimental activity requires up-grading
0
6/3 20/3 35/3 45/3 55/3 60/3 75/3 85/3 110/3
of the CEF 1 loop: most of work on the technical
specifications and on the design and construction activities
Ratio H 2 /H 2 O
has been done and will be concluded in early spring 2007. Fig. A3.10 – EUROFER experimental results in terms
ENEA Brasimone and the University of Palermo are of PRF as a function of the ratio H 2 /H 2 O
collaborating on validating a thermal-hydraulic code through
correlation with the experimental results for both types of
test. Once validated on the basis of the experimental results achieved, the
code will be used to predict the behaviour of the single components of the
ITER divertor as well as the integrated divertor cassette, in fully relevant
operative conditions and component geometry.
PRF
H permeation through EUROFER and heat exchanger
material (Incoloy, Inconel)
In 2006 the PERI 2 device was modified. A more precise quadrupole was
adopted and the mixing-gas system, with a mass flow meter on the Ar line,
a mass flow controller on the H/D line and a gas humidifier and humidity
measuring system, was moved from the low- to the high-pressure side. After
start-up of the modified device, tests with deuterium were begun, using a
ratio of three between deuterium and water. The experiment gave no
appreciable results: after a few seconds, there was a reduction in the Fig. A3.11 – Removal of the inner
permeated hydrogen flux but, continuing the experiment, this effect was vertical target
annulled and the flux reached the steady-state value. Tests were performed
with EUROFER in accordance with the test matrix, using hydrogen instead of deuterium. This solution was
adopted as the new quadrupole demonstrated high precision in measuring hydrogen concentration, unlike
the old instrument. A precise range of water/hydrogen mixtures in which the permeation reduction factor
(PRF) is appreciable was identified, and the campaign on EUROFER was concluded. Experimental results
obtained in terms of PRF are reported in figure A3.10.
Formal trials for the new ITER divertor cassette refurbishment
Since the divertor cassettes need to be replaced and updated several times during the ITER lifetime,
refurbishment of these components must be performed rapidly and with a high standard of safety. To
assess the feasibility of such refurbishment operations a test campaign consisting of two complete
assemblies and disassemblies (fig. A3.11) of the three PFCs, also called targets, was performed during
43
Progress Report 2006

A3 Technology Programme
A Fusion Programme
2006. The trials were aimed at validating the procedures already developed as well as evaluating the
suitability of the present cassette design (i.e., new ITER 2001 divertor cassette) for remote handling.
According to the results of the tests, the assembly process appears to be better than the
disassembly process. In fact the tool used for pin extraction was not correctly dimensioned and
hence target disassembling operations were carried out hands on. At present the tool design is
under revision. The other tools developed, such as the plasma-facing-component transporter
(PFCT), the pin expansion tool (PET) and the drilling machine, fulfil the specification requirements.
The activities were completed in July 2006.
A3.3 Breeder Blanket and Fuel Cycle
DEMO breeding blanket
Work continued on the development of the dual coolant lithium lead (DCLL) concept and the
possibility of having a vertical module segmentation (VMS) for the blanket [A3.6, 3.7]. The results
have pointed out the potential of the DCLL blanket to operate in the required DEMO environment
with allowable temperatures and stresses. The VMS studies showed a gain in reducing the electromagnetic
loads during disruption, therefore reducing the requirements for the support structure. The
results of dimensioning of the supports and studies on the kinematics in the vessel showed that it
might be possible to replace the whole blanket using a reasonable number of ports. The ports are
being studied in relation to the hypotheses on the DEMO magnets.
European Breeding Blanket Test Facility
Throughout 2006 ENEA was strongly involved in R&D activities for both the helium-cooled pebble
bed (HCPB) and the helium-cooled lithium-lead (HCLL) test blanket modules (TBMs) to be tested in
ITER. The work was focussed on i) experimental activities related to the development of relevant
technologies and ii) the design, construction and upgrading of the experimental facilities, which will
allow ENEA to retain the EU leadership in the field of experimentation on TBMs.
The construction of the liquid metal loop was started in April 2006. The reference parameters, fixed
in the design phase, are volumetric flow rate 0.03-0.9 m 3 /h; maximum temperature 550°C;
minimum temperature 300°C; thermal cycling 400 s T max , 1400 s T min ; liquid metal inventory in the
module 0.4 m 3 ; liquid metal inventory in the loop, including the TBM, 0.6 m 3 ; cover gas argon. The
main modifications foreseen to upgrade HEFUS3 are a new water heat exchanger of 900 kW and a
new electric power supply unit of 1 MW, in order to provide 250 kW of electrical power to the first
wall and 750 kW to the breeding region of the TBM mockups. HEFUS3 will be equipped with a new
He compressor capable of reaching a maximum He flow-rate of 1.4 kg/s with a head of 0.9 MPa.
Its installation is foreseen for late 2007. In 2006 the technical specifications for the HEFUS3
upgrading were prepared. The conclusion of the activity, with the installation of the above-mentioned
modifications, is scheduled for the end of 2007.
Thermo-mechanical characterisation of HCPB mockup
The thermo-mechanical behaviour of the breeder and neutron multiplier pebble bed in reactorrelevant
conditions is one of the main concerns in the design of the HCPB blanket for DEMO and
the TBM to be tested in ITER. Hence, experimental results and predictive models are of basic
importance in developing this blanket concept. During 2006 the HELICA mockup was dismounted
and the OSi pebbles recovered. The pebbles were “filtered” and the production of powder, after 34
thermal ramps, quantized in 4%. Scanning electron microscopy (SEM) examinations (fig. A3.12) of
the pebbles showed that they kept their physical integrity. A new experimental setup was designed
for thermo-mechanical characterisation of the HEFUS3 experimental cassette of the lithium-
Progress Report 2006
44

Fig. A3.12 – SEM of HELICA OSi pebbles
beryllium pebble bed (HEXCALIBER) mockup, designed and
manufactured to reproduce a portion of the former TBM-HCPB
with two OSi and two beryllium pebble bed cells, both heated
by couples of flat electrical heaters (figs A3.13). The mockup will
be tested in 2007 in the HEFUS3 facility, under appropriate
adjustment of bed temperatures, temperature gradients,
coolant temperatures, flow distributions and mechanical
constraints, to assess the thermo-mechanical performance of
the pebble beds under steady-state and cyclic-heat power
conditions. To perform the test campaign in safety, avoiding any
possible Be contamination, the whole experimental setup was
designed at a pressure of 2.0 MPa and a temperature of 500°C
and equipped with three independent helium circuits: one circuit
for the cooling plates of mockup, and two purge flow circuits for
OSi and Be beds, a vacuum system, and a double oil guard.
Each circuit will have units for filtering and monitoring the Be
powders. A preliminary study of the HEXCALIBER mockup
thermo-mechanical behaviour under steady-state conditions
was performed in the framework of the benchmark exercise to
select the best constitutive model for thermo-mechanical
prediction of pebble bed behaviour under blanket-relevant
conditions, among those developed by ENEA
Brasimone/University of Palermo, the Nuclear Research
Consultancy Group (NRG) Petten and Forschungszeuntrum Karlsruhe (FZK). In particular, a realistic 3D finite
element model (FEM) of HEXCALIBER (fig. A3.14), simulating a 1-cm-thick slice of the whole mockup, was
developed. A realistic set of loads and boundary conditions was applied, taking into account natural
convection with air, forced convection with helium coolant (T=300–400°C, p= 8 MPa) and distributed
electric heat generation within the heating plate electric resistors. A thermal contact model was implemented
at the interface bed-wall and bed-heater, where no mechanical sliding was assumed. Poloidal plain strain
was assumed to simulate the continuity of the mockup in that direction. The thermal field obtained matches
the prefixed goals, showing in each pebble bed the expected trapezoidal poloidal profile with the flat portion
located in the pebble bed layer
between the heating plates. A
decreasing radial profile from the
centre to the extremity of the bed
can be seen in each pebble bed.
Maximum temperatures of 819
and 563°C have been calculated
for the OSi and the Be pebble
bed, respectively. Analysis of the
mechanical volumetric strain field
within the beds shows that they
experience only compressive
strain states. The highest
mechanical strains are reached
within the OSi pebble beds and
are (≈0.17) one order of
magnitude higher than in Be
beds.
NT11
+8.300e+02
+7.934e+02
+7.569e+02
+7.203e+02
+6.837e+02
+6.471e+02
+6.106e+02
+5.740e+02
+5.374e+02
+5.009e+02
+4.643e+02
+4.277e+02
+3.911e+02
+3.546e+02
+3.180e+02
Max +8.292e+02
at node PART-1-1.15106
Min +3.197e+02
at node PART-1-1.958
Acc.V Spot Magn Det WD Exp 500 μm
20.0 kV 5.0 50x SE 15.9 823 PM13206 Li4SiO4 Helica 2>138 μm
A
A
Fig. A3.13 – HEXCALIBER mockup
z [m]
0.225
0.125
0.025
400 600 800
T (°C)
Fig. A3.14 – HEXCALIBER FEM model: thermal field and its poloidal profile along
the path A–A
[A3.6] C. Nardi, S. Papastergiou and A. Pizzuto, Development of DCLL blanket, ENEA Internal Report FUS-TEC BB MC R 0016 (2006)
[A3.7] C. Nardi, S. Papastergiou and A. Pizzuto, DEMO blanket segmentation, ENEA Internal Report FUS–TEC BB MC R 0017 (2006)
References
45
Progress Report 2006

A3 Technology Programme
A Fusion Programme
Table A3.II – EFDA-approved text matrix
Test Temperature Pb Li flow rate Stripping Ar flow rate
no (°C) (kg/s) (Nl/h)
TRIEX loop for studying
technologies for extracting
tritium from Pb-17Li
1 450 0.2 10
The first year of activity with the
2 450 0.5 100 (150)
TRIEX loop, delivered to
3 450 0.35 55 (80)
Brasimone at the end of 2005,
4 450 0.2 100 (150)
was dedicated to loop
5 450 0.5 10
acceptance tests and
6 450 0.35 55 (80)
qualification of the main installed
components (pump, gas
saturator, gas extractor by packed column). The first test campaign with a test matrix agreed on by
EFDA and other EU Associations will be performed in 2007. To verify the Pb-Li pump performance,
the loop was operated without the stripping gas flowing in the expected operative Pb-Li flow rate
range between 0.1 to 1 kg/s. Varying the Pb-Li mass flow rate, the attainment of 500°C as
maximum operative temperature was also evaluated to check the loop electrical heating system.
Then, the Ar gas system injection was operated to check the Pb-Li levels in the saturator and
extractor by using the gas mass flow control systems of the facility. After loop qualification the real
experimental activities will start with a first experimental test campaign and an optimised test matrix,
obtained by adopting a factorial method developed by CEA to better exploit each test, optimise the
test para meters and consequently to reduce the total number of tests. Table A3.II reports the test
matrix approved by EFDA. The experimental activities will be concluded at the end of 2007.
Conceptual design of auxiliary systems for HCPB-TBM
The possibility to recover with high efficiency the tritium generated in the HCPB blanket as well as
the fraction permeated into the He main cooling system is one of the main objectives of the blanket
test campaign planned in ITER. In summer 2006 ENEA was charged by EFDA with studying the
selection and conceptual design of the tritium extraction system (TES) and coolant purification
system (CPS) for the HCPB TBM. For both systems the activity consists in determining the inlet gas
composition during the different ITER operational phases, examining all the technological options
and selecting the most suitable and, finally, proposing a first conceptual design.
Both the TES and the CPS are based on the technology of physical adsorption on microporous
materials in different system configurations (pressure temperature swing adsorption (PTSA), TSA,
PSA), integrated with other systems which, depending on the process requirement, makes it possible
to reduce HTO in HT (Zn or Zn-Fe-Mn reactors) or, on the contrary, to oxidise HT to HTO (Cu 2 O-CuO).
Structural analyses during em loading
For the TBM with the HCPB concept, numerical models have been developed to take into account
the presence of pebble beds during electromagnetic (em) transient structural analyses. The em force
distribution increases to an asymptotic maximum and then drops exponentially to null. These loads
can produce oscillations in the TBM structures when the force disappears, or no oscillations if the
damping is large enough. Hence it is necessary to analyse the behaviour (stiffening and inertial) of
the beds during such a quick transient load (30-100 ms), differently from the “low-velocity” models
used up to now for thermal cycling modelling.
The presence of pebble beds inside the TBM is analysed though the definition of two representative
simplified models of the complete structure: 1) a submodel of the grid structure; 2) a submodel of a
breeder unit. The steel components are assumed to behave linearly elastic, while the modified
Drucker-Prager model is implemented for the mechanical characterisation of the pebble layers. The
results of oedometric tests for the beryllium pebble beds are used to calibrate the parameters of
such a constitutive relationship. On the basis of the numerical simulation it is possible make the
Progress Report 2006
46

following considerations: a) The pebble layers have a moderate effect on the structural behaviour of the
TBM components if shear deformation is dominant. In this case, the volume reduction of the TBM cavities
and, thus, the compaction of the pebble beds are negligible. As a consequence, the stress state in the
steel frame undergoes minor variations when the granular filler is considered in the numerical model. b)
When the deformation of the steel frame during em loading acts in such a way as to produce compaction
of the pebble beds, their effects become much more significant: the presence of a filler material produces
a sensible increase in the structure stiffness and, at the same time, the stress field is redistributed within
the steel frame. Usually, when the pebble beds are included in the model the stress intensity is less critical
compared with the case of the empty frame (without pebble material). Nevertheless, em loading can induce
high stresses in regions not designed to bear them.
VDS catalyst tests
Samples of Plexiglas, Polyvinyl chloride, vacuum pump oil and Teflon were burnt at 200°C in a 1 m 3 oven
(fig. A3.15) and the combustion fumes were sent onto catalytic beds consisting of platinum on Al 2 O 3 (Escat
26 furnished by Engelhard). The aim was to reproduce the case of a fire in the tritium laboratory of ITER
Fig. A3.15 – Materials used in the combustion tests: a-b) Plexiglas, c-d) PVC, e-f) vacuum pump oil with Teflon
47
Progress Report 2006

A3 Technology Programme
Fig. A3.16 – View of the PERMCAT module (above) and the
Pd-Ag thin–wall permeator tube (below)
A Fusion Programme
and to study the poisoning of the catalyst of the
vent detritiation system (VDS). The test results
show that the combustion fumes of PVC, pump oil and Teflon could affect the Pt-based catalyst
efficiency even if, under the operating conditions of the VDS, all the tritiated gases (HT) are
converted into tritiated water [A3.10].
Permeator tubes
The PERMCAT reactor module designed by ENEA has been completed (fig. A3.16). This device has
a special mechanical design in which two pre-tensioned metal bellows avoid any compressive and
bending stresses of the thin-walled (50 μm), long (500 mm) Pd-Ag permeator tube produced via
cold rolling and diffusion welding of metal foils [A3.8, A3.9].
A3.4 Magnet and Power Supply
ITER magnet casing welds
The tests performed at ASG Superconductors Genoa to verify the use of electron beam welding
procedures to perform the root weld for the ITER magnet casings (in AISI 316 LN modified with high
nitrogen content) showed that, using the welding apparatus in ASG and the proposed welding
procedures, it is not possible to weld a 40-mm thickness of this material [A3.11].
ITER pre-compression ring fibreglass composite material
A new batch (VR 5) of unidirectional fibreglass composite has been produced in the new kettle. The
increase in the active length of the kettle allowed the production of 650-mm-long samples, which
were tested with the use of the grip system (45’ fibreglass grips kept in place by 15 Inconel
compression rings in) at room temperature (RT) and 77 K (5 samples per temperature). The results
showed that the mean value of the ultimate tensile strength was 2200 MPa at RT and 2766 MPa at
77 K. In both test sets the dispersion in the values of the mechanical characteristics (ultimate
strength, elasticity modulus and fracture elongation) was very low, the maximum being lower than
3% of the mean value [A3.12]. Relaxation tests are envisaged for 2007.
High-frequency/high-voltage solid-state modulator for ITER gyrotrons
Activities regarding construction and testing of the solid-state modulator (EFDA contract 02-686)
were successfully completed during the first months of 2006. At the same time, a new task was
started on design support and digital simulation of the entire power supply system for the European
collector depressed potential (CDP) gyrotron test.
A3.5 Remote Handling and Metrology
In the sharing of the in-kind contributions to ITER, the EU is to procure the in-vessel viewing and
ranging system (IVVS), which has to be able to provide sub-millimetric 3D images inside the
activated machine. Results of the relative R&D performed during the last six years have shown that
Progress Report 2006
48

Fig. A3.17 – IVVS Probe
Table A3.III – IVVS operating conditions
Temperature 250°C
Vacuum
10 -9 mbar
Magnetic field
5 T
γ Radiation (rate/total) 1.5 kGy/h; 5 MGy
Viewing&ranging accuracy

A3 Technology Programme
A Fusion Programme
be noted that the
requested viewing
and ranging per -
formances are fully
met for both. The
Thermally stressed zones
749.0
viewing performance
795.8
was compared with a
standard resolution
chart, while ranging
accuracy was
identified as the
605.1
604.1
standard deviation of
all the ranging
measure ments. The
IVVS probe was
characterised by
defining ranging
accuracy vs the
Fig. A3.19 – Viewing and ranging test on ITER DVT metal side (d=2 m; ranging
backscattered power
accuracy standard deviation σ < 1 mm)
received by the
probe. The experimental results fully comply with the theoretical expectations. The ITER operator will
have a friendly tool to easily evaluate the expected accuracy according to the target position and
characteristics. Possible probe modifications (increasing laser power and modulating frequency)
were identified in order to upgrade the present performance by about a factor of 10.
mm
306.0 pixels
294.0 pixels
A3.6 Neutronics
Quality assurance for neutronics analysis for ITER
mm
342.0 pixels
348.0 pixels
Quality assurance (QA) procedures for neutronics analyses (task TW5-TDS-NAS1–D1) were drawn
up in collaboration with the ITER Responsible Officers for Neutronics and for the Management and
Quality Programme. The final version of the procedures, applicable to the ITER Central Team and to
external suppliers, was issued on the basis of feedback received, and loaded on the ITER
Documentation Management (IDM) system [A3.13]. As a test case the procedures were applied in
a neutronics task order implemented by EFDA on the ITER diagnostic plug analysis, and the
application was monitored. A series of actions was also undertaken to assess and provide
adequate instruments for the QA procedures. First, the status of the MCNP brand model for
neutronics analyses of ITER was reviewed. It was found that there were many outdated models that
contained significant differences as they had been developed for various specific purposes, so an
up-to-date reference model was produced and made available, and the reference materials
specified in the model were reviewed. According to QA procedures, computer software for
neutronics calculations has to be verified and validated prior to use. Codes were identified that had
already been verified and validated during the ITER EDA R&D activities and can be used in the ITER
neutronics analyses and calculations. The related documentation was collected. However, other
codes, or new versions of codes, may be developed for specific purposes: a verification/validation
procedure was worked out for these cases and applied to code packages under development, such
as CAD-MCNP interfaces, Attila code and D1S, R2S package (MCNP-FISPACT coupled) for dose
rate calculations. Three separate validation efforts were launched for the Fusion Evaluated Nuclear
Data Library (FENDL)-2.1, selected as the reference library for ITER. ENEA and the Japan Atomic
Energy Agency (JAEA) conducted the analysis of experimental benchmarks performed at the
Frascati Neutron Generator (FNG) and the Fusion Neutron Source (FNS), respectively, during ITER
Engineering Design Activities (EDA), and FZK conducted a computational benchmark on a simplified
ITER geometry.
Progress Report 2006
50

Finally, the ITER Nuclear Analysis Report (NAR) was reviewed and the parts that need to be updated were
identified. As a general result, it was found that new calculations are needed for many components, taking
into account the present design. A table of contents has been written for the next issue of the NAR.
Compared to the previous NAR structure, more emphasis will be given in the new issue to the components
rather than to the models used.
ITER systems: nuclear design
The activity was focussed on interfaces for divertor sensors and optical elements in the divertor cassette
and lower ports, the design and integration of the sensors and optical access and on a review of divertor
integration issues for the various diagnostic systems. In particular, work was carried out on the interface
and integration issues of the new “Divertor 2006 Design” of the neutron diagnostics (lower vertical neutron
camera and divertor neutron flux monitors) of the optical systems (divertor impurity monitor and divertor
Thomson scattering) and the residual gas analyser systems, taking into account possible locations in ITER
divertor, detectors/techniques, vacuum and magnetic field interferences, radiation, cooling and
cabling/power supplies. Calibration issues/schemes of the neutron systems and related integration
engineering aspects were studied to identify the ITER in-situ neutron calibration procedures. The rationale,
requirements, specifications concerning the neutron test area (NTA) and the NTA-hot cell system
integration issues were reviewed. The NTA is a laboratory for inspection trials, calibration, commissioning,
cross checking and support of neutron diagnostic sensitive devices and equipment during all the
functioning periods (assembly, operation, shutdowns, maintenance) of ITER [A3.14, A3.15].
TBM HCPB and HCLL neutronics experiments
In 2006 the neutronics experiment on a mockup of the European Union TBM, HCPB concept was
completed [A3.16, A3.17]. The aim was to validate the capability of nuclear data to predict nuclear
responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment was
carried out at the FNG 14-MeV neutron source in a collaboration between ENEA, the Technical University
of Dresden (TUD), FZK and the Joseph Stefan Institute of Ljubljana, with the participation of JAEA under
the International Energy Agency (IEA) Implementing Agreement on a Co-operative Programme on the
Nuclear Technology of Fusion Reactors. A slight underestimation was found in the calculation of tritium
production in the range (1–C/E)~5...10% on average. The resulting total uncertainties on C/E for the TPR
prediction were about 9 – 10% at 2σ level [A3.18]. The observed underestimation of the measured tritium
production by less than 10% on average is therefore at the lower bound of the assessed uncertainty
margin. Behind the mockup, the fast neutron flux (E>1 MeV) was found to be overestimated by
calculations by about 10–20%, while the gamma-ray flux is underestimated by about 10-20% [A3.19,
A3.20]. The slow neutron flux investigated by time-of-arrival spectroscopy is also underestimated by
[A3.13] P. Batistoni, Quality assurance in neutronic analyses (May 2006), https://users.iter.org/users/idm?document_id=ITER_D_23H9A4
[A3.14] G. Bonheure et al., Nucl. Fusion 46, 725 (2006)
[A3.15] A. Costley et al., The design and implementation of diagnostic systems on ITER, presented at the 21 st Inter. Atomic Energy Agency (IAEA)
Fusion Energy Conference (Chengdu 2006)
[A3.16] P. Batistoni et al., Fusion Eng. Des. 81, 1169 (2006)
[A3.17] U. Fischer et al., Neutronics and nuclear data for fusion technology - recent achievements in the EU programme, presented at the 21 st
IAEA Fusion Energy Conference (Chengdu 2006)
[A3.18] P. Batistoni et al., Neutronics experiment on a HCPB breeder blanket mock-up, presented at the 24 th Symposium on Fusion Technology
- SOFT-24 (Warsaw 2006), accepted for publication in Fusion Eng. Des.
[A3.19] K. Seidel et al., Measurement and analysis of neutron flux spectra relevant to the tritium breeding capability in a neutronics mock-up of
a test blanket module for ITER, presented at the Int. Workshop on Fast Neutron Detectors and Applications (Cape Town 2006)
[A3.20] K. Seidel et al., Measurement and analysis of the neutron flux spectra in a neutronics mock up of the HCPB test blanket module,
presented at the 24 th Symposium on Fusion Technology - SOFT-24 (Warsaw 2006), accepted for publication in Fusion Eng. Des.
References
51
Progress Report 2006

A3 Technology Programme
A Fusion Programme
C/E
Percent nuclide heat contribution
1.3
1.2
1.1
1.0
0.9
0.8
0.7
0.6
C/E
1.15
1.05
0.95
0.85
Beta heat
C/E
Error bars include only
exp. uncertanties
Error bars include only exp. uncertanties
0.65 0.75
10 2 10 4 10 6
10 2
10 4
10 6
0.5
0.5
Decay time (s)
Decay time (s)
10 2 10 3 10 4 10 5 10 6 10 2 10 3 10 4 10 5 10 6
Decay time (s)
Decay time (s)
100
10
1
Mo91
Mo99
Beta heat
Nb98m
Nb97
Nb96
Zr97
Tc99m
about 20%. The last results are consistent with the weak underestimation of the tritium breeding
also found for the main block. According to these results, the shielding performance of the mockup
is predicted within ±20% accuracy. The tritium production (most of which coming from 6 Li) is mainly
sensitive to the 9 Be cross sections for elastic scattering and, to a lower extent, for the 9 Be (n,2n)
reaction. The sensitivity/uncertainty analysis showed that the TPR from 6 Li changes by about 2%
per % change of the 9 Be elastic scattering integral cross sections, but the sensitivity with respect to
the angular differential cross section dσ/d Ω could be higher [A3.21, A3.22]. These results indicate
that the angular differential cross section for 9Be elastic scattering may require further improvement.
Results from the HCPB mockup experiment implied that for the HCPB TBM in ITER the tritium
production is underestimated by the calculations based on the European Fusion File (EFF) and
FENDL nuclear data (used in this analysis) by less than 10% on average, at the lower bound of the
assessed uncertainty margin, and that the neutron and gamma ray shielding performance is
predicted within ±20% accuracy [A3.23]. The pre-analysis of the next experiment on a mockup of
the TBM HCLL has also been completed.
Experimental validation of neutron cross sections for fusion-relevant materials
1.1
1.0
0.9
0.8
0.7
0.6
Percent nuclide heat contribution
C/E
1.15
1.05
0.95
0.85
0.75
Gamma heat
Gamma heat
Fig. A3.20 – Results from decay heat measurements for beta and gamma. Inserts: experimental
uncertainties
0.1
10 2 10 3 10 4 10 5 10 6
0.1
10 2 10 3 10 4 10 5 10 6
Decay time (s)
Decay time (s)
100
10
1
Mo91
Nb97
Nb96
Mo99
Mo101
Mo93m
Fig. A3.21 – Beta and gamma heat nuclide contributions - 65 nm range
Y89m
Tc99m
Nb92m
Nb98m
Zr89
Nb95
The neutron-induced decay heat on samples of molybdenum (99.99 %) irradiated at FNG in a firstwall-like
neutron spectrum was measured (European Activation File [EAF] Project). Three
molybdenum samples were irradiated for about 3.5 h at FNG. One sample was monitored with the
ENEA decay heat measuring system where gamma and beta decay heats are simultaneously
Progress Report 2006
52

measured. The other two samples were monitored with HPGe detectors. The decay times studied went
from a few minutes up to some days after irradiation. Comparison between experimental data and EASY
predictions is satisfactory for both beta and gamma heat (fig. A3.20). A discrepancy (C/E=0.85±0.1 exp
err.) found in the gamma heat for short decay times could be due to the reactions 92 Mo(n,2n) 91 Mo,
92 Mo(n,2n) 91m Mo m (IT)→ 91 Mo, and/or to decay data of 91 Mo. This nuclide is responsible for about 90% of
the heat produced for a short decay time (

A3 Technology Programme
A3.8 IFMIF
A Fusion Programme
Remote handling of the back-plate bayonet concept – bolted solution
The reference International Fusion Materials Irradiation Facility (IFMIF) target design is based on the
concept of a replaceable back-plate. At present two different design options for the back-plate
replacement are under investigation: the
reference design system, i.e., the so-called cut
and re-weld concept proposed by the Japanese
IFMIF team, and the alternative solution
developed in Europe and known as the backplate
bayonet concept. The latter concept is
based on the possibility of replacing the backplate
while working laterally to the target, thus
simplifying the sequence needed to perform the
operations and, as already demonstrated,
reducing back-plate-replacement operational
time. In addition the bayonet concept has a
major advantage in that the material for final
disposal is reduced. Two prototypes of the backplate
bayonet concept were manufactured in
2002. The first prototype is provided with a
Fig. A3.24 – IFMIF back-plate bayonet concept closing system based on a skate system and
based on bolted closing system
has already been successfully tested, whilst the
second (fig. A3.24) is characterised by a closing
system consisting of bolted closure. The
experimental activities carried out on the second
prototype were aimed at evaluating its suitability
for remote handling (RH). Comparison of the two
prototypes was performed from the RH
viewpoint.
Fig. A3.25 – New em pump for LIFUS III: a) initial
phase of mounting; b) final phase
In particular, the activities were articulated as
follows: development of the installation and
removal procedures; modification of the
prototype; adaptation of the bolting tool, RH
trials themselves; post-analysis of results. The
RH activities for the target prototype were
executed in the ENEA Brasimone divertor
refurbishment platform (DRP) and were
successfully completed in July 2006.
Lithium corrosion and chemistry:
LIFUS III facility
In the framework of the key action phase of
IFMIF development, the activities carried out
during 2006 included corrosion/erosion testing
of AISI 316 and EUROFER 97 in IFMIF
representative conditions; experimental
validation of the lithium purification strategy,
based on a single cold trap and two hot traps
having specific getters for nitrogen and
hydrogen; functional validation of the
Progress Report 2006
54

performance of the resistivity-meter for lithium
impurities, developed in collaboration with Nottingham
University. These activities have to be performed in the
LIFUS III loop, which will be the first liquid metal loop at
Brasimone to have liquid lithium as process fluid. The
whole loop was refurbished because the previously
chosen canned pump failed continuously and had to
be substituted with an em pump (fig. A3.25). Due to the
strong reactivity of Li with damp air and water, stringent
safety measures were carried out. In addition, new
pipes were installed, the test section was modified to
enhance the safe manipulation of the corrosion
specimens, the gas purification was enhanced by
interposition of a specific on-line getter, a glove box
was installed, the data acquisition software was
adapted, the experimental hall was refurbished to meet
the safety requirements, and the personnel were trained to deal with lithium safety issues. Moreover the
design calculations were revised to match new pump conditions. The complete thermo-mechanical
verification of the piping was successfully performed by the ANSYS code. Similarity calculations were
carried out to adapt the pump performance to the loop conditions in order to find a new reference
hydraulic working point (fig. A3.26).
Bar
4
3
2
1
0
Pump interdiction area
Line of minimum
velocity (10 m/s)
open valve
valve 67% open
Calculated
operational point
-1
0 10 20 30 40
Liter/min
valve 33% open
V=300 V
V=340 V
V=380 V
Fig. A3.26 – Re-calculated operational point
Preliminary remote handling handbook for IFMIF facilities
A preliminary remote handling handbook (PRHH) is to be produced for the target and test facilities. It is
based on the work already done and on the documentation available. So far, apart from the introduction
giving a description of IFMIF together with the methodologies adopted, the work has been focussed on 1)
defining a set of rules to distinguish components requiring RH maintenance from those that can be
maintained hands on; 2) identifying components requiring RH maintenance (not completed); 3) developing
RH procedures for each component; 4) evaluating the area accessibility and interference with other
components and 5) defining the technical requirements for the devices and equipment to be used for the
RH operations. The following components and devices have already been studied in the target and in the
test facilities:
1. Target assembly and replaceable back wall: both concepts (cut & reweld option and bayonet option)
were studied and compared. The procedures for back-wall replacement were already known as well as
the equipment and devices to perform these operations. A preliminary study for the feasibility of backwall
replacement through a lateral window was also performed and will be included in the PRHH.
2. Replacement of the quench tank and other components from the target area.
3. Main Li loop components have been and, still are, under investigation.
4. Requirements for the main devices and tools have been defined: robotic arm and bolting tool for the
bayonet concept; common manipulator system (CMS); transporter for the target assembly system. (No
data are available for the YAG machine).
5. The general layout of the access cell of the test facility has been defined. (The path for the cooling
systems is still missing).
6. The general procedures for the vertical test assembly (VTA) replacement from the test cell, including
separation and transportation of the test modules in the test module handling cell, have been
completed.
7. Requirements for the main devices to be installed in the access cell have been defined: the Universal
Robot System (gantry crane), the support for the removal of the test modules and the transporter of the
test modules from the access cell to test module handling cell.
The work is expected to be completed by the end of June 2007.
55
Progress Report 2006

A3 Technology Programme
Inventories and dose rates induced by deuterons and neutrons in the
accelerator system
A Fusion Programme
ANITA-DEUT is the newly developed deuteron activation code package dealing with deuteroninduced
transmutation and activation. The code can work with two deuteron cross-section libraries:
the first based on the ACSELAM library; the second, on the file EAF_D_GXS-2005.1 of the EASY-
2005-1 system. A methodology approach was set up to calculate deuteron/neutron-induced decay
gamma sources and evaluate dose rates along the accelerator line because of deuteron beam
losses. The first step in the sequence consists of deuteron and secondary-neutron transport
calculations via the MCNPX code (version 2.5b). The deuteron and neutron spectra obtained are
used in ANITA-DEUT and ANITA-IEAF activation codes to calculate the radioactive inventories of
materials and the corresponding decay gamma sources, which are then used for gamma transport
calculations via the VITENEA-IEF/SCALENEA-1 and MCNP-4C2 code systems to obtain the beamoff
dose rates around the various parts of the IFMIF accelerator (i.e., radiofrequency quadrupole
(RFQ) and drift tube linac (DTL). The decay gamma dose rates do not represent a hazard source for
workers (maximum value 5.6×10 -2 μSv/h on the surface of the RFQ section 10) [A3.24]. Preliminary
calculations were performed for the last tank of the DTL with source deuterons of 40 MeV,
considering a deuteron current loss of 130nA (8.11×10 11 d/s). On the basis of this value, the beamoff
total dose rate around the DTL is less than 10 μSv/h at 1 day’s cooling time.
Inventories and dose rates induced by deuterons and neutrons in the cooling
system
The structural materials of the high-energy beam transport (HEBT) section can be activated by
deuterons due to beam losses, by secondary neutrons produced by deuteron-induced nuclear
reactions and by back-stream neutrons coming from the lithium target. The deuteron source energy
is 40 MeV. HEBT activation due to neutrons was evaluated through the MCNP-4C2 code with the
McEnea neutron source, which is based on the measurements of neutron emission spectra
produced in Li(d,n) reactions for Ed=40 MeV performed at the Cyclotron and Radioisotope Center
(CYRIC), Tohoku University, Japan. Preliminary calculations were performed, considering a deuteron
beam-loss current of 865nA (5.4×10 -2 d/s). With this value the most relevant contribution to decay
gamma dose rates in the area around the HEBT is due to the activation induced by lost deuterons
(about 70%). The dose rate contribution of back neutrons is higher than that caused by secondary
neutrons due to beam losses.
A3.9 Safety and Environment, Power Plant Studies and
Socioeconomics
Failure mode and effect analysis for the European test blanket modules
A failure mode and effect analysis (FMEA) was done to study possible safety-relevant implications
arising from failures in the HCPB [A3.25] and HCLL [A3.26] TBMs for ITER. For both modules, six
postulated initiating events (PIEs) were selected for deterministic assessments:
• FB1 (loss of flow in a TBM cooling circuit because of circulator/pump seizure).
• LBB1 (loss of TBM cooling circuit inside breeder blanket box: rupture of a sealing weld).
• LBO3 (loss of coolant outside vacuum vessel because of rupture of tubes in a primary TBM-HCS HX).
• LBP1 (loss of coolant outside vacuum vessel because of rupture of a TBM cooling circuit pipe
inside port cell).
• LBV1 (loss of TBM cooling circuit inside vacuum vessel: rupture of TBM-FSW),
• TBP2 (small rupture from "TBM - tritium extraction system" process line inside port cell).
Progress Report 2006
56

Failure mode and effect analysis for remote handling transfer systems of ITER
A FMEA at component level was done to study possible failures while performing remote handling (RH)
operations [A3.27]. Two safety-relevant PIEs were selected: 1) break in “vacuum vessel + cask” isolating
boundary during RH operations, inducing release of radioactive products (fraction of dust and T implanted
in vessel) into the port cell; 2) cask stop and leakage during RH transportation of divertor cassette to hot
cell, inducing release of radioactive products (fraction of dust and T implanted in transported components)
into the gallery. Deterministic analysis could be required to evaluate the response of the safety systems
(e.g., efficiency of ventilation systems, isolation of heating, ventilation and air conditioning [HVAC] system)
and effectiveness of rescue operations in mitigating the consequences and risks for workers. Compliance
of the design features with the safety limits in the case of a fire triggered on board the transporter should
be required. Some concerns on recovery scenarios should dust or tritium be released inside the port cell
or gallery could arise from the use of the air cushion transportation system. Accident rescue scenarios were
also identified by the FMEA and grouped in seven families.
Validation of computer codes and models
New contributions were obtained for validation of the activation code package ANITA-2000 against the
Karlsruhe Isochronous Cyclotron (KIZ) and ENEA FNG experiments [A3.28, A3.29]. In the KIZ experiments
a saturation thick beryllium target was irradiated by 19-MeV deuterons. Samples of vanadium alloys, nickel,
copper, lithium orthosilicate, EUROFER 97 and tungsten were irradiated. Specific activities in Bq/kg for
each sample material for several gamma ray emitting activation products were obtained and compared
with the calculation. ANITA-2000 handled satisfactorily the activation channels induced by neutrons with a
smooth continuum spectrum. The discrepancies between calculated and experimental (C/E) activity values
are in the range 10-20%. The results of irradiation of samples of molybdenum and tantalum at the 14-MeV
FNG neutron source were also compared with ANITA predictions. For molybdenum the agreement
between C/E beta decay heats is good (within 10%) for all cooling times, while it is within 15% for the
gamma decay heats; for tantalum the agreement is very good (within 2%) for all cooling times for the beta
decay heat and lower than 10% for the gamma decay heat.
Time factors to be used for the JET shutdown dose-rate evaluation [A3.30, A3.31] in the direct one-step
(D1S) method were obtained with the ANITA-2000 code (FENDL/A-2.0 activation data library). The ANITA-
2000 calculations were performed using the data related to a) the JET materials composition for the
detector positions D1 (irradiation end) and D2 (Geiger Müller tube); b) the irradiation scenarios (DD and DT);
[A3.24] D.G. Cepraga, M. Frisoni and G. Cambi, Evaluation of activation inventories and dose rates induced by deuterons in the IFMIF
accelerator system, ENEA Internal Report FUS-TN-SA-SE-R-146 (2006)
[A3.25] T. Pinna, Failure mode and effect analysis for the European Helium Cooled Pebble Bed (HCPB) test blanket module, ENEA Internal
Report FUS-TN SA-SE-R-152 (2006)
[A3.26] T. Pinna, Failure mode and effect analysis for the European Helium Cooled Lithium Lead (HCLL) test blanket module, ENEA Internal
Report FUS-TN SA-SE-R-155 (2006)
[A3.27] R. Caporali and T. Pinna, Failure mode and effect analysis for remote handling transfer systems of ITER FEAT, ENEA Internal Report FUS-
TN SA-SE-R-156 (2006)
[A3.28] V. Massaut et al., Validation of European computer codes used for fusion safety analysis, presented at the 8 th IAEA Technical Meeting
on Fusion Power Plant Safety (Wien 2006)
[A3.29] D.G. Cepraga, G. Cambi and M. Frisoni, ANITA-2000 activation code packages: 2005 validation effort against Karlsruhe Isocyclotron and
FNG-ENEA experiments, ENEA Internal Report FUS-TN-SA-SE-R-136 Rev.1 (2006)
[A3.30] L. Petrizzi et al., Benchmarking of Monte Carlo based shutdown dose rate calculations applied in fusion technology: from the past
experience a future proposal for JET 2005 operation, Fusion Eng. Des. 81, 1417-1423 (2006)
[A3.31] M. Angelone et al., Neutronics experiment for the validation of activation properties of DEMO materials using real DT neutron spectrum
at JET, Fusion Eng. Des. 81, 1485-1490 (2006)
References
57
Progress Report 2006

A3 Technology Programme
A Fusion Programme
%
%
c) the relevant isotopes considered for the shutdown dose rates at the selected JET D1 and D2
positions [A3.32]. A first analysis was performed by considering the JET irradiation scenario (DD and
DT) up to March 2004. The gamma material decay sources were obtained (1.5 y cooling time) and
used in SCALENEA-1 to get the shutdown dose rate (September 2005, measurement time) in the
D1 position. The C/E ratio obtained is 1.015.
Finally, updated analyses were carried out on the possibility of clearance of the ITER vacuum vessel
materials, considering the new (August 2004) unconditional clearance levels given in the IAEA Safety
Guide RS-G-1.7. The relevant results from the updated analysis [A3.33] are:
• The 430 ferritic steel and the SS 304B4 steel of the outboard vacuum vessel zone (VVSHDO) are
clearable after a longer time compared with the previous analysis results when the TECDOC-855
clearance level data were employed. This is particularly true for the SS 304B4 steel, which
becomes clearable only after about 6000 years (with respect to the 90 years of the previous
analysis).
• The remarkable change for the VVSHDO(2) – SS 304B4 steel is due to the highest contribution
(at 100 years’ cooling time) from the Ni-63, which is now (i.e., with the RS-G-1.7) about 40%
compared to the older (i.e., with the TECDOC-855) value of about 10%.
Dust removal experiments in STARDUST
Dust removal inside the plasma chamber is a concern with regard to machine performance and to
safety. Experiments were carried out in the ENEA STARDUST facility in 2005 [A3.34] by using a
stream of air in the volume representing the vacuum vessel in which characterised carbon, tungsten
and stainless-steel dusts were placed. The capacity of dust mobilisation by means of the air inflow
was between a few percent and 100%, depending mainly on the type of dust and on the kind of
150
125 T=20°C
150
75
50
25
0
1234 5678 9101112
35
30
25
20
15
10
5
0
T=50°C
1234 5678 9101112
N. test
%
%
8
6
4
2
0
T=20°C
1 2 3 4 5 6 7 8 9 101112
T=50°C
0
1234 5678 9101112
N. test
Fig. A3.27 – Mobilisation factor and capture factor for carbon dust in hot and
cold conditions (red 10 m, blue 30 m, yellow 1 h
8
6
4
2
deposition (heap or flat
layer). Mobilisation is more
effective in cold conditions.
The efficiency of the system
to capture dust on the filter
reached a maximum of
about 7.5% for carbon in the
geometrical configuration of
the STARDUST facility.
Heavy dusts such as SS316
and W did not reach the filter.
Figure A3.27 shows the
carbon dust results. The
tested technique of removing
the vacuum vessel dust has
low efficiency in the
collection of powder
removed from the vessel and
deposited on appropriate
surfaces (i.e., filters). The use
of an air stream directly on the dust deposit can improve the effectiveness of the removal but, to
collect a significant amount of dust in the filter, the pressure in the volume must be increased, so
that conditions which are dangerous for the internal equipment can be avoided.
Feasibility study of a torus-shaped facility for dust mobilisation studies
A feasibility study was carried out for a toroidally shaped facility for dust mobilisation and removal
experiments [A3.35]. The facility, named STARDUST-U (fig. A3.28), should facilitate the extrapolation
Progress Report 2006
58

Fig. A3.28 – View of the upper part of STARDUST-U
to ITER of the experimental results obtained during
tests in which dusts are mobilised. It allows the
monitoring, by laser diagnostics, of the dust
concentration evolution in the different zones of the
machine. The windows will make it possible to view
the dust mobilisation in the zone with laser systems if
the concentration is below 1000 particles/cm 3 .
Although this concentration can be far from
accidental conditions in the ITER device, the
dynamics of the phenomena during mobilisation can be helpful in extrapolating the results at higher
concentrations, mainly to test the performance of the dust simulation codes.
The estimated cost of the whole facility is about 16,600 Euros (2006 evaluation).
Post-accident occupational exposure and radioprotection
The objective of this study [A3.36] was to provide an indication of occupational radiation exposure (ORE)
consequences associated with post-accident recovery operations. The accident analysis results
documented in Volume VII of the ITER Generic-Site Specific Safety Report (GSSR) were used. The focus
was on the actions that are needed to restore the machine to the operational state, and on the potential
impact of the actions on the collective worker dose. Even the release of one gram of tritium (in the form of
HTO), one gram of activated corrosion products, or one gram of tokamak dust can cause significant
contamination concerns. Airborne contamination is not a significant problem, as this can be easily removed
by the building/room ventilation system working in conjunction with the re-circulating detritiation system
and exhaust detritiation system. Surface tritium contamination is a bigger concern, as this takes
considerably more time to reduce, to acceptable levels, using the same systems. Surface aerosols from
activated corrosion products (ACPs) and dust contamination could be an even bigger concern if water
sprays are either not available, or not effective, for washing the deposited aerosols and dust in the active
drain system.
Integration of design modifications (in Rapport Préliminaire de Sûreté) to tritium
building and detritiation system
The tritium confinement strategy of the ITER design was compared with the safety requirements and the
safety standards and guidelines (ISO 17873) related to nonreactor nuclear facilities to find possible critical
issues in the design of tritium confinement [A3.37]. According to ISO 17873 the tritium plant has only two
confinement barriers, whilst the actual safety reports on the ITER tritium buildings claim that three lines of
defence are available. According to ISO standards the process equipment and related containment
[A3.32] M. Frisoni et al., ANITA 2000 activation code package calculation in support of the ENEA Direct 1-Step D1S method, ENEA Internal
Report FUS-TN-SA-SE-R-150 (2006)
[A3.33] G. Cambi, D.G. Cepraga and M. Frisoni, Summary results of 2005 activation calculation in support of ITER, ENEA Internal Report FUS-
TN-SA-SE-R-135 (2006)
[A3.34] M.T. Porfiri, S. Paci and N. Forgione, Experimental campaign 2005 for the dust removal in the STARDUST facility, ENEA Internal Report
FUS-TN-SA-SE-R-145 (2006)
[A3.35] M.T. Porfiri et al., Feasibility study for a torus shape facility aimed at dust mobilization and removal experiments, ENEA Internal report
FUS-TN-SA-SE-R-158 (2006)
[A3.36] A. Natalizio, L. Di Pace and T. Pinna, Post-accident recovery: a worker dose perspective, ENEA Internal Report FUS-TN SA-SE-R-149
(2006)
[A3.37] C. Rizzello and L. Di Pace, Tritium building and detritiation systems. Considerations on tritium confinement, ENEA Internal Report FUS-
TN-SA-SE-R-148 (2006)
References
59
Progress Report 2006

A3 Technology Programme
A Fusion Programme
enclosure form the first containment barrier, while according to ITER the process equipment
constitutes the first barrier and, in the specific case, glove boxes are part of the second barrier. The
ventilation flow rates chosen appear too low compared to ISO Standards and also to other related
guidelines (e.g., US Department of Energy standards).
An alternate concept of the ITER atmosphere detritiation has been proposed to mitigate accidental
tritium releases [A3.38]: a scrubber capable of contacting with a spray of water all the air effluent
from the areas where the tritium systems are located. If a tritium spill occurs in the room atmosphere,
the stream of scrubbing water will dilute the concentration of HTO in the effluent, thus reducing the
related environmental impact. A critical analysis of this unit is however required to demonstrate the
capability of such a concept to cope with the ITER safety requirements.
Collection and assessment of data related to JET occupational radiation
exposure
The scope of the work [A3.39] was to update the database of JET ORE experience up to the end
of 2005. The collective worker doses are the highest during the machine shutdown state, but are
due primarily to in-vessel work. The monthly collective worker doses accrued from ex-vessel work
during the shutdown state are comparable to those accrued during the non-shutdown state. The
maintenance group collective doses are the highest during the machine shutdown state, but are due
primarily to in-vessel work. The maintenance group monthly collective doses accrued from ex-vessel
work during the shutdown state are comparable to those accrued during the non-shutdown state.
The majority of the collective doses from ex-vessel work, with the machine in the shutdown or nonshutdown
state, is accrued by non-maintenance workers. Finally, there is no significant difference
for ex-vessel exposure time between the shutdown and non-shutdown state. The same is true for
work effort. It is possible to conclude that most of these results could be generally applicable to
ITER. In fact the ITER doses accrued during the non-shutdown state could be expected to be a
significant fraction of the total dose, as they are at the JET facility.
JET data collection on malfunctions and failures of ICRH system components
The data from operating experience of JET for the ion cyclotron resonance heating (ICRH) system
were gathered for the data collection on failures of components used in fusion facilities [A3.40].
Alarms/failures and malfunctions occurred during operations from March 1996 to November 2005.
Data related to crowbar events were also collected. About 3400 events classified as alarms or
failures related to specific components or sub-systems were identified. The ICRH was operated
during about 12000 plasma pulses from March 1996 to November 2005. Failure probabilities on
demand were evaluated with regard to the number of pulses operated. The highest number of
alarms/failures (1243) are related to erratic/no-output of the instrumentation and control (I&C)
apparatus. Tetrode circuits failed 829 times, the high-voltage power supply system 466 times and
the tuning elements 428 times. The maximum number of events related to I&C (595) led to
anomalous operations of CODAS, followed by 125 anomalous operations of stubs. The number of
failures/alarms of the ICRH system increases quite linearly with the number of pulses in which the
system is operated. A crowbar event happened on average every nine ICRH pulses. The rate of
failure on demand of an ICRH module is about 0.29/pulse.
JET dust in-vitro experiment: result assessment and in-vivo experiment
literature review
The work dealt with the analysis of in-vivo experiments and dosimetry models on the inhalation of
tritiated dust [A3.41]. The most consistent in-vivo experimental activity on the inhalation of metal
tritides was performed at the Lovelace Respiratory Research Institute (Albuquerque, NM, USA),
using Ti, Hf and Zr tritides with different size distributions. The aim of these experiments was to set
up a biokinetic and dosimetry model to better describe inhalation of T particles in a living being, and
Progress Report 2006
60

to derive suitable dose conversion factors. Analysis of the experimental results confirmed the concerns
about the inadequacy of the protection guidelines for workers exposed to tritium in particulate form (dusts,
flakes), if based on the radiotoxicity of tritium. The behaviour of tritiated dust in the human body is still not
well understood, considering the different size distributions and the variety of base materials (through
density and morphology). The in-vivo and in-vitro studies on tritiated dust have shown the dependence of
the tritium clearance and retention in the human body on their physico-chemical parameters. Tritium
absorption in the lungs from tritiated dust ranges from absorption type S (slow) to type M (moderate)
according to the International Commission on Radiological Protection (ICRP) classification, whereas HTO
and HT are classified as F (fast).
Study on recycling of fusion activated material
The study was devoted to the Power Plant Conceptual Study (PPCS) Model AB, based on the HCLL
breeder blanket concept using EUROFER as structural material and Pb-17Li as breeder material, neutron
multiplier and tritium carrier [A3.42]. For each main component the categorisation for two decay times (50
and 100 years) has been provided according to the following classification:
• Clearable, with clearance index CI < 1 (CI from IAEA-TECDOC-855).
• Specific activity SH”).
Comparing the categorisation results, given in terms of mass or volumes, there is a large increase in the
fraction that could be cleared and recycled without major complications, allowing 100 years of decay.
Passing from 50 to 100 years, there is a large transfer of material (~60% of the total mass) from class
“>SH” to class “SH”. This suggests that it would be convenient to extend the decay period up to 100
years. Furthermore, the extra decay period up to 100 years could be limited to SH and >SH categories,
as the overall inventory of clearable plus material with specific activity SH inventory of 65% in mass at 100
years of decay. Considering all the activated materials generated from
decommissioning and from operation, a conservative approach for their
management, based on clearance/recycling of lifetime components only,
Clearable
< 1000 Bq/g
SH
>SH
43.0%
32.8%
5.0%
19.2%
66.4%
9.4%
24.2%
0.0%
would lead to 29% mass cleared/recycled and the rest (71%) disposed of. If one takes into account the
scenario with LiPb reuse, and an effective recycling capability of ~50% of >SH and SH categories, the
cleared/recycled mass fraction would be ~69%, while the remainder should be disposed of.
[A3.38] C. Rizzello and L. Di Pace, Proposal of an atmosphere detritiation system for the ITER plant, ENEA Internal Report FUS-TN-SA-SE-R-
151 (2006)
[A3.39] A. Natalizio and M.T. Porfiri, JET radiation exposure analysis. Data relating to the years 1988-2005, ENEA Internal Report FUS-TN-SA-
SE-R-157 (2006)
[A3.40] G. Cambi and T. Pinna, JET data collection on component malfunctions and failures of ion cyclotron resonant heating ICRH system,
ENEA Internal Report FUS-TN SA-SE-R-143 (2006)
[A3.41] L. Di Pace, Literature study on in vivo experiments with tritiated dust, ENEA Internal Report FUS-TN-SA-SE-R-144 (2006)
[A3.42] L. Di Pace, Definition of components and materials involved in clearance and recycling for PPCS plant model AB, ENEA Internal Report
FUS-TN-SA-SE-R-153, TW5-TSW-001/ENEA/D1 (Rev. 1) (2006)
References
61
Progress Report 2006

A4 Superconductivity
A Fusion Programme
In 2006 activities were focussed basically on the ITER project and related tasks, as well as on some
important non-ITER tasks. In particular, work began on an important goal of the ITER parallel programme,
the so-called Broader Approach (BA), consisting of the design and construction of the NbTi toroidal field
coils of the new Japanese tokamak JT-60SA. Due to the complexity of the subject, the work is carried out
in close collaboration with other ENEA groups, and within an international framework including the French
and, of course, Japanese teams.
In the framework of an EFDA assignment, ENEA has been charged with following the construction of the
new European dipole conductor, a new test facility for ITER full-size samples. In this framework, the group
developed and patented a new type of joint between superconductive cables. ENEA is also in charge of
surveying the manufacturing of the new conductor samples for the toroidal field coils of ITER.
The activity related to high-temperature superconductors (HTSs) can be summarised as follows: 1) Metallic
textured substrate for YBe 2 Cu 3 O 7-x -coated conductors: texture and micro-structural evolution and control
of in Ni-5at.% W alloy and development of copper-based substrates, carried out in collaboration with the
Technical University of Cluj-Napoca (TUCN) Romania. 2) Chemical approach for YBCO film deposition by
the MOD-TFA technique and introduction of artificial pinning centres in YBCO films for critical current
improvement (in collaboration with TUCN and Roma Tre University). 3) Activities carried out in the framework
of the Frascati Laboratory of the National Institute of Physics (LNF-INFN) superconducting magnet
programs: i) magnetic characterisation of NbTi and Nb 3 Sn wires for the development of fast ramped
superconducting dipoles for the FAIR accelerators at Gesellschaft fu .. r Schwerionenforschung (GSI)
Darmstadt Germany, NTA_DISCORAP programme; ii) application of MgB 2 wires and tapes, MARIMBO
experiment. 4) Transport and thermal stability characterisation of commercially available HTS wires and
tapes, funded by the EFDA Technology Work Programme HTSPER task, carried out with the support of the
SuperMat National Research Council (CNR)-INFM Regional Laboratory facilities at Salerno Italy.
All these activities are leading ENEA toward deeper knowledge of superconducting-based magnet
technology.
A4.2 ITER and ITER-Related Activities
ITER toroidal field cable conductor
ENEA is a member of the international testing group for the ITER magnet R&D. At the end of 2005
the measurement campaigns started on the samples (toroidal field advanced strands [TFAS] 1
and 2), the first ITER-type full-size TF conductors, made with the recently developed “advanced”
Nb 3 Sn strands [A4.1, A4.2].
ENEA actively contributed to the definition of the testing programme for the conductors, attended
Progress Report 2006
62

the tests, which continued through the first half of 2006, and participated in the data analysis and
processing, in collaboration with the international testing group members.
In spite of the high-performance strands used in the cabling, the TFAS samples showed very unusual
behaviour, with very wide transitions and well before the expected current sharing temperature values.
Based on these results, EFDA promoted the fabrication of different toroidal field conductor prototypes
(TFPRO project) for testing the effect of mechanical stress on the Nb 3 Sn superconducting characteristics.
A total of four different cables was produced in 2006.
ENEA was assigned the task to supervise the Luvata (formerly OuktoKumpu) activities for conductor
manufacturing, under two different contracts (tasks TMSC-TFPRO-1298 and TMSC-LPTCON-
1525).signed with EFDA.
For the TFPRO task, two conductors were made with a bronze route Nb 3 Sn strand produced by EAS
Germany, while for the LPTCON task a second couple used a strand produced by Oxford Instruments
Superconducting Technologies (OST, England) with internal-tin technology (fig. A4.1).
Differently from the previous TF geometry, the four samples have mainly the same cable layout, based on
a starting triplet formed of two superconducting strands and one copper strand, but differing slightly in
twist pitch length and final cable diameter (i.e., different void fraction). This choice was made in order to
test the conductors under different mechanical stress conditions, to which the single strand is subjected
to at operating conditions. The four conductor samples were shipped to the Association Euratom-Swiss
Confederation Villigen [CRPP]) in November 2006 and are under test.
ENEA is also working on developing the functional dependence of cable stiffness as a function of
manufacturing parameters for the TF, through the use of computer codes based on finite elements models
(FEMs) and artificial neural networks (task TMSC-CABLST).
Fig. A4.1 – Cross section of the four cables ready for characterisation
[A4.1] P. Bruzzone et al., Test results of two ITER TF conductor short samples using high current density Nb 3 Sn strands, presented at the
Applied Superconductivity Conference - ASC (Seattle 2006)
[A4.2] R. Zanino et al., IEEE Trans. Appl. Supercond. 16-2, 886 (2006)
References
63
Progress Report 2006

A4 Superconductivity
Current redistribution study on ITER conductors
A Fusion Programme
Cable-in-conduit conductor (CICC) performance is affected by the current distribution among the
strands. To better understand this phenomenon and its implications, the results from the
experimental data taken on the NbTi BB-III sample, tested in the TOSKA facility at FZK, and on the
poloidal field insert sample tested in SULTAN (CRPP) in 2004 are being studied in collaboration with
the University of Udine [A4.3].
It has been shown that, during T cs measurements, a current re-distribution among the cable substage
bundles appears just before the conductor transition, i.e., well before any detectable voltage
development. Such a phenomenon, repeatable and depending on the overall transport current, has
been observed by Hall probe sets designed with ad-hoc sensitivity and geometry, to allow also the
reconstruction of the current distribution inside the CICC by means of the THELMA code.
EFDA dipole
As is well known, to reach the high field values requested for ITER operation, Nb 3 Sn
superconducting cables have to be used to wind the main magnets, the central solenoid and the
toroidal field coils. A fundamental step in the design and construction of these ITER magnets is to
test a lot of conductor samples in relevant operating conditions. The only facility available at the
moment in Europe for this purpose is SULTAN, which will not be able to withstand the huge duty
foreseen for ITER construction in the very near future. Thus the European Community decided to
build a new facility to share the test tasks with SULTAN.
The facility will be based on a wind and react (W&R) dipole magnet wound from the last generation
of Nb 3 Sn strands, the so-called “advanced strands”, and it will be the very first magnet based on
such strands, and also the first dipole ever made by using CICC. ENEA has been charged with
supplying the cable and following the manufacture of the entire amount of CICC for the dipole (task
TMSC-DIPCON-1316). A few meters of a prototype conductor were manufactured and tested
successfully in SULTAN in 2005. Unlike what was obtained in the first ITER TF full-size samples
made using the same kind of strands (TFAS samples), the performance of the conductors agrees
very well with expectations.
The activity in 2006 was carried out in close collaboration with EFDA and Luvata. During this period
the cable parameters were drawn, and a short dummy cable, made only of copper strands, was
produced in order to define the cabling process and allow Luvata to prepare the environment and
build the tools. The dummy consists of a short (50 m) copper cable, processed according to the
actual cable parameters, jacketed and compacted to the final dimensions. It was decided together
with EFDA to use a square cross section for this dummy cable, instead of the rectangular one used
in 2005.
A whole set of jacketed superconducting cables has been produced so far: two high-field units and
four low-field units, for a total length of about 600 m. Of the six cable lengths, the low-field units are
still uncompacted, while the two high-field units were compacted to the final dimensions and
delivered to BNG Industries Germany for further assembling tests.
Short lengths of each sample were prepared and sent to CRPP for characterisation. The results
obtained so far for square conductors are not encouraging, probably due to the different void
fraction and pressure on single strands during operation, so new rectangular cable samples are in
preparation (fig. A4.2), and additional tests are foreseen (task PITCON).
In the framework of the dipole design and construction, EFDA asked ENEA to develop a new type
of joint between CICCs. ENEA developed a new joint concept and designed and fabricated some
prototypes, whose test results showed a very low electrical resistance (< 1nΩ). The main
advantages of this new joint (ENEA patent) are the low room occupancy (only slightly higher than
the conductor size itself), the easy manufacturing procedure, and the low cost of realisation. The
Progress Report 2006
64

Fig. A4.2 – A short piece of the rectangular Nb 3 Sn
CICC for dipole application
ENEA joint was accepted by EFDA and used as the only type of
joint between all the different lengths of the dipole magnet.
Accompanying activities included providing mechanical analyses
that will support the engineering design and manufacturing phase of
the dipole procurement (contract TMS-EDDES4–1303, completed
in 2006). The cable jacket main deformation occurring during the compaction and winding phase, the peak
stress in the insulation, due to the large pressure excursion occurring during the magnet quench inside the
conduit, and the thermo-mechanical analysis of dipole assembly during the cool-down phase were all
evaluated. An experimental benchmark at the bending FEM analysis was also carried out. In the last part of
2006 ENEA undertook (contract TMSC-DICOMO-1480) to develop a code model that could help to investigate
the sensitivity of Nb 3 Sn superconducting properties to mechanical strain, which causes significant problems in
the accurate performance prediction of large multi-strand CICC. Empirical relationships between strain and
critical current have already been established, based on experimental measurements with known applied strain
fields. The problems arise in the prediction of the strands stress/strain state within a cable. A large CICC
includes hundreds of strands twisted with different pitches and in contact with each other and with the external
jacket. Moreover, individual strands exhibit non-linear average mechanical behaviour due to the plasticity of the
constituting components. At this point, the definition of a suitable numerical model for the mechanical analysis
of strands in cables is still an open problem. Simplified methodologies should be developed with the aim of
predicting the mechanical behaviour of cables and strands. The present activity deals with evaluating the strain
state of the strands in the dipole CICC during energization within a magnetic field.
In addition, EFDA charged ENEA with performing code simulations of the mechanical stress arising in the
dipole structure during cool down and in operating conditions. This work was successfully carried out, with
the help of L.T. Calcoli personnel, during the first half of 2006.
Barrel bending experiments
The EFDA task named “barrel bending experiments” (BARBEN) was completed during the first months of
2006. Its aim was to study the effect of a bending strain applied on relatively long lengths of Nb 3 Sn
“advanced strands”, initially inserted and compacted in stainless tubes before heat treatment.
The effect of a 0.5% peak bending strain on the performance of an internal tin strand developed by OST
for ITER was investigated. Comparison between the measured critical current data of the unbent samples
and the results computed by Durham’s scaling law showed that, for the analysed system, the differential
thermal contraction of stainless steel and superconducting strand corresponds to a –0.57%
pre–compression of Nb 3 Sn at 4.2 K. At 12 T the strand shows a performance decrease of about 10-20%
with the application of a 0.5% peak bending strain [A4.4].
A further activity outside the EFDA task itself concerned clarifying the influence of the twist pitch length on
the strand performance degradation, when submitted to bending strain. Hence similar experiments were
performed on Nb 3 Sn strands in which the superconducting filaments were not twisted.
Optimisation of NbTi strand for PF1/PF6 performance
The original strand specification for the high-field ITER PF coils (P1/P6) was based on the LHC strand and
was 2900A/mm 2 at 5 T and 4.2 K. Using the recommended scaling formula, this gave an acceptable
predicted critical current density at the P1/P6 critical conditions.
[A4.3] F. Bellina et al., IEEE Trans. Appl. Supercond. 16-2, 1798 (2006)
[A4.4] L. Muzzi et al., Pure bending strain experiments on jacketed Nb 3 Sn strands for ITER, presented at the Applied Superconductivity
Conference - ASC (Seattle 2006)
References
65
Progress Report 2006

A4 Superconductivity
A Fusion Programme
However, it seems that NbTi strands have been optimised for low-temperature performance at the
expense of the Jc at higher temperatures. The scaling formula is essentially an envelope of the
maximum achieved current density at each field/temperature and appears to be un-representative
of these LHC strands at 6 T and 6.5 K. The difference in Jc (measured vs predicted) is substantial
and only some of the variation may be due to measurement errors, as small errors at temperatures
above 6 K produce a large change in critical current.
At the end of 2006, ENEA was charged with developing and producing at least 50 kg of Ni-plated
NbTi strand according to the ITER P1/P6 strand specification, with optimised current carrying
capabilities at higher temperatures and fields (TW6-TMSC-NbTi). The minimum required non-Cu Jc
at 6.5 K and 6 T is 200A/mm 2 . The optimisation processes for NbTi are quite well understood and
the performance at 6.5 K and 6 T can be improved by changing the process parameters during
production, e.g., adjustments to the intermediate annealing steps. The activity will be carried out
during 2007.
A4.3 JT-60SA
The Broader Approach is a project related to the ITER Accompanying Programme and involves
cooperation between Japan and the EU for the construction of a new tokamak machine in Japan,
JT-60SA. In Europe, CEA and ENEA have been assigned the specific tasks to design, construct and
test the 18 toroidal plasma confinement
magnets (fig. A4.3) made of NbTi strands. ENEA
is in charge of coordinating all the related EU
activities and consequently has been involved in
the conductor and coil design
In 2006, a preliminary assessment of the toroidal
200 400 400
magnet characteristics in regard to the
expected operative conditions was carried out.
400
ENEA is working on a consistent conceptual
400
600
600
design concerning the strand choice as well as
the conductor and coil layout definition. At the
moment, the definition of toroidal-coil design is
still being discussed among the members of the
joint project. Related to this activity, the ENEA
Fig. A4.3 – Magnetic system of the Japanese tokamak Frascati facility for testing and characterising
JT60SA: in red the 18 toroidal coils to be designed, superconducting strands at variable
built and tested in the EU
temperatures (4.2 K – 20 K) and magnetic fields
(up to 12 T) has been considerably upgraded in
terms of accuracy, repeatability and signal-to-noise ratio, becoming one of the most reliable and
versatile among the few available in Europe for this kind of characterisation. It has allowed an
extended campaign of NbTi strand characterisation, focussed on the foreseen operative conditions
of JT-60SA.
A4.4 High–Temperature Superconducting Materials
Evolution and control of cube texture in Ni-W substrates for YBCO-coated
conductors
The realisation of high critical current density YBe 2 Cu 3 O 7-x -based coated conductors with the
rolling-assisted biaxially textured substrate (RABiTS) approach is primarily related to the sharpness
Progress Report 2006
66

of cube texture developed in the substrate.
Among the various Ni–based tapes proposed,
Ni-W alloys have attracted particular interest
because of the enhanced mechanical
properties and reduced magnetism with
respect to pure Ni, and the sharp and almost
pure cube texture that can be obtained after
recrystallization of cold rolled tapes.
a) 600°C b) 700°C c) 800°C
Fig. A4.4 – (111) pole figures for three Ni-W samples annealed at
a) 600, b) 700 and c) 800°C and quenched to room temperature
The texture of highly (>90%) deformed fcc
metals with a medium-high stacking fault energy (SFE) is concentrated in the so-called β-fibre, known as
the stable end position of lattice rotations occurring during cold rolling. The development of cube texture
through recrystallization is directly related to the deformation texture, as the stronger the β-fibre the sharper
the cube texture in the (111) pole figure. During annealing, the deformation texture evolves into the cube
texture. Annealing up to 600°C does not affect the deformation texture in Ni 5 at% W (Ni–W) tapes, as no
orientation difference with respect to the as-rolled samples can be detected (fig. A4.4).
Conversely, at 700°C a structural modification appears, with the coexistence of cube and deformation
textures, since four symmetric poles, at tilt angle χ=54.7° and azimuthal angles ϕ=45°, 135°, 225° and
315°, are superimposed on the pre-existent texture in the (111) pole figures. Finally, for temperatures higher
than 800°C the sample is cube oriented and no residual deformation texture is detectable; the only
identifiable poles other than cube are due to {221}, namely cube twins, which are intrinsically related
to the recrystallization of cube grains. However, indications of microstructural modifications already at
600°C are revealed by a Monte Carlo procedure on
θ–2θ x-ray diffraction peaks, since an evaluation of the
microstrain contribution to peak broadening is provided.
No remarkable modification is produced up to 500°C,
while above this temperature a decrease in microstrain
is evident, indicating a relaxing of the lattice defects by
the decrease in the dislocation density, i.e., the material
underwent the recovery phase (fig. A4.5). In fact, during
this stage, part of the energy stored during deformation
4×10 -3
3×10 -3
2×10 -3
400
300
200
annihilation and subgrain formation, leading to
1×10 -3 100
modification of several physical properties, such as
0 200 400 600 800 1000
is released through dislocation rearrangement/
hardness and electrical conductivity, without affecting
Annealing temperature (°C)
lattice orientation. Further microstrain reduction above
700°C is due to the growth of strain-free oriented
grains, namely recrystallization, which is complete
above 800°C.
Fig. A4.5 – Evolution of microstrain and Vickers
hardness for Ni-W samples annealed at different
temperatures and quenched to room temperature
After complete recrystallization the tapes may exhibit, to the naked eye, a more or less opalescent surface.
This feature is the result of the diffusion of light coming from pronounced grain boundaries, which are
normally high-angle boundaries, i.e., with a relative misorientation greater than about 15°. This is the case
of cube twins, often arranged in longitudinal bands. It was shown that the formation of components other
than cubes is related to the grain size of the bulk material before cold rolling (initial GS) (fig. A4.6). In
particular, the area of cube orientation decreases because of the increase both in cube twins and in noncube
orientation as the initial GS becomes larger. These data indicate that twin formation is related to both
the SFE and the deformed state (fig. A4.7, A4.8). The resulting direct relation between cube twins and
non-cube area densities suggests that their formation is controlled by a common parameter. This
correlation is supported by data from several samples of Ni-V, Ni-Cr and Ni-W. In particular, large non-cube
grain fractions correspond to samples subjected to a deformation degree below 95%, in which the few
cube grains were almost invariably twinned.
Both in- and out-of-plane distributions of the cube orientation measured by x ray are in agreement with
electron backscattering diffraction (EBSD) analysis in terms of cube texture coarsening, as an increase in
Microstrain
Vickers hardness HV 200
67
Progress Report 2006

A4 Superconductivity
a) b)
6
cube twin
non–cube
A Fusion Programme
100 μm
0 10 20 30 40 50 60
Deviation from {001} (°)
FWHM (degree)
9
7
Δω(RD)
Δω(TD)
Δφ
5
10 30 50 70
Initial grain size (μm)
Fig. A4.8 – FWHM of (200) rocking curves, along
both rolling (empty triangles) and transverse
directions (full triangles), and of (111) φ-scans
100 μm
0 10 20 30 40 50 60
Deviation from {001} (°)
Fig. A4.6 – EBSD misorientation maps for Ni-W samples with initial GS of
19 µm a) and 63 µm b)
the spread around the {001} ideal orientation for
larger initial GS is observed. As a consequence of this
behaviour, a broader distribution of the cube texture is
observed in substrates with reduced cube area fraction.
This kind of relationship seems to be a general feature
since it has been observed in several Ni-based substrates.
This result is of great conceptual and practical importance
because hindering cube twin formation not only provides
larger cube areas, but leads to sharper cube textures as
well [A4.5].
Nickel-copper alloys as textured substrates
for YBCO–coated conductors
(empty circles) for Ni-W samples with different
Ni-Cu-Co alloy tapes with different relative concentrations
initial GS
were studied as textured substrates for YBCO-coated
conductor application. A small amount of cobalt was
added in order to enhance the oxidation resistance of Ni-Cu alloy. 100-μm-thick tapes were
obtained through conventional cold rolling to a deformation degree of 97% followed by recrystalliza -
tion at high temperature. The use of different thermal treatments made it possible to obtain area
densities of cube orientation as high as 95% (figs. A4.9, A4.10). The substrate was thoroughly
characterised by means of x-ray diffraction, EBSD and scanning electron microscopy (SEM)
analyses. Electrical resistivity, mechanical properties and oxidation resistance of this substrate will
be compared with those exhibited by Ni, Ni-W and Ni-Cu tapes.
Area fraction (%)
4
2
0
10 30 50 70
Initial grain size (μm)
Fig. A4.7 – Non-cube and cube twin area density
drawn from EBSD measurements for Ni-W
samples with different initial GS
A Pd transient layer was epitaxially grown prior to depositing
conventional CeO 2 /YSZ/CeO 2 buffer layer architecture in
order to passivate the Ni-Cu-Co substrate. The deposition
conditions for the Pd layer were optimised in order to obtain
a particularly sharp out-of-plane orientation, so that the full
width at half maximum (FWHM) of the rocking curves in the
transverse direction (TD) through the (002) reflection drops
RD
TD
Fig. A4.9 – EBSD map for a recrystallized Ni-Cu-Co alloy substrate.
Red, green and blue colours refer to {100}, {110} and {111} planes
Progress Report 2006
68

{1,1,1}
32
RD
Fig. A4.10 – (111) pole figure obtained from EBSD
data for a recrystallized Ni-Cu-Co sample
16
8
TD
4
2
1
0.5
0.13
from about 9° of Ni-Cu-Co to 2.1° of Pd
layer; whereas in the rolling direction (RD)
these values attain about 6 and 1.7°,
respectively. This sharp texture is preserved
and both CeO 2 and YSZ films exhibit the
same out-of plane orientation (fig. A4.11).
The encouraging structural properties of the
buffer layer architecture obtained indicate
that this alloy is a promising alternative
substrate for the realisation of
YBCO–coated conductors.
Intensity (arb. units)
7×10 4
5×10 4
3×10 4
1×10 4
TD NiCuCo-Pd 1
TD NiCuCo-Pd 2
RD NiCuCo-Pd 1
RD NiCuCo-Pd 2
10
10 15 20 25 30 35
θ (degree)
Fig. A4.11 – Rocking curves around (002) reflection of Pd films grown at
different temperatures on Ni-Cu-Co substrate. A consistent sharpening
of the out-of-plane orientation is attained for higher deposition
temperatures
MOD-TFA YBCO films
It has been demonstrated that metal-organic deposition (MOD) using trifluoroacetate (TFA) precursors is
the most suitable for epitaxial YBCO deposition. In the MOD-TFA method, a fluorine containing coating
solution decomposes to fluorides which, in turn, undergo different chemical reactions during the hightemperature
firing process (700 – 800°C) in controlled atmosphere to convert to oxides.
The precursor solutions for YBCO were prepared by sonicating the mixture of Y, Ba and Cu acetates in a
1:2:3 cation ratio with a stoichiometric quantity of trifluoroacetic acid in de-ionized water. The resulting
solution was slowly dried at low temperature to form a glassy blue resin. The precursor solution was
deposited both on (00l)-oriented SrTiO 3 single crystals and on Ni-W/Pd/CeO 2 /YSZ/CeO 2 templates by
spin coating. The resulting gel films were treated in two heating stages to obtain the YBCO
superconducting films. The YBCO films obtained under these conditions are about 250 nm thick.
The x-ray diffraction (XRD) pattern of θ–2θ scans for YBCO/CeO 2 /YSZ/CeO 2 /Pd/Ni-W exhibits only the
(00l) YBCO peaks. No (h00) reflections due to a-axis oriented grains were observed. The presence of the
(111) reflection of YSZ and CeO 2 indicates a small fraction of (111) oriented grains in these films. The (002)
to (111) peak intensity ratio is of about 10 2 . The rocking curve through the (002)Ni-W, (002)YSZ, (002)CeO 2
and (005)YBCO peaks have an out-of-plane FWHM of 8.8°, 4.2°, 3.8° and 3.4°, respectively. The small
values of FWHM for the YSZ and CeO 2 with respect to the Ni-W substrate is correlated to the Pd film. The
in-plane crystallographic relationship of the structure is [100]YBCO||[110]CeO 2 ||[110]YSZ||[100]Ni-W.
The surface of YBCO/CeO 2 /YSZ/CeO 2 /Pd/Ni-W films is free of cracks but has some holes. In spite of the
voids, the c-axis oriented grains are well connected. Furthermore, YBCO grains are connected over pores.
[A4.5] A. Vannozzi et al., Supercond. Sci. Technol. 19, 1240-1245 (2006)
References
69
Progress Report 2006

A4 Superconductivity
A Fusion Programme
Intensity (counts/s)
100 nm
Fig. A4.12 – Film surface of YBCO TFA grown on
CeO 2 /YSZ/CeO 2 /Pd/Ni-W
2.0×10 5
1.5×10 5
1.0×10 5
0.5×10 5
BZO(100)
19 20 21 22 23 24
(100)
(200)
STO (100)
(400)
(500)
BZO(200)
40 42 44 46 48
STO (200)
Introduction of artificial pinning sites in YBCO films
(700)
The spherical particulates are
nanocrystallites of CuO
(fig. A4.12). The high quality
of the YBCO films is
confirmed by the T c values
(90.9 K) and the reduced
transition widths (ΔT∼1.5 K)
(fig. A4.13).
J c values as high as
1 MA/cm 2 are reported at
84 K, reaching 2.7 MA/cm 2
at 77 K for YBCO TFA films
deposited on SrTiO 3 single
crystals. The development of
MOD-TFA YBCO films on
long length CeO 2 /YSZ/
CeO 2 /Pd/Ni-W template is in
progress [A4.6].
One of the most effective ways to improve the pinning efficiency of magnetic flux vortices in YBCO
films is the introduction of epitaxial second–phase nanoinclusions in the YBCO matrix. This
technique has gained relevant interest due to the possibility of increasing the irreversibility field (H irr ),
which limits high magnetic field performance.
This goal has been pursued by growing YBCO thin films with the pulsed laser deposi tion (PLD)
method from com posite targets obtain ed by adding BaZrO 3 (BZO) powder in molar percents
ranging from 2.5 to 7%. The presence of BZO epitaxial inclusions inside the films has been checked
by XRD analysis (fig. A4.14).
As already reported in the literature, the introduction of second-phase nanoinclusions progressively
lowers the critical temperature T c of YBCO thin films (fig. A4.15).
Analysis of the transport properties shows the improvement of pinning efficiency in YBCO films with
BZO inclusions. Self-field critical current densities are increased by BZO addition, ranging from 1.23
MA/cm 2 for pure YBCO film to 2.22 MA/cm 2 recorded for 2.5 mol.% BZO-YBCO film. All the BZO
added films exhibit increased critical current densities in the whole magnetic field range inspected
and higher irreversibility field values, with the 5 mol.% BZO-YBCO film the best in field performances
(fig. A4.16a)). The improvement in the transport properties in BZO samples can be ascribed to the
introduction of extended defects elongated along the YBCO c-axis, as shown by a prominent peak
Resistance (Ω)
0
0 20 40 60 80 100 120
2θ (degree)
STO (300)
BZO(400)
20
10
8
6 T C =90.9 K
4
2
0
80 90 100
0
0 100 200 300
Temperature (K)
Fig. A4.13 – R(T) plot for YBCO TFA grown on
CeO 2 /YSZ/CeO 2 /Pd/Ni-W
STO (400)
94 95 96 97 98
Fig. A4.14 – X-ray θ-2θ diffraction spectrum showing the presence of
BaZrO 3 epitaxial second phase inside the YBCO matrix
Progress Report 2006
70

Normalised resistance
1.2
0.8
0.4
YBCO-STO
YBCO-BZO (2.5%)-STO
YBCO-BZO (5%)-STO
YBCO-BZO (7%)-STO
0
80 85 90 95 100
Temperature (K)
a)
Critical temperature (K)
90
88
86 YBCO-STO
YBCO-BZO (2.5%)-STO
YBCO-BZO (5%)-STO
YBCO-BZO (7%)-STO
84
0 2 4 6 8
BaZrO 3 nominal concentration (vol.%)
b)
Fig. A4.15 – Normalised
resistance as a function of the
temperature for YBCO films
with BZO molar concentration
ranging from 2.5 to 7% a).
Dependence of the critical
temperature T c on the BZO
molar concentration b)
at 0° (magnetic field parallel to the c-axis) in the critical current
density dependence on the angle between the magnetic field
direction and the direction normal to the film (fig. A4.16b)).
Measurements of the microwave complex resistivity in the mixed
state were carried out for films deposited on SrTiO 3 and sapphire
single crystal substrates. The pinning frequency ν p , which
represents a measure of the steepness of the potential well for the
flux lines, can be estimated from complex resistivity. Very high
values of about 50 GHz are attained between 60 and 80 K,
indicating extremely high vortex pinning and steep potential wells.
As expected, the ν p rapidly drops to zero as T approaches T c . It can
be concluded that the intragrain vortex pinning at high microwave
frequencies in YBCO films with BZO inclusion of nanometric size
has been greatly improved with respect to films free of BZO
inclusions.
Magnetic characterisation of superconducting wires
for fast ramped superconducting dipoles
The INFN Dipoli Super Conduttori Rapidamente Pulsati
2×10 5
(DISCORAP) programme originates from the new requirement of
μ 0 H=3 T
developing fast-ramped superconducting dipoles for the FAIR
μ 0 H=5 T
accelerators at GSI, Darmstadt, Germany. It is a four-year program
0
to develop a fully working bent dipole 3.8 m long in its horizontal
-100 -50 0 50 100
cryostat. The dipole has to generate a field of 4.5 T with a ramping
rate of 1 T/s.
Angle (degree)
Fig. A4.16 – a) Critical current density as a
function of the applied magnetic field at T=77K
Magnetic measurements in high magnetic field were carried out to
for YBCO films with BZO content ranging from
extract information about the intrinsic magnetization losses, critical
2.5 to 7 mol.%. b) Dependence of the critical
current, and filament size. Dissipation, when the magnetic field is
current density on the angle between the
rapidly changing, comes from the filament couplings, which are
magnetic field direction and the direction normal
connected through the metallic matrix. The magnetization M of
to the film at T=77K recorded for the 7 mol.%
NbTi and Nb 3 Sn wires was analysed with a vibrating sample
BZO-YBCO sample
magnetometer (VSM) operat ing in the range [300 - 4] K under a
maximum field up to 12 T. All the tests were carried out in the zero field cooling (ZFC) situation to be able
to record the purely diamagnetic response at low magnetic field, useful for studying the shielding regimes.
Figure A4.17 shows magnetic measurements for a 2-μm filament prototype NbTi wire.
[A4.6] A. Rufoloni et al., J. Phys.: Conf. Ser. 43, 199 (2006)
Critical current density (A/cm 2 )
Critical current density (A/cm 2 )
10 5
10 3
T=77K a)
YBCO-STO
BZO (F2.5%)-STO
BZO (F5%)-STO
10 1 BZO (F7%)-STO
0 4 8
Magnetic induction (T)
6×10 5
4×10 5
YBCO-BZO(F7%)-3 77K
μ 0 H=100 mT
μ 0 H=1 T
b)
References
71
Progress Report 2006

A4 Superconductivity
A Fusion Programme
Magnetic moment (emu)
0
-0.0005
50
-0.001
transverse field
parallel field
0
-0.0015
SL8979S
sample A-I=5.57 mm
-0.002
-50
4 8 12 16
T(K)
transverse
field 4.5 K
Fig. A4.17 – Magnetic moment for low filament size NbTi wire (2006)
MARIMBO experiment:
application of MgB 2
Activities regarding MgB 2
superconductors have been
devoted to fundamental
aspects, such as the influence of
the disorder introduced by
-6 -4 -2 0 2 4 6 neutron irradiation on
B(T)
polycrystalline MgB 2 material
and the MgB 2 phase nucleation
by means of MgB 2 /Mg multilayers,
and to more applicative features such as
the stability properties of a MgB 2 multifilamentary
tape.
10 5
The magnetic properties of polycrystalline MgB 2
10 5
10 4
J @T=5 K, B=4 T
0
exposed to different neutron fluencies were
P2
0 10 17 10 18 10 19
Fluence (cm 2 ) analysed to perform in-depth analysis of the
P3.5 critical field and current density behaviour and to
P3.7
identify what scattering and pinning mechanisms
P5
P4
P0
come into play (fig. A4.18).
10 4 P3
P6
P1
In the second study the chemical composition
3 6 9
μ
and electronic structure of the multi-layer films
o H(Tesla)
were analysed and compared with the
Fig. A4.18 – Critical currents measured in samples
corresponding MgB 2 bulk case to investigate the
after different neutron doses (P)
reasons for the low transition temperature typical
of low-temperature processed MgB 2 films. Short
straight samples of the Cu-stabilised, 14-filament MgB 2 tape, taken from a 1.6–km-length
production manufactured by Columbus Superconductors, Genoa, were used to produce a cryogenfree,
double pancake style, magnet. The conductor is a 3.6-mm-wide and 0.65-mm-thick tape,
fabricated with the powder-in-tube (PIT) method. The tape is composed of a copper inner region
delimited by an iron sheath and a nitrogen Niouter matrix where MgB 2 filaments are embedded. The
superconducting fraction is less than 10% of the whole section. The tape edges were welded over
3 cm on the bulk copper sample holder used for the tests, which were performed in a He gas flow
cryostat. Two brass counter flow cooled current leads, designed for 200 A, were used to bias the
tape. The wire was shielded by thick polystyrene from direct exposure to the cold gas. A 3-mm-wide
heater, made of NiCr wire, was wound and glued in the middle of the tape. Voltage contacts at
known positions were used to determine the presence of dissipative regimes. A calibrated cernox
thermometer was located on the tape, a few mm from the heater side. Figure A4.19a) reports the
heat propagation velocity ν p as a function of the delivered energy E at two temperatures and bias
currents, while figure A4.19b) reports ν p as a function of the temperature at two values of bias
current, each one triggered by a constant energy pulse.
J c (A/cm 2 )
V p (mm/s)
T= 5 K
J c
(A/cm 2 )
The ν p -vs.-energy curve indicates a fast increase in ν p at low heater energy followed by a weak
dependence for higher energy values. This ν p (E) behaviour at low E values may be ascribed to the
100
60
150
a) b)
μ 0 H=0 T
20K 200A
30K 100A
20
30
0 0.5 1 1.5 2 2.5 15 20 25 30
Heater energy (J)
Temperature (K)
90
200A 2.25J
150A 0.56J
μ 0 H=0 T
Fig. A4.19 – Normal zone
propagation velocity as a
function of a) heater energy and
b) temperature
Progress Report 2006
72

short distance between voltage taps and heater, where the equilibrium balance between the heat loss by
conduction and the generated heat is not yet achieved. The ν p increases with the temperature because
dissipation increases, narrowing the temperature margin T g -T 0 , where T 0 and T g are the operating the
generation temperature, respectively.
Transport and thermal stability characterisation of HTS wires and tapes: analysis of
quench propagation on YBCO-coated conductors
A 45-cm–long AMSC 344 conductor sample
was used, with a Ni-Cr resistive wire wound in
the middle of the tape as heat source. The
quench propagation was monitored by 12
voltage taps distributed along the sample
length. Figure A4.20 reports details of the
electrical connections on the tape. The
distance between voltage taps is 1 cm and the
total active length (distance between V+ and V)
is 32 cm. The Ni–Cr heater is in the region
delimited by ch0 voltage taps. Figure A4.21
shows a typical result for a set of
measurements with increasing energy at
T=80 K and I bias =35 A. Only ch0 and ch1
values are plotted for clarity. The energy was
varied by increasing the current, but keeping
the pulse duration at 0.1 s. Up to 0.36 J a sharp
increase in the ch0 voltage was revealed in
correspondence to the current pulse (t=0 s) and
then recovered after a few seconds. No other
significant variations in the voltage readings
were observed. For higher energy, propagation
sets up as revealed by the increase of the ch1
5 cm 6 cm 2.6 cm 2.5 cm 6 cm 5 cm
| | | | | | | | | | | |
-|--------|---|-------|---|----|---|----|---|-------|---|--------|-
V+ ch2 ch1 ch0 ch3 ch4 V-
Fig. A4.20 – Distribution and distance of voltage taps along the
active region of the tape
voltage. As expected, the process becomes faster as the energy increases. Heat propagation is stopped
when the I bias is switched off (sharp drops of both ch0 and ch1 marked by arrows in fig. A4.21).
Voltage (mV)
7×10 -3
5×10 -3
3×10 -3
T=80 K; l bias =35A (66% l c )
l bias off
1×10 -3 0
0 5 10 15 20 25
Time (s)
run 21 ch0
run 21 ch1
run 22 ch0
run 22 ch1
run 23 ch0
run 23 ch1
run 24 ch0
run 24 ch1
run 25 ch0
run 25 ch1
run 26 ch0
run 26 ch1
Heat-generation experiments were carried out at 75 and 80 K for different values of I bias . The heat
propagation velocity evaluated from ch1 as a function of the energy for both 75 and 80 K is reported in
figure A4.22a) and b). As can be seen, V 1 increases with I bias . It should be noted that the propagation
process in this tape is about two orders of magnitude slower than in typical NbTi multifilamentary wire and
one slower than in MgB 2 tape.
0.25J
0.36J
0.42J
0.49J
0.64J
0.81J
Fig. A4.21 – Time evolution of voltage along the tape
Signal velocity V 1 (m/s)
2.0×10 -2
1.5×10 -2
1.0×10 -2
T=75K
a)
53% l c
58% l c
64% l c
70% l c
76% l c
82% l c
85% l c
0.5×10 -2 0
0 0.2 0.4 0.6 0.8 1 1.2
Heater energy (J)
Signal velocity V 1 (m/s)
2.0×10 -2
1.5×10 -2
1.0×10 -2
0.5×10 -2 0
T=80K
b)
57%l c
66%l c
86%l c
0 0.2 0.4 0.6 0.8 1 1.2
Heater energy (J)
Fig. A4.22 – Normal zone propagation velocity in the HTS tape at two different temperatures :a) 75 K and b) 80 K
73
Progress Report 2006

A5 Inertial Fusion
A Fusion Programme
The Fast Ion Generation Experiment (FIGEX) proposed and designed by the ENEA Inertial Physics
and Technology Group [A5.1, A5.2] was performed at the Petawatt Facility of the Rutherford
Appleton Laboratory (UK) during the first two months of 2006. The aim was to study a possible way
to create by short laser pulses an ion source for inertial fusion energy (IFE) application. The Frascati
ABC facility was used to determine the required pre-pulse contrast in the experiment [A5.3]. Analysis
of the experimental results was mostly carried out at Frascati ENEA. A large fraction of the activity
had to be devoted to preparing within a few months a software package for the ion spectrometer
data processing. Preliminary results were available for an invited presentation at the European
Conference on Laser Interaction with Matter held in Madrid in June 2006.
Since FIGEX was designed for fast–ion generation (MeV/nucleon), a set of Thomson ion
spectrometers was used as the key diagnostic. The detectors were plastic CR39 plates where each
ion was registered as a pit. Ions on CR39 were registered along parabolas (one for each Z/A, where
Z and A are the ion charge and mass numbers):
V z 2
H 2 x
V2
H 2 ( z x )2
(A5.1)
(A5.2)
where (x, z) are the intrinsic coordinates taken with the origin in the point where ions with infinite
energy would impinge and with the x-axis parallel to the magnetic field H; V and E nucl are the voltage
applied to the plates and the energy per nucleon of the ion registered at the site (x, z). Equation A5.1
represents in the plane (x, z) a parabola with the vertex at x=z=0 and the axis parallel to x, whereas
A5.2 associates the specific energy to the coordinates (x,z) where the ion impinges.
In analysing the experimental data, to find the intrinsic position of the origin and the direction of the
axes it was sometimes useful to represent the pit positions in the plane E nucl ,Z/A) where parabolas
become straight lines parallel to the E nucl - axis (see an example in fig. A5.1).
A microscope driven by step motors was used to detect the position of the pits imprinted on the
CR39. The software associated with the equipment generated information about several features of
each pit, including their position with respect to a Cartesian coordinate system (u, v). These data
were released as a text file for each CR39 plate.
Rather complex software based on the Mathematica package was worked out and installed on a
laptop computer that makes the utility transportable when needed. The software was designed in
order to 1) find the intrinsic coordinate system where eqs. A5.1 and A5.2 hold; 2) recognise the Z/A
corresponding to each parabola; 3) count the number of pits registered on each parabola; and 4)
Progress Report 2006
74

Fig. A5.1 – Example of ion registration in the intrinsic plane and
in the (E nucl , Z/A) plane
(x, z) plane
associate to each pit the proper value of intrinsic
coordinates (x,z), the specific energy E nucl and the
distribution function in f( Enucl ) pertinent to each
parabola and the associated average energy values,
spread in energy, etc.
(E nucl , Z/A) plane
The code was designed to accomplish tasks 1, 2, 3,
4 and to perform ionic distribution functions, global
calculations relative to the complete set of
spectrometers (6) aligned along different directions
around the target and to study the angular
distributions for number and energy (fig. A5.2).
This programme was worked out as follows: First of
all the Z/A expected in the experiment (expected
contaminants included) were evaluated and the corresponding parabolas evaluated by eq. A5.1. Then the
intrinsic coordinates were found by superposing (by electronic handling) the experimental parabolas on the
theoretical ones given by eq. A5.1 as in figure A5.3. Figure A5.4 reports an example of species recognition
based on this method.
2
cosθ
- view
LARGE
θ
- view
1.5
1
θ
0
0.5
0
-1 -0.5 0 0.5
cosθ
1
Laser beam
Fig. A5.2 – Example of angular distribution calculations
Target surface
[A5.1] A. Caruso and C. Strangio, Laser Part. Beams 19, 295 (2001)
[A5.2] C. Strangio and A. Caruso, Laser Part. Beams 23, 33 (2005)
[A5.3] C. Strangio et al., A study for target modification induced by the prepulse in petawatt-class light-matter interaction experiments,
presented at the 28 th ECLIM Proceedings (2004)
References
75
Progress Report 2006

A5 Inertial Fusion
A Fusion Programme
Fig. A5.3 – The intrinsic coordinates are found by superposing the theoretical parabolas on the experimental. a) Initial
relative positions of the theoretical and experimental patterns. b) The two patterns have been superposed by electronic
handling and the species are identified
O8
C6
Si14
H N7
Si12
Si13 O7 N6
Si10
C5 Si11 O6 N5 C4 Si9 O5
a) b)
Si8
N4
Si7
C3
O4
Si6
N3
O3
Si5
C2
Si4
N2
O2
C1
Si3
Si2
N1
O1
Si1
Fig. A5.4 – Intrinsic coordinates and ion recognition by
the method of superposing the theoretical curves on the
experimental pattern. Green labels represent missing
elements
To count the pits for each Z/A it was necessary
to isolate the corresponding parabola from the
others and from the background. The code
performed this task by taking a strip around
each parabola having width assigned ad hoc
as input. The coordinates of the pits in each
parabola were recorded in a file and used to
evaluate the associated specific energy E nucl
through eq. A5.2. From the files the distribution
function of specific energy for each species
was evaluated (see an example in fig. A5.5).
Once the files pertinent to the number and
energy distribution for each Z/A are known all
the calculations relative to the ion charge
distribution functions are possible, with the
limits determined by the ion species mixing in
each Z/A (fig. A5.6).
Progress Report 2006
76

800
600
400
200
0
60
40
20
200
150
100
50
0
Si8-N4
175
Si9
150
800
C4
600
100
400
50
200
0
0
0.5 1 1.5 2 2.5 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 0.5 1 1.5 2 2.5 3
MeV/Nucl MeV/Nucl MeV/Nucl
Si10-N5
0
0.4 0.6 0.8 1 1.2 1.4
MeV/Nucl
20
15
10
5
C5
0.5 1 1.5 2 2.5
MeV/Nucl
Si14-N7-C6-O8
0
0.4 0.5 0.6 0.7 0.8 0.9
MeV/Nucl
140
O6
120
80
40
0
0.2 0.4 0.6 0.8 1 1.2 1.4
MeV/Nucl
30
20
10
Si12-N6
0
0.6 0.8 1 1.2 1.4
MeV/Nucl
30
20
10
0
Si11
0.5 0.6 0.7 0.8 0.9 1 1.1
MeV/Nucl
40
Si13
30
20
10
0
0.4 0.6 0.8 1 1.2 1.4
MeV/Nucl
Fig. A5.5 – Example of ion distribution function
calculations for a Thomson ion spectrometer
Fig. A5.6 – Example of ionization distribution functions
Analysis of the experiment is expected to be completed
within the first months of 2007 and part of the results
will be available for presentation as an invited talk at the
7 th Symposium on Current Trends in International
Fusion Research: A Review (5-9 March 2007,
Washington, DC, U.S.A.).
0.325
0.195
0.065
Si
N
2 6 10 14
Z
77
Progress Report 2006

A6 Publications, Patents and Events
A Fusion Programme
Articles
A6.1 Publications
G. VLAD, S. BRIGUGLIO, G. FOGACCIA, F. ZONCA, M. SCHNEIDER: Alfvénic instabilities driven by fusion
generated alpha particles in ITER scenarios
Nucl. Fusion 46, 1-16 (2006)
G. CARUSO, H.W. BARTELS, M. ISELI, R. MEYDER, S. NORDLINDER, V. PASLER, M.T. PORFIRI:
Simulation of cryogenic He spills as basis for planning of experimental campaign in the EVITA facility
Nucl. Fusion 46, 51-56 (2006)
A.A. TUCCILLO, F. CRISANTI, X. LITAUDON, YU.F. BARANOV, A. ECOULET, M. BECOULET, L. BERTALOT,
C.D. CHALLIS, R. CESARIO, M.R. DE BAAR, P.C. DE VRIES, B. ESPOSITO, D. FRIGIONE, L. GARZOTTI, E.
GIOVANNOZZI, C. GIROUD, G. GORINI, C. GORMEZANO, N.C. HAWKES, J. HOBIRK, F. IMBEAUX, E.
JOFFRIN, P.J. LOAS, J. MAILLOUX, P. MANTICA, M.J. MANTSINEN, D. MAZON, D. MOREAU, A. MURARI, V.
PERICOLI-RIDOLFINI, F. RIMINI, A.C.C. SIPS, O. TUDISCO, D. VAN EESTER, K.-D. ZASTROW AND JET-
EFDA WORK-PROGRAMME CONTRIBUTORS: Development on JET of advanced tokamak operation for ITER
Nucl. Fusion 46, 214-224 (2006)
F. SANTINI: Non-thermal fusion in a beam plasma system
Nucl. Fusion 46, 225-231 (2006)
P. BATISTONI, U. FISCHER, M. ANGELONI, P. BEM, I. KODELI, P. PERESLAVTSEV, L. PETRIZZI, M.
PILLON, K. SEIDEL, S. P. SIMAKOV, R. VILLARI: Neutronics design and supporting experimental activities
in the EU
Fusion Eng. Des. 81, 1169-1181 (2006)
M. ANGELONE, P. BATISTONI, M. LAUBENSTEIN, L. PETRIZZI, M. PILLON: Neutronics experiment for the
validation of activation properties of DEMO materials using real DT neutron spectrum at JET
Fusion Eng. Des. 81, 1485-1490 (2006)
M.T. PORFIRI, N. FORGIONE, S. PACI, A. RUFOLONI: Dust mobilization experiments in the context of the
fusion plants - STARDUST facility
Fusion Eng. Des. 81, 1353-1358 (2006)
T. PINNA, J. IZQUIERDO, M.T. PORFIRI, J. DIES: Fusion component failure rate database (ECFR-DB)
Fusion Eng. Des. 81, 1391-1395 (2006)
L. PETRIZZI, M. ANGELONE, P. BATISTONI, U. FISCHER, M. LOUGHLIN, R. VILLARI: Benchmarking of
Monte Carlo based shutdown dose rate calculations applied in fusion technology: from the past experience
a future proposal for JET 2005 operation
Fusion Eng. Des. 81, 1417-1423 (2006)
Progress Report 2006
78

M. ANGELONE, M. PILLON, A. BALDUCCI, M. MARINELLI, E. MILANI, M.E. MORGADA, G. PUCELLA,
A. TUCCIARONE, G. VERONA-RINATI, K. OCHIAI, T. NISHITANI: Radiation hardness of a polycrystalline chemicalvapor-deposited
diamond detector irradiated with 14 MeV neutrons
Rev. Sci. Instrum. 77, 023505/1-7 (2006)
C. CASTALDO, U. DE ANGELIS, V.N. TSYTOVICH: Screening and attraction of dust particles in plasmas
Phys. Rev. Letts 96, 075004/1-4 (2006)
S. TOSTI, L. BETTINALI, F. GIORDANO, E. SOLDANO, G. SCIOCCHETTI: A novel permeation method to measure
volumes
Measurement 39, 186-194 (2006)
S.E. SEGRE, V. ZANZA: Derivation of the pure Faraday and Cotton-Mouton effects when polarimetric effects in a
Tokamak are large
Plasma Phys. Control. Fusion 48, 339-351 (2006)
G. MICCICHÉ, G. COLLINA, L. MURO, B. RICCARDI: IFMIF repraceable backplate: remote handling activities,
rescue procedures and evaluation of a prototype reliability
Fusion Eng. Des. 81, 879-885 (2006)
M. SAMUELLI, L. RAPEZZI, M. ANGELONE, M. PILLON, M. RAPISARDA, S. VITULLI: Unconventional plasma
focus devices
IEEE Trans. Plasma Sci. 34, 1, 36-54 (2006)
A. BALDUCCI, M. MARINELLI, E. MILANI, M.E. MORGADA, G. PUCELLA, M. SCOCCIA, A. TUCCIARONE, G.
VERONA-RINATI, M. ANGELONE, M. PILLON, R.POTENZA, C. TUVÉ: Growth and characterization of single
crystal CVD diamond film based nuclear detectors
Diamond Rel. Mater. 15, 292-295 (2006)
F. ZONCA, L. CHEN: Resonant and non-resonant particle dynamics in Alfvén mode excitations
Plasma Phys. Control. Fusion 48, 537-556 (2006)
L. BERTALOT, B. ESPOSITO, Y. KASCHUCK, D. MAROCCO, M. RIVA, A. RIZZO, D. SKIPINTSEV: Fast digitizing
techniques applied to scintillation detectors
Nucl. Phys. B (Proc. Suppl.) 150, 78-81 (2006)
D. MAISONNIER, I. COOK, P. SARDAIN, L. BOCCACCINI, L. DI PACE, L. GIANCARLI, NORJAITRA PRACHAI, A.
PIZZUTO AND PPCS TEAM: DEMO and fusion power plant conceptual studies in Europe
Fusion Eng. Des. 81, 1123-1130 (2006)
M. ROMANELLI, F. BOMBARDA, C. BOURDELLE, M. DE BENEDETTI, B. ESPOSITO, D. FRIGIONE, C.
GORMEZANO, E. GIOVANNOZZI, G.T. HOANG, M. LEIGHEB, M. MARINUCCI, D. MAROCCO, C. MAZZOTTA, G.
REGNOLI, C. SOZZI, F. ZONCA: Confinement and turbulence study in the Frascati tokamak upgrade high field and
high density plasmas
Nucl. Fusion 46, 412-418 (2006)
P. BATISTONI: Il contributo italiano a ITER e al programma fusione
La Termotecnica, Anno LX, 5, 32-34 (2006)
S. TOSTI, A. BASILE, F. BORGOGNONI, L. BETTINALI, C. RIZZELLO: Pd membrane reactor design
Desalination 200, 676-678 (2006)
S. TOSTI, L. BETTINALI: Volumes measurement by means of membranes
Desalination 200, 140-141 (2006)
P. BURATTI, B. ALPER, S.V. ANNIBALDI, A. BECOULET, P. BELO, J. BUCALOSSI, M. DE BAAR, P. DE VRIES,
79
Progress Report 2006

A6 Publications, Patents and Events
D. FRIGIONE, C. GOMERZANO, E. JOFFRIN, P. SMEULDERS AND JET EFDA CONTRIBUTORS: Study of
slow n=1, m=1 reconnection in JET discharges with low central magnetic shear
Plasma Phys. Control. Fusion 48, 1005-1018 (2006)
A Fusion Programme
R. CESARIO, A. CARDINALI, C. CASTALDO, F. PAOLETTI, V. FUNDAMENSKI, S. HACQUIN: Spectral
broadening induced by parametric instability in lower hybrid current drive experiments of tokamak plasmas
Nucl. Fusion 46, 462-476 (2006)
F. ALLADIO, P. COSTA, A. MANCUSO, P. MICOZZI, S. PAPASTERGIOU, F. ROGIER: Design of the Proto-
Sphera experiment and of its first step (MULTI-PINCH)
Nucl. Fusion 46, S613-S624 (2006)
S. TOSTI, A. BASILE, L. BETTINALI, F. BORGOGNONI, F. CHIARAVALLOTI, F. GALLUCCI: Long-term tests
of Pd-Ag thin wall permeator tube
J. Membrane Sci. 284, 393-397 (2006)
M. MATTIOLI, G. MAZZITELLI, K.B. FOURNIER, M. FINKENTHAL, L. CARRARO: Updating of atomic data
needed for ionization balance evaluations of krypton and molybdenum
J. Phys. B: At. Mol. Opt. Phys. 39, 4457-4489 (2006)
A. BASILE, S. TOSTI, G. CAPANNELLI, G. VITULLI, A. IULIANELLI, F. GALLUCCI, E. DRIOLI: Co-current
and counter-current modes for methanol steam reforming membrane reactor: experimental study
Catalysis Today 118, 237-245 (2006)
A. BASILE, F. GALLUCCI, A. IULIANELLI, S. TOSTI, E. DRIOLI: The pressure effect on ethanol steam
reformig in membrane reactor: experimental study
Desalination 200, 671-672 (2006)
F. ZONCA, S. BRIGUGLIO, L. CHEN, G. FOGACCIA, T.S. HAHM, A.V. MILOVANOV, G. VLAD: Physics of
burning plasmas in toroidal magnetic confinement devices
Plasma Phys. Control. Fusion 48, B15-B28 (2006)
M. CIOTTI, A. NIJHUIS, P.L. RIBANI, L. SAVOLDI RICHARD, R. ZANINO: THELMA code electromagnetic
model of ITER superconducting cables and application to the ENEA stability experiment
Supercond. Sci. Technol 19, 987-997 (2006)
U. DE ANGELIS, G. CAPOBIANCO, C. MARMOLINO, C. CASTALDO: Fluctuations in dusty plasmas
Plasma Phys. Control. Fusion 48, B91-B97 (2006)
A. FRATTOLILLO: New simple method for fast and accurate measurement of volumes
Rev. Sci. Instrum 77, 045107 (2006)
F. ALLADIO, P. MICOZZI: Behaviour of perturbed plasma displacement near regular and singular X-points
for compressible ideal MHD stability analysis
Phys. Plasmas 13, 082505 (2006)
A. FRATTOLILLO: A simple automatic device for real time sampling of gas production by a reactor
Rev. Sci. Instrum. 77, 065108 (2006)
M.I.K. SANTALA, M.J. MANTSINEN, L. BERTALOT, S. CONROY, V. KIPTILY, S. POPOVICHEV, A. SALMI, D.
TESTA, YU BARANOV, P. BEAUMONT, P. BELO, J. BRZOZOWSKI, M. CECCONELLO, M. DE BAAR, P. DE
VRIES, C. GOWERS, J-M. NOTERDAEME, C. SCHLATTER, S. SHARAPOV AND JET-EFDA
CONTRIBUTORS: Proton-triton nuclear reaction in ICRF heated plasmas in JET
Plasma Phys. Control. Fusion 48, 1233-1253, (2006)
Progress Report 2006
80

S.E. SEGRE, V. ZANZA: Incident electromagnetic wave polarization and the resulting mode purity inside
magnetized plasma
Plasma Phys. Control. Fusion 48, 599-607 (2006)
C. STRANGIO, A. CARUSO, S. YU. GUS’KOV, V.B. ROZANOV, A.A. RUPASOV: Interaction of a smoothed laser
beam with supercritical-density porous targets on the ABC facility
Quantum Electr. 36, 3, 424-428 (2006)
R. BEDOGNI, A. ESPOSITO, M. ANGELONE, M. CHITI: Determination of the response to photons and thermal
neutrons of new LiF based TL materials for radiation protection purposes
IEEE Trans. Nucl. Sci. 53, 3, 1367-1370 (2006)
M. ANGELONE, M. MARINELLI, E. MILANI, A. TUCCIARONE, M. PILLON, G. PUCELLA, G. VERONA-RINATI:
Neutron detection and dosimetry using polycrystalline CVD diamond detectors with high collection efficiency
Radiat. Prot. Dosim. 120, 1-4, 345-348 (2006)
R. BEDOGNI, M. ANGELONE, A. ESPOSITO, M. CHITI: Inter-comparison among different TLD-based techniques
in a standard multisphere assembly for the characterisation of neutron fields
Radiat. Prot. Dosim. 120, 1-4, 369-372 (2006)
Articles in course of publication
P. BATISTONI: L’eredità di Chernobyl: i recenti rapporti del Chernobyl forum sulle conseguenze sulla salute
sull’ambiente e sul sistema socio-economico a vent’anni dell’incidente
Energia, Ambiente e Innovazione
J.R. MARTIN-SOLIS, B. ESPOSITO, R. SANCHEZ, F.M. POLI, L. PANACCIONE: Enhanced production of runaway
electrons during disruptive termination of discharges heated with lower hybrid power in the Frascati Tokamak
Upgrade
Phys. Rev. Letts
A. VANNOZZI, A. RUFOLONI, G. CELENTANO, A. AUGIERI, L. CIONTEA, F. FABBRI, V. GALLUZZI, U.
GAMBARDELLA, A. MANCINI, T. PETRISOR: Cube textured substrates for YBCO coated conductors:
microstructure evolution and stability
Supercond. Sci. Technol.
A. AUGIERI, G. CELENTANO, U. GAMBARDELLA, L. CIONTEA, V. GALLUZZI, T. PETRISOR, J. HALBITTER:
Analysis of angular dependence of pinning mechanisms on casubstituted YBa 2 Cu 3 O 7-δ epitaxial thin films
Superconductors Sci. Technol.
M. DE BENEDETTI AND JET EFDA CONTRIBUTORS: Observation of an intermediate rotation regime on JET
Nucl. Fusion
C. CASTALDO, S. RATYNSKAIA, V. PERICOLI, U. DE ANGELIS, L. PIERONI, E. GIOVANNOZZI, C. MARMOLINO,
A. TUCCILLO, G.E. MORFILL: Effects of dust on electrostatic probe signal in tokamak plasmas
Nucl. Fusion
S. TOSTI, A. BASILE, F. BORGOGNONI, L. BETTINALI, F. GALLUCCI, C. RIZZELLO: Design and process study of
Pd membrane reactors
J. Membrane Sci.
M. ROMANELLI, G.T. HOANG, C. BOURDELLE, C. GORMEZANO, E. GIOVANNOZZI, M. LEIGHEB, M. MARINUCCI,
D. MAROCCO, C. MAZZOTTA, L. PANACCIONE, V. PERICOLI, G. REGNOLI, O. TUDISCO AND THE FTU TEAM:
Parametric dependence of turbulent particle-transport in high-density electron heated tokamak plasmas
Plasma Phys. Control. Fusion
81
Progress Report 2006

A6 Publications, Patents and Events
A Fusion Programme
M. ROMANELLI, M. LEIGHEB, L. GABELLIERI, L. CARRARO, M.E. PUIATTI, M. MATTIOLI, L. LAURO-
TARONI, M. DE BENEDETTI, M. MARINUCCI, C. MAZZOTTA, L. PANACCIONE, G. REGNOLI, P.
SMEULDERS, O. TUDISCO, S. NOVAK, C. SOZZI, M. VALISA, AND THE FTU TEAM: Turbulent
transport of heavy impurities in tokamak electron heated high density plasmas: a study of FTU
discharges
Plasma Phys. Control. Fusion
D. FRIGIONE, L. GARZOTTI, C.D. CHALLIS, M. DE BAAR, P. DE VRIES, M. BRIX, X. GARBET, N. HAWKES,
A. THYAGARAJA, L. ZABEO, AND JET EFDA CONTRIBUTORS: Pellet injection and high density ITB
formation in JET advanced tokamak plasmas
Nucl. Fusion
Contributions to conferences
F. MIRIZZI, PH. BIBET, G. CALABRÒ, V. PERICOLI RIDOLFINI, A.A. TUCCILLO: PAM, MJ and conventional
grills: operative experience on FTU
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
G.L. RAVERA, C. CASTALDO, R. CESARIO, S. LUPINI, S. PODDA, G.B. RIGHETTI AND FTU TEAM: High
power RF components for IBW experiment on FTU
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
O. TUDISCO, C. MAZZOTTA, M.L. APICELLA, G.G. MAZZITELLI, G. MONARI, G. ROCCHI: Density profile
studies of plasmas with lithium limiter
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
C. CASTALDO, R. CESARIO, A. CARDINALI, M. MARINUCCI, P. MICOZZI, L. PANACCIONE, M. ANANIA, S. DI
FLAURO, B. EUSEPI, L. PAJEWSKI, G. SCHETTINI, G. GIRUZZI AND THE JET EFDA CONTRIBUTORS:
Modelling of experiments with ITER-relevant q-profile control at high β N by means of the lower hybrid current drive
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
F. ZONCA, S. BRIGUGLIO, L. CHEN, G. FOGACCIA, T.S. HAHM, A.V. MILOVANOV, G. VLAD: Physics of
burning plasmas in toroidal magnetic confinement devices
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006 (Invited Paper)
B. ESPOSITO, M. RICCI, D. MAROCCO, Y. KASCHUCK: A digital acquisition and elaboration system for
nuclear fast pulse detection
X Pisa 2006 Meeting on Advanced Detectors, La Biodola, Isola d’Elba (Italy), May 21-27, 2006
O. TUDISCO, G. GROSSETTI, C. SOZZI: Oblique ECE diagnostic on FTU
14 th Joint Workshop on “Electron Cyclotron Emission and Electron Cyclotron Resonance Heating,
Santorini Island (Greece), May 9-12, 2006
A. CARDINALI, B. ESPOSITO, F. RIMINI, M. BRAMBILLA, F. CRISANTI, M. DE BAAR, E. DE LA LUNA, P.
DE VRIES, X. GARBERT, G. GIROUD, E. JOFFRIN, P. JOFFRIN, P. MANTICA, M. MANTSINEN, A. SALMI,
C. SOZZI, D. VAN EESTER AND JET EFDA CONTRIBUTORS: Modeling and analysis of the ICRH heating
experiments in JET ITB regimes
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
M.L. APICELLA, G. MAZZITELLI, V. PERICOLI-RIDOLFINI, V. LAZAREV, A. ALEKSEYEV, A. VERTKOV, R.
ZAGÒRSKI AND FTU TEAM: First experiments with lithium limiter on FTU
17 th Conference on Plasma Surface Interactions (PSI), Hefei Anhui (China), May 22-26, 2006
Progress Report 2006
82

G. MADDALUNO, G. GIACOMI, A. RUFOLONI, L. VERDINI: Tungsten macrobrush sample exposure in FTU tokamak
17 th Conference on Plasma Surface Interactions (PSI), Hefei Anhui (China), May 22-26, 2006
V. MASSAUT, L. DI PACE, L. OOMS, K. BRODÉN, R.A. FORREST, M. ZUCCHETTI: The role of clearance in the
management of future fusion reactor radioactive materials
4 th Symposium Release of Radioactive Material from Regulatory Control, “Harmonisation of Clearance Levels and
Release procedures”, Hamburg (Germany), March 20-22, 2006
S. TOSTI, A. BASILE, F. BORGOGNONI, L. BETTINALI, C. RIZZELLO: Pd membrane reactor design
Euromembrane 2006, Taormina (Italy), September 24-28, 2006
U. DE ANGELIS, G. CAPOBIANCO, C. MARMOLINO, C. CASTALDO: Fluctuations in dusty plasmas
33 rd EPS Conference on Plasma Physic, Rome (Italy), June 19-23, 2006 (Invited Paper)
V. PERICOLI-RIDOLFINI, A. ALEKSEYEV, B. ANGELINI, S.V. ANNIBALDI, M.L. APICELLA, G. APRUZZESE, E.
BARBATO, J. BERRINO, A. BERTOCCHI, W. BIN, F. BOMBARDA, G. BRACCO, A. BRUSCHI, P. BURATTI, G.
CALABRÒ, A. CARDINALI, L. CARRARO, C. CASTALDO, C. CENTIOLI, R. CESARIO, S. CIRANT, V. COCILOVO,
F. CRISANTI, G. D’ANTONA, R. DE ANGELIS, M. DE BENEDETTI, F. DE MARCO, B. ESPOSITO, D. FRIGIONE, L.
GABELLIERI, F. GANDINI, E. GIOVANNOZZI, G. GRANUCCI, F. GRAVANTI, G. GROSSETTI, G. GROSSO, F.
IANNONE, H. KROEGLER, V. LAZAREV, E. LAZZARO, M. LEIGHEB, L. LUBYAKO , G. MADDALUNO, M.
MARINUCCI, D. MAROCCO, J.R. MARTIN-SOLIS , G. MAZZITELLI, C. MAZZOTTA, V. MELLERA, F. MIRIZZI, G.
MONARI, A. MORO, V. MUZZINI, S. NOWAK, F. ORSITTO, L. PANACCIONE, M. PANELLA, L. PIERONI, S.
PODDA, M. E. PUIATTI, G. RAVERA, G. REGNOLI, F. ROMANELLI, M. ROMANELLI, A. SHALASHOV, A.
SIMONETTO, P. SMEULDERS, C. SOZZI, E. STERNINI, U. TARTARI, B. TILIA, A.A. TUCCILLO, O. TUDISCO, M.
VALISA, A. VERTKOV , V. VITALE, G. VLAD, R. ZAGÓRSKI , F. ZONCA: Overview of the FTU results
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
G. MAZZITELLI, M.L. APICELLA, C. MAZZOTTA, V. PERICOLI RIDOLFINI. O. TUDISCO, V. LAZAREV, A.
ALEKSEYEV, A. VERTKOV, R. ZAGORSKI, AND FTU TEAM: Lithium as a liquid limiter in FTU
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
L. CHEN, F. ZONCA: Nonlinear equilibria, stability and generation of zonal structures in toroidal plasmas
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
L. CHEN, F. ZONCA: Theory of Alfvén waves and energetic particle physics in burning plasmas
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
F.P. ORSITTO, J–M. NOTERDAEME, A.E. COSTLEY, A.J. DONNÉ AND ITPA TG ON DIAGNOSTICS: Requirements
for fast particle measurements on ITER and candidate measurement techniques
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
F. CRISANTI, A. BECOULET, P. BURATTI, E. GIOVANNOZZI, C. GORMEZANO, E. JOFFRIN, A. SIPS, C.
BOURDELLE, A. CARDINALI, C. CHALLIS, N. HAWKES, J. HOBIRK, X. LITAUDON, G. REGNOLI, M. ROMANELLI,
A. THYAGARAJA, A. TUCCILLO, AND JET EFDA CONTRIBUTORS: JET hybrid scenarios with improved core
confinement
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
G. VLAD, S. BRIGUGLIO, G. FOGACCIA, K. SHINOHARA, M. ISHIKAWA, M. TAKECHI, F. ZONCA: Particle
simulation analysis of energetic-particle and Alfvén-mode dynamics in JT-60U discharges
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
F. ZONCA, P. BURATTI, A. CARDINALI, L. CHEN, J.–Q. DONG, Y.–X. LONG, A. MILOVANOV, F. ROMANELLI, P.
SMEULDERS, L. WANG, Z.–T. WANG: Electron fishbones: theory and experimental evidence
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
83
Progress Report 2006

A6 Publications, Patents and Events
V. PERICOLI-RIDOLFINI, M. L. APICELLA, G. MAZZITELLI, O. TUDISCO, R. ZAGÓRSKI AND FTU TEAM:
Edge properties with the liquid lithium limiter in FTU – experiment and transport modelling
Workshop on Edge Transport in Fusion Plasmas (ETFP), Cracovia (Poland), September 11-13, 2006
A Fusion Programme
V. PERICOLI–RIDOLFINI, P. BURATTI, G. CALABRÒ, M. DE BENEDETTI, B. ESPOSITO, L. GABELLIERI, G.
GRANUCCI, M. LEIGHEB, M. MARINUCCI, D. MAROCCO, C. MAZZOTTA, F. MIRIZZI, S. NOWAK, L.
PANACCIONE, G. REGNOLI, M. ROMANELLI, P. SMEULDERS, C. SOZZI, O. TUDISCO, A.A. TUCCILLO:
Internal transport barriers in FTU at ITER relevant plasma density with pure electron heating and current drive
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
A. CARDINALI, F. ROMANELLI: Simulation of burning plasma dynamics by ICRH accelerated minority ions
21 st IAEA Conference on Fusion Energy, Chengdu (China), October 16-22, 2006
A. BERTOCCHI, C. CENTIOLI, M. DI DONNA, F. IANNONE, M. PANELLA, L. PANGIONE, V. VITALE, L.
ZACCARIAN: The new FTU continuous monitoring system with Mac OS X technologies
Apple WWDC06 Conference, San Francisco (USA), August 7-11, 2006
G. VLAD, S. BRIGUGLIO, G. FOGACCIA, F. ZONCA: Interaction of fast particles and Alfvén modes in
burning plasmas
Joint Varenna-Lausanne International Workshop on Theory of Fusion Plasmas, Villa Monastero, Varenna
(Italy), August 28 - September 1, 2006
V. GALLUZZI, A. AUGIERI, L. CIONTEA, G. CELENTANO, F. FABBRI, U. GAMBARDELLA, A. MANCINI, T.
PETRISOR, N. POMPEO, A. RUFOLONI, E. SILVA, A. VANNOZZI: YBCO films with BZO inclusions for
strong-pinning in superconducting films on single crystal substrate
Applied Superconductivity Conference (ASC 2006), Seattle WA (USA), August 28 - Setpember 1, 2006
S. TOSTI, L. BETTINALI: Volumes measurement by means of membranes
Euromembrane 2006, Taormina (Italy), September 24-28, 2006
A. VANNOZZI, A. AUGIERI, G. CELENTANO, F. FABBRI, V. GALLUZZI, U. GAMBARDELLA, A. MANCINI, T.
PETRISOR, A. RUFOLONI: Cube textured substrates for YBCO coated conductors: influence of initial grain
size and strain conditions during tape rolling
Applied Superconductivity Conference (ASC 2006), Seattle WA (USA), August 28 - Setpember 1, 2006
A. CARDINALI, L. MORINI, F. ZONCA: Analysis of the validity of the asymptotic techniques in the lower
hybrid wave equation solution for reactor aplications
Joint Varenna-Lausanne International Workshop on Theory of Fusion Plasmas, Villa Monastero, Varenna
(Italy), August 28 - September 1, 2006
A. BERTOCCHI, M. DI DONNA, M. PANELLA, V. VITALE: The liquid lithium limiter control system on FTU
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
V. VITALE, C. CENTIOLI, F. IANNONE, M. PANELLA, L. PANGIONE, M. SABATINI, L. ZACCARIAN, R.
ZUCCALÀ: SA matlab based framework for the real-time environment at FTU
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
V. MASSAUT, R. BESTWICK, K. BRODEN, L. DI PACE, L. OOMS, R. PAMPIN: State of the art of fusion
material recycling and remaining issues
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
M. ANGELONE, L. PETRIZZI, M. PILLON, S. POPOVICHEV, R. VILLARI: Dose rate experiment at JET for
benchmarking the calculation direct one step method
24th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
Progress Report 2006
84

C. NERI, L. BARTOLINI, M. FERRI DE COLLIBUS, G. FORNETTI, F. POLLASTRONE, M. RIVA, L. SEMERARO: The
laser in vessel viewing system (IVVS) for ITER: test results on first wall and divertor samples and new developments
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
L. DI PACE, T. PINNA: Assessment of occupational radiation exposure (ORE) for hands-on assistance to the
remote handling at ITER ports and waste treatment
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
M. SANTINELLI, R. CLAESEN, A. COLETTI. T. BONICELLI, P.L. MONDINO, M. PRETELLI, L. RINALDI, L. SITA, G.
TADDIA: Solid state gyrotron body power supply, test results
24th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
P. BATISTONI, M. ANGELONE, L. BETTINALI, P. CARCONI, U. FISCHER, I. KODELI, D. LEICHTLE, K. OCHIAI, R.
PEREL, M. PILLON, I. SCHÄFER, K. SEIDEL, Y. VERZILOV, R. VILLARI, G. ZAPPA: Neutronics experiment on a
HCPB breeder blanket mock-up
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
T. PINNA, G. CAMBI, F. GRAVANTI: Collection and analysis of component failure data from JET systems
8 th IAEA Technical Meeting on “Fusion Power Plant Safety”, Wien (Austria), July 10-13, 2006
F. ALLADIO, P. COSTA, A. MANCUSO, P. MICOZZI, R. AKERS, G. CUNNINGHAM, M. GRYAZNEVICH, M. HOOD,
G. MC ARDLE, V. SHEVCHENKO, A. SYKES, F. VOLPE, A. DNESTROVSKIJ: Status and perspectives of MAST
start-up in the absence of solenoid flux
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
G. REGNOLI, M. ROMANELLI, C. BOURDELLE, M. DE BENEDETTI, M. MARINUCCI, V. PERICOLI, G. GRANUCCI, C.
SOZZI, O. TUDISCO, E. GIOVANNOZZI, ECRH, LH AND FTU TEAM: Microstability analysis of collisional plasmas
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
E. GIOVANNOZZI, C. CASTALDO, G. MADDALUNO: Evidence of dust in FTU from Thomson scattering diagnostic
measurements
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
G. FOGACCIA, S. BRIGUGLIO, M. ISHIKAWA, K. SHINOHARA, M. TAKECHI, G. VLAD, F. ZONCA: Particle
simulations of energetic particle driven Alfvèn modes in JT-60U
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
J.R. MARTIN-SOLIS, B. ESPOSITO, R. SANCHEZ, F.M. POLI, L. PANACCIONE: Runaway current plateau
formation during disruptions in the FTU Tokamak
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
B. ESPOSITO, G. GRANUCCI, S. NOWAK, P. SMEULDERS, J. BERRINO, J.R. MARTIN-SOLIS, R. SANCHEZ, L.
GABELLIERI, M. LEIGHEB, F. GANDINI, D. MAROCCO, C. MAZZOTTA, O. TUDISCO: Disruption mitigation
experiments in FTU using ECRH
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
M. ROMANELLI, G.T. HOANG, C. BOURDELLE, C. GORMEZANO, E. GIOVANNOZZI, M. LEIGHEB, M.
MARINUCCI, D. MAROCCO, C. MAZZOTTA, L. PANACCIONE, V. PERICOLI, G. REGNOLI, O. TUDISCO, AND
THE FTU TEAM: Parametric dependence of turbulent particle transport in high collisionality plasmas on the
Frascati Tokamak Upgrade FTU
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
M. ROMANELLI, M. LEIGHEB, L. GABELLIERI, L. CARRARO, M.E. PUIATTI, M. VALISA, M. MATTIOLI, L. LAURO-
TARONI, M. DE BENEDETTI, M. MARINUCCI, C. MAZZOTTA, G. REGNOLI, P. SMEULDERS, S. NOVAK, C.
85
Progress Report 2006

A6 Publications, Patents and Events
SOZZI: Investigation of turbulent transport of heavy impurities in FTU electron heated plasmas
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
A Fusion Programme
G. CALABRÒ, V. PERICOLI-RIDOLFINI, L. PANACCIONE AND FTU TEAM: Effect of the scattering from
edge density fluctuations on the lower hybrid waves in FTU
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
M. RIVA, B. ESPOSITO, D. MAROCCO: A new pulse-oriented digitial aquisition system for nuclear
detectors
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
M. PILLON. M. ANGELONE, D. LATTANZI, M. MARINELLI, E. MILANI, A. TUCCIARONE, G. VERONA-
RINATI, S. POPOVICHEV, R.M. MONTEREALI, M.A. VINCENTI, A. MURATI AND JET -EFDA
CONTRIBUTORS: Neutron detection at JET using artificial diamond detectors
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
C. CASTALDO, U. DE ANGELIS, V.N. TSYTOVICH: Screening and attraction of dust particles in plasmas
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
M.L. APICELLA, M. LEGHEB, M. MARINUCCI, G. MAZZITELLI, FTU TEAM, V. LAZAREV, A. ALEKSEYEV,
A. VERTKOV: Energy balance of FTU discharges with lithizated walls
33 rd EPS Conference on Plasma Physics, Rome (Italy) June 19-23, 2006
S.V. ANNIBALDI, F. ZONCA, P. BURATTI: Excitation of beta-induced Alfvèn eigenmodes in the presence of
a magnetic island
33 rd EPS Conference on Plasma Physics, Rome (Italy), June 19-23, 2006
E. VISCA, S. LIBERA, A. MANCINI, G. MAZZONE, A. PIZZUTO, C. TESTANI: Pre-brazed casting and hot
radial pressing: a reliable process for the manufacturing of CFC and W monoblock mockups
24 th Symposium on Fusion Technology (SOFT), Warsaw (Poland), September 11-15, 2006
Reports
RT/ 2006/35/FUS
RT/2006/69/FPN
RM2006A000429
R. CHIRICO
Studio sui rischi per la sicurezza e per la salute associati all’utilizzo di un limiter di
litio durante le sperimentazioni con FTU (Frascati Tokamak Upgrade)
R. CHIRICO
Teorie e parametrizzazioni per la ripartizione degli IPA su particolato atmosferico:
stato dell’arte
A6.2 Patents
A. DELLA CORTE, A. DI ZENOBIO
Procedimento per la realizzazione di un giunto tra cavi superconduttori di tipo
CICC a basso livello di ingombro, bassa resistenza elettrica e basso costo di
realizzazione
Progress Report 2006
86

RM2006A000314 L. BETTINALI, V. VIOLANTE, F. SARTO, C. SIBILIA, M. BERTOLOTT, E. CASTAGNA, I.
DARDIK, S.LESIN, T. ZILOV, M. TSIRLIN
Materiali laminati metallici con inclusioni di materiale dielettrico per l’amplificazione ed il
controllo del campo elettrico di interfase, e relativo processo di produzione
RM2006A000102
S. TOSTI, D. LECCI, C. RIZZELLO, A. BASILE
Procedimento a membrana per la produzione di idrogeno da reforming di composti
organici, in particolare idrocarburi o alcoli
A6.3 Conferences and Events
June 19-23, 2006
33 rd European Physics Society - Conference on Plasma Physics
Rome (Italy)
A6.4 Seminars
21/03/2006 S. ORTOLANI - Consorzio RFX - ENEA - Padova, Italy
Active MHD control experiments in RFX - mod
24/03/2006 J. KASAGI – LNS, Tohoku University - Tohoku, Japan
Low energy nuclear reactions in condensed matter
28/04/2006 S. MIRNOV - TRINITI - Troitsk, Russia
Test of the lithium capillary - pore system (CPS) as tokamak limiter and DEMO perspective of Li CPS
10/07/2006 M. TESSAROTTO - Università di Trieste - Trieste, Italy
Il problema di Debye per plasmi debolmente e fortemente accoppiati
25/09/2006 M. SHOUCRI - IREQ - Varennes, Quebec, Canada
Study of a turbulent spectrum at the edge of a 2D plasma slab in the gyrokinetic approximation
13/12/2006 P. SCARIN - Consorzio RFX - Padova, Italy
Edge turbulence evidence in RFX - mod with GPI diagnostic
13/12/2006 A. SANTUCCI - Università “Tor Vergata” - Roma, Italy
Reforming di etanolo in reattori a membrana
19/12/2006 S. GERASSIMOV - Technical Univ. of Münich and CERN - Münich, Germany
Use of ROOT to store large quantities of scientific data
13/12/2006 V. CAPALDO - Università “La Sapienza” - Roma, Italy
Reforming di etanolo in reattori a membrana
87
Progress Report 2006

B1 R&D on Nuclear Fission
B Fission Technology
B1.1 Innovative Fuel Cycles Including Partitioning and
Transmutation
Activities on partition and transmutation started under the 5 th European Framework Programme
(FP5) and have continued under FP6. The work on chemical partitioning was carried out under the
European Research Programme for the Partitioning of Minor Actinides (EUROPART) concerning the
partitioning of long-lived radionuclides (LLRNs) contained in the nuclear waste resulting from the
reprocessing of spent nuclear fuel. After separation, the LLRNs will be destroyed by nuclear means
so as to become short-lived or stable nuclides or conditioned into stable dedicated solid matrices.
Transmutation activities were carried out under the European Transmutation (EUROTRANS) project
submitted by ENEA, Commissariat à l’Energie Atomique (CEA), Forschungszentrum Karlsruhe (FZK)
and the Belgian Nuclear Research Centre (SCK-CEN). The objectives are to demonstrate
experimentally the accelerator driven system (ADS) operations and dynamic characteristics and then
to deliver a conceptual design for a European Transmutator Demonstrator (ETD), including its overall
technical feasibility, and to perform an economic assessment.
ENEA coordinates the European Virtual European Lead Initiative (VELLA) project, which has the
ambitious intent to homogenize the European research area in the field of leading technologies for
nuclear applications in order to produce a common platform of work that will continue also after the
end of the initiative. The issues of this activity are also of interest to evolutionary and innovative
reactor activities (see B1.2).
Studies on innovative uranium-free inert matrix and thorium fuels, aimed at in-reactor plutonium
incineration either in the current light-water reactors (LWRs) or in next-generation reactors, were
continued in 2006. ENEA researchers participated in a first evaluation of the experimental data from
the IFA-652 irradiation test performed up to the end of 2005 in the Halden Material Test Reactor
(Norway). A good response of the proposed fuel concept was found, with an under-irradiation
stability similar to that of UOX and MOX fuels, except for a somewhat higher fission gas release
(FGR) rate. In parallel, implementation of the inert matrix fuel (IMF) basic thermo-physical properties
and models on the fuel-rod performance code Transuranus was completed and a simulation of the
first-phase of IFA-652 irradiation in Halden was performed. Preliminary modelling with Transuranus
on CER-CER and CER-MET fuels for transmutation, started in 2005, was also continued.
Dispersion of the fissile phase based on Pu and MAs oxides in Mg oxide matrix (CER-CER) or in
molybdenum metal matrix (CER-MET) was considered in the study.
Partitioning technology
The principal operation of chemical partitioning (fig. B1.1, [B1.1]) is electrorefining, which takes place
in an electrochemical cell where dissolution of most of the fuel elements occurs. This is followed by
selective electrodeposition of the actinides onto a solid and/or a liquid cathode through application
Progress Report 2006
88

Spent
fuel
Disassembly
and
chopping
Duct Clad, N M
Melting
Consolidation
Metal waste
Gas waste (T,Xe,Kr)
Electrorefining
Spent
salt
TRU
extraction
TRU
Immobilisation
Salt waste (Cs,Sr,RE)
Salt, Cd
U, salt
U-TRU
-CD-salt
Cathode
processing
Crucible
of an electrochemical
difference among elements
in molten LiCl-KCl salt and
liquid cadmium (or
bismuth) under a highpurity
argon atmosphere at
773 K (pyro processing).
The ex perimental
campaigns concerned
electrorefining experiments
with the Pyrel II plant and
conditioning of chloride salt wastes arising from pyroprocessing of spent nuclear fuel.
Experimental campaigns. The Pyrel II facility [B1.2] was used to study the behaviour of lanthanum and
cerium loaded on different anodes and electrotransported to different cathodes. The current was varied
whenever possible and both salt and metal phases were sampled. Seven experiments were performed:
direct transportation from fuel dissolution basket (FDB) to solid steel cathode (SSC); anodic dissolution
(from FDB to Bi pool); transportation from FDB to liquid bismuth cathode (LBC); direct (chemical)
dissolution; transportation from Bi pool to SSC; transportation from Bi pool to LBC; salt clean-up between
Bi-Li anode and SSC.
Full evaluation of the results obtained is not easy at this stage, but a few considerations can be made:
• A cathode deposit is practically absent in any case.
• The electric current can be imposed only to a maximum specific value, depending on the type of
experiment.
• The fuel dissolution basket extracted from the salt bath after the experiments with La ingots shows that
La is still present in the FDB.
• Salt clean-up allows a significant amount of residual elements to be removed from the salt bath, with
the metals deposited at the Bi-Li anode, mainly around the magnesia vessel.
A real puzzle is the concentration of La and Ce in the chloride salt. It can be supposed that something
other than the electric current is involved in the process. Clarification of the above and other important
questions is mandatory for complete com prehension of
the phenomena which take place during the
2000
electrorefining experiments with Pyrel II, and for the
project of a new plant (Pyrel III) designed for
electrorefining with uranium ingots.
1000
Chloride waste treatment. The pyrochemical
process produces a salt waste containing Li, K, and FP
chlorides, which after several batches accumulate in
the molten salt media and represent an environmental
concern because of their high water solubility. Sodalite,
a naturally occurring mineral, is a major candidate for
conditioning salt waste as it can incorporate chloride
metals in its cage-like structure. Hence pure sodalite
was prepared for use as reference material.
Zr
TRU: Pu, Np, Am, Cm
RE: Rare earth
NM: Noble metal
New
fuel
Pin
casting
Mold
crucible
Counts/s
Fig. B1.1 – General schematic of spent
fuel reprocessing by pyrochemical
electrorefining (redrawn from ref. B1.1)
0
10 30 50 70 90
2θ
Fig. B1.2 – X-ray diffraction spectrum related to the
formation of sodalite from chloride salt, silica and sodium
aluminate after 50 h of reaction
[B1.1] T. Nishimura et al., Progr. Nucl. Energy 32, 3/4, 381-387 (1998)
[B1.2] G. De Angelis and E. Baicchi, A new electrolyzer for pyrochemical process studies, Presented at the GLOBAL 2005 (Tsukuba 2005)
References
89
Progress Report 2006

B1 R&D on Nuclear Fission
B Fission Technology
zos04
zos03
zos02
zos01
zeolite A
10 20 30 40 50 60 70
2θ
Time
Fig. B1.3 – X-ray diffraction spectra related to the
formation of sodalite starting from zeolite A, at increasing
reaction times
Tests performed to prepare the sodalite
starting from silica and sodium aluminate
(fig. B1.2) or from zeolite A (fig. B1.3) show
the formation of an intermediate phase,
nepheline, which is known to be more
leachable with respect to other phases, like
sodalite or pollucite. The role of microwaves
and their effect on the reaction yield were also
studied. Two different heating methods (inside
a microwave oven or in a tubular oven) proved
successful, even if the reaction conditions as
well as the starting materials need a clear
definition. The demonstration that the same
intermediate compound (nepheline) is present
in the synthesis of sodalite irrespective of the
method used was a success in itself.
The next step should be the synthesis of pure sodalite and localisation of the position of the various
cations inside the crystal lattice. While Li, Na, and K are presumably included in the structure of
sodalite, it is more difficult to identify the relative positions of ions such as Cs, Sr and Ba.
Transmutation systems and related technology
Research on transmutation was mainly focussed on:
• Neutronic design of a Pb-cooled European facility on an industrial-scale transmuter (EFIT – the
European Facility for Industrial Transmutation) (Domain Design).
• Study of the energetic gain expected in the Reactor-Accelerator Coupling Experiment (RACE)
and on a new neutron detector for characterisation of the neutron spectrum in subcritical devices
(Domain ECATS).
• Experimental activities to study the interaction between lead bismuth eutectic (LBE) and water
consequent to heavy leaks due to a cooling tube rupture inside the steam generator and on a
large-scale integral test (Domain DEMETRA).
EFIT core design criteria: the “42-0” approach. Work concerned the neutronic analysis of the
EFIT sub-critical reactor, in particular the preliminary definition of the core and fuel subassembly
(S/A). Experience gained through the ANSALDO-ENEA collaboration in the Preliminary Design Study
– Experimental Accelerator Driven System (PDS-XADS) project, a 80-MW th core cooled by LBE
[B1.3], was exploited as far as possible. However, since the fuel (uranium-free, fig. B1.4a)) as well
as the goal of maximum rate minor-actinide (MA: Np, Cm, Am) burning are quite different from the
PDS-XADS, an innovative approach was developed to deal with the core design and fuel cycle,
mainly aimed at minimising the cost per kg of MAs burnt. Hence, a twofold strategy was assumed
[B1.4]:
1. The so called “42-0” approach, i.e., the invariant 42 kg of fissioned material per TWh th have to
be MAs, whilst Pu is neither burnt (since it would be of low value in sub-critical reactors) nor bred
(since this would be inconsistent with the uranium-free choice) and acts as a “catalyser”. This
univocally leads to fuel enrichment at about 45.7% in Pu, which represents the main starting
parameter of the EFIT core design. In fact the K eff swing over the fuel cycle (which also mainly
depends on the fuel enrichment) has to be compatible with the proton accelerator performance.
2. The core size has to be optimised to obtain the minimum cost per fissioned kg of MAs. Since the
burning rate per TWh th does not depend on the core size, the optimisation criterion (in the 42-0
approach) becomes the minimum cost of the deployed power unit. Actually, the reactor unitary
cost decreases with core size (in a certain range), whilst the accelerator cost is likely to increase,
Progress Report 2006
90

mainly due to the loss of source
efficiency. As the information
needed for a possible trade-off
was missing, it was decided to
assume the largest core
compatible with the present
design of the spallation module,
as a simplified criterion.
The ANSALDO-develop ed [B1.4]
800-MeV/20 mA spallation module
which can evacuate 11.2 MW of
power (70% of the total beam
power) was assumed as reference.
The target is a windowless type (as
in PDS-XADS [B1.3]) with a
horizontal coolant flow in the
spallation region, mechanical
pumps and a heat sink below the
Pb free surface level. At room
temperature the circular target has
Pu
Pu238
Pu239
Pu240
Pu241
Pu242
Pu244
Pu238
Pu239
Pu240
Pu241
Pu242
Pu244
(w%)
3.737
46.446
34.121
3.845
11.850
0.001
an outer diameter of 782 mm: hosting it by replacing 19 S/As,
the fuel assembly (FA) wrapper flat to flat external distance is
186 mm (191 mm by considering the clearance between FAs).
In addition, the spallation module size together with the
maximum proton current (20 mA), the selected sub-criticality
level (K eff =0.97) and the maximum allowable linear power
(P L ≅200 [W cm -1 ]) give as output the core size and the overall
power (P th ).
Figure B1.4a) shows the adopted fuel isotopic composition (Pu
and MA vectors), which comes from a reprocessed MOX spent
fuel, irradiated at 60 GWd t -1 (i.e., 30 years [B1.5]) and after a
30–year ageing. The PuO 2 and MAO 2 were inserted in a
magnesium-oxide (MgO) matrix with different volume
Pu & MA isotopic compositions
MOX spent fuel after 30 years’ cooling (CEA)
Pu vector
Ma vector
91.8% Am
Np237
4.3% Cm
Am241
Am242
Am242m
Am243
Cm242
Cm243
Cm244
Cm245
Cm246
Cm247
Cm248
percentages in order to flatten the radial performances by using the same fuel enrichment in the whole
core.
Since a high MA content was assumed, particular attention was also devoted to the safety aspects. By
adopting a stochastic approach, it was demonstrated that similar uranium-free fuels, with cross-section vs
energy behaviour similar to that in figure B1.4b), are characterised by:
• “deterioration” of the delayed neutron effective fraction and kinetic parameters;
• lack of Doppler prompt reactivity feedback;
• deterioration of the void effect such that it does not guarantee in any case the desired sub-criticality
level.
Cross section (barns)
MA
Np237
Am241
Am242
Am242m
Am243
Cm242
Cm243
Cm244
Cm245
Cm246
Cm247
Cm248
(w%)
3.884
75.510
3.27E-06
0.254
16.054
2.37E-20
0.066
3.001
1.139
0.089
0.002
1.01E-04
10 5
10 4
b)
elastic scattering
1000
100
10
1
0.1
0.01
0.001
10 -4
fission
capture
10 -10 10 -8 10 -6 10 -4 10 -2 1 100
Energy (MeV)
Fig. B1.4 – a) Pu and MA vectors; b) capture, elastic
scattering and fission cross sections (from MCNPX code
and JEFF 3.1 neutron cross sections)
a)
[B1.3]
[B1.4]
[B1.5]
XADS 41 SNPX 042, Core configuration technical specification of the LBE-cooled XADS, Contractual Deliverable n° D10 FIKW-CT-2001-
00179, Technical Report Ansaldo Energia (2001)
Specialist Meetings on the Pb-EFIT core design, INPN Orsay – Paris, 24-25 October 2005; ENEA - Bologna, 22-23 February 2006; CEA
Cadarache, 9-10 March 2006; CEA Cadarache, 13-14 June 2006
G. Rimpault, Definition of the detailed missions of both the Pb-Bi cooled XT-ADS and Pb cooled EFIT and its gas back-up option,
Technical Report CEA SPRC/LEDC 05-420 (2005); and IP EUROTRANS – DM1 Design – WP 1.1 – Deliverable 1.1, Contract n° FI6W-
CT-2004-516520 (2006)
References
91
Progress Report 2006

45
B1 R&D on Nuclear Fission
B Fission Technology
0.6
7.2 mm
7.52 mm
8.72 mm
Fuel
Void
SS
Pb
PD Hom (W cm -3 )
0.16
Fuel inner
62.5% MgO
115
13.63 mm
4.91 mm
VF(Fuel pellet)=21.65%
Filling ρ = 0.9167
(750°C)
(480°C)
(440°C)
Fuel outer
50% MgO
ff rad =1.29
ff ax =1.14
191 mm
186 mm
178 mm
168+1 Fuel pins
(7+1 pin rows)
b)
a)
a)
EFIT two-zone model. A 395-MW th
reference core with two radial zones,
surrounded by dummy reflector elements in
Pb (fig. B1.5a)) was designed, and its
transmutation capability and overall core
performance were checked. To achieve this
configuration, while the active height (90 cm,
to limit the pressure drop) and the ΔT core (400-
480°C) were maintained fixed, the reactor
radial dimension (i.e., the number of FAs) and
the MgO volumetric fraction (VF) were varied
to flatten the core radial performance, at the
same time keeping the condition K eff ≤0.97
over the cycle.
95
This two-zone solution was analysed by a
ff rad =1.45
ff ax =1.15
cylindrical geometry model (fig. B1.5b)) in
75
which the different radii were assumed
Max INN (BOC)
equivalent to a certain number of FAs.
55 Max OUT (BOC)
Max INN (EOC)
Neutronic analysis was carried out with the
Max OUT (EOC)
ERANOS version 2.0 deterministic code [B1.6]
35
40 80 120 160 and the ERALIB1 nuclear data library [B1.7].
R (cm)
The spatial and energy distributions of the
external neutron source, generated by the
Fig. B1.6 – 395 MW th EFIT two-zone model: a) inner and spallation process of the 800-MeV protons
outer FA design; b) PD radial profiles at about half active impinging on the Pb windowless target, were
height (BOC and EOC)
obtained by Monte Carlo methods (Monte
Carlo N-particle transport code [B1.8]). The
angular distribution of the source neutrons was assumed isotropic in the laboratory frame.
The EFIT two-zone model (fig. B1.5) exhibits a suitable radial power distribution by using MgO VFs
of 62.5% and 50% (lowest MgO technological content) in the inner (48 FAs) and outer (174 FAs) fuel
zones, respectively. Figure B1.6a) shows the geometrical characteristics of the FAs with 168 fuel
pins and the 21.65% fuel VF. The 395 MW th power was reached by imposing the max P L (which
really depends on the MgO VF and the related pellet conductivity) at 213 and 180 [W cm -1 ] in the
inner and outer zones, respectively. Figure B1.6b) shows the power density (PD) radial distributions
330
265
215
185
170
140
125
75
Z (cm)
Beam line
15
Target
Internal lead
R t = 43.7
Top assembly
Plenum
Fuel inner
Plenum
ΔR
Fuel outer
50
50
ΔR 1
ΔR 2
Foot assembly
90
Box
dummy
Fig. B1.5 – 395 MW th EFIT two-zone model: a) hexagonal layout; b) cylindrical model
Pb Ext
200 252
b)
R (cm)
Progress Report 2006
92

on the homogenised core (at about half of the active
height where PD reaches maximum values). The orange
and red horizontal lines, which correspond to the
technological limits on the max P L , indicate that both at
the first and at the second year of irradiation
(corresponding to the beginning of cycle [BOC] and end
of cycle [EOC]) the safety limits are not exceeded. Figure
B1.6b) also shows the radial and axial form factor values
(ff rad and ff ax ), corresponding to the worst condition
(BOC, lowest K eff value).
The resulting sub-critical core exhibits very satisfactory
performance, since it has a very small K eff swing over the
cycle limited to about 200 pcm (fig. B1.7a)), requiring a
roughly constant proton current that never exceeds
16.3 mA.
As for the in-pile fuel cycle, a three-year maximum
residence time was assumed, due to Pb corrosion
constraint. Considering a refuelling pattern of 1/3 of the
core each year, some “ad hoc” hypotheses permit
burn–up calculations to be performed without any actual
refuelling, as follows:
kg K eff
0.975
0.973
0.971
BOL BOC EOC EOL
0.969
0 1 2 3
3100
2900
2700
ΔK eff swing
≅ 200 pcm
Tot Pu
b)
Tot MA
ΔMA/MA (BOL) = -12.95%
ΔPu/Pu (BOL) = -0.25%
2500
0 1 2 3
Time (years)
Fig. B1.7 – a) K eff (t) behaviour; b) absolute and relative
MAs and Pu burn-up performance
a)
• For the core performance it is sufficient to consider the reactor conditions at the first year (BOC) and at
the second year (EOC) of irradiation.
• For the transmutation performance, the third year of irradiation was considered as FA end of life (EOL).
Figure B1.7b) shows the burn-up capability (with the 45.7% fuel enrichment). By considering the mass
balances at the third year of irradiation (FA EOL), a relative MA and Pu burn-up of about 13% and 0.25%,
respectively, was obtained. The transmutation obtained for the MA and Pu isotopes is 40.6 and
0.7 [kg TWh th -1 ], respectively, which is almost in agreement with the 42-0 approach.
The main drawback of this two-zone core is the too high ff rad value in the outer part (1.45; fig. B1.6b)): a
RELAP thermohydraulic analysis [B1.9] showed that the limit on the maximum cladding temperature
(550°C) is reached in this zone. This is a consequence of the small difference between the Pb outlet
(480°C) and the maximum cladding temperatures allowed, which imposes working with very limited ff rad .
To avoid having a lot of different SA orifices, a core with three radial zones was considered. However the
main result is that the 42-0 approach for MA transmutation in a sub-critical system (without Pu burning and
production) is a viable strategy because the resulting K eff (t) swing (which depends on the fuel enrichment)
is compatible with a reasonable proton accelerator current range.
Reactivity worth, decay heat and fuel equilibrium analyses. The uranium-free choice (with high MA
content) is characterised by a lack of the Doppler effect, a low delayed neutron fraction and deterioration
of the coolant void effect (fig. B1.8a)). These results, obtained by MCNP [B1.8], suggest that some
[B1.6] G. Rimpault et al., Schema de calcul de reference du formulaire eranos et orientations pour le schema de calcul de projet, CEA
XT–SBD–0001 (1997)
[B1.7] E. Fort et al., Application a la realisation de ERALIB1, bibliotheque de donnes neutroniques pour le calcul des systems a spectre rapide,
Technical Report CEA SPRC/LEPh 97-002 (1997)
[B1.8] J.S. Hendricks et al., MCNPX Version 2.5.B, LA-UR-02-7086 (2002)
[B1.9] C.D. Fletcher and R.R. Schultz, Relap5/Mod3 Code Manual – User’s Guidelines, Vol. 5, Idaho National Engineering Laboratory;
NURG/CR-55 EGG-2596 (1992)
References
93
Progress Report 2006

B1 R&D on Nuclear Fission
B Fission Technology
Void effect total reactivity
worth per ring (pcm)
3000
2000
1000
0
-500
Cadarache meeting
core solution
Bologna meeting
core solution
1 2 3 4 5 6
Ring
2000
50%PU 50% MA-origen2
a) b)
113
22
MOX-origen2
MOX-fispact
30%PU 70% MA-fispact
30%PU 70% MA-origen2
50%PU 50%MA-fispact
neutronic design assumptions as well as the expected core performance should be reconsidered,
mainly from the safety viewpoint.
The decay heat problem deriving from the massive use of MAs [B1.10, B1.11] was also studied in
detail [B1.4, B1.12]. Compared to the standard MOX fuels, AnOx fuel has a higher decay heat rate,
the decay heat decreases less in time and is higher (by up to ∼45 times). Figure B1.8b) shows
performance and behaviour vs decay time for uranium-free fuel. From the reactor design viewpoint,
an important aspect is the long-term reliability of the decay-heat-removal components.
Finally the question of core (re)fuelling only by MA fuels, independently of the start-up fuel
composition, was approached [B1.4]. Preliminary results indicate that whatever the start-up (Pu,
MA)O 2-x fuel composition, an equilibrium (Pu, MA)O 2-x fuel composition, with different Pu/MA ratio
and Pu & MA vectors, is reached so that the equilibrium core can be (re)fuelled only by MAs. The
"potential reactivity" (or k ∞ ) seems to be sufficient both for the core reactivity and for sustaining BU
cycles.
EFIT thermohydraulic and safety analyses. Parallel to the activity for the core design, a
thermohydraulic numerical model of the EFIT reactor was developed with the system code RELAP5
[B1.9]. Besides representing the first step in the dynamic simulation of the sub-critical reactor
(coupled thermohydraulic and neutronic) for the safety analyses, the model allowed a preliminary
investigation of safety issues in order to confirm the neutronic design. A preliminary layout of the
primary system (fig. B1.9) that implemented all the basic design options for an industrial
Absolute DH (W/kg)
1000
0
1×10 -1 1×10 3 1×10 5 1×10 7 1×10 9
Time (s)
Fig. B1.8 – a) Coolant void effect; b) absolute decay heat (W/kg) vs time behaviour
175 176 (177/8/9) 171
TMDP JUN 5 1 (2/3/4)
SGs
Pumps
Pth
181 281 151
112
152
170
DHR
153
154
182
183
184
161
160
JUN 106
JUN 105 JUN 103
102
82 62
282
283
284
100
120
210 211 110 111
Outer Outer Inner Inner
Average Hot Average Hot
Core Core Core Core
Fig. B1.9 – Schematic of EFIT primary system and RELAP5 nodalization
Progress Report 2006
94

Fig. B1.10 – Unprotected loss-of-flow transient –
RELAP5 main parameters
transmutator (the use of pure melted lead as coolant,
elimination of intermediate loops, installation of heat
exchangers and mechanical pumps inside the primary vessel,
uranium-free fuel) was simulated starting from the neutronic
results of the EFIT two-zone cylindrical model.
The RELAP5 analyses were used to verify the capability of the
thermohydraulic design to support high temperature and
power density and to pass from MOX fuel to uranium-free
fuel, both in operational and in accidental conditions. Both
Design Basis and Design Extension Conditions (DBCs and
DECs) were considered so as to have a first evaluation of the
inherent safety behaviour of the plant. Figure B1.10 shows the
main parameter trends for an unprotected loss-of-flow
accident (100% power, natural circulation, full capability of
steam generators, low capability of decay heat removal
system, BOC conditions). The results show that a stable
natural circulation capable of limiting cladding, fuel and
coolant temperatures to acceptable values is quickly attained.
M/M0, P/P0
Temperature (°C)
1×10 0
6×10 -1
Core power
SG power
Core flow
2×10 -1 0
400 600 800 1000
1.4×10 3
1×10 3
8×10 2
Core mass flow and power
Inner core (hot) max temperature
a)
1429 1400
728
Tclad(hot)
Tfuel(hot)
684 Tlead(hot) 619
4×10 2 Time (s)
400 600 800 1000
b)
665
A preliminary analysis with the SIMMER-III code of the steam generator tube rupture accident (single and
multiple rupture) was performed for the preliminary design solution [B1.13]. A portion of the EFIT vessel
around the steam generator was modelled, using a simplified cylindrical 2D geometry centred on the tube
rupture location. The core structure was schematically represented in the model, and stagnant lead was
considered inside the vessel. The results of these calculations show that neither steam explosion effects
nor the risk of void formation inside the core are of concern in the SIMMER-III evaluation.
Electron vs proton accelerator–driven subcritical system performance using TRIGA reactor at
power. In the framework of the RACE project [B1.14] ENEA was concerned with studies on the energetic
gain expected in the RACE core. It is assumed that the thermal power dissipated by the W-Cu or uranium
RACE target during the high-power phase will be ~25 kW, which corresponds, according to preliminary
MCNPX calculations on a depleted uranium multi-disk target undergoing a 1.0 mA - 25 MeV electron
beam, to a source strength of ∼6×10 13 n/s.
Different MCNPX [B1.15, B1.16] and TRIPOLI4 [B1.17] calculations were performed to analyse the
coupling between a TRIGA (ENEA Casaccia) subcritical core and a photoneutron source and get a first
assessment of the RACE target-core power coupling coefficients (energetic gain) and compare them with
those obtained for the TRIGA Accelerator-Driven Experiment (TRADE) core configurations [B1.18]. Hence,
for some calculations it was assumed that an electron beam impinges on a W-Cu target surrounded by
[B1.10] NEA – OECD, Fuels and materials for transmutation. A status report. Nuclear Sci. ISBN 92-64-01066-1, NEA n. 5419, OECD (2005)
[B1.11] IP EUROTRANS, Actions list: decay heat benchmark (2006)
[B1.12] G. Glinatsis, Decay heat investigation on the U-free transmuter cores dedicated fuels, ENEA Internal Report in preparation
[B1.13] H. Yamano et al., Simmer III: a computer program for LMFR core disruptive accident analysis - Version 3.A: Model summary and program
description, O-arai Engeneering Center - Japan Nuclear Cycle Development Institute (2003)
[B1.14] D. Beller, Overview of the AFCI reactor-accelerator coupling experiments (RACE) project, Presented at the 8 th Information Exchange
Meeting on Actinide and Fission Product Partitioning and Transmutation (OECD/NEA) (Las Vegas 2004)
[B1.15] MCNP4C A general Monte Carlo nparticle transport code, J.F. Briesmeister Ed., Los Alamos National Laboratory report, LA-13709- M
(2000)
[B1.16] J. S. Hendricks et al., MCNPX, VERSION 2.5.d, LA-UR-03-5916 (2003)
[B1.17] J.P. Both et al., TRIPOLI4, a Monte-Carlo particles transport code. Main features and large scale application in reactor physics, Presented
at the Inter. Conference on Supercomputing in Nuclear Application - SNA’2003 (Paris 2003)
[B1.18] C. Rubbia et al., Nucl. Sci. Eng. 148, 103-123 (2004)
References
95
Progress Report 2006

B1 R&D on Nuclear Fission
B Fission Technology
1mm
Ø15
Ø28.5
17
18
19
24
6.5 cm
Ø1.3
60
771
504.7
72.7 209.7
105.3 83.5
16.493
4.311
7.351
7.351
348 20 107
Fig. B1.11 – TRADE-like conical shaped target. (dimensions in mm)
1.7 mm
12
mm
20 mm
Fig. B1.12 – Multi-plate target
Tantalum window
(thickness=1 mm)
Water
(thickness=2 mm)
Aluminium
(thickness=1 mm)
Depleted uranium
Fig. B1.13 – UT-NETL core with central
cylindrical uranium target
the subcritical Rc-1 TRIGA core. These cases
will be indicated in the following as TRADEelectrons
(TRADE-e).
Two core configurations, representative of the
SC0 (-500 pcm) and SC2 (-3000 PCM) TRADE
configurations, were coupled with three types of
target. The simulations took into account
electron, photon and neutron transport, using
the MCNPX code. The material considered was
a W-Cu alloy (75% wt and 25% wt respectively,
bulk density 14.7 g/cm 3 ) in two geometrical
shapes: raw cylindrical (h=8.89 cm, r=3.49 cm) and
conical (aperture angle 16.5°, h=34.8 cm,
r max =1.5 cm, radial thickness 0.19 cm), as shown in
figure B1.11 [B1.19]. A third target type was
represented by a set of coaxial disks of depleted
uranium with aluminium cladding and water coolant
(fig. B1.12). Uranium has the highest photoneutron
production [B1.20].
Given a fixed electron beam, the neutron source mainly
depends on the target material, but the feasibility of the
target configuration has to be considered. Some target
concepts were analysed and the cooling capability of
the system, choice of materials, thermomechanical
behaviour and safety issues were discussed.
Calculations by MCNPX were performed, mainly to
estimate the different neutron sources (besides the
power deposition distribution).
The first target configuration considered was a bare uranium cylinder with r=3.25 cm and h=8 cm.
This is not a feasible target solution as, for example, it does not allow proper cooling, but it is useful
for estimating the maximum achievable neutron yield (some geometrical constraints have to be
taken into account since the target is to be placed in the central channel of the core). The second
and third target configurations were based on the conical geometry shown in figure B1.11. The
materials considered were depleted uranium and tantalum, as for the TRADE target. The third
configuration was a hollow uranium cylinder irradiated by the electron beam in its inner surface,
where the power deposition was spread out to allow cooling. The photonuclear reaction data used
for the MCNPX simulations were the LA150u and the BOFOD libraries, both collected and reported
in the International Atomic Energy Agency (IAEA) photonuclear data library [B1.21].
Ø63
8.578
Ø43
Ø50.8
Ø46
Progress Report 2006
96

Fig. B1.14 – Geometrical models
Some analyses for the 1-MW TRIGA reactor were
performed at the Nuclear Engineering Teaching Laboratory
(NETL) of the University of Texas (UT). The UT-NETL core
was explicitly modelled using MCNP-5 in a coupled
electron/photon/neutron problem. Photonuclear data were
taken from T-16 at the Los Alamos National Laboratory
(LANL) and the electron source was a 25-MeV beam with a
1-cm–diameter beam spot. A 25-kW beam power and a
cylindrical uranium target in the central position were
assumed (fig. B1.13).
For the calculations performed by the TRIPOLI4 Monte
Carlo code [B1.16], the geometry was a simplified “clean”
Texas A&M University (TAMU) TRIGA core (fig. B1.14) having
the same fuel composition as TRADE. Two core
configurations, representative of SC2 (-3000 pcm) and SC3
(–5000 pcm) TRADE configurations, were considered, with
the external source located in central core position. The
external neutron source was considered to have a spectrum
similar to that of photoneutrons obtained through electron
interaction with a Pb target of a 20-MeV linear accelerator
(linac).
Figures B1.15 and B1.16 show the core power vs the
subcritical level for the different targets taken into account.
The results are normalised at 25-kW beam power. The
analysis shows the requirements for an electron-driven
coupling experiment aimed at providing significant validation
elements about the dynamic behaviour of an accelerator
driven system (ADS), in terms both of target performance
and of beam power characteristics. The results show that it
is necessary to have a U target in the central position of the
TRIGA reactor to obtain a core power greater than 50 kW
for K eff >0.98. Such a minimum power is required to have
SC2-K eff = 0.96816 SC3-K eff = 0.95037
feedback effects in the system responses in the presence of source/reactivity transients, as indicated by
preliminary analysis of the influence of thermal reactivity feedback in RACE.
Some results for TAMU TRIGA configurations indicate tighter target/core coupling than in the Casaccia
TRIGA RC-1 (TRADE-e), as can be seen by comparing the results in figure B1.16 relative to the same multiplate
target in a central position in TAMU TRIGA and TRADE-e. Further investigations are needed. In any
case, a final check should be performed for the actual core loading that could be envisaged for RACE-HP
(high power).
In-core test for the Piccolo-Micromegas neutron detector. One important step needed for approval
of a demonstration device is experimental validation of simulations. Of particular interest is determination
Core power (kW)
100
80
60
Ta conical in TRADE
40
W-Cu cylinder in TRADE
20
W-Cu conical in TRADE
0
0.95 0.97 0.99
K eff
Fig. B1.15 – Core power vs subcritical level for
non–fissile targets (beam power 25 kW)
Core power (kW)
100
80
60
40
U hollow cylinder in TRADE
U conical in TRADE
20
U multiplate in TRADE
U multiplate in TAMU
U in UT-NETL
0
0.95 0.97 0.99
K eff
Fig. B1.16 – Core power vs subcritical level for
uranium targets (beam power 25 kW)
[B1.19] P. Agostini et al., Neutronic and thermo-mechanic calculations for the design of the TRADE spallation target, Presented at the Inter.
Conference on Accelerator Applications (Venice 2005)
[B1.20] W.P. Swanson, Radiological safety aspects of the operation of electron linear accelerators, IAEA Technical Report, Series No.188,
STI/DOC/010/188 (1979)
[B1.21] Handbook on photonuclear data for applications: cross sections and spectra, IAEA-TECDOC-1178 (2000)
References
97
Progress Report 2006

B1 R&D on Nuclear Fission
B Fission Technology
a)
Detector
35 mm
Neutron/charged particle converter
B-10
Pads
160μm 1mm20 mm
Micromesh
P3
P4
3 mm
Th-232
HV1
U-235
HV1
HV2 P4
HV2
P1 P2 P1 P3
P2
Ceramic insulator
Ar+ (2%) Iso-butane (1 bar)
b)
Fig. B1.17 – a) Piccolo-Micromegas assembly used in
1–MW TRIGA Casaccia reactor test; b) schematic of the
principle of Piccolo-Micromegas detector (in horizontal
position) for neutron flux measurement
of the neutron spectrum (i.e., neutron flux as a
function of neutron energy) for different
configurations of the subcritical device. As is
well known, the neutron flux in an ADS consists
of neutrons produced via spallation reactions in
the target and fissions from the multiplying
blanket. Unfortunately neutron spectra cannot
be measured using only one type of detector. To
cover the complete energy range of neutrons
produced, a new neutron detector, named
Piccolo-Micromegas, based on Micromegas
technology has been developed in a
cooperation with CEA/DAPNIA/SEDI (Saclay
France) and CNRS/IN2P3 LPC (Caen France).
The principle of Micromegas is based on
detecting the electrons created by ionization of
the filling gas by charged particles. Operation of
Piccolo-Micromegas as a neutron detector
requires an appropriate neutron/charged particle
converter, which can be either the filling gas or
the target, with a suitable deposit on the
entrance window.
Fissile elements such as 235 U, 232 Th are used
simultaneously as neutron/charged particle
converters in addition to 10 B and recoil ions of
the gas (Ar + iC 4 H 10 quencher) filling the
detector. Using four converters with a unique
detector will permit practically on line extraction
of a large range of the neutron flux spectrum in
a specific position in the reactor. The large
dynamic range of Piccolo-Micromegas will
permit precise measurements and a detailed
scanning of the flux into the whole reactor
volume.
At very high counting rate (>100 MHz)
measurement will be performed on a current
Fig. B1.18 – a) The six BNC cables; b) Piccolo- mode basis. At low counting rate, the fast
Micromegas assembly inside the TRIGA reactor response of the detector will allow the incident
particles to be counted one by one by means of
a low-noise fast preamplifier. This will open up a
way to measuring the neutron flux at the peripheral part of the reactor and, in some cases, also
when full reactor power is not used.
After a first test with the CELINA 14-MeV neutron source at Cadarache, a second test was
performed with a sealed prototype placed inside the core of the TRIGA reactor at ENEA Casaccia
in the configuration shown in figure B1.17. The detector was placed inside a long sealed stainless
tube having the same dimensions as the empty reactor rod. The usual BNC (fig. B1.18a)) cables
Progress Report 2006
98

Configuration 250
Fuel
Graphite
Piccolo Micromegas
Source
Regulating rods
Shim 1 and 2 and
safety rod
Irradiation facility
Rabbit
Configuration 250
Fuel
Graphite
Piccolo Micromegas
Source
Regulating rods
Shim 1 and 2 and
safety rod
Irradiation facility
Rabbit
Fig. B1.19 – Position of Piccolo-Micromegas in the
periphery of the TRIGA reactor core
Fig. B1.20 – Position of Piccolo-Micromegas in the middle
of the TRIGA reactor core
were used and placed inside a
10-m watertight stainless tube
(fig. B1.18b)).
Piccolo-Micromegas has 6000
5×10 5
worked at different reactor
C4 (Th)
powers from 10 W to 400 kW
3×10 5
at two different positions inside
1×10 5
the reactor: on the periphery
0
4000
B-10
0 100 200 300 400
(fig. B1.19) and in the middle
Reactor power (kW)
(fig. B1.20). At low power,
measurements were per -
formed by simple counting on
the pads; at higher power, only 2000
the high-voltage current can
H recoil
be registered as the counting
rate is too high. This
experiment is aimed at
studying the behaviour of the
0
0 100 200 300 400
detector inside a nuclear
Reactor power (kW)
reactor. Several topics have
been tackled, such as Fig. B1.21 – Currents vs reactor power and fission fragment counts from 232 Th vs
response linearity vs the reactor power
reactor power or ageing. The
output energy spectra corresponding to the different converters have been also studied to get a thorough
understanding of the detector response.
Currents (nA)
Counts/100 s
1×10 6
9×10 5 Th-FF counts per 100 s vs reactor power
7×10 5
An example of the results is shown in figure B1.21, which shows clearly that the new Piccolo-Micromegas
can work in a nuclear reactor, which is a very aggressive experimental condition.
Interaction of lead alloys with water. The aim of the experimental campaign is to assess the physical
effects and possible consequences of interaction between LBE and the water from large leaks caused by
a cooling-tube rupture inside the steam generator of a reactor such as XT-ADS or EFIT and to provide data
for validation of the mathematical modelling.
The relevant parameters for the tests were selected according to the XT-ADS reference design. The
SIMMER code was adopted to simulate the experiments for the modelling activity. The ex LIFUS5 plant
(fig. B1.22), designed and constructed to simulate this kind of interaction in a wide range of conditions
(e.g., pressure up to 200 bar, temperature up to 500°C) was refurbished and re-arranged for the
experiments.
For Test n.1 (fig. B1.23), successfully carried out in March 2006, pressurized water was injected at 70 bar
into the reaction vessel containing LBE at 350°C. The most important of the experimental results
(fig. B1.23) in terms of pressure evolution in the reaction system is that a maximum value (78 bar) higher
U-235
99
Progress Report 2006

B1 R&D on Nuclear Fission
Fig. B1.22 – P&I of LIFUS5 plant
B Fission Technology
Pressure (bar)
80
60
40
20
0
0
TC
S4
Pb-17Li
S1
V10
1/2”
V8
V1 V2 LT2
H 2 3”
GS1
D1 FA
Al 1/2” PT1
Al 1”
V12 S5
V6 V15
Al 3”
Al 10
V13
TC
TC
1-30
PT11
PT S2 TC
PT7
6-8
H 2 O
S1 PT9
SP
DPT
TC3
1
PT5
Pb-17Li TC2
PT3
TC
PT
2-4
V11
V3
V5
Wa 1/2”
To vacuum
pump
V4
GS2 V14
Fig. B1.23 – Results of Test n.1
S5
1000 2000 3000
Time (ms)
than the water injection pressure (70 bar) was
reached in the reaction-expansion vessel of LIFUS5
during the test. This means that it is fundamental to
adopt suitable and reliable countermeasures in order
to avoid such pressure peaks in the reactor pool.
Integral Circulation Experiment activities. In the
framework of the Domain DEMETRA, ENEA is
strongly involved in the “Large-Scale Integral Test”
work package and is committed to performing an
integral experiment with the aim of reproducing the
primary flow path of the European Transmutation
Demonstrator (ETD) pool nuclear reactor, cooled by
LBE.
Heat
exchanger
CIRCE vessel
In 2006, ENEA worked on the design of a new test
section (fig B1.24) to install in the CIRCE facility at
ENEA Brasimone for the Integral Circulation
Experiment (ICE). To achieve the goals of the integral
test, a high thermal performance heat source (HS)
Fig. B1.24 – ICE test section
was required. The ICE heat source, consisting of a
pin bundle made up of electrical heaters with a total thermal power of 800 kW, was designed to
achieve a ΔT HS /L act value of 100°C/m, a pin power density of 500 W/cm 3 and an average liquid
Riser
Fitting
volume
TC
PT10
H 2 3”
V7
D2
S3
H 2
Drainage
1/2”
V9
V16
Dead
volume
Fuel pin
simulator
Flow meter
Progress Report 2006
100

Table B1.I – Overview of the experimental parameters adopted for the ICE activity,
compared with the ETD concepts foreseen
XT-ADS EFIT ICE
Coolant LBE Pure lead LBE
Primary loop circulation Mechanical pump Mechanical pump Gas lift technique
Fuel assembly lattice Hexagonal Hexagonal Hexagonal
Fuel assembly type Wrapper Wrapper Wrapper
Fuel assembly spacer Grid Grid Grid
Fuel pin diameter (D) [mm] 6.55 8.72 8.2
Pitch to diameter ratio (p/D) 1.41 1.56 1.8
Fuel heat flux q’’ [W/cm 2 ] 85-115 100-140 100
Fuel power density q’’’ [W/cm 3 ] 500-700 450-650 488
Average velocity fuel pin region [m/s] 1 1 1
Fuel pin active length [mm] 600 900 1000
Tin/tout core [°C] 300/400 400/480 300/400
ΔT HS /L act [°C/m] 167 88 100
Fuel pin cladding material T91 T91 AISI 316L
Secondary coolant Low pressure Water with Pressurized
boiling water superheated water
steam
metal velocity of 1 m/s, in accordance with the
reference value adopted for the ETD concepts (XT-
ADS, EFIT). A pin bundle was chosen to simulate
the HS in order to improve cooling of the heaters
and avoid overheating of the cladding material. The
HS was coupled to the test section by a suitable
mechanical structure, designed by ENEA. The
heaters and the mechanical structure which
surrounds them make up the so-called fuel pin
simulator (FPS). The main experimental para meters
characterising the heat source (fig. B1.25) and ICE
activity are reported in table B1.I.
Fuel pins simulated by electrical heaters
p
h
Assembly Hexagonal
Diameter 8.2 mm
Pitch/diam. 1.8
Active length 1000 mm
Active pins 31
Total pins 37
Fuel heat flux: 100 W/cm 2
Thermal power pin: 26 kW
A gas lift pumping system successfully tested and
Fig. B1.25 – ICE heating section
qualified during previous ex perimental campaigns
in CIRCE is used to perform the ICE activity. A pressure head of 40 kPa is available to promote the LBE
circulation along the flow path.
The ICE test matrix has been defined, with the following tests foreseen:
• Steady-state circulation: isothermal condition, LBE average temperature of 350°C, no power supply.
The aim is to get fluid-dynamics characterisation of the test section.
• Steady-state circulation: LBE average temperature of 350°C, full thermal power. The aim is to evaluate
the coupling of the HS and heat exchanger (HX) and analyse the thermal hydraulic behaviour of a heavy
liquid metal (HLM) pool system primary loop.
• Transient condition: loss of cold sink, starting from the nominal condition. The aim is evaluate the trend
of the average temperature through the HS and HX.
• Transient condition: loss of pumping system, starting from the nominal condition. The aim is analyse the
transition from forced to natural circulation and characterise the natural circulation flow regime in a HLM
pool system.
An appropriate cold sink was designed. Consisting of a prototypical LBE-pressurized water shell heat
101
Progress Report 2006

B1 R&D on Nuclear Fission
B Fission Technology
Tube side
Assembly: Triangular
Tubes: “U” - shape
Num. tubes: 13(3/4”)
Material: T91
Pressure: 6 bar
ΔT: 30°C
Flow rate: 6.5 kg/s
Velocity: 1.5 m/s
Max ΔT wall: 285°C
Fig. B1.26 – ICE heat exchanger
VELLA - Virtual European Lead Laboratory
exchanger made up of a seamless
U–tube, it will be placed in the upper
plenum of the main vessel (fig. B1.26).
The possibility of installing and testing
different prototypical HXs (i.e., helical
tubes) is under evaluation. In any case
the opportunity of adopting pressurized
water as a secondary fluid has to be
confirmed by the currently ongoing
safety analysis.
ENEA is responsible for coordinating the Virtual European Lead Laboratory (VELLA), which is an FP6
integrated infrastructure initiative started in October 2006. The ambitious intent to homogenise
European research in the field of lead technologies for nuclear applications thereby producing a
common platform of work suggested dividing VELLA into Networking Activities (NAs), Transnational
Access Activities (TAs) and Joint Research Activities (JRAs). The objectives of the NAs is to create
a wide “virtual" community of researchers, define common standards and protocols for the use of
the facilities and interact with other programmes and institutes operating in this field. The TA
objectives are to promote access by researchers, universities and companies to current
infrastructures and knowledge in order to increase the competitiveness of European industry. The
TAs would also provide a framework for training young researchers to use the EU infrastructures
during the three years of the project and for promoting mobility between the partners and the
laboratories of the consortium. Finally, the JRA goals are to improve current knowledge on lead
technologies, develop and operate heavy liquid metal (HLM) components and instrumentation,
especially in a neutron irradiation environment and, finally, study HLM thermal hydraulics.
ENEA, as coordinator, is involved in all the NAs, provides access to the infrastructures and
participates in three of the four JRAs.
In 2006 efforts were mainly devoted to management activities in order to rationally organise the work
to be carried out. The management structure of VELLA was approved and the technical and
scientific committees responsible for managing the JRAs and the access to infrastructures were set
up. The activities to be performed in the framework of the JRAs were planned in detail and an
appropriate quality control system established. ENEA’s activities also included financial
management. A lot of work was also devoted to creating the “virtual” community, by constructing
an official web-site [B1.22], intended to become a central point of information on HLM technologies.
B1.2 Evolutionary and Innovative Reactors
The main issue in this field in 2006 was the definition and launching of a three-year R&D national
programme based on “strategic funding devoted to the National Electric System R&D” and
focussed on participation in international initiatives such as the International Near-Term Deployment
(INTD) and Generation-IV Nuclear Systems. The programme is being managed through a specific
agreement between the Italian Ministry of Economic Development and ENEA, with the joint
involvement of major national organisations still active in the nuclear sector, i.e., Ansaldo Nucleare,
Ansaldo Camozzi, Del Fungo Giera Energia, Italian Universities Consortium for Research in Nuclear
Technologies (CIRTEN) and SIET (an ENEA subsidiary SME). The total funds for the first year amount
to 5.5 MEuro and comparable annual funds are expected for the rest of the programme. The main
goals of the programme are to
Progress Report 2006
102

• keep open the future nuclear energy option in the country;
• contribute to the development of innovative nuclear systems which promise to be “sustainable”,
acceptable by the public and economically interesting;
• sustain the growth of the necessary competences through participation in promising projects with solid
foundations;
• support the effort required by national industry to keep up with the pace at world and domestic level.
In particular, the national R&D programme supports experimental and analytical activities for the further
development of the GENIII+ International Reactor Innovative and Secure (IRIS) and GENIV Lead-Cooled
Fast Reactor (LFR) as well as some technological activities as support to the GENIV Very High Temperature
Reactor (VHTR) and to the GENIII AP1000 reactor.
This national programme is also intended to be synergic and coherent with the Generation-IV initiative as
well as with a number of the Sixth European Framework Programme (FP6) projects: the European Lead-
Cooled System (ELSY), coordinated by Ansaldo Nucleare; the Reactor for Process Heat, Hydrogen and
Electricity Generation (RAPHAEL); Roadmap for a European Innovative Sodium Cooled Fast Reactor
(EISOFAR); Assessment of Liquid Salts for Innovative Applications (ALISIA).
ENEA also participates in the “Coordination Action” Sustainable Nuclear Fission Technology Platform (CA
SNF-TP), which is in charge of developing a coherent European strategy on
nuclear fission and consolidating the European and Euratom position
within the GIF initiative and in the linked Coordination Action
Partitioning and Transmutation European Roadmap for
Sustainable Nuclear Energy (PATEROS), aimed at “delivering
a European vision for the deployment of the partitioning and
transmutation technology up to the scale level of pilot
plants for all its components”.
The following is a short summary of the main results
achieved over 2006 with reference to the innovative
nuclear systems developed within the abovementioned
programmes and projects.
Suppression
pools
Dry well
containment
International Reactor Innovative and Secure
IRIS (fig. B1.27) is designed as an advanced, modular
small-medium reactor. It is an integral-type pressurizedwater
reactor with a power level of 335 MWe, featuring an
integral primary system configuration with all the main
components (reactor coolant pumps, steam generators, pressurizer,
control rod drive mechanisms) located within the reactor vessel. This
configuration enables a simplified design with enhanced reliability and economics and supports its safetyby-design
approach, which results in exceptional safety characteristics. In addition to electricity-only
production, IRIS is also well suited for cogeneration, including water desalination, district heating, and
process steam generation.
Fig. B1.27
–
The
IRIS reactor
Furthermore, IRIS well fits the recently announced US Department of Energy initiative - Global Nuclear
Energy Partnership (GNEP) - aimed at supporting worldwide expansion of the use of nuclear energy in a
responsible and proliferation-resistant way. Within the GNEP framework, IRIS can - in the near term - offer
an advanced reactor design to satisfy the need for smaller, grid-appropriate reactors.
[B1.22] www.3i-vella.eu.
References
103
Progress Report 2006

B1 R&D on Nuclear Fission
Fig. B1.28 – Sketch of the SPES-3 integral testing facility
Containment
Primary pump
B Fission Technology
Long term
gravity makeup
system
Pressure
suppression
pools
Reactor cavity
Emergency
boration
tanks
Integral reactor
pressure vessel
Fig. B1.29 – Sketch of ELSY reactor block
Pb
IRIS is being developed by an international team, led
by Westinghouse, incorporating organisations from
ten countries, including Italy (ENEA, CIRTEN,
Ansaldo Nucleare, Ansaldo Camozzi and SIET). The
preliminary design has been completed and the
testing needed for design certification just started in
2006. The centrepiece of the testing programme is
the integral system testing to be performed at the
SIET facility in Italy. Since mid-2006, a multinational
group of experts coordinated by Westinghouse and
ENEA has been designing the SPES-3 experimental
facility, devoted to an integral testing campaign for
the IRIS reactor licensing process. The advanced
safety features of the IRIS reactor require a unique
test facility, where both the containment system and
the primary system are simulated (fig. B1.28).
Moreover, the scaling approach suggested the
adoption of an integral layout for the facility as well.
The SPES-3 integral test facility to be built at the
SIET labs is a full height, full pressure and
temperature, scaled volume facility (1:100 power
and volume ratio). The main components, i.e., the
helical coil steam generators, the core bundle
simulator, the pressurizer, are integrated in the tall
reactor pressure vessel as in the IRIS design. The
facility is designed both for integral testing and for
separate effect tests. A “simulation group” has been
set up to support both the design of the facility and
the pre-test and post-test analyses. Both bestestimate
system codes (RELAP, GOTHIC) and CFD
codes (Fluent) are adopted.
B4C rods
(36 As)
Fuel outer
(54 FAs;E pu =23.8%)
Fuel intermediate
(90 FAs;E pu =18.9%)
Fuel inner
(109 FAs;E pu =15.6%)
Fig. B1.30 – The open square
lattice ELSY core
European Lead-Cooled
Fast System
The ELSY reference design
(fig. B1.29) is a 600–MWe
pool-type fast reactor cooled
by pure lead. This concept has
been under development since
September 2006 and is
sponsored by the Euratom
FP6. The ELSY project,
coordinated by Ansaldo
Nucleare, is being performed
by a consortium consisting of
twenty organisations including
ENEA, CIRTEN and CESI
Ricerca from Italy. ELSY aims
to demonstrate the possibility
Progress Report 2006
104

of designing a competitive
and safe fast critical reactor
using simple engineered
technical features, whilst
fully comply ing with the
Generation-IV goal of MA
burning capability.
The activities carried out in H active fuel 1100 mm
2006 were mainly devoted
to defining requirements,
selecting options and verifying critical issues. The requirements reflect the GEN IV goals of sustain ability,
economics, safety, proliferation-resistant and physical protection. Sustainability is the leading criterion for
core design, which focusses on demonstrating the potential of the reactor to be self-sustaining in
plutonium and to burn its own generated MAs. Two different core configurations are being studied:
wrapperless assemblies in a square lattice where pins are arranged in square bundles as well (fig. B1.30),
or more conventional wrapped assemblies in a hexagonal lattice. Both the concepts assume the same
thermal power (1500 MWth), fuel (MOX), fuel residence time (5 years), BU (100 MWd/kgHM for the hottest
assembly), cladding (T 91 with a maximum allowable temperature of 550°C), cladding radiation damage
(100 DpA), inlet (400°C) and outlet (480°C) core temperature. The comparison focusses on conversion
factor and MA burning capability (sustainability); core dimensions, loading factor, fuel inventory, peak and
average power density (economics); coolant velocity (on which corrosion and natural circulation depend),
control-rod system, coolant void/density effect and reactivity coefficients (safety); use or not of axial
blankets (proliferation).
In order to provide the core design with some boundary conditions, a preliminary T/H analysis of the fuel
rod was also performed with the RELAP5 code. The parameter-set for open square SA in table B1.II was
fixed on the basis of engineering considerations, previous LFR designs and current knowledge on lead
technology. As the cladding temperature is considered
the most critical parameter to meet safety requirements
in heavy liquid metal (HLM) reactors, aluminized T91
steel was selected as cladding material to increase the
safety limit in operating conditions (T clad < 550 °C).
Preliminary parametric calculations allowed
determination of the maximum admissible linear power
to meet such a limit; the result was a maximum form
factor of the radial power distribution of 1.2 (fig. B1.31).
However, this preliminary evaluation could be too
conservative. Indeed, applying different empirical
correlations from the literature for single-phase heat
transfer in lead and LBE, the corresponding cladding
temperature distributions are pretty different
(fig. B1.32). In short, the correlations recently derived
for rod bundle geometry abate the heat transfer
resistance and may be beneficial for the design of an
LFR. For instance, using the Zhukov’ correlation
(developed in the framework of the BREST reactor), the
cladding peak temperature (red or brown line) is about
40°C lower than the value calculated with the
correlation implemented in the RELAP code
(yellow line).
Concerning lead technology, ENEA coordinates all the
activities to be performed in the work package and
dedicates a strong effort to investigating the physical
Table B1.II – Design parameters for LFR square open SA
Thermal power 1500 MW # FA 240
Av. linear power 200 W/cm FA Square 17x17
Inlet temperature 400 °C Power/FA 6.25 MW
Outlet temperature 480 °C Axial shape factor 1.16
Total mass flow 126157 kg/s D fuel pellet 7.14 mm
D pin 8.5 mm Gap thickness 0.115 mm
Pitch 13.6 mm Clad thickness 0.565 mm
Temperature (°C)
Elevation (m)
580
560
540
Acceptable clad max temp=550°C
520
0.9 1.1 1.3 1.5
Radial shape factor
Fig. B1.31 – Peak cladding temperature at different
radial shape factors
1.0
0.8
0.6
0.4
0.2
Lead temperature
Borishanski
Graber
Calamai
Zhukov (no spacers)
Zhukov (spacers)
Subbotin
Kirillov-Stromquist
Lubarsky
“clean”
“dirty”
0
380 420 460 500 540 580
Temperature (°C)
Fig. B1.32 – Effect of the different correlations on the
peak cladding temperature
105
Progress Report 2006

B1 R&D on Nuclear Fission
and chemical properties of lead and its interaction with the structural materials and the secondary
coolant, as well as evaluating corrosion protection-coatings and corrosion-resistant steels for
cladding, pump impellers, etc.
B Fission Technology
Following a critical review of the collection of existing data on lead thermophysical properties carried
out by ENEA, Bologna University and the Belgian Nuclear Research Centre (SCK-CEN), a consistent
database on the design-relevant properties is being compiled.
ENEA’s experience in LBE technology has been invaluable in extrapolating the technological
solutions (technologies, components, instrumentation) developed for lead bismuth to pure lead. In
addition, ENEA has contribut ed, with its knowledge on purification, to a critical review of LBE
properties and has participated in the collection of information on the oxygen control system (OCS),
instrumenta tion and procedures (filling, draining, component removal).
Very high temperature reactor
In consideration of the helium loop (HE-FUS3) available at ENEA Brasimone plus past experience in
the relative modelling, ENEA became a partner in the Reactor for Process Heat, Hydrogen and
Electricity Generation (RAPHAEL) Consortium in May 2006. RAPHAEL, coordinated by AREVA NP
SAS France, is an FP6 Integrated Project aimed at developing technologies for gas systems with
temperatures ranging between 850 and 1000°C.
ENEA’s contribution concerns three sub-projects: coupled reactor physics and core thermo-fluid
dynamics, component development and safety. In particular, ENEA will test a prototypical heat
exchanger (HEATRIC mockup) with helium at the primary and secondary sides in the HE-FUS3
facility (fig. B1.33). The main experimental conditions will reproduce the operating conditions
expected for the component: pressure 2.4 MPa and He flow rate 0.0475 kg/s in both sides, I/O
temperatures 508-127°C in the primary side and 108-488°C in the secondary side. The
experimental data from the tests at HE-FUS3 (several steady states and two loss of flow [LOFA]
transients (fig. B1.34) in a wide range of working conditions) will be used in a benchmark exercise
for validation of the thermal-hydraulics system transient codes.
GAS
ANALYSIS
HEATRIC
FV262
OUT IN
FV23ø
L263
MIXER
FV234
FV1
FV7
PSE265
VACUUM
FV261
FV231
PSE257 HV251
HV252
HV25ø
E219/1 E219/2 E219/3
HEATER HEATER HEATER
BY-PASS
E214
ECONOMIZER
PSV
269
FT
228
FV213
FV4
PSV268
FV235
FV5
L264
MIXER
FT212
COLD
TEST SECTION
E24ø
COOLER
PCV246
PCV248
FV249
PSE2ø9
FV6
HELIUM
DISCHARGE
SYS
V2ø5
TANK
HV289
S26ø
FILTER
PSV2ø8
VACUUM
FV9
FV8
PURIFICATION OUT
PURIFICATION
IN
FV1ø
K2øK2ø
COMPRESSOR
HV2
Fig. B1.33 – HE-FUS3 facility with HEATRIC
mockup
HE BOTTLES
PRV244 PCV246 FV247
HV243
HELIUM
FILLING
HV3øHV3ø SYS
Progress Report 2006
106

Fig. B1.34 – LOFA transient – test section temperature
at different positions and mass flow rate
Finally, in order to provide experimental data
for the validation of neutronics deterministic
codes, ENEA will perform benchmark
experiments in the fast source reactor
TAPIRO of ENEA Casaccia. These
experiments will make it possible to
Temperature (°C)
reproduce the strong changes in the neutron spectrum at the interface core/reflector, peculiar to hightemperature
gas reactor (HTGR) systems.
600
500
400
800
600
400
300
200
0 40 80 120 160 200
Time (s)
Mass flow rate (kg/h)
B1.3 Nuclear Safety
Nuclear safety studies are performed in the framework of international programmes. During 2006 the
activities addressed code validation and accident analysis, severe accidents, and reliability and risk
analysis.
Code validation and accident analysis
Mainly performed within a bilateral agreement funded by
the French Institute for Radioprotection and Nuclear Safety
(IRSN), the activities are summarised in the following.
Analysis of the BETHSY experiment 4.3b. The CESAR
thermal-hydraulic module of the Accident Source Term
Evaluation Code (ASTEC) V1 was validated against
experiment 4.3b at the CEA Grenoble BETHSY facility,
which simulates a multiple steam generator tube rupture in
a French PWR-900 reactor.
Comparison of the CESAR results with experimental data
confirms the capability of the code to well simulate
accident situations in such reactors and, in general, the
test parameters and phenomena are well reproduced. In
particular, a) the depressurization rate of the primary and
secondary sides is well calculated by the code
(fig. B1.35); b) in both test and calculation the restart of
the primary pump is effective in recovering the circulation
in the secondary side, leading to a rapid depressurization
towards stable and safe conditions; and c) the appearance
and disappearance of stratification phenomena in the
secondary side of the broken steam generator are well
reproduced by the code (fig. B1.36).
PWR-1300 H3 sequence analysis. A severe accident
sequence resulting from a station blackout with total
unavailability of auxiliary and safety systems after reactor
scram in a French PWR-1300 plant was calculated with
the integral ASTEC V1.2 up to core relocation and vessel
rupture. The ASTEC results before core degradation takes
place were compared with the results of the same
sequence calculated with CATHARE2 V2.5 code in order
Pressure (Pa)
Temperature (K)
1.6×10 7
1.2×10 7
8.0×10 6
4.0×10 6
560
540
520
0
Rapid
cooldown
Break opening
Spray on
Slow cooldown
Pump 2
restart
0 8000 16000
Time (s)
Fig. B1.35 – Primary and secondary pressure (solid lines)
compared with experimental data (dots)
Break flow reverses
500
0 4000 8000 12000 16000
Time (s)
Fig. B1.36 – Steam generator riser wall temperature (solid
line) at different heights compared with experimental data
(dots)
107
Progress Report 2006

B1 R&D on Nuclear Fission
120
Qliq CATHARE
Qliq ASTEC
Fig. B1.37 – Pressurizer safety valve mass flow rate
(ASTEC – CATHARE result comparison)
B Fission Technology
Flow rate (kg/s)
H(m)
80
40
Qvap ASTEC
Qvap CATHARE
0
6000 10000 14000
Time (s)
4.74
2.94
1.14
-0.657
-2.45
-3.6 -1.8 0 1.8 3.6
R(m)
3000
2500
2000
1500
1000
600
T(K)
Fig. B1.38 – Core melt relocation at transient end
(ASTEC code result)
to highlight the differences between the two
codes.
In spite of some discrepancies in the initial
phase, the time evolution of the main thermalhydraulic
parameters of the primary and
secondary systems calculated by ASTEC is, in
general, similar to that calculated by
CATHARE2. The largest discrepancy was found
in the pressurizer safety valve operation
modelling (fig. B1.37), which is much more
simplified in ASTEC than in CATHARE2.
In-vessel core melt progression and hydrogen
generation were evaluated with the DIVA
degradation module of ASTEC until corium
relocation in the lower plenum and lower head
vessel failure (fig. B1.38). A sensitivity study on
more uncertain core degradation parameters
highlighted their importance for in-vessel core
melt progression and hydrogen release. In
particular, in the fuel rod candling process,
relocation parameters, such as velocity and
minimal liquid fraction for the beginning of flow
down, may notably affect the timing and
amount of corium relocated in the lower head
and the in-vessel hydrogen mass produced.
QUENCH-11 post-test analysis. The boil-off QUENCH-11 experiment conducted at
Forschungszentrum Karlsruhe (FZK) was analysed with the ICARE/CATHARE code. At first, the
calculation was performed with the ICARE2 code in stand-alone mode for the participation in the
semi-blind QUENCH-11 benchmark promoted by the European Commission within the Severe
Accident Research Network (SARNET) project. Afterwards, the post-test analysis was carried out
using the more recent coupled version V2 of ICARE/CATHARE.
Temperature (K)
2000
1400
800
TFS 2/11
TFS 5/11
Tc3_75cm (icare2)
Tc3_75cm (IC_v2)
Reflood
Flow rate (kg/s)
0.0015
0.0009
0.0003
Boil-off phase
200
0 2000 4000 6000
Time (s)
Fig. B1.39 – Clad temperature at 0.75 m elevation
0
5400 5600 5800 6000
Time (s)
Fig. B1.40 – Hydrogen generation during reflood
ICARE 2 (green line), IC V2 (blue line), est. (red line)
Progress Report 2006
108

The ICARE2 and ICARE/CATHARE V2 codes were successfully applied in the post-test analysis of
QUENCH-11. Code-to-code result differences, depending on the thermal-hydraulic model used, were
pointed out and explained against experimental data. In spite of some deviations in the prediction of the
initial boiling rate and collapsed water level, in general, both codes simulate quite well the boil-off phase
(fig. B1.39). Both codes are also able to predict the large amount of hydrogen measured during reflood
(fig. B1.40): ICARE2 well predicts the total mass of hydrogen produced, but the timing of hydrogen
generation is notably delayed; whereas ICARE/CATHARE V2 predicts the timing of hydrogen release better
than ICARE2, but it underestimates the total hydrogen production. Finally, the sensitivity analysis with
ICARE/CATHARE V2 on some significant and uncertain code model parameters has highlighted the
importance of some code model parameters relative to hydrogen generation during reflood.
Spent fuel pool uncovery accident analysis. The consequences of an uncovery accident in an
irradiated fuel assembly during unload operations in the pool of the spent fuel building was simulated with
the ICARE/CATHARE code considering progressive pool draining, which is a more realistic scenario than
the instantaneous draining studied in 2005.
Two models were developed to simulate a water level decrease equal to 12.5 cm/min. One ("true draining")
considers that the system is initially filled with water and the progressive level decrease is obtained by
imposing as boundary condition a decrease in the pressure difference between the top and bottom of the
fuel assembly. The other ("water level simulated") takes into account the effects of the level decrease on
the fuel assembly and imposes a boundary condition on the surface of the fuel rods, which reproduces the
thermal transfer between the fuel rods and the water of the system (axial profile of the exchange coefficient
as a function of time).
The two models give very similar results. Figure B1.41 shows the axial temperature profiles in the fuel
assembly calculated with the two models during pool draining, 1000 s after the beginning of the accident.
The water level (true or simulated) is indicated by the arrow. The fuel assembly is completely uncovered
after 1850 s and the first temperature
escalation, driven by Zircaloy oxidation
under air atmosphere, occurs in the lower
part of the fuel assembly at around 1 m of
level, after about 4000 s of transient.
It is worth noting that minor code
“adjustments” to the calculations were
necessary with the true draining model in
order to obtain physical results during the
water level decrease. In particular, the lack
of a stratification model led to an erroneous
calculation of the heat transfer between the
structures and the fluid and, within the fluid,
between the liquid and the gas phase.
Modification of the dry-out criterion was
necessary to avoid non-physical behaviour.
Temperature (°C)
400
300
200
100
0
Row 1 (true draining)
Row 1 (water level simulated)
Row 2 (true draining)
Row 2 (water level simulated)
Row 3 (true draining)
Row 3 (water level simulated)
Row 4 (true draining)
Row 4 (water level simulated)
Row 5 (true draining)
Row 5 (water level simulated)
0 1 2 3 4
Elevation (m)
Fig. B1.41 – Temperature axial profiles 1000 s after the accident
beginning
Severe accident analysis
The severe-accident studies in progress within the SARNET project dealt with the following topics during
2006:
LOFT LP-FP-2 experiment analysis. The LP-FP-2 test, performed in the Loss-of Fluid Test (LOFT)
facility at the Idaho National Engineering Laboratory (INEL) USA to provide information on fuel rod
behaviour, hydrogen generation, and fission-product release during a loss-of-coolant accident scenario in
a pressurized water reactor (PWR) up to core reflood, was analysed with ASTEC V1 to assess the ability
of the code to simulate thermal-hydraulic conditions and core degradation phenomena. The ASTEC results
were then compared with the results of the ICARE/CATHARE code.
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B1 R&D on Nuclear Fission
B Fission Technology
ASTEC simulates reasonably well the transient phase of the experiment before the reflood phase,
that is, reactor system thermal-hydraulics, core uncovery and heatup, hydrogen generation and
fission-product release. The total hydrogen release is in good agreement with test measurements.
Instead the code needs some improvement in order to investigate the reflood phase because
temperature excursions and consequent heavy degradation of the fuel rods, hydrogen release and
primary pressure increase are not reproduced by ASTEC because of the inadequate modelling.
In general, the ICARE/CATHARE results confirm the validity of the ASTEC results.
MOZART experiment analysis. A preliminary comparison between the air oxidation model
actually implemented in the ICARE2 code that simulates the reaction kinetics between zircaloy and
oxygen with a parabolic law and the first isothermal experiments carried out in the MOZART facility
by IRSN and related to zircaloy-4 non-oxidized samples in the temperature range 800 to 1000°C
was performed for Work-Package WP9-3 (zircaloy oxidation by air and steam-air mixture).
Figure B1.42 shows calculated and measured (thermo-balance) mass gain vs time, at four different
temperatures (800, 900, 950 and 1000°C). The experimental data exhibit parabolic behaviour for a
very short time (roughly 30 min at 800°C). During this period, the calculated mass gain is
overestimated, except at
10000
1000°C. At this
temperature, the
experimental protocol
(iso thermal con ditions)
1000
cannot be completely
met because the
Mass gain/S (mg/dm 2 )
100
MOZART test 29 (1000°C)
MOZART test 30 (1000°C)
MOZART test 31 (1000°C)
ICA/CATH (1000°C)
MOZART test 33 (950°C)
MOZART test 34 (950°C)
ICA/CATH (950°C)
MOZART test 39 (900°C)
MOZART test 40 (900°C)
ICA/CATH (900°C)
MOZART test 44 (800°C)
MOZART test 45 (800°C)
ICA/CATH (800°C)
10
1 10 100 1000
Time (min)
Fig. B1.42 – Measured and calculated O 2 mass gain
totally inadequate to predict the mass gain (underestimation of the reaction kinetics).
oxidation power
produces a not negligible
temperature peak at the
beginning of the test.
After the loss of parabolic
behaviour (post breakaway
period), experi -
mental data indicate a
continuous increase in
the oxidation kinetics and
the code model becomes
The model limitations in the simulation of post break-away oxidation may be more or less important,
depending on the expected temperature evolution during accidental transients. However an
improvement in the code model to take into account the post break-away behaviour is necessary
to simulate uncovery accidents in the spent fuel pool, as the temperature increases gradually and
most oxidation occurs in the post break-away kinetics regime.
ASTEC reactor application and benchmarking. The work carried out in 2006 concerned a)
benchmarking of ASTEC V1.2 R1 and MELCOR 1.8.6. based on the accident reactor sequence H2
and b) identification of the most critical parameters and variables influencing the code response,
mainly for in-vessel processes. This activity was shared with AREVA-NP SAS and IRSN. AREVA
used the MAAP code for benchmark and successive comparisons.
Details explaining the differences that emerged during model comparison are briefly reported. The
main difference was found in some corium processes, mainly in candling. The candling model in the
MELCOR “COR” package is semi-mechanistic and refers to the downward flow of molten core
materials and subsequent refreezing of these materials as they transfer latent heat to cooler
structures below. ASTEC uses a different model and so some differences are now clearly
understood.
Progress Report 2006
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Concerning hydrogen production, at the moment there are still some unexplained questions. There are not
negligible gaps between the results provided by the codes involved in the benchmark. Very recently ENEA
made a simple comparison referring to the table with the timing of the main events of accident sequence
H2 and found very strong differences between new and old MAAP calculations, between ENEA-ASTEC
and AREVA-ASTEC calculations and between the latest MELCOR and MAAP calculations. Probably MAAP
and MELCOR users followed completely different approaches in modelling the main processes occurring
during corium production and mass relocation; perhaps they used different values in the most
representative coefficients governing some relevant equations.
Concerning the water inventory strong differences were found for water in the primary and secondary
circuits given by ASTEC and MELCOR calculations, due to a totally different approach of calculation inside
the codes. So far, code benchmarks have been performed without well-defined boundary and initial
conditions, as generally made (imposed) during the OECD ISPs. For this reason a new benchmark clearly
defining a list of still open issues has been recommended, with also a uniform protocol for calculations.
Reactor safety source-term activities. An assessment of UO 2 vapourisation in different atmospheres
was performed against some experiments conducted at Berkeley University [B1.23] with the aim of testing
the fuel oxidation/vapourisation model implemented in ELSA [B1.24], which is the fission product release
module of the European reactor ASTEC [B1.25].
The experiments were modelled as a simple bare UO 2 fuel mass inside a gas flow channel. As the model
departs from a fixed value of the equilibrium stoichiometrical deviation, the initial phase of experiments
leading to fuel oxidation was neglected and only the volatilisation phase due to steam ingress was
reproduced. Two steam-flow values of of 200 and 50 ccm were considered. The results show negligible
differences in vapourisation rates. This is in good
agreement with the fact that the ELSA vapourisation
model slightly depends on the inlet gas rate and on
the composition of the career gas. The calculated
percentage of volatilised mass notably increases with
temperature, which is further confirmation of the
code capability to correctly calculate vapourisation
rates.
Comparison of code results with the experimental
data normalised to mass fraction release, as given by
the code, shows reasonably good agreement
between calculation and data (fig. B1.43), thus
providing further confidence in the adequacy of the
fuel volatilisation modelling in ASTEC.
Volatilisation rate
1×10 -7
1×10 -8
Volatilisation rate of uranium in pure steam
at flow rate 200 ccm
Exp. (normalised to
fractional volat. rate)
1×10 -11
1×10 -9
Exp.
(mol/s/cm 2 )
1×10 -10
ASTEC-DIVA
(fractional volat. rate)
5 5.5 6 6.5 7
10 4 / T (1/K)
Fig. B1.43 – DIVA calculations compared with Hashizume
experimental data
Reliability and risk analysis
The following reliability and risk activities are performed within programmes promoted by international
organisations.
Ageing probability safety assessment. An official agreement has been signed between ENEA and the
Institute for Energy (IE) of the Joint Research Centre (JRC) of the European Commission, in Petten,
[B1.23] K. Hashizume et al., J. Nucl. Mater. 275, 277-286 (1999); and N. Davidovich, Validation of the ASTEC code fuel volatilization model on
the Hashizume et al. experiments - SARNET-ST-P20 (2006)
[B1.24] W. Plumecocq and G. Guillard, ELSA 2.1, ASTEC-V1/DOC/04-02 (2002); and N. Davidovich, Progress on synthesis modelling of UO 2
oxidation - SARNET-ST-P5, ENEA Internal Report FIS–P9G1–001 (2005)
[B1.25] W. Plumecocq and G. Guillard, ASTEC V1.2 code ELSA module, ASTEC-V1/DOC/05-06 (2006)
References
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B1 R&D on Nuclear Fission
B Fission Technology
Netherlands, for participation in an international collaboration denoted as Network on Incorporating
Ageing Effects into Probabilistic Safety Assessment. The network is devoted to developing reliability
and availability models of systems and components, incorporating the effects of aging, and is
expected to last till 2009.
During 2006 a study on the impact of ageing on passive systems and their performance was
undertaken, in addition to the definitions of basic ageing probabilistic safety assessment (APSA)
reliability models.
Passive system reliability. The activities performed mainly addressed issues related to passive
systems relying on natural circulation to accomplish their functions:
• development of an approach for integrating the passive systems within an accident sequence in
combination with active systems and human actions in a probabilistic risk assessment (PRA)
framework, based on fault tree and event tree techniques;
• development of a preliminary reliability physics model based on the fracture mechanics approach
to get the performance bounds to meet the reliability targets;
• evaluation of uncertainties associated with passive system reliability;
• risk study of a decay heat removal system based on failure mode, effects and critical analysis
(FMECA);
• participation in the IAEA Co-ordinated Research Project (CRP), denoted as “Natural circulation
phenomena, modelling and reliability of passive systems that utilise the natural circulation”,
launched in 2004. In this framework an activity aimed at the reliability assessment of the
Argentinean integral-type CAREM-like reactor passive features has been undertaken and results
are expected at the beginning of the 2007.
B1.4 Nuclear Data
General quantum mechanics
Scattering by PT-symmetric non-local potentials. Non-local potentials play an important role in
many applications of quantum scattering theory. In nuclear physics, they naturally arise from the
convolution of an effective nucleon-nucleon interaction with the density of a target nucleus. In
particular, a solvable non-local potential was proposed by Yamaguchi in 1954 in order to describe
bound and scattering states of the proton-neutron system. The present study was focussed on the
scattering properties of a PT-symmetric 1D version of the Yamaguchi potential, i.e., a non-Hermitian
potential invariant under the product of the parity operator P and the time reversal operator T, but
not under the separate actions of P and T: the transmission and reflection coefficients are worked
out by the Green’s function method and show aspects of unitarity breaking quite different from those
of PT-symmetric local potentials. The method of solution can be applied to large families of non-local
potentials with separable kernel and different behaviour under P and T transformations.
Group theory approach to transparent potentials. One-dimensional potentials with
transmission coefficients equal to one over the whole real axis occur in several domains of general
quantum mechanics: for instance, non-trivial reflectionless potentials can be derived by
supersymmetric techniques from the null potential, which is trivially reflectionless, or they can be
extracted by Lie-algebraic methods from the Casimir invariants of some non-compact groups. In the
present study the latter technique was applied to derivation of the general form of real potentials
appearing in Hamiltonians with underlying so(2,2) symmetry, which permits the solution of the
corresponding Schrödinger equation in terms of hypergeometric functions. The six-generator
so(2,2) algebra admits several decomposition chains and the corresponding potentials are, in
general, not transparent: reflectionless potentials are obtained in the so(2,2) → so(2,1) → so(2)
reduction chain when the solutions belong to discrete series representations of the so(2,1) sub-
Progress Report 2006
112

algebra appearing in the reduction. Hyperbolic potentials of the Pöschl-Teller type belong to this class. The
Inönü-Wigner contraction of so(2,2) to the pseudo-euclidean algebra e(2,1) yields solutions that are always
connected with reflectionless potentials. For the sake of simplicity, but without loss of generality, the
general form has been worked out for reflectionless potentials appearing in Hamiltonians with underlying
e(1,1) symmetry, where e(1,1) is a three-generator sub-algebra of e(2,1). The well-known reflectionless
potential V(x) ~ 1/x 2 belongs to this class.
Nuclear reaction theory and experiments
Neutron-induced fission of light actinides. Within
the work programme of theoretical activities of interest
to the n_TOF collaboration, a model has been
proposed to describe the coarse-grained resonant
structure in neutron-induced fission of light actinides at
sub-barrier excitation energies. The fission barriers are
either two-, or three-humped, depending on the
fissioning nucleus, and have an imaginary component
in the second (isomeric) well, simulating a partial
damping of class II vibrational states, while class III
states, corresponding to excitations in the third well,
are not damped. The sets of discrete transition states
include rotational bands built either on vibrational states
or on non-collective states. In the present
phenomenological version of the model, energies and
quantum numbers of transition states are not evaluated by means of a nuclear structure model, but are
adjusted on the experimental (n,f) cross sections, which can thus be reproduced with great accuracy, as
shown in figure B1.44, relative to the first-chance fission of 232 Th.
The present fission model has been incorporated in Version 19 (Lodi) of the EMPIRE-II code of nuclear
reactions, freely distributed by the National Nuclear Data Center, Brookhaven National Laboratory.
Measurements of neutron-capture cross sections at the n_TOF facility at CERN. After the end of
the experimental campaign in 2004, the two subsequent years were dedicated to analysis of capture
cross-section measurements and to publication of related papers, such as those on 232 Th(n,γ) in the
unresolved resonance region up to 1 MeV, 151 Sm(n,γ) in the energy range from 0.6 eV to 1 MeV, and
209 Bi(n,γ) in the resolved resonance region, already summarised in the ENEA UTS FIS 2005 Progress
Report. The new analysis completed and published in 2006 concerns the 207 Pb(n,γ) reaction in the
resolved resonance region. The measurement was performed with an optimised set up of two C 6 D 6
scintillator detectors, which permits reduction of scattered neutron background down to a negligible level,
by using the pulse height weighting technique. Resonance parameters and radiative kernels were
determined for 16 resonances in the neutron energy range from 3 to 320 keV. Good agreement with
previous measurements is found at low energies, while substantial discrepancies appear beyond 45 keV.
Maxwellian averaged cross sections were determined with an accuracy of ± 5%.
Cross section (barns)
0.15
0.10
0.05
0
Current work
2002 Shcherbakov
1991 Fursov
1986 Kanda
1983 Meadows
1982 Behrens
1978 Blons
1975 Blons
1.0 1.5 2.0 2.5
Incident energy (MeV)
Fig. B1.44 – 232 Th(n,f) near the fission threshold.
Solid line: present work. Experimental data are
taken from EXFOR
Nuclear data processing and validation
The cooperation between the ENEA Nuclear Data Group and the Organisation for Economic Co-operation
and Development/Nuclear Energy Agency (OECD/NEA) Data Bank (Issy-les-Moulineaux, France)
continued, in particular, within the Joint Evaluated Fission and Fusion (JEFF) Working Group on Benchmark
Testing, Data Processing and Evaluations. Several technical feedbacks were notified, dedicated to the
JEFF-3.1 European evaluated data files and their related processing through the NJOY nuclear data
processing system. A valuable collaboration with a specialist formerly working at the Institute of Physics
and Power Engineering (IPPE) Obninsk (Russian Federation) has been continued and extended.
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B1 R&D on Nuclear Fission
B Fission Technology
VITJEFF31.BOLIB and MATJEFF31.BOLIB generation. The VITJEFF31.BOLIB and
MATJEFF31.BOLIB coupled n-γ multi-group cross-section libraries for nuclear fission applications in
the VITAMIN-B6 American library energy group structure (199 neutron groups + 42 photon groups)
were completely reprocessed by means of a version of the NJOY-99.112 nuclear data processing
system, modified at ENEA. The present libraries, based on the JEFF-3.1 European evaluated
nuclear data files, were previously produced through the NJOY-99.90 system, but it was decided to
reprocess them completely after a detailed analysis of the list of modifications introduced by the
author in the recent NJOY–99.112 version. The THERMR and GROUPR modules of this latter
version were further modified at ENEA. In particular, a correction patch was prepared and
introduced in the THERMR module of NJOY-99.112 in order to solve a problem that emerged in the
processing of the JEFF-3.1 bound nuclides C (graphite) and Be (beryllium metal), where infinite loop
calculations were generated. A second relevant correction patch was prepared for the GROUPR
module of NJOY–99.112. The OECD/NEA Data Bank checked this patch and got positive results
and then diffused it freely. This initiative was taken in order to extend the group-wise data processing
capability to the evaluated data files including non-Cartesian interpolation schemes for secondary
neutron energy distributions (MF=5). The ENEA Nuclear Data Group proved that 69 JEFF-3.1
evaluated files could not be processed correctly through the GROUPR module of NJOY-99.112 or
through all previous NJOY versions officially released, as communicated to the NJOY User Group.
This set of data files contains, in particular, secondary neutron energy distributions (MF=5),
presented as arbitrary tabulated functions (LF=1) with the non-Cartesian unit base interpolation law
INT=22. The GROUPR module cannot process correctly the mentioned evaluated files because the
GETSED subroutine cannot deal with secondary neutron energy distributions with non-Cartesian
interpolation schemes (INT=11-15 and INT=21-25). Thus, the group-to-group scattering matrices
for the MT=16, 17, 22, 28, 32, 33, 91 reactions could not be produced in the GENDF output cross
section files of the 69 evaluated data files under consideration. The GROUPR problems described
above were autonomously identified, starting from analysis of unacceptably underestimated K eff
results obtained with criticality neutron transport calculations, performed through the XSDRNPM 1D
discrete ordinates module of the SCAMPI data processing system. Two ICSBEP (2004 Edition) fast
criticality benchmark experiments with 233 U (included in the previously cited 69–file set) were
simulated. The results obtained with the XSDRNPM code in the P5-S16 approximation were
obtained from JEFF-3.1 data, processed differently with the original GROUPR module of
NJOY–99.112 and with the GROUPR version modified at ENEA into the 199 neutron energy group
structure of the VITAMIN-B6 library. The results obtained with these deterministic transport
calculations were compared with the results obtained through the MCNP-4C Monte Carlo code
using JEFF-3.1 continuous-energy cross-section sets.
181 materials were processed: 175 for standard isotopes or natural elements and 6 for bound
nuclides. In the last group of materials, in particular, the H-Zr material was added to the set of 5
bound nuclide materials contained in the VITAMIN-B6 library. Only one material ( 46 Ca from
JEFF–3.1) could not be processed correctly. ENEA notified the OECD/NEA Data Bank of the fact
that the total and elastic cross-section values of the first officially released version of this 46 Ca file
below 1 keV, i.e., in the energy range 1.0×10 -05 - 1.0×10 03 eV, are set to zero, while capture crosssection
values differ from zero.
SCAMPI revision and updating. Many corrections and modifications were required for several
modules of the SCAMPI data processing system in order to process the JEFF-3.1 data for the
VITJEFF31.BOLIB library. In particular, the AJAX, MALOCS and SMILER modules were corrected.
The most interesting modification was made to SMILER and MALOCS in order to take into account
also the delayed component part (MF=5 and MT=455) of the fission spectrum, needed to obtain,
e.g., more correct results in fixed-source transport calculations. On the contrary, the original version
of SMILER can read only the prompt component (MF=6 and MT=18).
The following versions of the MALOCS module were compared, as taken from the SCAMPI nuclear
data processing and SCALE nuclear safety calculation systems:
• original version of MALOCS in the SCAMPI distributed by OECD/NEA Data Bank;
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• version of MALOCS/SCAMPI as modified by ENEA, called MALOCS/SCAMPI Bologna version;
• original version of MALOCS included in SCALE-4;
• original most recent updated version of MALOCS included in SCALE-5.
From the performance and feature comparison of the versions of MALOCS included in the SCAMPI,
SCALE-4 and SCALE-5 systems, the following conclusions were drawn:
• MALOCS/SCAMPI, MALOCS/SCALE-4 and MALOCS/SCALE-5 exclude the possibility of fission matrix
collapsing.
• MALOCS/SCALE-4 and MALOCS/SCALE-5 include the possibility to truncate the up-scatter crosssection
terms with options IOPT7=0, 1, 2, 3.
• MALOCS/SCAMPI includes only IOPT7=0.
• MALOCS/SCALE-5 is similar to MALOCS/SCALE-4, but it is rewritten in FORTRAN-90.
MALOCS/SCAMPI Bologna version includes the possibility of fission matrix collapsing and permits
truncation of the up-scatter terms with options IOPT7=0, 1, 2, 3. Taking into account both these
conclusions and the fact that the SCAMPI system includes functional modules all programmed in
FORTRAN-77, as for the MALOCS/SCAMPI Bologna version, it was preferred to avoid any potential
inconsistency in programming languages; therefore, this version was selected for the production of the
new BUGJEFF31.BOLIB collapsed working library from the multi-group general-purpose
VITJEFF31.BOLIB library in AMPX format. The GENDF cross-section files, obtained through a modified
version of GROUPR in NJOY-99.112, were used to generate VITJEFF31.BOLIB and MATJEFF31.BOLIB.
Extensive validation of the VITJEFF31.BOLIB library was performed through simulation of the same thermal
and fast-neutron criticality benchmarks, already prepared for VITJEF22.BOLIB. The results obtained with
the XSDRNPM 1D transport module of SCAMPI were compared with the results of Monte Carlo
calculations using the MCNP-4C code.
BUGJEFF31.BOLIB. Two preliminary versions (with and without up-scatter) of the cross-section working
library BUGJEFF31.BOLIB were collapsed from the VITJEFF31.BOLIB library in AMPX format, generated
with the ENEA modified version of NJOY-99.112. This collapsing work was done by means of the ENEA
revised SCAMPI system and, in particular, the modified version of the MALOCS module. The
BUGJEFF31.BOLIB working library for shielding and light water reactor (LWR) pressure vessel dosimetry
applications has the same group structure (47 n + 20 γ) and general features as the BUGLE-96 American
library. To complete the response function cross section collapsing in the BUGLE-96 neutron group
structure (47 n) from the most recent IAEA Reactor Dosimetry File IRDF-2002, a new tabulated weighting
function was obtained from XSDRNPM calculations in the 1/4T (T=PWR pressure vessel thickness) spatial
position, using the VITJEFF31.BOLIB multi-group library. The calculation chain was completely prepared
but, before starting the collapsing procedure to generate the final working library, further investigation will
be necessary to identify the inconsistencies and inaccuracies of the BUGLE-96 input data, which emerged
in 2005 in the ENEA feasibility analysis for a BUGLE-type library generation.
Computer code development
BOT3P is a set of standard FORTRAN-77 language codes developed by the ENEA Nuclear Data Group in
1997. The BOT3P Version 1.0 was originally conceived as a set of standard FORTRAN-77 language
programmes in order to give the users of the DORT and TORT deterministic transport codes (both included
in the Oak Ridge National Laboratory [ORNL USA] DOORS package) some useful diagnostic tools to
prepare and check their input data files for both Cartesian and cylindrical geometries, including mesh grid
generation modules, graphical display and utility programs for post-processing applications. Later versions
extended the possibility to produce the geometrical, material distribution and fixed neutron source data to
other deterministic transport codes such as TWODANT/THREEDANT (both included in the Los Alamos
National Laboratory [LANL] USA DANTSYS package), PARTISN (the updated parallel version of DANTSYS)
and the sensitivity code SUSD3D (distributed by the OECD/NEA Data Bank, Issy-les-Moulineaux, France)
and, potentially, to any transport code through BOT3P binary output files that can be easily interfaced (see,
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B1 R&D on Nuclear Fission
Fig. B1.45 – Simulation in Cartesian coordinates of a
complex geometry (120X, 88Y, 200Z)
B Fission Technology
e.g., the case of the 2D and 3D discrete-ordinates
neutron, photon and charged particle transport
codes KASKAD-S-2.5 and KATRIN-2.0, developed
at the Keldysh Institute of Applied Mathematics
Moscow, Russian Federation). Since BOT3P binary
output files can be easily interfaced, users can
potentially produce the geometrical and material
distribution data for any transport code starting from
the same BOT3P input. This makes it possible to
compare directly for the same geometry the effects on
transport analysis results, which stem from the use of
different data libraries and solution approaches.
BOT3P Version 5.1 was completed in 2006 and has been freely available from the OECD/NEA Data
Bank (F) since August 2006. This new version contains important additions specifically addressing
radiation transport analysis for medical applications. The new module CATSM allows users to
reduce the geometrical size of problems related to processed (already interpreted by physicians or
by proper software) computerised (axial) tomography (CT/CAT) scans with or without small detail
loss with respect to the original voxelized geometry. This permits problem sizes that can be more
easily managed by transport codes. CATSM can automatically generate tetrahedron mesh grids,
too, starting from the input voxelized geometry, even though the implemented algorithm is still rather
rough and to be improved in the future. BOT3P-5.1 contains new graphics capabilities that enable
users to visualise tetrahedron mesh grids in 3D and 2D cuts. As from Version 5.0, a general method
to conserve mass of geometrically complex material zones simulated on both Cartesian and
cylindrical mesh grids was implemented. BOT3P allows users to specify as refined a computation
as desired of the possible area/volume error of material zones due to the stair-cased geometry
representation, and automatically corrects material densities in order to conserve masses globally.
BOT3P can store on binary outputs the detailed material zone distribution map inside each cell of
the mesh grid, according to a sub-mesh grid refinement defined in input by the user and the
area/volume fraction distribution of the different material zones contained in meshes at zone
interfaces. This procedure allows a local (per cell) density correction as an alternative to the
approach of a uniform density correction on the whole zone domain and potentially makes it
possible to perform material zone homogenisation locally and transport analyses with more
accuracy. BOT3P allows users to model X-Y, X-Z, Y-Z, R-Θ and R-Z geometries in two dimensions
and X-Y-Z and R-Θ-Z geometries in three dimensions. BOT3P was successfully used not only in
some complex neutron shielding and criticality benchmarks, but also in power reactor applications
(Westinghouse AP1000 internals heating rate distribution calculations by Ansaldo Nucleare). BOT3P
is designed to run on most Linux/UNIX platforms. The plot of figure B1.45 gives an idea of the
complex modelling capabilities of BOT3P.
Radioactive ion-beam production for nuclear-structure studies
Intense neutron-rich isotope beams open many new fields of investigation, such as nuclear-structure
studies, in a yet unexplored region. Several laboratories are trying to produce high enough intensities
to warrant a new generation of experiments. The Study for the Production of Exotic Species (SPES)
project is an accelerator-based facility for the production of intense neutron-rich radioactive ion
beams, in the range of masses between 80 and 160. SPES is a new-generation ISOL facility
proposed in Italy at the Istituto Nazionale di Fisica Nucleare, Laboratori Nazionali di Legnaro (INFN-
LNL), able to represent a competitive intermediate step between the existing facilities and the longer
range high-performance EURISOL.
Progress Report 2006
116

Fig. B1.46 – Target configuration for the SPES project
Window
UCx disks
Carbon dump
The target system is one of the key issues for
such facilities. A target configuration has
been developed, in an ENEA/INFN-LNL
collaboration, consisting of a 40-MeV proton
beam (0.2 mA) directly impinging on the
fission materials, composed of uranium carbide (UC x ). The 238 U fission fragments constitute the source for
the exotic beams and, in order to extract them, the target is placed inside a graphite box at 2000°C. The
target is split into several thin disks to allow cooling of the system by thermal radiation (fig. B1.46). In this
way ∼10 13 fissions s -1 are obtained with a relatively simple system and at relatively low cost. All the main
parameters of the system have been analysed by means of calculation codes: the fission rates and fission
fragment distribution; power deposition and the thermo-mechanical behaviour of the disks.
Proton
beam
B1.5 TRIGA RC-1 and RSV TAPIRO Plant-Operation for
Application Development
The availability of the TRIGA RC-1 and RSV TAPIRO plants has permitted ENEA to acquire solid experience
in the development and management of research nuclear reactors and their application in programmes
that use ionizing radiation sources, in particular for the qualification of radiation damage to materials. With
the qualified TRIGA RC-1 neutron beams it is possible to develop highly technological neutron radiography
and tomography techniques by means of thin scintillator films in lithium fluoride. The neutron tomography
system located on the thermal column of the TRIGA reactor has been maintained in operation as a
propaedeutic to the installation of a new collimator in the TRIGA tangential channel to improve the L/D ratio
in a neutron flux of 10 8 n cm -2 s -1 .
The TRIGA RC-1 and TAPIRO reactors have been proposed as experimental support to the newgeneration
nuclear reactors that are nearing commercialisation (e.g., AP 1000) in order to check critical
components under thermal and fast neutron flux. In fact, the TRIGA core flexibility permits installation of an
experimental loop to continuously verify component performance under irradiation.
During 2006 the TRIGA and TAPIRO reactors operated for about 2000 h. The TAPIRO irradiation column
was also modified to permit boron neutron capture therapy (BNCT) for human brain tumour and melanoma
applications (see sect. B2.1).
117
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B2 Medical, Energetic and Environmental
Applications
B Fission Technology
B2.1 Boron Neutron Capture Therapy
In 2006 construction of the epithermal column EPIMED at the TAPIRO experimental nuclear reactor
was completed. The irradiation bunker to be used for beam characterisation was constructed and
the doses outside the bunker, both in and outside the reactor hall, were measured and compared
with calculations. A network of national groups involved in measuring the beam has been
established to coordinate the beam characterisation. Support equipment for BNCT clinical trials has
been acquired from the Swedish BNCT project (Hammercap S.p.A.).
The collaborative activity with the Study and Production of Exotic Species (SPES)-BNCT project of
the Legnano National Laboratory (LNL) of INFN and the University of Padua on using the thermal
column HYTHOR at TAPIRO in radiobiological and micro-dosimetric studies continued. HYTHOR
was also used for film irradiation in a collaboration with the University of Bremen (Germany) and for
the development of gel dosimeters (University of Milan). A collaboration with INFN Pavia and cofinanced
by the Ministry of Higher Education and Research (MIUR) was launched to study the
application of BNCT to lung tumours.
Design of a graphite configuration in the thermal column of the TRIGA reactor is under way. The
objective is to repeat the clinical experimentation carried out on an explanted liver at Pavia. Extensive
support in this activity has been provided by INFN Pavia.
The epithermal column EPIMED at TAPIRO
Human tissue has a relatively high tolerance to epithermal neutrons (in the BNCT context between
about 1 eV and 10 keV) which, unlike thermal neutrons, are able to penetrate some centimetres into
the tissue. However as the energy increases into the tens and hundreds of keV region the tolerance
Normalised neutron flux/unit
lethargy (cm-2 s-1)
1×10 0
1×10-2
"Epithermal" spectrum
Flux-to-Sievert rf (ICRP74)
1×10-4
1×10-8 1×10-6 1×10-4 1×10-2 1×10 0 1×10 2
Energy (MeV)
Fig. B2.1 – Comparison of epithermal neutron spectrum at TAPIRO
with a flux-to-sievert conversion factor
strongly decreases. Figure B2.1
compares the epithermal neutron
spectrum at TAPIRO with the
ICRP74 flux-to-sievert conversion
coefficient. Although this coefficient
is for stochastic doses, whilst in
therapy much higher systematic
doses are involved, this comparison
illustrates the critical importance of
designing and measuring the
neutron spectrum.
The different phases of assembling
the moderator and reflector are
shown in figure B2.2. The mounted
Progress Report 2006
118

Fig. B2.2 - Mounting the moderator and reflector
of EPIMED in TAPIRO
column outside and inside the
reactor (in the latter case
without the lithiated
polyethylene end neutron
shield) is shown in figure B2.3.
EPIMED provides a neutron
beam that directly enters the
reactor hall. The necessary
beam shielding consists of
Fig. B2.3 – The mounted column outside and
firstly a bunker of limited
inside the reactor
volume appropriate for beam
characterisation with the
reactor operating at a maximum 10% of nominal power and secondly an
irradiation room for patient therapy with the reactor at nominal power
(5 kW). The bunker has been designed and constructed and the doses
around the bunker in the reactor hall as well as outside the reactor hall
have been measured and compared with the predicted values.
Figure B2.4 shows a plan diagram of the bunker together with access maze and the shielding placed
outside the reactor hall. As the present shielding configuration is temporary, to save money and time, it has
been necessary to establish an exclusion zone outside the reactor hall in the direction of the neutron beam
(fig. B2.4).
The bunker shielding is composed of
standard (assumed density
2.3 g cm –3 ) 50-cm-thick concrete
blocks (so referring to figure B2.4 there
are two lines of concrete of total
thickness 1 m in the direction of the
control room). In addition the inner
walls of both the bunker and the first
part of the access maze are lined with
Lawn
Fence
External
concrete
shielding
Sliding door
Reactor hall
Entrance maze
bunker
43
41
42
Fig. B2.4 – Plan of bunker, access maze and
shielding outside the reactor hall and
exclusion zone (showing some of the points
used for dose comparison)
Cooling
system
room
Control
room
Air-lock
119
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B2 Medical, Energetic and
B Fission Technology
Fig. B2.5 – Construction of the
characterisation bunker and access
maze
Fig. B2.6 – The completed characterisation bunker and access maze
borated polyethylene to reduce the production of
high-energy prompt gamma rays as well as
activation. The ceiling is borated polyethylene just
over 10 cm thick covered by a thin layer of iron. As
a result, gamma rays from neutron capture in 10 B
provide quite high doses above the ceiling.
However at ground level the doses resulting from
reflection of these gammas from the walls and roof
of the reactor hall are relatively low. Figure B2.5
shows various stages in the construction of the
bunker, whilst figure B2.6 shows the completed
bunker.
The measured and calculated doses outside the
bunker are in reasonable agreement. As an
example, from [B2.1], figures B2.4 and B2.7 report
15
12
14 13
(SR1)
11
Reactor hall
9-10
7
6
8
(SR2)
2
3
4
5
1
Fig. B2.7 – Points for dose comparison, measurement/
calculation (see also fig. B2.4)
Control room
Progress Report 2006
120

Environmental Applications
Table B2.I – Comparison of selected measured and
calculated doses (see also figs. B2.7 and B2.4)
Point Gamma Dose (µSv/h) Neutron Dose(µSv/h)
Meas. Calc. Meas. Calc.
1 1.3 1.2 1.3 1.2
2 1.2 0.68 1.1 0.59
3 3.3 1.2 3.3 1.0
4 2.3 1.7 1.5 1.1
5 1.6 1.2 1.6 0.95
6 5.3 1.8 1.6 0.55
7 35.3 7.3 1.6 1.0
8 0.9 0.79 1.3 1.0
9 1.7 3.1 3 1.8
10 39 2.6 2.2
11 32 15 26 44
12 21 6.9 5 2.0
13 55 17 9 22
14 48 25 10 28
15 2.6 1.4 1.4 1.8
41 1.2 0.44 0.3 0.14
42 0.6 0.47 0.2 0.11
43 0.9 0.34 0.1 0.10
µSv/h
5.000E+2
Fig. B2.8 – Calculated total dose map in the reactor hall
at about 1 m above the floor
comparisons for selected dose points within
the reactor hall and outside the hall at the
margins of the limited access. The
comparison between calculated and
measured doses is shown in table B2.I.
Where the comparison is poor (e.g., point 7) the reason is known (in this case an unavoidable averaging
of the calculated dose over a volume that includes the concrete shielding as well as air). Having established
a degree of confidence on the calculated doses, it is possible consider the dose maps (e.g., in figure B2.8)
of the total (neutron and gamma) doses in the reactor hall at about 1 m above the floor.
Employment of the thermal column HYTHOR at TAPIRO
The thermal neutron experimental facility HYTHOR, designed by the LNL-INFN Padua, has been used by
LNL to carry out micro-dosimetric studies by means of specially designed tissue-equivalent proportional
counters (TEPCs). HYTHOR has also been utilised by the University of Padua for mouse irradiation in the
context of research into boron compounds for skin melanoma. Again in the BNCT framework, films for
neutron capture radiography have been irradiated in collaboration with the University of Bremen.
In collaboration with the Department of Physics of Milan University, Monte Carlo calculations were
compared with experimental results by means of gel dosimeters in order to investigate a) the spatial
distribution of the gamma dose and thermal neutron fluence and b) the accuracy at which the boron
concentration should be estimated in an explanted organ of a BNCT patient.
Study of BNCT applied to lung tumours
A collaboration has started with the multi-disciplinary group at INFN Pavia, previously involved in the
treatment of the explanted liver, to study the application of BNCT to lung tumours. Calculations concerning
[B2.1]
K.W. Burn, L. Casalini, E. Nava, Confronto tra calcoli e misure relativo al monitoraggio d’area del reattore TAPIRO per la caratterizzazione
della nuova colonna epitermica (EPIMED), ENEA Internal Report in preparation
References
121
Progress Report 2006

B2 Medical, Energetic and
Fig. B2.9 – Plan view a) and vertical cross section b) of the
TRIGA reactor core and thermal column with phantom
B Fission Technology
Fig. B2.10 – Plan view of the TRIGA reactor
analytical and voxel phantoms are being carried out,
employing the neutron beam from EPIMED.
Measurements will be made on a lung model
phantom placed in the EPIMED beam.
Design of a facility at TRIGA to treat
explanted livers
This activity was initiated in 2006 with the support of
INFN Pavia. A different thermal column [B2.2] to the
original one used for liver irradiation [B2.3] was
suggested by Pavia. The modified configuration has
the advantage of not requiring the liver to be rotated
by 180° during treatment. The MCNP model of TRIGA
at ENEA Casaccia [B2.4] was improved in the vicinity
of the thermal column and all the radial and tangential
experimental ducts were included. The new thermal
column design [B2.2] was incorporated with a liver
phantom [B2.3] and container (including a partial
screening layer of lithium fluoride to harden the
neutron spectrum) supplied by Pavia. The resulting
configuration is shown in figures B2.9 and B2.10.
The distribution of the therapeutic 10 B dose in the
phantom was calculated to verify that the graphite
configuration (fig. B2.9) together with the lithium
fluoride spectrum modifier gave a therapeutic dose
covering the whole phantom. An example of such a
distribution is shown in figure B2.11 (from the Pavia
TRIGA model, courtesy of the Department of Nuclear
and Theoretical Physics, University of Pavia and INFN,
Pavia). Figure B2.12 shows an axial profile of the
therapeutic dose along the central axis.
Dose (arb.units)
12
8
4
Boron dose distribution 2 nd z mesh
Fig. B2.11 – Typical qualitative distribution of
therapeutic dose within liver phantom (courtesy
of Department of Nuclear and Theoretical
Physics, University of Pavia and INFN, Pavia)
0
10
5
0
-5
5
10
-10 -10
-5 0
Axial mesh No
Radial mesh No
Progress Report 2006
122

Environmental Applications
Fig. B2.12 – Typical profile of the 10 B dose in the phantom along the axis
of the thermal column
28
26
Whilst results for the distribution of the dose within the phantom
agree quite closely with those obtained at the Pavia TRIGA, the
absolute values of the doses differ considerably. The differences
arise because lead is used as a gamma shield at Casaccia, while
bismuth is used at Pavia, and masonite is present in the thermal
column at Casaccia (some modelling differences are also evident).
A first conclusion is that the therapeutic doses at Casaccia will not
be very much larger than those at Pavia, notwithstanding the fact
that the nominal reactor power is four times higher at Casaccia than
at Pavia.
Dose (arb. units)
24
22
20
18
16
14
410 415 420 425 430 435
Axial mesh No
B2.2 Solar Thermal Energy
Experimental activities under the Solar Thermal
Energy Project are aimed at evaluating the
behaviour of several kinds of structural material
in stagnant molten nitrate environments. In
particular, the corrosion mechanism/rate and
weld resistance have been evaluated.
The experimental facility (fig. B2.13) is made up
of eight crucibles (fig. B2.14), coated inside with
Ti, in which the specimens are inserted (or
extracted) by means of a special device
(fig. B2.15). Each crucible is equipped with its
own heater and electronic control system for
monitoring and acquiring pressure and
temperature.
Fig. B2.13 – View of the experimental facility, special device
for introducing/extracting specimens and the monitoring
acquisition system
Tests were carried out on austenitic steel (AISI
321H). In selecting the specimen geometry the
type of test was taken into account: simple
rectangular slabs for the corrosion mechanism
tests and corrosion rate evaluation; rectangular
welding slabs for evaluation of the tungsten inert
gas (TIG) weld resistance in molten salt. The
tests were performed in a nitrate mixture of
sodium and potassium (40 wt.% NaNO 3 –
60 wt.% KNO 3 ).
Fig. B2.14 – External and internal view of the crucibles
[B2.2] S. Bortolussi, Neutron flux distribution in liver at the Pavia reactor, presented at the Workshop on Innovative Treatment Concepts for
Liver Metastases (University Hospital Essen 2006)
[B2.3] S. Bortolussi, TAOrMINA: una originale configurazione del campo neutronico per una migliore uniformità della dose nell’organo
espiantato, degree thesis, Department of Physics, University of Trieste (2002-2003)
[B2.4] N. Burgio et al., MCNP model of the 1 MW TRIGA MARK II at ENEA Casaccia, ENEA Internal Report FIS-P815-017 (2005)
References
123
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B2 Medical, Energetic and
Fig. B2.15 – Special device for
introducing/extracting specimens
B Fission Technology
The procedure followed was to
compare the pre- and post-test
analyses: visual inspection, surface
area and roughness measurement,
weight, x-ray analysis (only for the
welded specimens). In the post-test
analysis the optical microscopy
(scanning electron microscopy and energy dispersive x-ray spectroscopy) analysis was taken into
account. The static tests ended after 8000 h at 590°C.
B2.3 Development Activities for Antarctic Drilling
Talos Dome is an ice dome (72°48’S; 159°06’E, 2316 m) on the edge of the East Antarctic plateau
and adjacent to the Victoria Land Mountains in the western Ross Sea area. The firn core
temperature is -41°C, and average snow accumulation over the last eight centuries is 80 kg/m 2 /yr.
Airborne radar measurements indicate that the dome summit is situated above sloping bedrock (ice
thickness 1880 ± 25 m ), but there is relatively flat bedrock 5-6 km distant along the SE ice divide
(ID1 159°11’00”E, 72°49’40”S, 2315 m), about 770 ± 25 m in elevation and covered by
1545 ± 25 m of ice (fig. B2.16).
Wilkes Subglacial Basin
Weddel Sea
Legend
Core site
Wind direction
Main ice divide
10m contour line
Rock outcrop
Byrd
0 1000 2000
km
Southern Ocean
290 km
Reeves
Glacier
Kohnen
Berken Island
Ross Sea
Terra Nova
Bay Station
Dome Fuji
Vostok
Dome C
Talos Dome
Ross Sea
Ice thickness (m):
ID1=1545 ± 25m
TD summit=1880 ± 25m
1940000
1940000
1935000
1935000
1930000
1925000
North (m)
1930000
1925000
North (m)
1920000
1920000
Southern Ocean
Law
Dome
Dome C
EPICA
Vostok
500
0
1100 km
TD Summit
1915000
1910000
TD
Summit
1915000
1910000
550 km Ross
Taylor
Sea
Dome
1500 km
km
Siple
Dome
Talos
Dome
Bedrock elevation (WGS84 m):
ID1= 770 ± 25m
TD summit=440±25m
ID1
ID1
1905000
1905000
1900000
1900000
495000
490000
490000
505000
500000
500000
495000
515000
510000
505000
525000
520000
East (m)
515000
510000
East (m)
530000
530000
525000
520000
Fig. B2.16 – Talos Dome
geographic position
Progress Report 2006
124

Environmental Applications
Fig. B2.17 – Talos Dome remote camp
Sleeping and storages
Snow for water
Camp fuel
Drilling at Talos Dome should greatly increase
knowledge about the response of nearcoastal
sites to climate changes and the
Holocene history of accumulation rates in the
Ross Sea region. In addition, this ice record
will strongly contribute to understanding the
last glacial-interglacial transition when
different climatic features and trends are
observed between West-East Antarctica
(Byrd, Vostok, EPICA-Dome C, Dome Fuji,
Law Dome) and two near-coastal sites in the
Ross Sea sector (Taylor and Siple Dome).
Lastly it would give an idea of the future
variability of accumulation and dynamic
changes in this sensitive area.
Runway and
TO fuel
-600
-650
-700
Science trench and
drill generator
Main generator and living
Cargo line
29/11/06 START OF THE SEASON: -600.84M
3 Dec. 1° week: 32.49 m, -633.29 m
Problem with PLC-Inverter Stop 6-7 Dec.
10 Dec. 2° week: 52.04 m, -685.44 m
Trench
entrance
The Talos Dome Ice (TALDICE) project is
currently drilling to bedrock at the ID1 site,
and one glacial/interglacial period of usable
record is expected. The project is also aimed
at developing integrated instrumentation in
order to improve the Italian capability to drill
and measure the ice core and to plan and
manage both the mechanical parts and the
electronic control system of the new
perforation system.
The project started in the field in November
2004. In this first season one French and four
Italian technicians and a scientist were
involved for about 50 days in drilling activities
and in setting up a temporary field camp
(summer camp), using the vehicles, modules
and tents of the International Trans Antarctic
Scientific Expedition (ITASE) programme
(fig. B2.17).
During the second season, from November
2005 to January 2006, for about 80 days
eleven technicians and scientists (3 French
and 8 Italian) were involved in TALDICE
activity. The camp was opened on 7
Fig. B2.18 – Perforation progress graph season 2006-2007
November. The first 40 days of the season
were dedicated to re-building the roof of the
perforation trench and setting up both the drill facilities inside the trench and the camp infrastructures.
During the first drilling season from 17 December 2005 to 15 January 2006 the final depth of -607.74 m
was reached, equal to ~7500 years ago. The ice cores to 480 m depth were analysed by a dielectric
profiling instrument, then cut, put in plastic bags, packed in boxes and sent to Europe for further analyses.
The camp for the second campaign of perforation (fig. B2.18) was opened on 7 November 2006 during
Depth (m)
-750
-800
-850
-900
-950
-1000
-1050
-1100
-1150
-1200
-1250
-1300
26 Dec. M1 motor problem
power generator problem
31 Dec. 5° week: 117.83 m, -1098.54 m
7 Jan. 6° Week: 122.30m, -1220.84m
17 Dec. 3° week: 146.10 m, -831.64 m
11 Jan. end drill season: -1293.86 m
Logged Depth: -1300.58 m
24 Dec. 4° week: 149.17 m, -980.71 m
25 Dec. Christmas day off
01 Jan. new year day off
29/11/2006
01/12/2006
03/12/2006
05/12/2006
07/12/2006
09/12/2006
11/12/2006
13/12/2006
15/12/2006
17/12/2006
19/12/2006
21/12/2006
23/12/2006
25/12/2006
27/12/2006
29/12/2006
31/12/2006
02/01/2007
04/01/2007
06/01/2007
08/01/2007
10/01/2007
12/01/2007
14/01/2007
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B2 Medical, Energetic and
B Fission Technology
Fig. B2.19 – 27 December 2006: 1000 m of perforation
Fig. B2.20 – Tephra Layers
Table B2.II – Perforation season 2006-2007 final data
Start perforation 29 November 2006
End perforation 11 January 2007
Duration of perforation
36 useful days
Logged depth
1300.58 m
Drilling depth
1293.86 m
Drilling length this season 693 m
Packed, cut and sent depth 486 m (from 478.00 to 666.00 m
from 1001.00 to 1300.00 m)
Liquid level end of season
(density=0.958 g cm -3 ) 109 m
Daily average
19.25 m
Average liquid/m
17.41 m
Run number 386
Core length average/run 1.79 m
Total recovered chips ice 4950 kg (7 kg chips/m)
Perforation hours
501 h
Core length average/
perforation hours
1.37 m/h
Progress Report 2006
126

Environmental Applications
the ongoing Antarctic Campaign (2006-2007). The drilling season was started on 29 November and will
terminate 11 January 2007, with the final depth of –1300.58 m, corresponding to approxi mately 60,000
to 80,000 years ago. The original objective, 1200 m, has been surpassed, meaning that all the ice covering
the last deglaciation and the end of the last glaciation is up at the surface. The material will be studied in
the laboratories of the European countries (Italy, France, German, Switzerland and UK) involved in the
project (fig. B2.19).
The visible tephra layers observed during this season total 35, of which 10 are particularly dense
(fig. B2.20). These layers will be very useful for getting one exact dating of the ice extracted during
perforation. Table B2.II reports the final data of the 2006-2007 perforation season. The Antarctic bed rock,
situated at a depth of approximately-1550 m should be reached during the next perforation campaign
2007-2008.
127
Progress Report 2006

B3 Participation in International Working Groups
and Associations
B Fission Technology
In relation to ENEA’s institutional role as the national focal point and advisor for nuclear-energyrelated
scientific and technological issues, the department ensures that ENEA (and Italy as a whole)
is represented in the principal committees and bodies concerned with the pacific use of nuclear
energy, at national (Ministry of Economic Development [MSE]) and international (Nuclear Energy
Agency [NEA], International Atomic Energy Agency [IAEA], Euratom, etc.) levels. Representatives
and experts from the department are present in nearly all the NEA standing committees (NSC, NDC,
CSNI, RWMC, CRPPH) as well as in the steering committees, and in a number of IAEA permanent
Table B3.I - Bilateral agreements
Scientific & Technological
Agency/Institution
Commissariat à l’Energie Atomique (CEA)
Forschungszeuntrum Karlsruhe (FZK)
Oak Ridge National Laboratory (ORNL)
Belgian Nuclear Research Centre (SCK-CEN)
Joint Research Centre - Institute for Transuranium
Elements (JRC-ITU)
Paul Scherrer Institute (PSI)
Seoul National University (SNU)
Institut Laue-Langevin (ILL)
Institute of Mathematics and Mechanics,
Ural branch of the Russian Academic of Science
(IMM-RAS)
Country
FRANCE
GERMANY
USA
BELGIUM
EUROPEAN COMMISSION
SWITZERLAND
SOUTH KOREA
FRANCE
RUSSIA
Progress Report 2006
128

technical working groups (TWGs), for example, on fast reactors, advanced technologies for light water
reactors, on fuel performance and technology, etc. Finally, a senior researcher of the department acts as
national delegate in the Consultative Committee Euratom-Fission (CCE-Fission).
The department also administers several bilateral agreements with major international organisations (see
table B3.I) in the nuclear fission field to ensure that R&D activities of common interest are performed
synergically.
Field of Co-operation
Thematic area
Nuclear fission
Accelerator–driven systems &
transmutation
Advanced fission reactors
Accelerator–driven systems
Accelerator–driven systems
and partitioning and transmutation
Spallation neutron sources
Nuclear fission
Irradiation in high flux reactors
Nuclear and conventional
accident analysis
Physics, safety, technologies and
code developments for nuclear
reactors
Fuel cycle strategies, transmutation
systems and HLM Technologies
Experimental testing, computer
simulations, design and
performance of advanced reactors
Physics and technologies for ADS
ADS technologies and advanced
fuels
Physics and HLM technologies
HLM technologies
Nuclear Instrumentation, diagnostics
and materials
Exp. measurements, computer
simulations and code developments
129
Progress Report 2006

B4 Publications
B Fission Technology
Articles
B4.1 Publications
R&D on Nuclear Fission
C. HELLWIG, M. STREIT, P. BLAIR, F.C. KLAASSEN, R.P.C. SCHRAM, F. VETTRAINO, T. YAMASHITA: Inert
matrix fuel behaviour in test irradiations
J. Nucl. Mater. 352, 291-299 (2006)
M. STREIT, T. TVERBERG, W. WIESENACK, F. VETTRAINO: Inert matrix and thoria fuel irradiation at an
international research reactor
J. Nucl. Mater. 352, 263-267 (2006)
F. BIANCHI, C. ARTIOLI, K.W. BURN, G. GHEPARDI, S. MONTI, L. MANSANI, L. CINOTTI, D. STRUWE, M.
SCHIKORR, W. MASACHEK, H.A. ABDERRAHIM, D. DE BRUYN, G. RIMPAULT: Status and trend of core
design activities for heavy metal cooled accelerator driven system
Energy Convers. Manage. 47, 2698-2709 (2006)
S. ANDRIAMONJE, S. AUNE, G. BAN, S. BREAUD, C. BLANDIN, E. FERRER, B. GESLOT, A. GIGANON,
I. GIOMATARIS, C. JAMMES, Y. KADI, P. LABORIE, J.F. LECOLLEY, J. PANCIN, M. RAILLOR, R. ROSA, L.
SARCHIAPONE, J.C. STECKMEYER, J. TILLER: New neutron detector based on micromegas technology
for ADS project
Nucl. Intrum. Method A562, 755-759 (2006)
G. BENAMATI, A. GESSI, P.-Z. ZHANG: Corrosion experiments in flowing LBE at 450°C
J. Nucl. Mater. 356, 1-3, 198-202 (2006)
C. FOLETTI, G. SCADDOZZO, M. TARANTINO, A. GESSI, G. BERTACCI, P. AGOSTINI, G. BENAMATI:
ENEA experience in LBE technology
J. Nucl. Mater. 356, 1-3, 264-272 (2006)
P. AGOSTINI, L. SANSONE, G. BENAMATI, C. PETROVICH, S. MONTI: Neutronic and thermo-mechanic
calculations for the design of the TRADE spallation target
Nucl. Instrum. Method Phys. Res. 562, 849-854 (2006)
A. MATHIS, S. MONTI: Energia nucleare: l’opzione del futuro; prima e seconda parte
Termotecnica, Marzo N. 36-Aprile N.58 (2006)
L. BURGAZZI, R. FERRI, B. GIANNONE: Safety assessment of a liquid target
Nucl. Eng. Des. 236, 4, 359-367 (2006)
Progress Report 2006
130

L. BURGAZZI: Probabilistic safety analysis of an accelerator-lithium target based experimental facility
Nucl. Eng. Des. 236, 12, 1264-1274 (2006)
F. CANNATA, A. VENTURA: Scattering by PT-symmetric non-local potentials
Czech. J. Phys. 56, 943-951 (2006)
G. A. KERIMOV, A. VENTURA: Group-theoretical approach to reflectionless potentials
J. Math. Phys. 47, 082108, 1-16 (2006)
M. SIN, R. CAPOTE, A. VENTURA, M. HERMAN, P. OBLOŽINSKY: Fission of light actinides: 232 Th(n,f) and
231 Pa(n,f) reactions
Phys. Rev. C 74, 014608, 1-13 (2006)
G. AERTS, U. ABBONDANNO, H. ALVAREZ, F. ALVAREZ-VELARDE, S. ANDRIAMONJE, J. ANDRZEJEWSKI, P.
ASSIMAKOPULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN, F. BEČVÁR, E. BERTHOUMIEUX, F. CALVIÑO, D.
CANO-OTT, R. CAPOTE, A. CARRILLO DE ALBORNOZ, P. CENNINI, V. CHEPEL, E. CHIAVERI, N. COLONNA, G.
CORTES, A. COUTURE, J. COX, M. DAHLFORS, S. DAVID, I. DILLMAN, R. DOLFINI, C. DOMINGO-PARDO, W.
DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-SEGURA, L. FERRANT, A. FERRARI, R. FERREIRA-MARQUES,
L. FITZPATRICK, H. FRAIS-KOELBL, K. FUJII, W. FURMAN, I. GONCALVES, E. GONZALEZ-ROMERO, A.
GOVERDOSKI, F. GRAMEGNA, E. GRIESMAYER, C. GUERRERO, F. GUNSING, B. HAAS, R. HAIGHT, M. HEIL,
A. HERRERA-MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA, Y. KADI, F. KÄPPELER, D. KARADIMOS, D.
KARAMANIS, M. KERVENO, V. KETLEROV, P. KOEHLER, V. KONOVALOV, E. KOSSIONIDES, M. KRTIČKA, C.
LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S. LUKIC, J. MARGANIEC, L. MARQUES, S.
MARRONE, P. MASTINU, A. MENGONI, P.M. MILAZZO, C. MOREAU, M. MOSCONI, F. NEVES, H.
OBERHUMMER, S. O’BRIEN, M. OSHIMA, J. PANCIN, C. PAPACHRISTODOULOU, C. PAPADOPOULOS, C.
PARADELA, N. PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, M. T. PIGNI, R. PLAG, A. PLOMPEN, A.
PLUKIS, A. POCH, C. PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G.
RUDOLF, P. RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G. TAGLIENTE, J.L. TAIN,
L. TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN, M. C. VINCENTE,
V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M. WIESCHER AND K. WISSHAK (The
n_TOF Collaboration): Neutron capture cross section of 232 Th measured at the n_TOF facility at CERN in the
unresolved resonance region up to 1 MeV
Phys. Rev. C 73, 054610, 1-10 (2006)
S. MARRONE, U. ABBONDANNO, G. AERTS, F. ALVAREZ-VELARDE, H. ALVAREZ-POL, S. ANDRIAMONJE, J.
ANDRZEJEWSKI, G. BADUREK, P. BAUMANN, F. BEČVÁR, J. BENLLIURE, E. BERTHOMIEUX, F. CALVIÑO, D.
CANO-OTT, R. CAPOTE, P. CENNINI, V. CHEPEL, E. CHIAVERI, N. COLONNA, G. CORTES, D. CORTINA, A.
COUTURE, J. COX, S. DABABNEH, M. DAHLFORS, S. DAVID, R. DOLFINI, C. DOMINGO-PARDO, I. DURAN-
ESCRIBANO, M. EMBID-SEGURA, L. FERRANT, A. FERRARI, R. FERREIRA-MARQUES, H. FRAIS-KOELBL, K.
FUJII, W. I. FURMAN, R. GALLINO, I. F. GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F.
GRAMEGNA, E. GRIESMAYER, F. GUNSING, B. HAAS, R. HAIGHT, M. HEIL, A. HERRERA-MARTINEZ, S. ISAEV,
E. JERICHA, F. KÄPPELER, Y. KADI, D. KARADIMOS, M. KERVENO, V. KETLEROV, P. E. KOEHLER, V.
KONOVALOV, M. KRTIČKA, C. LAMBOUDIS, H. LEEB, A. LINDOTE, M. I. LOPES, M. LOZANO, S. LUKIC, J.
MARGANIEC, J. MARTINEZ-VAL, P. F. MASTINU, A. MENGONI, P. M. MILAZZO, A. MOLINA-COBALLES, C.
MOREAU, M. MOSCONI, F. NEVES, H. OBERHUMMER, S. O’BRIEN, J. PANCIN, T. PAPAEVANGELOU, C.
PARADELA, A. PAVLIK, P. PAVLOPOULOS, J. M. PERLADO, L. PERROT, M. PIGNATARI, M. T. PIGNI, R. PLAG,
A. PLOMPEN, A. PLUKIS, A. POCH, A. POLICARPO, C. PRETEL, J. M. QUESADA, S. RAMAN, W. RAPP, T.
RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G. RUDOLF, P. RULLHUSEN, J. SALGADO, J. C. SOARES,
C. STEPHAN, G. TAGLIENTE, J. L. TAIN, L. TASSAN-GOT, L. M. N. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A.
VENTURA, D. VILLAMARIN-FERNANDEZ, M. VINCENTE-VINCENTE, V. VLACHOUDIS, F. VOSS, H. WENDLER,
M. WIESCHER AND K. WISSHAK (The n_TOF Collaboration): Measurement of the 151 Sm(n,γ) cross section from
0.6 eV to 1 MeV via the neutron time-of-flight technique at the CERN n_TOF facility
Phys. Rev. C 73, 034604, 1-18 (2006)
131
Progress Report 2006

B4 Publications
B Fission Technology
C. DOMINGO-PARDO, U. ABBONDANNO, G. AERTS, H. ÁLVAREZ-POL, F. ALVAREZ-VELARDE, S.
ANDRIAMONJE, J. ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN,
F. BEČVÁR, E. BERTHOUMIEUX, F. CALVIÑO, D. CANO-OTT, R. CAPOTE, A. CARRILLO DE ALBORNOZ,
P. CENNINI, V. CHEPEL, E. CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M. DAHLFORS,
S. DAVID, I. DILLMAN, R. DOLFINI, W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-SEGURA, L.
FERRANT, A. FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, K. FUJII, W.
FURMAN, R. GALLINO, I. GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F. GRAMEGNA, E.
GRIESMAYER, C. GUERRERO, F. GUNSING, B. HAAS, R. HAIGHT, M. HEIL, A. HERRERA-MARTINEZ, M.
IGASHIRA, S. ISAEV, E. JERICHA, Y. KADI, F. KÄPPELER, D. KARAMANIS, D. KARADIMOS, M. KERVENO,
V. KETLEROV, P. KOEHLER, V. KONOVALOV, E. KOSSIONIDES, M. KRTIČKA, C. LAMBOUDIS, H. LEEB,
A. LINDOTE, I. LOPES, M. LOZANO, S. LUKIC, J. MARGANIEC, L. MARQUES, S. MARRONE, P.
MASTINU, A. MENGONI, P. M. MILAZZO, C. MOREAU, M. MOSCONI, F. NEVES, H. OBERHUMMER, M.
OSHIMA, S. O’BRIEN, J. PANCIN, C. PAPACHRISTODOULOU, C. PAPADOPOULOS, C. PARADELA, N.
PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, R. PLAG, A. PLOMPEN, A. PLUKIS, A. POCH, C.
PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G. RUDOLF, P.
RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G. TAGLIENTE, J. L. TAIN, L.
TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN, M. C.
VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M. WIESCHER, AND K.
WISSHAK (The n_TOF Collaboration): New measurement of neutron capture resonances in 209 Bi
Phys. Rev. C 74, 025807, 1-10 (2006)
C. DOMINGO-PARDO, U. ABBONDANNO, G. AERTS, H. ÁLVAREZ-POL, F. ALVAREZ-VELARDE, S.
ANDRIAMONJE, J. ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN,
F. BEČVÁR, E. BERTHOUMIEUX, S. BISTERZO, F. CALVIÑO, D. CANO-OTT, R. CAPOTE, C. CARRAPIC¸
O,P. CENNINI, V. CHEPEL, E. CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M.
DAHLFORS, S. DAVID, I. DILLMAN, R. DOLFINI, W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-
SEGURA, L. FERRANT, A. FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, K.
FUJII, W. FURMAN, R. GALLINO, I. GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F.
GRAMEGNA, E. GRIESMAYER, C. GUERRERO, F. GUNSING, B. HAAS, R. HAIGHT, M. HEIL, A.
HERRERA-MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA, Y. KADI, F. KÄPPELER, D. KARAMANIS, D.
KARADIMOS, M. KERVENO, V. KETLEROV, P. KOEHLER, V. KONOVALOV, E. KOSSIONIDES, M. KRTIČKA,
C. LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S. LUKIC, J. MARGANIEC, S. MARRONE,
P. MASTINU, A. MENGONI, P. M. MILAZZO, C. MOREAU, M. MOSCONI, F. NEVES, H. OBERHUMMER,
M. OSHIMA, S. O’BRIEN, J. PANCIN, C. PAPACHRISTODOULOU, C. PAPADOPOULOS, C. PARADELA, N.
PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, R. PLAG, A. PLOMPEN, A. PLUKIS, A. POCH, C.
PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G. RUDOLF, P.
RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G. TAGLIENTE, J.L. TAIN, L.
TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN, M. C.
VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M. WIESCHER, AND K.
WISSHAK (The n_TOF Collaboration): Resonance capture cross section of 207 Pb
Phys. Rev. C 74, 055802, 1-6 (2006)
R. ORSI: A general method of conserving mass in complex geometry simulations on mesh grids and its
implementation in BOT3P5.0
Nucl. Sci. Eng. 154, 247-259 (2006)
J.J. KLINGENSMITH, Y.Y. AZMY, J.C. GEHIN, R. ORSI: Tort solutions to the three-dimensional MOX
Benchmark, 3-D extension C5G7MOX
Prog. Nucl. Energy 48, 445 455 (2006)
E. BOTTA, R. ORSI: Westinghouse AP1000 internals heating rate distribution calculation using a 3-D
deterministic transport method
Nucl. Eng. Des. 236, 1558-1564 (2006)
Progress Report 2006
132

A. ANDRIGHETTO, C.M. ANTONUCCI, S. CEVOLANI, C.PETROVICH, M. SANTANA LEITNER: Multifoil UCx target
for the SPES project - an update
Europ. Phys. J. A 30, 591-601 (2006)
L. BURGAZZI: Failure mode and effect analysis application for the safety and reliability analysis of a thermalhydraulic
passive system
Nucl. Technol. 156, 2, 150-158 (2006)
Medical, Energetic and Environmental Applications
K.W. BURN, C. DAFFARA, G. GUALDRINI, M. PIERANTONI, P. FERRARI: Treating voxel geometries in
radiation protection dosimetry with a patched version of the Monte Carlo Codes MCNP and MCNPX
Radiat. Prot. Dosim. doi:10.1093/rpd/ncl150, OUP (2006)
Reports
R&D on Nuclear Fission
G. GLINATSIS: Safety aspects of the EFIT/MgO - Pb core, ENEA Internal Report FPN-P815-005 (2006)
M. SAROTTO, C. ARTIOLI: Possible solutions for the neutronic design of the two zones EFIT-MgO/Pb core, ENEA
Internal Report FPN-P815-001(2006)
M. SAROTTO, C. ARTIOLI, V. PELUSO: Preliminary neutronic analysis of the three zones EFIT-MgO/Pb core, ENEA
Internal Report FPN-P815-004 (2006)
M. SAROTTO, C. ARTIOLI, V. PELUSO: MgO/Pb core neutronic preliminary analysis, ENEA Internal Report FIS-
P815-021, EFIT (2006)
P. MELONI: A Neutronics-thermal-hydraulics model for preliminary studies on TRADE dynamics, ENEA Internal
Report FIS-P99R-006 (2006)
G. BANDINI: Interpretation of TRIGA experimental data with SIMMER-III code for RELAP5 model evaluation and
transient analysis, ENEA Internal Report FIS-P99R-007 (2006)
G. BANDINI, P. MELONI: Analysis of BETHSY experiment 4.3b with ASTEC V1.2 code for CESAR thermalhydraulic
module validation, ENEA Internal Report FPN-P9D0-001 (2006)
G. BANDINI: ICARE/CATHARE calculation of the QUENCH-11 experiment in the frame of the IRSN participation
in the SARNET benchmark, ENEA Internal Report FPN-P9D0-002 (2006)
G. BANDINI: Analysis of the OECD LOFT fission product experiment LP-FP-2 with ASTEC V1.2.1 code, ENEA
Internal Report FPN-P9G1-001 (2006)
G. BANDINI: Validation of CESAR thermal-hydraulic module of ASTEC V1.2 code on BETHSY experiments, ENEA
Internal Report FIS-P9D0-002 (2006)
R. CAPONETTI: Determination by thermochemical calculation of speciation and deposition of fission product and
structural material during the vercors HT 1 experiment, ENEA Internal Report FIS-P127-042 (2006)
R. ORSI: BOT3P Version 5.1: two/three-dimensional mesh generator and graphical display of geometry and results
for deterministic transport codes, ENEA Report RT/2006/34/FIS (2006)
133
Progress Report 2006

B4 Publications
R. ORSI: BOT3P Version 5.1: a pre-post-processor system for transport analysis, ENEA Internal Report
FIS-P9H6-014 Rev.0 (2006)
B Fission Technology
R. ORSI: CATSM: a pre-processor tool for medical applications, ENEA Internal Report FIS-P9H6-015 Rev.0
(2006)
S. CEVOLANI: Valutazione approssimata dell’effetto della frequenza di pulsazione del fascio sulla termica
del target sottile per SPES, ENEA Internal Report FIS-P815-022 (2006)
A. ANDRIGHETTO, C. ANTONUCCI, S. CEVOLANI, C. PETROVICH: ENEA contribution to the design of the
thin target for the SPES project, ENEA Internal Report FIS-P815-020 (2006)
S. CEVOLANI: Termica della camera del target lamellare per SPES, ENEA Internal Report FPN-P815-(2006)
Medical, Energetic and Environmental Applications
M. BASTA, E. NAVA, G. ROSI: Monitoraggio d’area del reattore TAPIRO. Proposta di modifica, ENEA
Internal Report FPN-TLE TAPIRO 06/02 (2006)
Contributions to Conferences
R&D on Nuclear Fission
R. CALABRESE, F. VETTRAINO, T. TVERBERG: Inert matrix fuel modelling: transuranus analysis of the
Halden IFA-652 first irradiation cycle
Inter. Workshop on Materials Models and Simulation for Nuclear Fuels (MMSNF-5), Nice (France), June 1-
2, 2006
R. CALABRESE, F. VETTRAINO, T. TVERBERG: Low burn-up inert matrix fuels performance: transuranus
analysis of the Halden IFA-652 first irradiation cycle
14 th Inter. Conference on Nuclear Engineering (ICONE-14), Miami (USA), July 17-20, 2006
S. BOURG, C. CARAVACA, E. WALLE, G. DE ANGELIS, R. MALMBECK, G.B. LEWIN, J. UHLIR, T. INOUE,
V. LUCA: Pyrochemistry within EUROPART from the acquisition of basic data to the processes for the
treatment of spent fuels
9 th IEM on Actinide and Fission Product Partitioning and Transmutation, OECD Nuclear Energy Agency (9-
IEMPT), Nimes (France), September 25-29, 2006
C. MADIC, M. J. HUDSON, P. BARON, N. OUVRIER, C. HILL, F. ARNAUD, A. G. ESPARTERO, J.F.
DESREUX, G. MODOLO, R. MALMBECK, S. BOURG, G. DE ANGELIS AND J. UHLIR: EUROPART.
European research programme for partitioning of minor actinides within high active wastes issuing from the
reprocessing of spent nuclear fuels. Some of the principal results obtained
9 th IEM on Actinide and Fission Product Partitioning and Transmutation, OECD Nuclear Energy Agency (9-
IEMPT), Nimes (France) September 25-29, 2006
F. BIANCHI, R. FERRI: Accident analysis of the windowless target system
Topical Meeting on Advances In Nuclear Analysis and Simulation (PHYSOR-2006), Vancouver (Canada),
September 10-14, 2006
F. BIANCHI, R. FERRI, V. MOREAU: Transient thermo-hydraulic analysis of the windowless target system
for the lead bismuth eutectic cooled accelerator driven system
14 th Inter. Conference on Nuclear Engineering (ICONE-14), Miami (USA), July 16-20, 2006
Progress Report 2006
134

P. AGOSTINI, M. CIOTTI, C. PETROVICH, M. CARTA, N. ELMI, L. SANSONE, D. BELLER, C. KRAKOWIAK, A.
BERGERON: Target study for the RACE HP experiment
14 th Inter. Conference on Nuclear Engineering (ICONE-14), Miami (USA), July 17-20, 2006
P. AGOSTINI, M. CIOTTI, C. KRAKOWIAK, C. PETROVICH, G. BENAMATI, A. BERGERON, N. ELMI, G. GRANGET,
L. SANSONE, M. SCHIKORR: Target study for the RACE HP experiment
8 th Inter. Workshop on Spallation Materials Technology (IWSMT-8), Taos (USA), October 16-20, 2006
M. CARTA, N. BURGIO, A. D’ANGELO, A. SANTAGATA, C. PETROVICH, M. SCHIKORR, D. BELLER, L. SAN
FELICE, G. IMEL, M. SALVATORES: Electron versus proton accelerator driven sub-critical system performance
using TRIGA reactors at power
Topical Meeting on Advances In Nuclear Analysis and Simulation (PHYSOR-2006), Vancouver (Canada),
September 10-14, 2006
W. AMBROSINI, G. BENAMATI, S. CARNEVALI, C. FOLETTI, N. FORGIONE, F. ORIOLO, G. SCADDOZZO, M.
TARANTINO: Experiments on gas injection enhanced circulation in a pool-type liquid metal apparatus
XXIV Congresso Nazionale UIT, Napoli (Italy), June 21-23, 2006
A. RENIERI: L’integrazione delle competenze nello sviluppo di sistemi nucleari innovativi
Convegno L’uso pacifico dell’energia nucleare da Ginevra 1955 ad oggi: Il caso italiano,
Rome (Italy), March 8–9, 2006
A. RENIERI, S. MONTI: Le attività di R&S dell’ENEA nel contesto europeo ed internazionale del nuovo nucleare da
fissione: sinergie e collaborazioni in Italia e all’estero
Convegno Nazionale AEIT, Capri (Italy), September 16-20, 2006
S. MONTI: IRIS integral test: experimental investigation of small break LOCAs in coupled vessel/containment
integral reactors
15 th IRIS Team Meeting, Pittsburgh (USA), April 25-27, 2006
S. MONTI: IRIS activities in the framework of the Italian national program on nuclear fission
16 th IRIS Team Meeting, Santander (Spain), November 7-9, 2006
L. CINOTTI, C. FAZIO, J. KNEBEL, S. MONTI, H. AIT ABDERRAHIM, C. SMITH, K. SUH: LFR lead-cooled fast
reactor
Conference on EU Research and Training in Reactor Systems (FISA 2006), Kirchberg, Luxembourg, March 13-16,
2006
S. MONTI: Status and perspectives of Italian activities in the field of fast spectrum nuclear systems
IAEA Technical Meeting on Review of National Programmes on Fast Reactors and Accelerator Driven Systems,
39 th TWG-FR Annual Meeting (CIAE), Beijing (China), May 15-19, 2006
S. MONTI: Planned R&D and technology activities in Italy for the development of the GENIV lead-cooled fast
reactor
IAEA Technical Meeting, Vienna (Austria), December 6-8, 2006
P. MELONI: Overview of helium cooled system applications with RELAP at ENEA
2006 Inter. Congress on Advances in Nuclear Power Plant (ICAPP ’06), Reno (USA), June 4-8, 2006
G. BANDINI, P. MELONI, N. TRÉGOURÈS, J. FLEUROT: Post-test analysis of the BETHSY experiment 9.1b with
ASTEC V1.2 Code for CESAR thermal-hydraulic module validation
NENE International Conference, Portoroz (Slovenia), September 18-21, 2006
G. BANDINI, G. GUILLARD, J. FLEUROT: Participation in the SARNET benchmark: analysis of the QUENCH-11
experiment with ICARE/CATHARE code
12 th Inter. QUENCH Workshop, Forschungszentrum Karlsruhe (Germany), October 24-26, 2006
135
Progress Report 2006

B4 Publications
S. EDERLI: ENEA activity in the WP9.3,
SARNET CORIUM Topic 2 nd Annual Review Meeting, Villigen (Switzerland), January 30-31, 2006
B Fission Technology
G. REPETTO, S. EDERLI: Assessment of the heat transfer and late phase model of the ICARE/CATHARE
code against debris bed in pile experiments
18 th National & 7 th ISHMT-ASME Heat and Mass Transfer Conference, Guwahati (India), January 4-6, 2006
L. BURGAZZI, M. MARQUES: Integration of passive system reliability in PSA studies
14 th Inter. Conference on Nuclear Engineering (ICONE-14), Miami (USA), July 16-20, 2006
L. BURGAZZI: Probabilistic design of a passive system
2006 ANS Winter Meeting, Albuquerque (New Mexico), November 12-16, 2006
L. BURGAZZI: Development of probability distributions of passive system failure
3 rd Inter. Symposium on Systems & Human Science: Complex Systems Approaches for Safety, Security
and Reliability (SSR 2006), Vienna (Austria), March 6-8, 2006
L. BURGAZZI: Reliability aspects of passive systems
EC Enlargement and Integration Workshop on Use of Probabilistic Safety Assessment (PSA) for Evaluation
of Impact of Ageing Effects on the Safety of Nuclear Power Plants, Bucharest (Romania), October 2-4,
2006.Invited talk
L. BURGAZZI: Incorporation of ageing effects into component reliability and availability models
Europ. Safety and Reliability Conference (ESREL ’06), Estoril (Portugal), September 18-22, 2006
M. HEIL, U. ABBONDANNO, G. AERTS, H. ÁLVAREZ-POL, F. ALVAREZ-VELARDE, S. ANDRIAMONJE, J.
ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN, F. BEČVÁŘ, E.
BERTHOUMIEUX, S. BISTERZO, F. CALVIÑO, D. CANO-OTT, R. CAPOTE, C. CARRAPICO, P. CENNINI, V.
CHEPEL,E. CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M. DAHLFORS, S. DAVID, I.
DILLMAN, R. DOLFINI, CÉSAR DOMINGO PARDO, W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-
SEGURA, L. FERRANT, A. FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, K.
FUJII, W. FURMAN, R. GALLINO, I. GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F.
GRAMEGNA, E. GRIESMAYER, C. GUERRERO, F. GUNSING, B. HAAS, R. HAIGHT, A. HERRERA-
MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA, Y. KADI, F. KÄPPELER, D. KARAMANIS, D.
KARADIMOS, M. KERVENO, V. KETLEROV, P. KOEHLER, V. KONOVALOV, E. KOSSIONIDES, M. KRTI ČKA,
C. LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S. LUKIC, J. MARGANIEC, S. MARRONE,
P. MASTINU, A. MENGONI, P.M. MILAZZO, C. MOREAU, M. MOSCONI, F. NEVES, H. OBERHUMMER, M.
OSHIMA, S. O’BRIEN, J. PANCIN, C. PAPACHRISTODOULOU, C. PAPADOPOULOS, C. PARADELA, N.
PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, R. PLAG, A. PLOMPEN, A. PLUKIS, A. POCH, C.
PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G. RUDOLF, P.
RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G. TAGLIENTE, J.L. TAIN, L.
TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN, M. C.
VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M. WIESCHER, K.
WISSHAK (The n_TOF Collaboration): Neutron capture cross section measurements for nuclear
astrophysics at n_TOF
9 th Inter. Symposium on Nuclei in the Cosmos (NIC-IX), CERN (Geneva), June 25-30, 2006, SISSA
Proceedings of Science (http://pos.sissa.it/), PoS (NIC-IX) 053
M. MOSCONI, M. HEIL, F. KÄPPELER, R. PLAG, A. MENGONI, K. FUJII, R. GALLINO, G. AERTS,
R.TERLIZZI, U. ABBONDANNO, H. ÁLVAREZ-POL, F. ALVAREZ-VELARDE, S. ANDRIAMONJE, J.
ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN, F. BEČVÁŘ, E.
BERTHOUMIEUX, S. BISTERZO, F. CALVIÑO, D. CANO-OTT, C. CARRAPIÇO, R. CAPOTE, P. CENNINI, V.
CHEPEL, E. CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M. DAHLFORS, S. DAVID, I.
DILLMAN, R. DOL NI, C. DOMINGO PARDO W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-
Progress Report 2006
136

SEGURA, L. FERRANT, A. FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, W.
FURMAN, I. GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F. GRAMEGNA, E. GRIESMAYER, C.
GUERRERO, F. GUNSING, B. HAAS, R. HAIGHT, A. HERRERA-MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA,
Y. KADI, D. KARAMANIS, D. KARADIMOS, M. KERVENO, V. KETLEROV, P. KOEHLER, V. KONOVALOV, E.
KOSSIONIDES, M. KRTI CKA, C. LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S. LUKIC, J.
MARGANIEC, S. MARRONE, P. MASTINU, P.M. MILAZZO, C. MOREAU, F. NEVES, H. OBERHUMMER, M.
OSHIMA, S. O'BRIEN, J. PANCIN, C. PAPACHRISTODOULOU, C. PAPADOPOULOS, C. PARADELA, N.
PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, A. PLOMPEN, A. PLUKIS, A. POCH, C. PRETEL, J.
QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G. RUDOLF, P. RULLHUSEN, J. SALGADO,
L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G. TAGLIENTE, J.L. TAIN, L. TASSAN-GOT, L. TAVORA, G.
VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN, M. C. VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S.
WALTER, H. WENDLER, M. WIESCHER, K. WISSHAK (The n_TOF Collaboration): Experimental challenges for the
Re/Os Clock
9 th Inter. Symposium on Nuclei in the Cosmos (NIC-IX), CERN (Geneva), June 25-30, 2006, SISSA Proceedings
of Science (http://pos.sissa.it/), PoS (NIC-IX) 055
C. DOMINGO PARDO, U. ABBONDANNO, G. AERTS, H. ÁLVAREZ-POL, F. ALVAREZ-VELARDE6, S.
ANDRIAMONJE, J. ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN, F.
BEČVÁŘ, E. BERTHOUMIEUX, S. BISTERZO, F. CALVIÑO, D. CANO-OTT, R. CAPOTE, C. CARRAPIÇO, P.
CENNINI, V. CHEPEL, E. CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M. DAHLFORS, S. DAVID,
I. DILLMAN, R. DOLFINI, W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-SEGURA, L. FERRANT, A.
FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, K. FUJII, W. FURMAN, R. GALLINO, I.
GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F. GRAMEGNA, E. GRIESMAYER, C. GUERRERO, F.
GUNSING, B. HAAS, R. HAIGHT, M. HEIL, A. HERRERA-MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA, Y.
KADI, F. KÄPPELER, D. KARAMANIS, D. KARADIMOS, M. KERVENO, V. KETLEROV, P. KOEHLER0, V.
KONOVALOV, E. KOSSIONIDES, M. KRTIČKA, C. LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S.
LUKIC, J. MARGANIEC, S. MARRONE, P. MASTINU, A. MENGONI, P.M. MILAZZO, C. MOREAU, M. MOSCONI,
F. NEVES, H. OBERHUMMER, M. OSHIMA, S. O’BRIEN, J. PANCIN, C. PAPACHRISTODOULOU, C.
PAPADOPOULOS, C. PARADELA, N. PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, R. PLAG, A.
PLOMPEN, A. PLUKIS, A. POCH, C. PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C.
RUBBIA, G. RUDOLF, P. RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G.
TAGLIENTE, J.L. TAIN, L. TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D.
VILLAMARIN, M. C. VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M.
WIESCHER, K. WISSHAK (The n_TOF Collaboration): Neutron capture measurements on the s-process
termination isotopes, lead and bismuth
9 th Inter. Symposium on Nuclei in the Cosmos (NIC-IX), CERN (Geneva), June 25-30, 2006, SISSA Proceedings
of Science (http://pos.sissa.it/), PoS (NIC-IX) 058
S. MARRONE, U. ABBONDANNO, G. AERTS, H. ÁLVAREZ, F. ALVAREZ-VELARDE, S. ANDRIAMONJE, J.
ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN, F. BEČVÁŘ , E.
BERTHOUMIEUX, F. CALVIÑO, D. CANO-OTT, R. CAPOTE, C. CARRAPIÇO, P. CENNINI, V. CHEPEL, E.
CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M. DAHLFORS, S. DAVID, I. DILLMANN,
R.DOLFINI, C. DOMINGO-PARDO, W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-SEGURA, L. FERRANT,
A. FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, K. FUJII, W. FURMAN, R. GALLINO,
I. GONCALVES, E. GONZALEZ-ROMERO, A. GOVERDOVSKI, F. GRAMEGNA, E. GRIESMAYER, C. GUERRERO,
F. GUNSING, B. HAAS, R. HAIGHT, M. HEIL, A. HERRERA-MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA, Y.
KADI, F. KÄPPELER, D. KARAMANIS, D. KARADIMOS, M. KERVENO, V. KETLEROV, P. KOEHLER, V.
KONOVALOV, E. KOSSIONIDES, M. KRTIČKA, C. LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S.
LUKIC, J. MARGANIEC, P. MASTINU, A. MENGONI, P.M. MILAZZO, C. MOREAU, M. MOSCONI, F. NEVES, H.
OBERHUMMER, S. O'BRIEN, M. OSHIMA, J. PANCIN, C. PAPACHRISTODOULOU, C. PAPADOPOULOS , C.
PARADELA, N. PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, M. PIGNATARI, R. PLAG, A. PLOMPEN, A.
PLUKIS, A. POCH, C. PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C. RUBBIA, G.
137
Progress Report 2006

B4 Publications
B Fission Technology
RUDOLF, P. RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, G. TAGLIENTE,
J.L. TAIN, L. TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN,
M.C. VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M. WIESCHER,
K.WISSHAK (The n_TOF Collaboration): Astrophysical implications of the 139 La(n,γ) and 151 Sm(n,γ) cross
sections measured at n_TOF
9 th Inter. Symposium on Nuclei in the Cosmos (NIC-IX), CERN (Geneva), June 25-30, 2006, SISSA
Proceedings of Science (http://pos.sissa.it/), PoS (NIC-IX) 138
G. TAGLIENTE, U. ABBONDANNO, G. AERTS, H. ÁLVAREZ, F. ALVAREZ-VELARDE, S. ANDRIAMONJE, J.
ANDRZEJEWSKI, P. ASSIMAKOPOULOS, L. AUDOUIN, G. BADUREK, P. BAUMANN, F. BEČVÁŘ, E.
BERTHOUMIEUX, F. CALVIÑO, D. CANO-OTT, R. CAPOTE, A. CARRILLO DE ALBORNOZ, P. CENNINI, V.
CHEPEL, E. CHIAVERI, N. COLONNA, G. CORTES, A. COUTURE, J. COX, M. DAHLFORS, S. DAVID, I.
DILLMANN, R. DOLFINI, C. DOMINGO-PARDO, W. DRIDI, I. DURAN, C. ELEFTHERIADIS, M. EMBID-
SEGURA, L. FERRANT, A. FERRARI, R. FERREIRA-MARQUES, L. FITZPATRICK, H. FRAIS-KOELBL, K.
FUJII, W. FURMAN, C. GUERRERO, I. GONCALVES, R. GALLINO, E. GONZALEZ-ROMERO, A.
GOVERDOVSKI, F. GRAMEGNA, E. GRIESMAYER, F. GUNSING, B. HAAS, R. HAIGHT, M. HEIL, A.
HERRERA-MARTINEZ, M. IGASHIRA, S. ISAEV, E. JERICHA, Y. KADI, F. KÄPPELER, D. KARAMANIS, D.
KARADIMOS, M. KERVENO, V. KETLEROV, P. KOEHLER, V. KONOVALOV, E. KOSSIONIDES, M. KRTIČKA,
C. LAMBOUDIS, H. LEEB, A. LINDOTE, I. LOPES, M. LOZANO, S. LUKIC, J. MARGANIEC, L. MARQUES,
S. MARRONE, C. MASSIMI, P. MASTINU, A.MENGONI, P.M. MILAZZO, C. MOREAU, M. MOSCONI, F.
NEVES, H. OBERHUMMER, S. O'BRIEN, M. OSHIMA, J. PANCIN, C. PAPACHRISTODOULOU, C.
PAPADOPOULOS, C. PARADELA, N. PATRONIS, A. PAVLIK, P. PAVLOPOULOS, L. PERROT, R. PLAG, A.
PLOMPEN, A. PLUKIS,A. POCH, C. PRETEL, J. QUESADA, T. RAUSCHER, R. REIFARTH, M. ROSETTI, C.
RUBBIA, G. RUDOLF, P. RULLHUSEN, J. SALGADO, L. SARCHIAPONE, I. SAVVIDIS, C. STEPHAN, J.L.
TAIN, L. TASSAN-GOT, L. TAVORA, R. TERLIZZI, G. VANNINI, P. VAZ, A. VENTURA, D. VILLAMARIN, M.C.
VINCENTE, V. VLACHOUDIS, R. VLASTOU, F. VOSS, S. WALTER, H. WENDLER, M. WIESCHER,
K.WISSHAK (The n_TOF Collaboration): Measurement of the 90,91,92,94,96 Zr neutron capture cross sections
at n_TOF
9 th Inter. Symposium on Nuclei in the Cosmos (NIC-IX), CERN (Geneva), June 25-30, 2006, SISSA
Proceedings of Science (http://pos.sissa.it/), PoS (NIC-IX) 227
C. GUERRERO R. CAPOTE, A. MENGONI et al. (The n_TOF Collaboration): Measurement at n_TOF of the
237 Np(n, γ) and 240 Pu(n,γ) cross sections for the transmutation of nuclear waste
ANS Topical Meeting on Reactor Physics (PHYSOR-2006), Vancouver (Canada), September 10-14, 2006
W. DRIDI ET AL. (The n_TOF Collaboration): Measurement of the neutron capture cross section of 234U in
n_TOF at CERN
ANS Topical Meeting on Reactor Physics (PHYSOR-2006), Vancouver (Canada), September 10-14, 2006
F. GUNSING et al. (The n_TOF Collaboration): Measurement of the neutron capture cross section of 236 U
ANS Topical Meeting on Reactor Physics (PHYSOR-2006), Vancouver (Canada), September 10-14, 2006
A. ANDRIGHETTO, C. ANTONUCCI, M. BARBUI, S. CARTURAN, F. CERVELLERA, S. CEVOLANI, M.
CINAUSERO, P. COLOMBO, A. DAINELLI, P. DI BERNARDO, F. GRAMEGNA, G. MAGGIONI, G.
MENEGHETTI, C. PETROVICH, L. PIGA, G. PRETE, V. RIZZI, M. TONEZZER, D. ZAFIROPOULOS, P.
ZANONATO: The SPES direct UCx target
7 th Inter. Conference on Radioactive Nuclear Beams (RNB-7), Cortina d’Ampezzo (Italy), July 3-7, 2006
A. ANDRIGHETTO, C. ANTONUCCI, M. BARBUI, S. CARTURAN, F. CERVELLERA, S. CEVOLANI, M.
CINAUSERO, P. COLOMBO, A. DAINELLI, P. DI BERNARDO, F. GRAMEGNA, G. MAGGIONI, G.
MENEGHETTI, C. PETROVICH, L. PIGA, G. PRETE, V. RIZZI, M. TONEZZER, D. ZAFIROPOULOS, P.
ZANONATO: The SPES direct UCx target
IX Inter. Conference on Nucleus-Nucleus Collisions, Rio de Janeiro (Brazil), August 28-September 1, 2006
Progress Report 2006
138

F. GRAMEGNA, A. ANDRIGHETTO, C. ANTONUCCI, M. BARBUI, L. BIASETTO, G. BISOFFI, S. CARTURAN, L.
CELONA, F. CERVELLERA, S. CEVOLANI, F. CHINES, M. CINAUSERO, P. COLOMBO, M. COMUNIAN, G.
CUTTONE, A. DAINELLI, P. DI BERNARDO, E. FAGOTTI, M. GIACCHINI, M. LOLLO, G. MAGGIONI, M.
MANZOLATO, G. MENEGHETTI, G.E. MESSINA, A. PALMIERI, C. PETROVICH, A. PISENT, L. PIGA, G. PRETE,
M. TONEZZER, M. RE, V. RIZZI, D. RIZZO, D. ZAFIROPOULOS, P. ZANONATO: The SPES direct target project at
LNL
Zakopane Conference on Nuclear Physics, Zakopane (Poland), September 4-10, 2006
Medical, Energetic and Environmental Applications
K. W. BURN, L. CASALINI, S. MARTINI, D. MONDINI, E. NAVA, G. ROSI, R. TINTI: Final design and construction
issues of the TAPIRO epithermal column
12 th Inter. Congress on Neutron Capture Therapy (ISNCT-12), Takamatsu (Japan), October 9-13, 2006, ISNCT
Proceedings, p. 564 (2006)
G. GAMBARINI, S. AGOSTEO, S. ALTIERI, S. BORTOLUSSI, M. CARRARA, S. GAY, M. MARIANI, C. PETROVICH,
G. ROSI, E. VANOSSI: Dose imaging with gel dosimeters in phantoms exposed in reactor thermal columns
designed for BNCT
12 th Inter. Congress on Neutron Capture Therapy (ISNCT-12), Takamatsu (Japan), October 9-13, 2006, ISNCT
Proceedings, p. 417 (2006)
P. FERRARI, G. GUALDRINI, E. NAVA, K. W. BURN: Preliminary evaluations of the undesirable patient dose from
a BNCT treatment at the ENEA-TAPIRO reactor
10 th Symposium on Neutron Dosimetry, Uppsala (Sweden), June 12-16, 2006
G. GAMBARINI, S. AGOSTEO, S. ALTIERI, S. BORTOLUSSI, M. CARRARA, S. GAY, E. NAVA, C. PETROVICH, G.
ROSI, M. VALENTE: Dose distributions in phantoms irradiated in thermal columns of two different nuclear reactors
10 th Symposium on Neutron Dosimetry, Uppsala (Sweden), June 12-16, 2006
J. ESPOSITO, G. ROSI, S. AGOSTEO: The new hybrid thermal neutron facility at TAPIRO reactor for BNCT
radiobiological experiments
10 th Symposium on Neutron Dosimetry, Uppsala (Sweden), June 12-16, 2006
K.W. BURN, L. CASALINI, E. NAVA, G. ROSI, R. TINTI: The epithermal column for BNCT at the TAPIRO reactor
Workshop on Innovative Treatment Concepts for Liver Metastases, University Hospital Essen (Germany),
December 7-9, 2006
139
Progress Report 2006

C1 Radioactive Waste Management and
Advanced Nuclear Fuel Cycle Technologies
C Nuclear Protection
In 2006 ENEA’s Department of Nuclear Fusion and Fission, and Related Technologies (Dipartimento Fusione,
Tecnologie e Presidio Nucleare [FPN]) acted according to national policy and to the role assigned to ENEA
FPN by Law 257/2003 with regard to radioactive waste management and advanced nuclear fuel cycle
technologies.
C1.2 Entrustment of ENEA’s Fuel Cycle Facilities and
Personnel to Sogin
The management of ENEA’s fuel cycle facilities (EUREX Saluggia - spent fuel reprocessing; ITREC
Trisaia - spent fuel reprocessing; IPU Casaccia - fuel element fabrication; OPEC Casaccia - postirradiation
analysis) has been assigned to the Società Gestione Impianti Nucleari SpA (Sogin)
through the “Entrustment of Management Act” and integrating annexes, signed by the ENEA
Director General and the Sogin Executive Manager on 23 December 2005. ENEA has seconded its
expert personnel to Sogin to enable full operability of the facilities and to ensure that all the technical
prescriptions be achieved and that the activities concerning site decommissioning be fully exploited.
C1.3 Characterisation, Treatment and Conditioning of
Nuclear Materials and Radioactive Waste
The Laboratory for the Characterisation of Nuclear Materials at ENEA Casaccia, in collaboration with
the universities, carries out nuclear and radioactive material analyses, R&D on reprocessing fuel
used in new-generation reactors (e.g., PYREL project, see sect. B1.1) and is the reference lab for
characterisation of conditioned/non-conditioned radioactive waste. Above all the laboratory has to
guarantee Italy the functions of radioactive-material characterisation and process qualification. The
laboratory manages four operative areas: two classified areas (C-43 Radiochemical Laboratory and
C-25 Technological Hall - Zone A) and two cold areas (CETRA Laboratory and C-25 Technological
Hall - Zone B).
The C-43 Radiochemical Laboratory is authorised through a Category A decree to carry out nondestructive
measurements of radioactive waste and materials by means of gamma spectrometry
systems. Table C1.I briefly describes the available techniques.
The first two techniques, implemented on the SEA radioactive-waste gamma analyser
Progress Report 2006
140

Table C1.I - Techniques used for nondestructive measurements of radioactive wastes & materials
Measurement technique Field of application Input Output
Emission & transmission Characterisation of 220–and - Volume of the drum Total Activity and relative
axial scan segmented 400–litre drums containing - Collimation axial distribution for each
gamma scanner (SGS) gamma emitter radionuclides - Radionuclide library radionuclide identified
- Quantitative analysis reports
Software: Segment 2.1
given by the spectroscopy
software Genie 2k
Angular scanning (AS) Characterisation of 220–and - Detection efficiency curve Number, position and
400–litre drums containing for point source activity of each identified
Software: Ascanio 1.1 gamma emitter radionuclides - Linear attenuation factor radionuclide
- Radionuclide library
- Quantitative analysis reports
given by the spectroscopy
software Genie 2k
Low–resolution Characterisation of 220 and - Detection efficiency curve - Spatial attenuation
emisssion & transmission 400 litre drums containing for point source factor distribution
tomography ((ECT/TCT) gamma emitter radionuclides - Linear attenuation factor - Spatial activity distribu-
- Radionuclide library tion for each
Software Plinius 1.1 - Quantitative analysis reports radionuclide identified
given by the spectroscopy - Total activity for each
software Genie 2k
radionuclide identified
In-situ object counting Characterisation of various - Accurate description of - Total activity for each
system (ISOCS) objects containing gamma measurement configura- radionuclide identified
emitter radionuclides
tion
Software: Genie2k
- Radionuclide library
ISOXSW
(SRWGA, fig. C1.1), have the same field of application but are characterised by different levels of
accuracy according to activity and density distribution of the matrix, while for the electrical capacitance
tomography/transmission computer tomography (ECT/TCT) techniques these characteristics have no
influence on the reliability of results. The weakness of ECT/TCT is only the measuring time (typically
18 h/drum). It is also worth noting that when the segmented gamma scanner (SGS) is used, variations in
matrix density and in the activity distribution inside the matrix could lead to overestimation or
underestimation of the real activity up to a factor of 10.
The ISOCS (fig. C1.2) is used in a wide variety of measurement applications. The most important
characteristic of the ISOCS is its capability to obtain radionuclide activity by applying pre-defined geometry
templates in the analysis software: defining the
template, the user obtains an evaluation of the
overall efficiency curve without needing
experimental calibration. However, the
measurement configuration has to be defined and
reproduced by the user with good accuracy, even
though this is not always allowed because of
practical problems.
In 2006 experimental measurement campaigns
were carried out to validate the ISOCS and assess
Fig. C1.1 - SRWGA gamma system
141
Progress Report 2006

C1 Radioactive Waste Management
Fig. C1.2 - ISOCS detector
C Nuclear Protection
its performance. Measurements were performed on several
220-litre drums (with differing matrix density and distribution)
equipped with various configurations of certified gammaemitting
sources placed at different positions. The aim was to
illustrate the problems that can compromise the application
of this measurement technique and to simulate the following
real situations:
• characterisation of 220-litre raw waste drums and
conditioned waste drums;
• localisation and quantification of buried and covered
activity;
• quantification of activity in sealed containers.
The results obtained have clearly defined the field of application of ISOCS (for small samples or, with
appropriate measurements and analysis procedures, for large samples and buried activity) and will
be the basis for future research activities and for setting up the technical procedure to be accredited
according to ISO-17025 in autumn 2007. The same activity was foreseen for 400-litre drums, but it
was put forward to 2007 because of mechanical problems with the SRWGA.
The CETRA Laboratory is specialised in the formulation and characterisation of matrices for
conditioning toxic and/or radioactive wastes. In accordance with the Technical Guide 26 of the
Italian Agency for Environmental Protection and Technical Services (APAT) for the management of
radioactive waste, the laboratory studies, qualifies and sets up processes for treating and
conditioning radioactive wastes and performs the chemical and physical-mechanical
characterisation of the conditioned products, obtained via the employment of chemical elements
simulating real waste.
Fig. C1.3 - Tensile strength tester
C1.4 Radioprotection and Human Health
The qualification of the conditioning
processes consists of a series of activities
aimed at demonstrating that the matrix
resulting from the conditioning process
complies with the minimum requirements for
interim storage, transport and clearance of
waste. The major tests performed are tensile
strength (fig. C1.3); cyclic temperature
gradient resistance; radiation damage
resistance; fire resistance; leaching test; free
liquid absence; bio-degradation resistance
and immersion resistance. Some tests (biodegradation
resistance, leaching test and
radiation damage resistance) are performed
in cooperation with other ENEA laboratories.
Methodological proposal for the evaluation of a physiological comfort index in
indoor environments
Carried out in the framework of a research activity supported by the National Institute of
Occupational Safety, Health and Prevention (ISPESL), the aim is to propose a physiological comfort
Progress Report 2006
142

and Advanced Nuclear Fuel Cycle Technologies
index for workers wearing personal protective equipment. Many attempts have been made to combine
environmental with physiological parameters in order to develop a single index. Currently there are many
indices, but none of them is widely accepted. The main reason lies in the great complexity and plurality of
interactions among the main factors to take into account when defining the index. A survey on
physiological comfort indices showed that the Physiological Strain Index (PSI) is the most appreciated as
individual reactions to it are only based on core temperature and heart rate. Moreover, the PSI can assess
in real time physiological responses both to heat and heat strain between any combination of climate,
clothing and work rate. The PSI does not consider sweat rate because of the intrinsic difficulty in
performing an on-line measurement: nevertheless this term should be taken into account, especially in the
case of short and repeated operations.
The work, carried out with La Sapienza University of Rome, proposes the use of two physiological comfort
indices: the first concerns long-lasting operations; the second, short and repeated operations. In the first
case the suggested index is like the PSI index; it is possible to take into account the effects of cooling
devices operating with personal protective equipment by reducing the index value. In the second case
another term linked to sweat rate can be introduced. To calculate the value of coefficients in the new
physiological comfort index, it is necessary to carry out a measurement campaign on a sufficiently wide
population. These measurements should lead to a quantitative evaluation of the importance of the term
considering the presence of cooling systems in personal protective equipment.
LCA of strippable coating and the principal competing technology used for nuclear
decontamination
Life cycle assessment (LCA) is a systematic way to evaluate the environmental impact of products or
activities throughout their entire life cycle by following a “cradle to grave” approach. This approach implies
the identification and quantification of emissions, materials and energy consumption, which affect the
environment at all stages of the entire life cycle of the product. The application of strippable coatings is an
innovative technology for decontamination of nuclear plants and for any decontamination project where the
purpose is to remove surface contamination (such as polychlorobiphenyls (PCBs), asbestos particles, etc).
It effectively reduces hazardous residuals, at low cost. An adhesive plastic coating is applied on the
contaminated surface. The strippable coating is allowed to cure for up to 24 h, after which it can be easily
peeled. The coating traps the contaminants in the polymer matrix. Strippable coatings are non-toxic and
do not contain volatile compounds or heavy metals. Since the coating constitutes solid waste, disposal is
easier than treating contaminated liquid wastes produced by the baseline technology.
The competing baseline technology is the steam vacuum cleaning technology based on superheated
pressurised water, used to remove contaminants from floors and walls. The LCA was carried out to
compare the strippable coating with the steam vacuum technology. The functional unit of the study is
represented by a surface of 1 m 2 to be decontaminated. The results of LCA achieved using Sima
Pro 5.0 ® software confirmed the good environmental performance of strippable coatings. The storage
phase is the phase showing the most important differences between the two technologies. For this reason
this phase was studied in detail, even from the economic point of view. A simplified economic analysis of
only the storage phase showed higher costs for the steam vacuum technology. In a further development
of the work, the cost of all the phases could be examined in order to confirm or not the best behaviour of
the strippable coating, also from the economic point of view.
C1.5 Integrated Service for Non-Energy Radwaste
Radioactive waste is generated in a broad range of activities involving the use of radioactive material in
different conventional fields, such as medicine, industry, agriculture, research and education. In general the
waste generated in these fields is often limited in volume and activity; however it has to be managed like
the other radwaste from nuclear power plants.
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C1 Radioactive Waste Management
C Nuclear Protection
Italy has no radwaste national repository, so Government has entrusted ENEA with the management
of radioactive waste coming from small producers (collection, transport, characterisation, treatment,
conditioning, interim storage, or release for waste with short life radio-nuclides, after their
radioactivity decay). ENEA has organised a special technical-operative service, called “Integrated
Service”, and is responsible for the guidance, supervision and control of the whole cycle of waste
management. ENEA has entrusted NUCLECO SpA with the operative and commercial tasks, and
offered the company access to specific facilities and infrastructures, located at the Casaccia
Research Centre. The two parties drew up a special agreement laying out mutual duties and
responsibilities.
Integrated Service has also collected thirty disused sealed radioactive sources with Cs-137 and
Co–60 and about seventy grams of Ra-226 no longer used in medical therapy. Except for this last
type of waste, ENEA becomes the owner of all the radwaste collected and deals with the final
release, leaving the waste producers free of any juridical responsibility. Integrated Service is available
to private companies operating in this sector. The companies supply collecting services and
temporary storage. ENEA provides qualification for the companies and gives them specific
technical-operational procedures.
C1.6 Transport of Nuclear Material
Packaging for transport of radioactive material
The Laboratory for Characterisation of Nuclear Materials is a permanent member of the European
Network of Testing Facilities for the Quality Checking of Radioactive Waste Packages. (EN-TRAP:
created in 1992 on the initiative of the European Commission.) The objectives are to promote
collaboration on the development, application and standardisation of quality checking for waste
packages. The network involves the reference laboratories of the European Union member states
that verify the regulatory issues on waste packages. In this framework the laboratory participates in
steering committee meetings and in technical-scientific activities regarding the characterisation of
the radioactive wastes, promoted by the working groups. In 2006 one meeting of the steering
committee took place in April at Winfrith (UK) and one meeting of working group D (Quality
Checking of ILW/HLW), in September in Brussels.
As ENEA-NUCLECO (see sect. C1.5) has to store, transport and dispose of radio needles and some
large radioactive sources, the following activities have been carried out:
• Preparation of a management system for quality assurance. This is fundamental for the approval
process of a package model by the competent authority (APAT) and the emission of the
corresponding certificate.
• Design of a new dual-purpose (storage and transport) package using the scale factors already
developed in the past for the CF66. The first instance will contain only the radium needles. The
objective is to improve warehouse safety. Once the certificate of approval type B(U) is released
by APAT, the transport modality will be studied, with the inclusion of cobalt and caesium sources.
• Obtaining authorisation to dismount the irradiation heads and their packing according to the IAEA
radioactive sources catalogue.
• Once the management system for quality assurance is set up, application will be made to renew
the certificate of approval for CF6 and CF66 packaging.
• Qualification of industrial packaging (IP) of type A for transport of radioactive material.
• Contributions to the updating and revision of national and international transport regulations.
• Participation in the IAEA group for “Regulations for the Safe Transport of Radioactive Material”
(Transport Safety Standards Committee).
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and Advanced Nuclear Fuel Cycle Technologies
C1.7 Disposal of Radioactive Waste
Artificial barriers for disposal units
The research regarded mainly cementitious materials and was carried out in collaboration with the
Departments of Structural Engineering and of Chemistry-Physics of Milan Polytechnic, EN.CO. Srl a wellknown
laboratory working with concrete and the Experimental Testing Laboratory of CESI-ISMES SpA.
The work was divided into three steps: identifying the optimal characteristics for structural concrete and
grout; investigating their properties through artificial aging tests; testing their efficiency in module
prototypes.
A systematic study was performed to establish a reference mix-design, taking into account all possible
environmental attacks in Italy for a period of 300 years. A series of tests simulating the chemical and
physical attacks forecasted was then carried out on several specimens compounded with aggregates from
four different Italian regions. In addition a suitable concrete mix-design was obtained through the
ANSI/ANS 16.1.1986 leakage tests, carried out on a 300x300-mm-wide concrete slab. Four full-scale
module prototypes were then built, within the framework of the project on designing a final repository for
low-activity radioactive waste, from the same concrete mix and submitted to a waterproofing test at the
Ismes Laboratory. The main object of the study was leaching, i.e., the selective transport of particles
occurring in a material once water seepage is established, which can be activated by weathering of the
material in the longer term. One prototype was also submitted to a seismic test of 1 g.
All the trials and tests confirmed the adequacy of the design and the materials chosen, and analysis of the
achievements indicated ways to improve the module performance, e.g., by using fibre-reinforced concrete
with new-generation additives and considering minor strength requirements under dynamics stresses.
The theoretical studies were completed by using a Monte Carlo simulation method based on the theory of
branching stochastic processes. Also addressed was the issue of radionuclide transport through the
artificial porous matrices constituting the engineered barriers: the complexity of the phenomena involved,
augmented by the heterogeneity and stochasticity of the media in which transport occurs, renders classical
analytical-numerical approaches scarcely adequate for a realistic representation of the system of interest.
This approach, applied in the artificial porous matrices hosting the waste (near field), can certainly be
usefully extended to study the phenomena of advection and dispersion of radionuclides in the natural rock
matrix of the host geosphere (far field).
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Progress Report 2006

D1 Advances in the IGNITOR Programme
D Miscellaneous
The IGNITOR project has continued to progress both in the machine engineering design and related
auxiliary systems and in the definition of the physics programme. The validity of the objectives of the
IGNITOR programme and of its design solutions were reaffirmed by PH. Rebut at the latest EPS
meeting in Rome: i) in order to prove the scientific feasibility of relevant fusion reactors, burning
plasmas with Q > 50 should be produced and studied; ii) copper magnets are the most convenient
solution for machines capable of reaching this objective; iii) experiments that do not include a
divertor are the most efficient at producing the highest plasma currents with the best confinement
parameters.
While ignition scenarios in IGNITOR have been extensively analysed in the past, plasma regimes and
the physics objectives that can be achieved when operating the machine at lower parameters have
been further explored. In particular, at B T ≅9 T, and I p ≅7 MA in the “extended limiter” configuration or
I p ≅6 MA in the double null configuration, in D-T plasmas, simulations performed with the JETTO
code show that the ideal ignition temperature can be attained. This is the point where the energy
loss by Bremsstrahlung emission is compensated for by α-particle heating, and the density can be
raised further without encountering a radiation barrier. These regimes require a certain amount of ion
cyclotron resonance heating (ICRH) (5-8 MW at ~90 MHz). A parametric study of the power
deposition profiles as a function of minority concentration, minority species, and frequency range
was carried out by solving the 2D full wave equation, which describes the plasma-wave interaction,
in toroidal geometry.
IGNITOR can operate with a double-null configuration at B T ≅13 T, and I p ≅9 MA. On the basis of
recent scalings, the power threshold to access the H-mode regime is considerably lower than
previously estimated. The expected plasma parameters were evaluated by using a 0-D model,
which showed the existence of a relatively broad region of operation corresponding to Q ≅ 10, even
in the pessimistic case of rather flat density profiles. With more peaked profiles (e.g., n 0 / ≅ 1.5),
of the type being observed in some JET experimental data, the attainable plasma parameters are
found to improve considerably and values of Q much larger than 10 can be attained. The
construction of the four-barrel two-stage IGNITOR
pellet injector, in collaboration with the Oak Ridge
National Laboratory (ORNL), is nearly completed
(fig. D1). The development of new, fast pulse
shaping valves will make it possible to reach pellet
velocities of 4 km/s. The propulsion system, built in
Italy, will be shipped to ORNL for final integration
with the ORNL cryogenic and control systems, and
pellet performance characterisation. The possible
application of the injector to JET has been
explored, but other large existing devices are also
being considered.
The design of the full set of electromagnetic
diagnostics for the IGNITOR experiment and their
integration with the plasma chamber has been
completed. Because the estimated neutron flux at
the first wall during high-performance discharges
is expected to cause a sensible, although
Fig. D1 – a) The ENEA Frascati sub-system of the Ignitor pellet
injector during testing at Criotec Impianti in Chivasso (Turin, Italy). b)
In-flight picture of a 3-mm D2 pellet, travelling at about 1.2 km/s
Progress Report 2006
146

Fig. D2 – a) Lateral and b) top views of the IGNITOR machine obtained from
integration of the CATIA CAD drawing of the detailed design of all the
components of the machine itself
a)
reversible, degradation of the inorganic insulator surrounding the
conductors, an R&D program aimed at selecting insulator
materials and fabrication procedures has been established. Two
prototype coils made of pre-insulated nickel wire immersed in a
magnesium oxide weakly bonded powder were manufactured in
collaboration with the University of Lecce (Italy) and SALENTEC.
Vacuum tightness is provided by sintered alumina cases or by
oxide ceramic composite wrapping layers.
An alternative diagnostic method for plasma position control has
been proposed: using a multilayer mirror as the dispersing
element for the soft x-ray radiation emitted from the plasma outer
region and a gas electron multiplier detector would allow the
radiation from the lower or upper part regions to be diffracted to
the 2D detector placed outside one of the machine horizontal
ports, not in direct view of the plasma, to minimise the
background radiation noise. This system should measure the
plasma position and detect any movements with sufficient spatial
and time resolution to be used for real-time control of the vertical
position.
b)
The detailed design of the machine was completed during 2006,
taking as reference the maximum performance scenario
(11 MA/13 T/extended limiter). This design will be checked,
referring to the double-null configuration at B T ≅13 T, and
I p ≅9 MA.
The machine integration was also completed during the reporting
year, starting from the CATIA CAD of each component of the
machine. As far as known, this is the first time that a detailed CAD-based integration has been performed
for such a complex apparatus. Figure D2 shows an example of the integration results.
D2 Ultra-Pure Hydrogen Production
Project (FIRB RBAU01K4HJ) funded by the Italian Ministry of Education, University and
Research. Long-term tests (more than one year) of thin-wall Pd-Ag permeator tubes produced at ENEA
Frascati laboratories were carried out and the capability to produce ultra-pure hydrogen as well as the
durability of the permeators were demonstrated [D1].
A membrane process for producing hydrogen from hydrocarbon and alcohol reforming was developed
[D2–D4] and a Pd-Ag multi-tube membrane reactor capable of producing 6 L/min of pure hydrogen was
[D1] S. Tosti et al., Long-term tests of Pd–Ag thin wall permeator tube, J. Membrane Sci. 284, 393–397 (2006)
[D2] S. Tosti et al., Procedimento a membrana per la produzione di idrogeno da reforming di composti organici, in particolare idrocarburi o
alcoli, Domanda di brevetto per invenzione industriale n. RM2006A000102 del 01.03.2006
[D3] S. Tosti et al., Design and characterization of membrane reactors for producing hydrogen via ethanol reforming, ENEA Internal Report
FUS TN BB-R015 (2006)
[D4] S. Tosti et al., Pd membrane reactor design, Desalination 200, 676-678 (2006)
References
147
Progress Report 2006

Miscellaneous
Fig. D3 – The multi-tube membrane reactor
Fig. D4 – Particulars of
the flange supporting
the Pd–Ag permeator
tubes
built (figs. D3 and D4). A lot of
experimental work concerning the
methanol and ethanol steam reforming
reactions in Pd-Ag membrane reactors
was performed [D5–D7] and it was
demonstrated that the membrane is able
to promote the reaction conversion
beyond the thermodynamic limit. In
particular, at 450°C a high hydrogen yield
was attained via the ethanol steam
reforming on a Ru–based catalyst.
Figure D5 reports the hydrogen yield
measured in the shell side of the Pd–Ag
membrane reactor.
Shell side
hydrogen yield (%)
100
80
60
40
200k Pa
150k Pa
20
0
0 5 10 15 20 25
Feed flow rate (g h -1 )
Fig. D5 – Shell-side hydrogen percent yield at 450°C and
feed molar ratio H 2 O/EtOH=13
D3 Non-ITER Activities
Cryogenic testing of superconductive current leads for CERN. Since September 2004 ENEA
has been responsible for the cryogenic tests of the complete series of 6000 A and 13000 A Large
Hadron Collider (LHC) HTS current leads, consisting of 333 units.The whole job also included, as a
first step, testing of the pre-series lead production, manufactured and assembled at CERN.
The main campaign of tests involving the current leads produced by the BINP laboratory, Russian
Federation (6 kA) and by CECOM, Italy (13 kA), started in the second half of 2005, and proceeded
throughout 2006, thereby meeting the tight time schedule requested by CERN.
Thanks to ENEA’s dedicated measurement facility, characterised by a high-precision signal
acquisition system, the results showed very good reproducibility of both the electrical and the
thermo-hydraulic performances of the leads in LHC-relevant operating conditions and all the tested
samples fully met the requirements of the CERN technical specifications.
The measurement campaign is foreseen to be completed within the first months of 2007.
Progress Report 2006
148

D4 Condensed Matter Nuclear Science
Material science and calorimetry. Cold fusion matter, now more properly renamed “condensed matter
nuclear science”, has been debated over for the last two decades [D8]. Prestigious institutions have been
working in this field and some have cooperated successfully. It was discovered [D9, D10] that the
phenomenon of excess power production was a threshold effect occurring only if the average deuterium
concentration in the palladium lattice was not less than 0.9 (atomic fraction). Studies performed at ENEA
Frascati highlighted the fact that the high loading of deuterium in the lattice was not reproducible when
using commercial palladium. Hence, a wide material-science study was carried out to produce a metal with
a proper metallurgical structure, capable of giving a very high deuterium concentration during
electrochemical loading.
Under contract agreements ENEA delivered cathodes prepared with such a particular palladium to SRI
International (California USA) and Energetics Ltd. (US company with a research centre in Israel). A
reasonable level of transferred reproducibility was achieved by the three groups and this was one of the
reasons for promoting a two-phase research project with government funding in the USA to revisit the
“cold fusion effect”. ENEA was involved in the programme as ENEA cathodes were selected for the
research. During the first phase SRI International was charged with replicating the results obtained with
ENEA’s cathodes and with the calorimeters used by Energetics Ltd. Phase 1 was concluded at the
beginning of 2007 with results well above the objectives defined by the US Government referees, and
continuation of the project towards Phase 2 was approved. In the second phase, the US Naval Research
Laboratory is also involved in replicating the experiments.
The Italian Ministry of Economic Development (MSE) supported a two-year project (Produzione di Eccesso
di Potenza in Metalli Deuterati) to improve the material science study and to gain an enhanced signal/noise
ratio. Material science studies have been extended to
surface physics aspects and to interphase physics, with the
involvement of the University of Rome La Sapienza. The
Italian project began in January 2006 and overlapped Phase
1, so the two projects have been developed in parallel.
During this period both ENEA and SRI International gained a
reproducibility not less than 60% with a signal/noise ratio well
above the measurement uncertainty. Figure D6 shows the
ENEA flow calorimeters; figure D7 the excess power
observed during the experimental campaign performed at
ENEA (experiment L17) and figure D8 the increase in the
electrolyte temperature associated with the excess.
The power gain was 500% with an input and an output
power of 0.1 W and 0.6 W, respectively (measurement
Fig. D6 – Calorimeter room at ENEA Frascati
[D5] F. Gallucci et al., Methanol and ethanol steam reforming in membrane reactors: An experimental study, Int. J. Hydrogen Energy (2006),
doi: 10.1016/j.ijhydene.2006.11.019
[D6] A. Basile et al., Co-current and counter-current modes for methanol steam reforming membrane reactor: Experimental study, Catalysis
Today 118, 237–245 (2006)
[D7] A. Basile et al., The pressure effect on ethanol steam reforming in membrane reactor: experimental study, Desalination 200, 671–672
(2006)
[D8] M. Fleishmann and S. Pons, J. Electroanal. Chem. 261, 301 (1989).
[D9] M. McKubre et al., Excess power observation in electrochemical studies of the D/Pd system; the influence of loading, Proc. 3 rd Inter.
Conference on Cold Fusion (Nagoya 1992) p. 5
[D10] K. Kunimatsu et al., Deuterium loading ratio and excess heat generation during electrolysis of heavy water by a palladium cathode in a
closed cell using a partially immersed fuel cell anode, Proc. 3 rd Inter. Conference on Cold Fusion (Nagoya 1992), p. 31
References
149
Progress Report 2006

Miscellaneous
0.6
Fig. D7 – Input and output (upper curve) power evolution in
the experiment L17
W out
W in
Power (W)
0.4
0.2
uncertainty ±20 mW). Of the many experiments
performed with hydrogen, not one has produced
excess power.
Temperature (°C)
0
200000 240000 280000
Time (s)
30.0
T cell
29.0
28.0
27.0
T box
26.0
25.0
200000 240000 280000
Time (s)
Fig. D8 – Electrolyte temperature evolution:
temperature increase is well correlated with the
excess of power
Figure D7 shows that during the excess the input
power decreases due to the power supply
operating mode (galvanostat) because the strong
excess of power (up to 620 mW) caused the
temperature of the electrolyte to increase
(fig. D8), which was responsible for reduced
electrolyte resistivity and also of the cathode
interfacial impedence, so a lower voltage was
required to maintain the set point current. The
consequence was a reduction in input power
during the burst. The conclusion was 620 mW of
output with an input of 125 mW, hence an output
gain of 500%.
Similar results were observed with ENEA’s
cathodes at SRI International. The statistics
revealed that the cathode lots producing excess
power at ENEA had, in general, the same
behaviour at SRI. On the contrary, any lot that did
not produce excess power at ENEA did not
produce it at SRI International either.
Despite the very high excess of power observed, the most relevant point is the energy gain
associated with power gain. Energy gains up to some MJ have been observed in ENEA’s cathodes
(6.25 keV per atom into the electrode). Energy gain is a crucial point because a large amount of the
energy produced cannot be simply justified as a chemical effect if sharing the energy between all
the atoms embedded in the electrode produces, as already reported, an energy per atom well above
a few eV. A possible explanation is that there is a mechanism accumulating energy in the system at
a very slow rate so that no negative power gain is detected by the calorimeter because outside the
detection limits. In the case of a fast energy release, as in the Wigner effect, an apparent excess of
power would be revealed by the calorimeter; however, such an energy gain would have to be of the
order of a few eV/atom in order to be ascribed to a chemical effect, which is in contrast with the
experimental observations.
The amount of energy gain and the occurrence of the effect with deuterium and not with hydrogen
point in the direction of a nuclear fusion reaction between two deuterons producing, in the lattice,
4 He and heat. This is in agreement with preliminary measurements of 4 He [D11-D14], which reveal
an increase in the concentration above the ambient level, consistent with the energy gain.
In 2005 a very positive co-operation was started in the field of materials science with the Materials
Branch of the Naval Research Institute of Washington DC. This ongoing research activity is funded
by the Office of Naval Research Global (ONRG), London UK, and an important experiment has
already been carried out at the Brookhaven National Laboratory, USA. X-ray diffraction was
performed during electrochemical loading of cathodes prepared at ENEA in order to study the
palladium hydride (deuteride) in the so far unexplored region of loading above H(D)/Pd>1. The
experiment was concluded successfully by collecting more that 240 spectra.
Progress Report 2006
150

The support received by MSE has made it possible to extend
the material science study by performing a systematic
characterisation of the surface of cathodes on the basis of
the atomic force microscopy (AFM) and scanning electron
microscopy (SEM) analyses.
Microscopic characterisation of finished electrodes
before electrolysis. Microscopic characterisation of the
electrodes was performed in order to correlate the
characteristics of the cathodes before loading and the
excess power production during electrochemical deuterium
loading.
Three cathodes produced from three different lots of rough
materials, but with similar rolling, thermal annealing and
chemical etching processes were analysed and showed
different behaviour in heat production. The deuterium loading
was fairly similar in all three samples and above the threshold
(D/Pd>0.9). The nominal purity of the rough foils is 99.95 for
L25b and L35 and 99.98 for L40.
L25b
100 µm EHT=20.00 kV Signal A=CZ BSD
WD=9.5 mm Mag=200x
L35
a)
b)
SEM analysis. SEM analysis showed different
characteristics in grain size distribution, grain boundary
shape and surface morphology. Figure D9 reports a
comparison of the SEM images of the three samples. Sample
L40 shows an average grain size smaller than the other two
samples; sample L25, the larger dimensions of the grains.
Furthermore, in L40 some of the grain boundaries have a
particular “crest-like” shape, while in L25b and L35 the
boundaries have the more usual “valley-like“ shape. These
characteristics were checked by AFM recording of the
surface profile as it is well known that SEM images can be
misleading in identifying peak or kink features.
EHT=20.00 kV
WD=10.0 mm
Signal A=SE1
Mag=200x
Fig. D9 – SEM images of samples L25b a), L35 b)
The SEM and AFM analyses revealed some differences in the and L40 c)
samples. A specific work devoted to identifying the
correlation between excess of heat and the characteristics of the samples, now in progress, should lead
to identification of the characteristics of the rough material capable of producing Pd cathodes with a
further increasing of the reproducibility of excess power production.
100 µm
L40
100 µm EHT = 20.00 kV Signal A=SE1
WD = 10.0 mm Mag=200x
c)
[D11] V. Violante et al., Some recent results at ENEA, Proc. XII Inter. Conference on Cold Fusion (Yokohama 2005), p. 117
[D12] D. Gozzi et al., J. Electroanal. Chem. 452, 253 (1998)
[D13] M. McKubre et al., The emergence of a coherent explanation for anomalies observed in D/Pd and H/Pd systems: evidences for 4 He and
3 H production, Proc. VIII Inter. Conference on Cold Fusion (Lerici 2000), p 3
[D14] M. Miles et al., J. Electroanal. Chem. 346, 99 (1993)
References
151
Progress Report 2006

December 2006
Organisation Chart
Coordinamento Trasferimento
Tecnologico
Francesco De Marco
Nucleo di Agenzia
Paola Batistoni
Unità Supporto Tecnico Gestionale FUS
Nicola Manganiello
DIREZIONE
Alberto Renieri
*
*
*
Coordinamento Funzionale
Giovanni Coccoluto
Unità Supporto Tecnico Gestionale RAD
Pasquale Di Giamberardino
*EURATOM - ENEA Association
Progress Report 2006
152

Sezione Fisica della Fusione a Confinamento
Magnetico
Alberto Renieri a.i.
Sezione Tecnologie della Fusione
Aldo Pizzuto
Sezione Ingegneria Elettrica ed Elettronica
Alberto Coletti
Sezione Gestione Grandi Impianti Sperimentali
Giuseppe Mazzitelli
Sezione Superconduttività
Antonio della Corte
Gruppo Fisica e Tecnologie del Confinamento
Inerziale
Carmela Strangio
Sezione Ingegneria Sperimentale
Gianluca Benamati
*
*
*
*
*
*
*
Sezione Esercizio Impianti - Casaccia
Corrado Kropp a.i.
Sezione Esercizio Impianti - Trisaia
Corrado Kropp a.i.
Sezione Esercizio Impianti - Saluggia
Corrado Kropp a.i.
Sezione Sorgenti Radiazioni e Applicazioni di
Radiazioni Ionizzanti
Armando Festinesi
Sezione Sistemi Nucleari Innovativi e Chiusura Ciclo
Nucleare
Renato Tinti
Laboratorio Caratterizzazione Rifiuti Radioattivi
Natale Sparacino
153
Progress Report 2006

Abbreviations and Acronyms
ACP
AFM
ALE
ALISIA
APSA
AS
ASDEX
ASDEX-U
ASTEX
BA
BAE
BNCT
BOC
CD
CDP
CEA
CERN
CFC
CHF
CICC
CIRTEN
CMS
CNR
COMPASS-D
CODAS
CPS
CPS
CRPP
CS
CSU
CT/CAT
Cyric
CVD
DAS
DEMO
DCLL
DIS
DISCORAP
DL
DRP
DIII-D
DTL
DVT
DW
activated corrosion product
atomic force microscopy
abrupt large-amplitude event
Assessment of Liquid Salts for Innovative Applications
ageing probabilistic safety assessment
angular scanning
Axisymmetric Divertor Experiment - Garching - Germany
Axisymmetric Divertor Experiment Upgrade - Garching - Germany
Advanced Stability Experiment
Broader Approach
beta-induced Alfvén eigenmode
boron neutron capture therapy
beginning of cycle
current drive
collector depressed potential
Commissariat à l’Energie Atomique - France
Organisation Europeénne pour la Recherche Nucléaire- Geneva
carbon fibre composite
critical heat flux
cable-in-conduit conductor
Consortium for Research in Nuclear Technologies
common manipulator system
Consiglio Nazionale delle Ricerche - Italy
is a highly flexible, medium-sized tokamak - Culham
control and data acquisition system
coolant purification system
capillary porous system
Centre de Recherches en Physique des Plasmas - Villigen - Switzerland
central solenoid
Close Support Unit
computerized (axial) tomography
Cyclotron and Radioisotope Centre - Tohoku University - Japan
chemical vapour deposition
data acquisition system
demonstration/prototype reactor
dual coolant lithium-lead
data one-step
Dipoli Super Conduttori Rapidamente Pulsati (INFN) - Frascati
dome liner
Divertor Refurbishment Platform - ENEA - Brasimone
Doublet III - D-shape. Tokamak at General Atomics - San Diego - USA
drift tube linac
divertor vertical target
drift waves
EAF
EBSD
EC
ECCD
European Activation File
electron backscattering diffraction
electron cyclotron
electron cyclotron current drive
Progress Report 2006
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ECH
ECR
ECRH
EC WGB
ECT/TCT
ECT/TCT
EDA
EDS
EFDA
EFIT
EFF
EISOFAR
ELD
ELM
ELSY
em
EN-TRAP
EOC
EOL
EP
EPM
ETD
EUROPART
EUROTRANS
electron cyclotron heating
electron cyclotron resonance
electron cyclotron resonance heating
electron cyclotron wave Gaussian beam
emission & transmission tomography
electrical capacitance tomography/transmission computer tomography
Engineering Design Activities
electron dispersion spectroscopy
European Fusion Development Agreement
European facility on an industrial-scale transmuter
European fusion file
European Innovative Sodium-Cooled Fast Reactor
electron Landau damping
edge localised modes
European lead-cooled system
electromagnetic
European Network of Testing Facilities for the Quality Checking of Radioactive Waste Packages
end of cycle
end of life
Enhanced Programme (JET)
energetic particle mode
European Transmutation Demonstrator
European Research Programme for the Partitioning of Minor Actinides
European Transmutation
FC
FCS
FDB
FEB
FEM
FIGEX
FMEA
FMECA
FNG
FNS
FPS
FRTC
FTU
FWHM
FWP
FZK
fission chamber
flux-core spheromak
fuel dissolution basket
fast electron bremsstrahlung
finite-element method/model
Fast Ion Generation Experiment
failure mode and effect analaysis
Failure Mode, Effects, and Criticality Analysis
Frascati neutron generator - ENEA
Fusion Neutronics Source - JAERI - Japan
fuel pin simulator
fast ray tracing code
Frascati Tokamak Upgrade - ENEA
full width at half maximum
first-wall panel
Forschungszeuntrum - Karlsruhe - Germany
GAM
GNEP
GRTN
GSI
GSSR
geodesic acoustic mode
Global Nuclear Energy Partnership
National Grid Regulator
Gesellschaft fuer Schwerionenforschung - Darmstadt, Germany
Generic-Site Specific Safety Report
HCLL
HCPB
HEBT
HHFT
HLM
helium-cooled lithium-lead
helium-cooled pebble bed
high-energy beam transport
high heat flux testing
heavy liquid meta
155
Progress Report 2006

Abbreviations and
HL-1Ml
HRP
HRTS
HS
HTS
HX
Circular cross section tokamak modified from HL-1 - Centre for Fusion Science - China
hot radial pressing
high-resolution Thomson scattering
heat source
high-temperature superconductor
heat exchanger
IAEA
I&D
IBW
ICE
ICRH
IDM
IE
IEA
IFE
IFMIF
IMF
INFN-LNL
INTD
IP
IRIS
ISPESL
ISOCS
ITASE
ITB
ITER
ITG
IVT
IVVS
International Atomic Energy Agency - Vienna - Austria
instrumentation and control
ion Bernstein wave
Integral Circulation Experiment
ion cyclotron resonance heating
ITER Documentation Management
Institute for Energy - Petten - the Netherlands
International Energy Agency
inertial fusion energy
International Fusion Materials Irradiation Facility
inert matrix fuel
Istituto Nazionale di Fisica Nucleare, Laboratori Nazionali di Legnaro
International Near-Term Deployment
industrial packaging
International Reactor Innovative and Secure
Institute of Occupational Safety, Health and Prevention
In-situ object counting system
International Trans Antarctic Scientific Expedition
internal transport barrier
International Thermonuclear Experimental Reactor
ion temperature gradient
inner vertical target
in-vessel viewing and ranging system
JAEA
JET
JIPPT-IIU
JRAs
JRC
JT-60U
Japan Atomic Energy Agency - Japan
Joint European Torus - Abingdon - U.K.
Japanese Institute of Plasma Physics Torus-II Upgrade
Joint Research Activities
Joint Research Centre - Ispra - Italy
JAERI Tokamak 60 Upgrade, Naka, Japan
KH
KIZ
Kelvin Helmholtz
Karlsruhe Isochronous Cyclotron
LANL
LBC
LBE
LCA
LED
LFR
LH
LHC
LHCD
LHW
LIDAR-TS
LLL
Los Alamos National Laboratory
liquid bismuth cathode
lead bismith eutectic
life cycle assessment
light emitting diode
Lead-Cooled Fast Reactor
lower hybrid
Large Hadran Collider (CERN)
lower hybrid current drive
lower hybrid wave
laser imaging detection and ranging - Thomson scattering
liquid lithium limiter
Progress Report 2006
156

Acronyms
LLRN
LNL
LOFT
LWR
long-lived radionuclides
Legnano National Laboratory
loss-of fluid test
light-water reactor
MA
MARFE
MAST
MHD
MIUR
MOD
MSE
MSE
minor actinides
multifaceted asymmetric radiation from the edge
Mega Ampère Spherical Tokamak
magnetohydrodynamic
Italian Ministry of Higher Education and Research
metal-organic deposition
motional Stark effect
Italian Ministry of Economic Development
NAR
Nuclear Analysis Report
NAs
Networking Activities
NBI
neutral beam injection
NEA
Nuclear Energy Agency (Paris, France)
NEA
standing committees (NSC - Nuclear Science; NDC - Nuclear Development; CSNI - Safety of Nuclear Installation; RWMC
- - Radioactive Waste Management; CRPPH - Radiation Protection and Public Health)
NETL
Nuclear Engineering Teaching Laboratory - Texas - U.S.A.
NNB
negative neutral beam
NTA
neutron test area
NTM
neoclassial tearing mode
NRG
Nuclear Research Counsultancy Group - Petten - The Netherlands
OCS
ODE
ONRG
ORE
ORNL
OVT
oxygen control system
ordinary differential equation
Office of Naval Research Global
occupational radiation exposure
Oak Ridge National Laboratory - Tennessee - U.S.A.
outer vertical target
PATEROS
PCB
PBC
PD
PDE
PDI
PET
PF
PFC
PFCT
PFW
PIE
PIT
PLD
PMT
PPCS
PRA
PRF
PRHH
PSI
PWR
Partitioning and Transmutation European Roadmap for Sustainable Nuclear Energy
polychlorobiphenyls
pre-brazed casting
power density
partial differential equation
parametric decay instability
pin expansion tool
poloidal field
plasma-facing component
plasma-facing component transporter
primary first-wall
postulated initiating event
powder-in-tube
pulsed-laser deposition
photomultiplier tube
Power Plant Conceptual Studies
probabilistic risk assessment
permeation reduction factor
preliminary remote handling handbook
Physiological Strain Index
pressurised water reactor
157
Progress Report 2006

Abbreviations and
QA
quality assurance
RABiTS
RACE
RAPHAEL
RD
rf
RFQ
RFX
RH
RO
RT
rolling-assisted biaxially texture of substrate
Reactor-Accelerator Coupling Experiments
Reactor for Process Heat, Hydrogen and Electricity Generation
rolling direction
radiofrequency
radiofrequency quadrupole
Reversed Field Pinch Experiment - Padua - Italy (Association EURATOM-ENEA)
remote handling
responsible officer
room temperature
S/A
SCD
SEM
SFE
SGS
SOL
SP
SPES
SRWGA
SSC
SSQLFP
ST
subassembly
single crystal diamond
scanning electron microscopy
stacking fault energy
segmented gamma scanner
scrape-off layer
screw pinch
Study for the Production of Exotic Species
SEA radioactive-waste gamma analyser
solid steel cathode
steady-state quasi-linear Fokker-Planck
spherical torus
TAa
TAE
TALDICE
TBM
TES
TEPC
Transnational Access Activities
toroidicity-induced Alfvén eigenmode
Talos Dome Ice
test blanket module
tritium extraction system
tissue-equivalent proportional counters
TEXTOR Torus Experiment for Technology Oriented Research. Tokamak at Jülich Germany (Association EURATOM –
FZJ)
TFA
trifluoroacetate
TFAS
toroidal field advaced strands
TFC
toroidal field power
TIG
tungsten inert gas
TITG
trapped ion ITG
TOFOR
time-of-flight neutron spectrometer optimized for high counting rate
TPR
tritium permeation rate
TRADE
TRIGA Accelerator-Driven Experiment - ENEA- Casaccia
TS
Thomson scattering
TUCN
Technical Universitry of Cluj-Naoica - Romania
TUD
Technical University of Dresden - Germany
UCI
UKAEA
UT
University of California at Irvine - Usa
United Kingdom Atomic Energy Agency
University of Texas - USA
VELLA
VDS
VHTR
Virtual European Lead Initiative
vent detritiation system
Very High Temperature Reactor
Progress Report 2006
158

Acronyms
VMS
VSM
VTA
vertical module segmentation
vibrating sample magnetometer
vertical target assembly
WKB
Wenzel, Kramer, Brillouin code
XRD
ZF
ZFC
x-ray diffraction
zonal flow
zero field cooling
159
Progress Report 2006

2006
PROGRESS REPORT