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PROGRESS REPORT
Euratom–ENEA Association
fusion activities
2010
ITALIAN NATIONAL AGENCY FOR NEW TECHNOLOGIES ENERGY AND SUSTAINABLE
ECONOMIC DEVELOPMENT
Nuclear Fusion Unit

This report was prepared by the Scientific Publications Office from contributions provided by the scientific and technical
staff of ENEA’s Nuclear Fusion Unit
Scientific editors: Paola Batistoni, Gregorio Vlad
Cover, artwork and design/composition: Marisa Cecchini
English revision: Laura Bocci
See http://www.fusione.enea.it for copy of this report
FNG with an assembled ITER mock–up
Radiography of a plastic
foil with holes
ITER divertor IVT qualification
prototype manufactured by
ENEA–Ansaldo
Pre-compression ring
dismantling after test
Published by:
ENEA - Servizio Diffusione Tecnologie
Edizioni Scientifiche,
Centro Ricerche Frascati, Tel.: +39(06)9400 5016
C.P. 65 Fax: +39(06)9400 5015
00044 Frascati Rome (Italy) e-mail: marisa.cecchini@enea.it

progress report
2010
001
2010
Progress Report
Euratom - ENEA Association
Fusion Activities

Contents
006 Preface
008 Anniversary
010 Magnetic Confinement
010 Introduction
011 FTU Facility
011 Summary of the machine operation
013 FTU Experimental Activity
013 Lower hybrid
016 Alfvén eigenmodes: modeling, experimental results
and future plans
017 Electron cyclotron
021 Liquid lithium experiments
023 Diagnostic developments
029 Plasma–wall interaction
Euratom–ENEA
Association
2010annual
fusion
activities
030 Plasma Theory
030 Alfvén modes in the presence of a magnetic island
031 Nonlocal theory of energetic–particle–induced
geodesic acoustic mode
032 Kinetic structures of shear Alfvén and acoustic wave
spectra in burning plasmas
032 ITPA benchmarking activity using HMGC
033 TAE eigenfunctions for ITER magnetic diagnostics
033 ITM–TF activities
033 The new hybrid MHD–gyrokinetic code: HYMAGYC
034 Applications and further implementation of the
eXtended HMGC hybrid code
035 Parametric form of equilibrium particle distribution
functions for implementations in HMGC
035 Design of energetic particle equilibrium distribution
functions
036 A new model for self–organized critical dynamics, with
implication of the fishbone–like instability cycle
037 Asymptotic techniques in the study of the propagation
of high frequency (lower hybrid range) electromagnetic
waves
037 Analytical and numerical studies of the cold
electromagnetic LH wave equation in the mode
conversion regimev
037 Hamiltonian perturbation theory in studying the LH
wave propagation in burning plasmas
038 The propagation and absorption of lower hybrid
waves in H–mode plasma with very sharp density
gradient at the pedestal
038 Tritium minority heating with mode conversion of fast
waves
038 Stellar winds or matter–jet seeds from disk plasma
configuration

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2010
003
038 JET Collaboration
039 Pellet studies
039 High beta operation
040 Dust
040 L to H mode transition
040 Safety factor profile determination
041 Upgrade of neutron profile monitor (KN3N)
041 Elaboration of JET programme for 2011
041 Coordination of other Italian partners
042 Plasma control
042 References
046 Fusion Advanced Studies Torus
046 Introduction
047 Physics
047 Transport simulations
048 Energetic particle physics in H–mode scenario with combined NNBI and ICRH
049 Plasma–Wall interaction activities and optimization of the divertor geometry
049 Diagnostics
050 Design Description
050 Divertor design
051 First wall design
052 Vacuum vessel design
052 Toroidal field coil system
053 Toroidal field ripple reduction
053 Poloidal field coil system
054 Poloidal field coil system flexibility and plasma control
054 Cooling system
054 Remote handling
055 References
056 Technology Programme
056 Introduction
057 Divertor, First Wall, Vacuum Vessel and Shield
057 Manufacturing technology for the ITER inner vertical target
057 Qualification of ultrasonic non–destructrive testing method for plasma facing
components
058 Analysis of the ITER divertor cassettes
060 Breeder Blanket and Fuel Cycle
060 European Breeding Blanket Test Facility design and construction
060 TRIEX loop for studying technologies of tritium extraction from Pb–17Li
060 Electromagnetic load analysis for the design of the blanket manifold pipe
concept for ITER
061 Development of method for highly tritiated water handling in ITER tritium plant
061 Training activities
062 Magnet and Power Supply

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2010
004
062 Optimization of the toroidal field ripple reduction system
062 Three–dimensional magneto–static analyses for ITER
063 Analysis of the ripple effects due to the ferromagnetic materials in the NBI
magnetic shields and the test blanket modules
063 Assessment of the NBI magnetic field reduction system
064 Assessment of the effects of the ferromagnetic material in the ITER building
structures
065 Pre–compression rings final design qualification
065 Remote Handling and Metrology
065 ITER in vessel viewing system
068 Neutronics
068 Neutronics shielding experiment on a mock–up of ITER: dose measurement in
the magnet coils
069 ITER diagnostic port integration: development of the ITER MCNP 40° model
069 Neutronic calculations in support of the design of the ITER high resolution
neutron spectrometer
070 Three–dimensional neutronic analysis of the ITER in–vessel coils
072 Neutron and gamma spectra behind ITER blanket modules and in divertor
072 Neutronic analysis of ITER divertor rails
073 Development of a high resolution compact neutron spectrometry using
diamond detectors
073 Measurement of Ti profiles with the ITER radial neutron camera
074 Measurement of fuel ratio profiles with the ITER radial neutron camera
075 Materials
075 Development of a multi–scale methodology for composite structural modelling
and validation of modelling procedure by mechanical testing
075 Safety and Environment, Power Plant Conceptual Studies and Socio Economics
075 Dust mobilization studies
076 Validation of the PACTITER computer code and related fusion specific
experiments in CORELE loop
077 Safety analyses by hazard and operability studies
078 Broad Approach
078 JT–60SA
079 International Fusion Materials Irradiation Facility
082 DEMO R&D in the Broader Approach activities
083 JET Fusion Technology
083 JET housekeeping wastes detritiating
084 GEM–based neutron detector for 2.5 and 14 MeV
084 Compact neutron spectrometer
085 Cryogenics
085 Liquid helium service
086 References
088 Superconductivity
088 Introduction
089 Cable–in–Conduit Conductor
089 Manufacturing of ITER poloidal field conductor samples

progress report
2010
005
089 Compaction of ITER CS empty tubes
090 Feasibility verification of the superconducting proposal of the TF magnet
system of FAST
090 Heat exchanger design for the 30 kA current leads of ENEA CICC upgrade
facility
090 Study on the effect of strand bending on the voltage–current characteristic of
Nb 3 Sn CICC
091 Role of the cross section geometry in rectangular Nb 3 Sn CICC performances
091 R&D on Superconducting Materials
091 Study on Nb 3 Sn strand
092 Study on NbTi strands
092 Quality control monitoring of NbTi strands for JT–60SA TF coils
093 ITER NbTi strand benchmarking tests
093 High Temperature Superconductors
093 Study of Ni–Cu based alloy tapes for YBCO coated conductor application
094 Oxidation behaviour of the Ni–W and CeO 2 interface: role of Pd inter–layer
094 Low fluorine YBCO MOD
094 Transport properties improvement in low fluorine YBCO MOD with artificial
pinning sites
095 Conceptual design of YBCO coil for the toroidal magnetic system of ISTTOK
tokamak
096 Inertial Fusion
097 Fusion & Industry
099 Quality Assurance
100 Publications and Events
100 Publications
100 Articles
106 Articles in course of publication
109 Contributions to conferences
116 Workshops and Seminars
116 Workshops
117 Seminars
117 Patents
118 Miscellaneous (*)
118 The Fleischman&Pons Effect Through the Materials Science Development
118 Pd–Based Membranes
119 AGILE and LOFT
120 Digital Phase Detector for the RF Interferometer for Mini Helicon Space Thruster
121 References
122 Organization Chart
124 Abbreviations and Acronyms
(*) Not in Association framework

progress report
2010
006
Preface
Euratom–ENEA
Association
2010annual
fusion
activities

008
progress report
2010
Anniversary
50 th Anniversary EURATOM–ENEA Association on Fusion
Spurred by Edoardo Amaldi and with the collaboration of Enrico Persico and Franco Rasetti, starting in 1957
Bruno Brunelli had gathered a small group of scientists of CNEN (today ENEA) and of Euratom who began
a research activity on plasma physics and on problems related to thermonuclear fusion at the Institute of
Physics of the University of Rome. In 1960 this activity was transferred to the CNEN Research Centre in
Frascati where a new building was erected for the new Laboratorio Gas Ionizzati. The contract of association
Euratom–CNEN was signed in the same year in attachment to the preceding French contract of association
Euratom–CEA. In this way the Italian Fusion Program was started.
After 50 years, the past and present leading actors of fusion research have gathered on 8 July 2010 at ENEA
Centre in Frascati to celebrate this important anniversary. Under–secretary of State – Ministry of Economic
Development, Stefano Saglia, ENEA Commissioner, Giovanni Lelli, and Euratom Director, Octavi Quintana
Trias, attended the ceremony. In their addresses they emphasized the continuously increasing role, both in
fusion physics and technology, played by Italy in the
European Fusion Program, thanks to strong
determination and capability to adapt to the new
challenges.
The volume describing the activities carried out by
EURATOM–ENEA, written in the occasion of the 50th
anniversary of the Association
Speeches were delivered by the Italian
representative in CCE–FU, Romano Toschi, the
EFDA leader Francesco Romanelli, the Chairman
of the Governing Board of Fusion for Energy,
Carlos Varandas, the Head of ITER Department
of Fusion for Energy, Maurizio Gasparotto, the
President of Consorzio RFX, Giorgio Rostagni, the
Director of Istituto di Fisica del Plasma – CNR,
Maurizio Lontano, and by the Director of ENEA
Fusion Technical Unit, Aldo Pizzuto. Looking over
the past 50 years, at the experiments that have
marked the more significant and outstanding results
obtained, they testified how fusion science has
achieved its success through the ability to evolve
from pure research activity confined in few
laboratories, into a complex system involving
physics, technology and engineering research, and
close collaboration also with university and
industrial worlds. Thanks to this capability, it has
been possible to face increasingly demanding
challenges, such as FT, FTU and RFX, and support

progress report
2010
009
the involved industry with the necessary
technological know – how. The wealth of
experience acquired has proved to be
extremely important in the development of
JET and later in ITER, and is now the
source of technological activities for the
fusion carried out in ENEA today.
In the recent years, as a new phase has
begun with the start of ITER construction,
the EURATOM – ENEA association has
launched a new, high level programme for
our Country to maintain its position in the
forefront, based on a strengthened
collaboration among the national partners,
ENEA, CNR, Consorzio RFX and many
Universities, and on a full integration with
the activities carried out in the other
European Countries. A closer relationship
with the national industry has been
established, with initiatives to promote its
participation in ITER construction, and
with provisions for participation in
procurements using technologies developed
in our laboratories.
The celebration of 50th anniversary EURATOM-ENEA Association
on Fusion at the ENEA Frascati centre on 8th July 2010
The future relies on a new experiment,
FAST, that can serve as a European and
global research infrastructure on which to
develop new knowledge of both physics and
technology and to train scientists and
technologists in view of ITER exploitation
and the design of the demonstration
reactor.
1960–2010, fifty years characterized by
passion and enthusiasm. The wish is that
many others with enthusiasm and passion
will follow with the awareness that fusion
energy is a goal too important for future
generations to miss.
Yvan Capuet for EURATOM and Giovanni Lelli Commissioner for
ENEA slicing the celebrating cake

010
progress report
2010
chapter 1
magneticconfinement
The alternation of extremely scientifically productive years with technically critical ones seems to continue on FTU.
Following the very productive year 2009, in 2010 the FTU time effectively used for the experimental activity has been
extremely limited. The spring campaign was devastated by a series of vacuum leaks and eventually ended by a failure on
the flywheel generator feeding the toroidal magnet. The autumn campaign was affected by several failures, mostly linked
to poloidal circuits, producing at the end only a few shots useful for experimental programmes.
As a consequence, during the 2010 year, more emphasis was given to completion of analyses of existing data and of new
realisation on diagnostic and heating systems.
However reduced in number, plasma operations were always benefitting of liquid lithium limiter, either used to limit the
plasma or to condition the walls through depositing lithium on plasma facing surfaces. In such discharges pellet were
successfully injected for the first time, obtaining extremely peaked density profiles at very high density values as high as
6×10 20 m –3 . Detailed linear microstability analysis was performed on lithized discharges, proposing a possible mechanism
for generating the density peaking. A detailed analysis has also been completed for the discharges were, thanks to edge
optimisation, it was demonstrated that lower hybrid (LH) waves can penetrate plasmas with ITER like density profile.
These results, relevant for future applications of LH waves have been published in Nature Communication and have
triggered a multi-machine joint experiment coordinated by International Tokamak Physical Activity (ITPA)– Integrated
Operation Scenario (IOS) group. More experiments, at higher power, are planned on FTU to quantify the LH current
drive (CD) in these conditions as well as to complete the systematic comparisons between experiments and theory of
electron fishbones in the frame of the general fishbone–like dispersion relation. Looking forward to a positive decision for
future application of LH to ITER, a complete revision of the system design for ITER has been also completed in
collaboration with CEA Association.
Experimental activities on electron cyclotron (EC) physics have been conducted at high priority on FTU. Runaway
electrons have been suppressed; applying EC waves in heating scheme during the flattop phase of the discharge, at an
electric field value a factor ∼2 larger than the predicted collisional threshold field, E R ∼0.09×n e (10 20 m –3 ). EC on FTU
has contributed to ITPA joint experiments on assisted breakdown and on MHD control in particular controlling sawteeth
period through the localization of EC waves (injected in heating and CD scheme) with respect to q=1 surface. These
experiments will greatly benefit from the installation of the new real time steerable antenna, that will inject 0.5 MW of
EC waves through two front mirrors real time steerable poloidally and toroidally. The launcher has been fully tested in
laboratory, achieving all design performances, and was ready to be installed on FTU in the 2010–2011 shutdown.
By end of 2010, also two new diagnostics were completed and fully tested in laboratory. A new electro–optic probe able
to simultaneously detect the electric spike and the ionization radiation generated by the impact of dust particles and a
fast camera, which, able of up to 500000 frames/s, will allow diagnosing fast plasma evolution from breakdown to
disruptions.
Plasma theory progressed along the traditional lines of Frascati theory team. Beta induced Alfven eigenmodes (BAE),
observed on FTU during the first part of islands growth, were found to agree with theoretical predictions for sufficiently
low magnetic island amplitudes, consistently with perturbative theory. Significant efforts have been devoted to analysing
nonlinear behaviours in burning plasmas of fusion interest, their complex dynamics and the issues that arise when
modelling these phenomena with increasingly more realistic physics and equilibrium descriptions. Many of these
theoretical activities significantly contributed to advancing the FAST conceptual design; the activities specifically carried
out within the framework of this project are summarized in the specific FAST section.
Being JET operations suspended for the installation of the new ITER like wall (ILW), ENEA scientists have focused their
activity in the analysis of the data gathered during last campaigns and finalized journal papers and contribution to
international conferences, such as EPS, SOFT, IAEA and others. They also actively participated in the elaboration of the
JET programme for 2011

magnetic confinement (cont’d.)
progress report
2010
011
1.1 FTU Facility
Summary of the machine operation
The first experimental campaign started at the end of March but it was immediately stopped due to a leak
in the vacuum chamber. After that a long series of vacuum problems prevented operations from being
restarted until mid–June. The experimental campaign went on until the end of June, when it was closed due
to problems to the motor flywheel generator (MFG1) chiller.
The restart of the machine, originally scheduled for the end of October, began mid November, but,
unfortunately, the 141 pulses done (51 of which for plant tests) turned out to be useless for the experimental
campaign, because of several hardware failures. The main problem encountered was a broken electronic card
of the Programmable High Speed Controller (PHSC) which controls the poloidal power supplies: since it
required long time for reparation, this led to the end of the experimental campaign.
Therefore, the statistics considered for 2010 are those of the first experimental campaign: 320 shots were
successfully completed, out of a total of 349 performed in 12.5 experimental days. The average number of
successful daily pulses was 25.60. Table 1.I reports the summary data. Figure 1.1 reports the source of
downtime in 2010: Power Supplies is the greatest cause of delay with 31% of the total.
Control and Data Acquisition System. The project of replacing ageing hardware on Frascati Tokamak
Upgrade (FTU) and improving the software infrastructure in the control and data acquisition area, is still
under way. It is mainly based on the principles of the open source software and commodity hardware.
The physical link between the lower and the medium level in the FTU
control system three–level standard model – historically relying on
RS232 protocol on fiber optics – was replaced with serial
communication on standard Ethernet by using a new device
which provides network access to industrial instruments through
a serial connection. The technology has also allowed dedicated
hardware to be replaced by using machine virtualization on a
single Linux host.
A high multiplexed channel density and few kHz lower hybrid
(LH) power data acquisition systems were developed in
Compact PCI standard. Moreover, the legacy Soft–X VME
data acquisition system CPU was renewed by developing at the
same time a new software interface based on open source
Universe II libraries, and a more user friendly management
system.
Figure 1.1 – Source of downtime in 2010.
Power supplies is the greatest cause of delay
with 31% of the total

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progress report
2010
Table 1.I – Summary of FTU operations in 2010
Jan. Feb. March April May June July Total
Total pulses (p) 11 0 0 22 0 283 33 349
Successful pulses (sp) 9 0 0 19 0 262 30 320
I(sp) 0.82 0.86 0.93 0.91 0.92
Potential experimental days 3.5 0.0 3.5 0.0 10.0 6.0 23.0
Real experimental days 0.5 0.0 1.0 0.0 10.0 1.0 12.5
I(ed) 0.14 0.29 1.00 0.17 0.54
Experimental minutes 134 0 287 0 4476 500 5397
Delay minutes 119 0 345 0 1735 155 2354
I(et) 0.53 0.45 0.72 0.76 0.70
A(sp/d) 18.00 19.00 26.20 30.00 25.60
A(p/d) 22.00 22.00 28.30 33.00 27.92
Delay per system (minutes)
Jan. Feb. MarchApril May June July Total %
Machine 94 0 0 12 0 339 42 487 20.7
Power supplies 10 0 0 83 0 599 33 725 30.8
Radiofrequency 0 0 0 0 0 0 0 0 0.0
Control system 0 0 0 0 0 108 0 108 4.6
DAS 0 0 0 0 0 44 0 44 1.9
Feedback 0 0 0 0 0 84 0 84 3.6
Network 0 0 0 0 0 0 0 0 0.0
Diagnostic systems 0 0 0 39 0 115 30 184 7.8
Analysis 12 0 0 0 0 283 50 345 14.7
Others 3 0 0 211 0 163 0 377 16.0
TOTAL 119 0 0 345 0 1735 155 2354 100
An upgrade of the present FTU feedback control system was started, also envisaging a possible reutilization
in the proposed FAST experiment. A pre–existent framework called multithreaded application real–time
executor (MARTe), already successfully used in other European tokamak devices, was adopted. MARTe has
been integrated in the current control structure and pulse programming interface, thus providing a common
environment with other MARTe systems already running in FTU, such as the real time version of the plasma
equilibrium reconstruction RT–ODIN and the satellite station that minimizes the LH reflected power.
50
40
30
20
10
0
Figure 1.2 – 2010 ITM–TF users grouped by country
F D I UK FINCH PL P S EU A CZ GR NL B DK RO CY E H IRLRUSLO US NaN
ITM Task Force Gateway. The
Gateway operations have continued to
provide the full scale support to
Integrated Tokamak Modelling–Task
Force (ITM–TF) users, thus allowing
them to develop ITM software and
data access facilities. It has been
carried out by means of the Gateway
trouble ticket system and direct
support by email.
A number of 223 ITM–TF users have
got an account onto the Gateway in
2010 as depicted in figure 1.2. The
development & testing activity of the
ITM–TF users is reported as CPUs

magnetic confinement (cont’d.)
progress report
2010
013
% CPU usage
100
EFDA-ITM batch cluster yearly usage
0
Jan 01 Mar 01 May 01 Jul 01 Sept 01 Nov 01
Total CPU load % Current: 2% Average: 9% Maximal: 100%
Figure 1.3 – CPUs yearly usage of the
Gateway batch cluster
yearly usage of the Gateway batch cluster in 2010 and shown in figure 1.3. The upgrade activity has been
carried out involving the base software (operating system, parallel filesystem, distributed filesystem) and
scientific tools as well.
The Storage Data Area used in 2010 is as follow: 4.3 TB for users and software repositories stored under the
distributed filesystem, 14 TB for simulation data stored under the parallel filesystem.
The RAM resources of Front–End nodes have been updated to 32 GB for each one, to take into account the
heavy CPU usage due to the interactive sessions during Code Camp meeting.
Several EU Code Camp 2010 meetings were supported by using hardware/software resources of the Gateway.
The Inter–operability between ITM Gateway and High Performance Computing For Fusion (HPC–FF) has
been implemented fully. It allows both HPC facilities to be used in a common operating environment for
sharing codes, data exchange and user access, as well as joint tools and libraries to facilitate the creation of
community fusion codes.
1.2 FTU Experimental Activity
Lower hybrid
Lower hybrid current drive experiments in plasma regimes of interest for ITER. The need for non–ohmic
steady–state current drive is a fundamental weakness of the tokamak as a magnetic fusion reactor. Low
frequency microwaves, in the range of a few GHz, have been successfully used to drive current in tokamak
core plasmas at relatively low densities [1.1–1.3], but extrapolation of these results to the higher densities of
ITER and subsequent reactor grade plasmas has proved to be problematic, because the coupled LH power
stubbornly localizes in the plasma edge region [1.4,1.5].
A thermonuclear plant needs to maximise the bootstrap current for saving current driven by external power
sources. For this reason ITER will need to operate with almost flat radial profiles such that high pressure
gradient occur in the external region of the plasma column with high densities even at the edge (at normalized
minor radius r/r LCMS
∼0.9: n e_0.9
≥0.7×10 20 m –3 , and in the centre: n e0
≥0.8×10 20 m –3 , where r LCMS
is the
plasma minor radius at the last closed magnetic field surface (LCMS)). The operation with relatively high
plasma density in the region of the periphery is thus an important constraint to be considered for designing a
tool for driving actively non–inductive current in the plasma, which is essential for obtaining the necessary
requirements of stability and confinement.
Theoretical and experimental works have been extensively carried out on FTU, which have evidenced the role
of parametric instabilities (PI) in inhibiting penetration of the lower hybrid wave to the plasma core when
operating at relatively high plasma densities [1.4,1.5]. These works have produced the guidelines used in the
FTU experiments, based on theoretical predictions that the PI–produced spectral broadening would diminish
under operations with higher electron temperatures at the periphery (T e_outer
) of the plasma column [1.3]. As
a result of operating in such a regime, the spectral broadening measured by radio–frequency (rf) probe has
been indeed observed to diminish. It has also been observed that the LH power penetration into a high density
core took place as expected.
The following options, available on the FTU device, have been exploited for producing different plasma edge
temperatures in high–density plasmas, and different plasma targets for LH power coupling: i) Toroidal or
poloidal limiter operations. With toroidal limiter operations, a stronger plasma–wall interaction occurs due to
the larger plasma–wall contact area. In these conditions lower recycling of particles from the vessel walls and

014
progress report
2010
relatively low temperature at the plasma periphery generally occur. Conversely, in operations with the plasma
displaced towards the poloidal limiter, the D–alpha emission level is about ten times smaller, and slightly higher
electron temperatures are observed at the plasma edge than in similar experiments using the toroidal limiter.
ii) Vacuum vessel covered with boron or lithium. With lithium, sprayed on the walls from the limiter, where it
is present in liquid form (because of the liquid lithium limiter (LLL) available in FTU), a level of recycling
significantly lower than with a boron–coated vessel is produced. iii) Plasma fuelling assisted by pellet injection.
FTU has a pneumatic single stage multibarrel pellet injector, capable of firing up to eight pellets per plasma
discharge with a typical velocity of 1.3 km/s and a mass of the order of 10 20 deuterium atoms. As critical
aspects of the pellet technique, in order to achieve a deep plasma core fuelling for producing high electron
temperature at the plasma periphery, proper velocity and mass of pellets should be set. In addition, the LH
power switch-on time point should occur with some delay (of the order of 10 ms) as compared to the pellet
injection, in order to prevent the ablation of pellet by the coupled rf power. iv) In order to further reduce the
recycling, the technique of extra–gas fuelling in the early discharge phase has been used. In the standard gas
fuelling technique, the requested density value at the start of the lower hybrid current drive (LHCD) pulse is
set by the plasma density feedback control, which generally produces a continuous gas injection during the
whole plasma discharge. Relatively high levels of recycling and low temperatures at the plasma periphery were
obtained on FTU with this operation. In the technique of extra-gas fuelling in the early discharge phase a
large amount of gas should be injected in the early discharge phase (but already during the plasma current
flat–top), which transiently produces a plasma density slightly higher than the value required for the LHCD
pulse. The density then falls to the required value after a delay during which a pause in the gas injection is
programmed, so that low recycling occurs. By firing pellets during the gas injection pause, plasmas with very
high–density and low recycling should be produced.
The experiments results shown in ref. [1.4] have produced conditions for LH current drive that are transient,
and the extrapolation of the method obtained to stead–state regime would appear problematic. Further data
of the FTU experiment obtained in 2009 support the conclusion that a relatively high electron temperature at
the periphery of the plasma column, produced with tools useful for maintaining the edge plasma conditions
in steady–state, has the effect of reducing the spectral broadening of the signal as measured by rf probe and
enhancing the hard–x ray emission from LH–generated fast electrons, i.e., the LHCD occurring in the core.
Plasma discharges have been produced with similar density profiles at medium values of density
(n e_av
∼0.8×10 20 m –3 ), keeping the same radial profile, but with different electron temperature at the plasma
periphery. The behaviour of the spectral broadening measured by rf probe during FTU experiments
performed in similar operating conditions but different temperatures at the edge is shown in the figure 1.4.
The new important information obtained by the FTU results is that higher electron temperatures at the
plasma periphery favour the occurrence of LHCD effect in the core also in case without pellet injection and
during the whole phase of LH power coupling. This circumstance support the hypothesis that LHCD effects
can be produced and sustained also in steady–state at high densities provided that suitable conditions of high
temperatures at the edge should be maintained by means of an active tool of local plasma heating, provided
by electron cyclotron resonant heating (ECRH). Available
data from simulations show that the LHCD effect, already
=0.57×10 20 m –3
14 obtained during the transient phase of pellet fuelling of
1×10 4
=0.8×10 20 m –3
FTU plasmas, should be further sustained in steady-state
condition by utilising ECRH power.
LH–hard X (counts/s)
1.2×10 4
8000
4000
0
0.2 0.3 0.4 0.5
T e–edge (@ r/a=0.7) KeV
10
6
Δf(@–35 dB MHz)
Figure 1.4 – Hard–x ray level (curve in red) produced
by LH–accelerated electrons and line frequency
spectral broadening (curve in blue) plotted versus
the electron temperature at normalised minor radius
r/a=0.7, for FTU discharges with same plasma
density profiles and coupled LH power (0.35 MW)
Revision of the LHCD system for ITER. In the frame of
the European Fusion Development Agreement (EFDA)
task WP08–HCD–03–01 "LH4IT" (WP VI–2–1): EU
Contribution to the ITER LHCD Development Plan, the
conceptual design of the LHCD system for ITER,
elaborated in the years 2001–2002, has been revised by
taking into account all the physical and technological
progresses and innovations.
Experimental results obtained in the last few years, such as
the successful tests of a passive–active multijunction (PAM)
launcher on FTU and Tore Supra, have definitely indicated
this concept as a possible candidate for the LHCD launcher
for ITER. The successful development of a prototype

magnetic confinement (cont’d.)
progress report
2010
015
klystron at 5 GHz with a target rf power of 500 kW continuous wave (CW) has definitely set this power as the
upper limit to the present high vacuum electron tubes technology at this frequency. From the physics point of
view, the ITER operational scenarios have been well defined, while the latest experiments have pointed out the
effectiveness of the LHCD waves at high plasma density.
Following the recommendations of the 4th meeting (19–22 May 2008) of the ITER STAC, the assigned
objectives of the task were "to perform the conceptual design and to initiate the urgent R&D activities in order
to bring the ITER LHCD system up to the point at which such activities could be transferred to the Fusion
For Energy (F4E)".
The "LH4IT" task has been performed in close collaboration between ENEA – Frascati and Commissariat à
l’Energie Atomique (CEA) – Cadarache as well as with ITER Organization (IO) and several other ITER
partners.
A 5 GHz prototype klystron (fig. 1.5) developed by Toshiba for the LHCD system of the Korean tokamak
KSTAR has demonstrated a CW output power of 350 kW, a 10s pulsed power of 455 kW, and a peak power
of 510 kW for 0.5 s.
These promising results confirm the possibility of realizing a 500 kW CW klystron with an operational
frequency of 5 GHz in a relatively short time. On the other hand these results also demonstrate the 500 kW
as the upper power limit for klystrons in this range of frequency. The required 20 MW of rf power coupled
to the plasma can be therefore obtained by 48 500 kW klystrons. Each klystron feeds a PAM module (fig. 1.6)
made of six toroidal rows and four toroidal columns of elemental alternated
active/passive waveguides. To launch an optimum rf spectrum with N 0
=1.9 with
a phase shift of 270° between active waveguides, the active waveguide width is set
to 10 mm, with a passive waveguide width of 8mm and separation wall thickness
of 3 mm at the launcher mouth.
The rf power of each klystron is carried by moderately oversized circular
waveguides (the C 16 standard with inner radius R=67.05 mm) which transmit the
TE 01
circular mode. The corresponding evaluated transmission losses are about
0.22 dB/100 m.
Suitable mode filters (fig. 1.7) have been taken into consideration and studied in
order to suppress potentially detrimental spurious modes generated by the four 45
deg and one 100 deg bends included in the main transmission lines (MTLs). These
filters are based on the attenuation introduced by the periodic corrugation of the
waveguide wall, enhanced by the presence of absorbing material silicon carbide
Active
waveguide
1st BJ
Figure 1.6 – Schematic drawing of a PAM module
Figure 1.5 – The
500 kW prototype
klystron
2nd BJs
d
a
400 mm
Base circular waveguide
P
Corrugations
SiC
100 mm
Passive
waveguide
Figure 1.7 – Longitudinal cut of a corrugated mode
filter in circular waveguide

016
progress report
2010
Ceramic
window
Input WR284
(72.14 x 34.04 mm)
Reference plane
position
Short circuit
WR 284
Figure 1.8 – Pillbox ceramic window
Output
(58 x 34.04 mm)
Figure 1.9 – The 3dB hybrid junction
Figura 1.10 – Electric field in the
TE 10 to TE 30 mode converter
(SiC), on the TE mn
modes (m≠0). The
propagation of TE 0n
modes with n>1 is
avoided by an accurate dimensioning of the
base waveguide diameter.
In the following figures the most critical microwave components of the transmission system are shown, as, e.g.,
the 500 kW, CW pillbox ceramic window (fig. 1.8) separating the pressurized MTL from the vacuum region
of the launcher; the 3 dB hybrid junction (fig. 1.9) to initially split by two the rf power generated by each single
klystron; the TE 10
to TE 30
mode converter (fig. 1.10), which splits into three each output branch of the
previous hybrid junction in order to suitably feed the upper and lower parts of a single PAM module. The final
power distribution among the four active waveguides of each row of a PAM module is obtained by classical
cascaded bijunctions.
Alfvén Eigenmodes: modeling, experimental results and future plans
Electron–fishbones studies. Electron–fishbones are internal kink mode instabilities induced by supra–thermal
electrons. The knowledge of the supra–thermal electron dynamics, in presence of fishbone fluctuations, is very
important in burning plasmas. Indeed, energetic particles in the MeV range are present in ignited plasmas, as
either fusion products (alpha particles) or supra–thermal tails induced by additional heating systems. In
particular, good confinement and slowing down of alpha particles (electron collisional heating) are very
important issues. Perpendicular transport of magnetically trapped supra–thermal electrons is similar to that
of fusion alphas, since both have small orbit as compared to the plasma minor radius. Thus, the investigation
of such dynamics may shed new light on physics issues related to alpha particles. The stability analysis of
e–fishbones can be carried out within the framework of the so called general fishbone–like dispersion relation,
which can be thought of as a kinetic energy principle. Previous experiments on e–fishbone in FTU have been
conducted with lower hybrid radio frequency heating and q min
>>1. They suggest that the level of the rf power
determines the energy content of the supra–thermal electron tail and is crucial in controlling the transition
from nearly steady state non–linear oscillations to bursting regime; moreover, experimental observations
confirm the theoretical conjecture that the e–fishbone in the bursting regime is a continuum resonant mode
[1.6]. FTU experiments, as well as experiments conducted at Tore–Supra [1.7], underline the necessity of
systematic comparisons between experiments and theory by means of the general fishbone–like dispersion
relation and appropriate simulations codes. MARS is a resistive–magnetohydrodynamic (MHD) stability code
that can be used to this purpose to calculate the part of the plasma response entering the fishbone–like
dispersion relation, i.e. the MHD potential energy contribution. With this information, systematic stability
studies are underway, which can give insights about the FTU plasma conditions where the kinetic response due
to supra–thermal electron tails is most likely to yield the fishbone instability. In particular, q–profile sensitivity
studies will be carried out as well, for they will clarify the role of current profile in the e –fishbone dynamics.

magnetic confinement (cont’d.)
progress report
2010
017
Experimental observation of beta–induced Alfvén Eigenmodes.
FTU activities in 2010 were focused on the observation of long
(>200 ms) beta–induced Alfvén eigenmodes (BAEs), which are
high frequency oscillations (30–70 kHz) in tokamak plasmas,
identified in the spectrum of the Alfvénic modes as discrete
eigenmodes located in the low frequency gap of the Alfvén
continuum produced by the geodesic curvature and finite beta
effect. BAEs have been first identified as waves excited by
circulating fast ions, however they were also observed in ohmic
plasmas with the simultaneous presence of large magnetic islands
[1.8]. BAE oscillations from 30 to 50 KHz have been observed in
FTU plasmas with the simultaneous occurrence of tearing modes
(characterized by frequencies from 1 to 5 KHz). BAEs have
intensities two order of magnitude less than tearing modes, and
are characterized by two main lines, which merge in a single line
when the island oscillation frequency becomes very low. Mode
analysis for tearing mode shows that poloidal and toroidal mode
numbers are (m, n)=(–2,–1). On the other side, the higher BAE
frequency is characterized by (m, n)=(–2,–1), i.e., propagates in
the same direction as the island, while the lower BAE frequency
has (m, n)=(2,1), i.e., propagates in the opposite direction. The
difference in frequency of the two BAE lines is exactly twice the
fundamental frequency of the tearing mode, and the frequency
difference can be explained by the doppler shift due to island
rotation.
In the following figure 1.11 the comparison between the
theoretical prediction [1.9] of BAE frequency as a function of
magnetic island amplitude δB r
/B p
and the experimental
observations is shown. During the first part of the island growth,
BAE frequency increases roughly linearly, in good agreement
with theoretical predictions; on the other hand, a discrepancy is
found for δB r
/B p
>2 ×10 –3 where BAE frequency remains rather
constant. Therefore, we can state that, as expected, the
perturbative theory used gives consistent results only for
sufficiently low magnetic island amplitudes.
BAE frequency (kHZ)
55
45
Observed frequency (N 23184)
Observed frequency (N 26644)
Observed frequency (N 25877)
Predicted frequency (N 23184)
Predicted frequency (N 26644)
Predicted frequency (N 25877)
35
0.5 1.5 2.5 3.5 4.5 ×10 -3
δBr(rs)/Bp(rs)
Figure 1.11 – BAE frequency as a function of
magnetic island amplitude δB r /B p (curves
are the theoretical predictions and markers
are the experimental points)
(V)
T e (keV)
n/s
Counts
Energy (MeV)
1.5
0.5
0.5
0
4
2
0
10 11
10 10
10 3
10 2
8
4
Experimental V I
Collisional threshold
NE213
BF 3
a)
b)
c)
Total γ - counts
#22472
0
0 0.4 0.8 1.2
Time (s)
d)
e)
Electron cyclotron
Runaway electrons in FTU. Runaway electrons constitute a
well–known phenomenon in tokamaks: due to the decrease in the
Coulomb collision frequency with energy, electrons with energy
larger than some critical value are continuously accelerated by
the toroidal electric field. Calculations including relativistic effects
indicate that below a critical or threshold electric field,
E R
=(n e
e 3 lnΛ)/(4πε 0
m e
c 2 ) absolutely no runaway electrons are
produced [1.10]. Experiments carried out in FTU show that
Figure 1.12 – Runaway suppression in
discharge 22472: a) loop voltage; b) electron
temperature (the ECRH time window is
represented by the shaded area); c)
comparison between neutrons (BF 3 ) and
gamma & neutron detectors (NE213): when
the two signals coincide runaways are
suppressed; d) total counts and e) maximum
measured energy from γ–ray NaI
spectrometer
synchrotron radiation losses should also be taken into account, since they lead to an increase in the critical
electric field. These results are important when making predictions on runaway generation and mitigation
during disruptions in next–step devices such as ITER.
Suppression of runaway electrons has been achieved in FTU (fig. 1.12) by application of ECRH) during the
flat–top phase of the discharge, at electric field values substantially larger (by a factor ∼2) than the predicted
collisional threshold field, E R
∼0.09 n e
(10 20 m –3 ). The parameters used in the experiments are the following:
I p
=0.3–0.5 MA, B t
=4–6 T, central line averaged density n e
=4–8×10 19 m –3 and electron cyclotron (140 GHz)
input power P ECRH
= 0.3–1 MW. The interpretation of these observations, based on calculations including the
electron synchrotron (radiation) term (fig 1.13) leads to the conclusion that the radiation losses are responsible

018
progress report
2010
dW/dt (Arb. units)
6
5
4
3
2
Collision stopping power
Radiative stopping power
Total stopping power
eE R
rad νII
e E II ν II
1
W c1
W c2
0
0 1 2 3 4 5
Energy (MeV)
Figure 1.13 – Collision, radiation, and total stopping
powers for an electron. For an energy gain below
(eE ν ) no runaways can be found. Electrons with
energies W c1

magnetic confinement (cont’d.)
progress report
2010
019
(see fig.1.15, blue symbols). Moreover we observe that, at high ν eff
values, a well detectable delay between
density and temperature rise (re)appears.
The interaction between ITER like plasmas and the EC heating have to be investigated also under the profile
of the effect of the heating system on the particle–energy confinement time with dedicated experiments
extended to H–mode regimes and in conjunction with different heating systems.
Sawteeth (de)stabilization by ECH and ECCD in FTU. In FTU the destabilization of sawteeth (ST) has been
investigated using ECH and electron cyclotron current drive (ECCD) as a powerful tool for the shortening of
ST period τ saw
. This is a critical issue for the plasma confinement in fusion devices. In fact, long–period
sawteeth give rise to large crashes that can trigger seed islands, thus destabilizing neoclassical tearing modes
(NTM) that, on turn, degrade the plasma performances at high β. This programme was included in the work
plan of the International Tokamak Physical Activity (ITPA) MHD working group on ST control for the
empirical scaling of power requirement for avoiding the onset of NTMs. The sawteeth crash induced by ECH
and co–ECCD was performed by using 500 ms of
repetitive pulses (10 ms EC on, 40 ms EC off) from two
12
8 ×10 5
gyrotrons, up 0.8 MW total power.
As the ST crash happens when the local magnetic
shear s=rq’/q exceeds a critical value,
ECH/co–ECCD, increasing the current density inside
the inversion radius (q=1), can shorten the ST period
by sharpening the q profile and enhancing the local
shear. ST destabilization has been performed in
plasmas at 500 kA with toroidal field ramping from 5.1
to 5.9 T in order to scan the EC absorption radius r abs
from the plasma centre to well outside the q=1 radius
and find the magnetic field value at which the sawteeth
crash is induced by the EC trigger. Preliminary results
are shown in figure 1.16, where the time delay
t EC,on
– t ST,crash
is plotted during the EC modulation.
For EC absorption inside the q=1 radius the ST
periods are reduced from 6 ms (ohmic value) to about
2–4 ms, while for EC absorption outside the q=1
(t>0.8 s) the EC power plays an opposite effect, thus
lengthening τ saw
up to 10 ms.
These results show that 0.8 MW of EC power can
destabilize the ST by shortening their periods when EC
absorption occurs inside the q=1 radius, where the
ECH/co–ECCD increase the magnetic shear thus
inducing the sawteeth crash. The next experiments
should be devoted to investigating the EC power
threshold for destabilization.
Tests of the new ECRH launcher for FTU. The new
EC antenna for real time control experiment on FTU
tokamak was tested at the Istituto di Fisica del Plasma
(IFP)–Consiglio Nazionale Ricerche (CNR) – Milano
at the beginning of 2010. A series of tests concerning
specific design issues and the launcher assembly were
performed in order to validate the technical solutions
adopted and to characterize the antenna in its final
configuration, with low power measurements. A test
facility was installed for the launcher tests (fig. 1.17).
The mechanical tests involved the cardanic joints of
the driving system, based on rotational movements,
which are transferred, via vacuum–compatible
rabs/rinv τsaw(ms) B t (T)
10
8
6
6 ×10 5
4 ×10 5
4
2 ×10 5
2
q=1
0
0
0.5 0.6 0.7 0.8 0.9 1
Time (s)
ECRH
Figure 1.16 – Combined effects of ECH/co–ECCD on τ saw
(green circles) in the discharge with ramped B t (black
solid line). For B t > 5.4 T the absorption is outside q=1
and the ST periods are greater than the ohmic ones
(dashed line). The ratio between the EC power
absorption minor radius rabs and the inversion radius is
also marked (red line) during the 500 ms EC power
modulation (10 ms on/40 ms off, blue line)
Figure 1.17 – Experimental setup used to test the plug-in
launcher at IFP mm-wave laboratory. The two launching
lines (symmetric as to the equatorial plane) are visible
and the low power mm–wave source can be seen on the
top left

020
progress report
2010
Power
(Linear amplitude)
Power
(Linear amplitude)
1.0
0.8
0.4
0
1.0
0.8
0.4
0
W tor =18.0 mm d in =800 mm
W tor =20.8 mm d in =700 mm
W tor =24.5 mm d in =600 mm
a)
-20 0 20 60
Probe position (mm)
ν pol =19.2 mm d in =800 mm
ν pol =22.1 mm d in =700 mm
ν pol =27.6 mm d in =600 mm
b)
0 20 40 60 80
Probe position (mm)
Figure 1.18 – Output beam profiles in the toroidal a) and poloidal b) direction,
obtained with the upper antenna (W is the beam radius in the toroidal and poloidal
direction, respectively; d in is the input distance between the input beam waist and
the focusing mirror). Dots correspond to the acquired data, curves to the Gaussian
fit of the beam profile
Vertical probe position (mm)
80
60
40
20
0
Power (dB) - angles: α=0°, β=0°
Speed ref.
motor 2
Speed ref.
motor 1
MIRROR POSITION CONTROLLER
(with real–time plasma ray tracing)
Plasma
feedback
-20 0 20 20 60
Horizontal probe position (mm)
Figure 1.19 – Beam amplitude pattern (dB) obtained from
the upper antenna set in the reference configuration:
contour levels are shown, where the –8.7 dB level
corresponds to the definition of beam radius w
PROTECTION SYSTEMS
Alarm
MOTOR 1 + DRIVE
(speed and torque controller)
MOTOR 2 + DRIVE
(speed and torque controller)
Position
motor 1
Position
motor 2
Figure 1.20 – Scheme of mirror control and
protection systems
coupling rods, to the external motors, put at the rear of
the port. After these tests, a revision of the joints design
was necessary. The new solution ensured better results in
terms of the overall performances and reliability in the
mechanical transmission of the movement, from the
motor to the steering mirror unit.
Low power tests were carried out after an optical
alignment of the antenna, performed by using a laser
mounted outside the port flange with beam aligned with
the microwave beam axis and markers used to check the laser spot position on each mirror of the beam line.
The tests have been made by using a 140 GHz horn–mirror antenna to launch a Gaussian beam matching a
field close to the one measured at the aperture of the FTU transmission line. A vector network analyzer (VNA)
was used to perform the tests, with a receiving antenna made with a properly shielded truncated waveguide.
The upper and lower antennas output beams were measured. The beam zooming effect was verified by
measuring the launched beam dimension (fig. 1.18) for different positions of the sliding mounting, and
comparing them with the expected values. Also 2–D beam pattern measurements and 1–D scans were
acquired, to evaluate the diffraction effects (fig. 1.19).
The functionality of the brushless motors chosen for the last steering mirror movement, was tested in presence
of an external magnetic field (B MAX
=0.1 T) estimated at the location foreseen for the motors in the FTU hall,
in real conditions of typical plasma operations. A 3–D model provides maps of the local value of the stray
field surrounding FTU during the shot (fig. 1.20). Such a model gave a maximum of ≈0.058 T for the stray
magnetic field expected at the motors location during typical plasma discharges. A permanent magnetic field
source, able to produce 0.1 T as a maximum field in correspondence of the poles and 0.05 T in the air gap,
was used for dynamical tests: they confirmed that such an external stray magnetic field does not affect the
correct operations of either the brushless motors or the resolver system.
The control and protection systems (shown in fig. 1.20 for one mirror) were tested in order to identify the
parameters of the drive and the mechanical system. Several tests have been repeated with this open-loop
configuration, with and without any mechanical load. In particular, the tests aimed to acquire the system step
response, so the speed reference input was a step with finite length; different step amplitudes have been used
to test the linearity of the system (fig. 1.21). The responses obtained showed that the non–linearity is not

magnetic confinement (cont’d.)
progress report
2010
021
relevant. Simulation results, together with the speed controller
response verified experimentally, seem to confirm that the
target specifications can be achieved.
Liquid lithium experiments
Pellet injection in lithized wall discharges. High density
discharges were obtained through the first successful injections
of D 2
pellets in presence of a significant wall lithization.
Density profiles, for transport analysis, were obtained from the
inversion of the FTU CO 2
scanning interferometer [1.12],
which is a very powerful tool for studying fast evolution of
density profiles and very strong gradients. Density profiles were
taken every 62.5 μs with a spatial resolution of 1 cm and a line
integrated density resolution Δn¯e/n¯e∼2%. After a prompt
increase in central density, typical of pellet fuelled discharges, a
further peaking is observed following the first pellet, which
suggests the existence of an inward pinch, as seen also in gas
fuelled discharges (figs. 1.22 and 1.23). A fully interpretative
transport analysis of the density profile decay (1st pellet) has
been performed in the region where there are no particle
sources upon completion of pellet ablation (r/a

022
progress report
2010
0.5
#32552, 0.703-0.765 s, r=0.09 m
a)
#32552, 0.703-0.765 s b)
3
D
V
0.3
0.1
2
Γ/n(m/s)
0.1
D(m2/s)
1
V(m/s)
-0.1
0
0
-0.3
0
2 4 6 8 10
d log(n)/dr (1/m)
0
0.05 0.1 0.15
Minor radius (m)
Figure 1.24 – Experimental particle flux at r=9 cm in the post pellet phase derived from density profiles evolution a). Diffusion
coefficient and pinch velocity versus minor radius b)
γR ref /c s,D
40
20
0
t=0.3 s,with Li
t=0.3 s,no Li
t=0.8 s,with Li
t=0.8 s,no Li
#30582 r/a=0.6
a)
γRref/cs,D
-0.5
0
20 40
0 0.5 1 1.5 2
k θ ρ D
Figure 1.25 – Growth rates of FTU #30582 at t=0.3 s (crosses) and t=0.8 s (circles). a) ETG modes, b) ITG–TEM modes up to
k θ ρ D =2.0 (here k θ is the wave vector in the θ direction, ρ D is the Larmor radius of deuterium ions). Results with the same
parameters but without Li impurity are also shown (blue dashed). A concentration of n Li /n e =15% is found to stabilize the ETG
mode, ((a) crosses) and TEM modes ((b) crosses)
2
1
0
t=0.3 s,with Li
t=0.3 s,no Li
t=0.8 s,with Li
t=0.8 s,no Li
k θ ρ D
#30582 r/a=0.6
a)
Linear microstability analysis of lithium doped ohmic plasmas. Improved energy and particle confinement in
the presence of low–Z impurities has been observed in many tokamaks under various experimental conditions.
In particular, peaked electron–density profiles have been obtained in FTU ohmic plasmas where high
concentrations of lithium have been detected following the installation of the LLL. A gyrokinetic study on the
effects of lithium and other low–Z impurities on the linear stability of deuterium and electron temperature
driven modes and their associated fluxes has been carried out for plasma parameters such as those found in
the core of LLL–FTU plasmas. Simulations (fig. 1.25) show that a lithium concentration in excess of
n Li
/n e
=15%, as estimated in the initial phase of a reference FTU discharge, is found to have a strong
stabilizing effect on the trapped electron mode (TEM) and high frequency electron temperature gradient
(ETG) modes. A significant stabilization of the electron driven modes can still be observed when the lithium
concentration is reduced to 3%. In the presence of a significant impurity concentration (n Li
/n e
=3%–15%) the
long wavelength ion temperature gradient (ITG) modes drive an inward electron and deuterium flux and
outward lithium flux.
The microstability analysis carried out indicates (fig. 1.26) that, as far as the plasma parameters considered are
concerned, the presence of a lithium concentration of n Li
/n e
=3%–15% in the initial phase of the discharge
is enough to trigger an ITG–TEM driven inward flux of deuterium ions and electrons and an outward flux of
the impurity ions. This dynamics eventually leads to the observed peaked electron density profile and clean
plasma. The different sign of the fluxes is due to the phase difference between δn e
, δn i
and δφ allowed by the
presence of the third plasma species (lithium) in the quasi–neutrality condition.

magnetic confinement (cont’d.)
progress report
2010
023
×10 -3
4
#30582 r/a=0.6
a)
×10 -3
12
#30582 r/a= 0.6
b)
ΓspRref/(ρ 2 nsp vth,ref)
*
0
−4
−8
ref, D
ref, e −
ref, Li
no Li, D
no Li, e −
0 1 2 3 4 5
k θ
ρ D
ΓspRref/(ρ 2 nsp vth,ref)
*
8
4
0
D
e −
Li
-4
0 1 2 3 4 5
k θ ρ D
Figure 1.26 – Linear E × B particle flux spectrum of the species (red: deuterium, blue: electron, green: lithium) at a) t=0.3 s,
(n Li /n e =15%) and b) t=0.8 s, (n Li /n e =1%) at r/a=0.6. Results with the same physical parameters but without the presence on
lithium impurities are also shown (dashed). The numerical values do not correspond to the saturated flux levels
Diagnostic developments
Electronics equipment for the FTU
bolometry. The bolometric diagnostic
installed in FTU measures the total
radiation losses and the neutral emission
from the plasma, from visible ultraviolet
(VUV) to soft x–rays range. The
bolometer head, developed by
Max–Planck–Institut für Plasmaphysik
(IPP) – Garching, is constituted by two
absorbers, one exposed to the radiation
and the other one ‘blind’, inserted into a
Wheastone bridge: the rising of
temperature, due the incident radiation,
causes a variation of the absorber
resistance and, consequently, an
unbalance of the bridge.
Figure 1.27 – New layout of the system, which is extremely more
compact than the original one and has increased functionality
A new activity started in 2010 to
completely revise, design and develop the
new electronics measurement system of the FTU bolmetry, which is foreseen to be completed in 2011 year
(fig.1.27).
The system main characteristics are:
• Reduced analog hardware. Only amplification and offset removal with high precision components
• 16 bit over sampled (250 KHz) digitalization. Each output sample assuming 1 KHz depends on 250 samples
giving a noise reduction of ≈16 in the complete analog/acquisition process.
• References acquisitions. Further the bridge signal also the bridge reference is acquired.
• Vector voltmeter algorithm. Fully digital and numeric algorithms, the vectorial components of the signal are
evaluated. This allows the complex offset removal of the bridge avoiding the manual hardware
compensation of the cable lenght.
• Digital synthesis of bridge excitation frequencies. The excitation frequencies are digitally synthesized with 16 bit
D/A at 3 MHz rate.
• 4 contemporary excitation frequency in each bolometer head to remove cross talk in cables, bridges and electronics
• Compactness and network integrated acquisition system any additional acquisition system is necessary
• Full remote control. All the bolometer functions can be remotely operated and all the parameter can be
remotely visualized.

024
progress report
2010
New diagnostics for soft x–ray imaging and tomography. Magnetic fusion plasmas (MFP) are extended sources
of x–rays and these emissions could reveal a lot of information about the processes occurring inside the
plasmas.
The aim of this project is the development of a 2–D detector with independent energy discrimination
capability for each pixel. Plasma diagnostics based on soft x–ray (SXR) tomography and/or imaging for
magnetic confinement fusion plasmas could be greatly enhanced if different energy bands (which are
representative of different plasma zones and their impurity content) could be selected dynamically. A gas
detector with 2–D pixel read–out is being proposed for such a diagnostics.
Moreover, the constraints posed by toroidal devices (highly radiative background, extremely high
radiofrequency powers, high magnetic fields, optical limitations and so on) are very severe and strongly limit
the possibility to install x–ray detectors directly into, or close to the machine. Therefore, it becomes mandatory,
in particular for future burning plasma experiments, to study the possibility of transporting the SXR radiation
far from the machine. Polycapillary lenses appear promising for these purposes and suitable to be used for
x–ray imaging and tomography in MFP. All these activities have been carried out in collaboration with the
National Institute of Nuclear Physics (INFN) – Frascati National Laboratory (LNF) and CEA laboratory of
Cadarache, under the auspices of EFDA (Work Program 2010 (WP2010)).
1) Realization and characterization of a triple GEM gas detector. A triple gas electron multiplier (GEM) detector
has been designed and built by the LNF–INFN laboratory of Frascati and subsequently installed in the SXR
laboratory of CEA for this preliminary characterization [1.13]. It is based on a gas detector with triple GEM
as amplifying structure and with a two–dimensional read–out, coupled to the integrated front–end electronics.
The energy discrimination in bands, at the level of individual pixel, has been studied. The front-end
electronics of the GEM detector, working in photon counting mode with a selectable threshold for pulse
discrimination, is optimized for high rates.
The energy resolution of the detector has been accurately studied in laboratory with continuous SXR spectra
produced by an electronic tube (continuous spectra with a Moxtek 40 kV Bullet) and line emissions produced
by fluorescence (K, Fe, Mo), in the range 3–17 keV (fig. 1.28). For the measurements presented, the detector
has been filled with a gas mixture at atmospheric pressure with
10 2 counts/s
6
4
2
K
Fe
Mo
0
0 10 20
Energy (keV)
Figure 1.28 – SXR fluorescence of K, Fe and Mo
samples measured with the Si–PIN detector
Counts (Arb. units)
0.6
0.4
K
3.3 keV
6.39 keV
0
0 100 200 300 400
Energy (mV)
Figure 1.29 – Reconstructed spectra for
samples of K and Fe sources
Fe
Ar (70%) and CO 2
(30%). In order to assess the intrinsic
energy resolution of the detector, the distribution of the pulse
amplitude of the signals collected on a single pixel has been
indirectly derived, for different K α
line radiations (K, Fe, Mo).
The integral of all the counts whose amplitude is greater than
the threshold has been measured instead of the effective
distribution of the counts as function of the peak amplitude.
The distribution of the pulse amplitude has been indirectly
derived by means of scans of the threshold and by fitting it
with a guess function with three free parameters: position of
maximum and half width at half maximum of the Gaussian
peak and decay length of the tail (fig. 1.29). The distribution of
the pulse amplitude, in the range 3–17 keV, was found nearby
a Gaussian. The best agreement is found for a total detector
gain of 500 and an energy resolution of about 30% and a tail
at higher energy. The pulse broadening is entirely due to the
poor energy resolution of the detector. Scans in detector gain
have been also performed to assess the capability of selecting
different energy ranges. Combining together two samples, Fe
and Mo, fluorescence spectra with two lines have been
generated, to investigate the sensitivity of the threshold scan to
a more complex spectrum.
The reconstruction, by means of a threshold scan, of the
incident spectrum has been demonstrated in case of simple
spectra, single or double line emissions or bell shaped features.
In case of more complex spectra, with different features, the
only scan in threshold might be not sufficient. In this case a

magnetic confinement (cont’d.)
progress report
2010
025
scan in gain should also be applied, though this technique has to be
fully proven in laboratory.
256
Now the gas detector is at the Frascati laboratories in order to continue
the characterization and prepare irradiation tests under 2.5 MeV
neutron flux using the Frascati Neutron Generator (FNG) facility and
in environment with radiofrequency interference (the FTU’s hall).
Since this detector has three GEM foils for the electron amplification,
it could provide the advantage to be less sensitive to neutrons and
gammas as compared with the single stage one. These tests are
scheduled for February 2011.
2) Feasibility studies of polycapillary lenses. Polycapillary lenses [1.14]
have been studied in laboratory for preliminary characterization, with
the aim of using them in SXR diagnostics such as imaging and
tomography [1.15]. The first tests were performed to characterize the
polycapillary lenses (convergence, divergence, efficiency, spectral
dispersion, aberrations and so on) in the SXR range 5–25 keV and for
distances much larger than the optical focal length of the lenses, both
for the detector and the source. A silicon based C–MOS imager
(Medipix 2) has been used as 2D detector and the micro focus x–ray
tubes as “point–like” sources.
The full lens has been characterized by using an electronic tube with a
Mo anode powered at 25 kV, 150 μA (continuous spectra with the K α
line of Mo at 17.4 keV in addition), while the half lens by using an
x–ray tube with a Cu anode and a Ni filter (K α
line of Cu at 8.04 keV),
powered at 15 kV, 300 μA. The images have been acquired with the
Medipix 2 [1.16]. In particular, the output focus of the full lens has
been found by a scan of the distance detector–lens. The intensity of the
spot is found to be fairly Gaussian and the spot broadening is linear
with the distance (60 cm) much larger than the focal distance (4.4 cm).
The total geometry divergence results to be approximately equal to 2.6°. The same measurement has been
done with the half lens: set at 25 kV and 150 μA respectively. A radiography of the samples has been done at
first (fig. 1.30) by putting it just in front of the detector, to check structure, contrast and dimensions and so on,
as reference. The image through the lens is obtained by putting the samples just before (roughly 5 mm) the
optics. These samples have been perfectly reproduced (fig. 1.31). There is only a decrease in the intensity at
the edges of the lens, but it could be corrected by using an appropriate factor.
Imaging properties have been therefore demonstrated for full–lens, with a resolving power at least of about
100 and for distances much longer (15 times) than the focal distance. Efficiency is found to be progressively
lower at the edge, as expected, and dependent on the energy of the x–ray photons. Half–lens revealed as an
excellent light collector, having an output beam with a very low divergence (quasi parallel).
New diagnostic for dusty plasmas. The research activity on dust in tokamak plasmas during the year 2010 was
in part devoted to the development of a final design and construction of a specific diagnostics for the detection
of micrometer size dust particles, which impinge on the vacuum chamber wall at velocities larger than a few
km/s, i.e. larger than the velocity of the compressional waves of the wall material. Such particles are particular
harmful for fusion reactors because, depending on their size and concentration, they may significantly
contribute to the wall erosion due to the impact craters produced (the erosion rate evaluated on the basis of
the FTU results is indeed of the order of 10 –4 –10 –3 mm/s). Their presence in FTU was suggested by the
detection of spikes in the ion saturation current collected by an electrostatic probe, interpreted in terms of
impact ionization events, and by the impact craters found on the molybdenum probe tip [1.17, 1.18]. The new
diagnostics, developed to confirm and extend such observations, is based on the simultaneous detection of the
spikes in the ion saturation current collected by an electrostatic, tungsten probe and the light flashes emitted
by the tungsten vapour cloud formed by the impact. This coincidence allows indeed to discriminate between
events due to plasma fluctuations, which are not accompanied by line emission of tungsten, and events due to
hypervelocity impacts. A detailed study of feasibility of the new electro–optical (EO) probe has been
performed [1.19]. The main concern was represented by the background plasma emission, mainly due to the
X
1
256 Y (Row number) 1
0
19.18 38.35 57.53 76.7
Figure 1.30 – Radiography of a
plastic foil with holes
1
X
256
1 Y (Row number)
256
0 36.83 73.66 110.5 147.3
Figure 1.31 – Image with a full lens
of a plastic foil with holes

026
progress report
2010
S(μ W/sr nm)
1.5
0.5
-0.5
0
200 400
Time (ns)
Figure 1.32 – Emissions observed by
four spectral channels (blue 1020–1050
nm, green 950–1020 nm, orange
850–950 nm, cyan 670–850 nm)
Bremsstrahlung radiation. Accurate evaluations suggest that the
tungsten line emission at 401 nm (selected by an interference filter
with full width at half maximum (FWHM) ≤1 nm) by the impact
vapour cloud is expected to be larger than the background radiation
by two orders of magnitude during a time interval of the order of a
few hundred ns after the impact. The construction of the mechanical
parts of the probe was completed in 2010. The installation of the
probe in the FTU port 5, as well as the final assembly of the optical
and electrical detection systems (including data acquisition) should
be completed by 2011. It is worth noticing that the project was
developed with a strong support, both financial and scientific,
provided by the IFP of the CNR, Milan, also in the frame of a
collaboration with the University of Naples “Federico II”, the University of Molise and the Royal Institute of
Technology (KTH) of Stockholm. The EO probe development was also supported by Ministero
dell’Università e della Ricerca (MiUR) under grant PRIN–2007L4YEW4 and by EFDA baseline and priority
supports.
Another important issue considered in the framework of dust particle studies in tokamak environments was
the detection of dust mobilized after disruptions. Such dust particles were observed by the new high resolution
Thomson scattering system on Joint European Torus (JET) [1.20]. The system consists of filter spectrometers
that analyze the Thomson scattering spectrum from 670 to 1050 nm in four spectral channels. The laser source
was a 5 J Q–switched Nd:YAG laser. Without a spectral channel at the laser wavelength, the dust elastic
scattering cannot be observed and only dust particles that emit broadband light could be detected. The time
behaviour of their emission is clearly different from that expected for a Thomson scattering pulse. As shown
in figure 1.32, the emission peak is actually delayed by about 10 ns as compared to the laser pulse (15 ns
duration). Moreover, the typical emissions observed by the spectral channels have almost the same amplitude,
consistent with a black body emission of dust particles with a diameter of 10 μm at a temperature of 4000 K.
The fast decrease of the signal after the peak (10 ns time scale) suggests the occurrence of ablation
phenomena, since the radiative cooling is expected to be much slower. These studies concerning the detection
of dust during disruptions at JET are preliminary, further investigations are planned for the next JET
campaigns C28 and C29.
It is worth noticing that the characterisation of size, composition and origins of dust in fusion devices is the
subject of a specific International Atomic Energy Agency (IAEA) Coordinate Research Project (CRP), and the
FTU group is involved in this project by providing the chief scientific investigator. The report of the second
research coordination meeting of this project has been published in November 2010 [1.21].
These researches can benefit from collaboration with scientists involved in different fields, such as the study of
dust in laboratory or space plasmas [1.22].
Ion cyclotron antennas. The collaboration between IPP – Garching and ENEA–Frascati started more than one
year ago, but has been effective since spring 2010. It is essentially focused on the solution of one major open
issue related to ion–cyclotron (IC) antennas, namely the capability to reach both good plasma coupling and
low sheath potentials in front of the launchers. In particular, the sheath potentials are responsible for high
impurity production and their reduction is still a huge challenge for the design of high power ion cyclotron
resonance frequency (ICRF) antennas, with specific relevance to ITER and FAST experiments.
In this context Axially Symmetric Divertor EXperiment Upgrade (Asdex–U) stands out to be one of the best
candidates to test new IC antenna solutions due to its H–modes and to the fully tungsten wall. Furthermore,
the Asdex–U team has a remarkable experience in terms of antenna design and sputtering yield analysis, as
proved by its long list of references and international collaborations ([1.23, 1.24] to cite only a few) on these
topics.
The present work is focused on two main activities, i.e. the modelling and simulation of the actual AUG IC
antennas, and the design of a new launcher to be installed on the machine by mid 2012. Two main prediction
tools are being used for this purpose, HFSS and TOPICA [1.25], whose results have been compared as far as
rectified potentials are concerned (the integral of the parallel electric field component along magnetic field
lines in front of the antenna).
HFSS is a 3D full–wave frequency domain electromagnetic field solver based on the finite element method,

magnetic confinement (cont’d.)
progress report
2010
027
Fields are predominantly real
Re E II
(kV/m)
HFSS
-0.5
Rectified potentials
voltage (kV)
0.6
0.4
0.2
0.0
Z
0.0
(m)
(m)
0.6
0.3
0.0
-0.3
HFSS
TOPICA
-0.5 0 0.5
Z (m)
Z
TOPICA
-0.6
0.5 -0.5 0.0 0.5
(m)
Figure 1.33 – Comparison
between E–field distribution
and rectified potentials
calculated with TOPICA and
HFSS for a 2–strap flat antenna
loaded with a lossy dielectric.
Values are normalized to 500
kW per strap
Figure 1.34 – Reference and broad limiter
antennas together with their TOPICA
models (left Faraday screen has been
hidden to make the interior visible)
Figure 1.35 – Preliminary CAD
drawing of the 3–strap
antenna
while TOPICA is a 3D full–wave boundary–value
problem using a hybrid spatial–spectral domain
formulation. In addition to the different formulation
and other minor differences, the former can only take
into account dielectric loadings, while the latter also
accounts for magnetized, inhomogeneous and hot
plasmas solved in slab approximation with the finite
element method.
Up to now modelling and simulation of the present Asdex–U IC antennas has been following a stepwise
approach. Firstly the consistency of results between HFSS and TOPICA has been successfully verified in case
of dielectric loaded antennas for flat approximated geometries. Good agreement has been found between the
different tools both in terms of coupled power and electric field distribution (fig. 1.33) in front of the launcher.
Then, the real 3D geometry of the IC antennas has been imported in TOPICA for realistic simulations. The
“reference” antenna – a 2–strap antenna with narrow side limiters – has been modelled both as it is now and
as it was before 2010, since the top limiter underwent some small changes between the latter and the current
experimental campaigns. Furthermore, the broad limiter antenna – an antenna with broader limiters, thinner
straps and two additional poloidal side plates – has been modelled too (fig. 1.34). The latter is expected to
achieve a moderate reduction of the rectified potentials as compared to the aforementioned reference antenna.
Two further steps are presently being envisaged. First, TOPICA simulations of the flat and curved antenna
geometries have to be compared for several plasma profiles. Second, an assessment of TOPICA predictions
as far as Asdex–U experimental data are concerned has to be performed as soon as such data become
available.
As far as the design of a new IC antenna is concerned, different solutions have been proposed and simulated
with HFSS and/or TOPICA. In quite recent times, even though at a preliminary stage, much attention has
been payed to a 3–strap antenna solution (fig. 1.35), where image currents, circulating on the limiters and
usually causing high rectified potentials, are strongly reduced by a compensation mechanism between central
and outer straps. Unfortunately, to achieve this result, the requirements in terms of phase tuning and power
balance between straps are very demanding and they have to cope with severe mechanical constraints.

028
progress report
2010
Transmission and relative reference
spectrum (red)
1.2
0.8
0.4
0
290 GHz
nominal low-pass
Wire grid polariser, E ⊥ wires
Fabry-Perot
Mylar Mesh
Reference spectrum
0.4 1.2
Frequency/THz
WG polariser:
E// wires
Figure 1.36 –Spectral response of different
materials and components measured with the
TDS–THz spectrometer. WG is short for
Wire–Grid polarizers, the brown line
corresponds to a 250 GHz low–pass filter. Note
that the SNR in the measurements is unchanged
along the spectral range under investigation,
although the relative intensity of the reference
spectrum (red) ranges across a factor 10
2
Novel approach to the submillimeter and THz plasma
diagnostics. The Terahertz (THz) band of the
electromagnetic spectrum is defined as the range of
frequencies between microwaves and mid–infrared light,
covering the region where electronic and optical technologies
overlap. Recently, femtosecond laser pulses have been used to
generate extremely broadband (100 GHz to 30 THz)
single–cycle THz pulses. These pulses can be detected
coherently with very high sensitivity, and may be used for
time–domain spectroscopy (TDS) experiments on a variety of
systems, with a great potential for plasma diagnostics in this
area of the electromagnetic spectrum, i.e. electron cyclotron
emission (ECE), reflectometry, interferometry. A
Collaboration between ENEA Frascati and Oxford Terahertz
Photonics Group on this subject has started in 2010, and a
memorandum of understanding has been signed by the Head
of UTFUS–MAG Frascati and the leader of Oxford
Terahertz Fotonics, an excellence research group within the
sub–faculty of Physics at the University of Oxford.
The introduction of the reliable THz–TDS cutting–edge
technology in the tokamak environment will allow high
sensitivity, due to the background rejection potential of this
technique. With this system background radiation can be
rejected extremely well, to easily detect sub–micro watt average power. A great advantage will be the use of
sources and detectors intrinsically broadband, like a Fourier Transform Spectrometer, but without the
beamsplitter restrictions, and, not negligible, they are solid–state devices, operating at room temperature.
The diagnostic potential of the existing instrumentation has been assessed at the Oxford Clarendon
laboratory. Preliminary laboratory tests have been carried out, for phase measurements, windows and
materials characterization, and signal to noise ratio (SNR) (fig. 1.36). A new type of THz spectrometer is being
jointly developed: this system operates entirely with optical fibers, which would allow the usual difficulties of
installing receiving and transmitting antennae in a tokamak to be overcome.
The conceptual design of the new plasma diagnostic system, with quantities to be measured, parameters
range and measurement precision is in progress.
On the same front, a complementary line of work is in progress in collaboration with General Atomics, San
Diego. This action takes place under the Large Tokamak Cooperation (LTC) scheme. The use of a high
resolution Fabry Perot interferometer for plasma real–time control purpose is being studied, and laboratory
tests took place both in Frascati and at DIII–D. Two instruments of this kind are being realised simultaneously
in the two laboratories. On this front the work will merge with the THz device line of research, with the
combination of quasi–optical and electronics devices, which has never been implemented in Fusion, and also
offers a great range of applications in other areas.
We aim to improving the diagnostic capabilities beyond the routine plasma density and temperature
measurements, thus moving towards the measurement of fluctuations, charge density and conductivity, with
unprecedented temporal, spatial and spectral resolution.
Other diagnostics development for FTU
Study of fast events. In order to study events that occur on time scales shorter than milliseconds (such as the
quenching phase following plasma disruptions, the dust behaviour, correlations between density oscillations
and Marfes, etc), some upgrade of the FTU diagnostic system has been introduced:
• A fast camera Photron SA4 has been installed on Port 6. The new camera can operate up to 3600 frames/s
at full resolution or up to 500000 frames/sec at the smallest resolution). Image transport through the Port
plasma is obtained with a high quality fibre bundle.
• New spectral line monitors with fast ADC (CAEN DT5724 4 channel 14 MS/s Digitizer) allow observation
of H α
(6563Å, MoI (3864 Å, 5503 Å), LiI 6104 Å, BII (7031 Å), CIII (4647 Å) to be performed with time
resolutions up to 10 μs.

magnetic confinement (cont’d.)
progress report
2010
029
• 2 existing visible polychromators have been set on higher acquisition rates
• The multichord interferometer has been upgraded to 1.5 MHz acquisition frequency.
New acquisition system for magnetic probes. The new system based on analog–digital converter (ADC) on the
PXI crates is presently being tested on 16 poloidal sensors and 10 saddle coils. The final system will acquire
12 bit samples at a maximum rate of 500 Hz on 192 channels.
Tests of compact MHD sensors. The new sensors, low temperature co–fired ceramic technology (LTCC) coils
developed by Centre de Recherches en Physique des Plasmas (CRPP) – Lausanne, are constituted by 3×3 cm 2
ceramic plates, which embed the conducting coils, featuring a wide effective area and an extreme compactness.
A linear response up to 500 kHz and a sensitivity 7 times higher than the conventional Mirnov coils in use on
the FTU poloidal limiter have been measured. Two of these sensors have been mounted aside a conventional
MHD coil on the extractable poloidal limiter, with the same orientation, in order to be tested during FTU
operation. The resistance to radiation and thermal stresses will also be investigated.
Fast equilibrium reconstruction code. A real time version of the equilibrium code ODIN is currently under
implementation; among its advantages, a better behaviour of the control feedback, with additional
information on internal flux surfaces and the possibility to track a flux surface for ECRH heating can be
counted. As a result of the optimization, the code is now able to obtain a reconstruction in 10 ms on average.
Parallelization of the code is under examination; this is expected to reduce the elaboration time to about 2 ms
per each time frame. The code will be tested on FTU discharges on the next campaign.
The time–of–flight refractometer with double frequencies. A pulsed double frequency refractometer, realized by
the Russian company SPHERAtec, has been installed in FTU to measure the line average density from the
time of flight of the wave packet through the plasma. The two frequencies employed are 51.5 GHz and 60.5
GHz. The pulses alternatively produced by the two generators, each with a length of 8 ns and a repetition
rate of 250 kHz, are channelled in a single waveguide and sent to the plasma. Part of the signal is spilled by
a directional coupler for the reference signal. The oscillators are synchronized by a master clock. Once the
pulses are launched to the plasma, they are reflected back by the vacuum vessel and detected by the receiver
unit. The output signals from the reference and the receiver units are send to a pulse former PF and then to
the time–to–voltage converter, which provides a voltage proportional to the time delay between the two pulses.
Plasma–wall interaction
Activities concerning the plasma-wall interaction have been continued during the year 2010 on several issues,
in the frame of EFDA task agreements (TAs) and collaboration between FTU and ASDEX–U. Concerning
EFDA TAs, in 2010 ENEA was in charge of the coordination of the task concerning the “Development of the
PWI basis in support of integrated high–Z scenarios for ITER. Demonstration of liquid plasma–facing
components” (activity TA–WP10–PWI–05–00/ENEA/PS). Task WP09–PWI–05–01/ENEA/BS and PS:
“Injection of noble gases (Ne and possibly Ar) in the FTU discharges for testing its effect on the production of
high–Z impurities (Mo from the toroidal limiter)” was concluded in 2010 by further analysis of the data
collected in 2009. No new experiment was possible due to the FTU inactivity in 2010. In the examined ohmic
discharges with B T
=6 T, I p
=500 – 800 kA, n e
=0.7–1.6×10 20 m –3 , characterized by negligible pre–Ne injection
Mo influx, no Mo influx following Ne injection was found: in the few cases where more consistent Mo influx
was present before the Ne injection, no increase, and often a decrease, in its amount was observed.
Collaboration with ASDEX–U. In the first half of 2010, a proposal for testing the capability of CREATE
software tools to mitigate the heat load on the divertor tiles by decreasing the inclination angle of the
separatrix at the outer strike point was submitted to the IPP
management. The upper part of the plasma was frozen
and the separatrix was moved just in the neighbourhood of
the inner strike point (ISP) and outer strike point (OSP). In
figure 1.37 the reference and the modified equilibrium for
an ASDEX–U shot are reported. Dedicated plasma shots
will be allocated in the 2011 ASDEX–U experimental
program.
Z(m)
-0.7
-0.9
-1.1
Reference
+7 cm (ISP),
-6 cm (OSP)
Figure 1.37 – Reference (blue) and modified (green)
equilibrium for an ASDEX–U shot
-1.3
1
1.2 1.4 1.6 1.8 2
R(m)

030
progress report
2010
1.3 Plasma Theory
Significant efforts of theoretical research activities have been devoted to analyzing nonlinear behaviors in
burning plasmas of fusion interest, their complex dynamics and the issues that arise when modeling these
phenomena with increasingly more realistic physics and equilibrium descriptions. Many of these theoretical
activities significantly contributed to advancing the FAST conceptual design; the activities specifically carried
out within the framework of this project are summarized in the specific FAST section.
Several activities of the plasma theory group have been pursued in the framework of international
collaborations with the University of California at Irvine (UCI) and the Institute for Fusion Theory and
Simulation (IFTS), Zhejiang University (ZU), Hangzhou (China), namely: the nonlinear dispersive properties
of low–frequency shear Alfvén waves in the presence of magnetic islands which have been successfully used
for interpreting experimental observations in FTU, and the analyses of geodesic acoustic modes (GAM)
excitation by supra–thermal particles and of kinetic structures of low–frequency shear Alfvén as well as the
acoustic wave spectra with their interactions with fast ions. Other examples of fruitful international
collaborations and local research efforts that are impacting international research projects are given by the
continuing benchmarking activities of the hybrid magnetohydrodynamic gyrokinetic code (HMGC) within the
framework of the ITPA – Topical Group (TG) on energetic particle physics, the investigation of theoretical
issues to be applied for optimization of ITER magnetic diagnostics and the work carried out as contribution
to the ITM–TF.
The development of advanced numerical simulations tools, which exploit the most modern, massively parallel
architectures for high performance computing, continued with the implementation and testing of the new
hybrid magnetohydrodynamic gyrokinetic code (HYMAGYC). At the same time, further implementation and
applications of the extended HMGC (XHMGC) hybrid code have been pursued in collaboration with UCI,
IFTS and the US Scientific Discovery trough Advanced Computing (SciDAC) Gyrokinetic Simulation of
Energetic Particle (GSEP) turbulence and transport Project. XHMGC is now a very flexible and versatile tool
for investigations of burning plasma physics issues, which are made possible thanks to the increasingly more
realistic description of the supra–thermal particle dynamics (in collaboration with the Japan Atomic Energy
Agency (JAEA)) and are mostly focused on fundamental physics studies and applications to FAST (see the
FAST section). As a further example of fundamental physics studies being pursued in burning plasmas of
fusion interest, there is the characterization of complex dynamics and fishbone instability for self-organized
critical systems.
Fundamental properties of LH wave interactions with burning plasmas have been addressed, both with
explorations based on analytic techniques as well as with Hamiltonian perturbation theory methods and with
a full wave equation approach, for investigating wave propagation properties in regions with sharp density
gradients, such as those characterizing the H–mode plasma edge. Meanwhile, a new ion–heating scenario,
based on cyclotron damping of ion Bernstein waves (IBW) by tritium minority has also been addressed. Finally,
with emphasis on broader applications and implications of plasma physics, its role in the morphology of
accretion discs and on the mechanism for the generation of stellar winds or matter-jet seeds has been
addressed, in collaboration with the University of Rome “La Sapienza” and International Center for
Relativistic Astrophysics Network (ICRANet).
Alfvén modes in the presence of a magnetic island
The dynamics of Alfvén modes in plasma equilibria with a quasi–static magnetic island has been investigated
in the framework of an ideal MHD model [1.26] (in collaboration with Physics Dept. of the University of Pisa,
and Department of Physics and Astronomy of the UCI). A generalized safety factor has been defined inside
the magnetic island flux tube, giving information on the rational flux surfaces where different modes couple
and create gaps in the shear Alfvén wave (SAW) continuous spectrum. It has been found that a SAW
continuous spectrum exists, which is similar to that calculated in tokamak equilibria. A typical size magnetic
island is shown to induce wide gaps in the continuous spectrum, due to the strong eccentricity of the closed
flux surfaces inside the island. The central frequency of the gap, near the O–point, is [1.27, 1.28]
f MiAEgap =f BAECAP
2 2 2
1+Mn islfA
/fBAECAP
3)

magnetic confinement (cont’d.)
progress report
2010
031
where M=(q 2 0 s2 /2)(W isl
/r o
) is the parameter accounting for the
magnetic island amplitude, f BAE–CAP
is the BAE continuum
accumulation point and f A
is the Alfvén frequency (the MiAE
subscript standing for magnetic island induced Alfvén eigenmode).
Here q 0
,s,r o
are, respectively, the safety factor, the magnetic shear
and the radial position at the mode rational surface, and W isl
is the
magnetic island half width. This phenomenon is analogous to the
formation of ellipticity–induced Alfvén eigenmode (EAE) gaps in
tokamaks [1.29] but, here, it is localized inside the magnetic island
separatrix. When considering modes with the same helicity as the
magnetic island, the frequency of the nonlinearly modified BAE
continuum accumulation point (BAE–CAP) has been calculated.
The existence of MiAE, namely the global Alfvén eigenmode (AE)
localized inside the magnetic island, has been investigated. The
case of small eccentricity of the flux surfaces has been considered,
for this is the case that can be investigated with analytical
techniques. As a result, the dispersion relation of eccentricity
induced MiAE (or MiEAE) has been calculated. Moreover, the
contribution of finite Larmor radius terms was found to be crucial
to create the potential well for other AEs. These modes, with
characteristic length scale of the order of the ion Larmor radius,
have been identified as kinetic MiAE (or MiKAE). Their frequency
is predicted to be above the MiAE gap and they are localized in a
region close to the rational surface (fig. 1.38).
It has also been shown that the SAW continuous spectrum in
finite–beta tokamak plasmas and in the presence of a magnetic
island is still characterized by the beta–induced gap at low
frequencies. This window in the Alfvén continuum, namely
0

032
progress report
2010
1.4
1
0.6
0.2
-0.2
0 0.2
ω G (r)/ω tr,b
Im (δE)
Re(δE)
0.4 0.6 0.8 1
r
Figure 1.40 – The global mode structure
of EGAM in the case of exponentially
small EGAM tunneling coupling to KGAM.
Here, the GAM frequency at the beam
intensity peak is slightly larger than the
beam transit frequency; ω G (r b ) ≅ 1.1ω tr,b
That EPs can drive GAMs has been observed in recent experiments
[1.39,1.40] and theoretical stability properties of this
energetic–particle–induced GAM (EGAM) were presented in
[1.41]. While the continuous spectrum is one of the key features of
GAMs, previous theoretical treatments [1.37,1.39,1.41–1.45] often
ignored it by either focusing on the derivation of local dispersion
relations or introducing the ad hoc assumption that modes are
localized near an extremum of the continuous spectrum [1.46,1.47]
and, thereby, neglecting its associated radial structures.
Reference [1.48] first investigated the excitation of EGAM
accounting for the GAM continuous spectrum, thus demonstrating
the analogy between EGAM and EPM. The corresponding
dispersion relation exhibits the finite threshold condition in EP drive
due to the continuum damping of the global EGAM when it is
strongly coupled to the GAM continuum. In this work, the
excitation of EGAM has been considered when its coupling to the
GAM continuum is formally exponentially small, analogous to the corresponding EPMs excited by localized
EP sources [1.49]. The mode, meanwhile, is excited via the transit frequency resonance of EPs in the small
magnetic drift orbit limit. This case has been analyzed as being complementary to the global EGAM stability
problem analyzed earlier, assuming finite size magnetic drift orbits and strong coupling to the GAM
continuous spectrum [1.49]. Since the frequency and mode structure of EP modes are determined by the EPs,
EPs have been treated on the same footing as the bulk plasma, i.e. nonperturbatively. Both local and nonlocal
dispersion relations of EGAM are derived by assuming single pitch angle slowing–down energetic ion
equilibrium distribution function. For a sharply localized EP source, it is shown that the mode is self–trapped
where the EP drive is strongest [1.50], with an exponentially small damping due to the tunneling couplied to
the GAM continuous spectrum, where the EGAM is mode converted to a short wavelength kinetic GAM
(KGAM) [1.47], which propagates outward, as shown in figure 1.40 [1.50] and consistent with
observations [1.51].
Kinetic structures of shear Alfvén and acoustic wave spectra in burning plasmas
The low frequency fluctuation spectra of shear Alfvén and acoustic waves are important to burning plasmas
of fusion interest, since they can be excited by both thermal as well as supra–thermal particles in various
parameter regimes. There is an increasing consensus that accurate treatments of key physics processes of these
spectra do require a kinetic approach [1.46,1.52–1.60]; although whether a fully kinetic approach is really
necessary is still being debated [1.61].
In this work [1.62], recently derived analytic theories are briefly summarized, which demonstrate the crucial
importance of employing kinetic analyses for fundamental processes in collisionless plasmas of fusion interest
in the low frequency regime, i.e., where geometry effects, magnetically trapped particles [1.59,1.60], wave
particle interactions and finite (kinetic) plasma compressibility play major roles. Analytic expressions show that
the acoustic branch is generally more strongly damped than the shear Alfvén branch, which is the one of
practical interest for interpreting experimental observations. Meanwhile, analytic calculations can be used to
benchmark results of numerical codes in the relevant limits [1.60,1.63], thus helping their verification
processes. Possible further extensions of present analytic theories, which are already underway, have also been
considered.
ITPA benchmarking activity using HMGC
The benchmark activity, proposed by the ITPA topical group on energetic particle physics, between the hybrid
MHD–gyrokinetic particle in cell codes HMGC [1.64] and MEGA [1.65], and the gyrofluid code TAEFL
[1.66] has been continued. In particular, the shear Alfvén mode stability of a model equilibrium characterized
by aspect ratio R 0
/a=0.1, safety factor q(s) =1.1 + 0.8116 s 2 , with s=(1–ψ/ψ 0
) 1/2 , β bulk
=0 and normalized
bulk density profile n i
(s)/n i0
=[q 0
/q(s)] 2 , has been analyzed in the presence of an energetic particle distribution
function given by F(P ϕ
,E)= (m H
/2πT H
) 3/2 n H0
exp(k m H
cP ϕ
/e H
|ψ 0
|) exp(–E/T H
), with T H
=T H0
and
P ϕ
=–(e H
/m H
c) |ψ 0
| s 2 +v
R. A energetic particle profile (k=2.5) steeper than that of the previous benchmark

magnetic confinement (cont’d.)
progress report
2010
033
test case (k=2.0) has been considered in order to emphasize the core
plasma dynamics.
Figure 1.41 compares codes’ results showing the growth rates of
different modes (internal, half–radius EPM and outer gap mode)
versus n H0
/n i0
.
γ/ω A
1
0.06
HMGC-internal
HMGC-gap
HMGC-epm
MEGA-epm
MEGA-internal
TAEFL-internal
TAE eigenfunctions for ITER magnetic diagnostics
In the frame of the F4E task “F4E–2009–GRT–047 System-level
optimization of the ITER magnetic diagnostics”, the full MHD
code MARS [1.67] has been used to generate the magnetic field
perturbation at the plasma surface as generated by a tipical toroidal
Alfvén eigenmode (TAE). An ITER equilibrium (reference scenario
2) has been generated by the code FIXFREE, which can produce an
-0.02
0.0018 0.0022
EQDSK file as output, which, in turn, has been given as input to the high resolution equilibrium code
CHEASE, in order to generate a suitable equilibrium input for the stability code MARS. MARS has been used
first to found the Alfvén continuous spectra for a variety of toroidal mode numbers, in order to identify the
frequency range were the toroidal gap appears. Secondly, the MARS code has been used to obtain the global
eigenmodes (TAEs), which exist with frequencies that lie in the toroidal gap; moderate toroidal mode numbers
(n∼10) have been used, and the 2D spatial structure of the magnetic field perturbations at the plasma/vacuum
interface for a specific toroidal location have been produced. Such perturbations will be used as a benchmark
to test the goodness of the magnetic measurements envisaged for ITER in reconstructing mode structures.
0.02
0
0.0026 0.003
n H0 /n i0
Figure 1.41 – Growth rates vs n H0 /n i0 for
shear Alfvén modes found by HMGC,
MEGA and TAEFL codes
ITM–TF activities
The activities of the Frascati theory and modeling group in the frame of the ITM–TF in year 2010 were
concentrated on the Integrated Modelling Project 5 (IMP5) “Heating, current drive and fast particles”. Beside
continuing the development and testing of the new Hybrid MHD–Gyrokinetic Code HYMAGYC (see next
section), the main activity of this year was concentrated on porting the codes on the Gateway and interfacing
them with the ITM environment (Consistent Physical Objects (CPOs), Kepler actors, Kepler workflows).
The codes involved in this activity are, by now, HYMAGYC, HMGC and RAYLH. During the year 2010 a
strong activity has been carried out for critically revising the completeness and the definitions of several
quantities stored in the equilibrium and the coreprof CPOs (the first usually written by equilibrium solvers,
the second usually written by a transport solver, as, e.g., the European Transport Solver (ETS) developed
within the ITM–TF, or, alternatively, both obtained by experimental data): this effort has been mainly pursued
by using the MHD code from which the MHD solver used in HYMAGYC comes from (the MARS code
[1.67]), with the collaborative effort of other ITM projects (mainly IMP12 and IMP3). The result of this year’s
activity has been very fruitful, and, as a by–product, the full MHD, resistive, general geometry code MARS
has been fully ported on the Gateway, i.e., fully interfaced with the CPOs (release 4.08b); furthermore, a
Kepler actor and a Kepler test workflow have been generated. This work will be propaedeutical to the porting
of the full HYMAGYC, and has been crucial for the progress of the activities of the full IMP5 group.
Regarding the code HMGC and its extended version (see next), the routines that produce the simplified
equilibrium (circular cross–section geometry, shifted magnetic surfaces, low bulk plasma pressure
approximation) required by the code are presently under modification, and will be completed early next year.
Regarding the code RAYLH (a ray–tracing code for lower hybrid propagation), the activity of porting the code
on the Gateway and substituting the original equilibrium and plasma description (obtained by a EQDSK file)
required to solve the propagation problem with the ITM CPOs has began, and will be finished early next year,
together with the interfacing of the code with the complete ITM environment, with the ultimate goal, e.g., of
producing a Kepler actor to be used by the ETS.
The new hybrid MHD–gyrokinetic code: HYMAGYC
The HYMAGYC [1.68] was built by interfacing an equilibrium module (a modified version of the CHEASE
code [1.69]), a MHD module (an initial–value version of the original eigenvalue MHD stability code MARS

034
progress report
2010
δnm-1,n
4×10-11
2×10 -11
0
-2×10 -11
a)
3×10 -10
0
-1×10 -10
δn m,n1×10-10
b)
δn m+1,n
6×10-11,
2×10 -11
-2×10 -11 0
c)
-4×10 -11
0
40 80 120
Time
-3×10-10
0 40 80 120
Time
-6×10 -11
0 40 80 120
Time
Figure 1.42 – Fourier components of energetic particle perturbed density for a given electromagnetic field with m=4, n=4. Blue
and black curves are, respectively, numerical and analytical results for the real part of density Fourier component; red and
green curves are numerical and analytical results for the imaginary part. a) δn m-1,n b) δn m,n c) δn m+1,n
[1.67], adapted for the computation of the perturbed scalar and vector potentials, besides the perturbed
magnetic and velocity fields) and a gyrokinetic particle–in–cell module (yielding the energetic ion pressure
tensor returned to the MHD solver). The gyrokinetic module had been previously tested as to the single
particle orbits in equilibrium fields.
In order to test the response of energetic particles to a given electromagnetic field, an analytical model has
been developed for large aspect ratio equilibrium with circular magnetic flux surfaces and flat q profile, and
for circulating energetic particles with ρ H
/a

magnetic confinement (cont’d.)
progress report
2010
035
|φ(r)| (Arb. units) |φ(r)| (Arb. units)
1
0.8
0.6
0.4
0.2
0
0
1
0.8
0.6
0.4
0.2
a)
1
(1,3)
b)
1
(2,3)
c)
(5,3)
(6,3)
(7,3)
0.8
0.6
(8,3)
(9,3) 0.6
0.2
(10,3)
(11,3)
0
(12,3) 0.4
-0.2
ω/ω A0 ω/ω A0
0.2
0
-1
0.4 0.8
0 0.4 0.8
-1
1
1
a) b) c)
0.8
0.6
0.4
0.2
Z/a Z/a
-0.6
0.6
0.2
0
-0.2
-0.6
0 1
0
0
-1
0 0.4 0.8
0 0.4 0.8
-1
0 1
r
r
(R-R 0 )/a
Figure 1.44 – Simulation results of KBAE excitations by energetic particles for β H = 0.009. The top panel shows the case with
no kinetic thermal ions, the bottom panel shows the one with kinetic thermal (core) ions where β ic =0.0128. Column a) is the
radial mode structure, column b) is the frequency spectrum of the electrostatic component of the fluctuating em field, and
column c) is the poloidal mode structure
ion frequency gap but also discretizing the BAE–SAW continuum. In that case, the discrete KBAEs are readily
excited by the EP drive.
The extended hybrid MHD gyrokinetic model for XHMGC has been further generalized. In order to account
for the finite parallel electric field due to parallel thermal electron pressure gradient, the parallel Ohm’s law is
modified by accounting for kinetic bulk plasma responses. Meanwhile, electron pressure gradient is added into
the vorticity equation and finite parallel electric field is taken into account in the particle motion
self–consistently. The further extended model for XHMGC allows us to confirm the theoretical predictions –
based on the generalized fishbone–like dispersion relation [1.56] – that the Alfvénic fluctuation is always the
least damped among the kinetically modified shear Alfvén and sound fluctuation branches and, thus, in
general the experimentally most relevant one[1.62].
Parametric form of equilibrium particle distribution functions for implementations in HMGC
The hybrid MHD gyrokinetic code HMGC and its extended version XHMGC [1.70] are capable of loading
a parametric form of equilibrium particle distribution functions, given in the space of particle constants of
motion. In this way is possible to use a numerically effective “delta–f scheme” without introducing any
spurious effects and numerical noise caused by a prompt relaxation of the initial distribution function. The
parametric distribution function adequately represents energetic particle populations in different situations: α
particles and supra–thermal particle tails produced by either minority ion cyclotron resonance heating (ICRH)
or (negative) neutral beam injection ((N)NBI). When FOW effects are ignorable, the parametric distribution
function readily reproduces the “slowing down” (SD) for αs, the anisotropic SD for (N)NBI and the
bi–Maxwellian for ICRH minority heating. It is possible to choose parametric dependences to fit known
profiles, such as temperature or density profiles, from experimental data or other simulation results. This
distribution function has been applied to simulate the behavior of energetic particles due to ICRH minority
heating in FAST [1.71].
Design of energetic particle equilibrium distribution functions
To the purpose of implementing an equilibrium particle distribution function in more general conditions than
those described just above, a numerical method has been developed (in collaboration with the Naka Fusion
Institute, JAEA), which allows particle simulations of toroidally confined plasmas to be initialized with a
particle distribution function F eq that owns the following properties: (i) having been constructed from
unperturbed particle orbits, F eq is an exact equilibrium that properly takes into account the effect of magnetic

036
progress report
2010
drifts and finite system size; (ii) chosen moments of f eq match given reference profiles as much as physical
constraints permit; (iii) the process of designing the distribution in velocity space is made largely independent
of any spatial profile fitting. The method is particularly useful for simulations where magnetic drifts across
magnetic flux surfaces are significant. In such cases, velocity and spatial coordinates are intertwined and the
above three properties are difficult to be simultaneously obtained with analytical models. The new numerical
method was implemented in a tool called VisualStart, which is written in Matlab and provides for interactive
graphical user interfaces in order to assist the user with the design of a complete initial snapshot for
MHD–kinetic hybrid codes.
The above analytical and numerical methods will now be applied to represent more realistically and more
accurately than before the energetic ion populations created by NBI and ICRH. Through the new distribution
functions, simulations of energetic–ion–driven instabilities will be carried out by using XHMGC [1.70] and
applied in plasma conditions that are relevant for burning plasma physics studies, such as those of FAST.
A new model for self–organized critical dynamics, with implication of the fishbone–like
instability cycle
The phenomena of self–organized criticality (SOC) originally proposed by Bak, Tang, and Wiesenfeld [1.72]
have been investigated in a new lattice model, dubbed dynamic polarization random walk (DPRW) model, in
which the transported quantity is taken to be the electric charge of electrostatically interacting particles of
different kinds. For convenience, the transport problem is defined in a capacitor-like geometry as shown in
figure 1.45.
The model is motivated by the well–known problem [1.73] that, by increasing the probability of site
occupancy, a lattice is brought to a critical point characterized by fractal geometry of the threshold
percolation, i.e., self–similar distribution of arbitrary large finite clusters, each presenting the same fractal
geometry of the infinite cluster. This fractal geometry is made self–organized via mechanisms of random
walk–like hopping of the lattice defects («holes») between the nearest-neighbor conducting sites. The
concentration of the conducting sites, in its turn, is self–consistently controlled by the varying potential
difference on the capacitor. That means that the dynamical state of the lattice is coupled with the particle
injection and loss processes. The system is shown to be fluctuating near the state of the threshold percolation
in response to a small white-noise perturbation applied to the charged plate. The DPRW model gives critical
exponents consistent with the results of numerical simulations of the traditional «sand–pile» SOC models
[1.72,1.74,1.75], and stability properties, associated with the scaling of the control parameter versus distance
to criticality. It is found that relaxations to SOC of a slightly supercritical state are described by the Mittag-
Leffler function, since this is the solution to the fractional relaxation equation [1.76], and not by a simple
exponential function as for standard relaxation. The durations of relaxations events are power-law distributed,
with the diverging upper cut–off relaxation scale. The model belongs to the same universality class as the Bak,
Tang, and Wiesenfeld sand–pile [1.72], and should be distinguished from the directed-percolation-based SOC
models [1.77].
One important result worth noticing is that the behavior of the DPRW system crucially depends on the
strength of external forcing.
Sub-critical: pp c
Indeed, a self–organized state
mimicking a pure SOC state only
occurs in the limit of infinitesimal
driving when the random–walk
motions in the lattice are allowed
-I
to relax before input conducting
sites are again introduced. This
being not the case, the system is
– Free charge (electron)
– Polarization charge
– Hole or lattice defect
found to become unstable. The
instability manifests itself as sharp
periodic bursts of the particle loss
Figure 1.45 – DPRW model. Grey, blue, and azure particles show respectively the current in the ground circuit. The
free charges, polarization charges, and holes. a) System below the percolation bursting mode is excited because
point. Free charges are built by external forces on the capacitor's left plate. b) the dc conductivity of the lattice
System slightly above the percolation point, with an illustration of hopping nonlinearly changes with the
activities on the lattice. Adapted from ref. [1.79]

magnetic confinement (cont’d.)
progress report
2010
037
driving strength. The instability is «resonant» in that the particle loss process is directly proportional to the
electric current in the ground circuit. This «resonant» property dictates a very specific nonlinear twist to the
bursting dynamics, thus enabling to associate this instability of SOC with the «fishbone» instability in
magnetic confinement systems with high–power beam injection [1.52] (the lattice occupancy per site
corresponds to the effective resonant beam–particle normalized pressure within the q=1 surface; the
percolation point corresponds to the mode excitation threshold; and the particle loss current in the ground
circuit corresponds to the amplitude of fishbone). The transition to bursting dynamics signifies the increased
role of global and nonlinear behaviors in the strongly driven DPRW system as compared to a pure SOC
system. The results of the present study also suggest that the nonlinear science of fishbone [1.46] might be
thought of as being connected with those properties generic to strongly-driven, interaction–dominated,
thresholded dynamical systems such as the DPRW supercritical system rather than unique to toroidally
confined fusion plasmas. An account of some of these investigations have been reported recently [1.78,1.79].
The comprehension of fishbone–like instability of SOC when account is taken for the global character of the
bursting dynamics suggests a type of mixed SOC–coherent behavior [1.80] in which the multi–scale features
due to SOC can coexist along with the global or coherent features. One example of this coexistence is
speculated [1.79] for the coupled solar wind-magnetosphere-ionosphere system. These theoretical fidings are
under way for being discussed as part of the scientific rational of the ROY project [1.80], a state–of–the–art
multi–satellite project in space research aiming to investigate the Earth's magnetosphere as a complex
multi–scale system interacting with its solar wind drive.
Asymptotic techniques in the study of the propagation of high frequency (lower hybrid range)
electromagnetic waves
The propagation and absorption of high frequency electromagnetic waves in laboratory plasma and in
presence of an external magnetic field, is modeled by an integro-differential system of equations: the
Maxwell–Vlasov system (in collaboration with IFTS– ZU). Starting from the Maxwell–Vlasov system, a
simplified equation has been deduced that describes the propagation of an electrostatic pulse in cold plasmas
and general magnetic equilibrium, which has been analytically studied by means of a multiple spatial scale
approach. This technique is strictly related with that discussed earlier in [1.81], and, to the lowest order,
reduces to the well–known “ballooning formalism” [1.82]. A simplified equation is derived and studied for the
scalar potential in the cold plasma limit, by applying the WKB asymptotic technique to describe the slow radial
dependencies of the wave envelope, while the full–wave equation is considered along the magnetic field lines.
This Ansatz can be entirely justified on the basis of spatial scale separation in the radial direction and, thus,
this approach could be viewed as a mixed WKB–full–wave technique.
Analytical and numerical studies of the cold electromagnetic LH wave equation in the mode
conversion regime
The purpose of this study was to investigate the effect of mode conversion (n z

038
progress report
2010
non–integrability, due to the presence of two or more degrees of freedom, may lead to Hamiltonian chaos.
These aspects have been studied by producing phase portrait plots, Poincare maps, and computing numerically
the Lyapounov exponents, thus elucidating some interesting features of the LH propagation in the external
plasma.
The propagation and absorption of lower hybrid waves in H–mode plasma with very sharp
density gradient at the pedestal
Flat density profiles in tokamak plasmas are characterized by a density rise at the plasma periphery, which may
prevent LH waves from penetrating deeply into the center. The reasons may be the deviation of the ray
trajectories toward the plasma edge or the lack of validity of the linear optics approximation (WKB). In this
work the wave equations has been solved in full wave limit, in order to analyze whether outward reflections
take place. The results indicate that, unless the density rise is a very sharp jump, the reflection is negligible and
the linear WKB wave equations are a valid tool to describe the wave ray tracing. These trajectories result to
be deviated across the density rise, but their deviation depends on the central density value and scarcely on the
edge gradient strength.
Tritium minority heating with mode conversion of fast waves
A new ion–heating scenario in tokamak plasmas is proposed, based on cyclotron damping of IBWs by tritium
minority at the first ion cyclotron harmonic (i.e. ω=2Ω cT
). The IBWs are coupled by mode conversion of fast
magneto–sonic waves (FMW) in a D–H(T) (tritium minority in hydrogen-deuterium) plasma. The mode
conversion layer is located near the centre of the plasma column as well as the resonant layer of the tritium
minority. A possible scenario for the JET tokamak [1.84], based on the present idea, has been analyzed by
means of the numerical codes toroidal ion cyclotron (TORIC) & steady state Fokker Planck quasi linear
(SSFPQL) [1.85,1.86]. As a result, Tritium ions are accelerated up to energies close to the peak value of the
DT cross–section and a significant increase in Q has been found [1.87].
∼
ν z
2000
2.0
1000
0
u
0
1.0
2
4
x
6
8 0.0
Figure 1.46 – Behavior of the vertical velocity ν ∼ z as a
function of the dimensionless radial x and vertical u
coordinates
Stellar winds or matter–jet seeds from disk
plasma configuration
The profile of a thin disk configuration was studied
[1.88] as described by an axisymmetric ideal MHD
steady equilibrium (in collaboration with the University
of Rome “La Sapienza” and ICRANet). The disk was
considered as a differentially rotating system dominated
by the Keplerian term, but allowing for a non–zero radial
and vertical matter flux. In this scheme, the steady state
allows for the existence of local peaks for the vertical
velocity ν ∼ z
of the plasma particles (fig. 1.46), though it
prevents the radial matter accretion rate from taking
place. This ideal MHD scheme is therefore unable to
solve the angular momentum–transport problem, but it is
suggested that it provides for a mechanism for the
generation of stellar winds or matter–jet seeds. A
visco–resistive scenario has also been set up [1.89], which generalizes previous two–dimensional analyses by
reconciling the ideal MHD coupling of the vertical and the radial equilibria within the disk with the standard
mechanism of the angular momentum transport, relying on dissipative properties of the plasma configuration.
1.4 JET Collaboration
In October 2009, JET operations were suspended to start the installation of the new ITER like wall (ILW)
using beryllium and tungsten as plasma facing components (PFC). The related shut–down is still continuing
and the first plasma is now foreseen for the beginning of August 2011.

magnetic confinement (cont’d.)
progress report
2010
039
ENEA Frascati scientists have participated in the analysis of the data gathered during the last campaigns and
finalized journal papers and contribution to international conferences, such as EPS, SOFT, IAEA and others.
On the basis of data taken during the last experimental campaigns, significant progress was made in terms of
analysis, interpretation and diagnostic commissioning. Highlights regarding the main areas of interest are
reported below.
Pellet studies
The analysis of pellet ablation and particle deposition experiments has been completed and presented at the
2010 IAEA Conference [1.90]. Data show a clear evidence of ∇B drift effects in agreement with code
simulations (fig. 1.47). These simulations were performed within the JINTRAC code suite by using JETTO as
a core transport code and HPI2 for pellet ablation and ∇B drift effect calculations. In L–mode or moderately
heated H–mode this effect seems to be weak. At high heating power instead, injection from the vertical high
field side (VHFS) has proven to be much more effective than from low field side (LFS) in depositing particles
beyond the pedestal. In the latter case,
ablated particles are quickly lost due to a
combination of ∇B drift and edge
instabilities. The fuelling performance,
although investigated only in a limited
range of plasma and pellet parameters,
seems to be promising for the capability of
pellets to raise the density without
increasing the neutral pressure in the main
chamber. This latter feature is important
for ITER, since it confirms that pellets are
less demanding in terms of pumping
capabilities.
High beta operation
Stability of high–beta advanced scenarios
has been analyzed for a wide range of
q–profiles and normalized beta values up
to β N
=4. Global n=1 instabilities limit the
achievable β N
or degrade confinement.
Most recent results were presented at the
2010 IAEA Conference [1.91]. Stability
boundaries in terms of q min
and pressure
peaking have been determined. For
relatively broad pressure profiles the limit
decreases from β N
=4 at q min
=1 to β N
=2
at q min
=3, while at fixed q min
it decreases
with increasing pressure peaking. Bursting
and continuous n=1 instabilities have also
been analyzed. A new form of instability
that grows on typical resistive time-scales
but has kink internal structure has been
identified. This instability systematically
occurs above the no–wall ideal stability
limit as calculated for n=1 modes by the
MISHKA code. The measured mode
structure is in good agreement with
eigenfunctions given by the code. An
overview of the mode analysis results is
given in figure 1.48 on a β N
vs q min
plane.
Δn/Δn peak
1.0
0.6
0.2
JPN 77863 – LFS hybrid
-0.2
0 0.4
r/a
a) b)
JPN 77864 – VHFS hybrid
0.8 0 0.4 0.8
r/a
Figure 1.47 – Particle deposition code simulations (blue) show a clear
effect of ∇B drift on pellets injected from a) the LFS, and b) VHFS into
high power (∼21 MW) hybrid discharges. Particles injected from LFS
are rapidly displaced outwards as compared to ablation (red) and
leave the main plasma, while in the VHFS case they are shifted inwards,
thus producing a significant deposition (black), as measured by the
HRTS
βN
4
3
2
1
0.5 1.5 2.5
q min
no n=1
Broad b.
Chirping
Disrupt.
Cont.
Figure 1.48 – Distribution of stability characteristics in the β N –q min
diagram for MHD modes with toroidal number n=1. Circles and
squares are taken at maximum β N in discharges without any n=1 mode
and with weak broadband n=1 activity respectively. Triangles
represent discharges with bursts of short–lived (10 ms) intense
chirping modes, in particular downward pointing triangles represent
disruptive cases. Stars represent the onset of long–lived n=1 modes

040
progress report
2010
2
Pulse # 73744
a)
6
S(μW/sr nm) Ipla (MA)
1
0
0.4
0.2
0
0
4
Time (s)
Figure 1.49 – After a disruption, in the absence of plasma, the
extra light seen by the Thomson scattering detectors is thought
to be due to the interaction between the laser pulses and dust
particles later released by the wall. a) plasma current, b)
average dust signal of the 4th spectral channel of every core
spectrometer
b)
8
P thr (MW)
4
2
0
0
H e conc scan shots
P thr, scal 08
D references
40 80
H e concentration (%)
Figure 1.50 – Power threshold (P thr ) obtained with NBI
as function of helium concentration for both He and D
reference discharges at 1.7 MA/1.8 T. All the
configurations had the same low δ≈0.25 shape, and
densities variations 2.3–2.8×10 19 m –3
Dust
Radiation has been observed by the high resolution Thomson scattering (HRTS) system after plasma
terminations, which reveals the presence of impurities along the beam path released by the wall during the
disruptive power quench, known as dust [1.92]. This radiation has been carefully analyzed and, due to its
spectral features, seems not to be due to pure scattering but to a more complicated laser–matter interaction.
The intensity is compatible with the expected dimension of dust grains and the temporal decay shows two
timescales (fig. 1.49). Averaging the values of the peaks in the signal traces over all spectrometers, one can have
a rough estimate of the dust content. The dust lasts for about 1.5 seconds after the disruption. A preliminary
distribution of the signals has been obtained and compared with a compound Poisson distribution.
L to H mode transition
Data taken during the 2009 JET helium campaign have been fully analyzed and presented at the 2010 EPS
Conference [1.93]. L–H in helium has been observed as a transition to Type III ELMs with a small
confinement improvement associated with a small edge pedestal. Helium concentration varied from 1 to 87%
and was found to have little impact on the power threshold (fig. 1.50). This is in line with recent ASDEX–U
studies, but in contrast with JET 2001 results, which showed that He plasmas had a 40% higher threshold than
D equivalents. However, it should be noted that the 2001 comparison was performed at lower density
(∼2.0×10 19 m –3 ) than the more recent data base (∼2.3–2.8×10 19 m –3 ). If a slope inversion of the density
dependence were at play, the 2009 and the 2001 results might be reconciled. Unfortunately, helium data at low
density are not available in the 2009 database. Although physics understanding still remains to be improved
and further experiments will have to be carried out, recent results seem to encourage the helium choice for the
non–nuclear phase of ITER.
Safety factor profile determination
Profiles of the safety factor are determined by the motional stark effect (MSE) diagnostic combined with
equilibrium reconstruction. The positions of low–order rational surfaces q=m/n (from 2/1 to 4/3, m and n
being the poloidal and toroidal mode numbers, respectively) have been systematically compared with the ones
inferred from the frequency of MHD modes. The connection between mode frequency and rational surface
location depends on the velocity of the magnetic island produced by the mode at the corresponding rational
surface. A previous assumption of island propagation at the toroidal velocity of the carbon impurity resulted
to be inconsistent. Much better agreement was found by assuming that the island rotates with the deuterium
ions fluid; the latter was calculated by radial force balance from carbon toroidal rotation and ion temperature
profiles neglecting poloidal rotation. If neoclassical carbon poloidal rotation is included in the force balance,

magnetic confinement (cont’d.)
progress report
2010
041
the agreement between MSE and MHD mode locations remains good, whereas the agreement is spoiled if
measured poloidal rotation is used. More investigation is needed to understand this discrepancy. For reversed
profiles, uncertainties in q values near the axis are higher than those for monotonic profiles, although in some
cases they are seen to be corresponding to experimental data from x–ray tomography and the observation of
Alfvén Cascades [1.94].
Upgrade of neutron profile monitor (KN3N)
The performances of the new digitizing system for the neutron profile monitor (NPM) has been assessed by
means of the analysis of data collected during JET plasma discharges [1.95]. The signals from the NPM
NE213 scintillators are acquired by a set of five digital pulse shape discrimination (DPSD) boards, each one
with 4 acquisition channels (20 channels in total: 19+1 spare) with 14 bit resolution and 200 MHz sampling
rate. Each DPSD acquisition board is fully configurable: thanks to the ethernet port and the embedded Linux
operating system integrated in the field programmable gate array (FPGA), it is possible to set the parameters
of the acquisition, to poll the state of the system as well as to start/stop the acquisitions.
The system has been demonstrated to be capable to provide simultaneous 2.5 (DD) and 14 MeV (DT) neutron
count rates. An example is given figure 1.51 for line of sight #4 (discharge #79698); the plot also shows
neutrons due to fast ion tails (FAST), corresponding to the proton energy range 3.5–10 MeV. For the same
discharge, figure 1.52 shows the line–integrated neutron emission
profiles for the DD and DT cases. Note the different rise times of
LOS#4 #79698
the DD and DT profiles at 9.5 s and 10 s due to the triton slowing
800
down time.
6×104
Work on the implementation of the KN3N data acquisition and
processing within the JET CODAS system has been started in 2010.
Elaboration of JET programme for 2011
The ENEA–Frascati team has also actively participated in the
elaboration of the JET programme for 2011 by sending its
representatives to the main Planning Meetings held in Culham, and
by joining many working group meetings from remote. A number
of experiments have been proposed in view of the new wall
conditions, which are now in the process of being selected and
rationalized in an executable time-line. Main areas are of interest
include: current profile and recycling control to access hybrid and
steady state regimes, pellet studies, detection and control of intrinsic
impurities, bulk tungsten tile response to heat loads, impact of edge
parameters on LHCD efficiency, threshold for the L to H–mode
transition with Be/W versus carbon wall, dust detection following
disruptions, integration of MHD, MSE and polarimetry data for
q–profile reconstruction.
Coordination of other Italian partners
ENEA has also coordinated the contribution of other Italian
partners to JET activity and taken an active part in their execution.
During the shut–down, scientific papers have been produced, based
on former Campaigns and also some enhancements have been
carried on mainly in the field of plasma diagnostics (University of
Rome Tor Vergata) and plasma control (CREATE Consortium).
In the field of diagnostics, in collaboration with Tor Vergata Rome
University, a detailed statistical analysis of the polarimetric
measurements, acquired during the campaigns 2003–2009, has
been performed [1.96] thus highlighting some calibration problems
Counts DD (s–1)
2×10 4
0
48
DD
DT
DT
52 56
Time (s)
400
0
Counts DT,FAST (s –1 )
Figure 1.51 – Neutron count rates for DD,
DT and FAST energy windows (pulse
#79698)
Counts/s
Counts/s
1×105
6×10 4
2×104
0
0
400
200
49.5 s DD (1.8-3.5 MeV)
50.0 s
a)
50.5 s
53.0 s
55.0 s
#79698
10 20
Channel
49.5 s
DT (10-16 MeV)
50.0 s
b)
50.5 s
53.0 s
55.0 s
#79698
0
0 10
Channel
Figure 1.52 – Neutron profiles for pulse
#79698: DD a), DT b)
20

042
progress report
2010
mainly related to the experimental data of the most recent campaigns (2008–2009), particularly those of the
high current discharges. During 2010 a new calibration algorithm [1.97] was developed and completely
validated for offline analysis to the purpose of to overcoming this problem.
Plasma control
The plasma pontrol activity was carried out in collaboration with CREATE Consortium. Fusion relevant
theoretical studies, tool development and technology studies were carried out in the framework of the general
support to physics activities and underlying technology. A significant fraction of the activity has been addressed
to the following fields:
Modelling of resistive wall modes. The dispersion relation of resistive wall modes (RWM) has been derived
when a cylindrical plasma circumvented by a thin resistive shell and an outer, non–conducting, ferromagnetic
wall is considered [1.98]. It has been demonstrated that the resulting growth rate is always higher than that in
the absence of any ferromagnetic materials, hence it always provides a destabilizing effect, which is consistent
with previous results similarly obtained for somewhat different arrangements. The growth rate increase does
not depend on the plasma configuration but only on the poloidal mode number as well as on the physical
and geometrical properties of the ferromagnetic shell. Several calculations have been made to verify these
findings on the JET tokamak with the CREATE L code. Although the JET geometry is very different from the
simplified case treated analytically, the main qualitative conclusion is absolutely confirmed.
Development of a real–time framework for data acquisition and control in fusion experiments. The CREATE
group has contributed in the development of the MARTe [1.99]. MARTe is a modular real–time
multiplatform C++ framework tailored at, but not restricted to, the development and deployment of control
systems. This framework has been recently adopted to drive the vertical stabilization of the JET tokamak
[1.100,1.101].
Breakdown analysis. A modeling activity has been carried out by using a refined CREATE L state space model
[1.102] in order to analyze the electromagnetic conditions of JET breakdown, also for what concerns the
possible impact of the new radial field system on plasma breakdown [1.103]. This work identified for the first
time the radial field due to the up-down asymmetric gaps in the iron core and the relative influence on the
field null.
Current limit avoidance. The preliminary activity for the implementation of a current limit avoidance system
within the existing architecture of the shape controller has been carried out [1.104]. In particular, the detailed
design document and the Level–1 interface documents have been issued at the end of the year.
Presentation of the main achievements of the PCU2 projects. In 2009 the PCU2 project was mainly focused
on designing and commissioning the new vertical stabilization system of JET. Therefore, little time was
dedicated to the preparation of the report, the methodic and systematic analysis of the experimental results,
and the presentation of the achievements to the scientific community. This has been effectively done in 2010,
with the compilation of the final report, the presentation of the commissioning phase, and the illustration of
the main achievements [1.105].
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046
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2010
chapter 2
fusionadvancedstudiestorus
It is presently widely accepted that a successful exploitation of ITER (“the way” in Latin) and a reliable as well
as early design of demonstration/prototype reactor (DEMO) needs a strong accompanying program. Along
this roadmap, a key role should be played by the so–called “Satellite Experiments”: Japan Tokamak 60 Super
Advanced (JT–60SA) [2.1] and possibly Fusion Advanced Studies Torus (FAST) [2.2]. The two proposals are
complementary with each other and hold the capability of accessing overlapping regions in the operational
space; this may allow scientific achievements and experimental results to be compared. The primary aim of
FAST is studying integrated plasma scenarios to the broadest possible extent.
a) Plasma wall (PW) problems that ITER is expected to face, with an outlook on possible scenarios of
relevance to DEMO; consequently, a large power load is foreseen (P/R∼22 MW/m –P/R∼12 MW/m),
with actively cooled divertor and first wall (FW) in full tungsten; in addition, the very high operational
density ( up to ∼ 6×10 20 m –3 ) will allow for experiments with high density and radiative plasma edge
(up to ∼90%) even at low collisionality, as in ITER and unlike in other devices.
b) ITER and DEMO will necessarily tackle severe operational problems, such as the presence of large edge
localized modes (ELMs) (and the need of mitigating them), and the necessity of completely integrated
plasma control tools; FAST will have very large ELMs (up to few MJ) and a complete set of systems to
control the plasma operations in an integrated environment.
c) Burning plasma stability and mutual feedbacks between thermal plasma and energetic particle populations
are among the most interesting and unexplored physics aspects in view of ITER and, more importantly,
DEMO. The possibility of performing ITER–relevant integrated experiments in satellites machines relies
on the similarity argument based on the existence of three dimensionless parameters: ρ ∗ , β and ν ∗ [2.3]
and in the so called “weak scaling” [2.2].

fusion advanced studies torus (cont’d.)
progress report
2010
047
2.2 Physics
Transport simulations
Table 2.I shows the main FAST operating scenarios foreseen, which should cover all the ITER scenarios and
should be capable to address some of the predictable DEMO necessities. Several different core transport
models have been used to simulate the FAST reference scenario, some of them based on first principle
(Weiland and GLF23) and some on semi–empirical models (mixed Bohm–gyroBohm (BgB) and critical
gradient model (CGM)).
In these simulations the density profile has been
assumed to be either flat (as it usually happens in the
present H–modes) or peaked (as scaled by the
present database for the FAST low collisionality
case); in the case of simulations in which the plasma
density is left to evolve (as, e.g. bu using the GLF23
transport model), it results to be peaked. The
simulations have been carried out by using the
JETTO code; for ion cyclotron resonance heating
(ICRH) heating profiles we have either used the
PION code called self–consistenly by JETTO or by
TORIC that is run outside JETTO and requires a
few iterations. The final result is that all models
predict about the same electron temperature, but
there is a larger range of variation in the T i
profiles.
Table 2.I – Operating scenarios
In particular a careful attention has been devoted to Q 0.65 1.5 0.32 0.06
studying the alternative scenario where only 15 MW t discharge (s) 20 13 55 170
ICRH are used together with 15 MW electron
t flat–top (s) 13 2 45 160
cyclotron resonance heating (ECRH) at 170 GHz.
I
The ECRH heating profiles have been provided by NI /I p (%) 15 15 60 >100
the GRAY code, that requires iterations with
JETTO. In this scenario the ICRH has been
P ADD (MW) 30 40 40 40
reduced up to the minimum level sufficient to generate a β H
∼1%. Since the ECRH resonance is at 6.1 T,
B T
=6.0 T and I p
=5.5 MA were used for this simulation; however, as a consequence of the Shafranov shift, the
actual experiment could be planned at B T
=6.7 T with a plasma current of around 6.0MA. In figure 2.1 the
electrons and ions temperature, as predicted by using the different models, are shown for the case with only
ICRH and for the case with ICRH+ECRH. In the latter case the electron temperature always remains higher
than the ion temperature (T e
(0)∼15 keV, T i
(0)∼9 keV). Although the ECRH deposition has been spread out
on Δρ∼0.2, the fact that T e
>T i
can be justified by the much larger input power density on the electrons when
FAST
H–mode
reference
HMR
H–mode
extreme
HME
AT
Full
NICD
I p (MA) 6.5 8.0 3.5 2
q 95 3 2.6 5 5
B T (T) 7.5 8.5 6 3.5
H 98 1 1 1.2 1.2
(m -3 ) 2 5 1.4 1
β N 1.3 1.7 2.2 3.4
τ E (s) 0.4 0.65 0.20 0.10
τ res (s) 5.5 5 3 2÷5
T 0 (keV) 13.0 9.0 12 7.5

048
progress report
2010
T(eV)
ICRH BgB
ICRH W
ICRH GLF
ICRH CGM (chisi=2)
ICRH CGM (chisi=4)
ECRH+ICRH BgB
ECRH+ICRH W
ECRH+ICRH GLF
ECRH+ICRH CGM (chisi=2)
ECRH+ICRH CGM (chisi=4)
ICRH BgB
ICRH W
ICRH GLF
ICRH CGM (chisi=2)
ICRH CGM (chisi=4)
ECRH+ICRH BgB
ECRH+ICRH W
ECRH+ICRH GLF
ECRH+ICRH CGM (chisi=2)
ECRH+ICRH CGM (chisi=4)
5000
T e
n e
1×10 4 2×10 20
T i
6×10 20
4×10 20
n e (m –3 )
ωφ(rad/s)
1×10 5
8×10 4
6×10 4
6×10 4
n e
v f
4.5×10 20
1.5×10 20
4×10 4 3×10 20
n e (m –3 )
2×10 4 0.4 0.8 0 0.4 0.8
0
0
ρ
ρ
Figure 2.1 – Ion and electron temperature and density profiles shown
for the case of ICRH +ECRH (full line) and full ICRH (dotted line). Red
lines are for old BgB, blue for Weiland, black for GLF23 and green for
CGM. The assigned density profile is shown with dashed line
2×10 4 0
0
0.4 ρ
0.8
Figure 2.2 – Rotation profile for NICD scenario
compared to the ions one. This causes a negative loop to take place, where higher T e
/T i
decreases the ion
temperature gradient (ITG) micro instabilities threshold, with consequent colder ions and not significantly
hotter electrons than the full ICRH case.
As shown in table 2.I the non inductive current driven (NICD) scenario has, in principle, the unique capability
of studying a full non inductive regime at high β N
, with a reactor relevant FW in tungsten. However, FAST
will only have a small input of external momentum (like ITER) and only in the framework of the negative
neutral beam injection (NNBI) scenarios. From recent experimental results it seems that the toroidal rotation
plays a key role in achieving improved ion core confinement, not only through the well–known threshold up
shift, but through a significant reduction of the ion stiffness. The rotation has been included in the simulations
by modeling the momentum transport with physical assumptions consistent with recent theoretical
developments. Due to the inward pinch, core rotation in FAST can be driven by intrinsic rotation edge sources.
Given the present lack of understanding and the theory–based predictive capability on intrinsic rotation, we
have assumed for FAST an edge rotation value ω φ
=30 krad/s. Considering the high intrinsic rotation values
measured in C–MOD, which is a high field compact machine conceptually similar to FAST, it may still be
legitimate to assume an edge rotation value as predicted by the existing C–MOD driven empirical scaling. The
NICD scenario, with I p
=2 MA, has been simulated by using the BgB model, without including any torque
source and by adding 4 MW of LHCD at 5 GHz (n
=2.3). In figure 2.2 the obtained/used rotation profile is
shown. An ion temperature profile with an internal transport barrier (ITB)–like gradient, around ρ∼0.5÷0.6,
has been obtained with a reversed q profile (q min
∼2). The ion and electron temperature are very close with
T i0
∼20 keV, T e0
∼15 KeV and with a density n e0
∼2×10 20 m –3 . These parameters have to be regarded as
overestimated due to the simplistic assumptions of the BgB model.
Energetic particle physics in H–mode scenario with combined NNBI and ICRH
The combination of ICRH+NNBI in FAST allows the generation of fast ion populations to occur with
different velocity space anisotropy and radial profiles. These energetic ion populations can excite meso–scale
fluctuations with the same characteristics as those expected in reactor relevant conditions and, for this reason,
FAST can address a number of important burning plasma physics issues. Numerical simulation and modeling
of energetic particle physics are based on the use of various transport codes that are iteratively coupled with
a bi–dimensional full wave–quasi–linear solver for ICRH, which also includes the solution of the
Fokker–Planck equation for NNBI–plasma interactions. Self–consistent profiles evolution is obtained by the
suite of codes CRONOS with combined ICRH/NNBI heating. The ICRH frequency is in the range
80–85 MHz and the NNBI energy from 0.7 to 1 MeV depending on species (hydrogen or deuterium). A
parametric study of the normalized supra-thermal population pressure β hot
is being presented here and
discussed in terms of the RF+NBI power deposition profiles with various minority concentrations ( 3 He 1–3%).
The value of β hot
as well as the energetic particle distribution functions can be used as an initial condition for
numerical simulation studies, thus investigating the destabilization and saturation of fast ion driven Alfvénic

fusion advanced studies torus (cont’d.)
progress report
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049
ω/ωA0
1
0.5
1
0.8
0.6
0.4
ω/ω A0
1
0.5
1
0.8
0.6
0.4
0.008
β hot⊥
0.004
Equil.
n=8
n=16
0.2
0.2
0
0 0
0 0
0 0.4 0.8 0 0.4 0.8 0
ρ pol
ρ pol
0.4 0.8
ρ pol
Figure 2.3 – a) Intensity contour plot (of the electrostatic component of the fluctuating em field) of the n=8 Alfvénic mode in
the space of normalized frequency and radial position. The shear Alfvén continuous spectrum for n=8 is indicated by the black
solid line; b) Intensity contour plot of the n=16 Alfvénic mode in the space of normalized frequency and radial position. The
shear Alfvén continuous spectrum for n=16 is indicated by the black solid line; c) Effect of n=8 and n=16 saturated EPM on the
β hot⊥ radial profile
Figure 2.4 – One of the divertor geometries
used for the SOL/edge simulations with
EDGE2D/EIRENE
modes by means of a recently extended version of the hybrid
-1.30
Pump B
FAST divertor n. 4
magnetohydrodynamic gyrokinetic code (HMGC) [2.4],
which is able to simultaneously handle two generic initial
-1.40
particle distribution functions in the space of particle
1.1 1.3 1.5 1.7
constants of motion. The crucial role played by radial
Major radius R (m)
non–uniformities and by the shear Alfvén continuous
spectrum, shown as solid lines in figures 2.3a and 2.3b, is
consistent with the theoretical framework discussed in [2.5,2.6]. The effect of n=8 and n=16 saturated EPM
on the β hot⊥
radial profile, shown in figure 2.3c, suggests that significant radial redistributions of energetic
particle are expected with limited global losses.
Heights z(m)
-1
-1.10
-1.20
Pump A
Plasma–wall interaction activities and optimization of the divertor geometry
The modelling of the FAST scrape–off layer (SOL)/edge plasma with the software package EDGE2D +
EIRENE was carried out in 2010. Simulations have been run for the basic H–mode scenario: different
preliminary divertor designs (fig. 2.4) have been examined by varying the density at the separatrix over the
range n sep
=0.7–1.0×10 20 m –3 [2.7]. The input power into the SOL has been set to 20 or 16 MW and no
impurity seeding was allowed. Besides the importance of an as shallow as allowed strike angle of the separatrix
on the divertor target, simulations highlight that significant detachment is attained only if the divertor
geometry and pump location are optimized to allow the formation of a consistent neutral cloud close to the
strike point. At the lower separatrix densities the help of strongly radiating impurities would appear to be
unavoidable.
Diagnostics
A first assessment of the extent to which the 3 He fast ion population can be diagnosed in FAST with a set of
dedicated diagnostics has been performed [2.8], in order to show the capabilities of FAST in studying fast ion
physics in conditions relevant to a burning plasma. These energetic confined fast particles, with dimensionless
parameters (normalized Larmor radius and plasma pressure) close to those of fusion-born alphas in ITER,
will be produced in deuterium FAST plasmas by means of 30 MW ICRH minority heating. Neutron emission
spectroscopy (NES), gamma ray spectroscopy (GRS) and collective Thomson scattering (CTS) diagnostics have
been reviewed with a description of the state–of–the–art hardware and a preliminary analysis of the required
lines of sight.
Taking as an input the spatially resolved numerical simulations of the energy distribution of ICRH accelerated
particles, the corresponding spectra observed by the set of diagnostics have been simulated for specified lines

050
progress report
2010
of observation and the observables related to changes in the tails of the fast ion energy distribution have been
determined and studied. As for NES, the neutron emission (from the fusion reaction) energy spectrum reflects
the energy distribution of the deuteron population with the possible suprathermal components caused, for
instance, by external heating. The GRS measured gamma ray emission spectra, from reactions between fast
ions and impurities, allow different information to be extracted: fast ions exceeding the threshold energies of
the reactions (by the identification of characteristic gamma ray peaks), combined fast ion density, temperature
and impurity concentration (by peak intensities), fast ion temperature (by doppler broadening of the emission
lines observed). At last, in CTS the doppler broadened scattered radiation provides space and time resolved
information on the velocity distribution of the plasma ions, including the fast–ion population generated by
ICRH. The detailed characterization of the fast particles dynamics poses demanding requirements since all
the relevant plasma parameters should be ideally measured with spatial resolution close to the fast particles
Larmor radius (of the order of 6 cm) and time resolution of the order of the interaction time of Alfvén modes
with the fast particles (of the order of 30 ms) in the FAST H–mode reference scenario.
The results of the analysis, based on numerical simulations of the spatial and energetic particle distribution
function of the ICRH accelerated minority 3 He ions, have shown that NES and GRS measurements can
provide information on the anisotropy of the fast 3 He population and a measurement of its effective tail
temperature, with time resolutions in the range 20–100 ms, and that the proposed CTS diagnostics can
measure the fast ion parallel and perpendicular temperature with a spatial resolution of 5–10 cm and a time
resolution of 10 ms.
2.3 Design Description
The FAST project significantly advanced in the year 2010 [2.9] as for load assembly design, vacuum vessel
and plasma facing components (fig. 2.5), with close attention to the remote maintenance issues. Extensive
studies were carried out to investigate several divertor design options.
Divertor design
The divertor optimization of the divertor from the physics and engineering point of view significantly
advanced in 2010; at the same time, a whole divertor design has been completed [2.10], able to withstand a
power load up to 20 MW m –2 by using W monoblocks tiles [2.11,2.12]. Further studies will be carried out to
complete the optimization in the next year.
The divertor, as currently designed, is composed by a cassette body (CB) that supports all the plasma facing
components (PFC). The cassette body routs the water coolant and is mounted onto the vacuum vessel inboard
and outboard rails via a mechanical locking system. Each
divertor module spans over 5 degrees for a total of 72
modules in the whole tokamak and each module is made of
6 rows of W monoblocks, individually cooled by pressurized
water flowing in CuCrZr pipes, 12 mm diameter wide. The
dome geometry is provided with a large opening in front of
the strike zones in order to allow neutrals to flow into the
private flux region and to be removed from there by pumps
located in the divertor port (fig. 2.6). Slots drilled in the
divertor body cassette provide for the needed pumping
conductance. The plasma facing components are expected
to be able to remove the heat load coming from the plasma
via conduction and convection during the normal and
transient operation. The outboard vertical target poloidal
profile has been chosen in order to minimize the impinging
heat loads by assuring ∼20° striking angle.
Figure 2.5 – Complete CATIA5 model showing the
main components of FAST: from outer to inner, the
cryostat, the poloidal and toroidal field coils, the
vessel with the access ports, the first wall with the
divertor and the central solenoid
The hydraulic scheme of the divertor cooling foresees a
manifold which feeds the coolant into each tiles row from
the top of the outer vertical target, the coolant than flows

fusion advanced studies torus (cont’d.)
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051
Pressure
4.02×10 06
3.77×10 06
3.51×10 06
3.25×10 06
3.00×10 06
2.74×10 06
2.48×10 06
2.23×10 06
1.97×10 06
1.72×10 06
1.46×10 06
Pa
Figure 2.7 – Pressure in the divertor pipes
Figure 2.6 – FAST divertor design with ~20° outward strike
point angle
along the dump plate and the dome, continues through
the inner dump plate and vertical target and finally
returns into the cassette manifold through the back of
the inner vertical target. A preliminary 3D
thermo–hydraulic analysis has been carried out for the
divertor module by using the ANSYS CFX fluid
dynamics code. The total heat power on each individual
divertor module a single module (the total power on the
whole divertor is 22.7 MW) has been splitted between
the two vertical target surfaces (2/3 in the outer and 1/3
in the inner vertical target) and applied as heat flow into
the model. Assuming the water flowing with 10 Kgs –1
rate and 4 Mpa pressure, 120°C as inlet conditions, a
2.2 MPa pressure drop (fig. 2.7) and 11°C temperature
increase have been computed in a single divertor
module. The temperature distribution along a single
tiles row in the straight part of the outer vertical target
has been evaluated by assuming the same inlet condition
as before (1.67 Kg s –1 mass flow rate, 4 MPa inlet
pressure, 120°C inlet temperature). The model includes
the presence of the swirl tape as foreseen in the bottom
part of the vertical target. It was supposed that the W
monoblocks are bonded on the CuCrZr pipe by a Cu
OFHC interlayer. The heat load has been assumed to
vary exponentially with an energy decay length of 5 mm
at the outer midplane and a factor 5 for the expansion
at the target, thus resulting in ∼20 MWm –2 on the strike
point. The average heat transfer coefficient at the
interface between water and pipe has been computed as
∼110 kW m –2 K –1 . The maximum stationary
temperature reached is ∼1700°C on the W tile facing
the plasma and ∼490°C on the Cu OFHC layer
(fig. 2.8).
Temperature
1.7×10 03
1.6×10 03
1.4×10 03
1.3×10 03
1.1×10 03
9.3×10 02
7.7×10 02
6.1×10 02
4.4×10 02
2.8×10 02
1.2×10 02
[C]
Figure 2.8 – Temperature contour on the divertor outer
vertical target
Figure 2.9 – FW and divertor in the FAST CATIA5
model
First wall design
The first wall design advanced in the last year, based
upon a solution consisting in a bundle (fig. 2.9) of
poloidal coaxial pipes (fig. 2.10) armoured with 4 mm
thick plasma-sprayed W [2.10]. This configuration
Figure 2.10 – Section of the FW model showing the
poloidal coaxial pipes bundle

052
progress report
2010
allows a very uniform temperature distribution to be achieved by the pipes and gives the opportunity to have
a single manifold for each segment, thus minimizing the eddy current effects on the structure and easing the
system remote handling. The heat load impinging on the FW is, on average, 1 MW m –2 , with about
3MWm –2 peak. The First Wall is designed to be actively cooled down by pressurized water flowing with a
rate of 5 m s –1 , able to keep the temperature around 200°C in order to avoid any impurities adsorption. The
design will be compatible with a completely remote handling and maintenance.
Vacuum vessel design
The vacuum vessel (VV) provides for vertical (up, down), oblique (up, down) and equatorial access ports (90 in
total) for the plasma diagnostics, the vacuum, the auxiliary heating, the in–vessel remote handling (RH)
maintenance and all of overall the systems that get to the vacuum vessel [2.10]. Special machine ports were
designed to accommodate a 10 MW (45° inclined on the plasma cord) NNBI system (fig. 2.11) [2.13].
The equatorial ports are characterized by an aperture shape that is relatively high and rather narrow. Two
ports are one side enlarged to accommodate the NNBI beam. Since the beam are injected 45° on the magnetic
axis the port available space is reduced. This configuration has been chosen to assure the best compromise
between the narrow spaces and the designed 10 MW of power input to be supplied. Although at high density
practically no shine–through is predicted, in case of low density operation a dump plate protection must be
foreseen in the inner and outer first wall.
Neutral beam
Critical
points
NBI equatorial port
Calorimeter
Beam line vessel
Figure 2.11 – Horizontal section of the NNBI system
access port at FAST
Figure 2.12 – FAST toroidal module
The vacuum system has to reach a base pressure of
about 1×10 –7 Pa, in order to minimise the presence of
any possible residual impurities in the vacuum vessel
before the plasma discharge. By allowing for a specific
outgassing rate (after all cleaning procedures) of
6.7×10 –9 Pa m 3 s –1 m –2 , the needed effective pumping
speed is about S eff
=2.2 m 3 s –1 . The conductance
between a pump and the vacuum vessel is dominated
by the conductance of the vacuum line between the
pump and the port, the dimensions of the latter being
rather large. As a result the overall pumping speed
must be about 4.25 m 3 s –1 and therefore 4
turbomolecular pumps, with pumping speed of
1.5 m 3 s –1 will have to be used.
Toroidal field coil system
The toroidal field coils (TFC) system has a 20°
modular configuration (fig. 2.12), with a total of 18
coils. Each coil consists of 14 oxygen free copper plates
suitably worked to realize 3 turns in radial direction.
The turns of each coil are welded on the most external
region in order to obtain a continuous helix and the
plates are tapered at the innermost region to realise the
wedged shape. A conceptual design of the feeding bars
has been done, keeping in account the “return
currents” for each coil: the relevant preliminary 3D
electromagnetic analysis has not indicated any
noticeable perturbation introduced by the bars. In the
highest performances H–mode scenario, a field of
8.5 T is foreseen at the major radius R 0
=1.82 m; this
corresponds to a total of 76.5 MA–turn, i.e., to
101.1 kA per turn. The maximum radial inward force
acting on each TFC is 66.5 MN, while the vertical
force on half of the TFC system is 690 MN, for the
highest performance case. An appropriate structural

fusion advanced studies torus (cont’d.)
progress report
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053
analysis has been carried out, thus allowing to predict a stress lower than 300 MPa for the worst case. The
TFCs are kept together by a steel open structure surrounding them (fig. 2.13), a structure that should be quite
elastic in order to be able to accommodate TFC thermal expansion with reasonable stresses. The same
structure is also used to position the poloidal coils, which surround the TFC, and to fix the VV supports. Two
preloaded rings, in the upper and lower zones of the structure, keep the whole TFC structure in wedged
configuration.
In order to increase the plasma duration in the NICD scenario, currently
imposed by the TFC heating, and to address the DEMO conditions, a
feasibility study has been worked out to the purpose of analyzing the
possibility of realizing FAST with a complete set of superconductor coils
[2.14]. The toroidal field superconductor coil (TFSC) has been designed
by filling up only the room available in the present copper conductors
design. Due to the maximum magnetic field over the winding pack around
14 T, in the innermost layer of the inboard straight leg on the equatorial
plane for the reference H–mode scenario, the superconductor chosen is a
Nb 3
Sn cable–in–conduit–conductors (CICCs). The design envisages a
Double–Pancake Winding Pack, wound from a rectangular conductor
with an aspect ratio lower than 2, long twist pitch values, low void fraction
and without any central channel. The nuclear heating for this option has
been determined by using data provided by a MCNP code analysis [2.15].
As the presence of a nuclear shield would considerably impact the
machine geometry, it has been considered a solution not to envisage any
neutron shield but with an actively He cooled steel casing, by means of adhoc
designed channels. With this design, the conductor temperature
margin is around 0.9 K for a nuclear heating of about 5 mW/cm 3 . Then
superconductors coils could be introduced in the current machine load
assembly design without any major modifications [2.16].
Figure 2.13 – Model of the TFC
supporting structure
Toroidal field ripple reduction
The plasma boundary in FAST is close to the TFCs, thus resulting in a
relatively high (as in ITER) deviation from the toroidal symmetry of the
toroidal magnetic field, known as toroidal field ripple (TFR). The TFR
maximum value in FAST is, without any correcting solution, about 1.5%
on the plasma separatrix region at 45% from the equatorial plane. The
first solution, analysed to the purpose of reducing the TFR as much as
possible, was based on the insertion of ferrite elements between the TFCs
and the plasma. Nevertheless, this solution could produce the appearance
of a TFR with opposite sign when operating at lower toroidal field, as it
is widely foreseen in FAST. A conceptual study was then performed to
analyze the option of realizing an active control of the ripple amplitude
[2.16] by means of proper coils located outside the VV in front of every
TFC (fig. 2.14) and fed with a current in the opposite direction as to the
Figure 2.14 – Schematic view of the
active coils designed to control the
ripple amplitude
TFC current. The maximum current foreseen for these coils is the same as for the TFC system, thus
simplifying then the power supply and feeding issues. By varying the current in coils the ripple value can be
actively controlled up to reverse its sign, thus allowing not only the required reduction to values compatible
with plasma operations but also the flexibility needed to study the ripple role in the pedestal behaviour and in
the H–mode performance. Preliminary structural studies indicated that the forces exerted on these coils
require the design of proper support structures to be integrated in the vacuum vessel.
Poloidal field coil system
The poloidal field (PF) coils system, consisting of a vertically segmented (in 6 modules) central solenoid (CS)
and of 6 external coils, allows for the needed poloidal flux swing requested to create and maintain the plasma
as well as and for shaping flexibility. In the present design the CS, the external PFCs and the busbars are made
of hollow copper conductors for cooling.

054
progress report
2010
Although the PFC system is not the main element constraining the plasma duration, a feasibility study has
been performed to analyze the opportunity to realize it with superconductors [2.14]. The large flux necessary
for the plasma sustainment brings to maximum magnetic fields around 17.7 T in the CS, on the equatorial
plane, which represents a difficult challenge for the superconductor coils design. In order to investigate its
feasibility, a design of the 6 CS modules by using Nb 3
Sn strand, characterized by high critical current density
performance, has been carried out while maintaining at the same time such stringent operative conditions and
taking up the same room foreseen for the copper in the reference design. The high value of the maximum
magnetic field in the CS conductor will require future modifications of the coil design to be performed, mainly
by introducing conductor grading aimed at optimizing the room occupancy and thus the current density level.
Moreover the large magnetic field variation during the scenario, in the order of 1÷2 T/s, will require a careful
AC losses analysis to analyze their impact on the strand choice.
As for the 6 external PFCs superconductor option, the relatively low maximum magnetic field suggested the
use of NbTi CICCs, which is affected by degradation of performance due to the EM loads lower than the
Nb 3
Sn strands. Also these external coils were designed with a rectangular shape to permit a higher packing
factor without degrading the single strand performance.
Poloidal field coil system flexibility and plasma control
FAST has been designed to allow a large flexibility in the plasma shape to be achieved: to investigate and
optimize this feature of the machine, the extreme shape control (XSC) algorithm, developed for Joint
European Torus (JET) [2.18], has been also applied to FAST [2.19]. The XSC Tools implementation for
FAST has shown the possibility to achieve plasma shapes significantly different from the reference ITER–like
shape, for instance with an increased aspect ratio R/a up to 4. Another application enabled by XSC Tools is
the possible sweeping of the plasma boundary strike points on the divertor plates, with a fixed plasma shape,
to spread out the energy deposition on the plates. A strike point movement around 12 cm, on the vertical
target, has been shown to be possible by leaving unchanged the reference plasma boundary unchanged.
Cooling system
Helium gas is used for cooling the magnets down to 30 K. The refrigerator is required to operate in a steady
state cooling mode. However the heat load of the magnet system is delivered in pulses and then the smoothing
of the pulsed heat load will be taken into account for the refrigerator design by including details of buffer tank
and heat transfer in the magnet system. The inner surface of the cryostat and the VV are covered with 80 K
thermal shield to prevent high heat radiation from the cryostat to the machine and from the VV to the toroidal
field coils. The total heat load required to maintain the machine at 30 K temperature has been assessed as 5
kW. For the most demanding scenario (pulse length up to 170 s) the inter–pulse cooling time is below 1 h. The
refrigerator concept makes reference to the detailed study carried out by Linde Kryotechnik AG for a similar
plant.
Remote handling
Due to neutron activation [2.15], the repair, inspection and maintenance of FAST in–vessel components
should be performed remotely. The scheme of the divertor maintenance operation is foreseen as in ITER, with
the frame acting as a carousel around the machine to move the divertor modules by mean of an ad-hoc
developed cassette tractor, capable to grasp, pull and push the cassettes through the oblique lower port.
Collaborations with the ITER Divertor Test Platform in Tampere (Finland) and with the Department of
Mechanics, Virtual Reality Laboratory, of the Napoli II University have been started for this purpose and a
first simulation of the divertor modules movement has been performed by using the FAST full model. As for
the first wall remote assembly and disassembly, a classical articulated boom plus a front end manipulator has
been considered.

fusion advanced studies torus (cont’d.)
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055
References
[2.1] N. Hosogane et al., Fusion Sci. Tech. 52, 375 (2007)
[2.2] A. Pizzuto et al. , Nucl. Fusion 50, 095005 (2010)
[2.3] B.B. Kadomtsev, Sov. J. Plasma Phys. 1, 295(1975)
[2.4] X. Wang, et al., An extended hybrid magnetohydrodynamics gyrokinetic model for numerical simulation of
shear Alfvén waves in burning plasmas, accepted for publication in Phys. Plasmas
[2.5] L. Chen and F. Zonca, Nucl. Fusion 47, S727 (2007)
[2.6] F. Zonca et al., Plasma Phys. Control. Fusion 48, B15 (2006)
[2.7] V. Pericoli–Ridolfini et al., Simulations of the SOL plasma for FAST, a proposed ITER satellite tokamak, Presented
at the 26th Symposium on Fusion Technology – SOFT (Porto, 2010) and to appear in Fusion Eng. Des.
[2.8] M. Tardocchi et al., Production and diagnosis of energetic particles in FAST, Proceedings of the 23rd IAEA
Fusion Energy Conference (Daejon 2010), CN–180 – Paper EXW/P7–26 (2010)
[2.9] F. Crisanti et al., FAST: a European ITER satellite experiment in the view of DEMO, Presented at the 26th
Symposium on Fusion Technology – SOFT (Porto, 2010) and to appear in Fusion Eng. Des.
[2.10] A. Cucchiaro et al., Engineering evolution of the FAST machine, Presented at the 26th Symposium on Fusion
Technology – SOFT (Porto, 2010) and to appear in Fusion Eng. Des.
[2.11] E. Visca et al., Manufacturing, testing and post–test examination of ITER divertor vertical target W small scale
mock–ups, Presented at the 26th Symposium on Fusion Technology – SOFT (Porto, 2010) and to appear in
Fusion Eng. Des.
[2.12] S. Roccella et al., Non–destructive methods for the defect detection during the manufacturing of ITER high
heat flux components, Presented at the 26th Symposium on Fusion Technology – SOFT (Porto, 2010) and to
appear in Fusion Eng. Des.
[2.13] M. Baruzzo et al., Requirements specifications for the neutral beam injector on FAST, Presented at the 26th
Symposium on Fusion Technology – SOFT (Porto, 2010) and to appear in Fusion Eng. Des.
[2.14] A. Di Zenobio et al., FAST: Conceptual design for a completely superconducting magnet system, to appear in
IEEE Trans. Appl. Supercond.
[2.15] R. Villari et al., IEEE Trans. Plasma Sci. 38, 406-413 (2010)
[2.16] G.M. Polli et al., A 2D thermal analysis of the superconducting proposal for the TF magnet system of FAST,
Presented at the 26th Symposium on Fusion Technology – SOFT (Porto, 2010) and to appear in Fusion Eng.
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[2.17] G. Calabrò et al., Active toroidal field ripple reduction on FAST, Presented at the 26th Symposium on Fusion
Technology – SOFT (Porto, 2010) and to appear in Fusion Eng. Des.
[2.18] R. Albanese et al., Fusion Eng. Des. 74, (2005).
[2.19] F. Maviglia et al., Poloidal field circuits sensitivity studies and shape control in FAST, Presented at the 26th
Symposium on Fusion Technology – SOFT (Porto, 2010) and to appear in Fusion Eng. Des.

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chapter 3
technologyprogramme
The technology program performed by the Association in 2010 cover the most crucial activities related to
ITER, the European Fusion Development Agreement (EFDA) program and Broader Approach (BA)
agreement.
Many of these activities were performed within Consortia established among Association and collaboration
with industry.
The most relevant results achieved in the period are the following.
The ENEA technology developed for the construction of the divertor monoblock plasma facing units has been
successfully applied, in collaboration with Ansaldo Nucleare, for the construction of the qualification
prototype.
An important irradiation experiment has been carried out to assess the design of the ITER shielding blanket.
A mock up of the shielding blanket plus vacuum vessel and the toroidal field (TF) magnet (utilizing actual
cable–in–conduit (CIC) conductor) has been irradiated. The test results were very useful to reduce the
uncertainties on the heat deposited on the superconductors and to better define the shield block design. A
number of MCNP models have been developed in the frame of the design of the high resolution spectrometer,
the radial neutron camera and the in vessel and divertor coils. Concerning sensor development, further to the
continuation of the diamond sensors, gas electron multiplier (GEM) were utilised as neutron detectors.
The in vessel viewing system (IVVS) developed by ENEA for ITER has been further improved by adding a
vibration filtering system.
Safety studies have been performed for studying the mobilization of dusts in the event of a loss of vacuum
accident. Experimental and modelling activities were accomplished with also with the aim to develop the
proper diagnostic system.
In the field of materials, a multi-scale method for modelling ceramic composites has been implemented
developing a special subroutine to be utilised in ABAQUS code.
In Brasimone the breeder blanket test facilities were further improved with the commissioning of TRItium
EXtraction (TRIEX) loop for the simulation of the lithium lead purification system.
The Broader Approach activities were mainly focussed on the International Fusion Materials Irradiation
Facility (IFMIF) and materials. Target design based on an alternative remote handling scheme has been further
developed and the preparation of the experimental activity in EVEDA lithium test experiment (ELITE) facility
in Japan has been started. Test facility to investigate the compatibility issues of SiC/SiC in high temperature
flowing lithium has been realised. In this frame the construction of lithium purification loop has been
completed. As for the Japan Tokamak 60 Super Advanced (JT60–SA), the technical specification of the TF
magnet have been finalised and those for the power supply and switching system are in progress.
Training activities have been performed in the field of liquid metal, magnet and divertor technology in the
frame of the Euratom training scheme.

technology programme (cont’d.)
progress report
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057
3.1 Divertor, First Wall, Vacuum Vessel and Shield
Manufacturing technology for the ITER inner vertical target
ENEA has been deeply involved in the ITER R&D activities for the manufacturing of high heat flux
plasma–facing components, and in particular for the inner vertical target (IVT) of the ITER divertor.
This component has to be manufactured by using both armour and structural materials whose properties are
defined by ITER. Their physical properties prevent standard joining techniques from being applied. The
reference armour materials are tungsten and carbon/carbon fibre composite (CFC), and, for the cooling pipe,
a copper alloy (CuCrZr).
During the last years ENEA, in collaboration with Ansaldo, has been manufacturing several actively cooled
mock–ups and prototypical components of different length, geometry and materials, by using innovative
processes: hot radial pressing (HRP) and pre–brazed casting (PBC).
The optimization of the processes started from the successful manufacturing of both W and CFC small scale
mockups and successful testing in the worst ITER operating condition (20 MW/m 2 ) through the achievement
of record performances obtained from a medium scale IVT CFC and W armoured mockup: after ITER
relevant heat flux fatigue testing (20 MW/m 2 for 2000 cycles CFC part, 15 MW/m 2 for 2000 cycles W part)
it reached a critical heat flux of 35 MW/m 2 at ITER relevant thermal–hydraulic conditions.
On the basis of these results, ENEA–ANSALDO participated in the European program for the qualification
and manufacturing of the divertor IVT, according to the Fusion For Energy (F4E) specifications. A divertor
IVT prototype (400 mm total length) with three plasma facing component units (fig 3.1) was successfully tested
at ITER relevant thermal heat fluxes (20 MW/m 2 for 3000 cycles CFC part, 15 MW/m 2 for 3000 cycles W
part).
Now, this technology is ready to face the challenge of the ITER
IVT production, thus transferring the experience gained in the
development, optimization and qualification of the PBC and
HRP processes to an industrial production line.
Qualification of ultrasonic non–destructrive testing method
for plasma facing components
Under the “EURATOM Training Network for Plasma–facing
Materials” (ETN_PFM) (Contract No. 042314 (FU 06)), ENEA
was assigned with a trainee to be prepared in the area of
“Design, manufacturing and acceptance procedures of PFC”.
This area is strictly linked to the activities related to the
Figure 3.1 – ITER divertor IVT qualification
prototype manufactured by ENEA–Ansaldo

058
progress report
2010
development and manufacturing of plasma facing compontens
(PFCs) for the ITER tokamak and to the qualification of the
manufacturing technologies for the ITER divertor procurement.
Figure 3.2 – Plane CFC–Cu joint sample
Amplitude
40
30
20
10
0 0 0 20
C-scan 0
C-scan 1
C-scan
REPORT
100
0
-100
Nome del file
30 40 50 60
13.39 17.39 22.00
-19.28 43.75 0.00
120
80
40
0
DX mm
32.67
DY mm
26.36
TXT
OK
D:\Lavoro\ULTRASUONI\CAMPIONE01ULTRAN\ultime provecd
plane15.rf1
One of the main issues in the manufacturing of the plasma facing
units is the reliability of the non destructive controls that are
necessarily performed during the manufacturing process. ENEA has
developed a suitable ultrasonic technique (UT) for the control of all
the joining interfaces of the ITER divertor IVT plasma facing units,
but the defect detection capability of the method has to be proved for
both metal to metal and metal to CFC joints, since both types of
joints are present. Within this activity, the UT results coming from
the investigation being performed during the manufacturing, but also
after the thermal fatigue testing (up to 20 MW/m 2 ) of mock–ups
manufactured in ENEA labs by using the HRP technology were
studied and compared with the evidences coming from the final
destructive examination in order to qualify the method. Regarding
the Cu/CFC joint, the effectiveness of the ultrasonic test has been
deeply studied due to the high acoustic attenuation of CFC to
ultrasonic waves. For these purpose an ‘ad hoc’ plane Cu/CFC joint
sample, that reproduces the actual annular joint interfaces, was
manufactured. This plane sample has the advantage of being easily
tested by probes with different geometry and ultrasonic
characteristics. UT testing results were compared with x–ray and
Eddy current of the same sample.
The results confirmed that the evidences detected by UT can be
easily correlated to lack of adhesion of the copper to CFC; in fact,
the position of the defective zones coincides with the points where
the brazing alloy deposition did not succeed.
Figure 3.2 shows the CFC–Cu sample being used for the testing and
figure 3.3 compares the images obtained by the different techniques:
UT, x–ray, Eddy current.
C-scan
Height
0
10
20
30
0 50 100 150
Figure 3.3 – Images obtained by UT,
x–ray, and Eddy current
Analysis of the ITER divertor cassettes
ENEA completed the activities related to the performing of a new set
of 3D electromagnetic (EM) and mechanical analyses of the revised
design of the ITER divertor cassettes to the purpose of checking the
fulfillment of the requirements and better assessing the merits of
each envisaged alternative during several off–normal events,
including category II and III events. These activities, in the frame of
the European Fusion Development Agreement (EFDA) Contract 07–
1702/1596 (TW6–TVD–DIAGAN) [3.1], were performed with the
support of L.T. Calcoli, an Italian company specialized in
electromagnetic and structural finite element analysis.
The Divertor is one of the most challenging components of the ITER machine: it is designed to sustain the
heat load and reduce the impurity in the plasma: it consists of the PFCs and a massive structure called the
cassette body (CB) (fig. 3.4). The PFCs are actively cooled thermal shields required to sustain the heat and
particle fluxes during normal and transient operations as well as during disruption events. The CB is needed
for supporting the PFCs, routing the water coolant into them and providing neutron shielding: it is mounted
onto the vacuum vessel’s (VV) inboard and outboard rails via a mechanical locking system.
The performed 3D EM numerical analysis allowed the EM loads to be calculated, including integral forces
and moments on the PFCs, CBs and components (pipes, manifolds and multilinks in figure 3.5) due to halo
and Eddy currents by taking into account both the thermal and current quench. The EM loads thus obtained
were then transferred for the corresponding dynamic mechanical analysis that allowed the stresses and

technology programme (cont’d.)
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059
Inner vertical target
Outer vertical target
Dome
Figure 3.4 – ITER divertor
location and components
Cassette body
Figure 3.5 – Multilinks between
the CB and the PFCs
Figure 3.6 – Halo
current distribution
for the central halo
current path
73.387 0.216×10 07 0.432×10 07 0.648×10 07 0.864×10 07
0.108×10 07 0.324×10 07 0.540×10 07 0.756×10 07 0.972×10 07
displacements generated in the divertor structural parts to be
calculated and the stresses generated during off–normal events with
the ITER structural design criteria to be compared.
Several EM finite element models (FEMs) were developed to
analyze the effects due to the poloidal halo currents (fig. 3.6) and
the poloidal and toroidal field variation during both fast and slow
plasma vertical displacement events (VDE) which lead to category
II and III plasma disruptions. The Slow Downward VDE–III with
inward halo currents configuration has been selected, on the basis
of the resultant EM loads, as the worst condition to be used for the
next structural assessment. Both the standard cassette in front to the
cryo–pump port (fig. 3.7) and that between ports were modeled also
by taking in account the possible bridging on the plasma–facing
units (PFUs). These analyses showed that no significant difference
on the EM loads is produced by the port hole and that the bridged
configurations generate loads higher than the unbridged ones on
the CB.
Figure 3.7 – FEM of the standard cassette
in front of the cryo–pump port
The structural assessment required for the development of a more detailed FEM (fig. 3.8) in order to
perform proper static and dynamic analyses under the selected load conditions, obtained by combining the
EM forces with the additional loads due to dead weight, hydraulic pressure, initial preload of the CB and
thermal stress from normal operating conditions.
The elastic and elasto–plastic mechanical analyses allowed stresses, reaction forces, moments and
displacements produced in the divertor structural parts and attachments to be calculated, both for the
original and the updated geometry, with and without bridging of the PFUs. The structural integrity of the
divertor components and the compliance with the ITER structural design criteria was then assessed for the
updated Dome geometry both for category II and III loads.

060
progress report
2010
Figure 3.8 – Details of the structural FEM of the divertor system: CB with locking
system, dome and VT
3.2 Breeder Blanket and Fuel Cycle
European Breeding Blanket Test Facility design and construction
The European Breeding Blanket Test Facility (EBBTF) is the reference European facility for the qualification
of test blanket modules (TBMs). It is constituted by the Integrated European Lead Lithium LOop (IELLLO)
and the HElio for FUSion loop (HEFUS 3).
In 2009 the lead lithium to fill the loop and complete the commissioning was received in ENEA Brasimone.
The acceptance procedure took a long time due to large differences in the lithium content, and the procedure
was concluded in November. The loop was filled in February 2010. The new HEFUS 3 circulator was
delivered in ENEA Brasimone in the first weeks of 2010 and the installation completed in the first semester
of 2010.
TRIEX loop for studying technologies of tritium extraction from
Pb–17Li
The main objective to be achieved in the TRItium EXtraction (TRIEX)
facility is to verify the efficiency of the hydrogen extraction system from
the liquid metal loop in the range of operating conditions of the European
TBM (fig. 3.9).
In order to carry out the tests, in 2010 the mechanical pump was requalified
and inserted into the TRIEX loop. In April the pump was not
characterized in lead lithium due to the mechanical pump failure after 12h
of work. On the basis of this experience, an upgrading of the TRIEX loop
was planned and a new pump was selected. The new loop is expected to
start in November 2011.
Figure 3.9 – TRIEX loop
Electromagnetic load analysis for the design of the blanket manifold pipe
concept for ITER
ENEA completed the activities related to the 3D EM load analysis of the new manifold concept (NMC) for
the ITER blanket and shield system, according to the EFDA Contract 07–1702/1620 (TW6–TVB–MANEM)
[3.2]. The activities were performed with the support of L.T. Calcoli company.

technology programme (cont’d.)
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061
The ITER blanket and shield system is the
innermost part of the reactor; it is directly
exposed to the plasma and provides the main
thermal and nuclear shielding to the vacuum
vessel and external reactor components. Its
concept is based on a modular configuration with
blanket modules consisting of water–cooled
austenitic stainless steel shield blocks and
separable first–wall (FW) panels, mechanically
attached to the shield blocks. The blanket
modules have typical dimensions of
1m× 1.5 m × 0.5 m and are mechanically
attached to the VV. The water coolant is supplied
to the modules by a set of inlet and outlet
manifolds attached to the inner wall of the VV.
Figure 3.10 – Proposed design and detail of the FEM of the
inboard ITER shielding blanket NMC including the inner vessel
shell, the modules #6 and #7 and part of the single pipe
assembly
This work was performed to evaluate the eddy currents and the forces induced by fast variations of the
magnetic fields due to plasma off–normal events in the NMC where the original welded structure supplying
the cooling water to each module has been replaced by single pipes. This evaluation was needed to address
some concerns raised during the ITER design review about the capability of repairing the components
remotely and the high operating stresses and possible water leakages that might be produced at some locations
on the manifolds.
The FEMs developed for these analyses (fig. 3.10) include a 10° toroidal sector of the ITER machine with the
double shell vessel, the divertor, the blanket modules and a quite detailed model of the pipe manifold where
each pipe is insulated from the other pipes, the blanket and the vessel except in the points where the pipes are
bundled together and attached to the vessel.
The 3D EM load analyses, performed by using the ANSYS code, allowed the time evolution of the current
and bending force to be evaluated per unit length in the whole bundle as well as the current sharing between
the whole pipe bundle and the vessel, for several plasma off–normal events. The major disruption type II with
36 ms linear current quench was proved to be the most severe event as far as the force per unit length on a
single pipe, even if the total force on the manifold is larger during the thermal quench, because the eddy
current is not shared among the pipes but is concentrated in the first part of a single pipe at each manifold
section. The EM loads induced on the NMC are lower as compared to the reference concept, but this
advantage has to be confirmed in terms of mechanical stresses due to the significant differences among the
mechanical structures. In any case, the large resulting EM loads must be carefully considered in the detailed
design of the manifold components and their supporting systems.
Development of method for highly tritiated water handling in ITER tritium plant
A process based on a combination of permeator catalyst (PERMCAT) (Pd–based membrane reactor) and
vapor phase catalytic exchange (VPCE) has been studied for processing higly tritiated water (HTW) (ITER
contract ITER–CT–09–4300000087). The simulation of the PERMCAT and VPCE systems has been carried
out and the process flow diagram has been prepared [3.3]. As a main advantage, the process being proposed
permits to combine the PERMCAT and VPCE in several ways accordingly to the characteristics of the
different HTW streams to be processed. Furthermore, both PERMCAT and VPCE can use the mixture of
hydrogen isotopes produced by the electrolyzers of the water detritiation system as sweep gas, thus reducing
the impact on the isotopic separation system and avoiding the generation of secondary wastes.
Training activities
ENEA Frascati laboratories have had in charge a Trainee (EURATOM Research Training Network
"Preparing the ITER Fuel Cycle" – Contract No. 042862 (FU 06)) to be prepared in the area of the fuel cycle
of ITER. All the activities have been completed by achieving the milestones of the task: study of cold–rolling
and diffusion welding techniques, long–term testing of Pd/Ag membranes, participation to the CAPER R&D
program, tritium confinement study [3.4], analysis of tritium release from the neutral beam injector [3.5] and
inactive tests for new PERMCAT prototypes [3.6].

062
progress report
2010
ENEA Frascati laboratories have also had in charge an Early Stage Researcher (EFDA Goal Oriented Training
Programme “Tritium Technologies for the Fusion Fuel Cycle” – TRI–TOFFY) to be prepared in the
deuterium–tritium fuel cycle area for ITER. Main research activities have consisted in characterizing thin wall
Pd–based permeator tubes in terms of hydrogen permeability and selectivity and the hazard and operability
study (HAZOP) for the HTW process system of ITER. For the treatment of HTW three processes have been
studied in details [3.7]: water decomposition by using the water gas shift reaction, high temperature electrolysis
and water splitting through reduction on metals. Particularly, the use of Pd–based membrane reactor for
carrying out the water gas shift reaction of tritiated water permits high reaction yields to be reached and pure
hydrogen isotopes to be recovered, which can be directly sent to the isotopic separation system.
3.3 Magnet and Power Supply
Optimization of the toroidal field ripple reduction system
The ITER toroidal field coil (TFC) system is made of 18 D–shaped coils spaced by 20° in toroidal angle. The
discontinuity produces a deviation (ripple) from the toroidal direction of the magnetic flux surfaces that can
cause significant losses in the confinement of high energy particles (α–particles or high–energy ions from
neutral beam injectors) due to their trapping inside the “ripple valleys” and unwanted peaking in the heat loads
on the FW. Due to these reasons, an accurate evaluation of the toroidal field ripple (TFR) was performed in
various operation conditions. This evaluation was carried out by means of finite element models built only by
using structured meshes in order to obtain a very high field precision on a regular spaced grid extended to the
entire region enclosed in the FW. The model was built by taking into account the real 3–D shape of the TFC
and modelling three nested D shaped coils capable of carefully reproducing the real geometry of the TFC.
The analysis showed a high value of TFR (exceeding 1% in the outboard plasma region near the equatorial
plane) and this confirmed the need for introducing some correcting elements. The implementation of
ferromagnetic SS430 steel inserts in the outboard region between the inner and outer vessel shells, properly
optimized in shape, size and location, then allowed the maximum ripple at the plasma separatrix to be reduced
to 0.19% (fig. 3.11); this value could increase up to 0.38% when the number of inserts was limited by the filling
factor required for ITER design.
The analysis also confirmed that the introduction of ferromagnetic inserts into the equatorial region between
equatorial ports is essential to reduce the TFR to acceptable levels in the plasma region. The optimization of
the ferromagnetic inserts was performed by taking care of limiting the ripple over–compensation under 0.6%
during plasma operation at half toroidal field.
0.268×10 -3 0.736×10 -3 0.001205 0.001674 0.002142
0.502×10 -3 0.001908
SMN=0.357×10 -5
0.971×10 -3 0.001439
Optimized insert plate
SMX=0.010749
distributions each plate is 4.4 cm
thick with a filling factor of 4.4
0.357×10 -5
0.001198
Ripple in the plasma
region with optimized
insert distribution around
ports without NBI
Enlarged
scale
0.002392 0.004779 0.007167 0.009555
0.003585 0.005973 0.008361
Maximum ripple
at the separatrix
0.188%
Figure 3.11 – ITER ripple map at full toroidal field with the optimized distribution of
the inserts
Equatorial port position
Three–dimensional
magneto–static analyses
for ITER
ENEA completed the
activities related to several
magneto–static analyses
in ITER according to the
EFDA Study Contract
07–1702/1602
(TW6–TPO–3DMAGS):
the optimization of the
shape, size and location of
the ferromagnetic plates
used for the reduction of
the TFR; the evaluation
of the Maxwell’s forces on
these plates and the
analysis of the effects on
the ripple due to the
ferromagnetic materials
in the magnetic shields of

technology programme (cont’d.)
progress report
2010
063
the neutral beam injection (NBI) and the TBMs; the optimization of the NBI magnetic field reduction system
in reducing the residual field to the requested level for operation; the evaluation of the effects of ferromagnetic
material present in the building concrete surrounding the torus on the stray field in the NBIs system, over the
plasma region and outside the bio–shield structure. These analyses were performed with the support of L.T.
Calcoli by developing in ANSYS several 3D magneto–static FEMs by using a parametric approach which
allows an easy change of the models geometric main features to match different design options.
Analysis of the ripple effects due to the ferromagnetic materials in the NBI magnetic shields and the test
blanket modules
Some ferromagnetic components foreseen in the ITER design are not uniformly spaced along the toroidal
angle and can then dramatically increase the ripple in the region where they are placed. To evaluate this
increase, two TBMs were modelled as blocks of EUROFER ferromagnetic material (2700 kg each one) and
placed in two ITER equatorial ports: their complete saturation, at about 1.9 T during ITER operation,
induced a toroidal field perturbation at the plasma separatrix up to 1.0%. Moreover, the equatorial ports
devoted to NBIs cannot allocate the same ferromagnetic inserts shape foreseen for standard ports, due to their
different geometry: properly optimized in shape and thickness inserts are required around these ports and
induce a further perturbation of the toroidal ripple map.
Detailed TFR maps of the plasma region were produced (fig. 3.12) by considering the real geometry at the
operating temperature, with a relative precision in the error field better than 1%.
Assessment of the NBI magnetic field reduction system
The residual magnetic field inside the NBI region is required to be very low in order to avoid the deflection of
the ion beam before neutralization, and to achieve the requested performances. In ITER, both passive and
active shielding systems are foreseen to the purpose of reducing by several order of magnitude the stray field
inside the NBI region due to the plasma, the currents in the magnetic field coils and all the magnetized
components around it.
A detailed FEM of the whole NBI system was developed (fig. 3.13) by using a fully parametric approach to
allow an easy change of the iron shielding thickness (15 cm at the present), coil shapes and NBI shielding box
geometry during the optimization. The effects on the stray field of the ferromagnetic materials in the building
were roughly reproduced in the model by adequately rearranging the currents in the PF coils and the plasma.
TBM
region
TBM+new optimezed Fe
inserts isosurfaces of dBtor
40° sector of plasma region
A=-0.016903 C=-0.008637 E=-0.371×10 -3 G=-0.007896 I=-0.016162
B=-0.01277 D=-0.004504 F=-0.003763 H=-0.012029
Figure 3.12 – TFR isosurfaces on the plasma separatrix region for a
40° sector with TBM and optimized ferromagnetic inserts
Figure 3.13 – Detailed model used in the
optimization of the ITER NBI magnetic field
reduction system

064
progress report
2010
Field (T)
6.69
4.63
3.91
2.58
Field (T)
5.15
3.96
2.76
1.56
×10 -3 RID
Neuralizer
Gap
Beam source
Figure 3.14 – Orthogonal to the beam
residual magnetic field components at end
of burning time and relating constraints
along the ion beam path for the optimized
NBI magnetic field reduction system
×10 -3 26.2
1.21
By
0.37
Bz
-0.15
19.2 21.9 24.7 27.5 30.3
Distance from yz plate (m)
-0.82
21.7 23.9 28.5 30.8
Distance from yz plate (m)
View from the top
External walls
Columns outside bioshield
Iron doors
NBI boxes
Bioshield
2° cylindrical wall
NBI floor
PFC TFC and AMFRS coils
The magneto–static analysis was carried out by
using a mixed approach which adopts the
Biot–Savart integration for the plasma, PF coils and
active coils circuital elements and employs a FEM
approach for the detailed model of the passive
ferromagnetic shielding plates. The performed
optimization of the active and passive magnetic field
reduction system allowed the residual magnetic field
along the beam path below the required level for
operation to be reduced (fig. 3.14).
Figure 3.15 – EM
FEM of the ITER
building complex
Figure 3.16 – Magnetic field contour at EOB inside the ITER building at
a vertical quota equal to the NBI axis level
Assessment of the effects of the ferromagnetic
material in the ITER building structures
The stray magnetic field due to the ferromagnetic
materials in the building surrounding the torus was
evaluated inside the NBI system, the
plasma region and outside the
bio–shield structure by developing a full
360° 3D electromagnetic model of the
building complex (EMOBC) (fig. 3.15),
including the PF coils, the plasma, the
two iron boxes and the active coils of the
NBI magnetic field reduction system,
the main building components with
ferromagnetic materials (iron doors at
the end of the port corridors, rebars in
the building walls, floors, roof, basemat
and seismic pit concrete).
The analysis allowed the currents in the
NBI active coils to be adjusted at several
plasma scenario times, taking into
account the effects due to the other NBI
and the ferromagnetic materials all
around. Moreover, the stray field due to
the ferromagnetic content outside the
vessel was calculated and the field
perturbation produced on the plasma
q=2 surface was evaluated with great accuracy in order to estimate the perturbation on the magnetic field from
the axial symmetry that might produce unwanted plasma modes locking. Indeed, the evaluation of the very

technology programme (cont’d.)
progress report
2010
065
low (some gauss) stray field produced by the magnetization induced in the ferromagnetic materials on the
plasma region required a careful analysis due to the presence of the high poloidal field produced by the plasma
itself and by the PF coils that could not be excluded from the analysis itself.
The stray field inside the whole main ITER building was also evaluated for several plasma scenario times
(fig. 3.16) and the maximum magnetization of all the main ferromagnetic components was calculated (1.6 T
for the NBI shielding boxes, 0.55 T for the iron doors, 0.5 T/m 3 for the homogenised material including the
rebars and the concrete of the building).
Pre–compression rings final design qualification
The ITER pre–compression rings activities continued in ENEA Frascati with the fifth ultimate tensile strength
(UTS) test performed on a ring scaled mock–up with a diameter of 1 meter (fig. 3.17) (ITER Contract
09–4300000015) [3.8,3.9]. This was the latter of six UTS tests on six different mock–ups. Four rings were
manufactured with the vacuum pressure impregnation (VPI) technique developed and optimized in ENEA
while the other two were produced by the filament wet winding
(WW) industrial conventional process.
The mock–ups were dimensionally checked and x–rays surveyed
before tests. Then UTS tests were carried out by loading the rings
with radial displacement increments of 0.1 mm by means of the
ring hydraulic testing facility in ENEA Frascati and following as
close as possible the standard test method ASTM D3039 for
tensile properties of polymer matrix composite materials.
UTS tests showed an average strength of 1550 MPa (mean hoop
stress in the cross section) and constant tensile modulus of
elasticity up to failure. UTS obtained on the VPI rings was
1584 MPa, higher than UTS on the WW rings, 1485 MPa.
Figure 3.17 – Pre–compression ring
dismantling after test
The volumetric glass content of the rings was measured on some of the rings after test, resulting in an average
of 70% both for VPI and WW rings.
The testing activity continued with a stress relaxation test performed on a VPI ring mock–up for 210 days at
a stress level of 950 MPa. The mock–up showed a stress relaxation of less than 0.5% and a residual strain of
0.02% after test. No defects were detected by x–rays after test.
Another WW ring was then manufactured and will be stress relaxation tested during 2011. Characterization
of the ring composite material has been completed with creep, shear and compression tests.
On the basis of the ENEA R&D activity, at the end of 2010 F4E launched a call for tender for the
procurement of the ITER full scale pre-compression rings where ENEA will be involved to qualify preliminary
scaled rings.
3.4 Remote Handling and Metrology
ITER in vessel viewing system
ENEA developed and tested a prototype of a laser in vessel viewing and ranging system (IVVS), that uses the
amplitude modulated laser radar concept and is based on an intrinsically radiation resistant concept and
architecture to withstand the severe ITER conditions. It already approaches the target specification requested
for ITER, although its present layout is not capable to withstand all the ITER environmental conditions.
In late 2008, ENEA won a grant launched by F4E for the conceptual design of the final ITER in–vessel
inspection prototype and the assessment of the present IVVS prototype. The activity started in April 2009 and
continued in 2010 to evaluate the potential application of the ENEA IVVS prototype for ITER in–vessel
inspection and then produce the conceptual design of an IVVS system compliant with all the ITER
requirements and the related test bed. The work has been divided into three main tasks.

066
progress report
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• The execution of laboratory tests with the ENEA IVVS mock–up in order to verify that the performance
of the mock–up are matching (or at which level are approaching) those required for the final application in
ITER
• The assessment of the expected viewing/metrology capability of the IVVS in the ITER conditions,
• The conceptual designs of an IVVS probe prototype and related test bed, in preparation of the future
procurement and testing activities. They take into account all the very severe environmental conditions of
ITER during the inspection.
The 2010 year was mainly devoted to the activities described in the following.
Vibration effects on IVVS images and vibration correction
method. As the IVVS probe will be installed on a dedicate
deployer to be inserted in ITER, vibrations may arise in the
probe due to the tilt movement of the scanning prism. A
dedicate test campaign was carried out to evaluate how
viewing and metrology are affected by probe vibration at low
frequency (f

technology programme (cont’d.)
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067
The mechanical design of the IVVS probe was also revised and finalized
to the ITER environmental requirements. It has been improved with a
different actuating system for the scanning prism, a new step focus
actuating system and a revision of materials and components due to the
stringent ITER requirements. The present gearing chain designed and
realized to transmit the tilt and the pan movements to the prism by
means of step motors has be completely removed in the new design and,
in order to meet the high magnetic field requirement, it has been
replaced with rotation actuators driven by a couple of ultrasonic
piezo–motors (see fig. 3.22). The softness of movement reachable with
such motors is expected to sidestep the vibrations due to the scanning
prism movement. Commercial ultrasonic piezo–motors are compact,
maintenance–free, intrinsically non–magnetic and vacuum–compatible
but they must be customized to work at 120 °C and to be backed at
200–240 °C. Linear piezo–motors will be also adopted to operate the
new step focus system.
The dimensional stability of the mechanical structure will be obtained
by using antimagnetic steel alloys such as Invar or Kovar, which have a
thermal expansion close to fused silica, which is the material selected to manufacture prism and lenses.
Ceramic materials will also be used for particular components, such us bearings and collars.
Assessment of the expected viewing/metrology capability of the IVVS in the ITER. In order to realize an
assessing tool, a 3D FEM based software has been realized to satisfy the following requirements:
• to import from F4E “CAD” data of ITER in vessel surfaces
• to originate all the positions and orientations of the IVVS heads
• to find the ITER in vessel surfaces elements visible from the scanning heads
• to distinguish them from the elements hidden to a scanning head because of its position and orientation
• to numerically implement the quality functions describing the IVVS performance
• to find the quality of each element of ITER in vessel surfaces
• to produce quality maps with expected IVVS viewing and metrology capabilities
• to optimize number and position of scanning heads necessary to best cover the whole inner surface of the
ITER vessel.
A contract has been signed with Enginsoft Spa to develop this software, which was delivered to F4E the last
September 23th.
The beta release of this software had been tested by ENEA in Frascati,
and produced quality maps of ITER in vessel surfaces.
In the meantime, ENEA had developed an optimization strategy, whose
specification was emitted in the first half of july.
The beta release of this software had been installed in F4E Barcelona on
August the 16th, after which a training on job was provided by ENEA
Frascati scientists, the first week in F4E Barcelona, the following ones at
the Frascati Research Center.
At the end of the debug activities, the final release of this software took
place.
Since then, this software is being employed by ENEA in conjunction with
F4E to the purpose of producing quality maps (see fig. 3.23) and
optimizing the IVVS prototype and its placement in ITER.
Figure 3.22 – Mechanical design of
the new IVVS scanning head
Figure 3.23 – Example of quality map showing the standard deviation
of the IVVS range measurement on the ITER geometry for a given head
position

068
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Figure 3.24 – Picture of the ITER Mock–up
assembled at FNG
Front casing
Front insulation
1st winding
2nd winding
Side and rear
casing
1st layer SB
1981 1982
Back of SB
2041 2047
CuCrZr
Rest of coil
Thermal shield IVVS
Side and rear insulation VV outer shell
VV inner shell
3rd layer
FW
Manifolds
First wall
Figure 3.25 – The radial section on the inboard
side of the latest ITER Alite model
Figure 3.26 – MCNP model (equatorial section
(z=0))
3.5 Neutronics
Neutronics shielding experiment on a mock–up of ITER:
dose measurement in the magnet coils
In the ITER design it is important to minimize the
uncertainty in the estimates of the nuclear loads; in
particular, the nuclear heating of the TF coils in the
inboard leg is the most critical. To this purpose, accurate
radiation transport calculations of the shielding is
requested. These calculations are very challenging, since the
radiation attenuation from the first wall to the TF coils can
be many orders of magnitude and, at the same time,
accuracy of the order of ±10% or better is required. The
calculation must be benchmarked as far as possible against
suitable experiments to attain the necessary validation of
nuclear data and codes used.
To check whether the present design calculations are able to
evaluate the shielding properties of the ITER shield at
inboard side with sufficient accuracy and reliability, an
experiment was realized at the Frascati Neutron Generator
(FNG) of ENEA–Frascati. The experiment was
commissioned by ITER IO.
In this experiment, a mock–up of ITER inboard shield,
vacuum vessel and TF coils was replicated and irradiated by
14–MeV neutrons (fig. 3.24). The mock–up also included
the borated steel plates presently foreseen by the design as
well as some pieces of the actual superconducting cables,
which represented the actual experimental region (TF coils).
The final mock–up dimensions and materials compositions
were based upon the dimensions and materials of the latest
version of the Alite Monte Carlo n–Particle (MCNP) model
of ITER (fig. 3.25).
The resulting nuclear heating in the TF coils was measured
by using state–of–the–art experimental techniques (high
sensitivity thermoluminescent dosimeters), and compared
with calculations performed with the MCNP code and the
ITER reference nuclear data library FENDL.–2.1. The
experiment also included the measurement of selected
reaction rates along the central axis of the mock–up as well
as in the coil region. These measured quantities were
compared with the results of calculations too.
A very detailed model of the experimental set–up was used (fig. 3.26). In this way, the highest level of accuracy
was attained in the ratio between the calculated and the experimental quantities (C/E ratio), thus providing
fundamental information about the dose absorbed by the superconducting inner coils. For example, for the
nuclear heating a slight overestimation is observed within the C/E error (±10%) in the coil region. This
accuracy in the C/E ratio was never reached in previous experiments.
Moreover, regarding the uncertainty margins in the FENDL–2.1/MCNP–5 prediction, it can be concluded
that:
• the fast neutron flux is calculated within an uncertainty margin of about ±15% in the ITER shielding
blanket and the magnet region.
• the thermal neutron flux is calculated within an uncertainty margin of about ±15% in the ITER shielding
blanket, vacuum vessel, up to the toroidal field coils.

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069
• the nuclear heating is calculated within an uncertainty margin of about ±10% in the spatial region of the
ITER shielding blanket and vacuum vessel, and within an uncertainty margin of about ±10% in the region
of the toroidal field coil with a total uncertainty on the C/E ratio which is always

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15 cm radius); b) neutron and gamma spectra at the
detector location; c) analysis of the components of the
collided neutron spectrum.
The estimation of neutron spectra and
collided–to–uncollided ratio(C/U) is useful to evaluate
the neutronic performance of the HRNS collimator in
both configurations (i.e. variation of the background with
the distance inside the cavity) and to characterize the
radiation field inside it. Figure 3.29 shows the C/U values
for 15 cm and 5 cm radius: the latter option is more
effective in the collimation, especially in the front regions.
Neutron spectra for 15 cm and 5 cm radii at the detector
position are shown in figure 3.30.
Figure 3.28 – The HRNS inside the port plug: in pale
blue the concrete collar
Collided/uncollided
1.6
1.2
0.8
0.4
0.0
800
5 cm radius
15 cm radius
1200
1600
Distance from the torus axis (cm)
Figure 3.29 – Collided to uncollided ratio at different
positions along the HRNS collimator: 15 cm (red) and
5 cm (black) radius
Neutron flux (n cm2/MeV/source n)
10 -10
10 -12
10 -14
0 10 20
Energy (MeV)
Figure 3.30 – Neutron spectra at the detector position
D for 15 cm (dotted blue) and 5 cm (straight black)
radius
In order to fully characterize the radiation field at the
detector position D, photon flux and spectra have been
evaluated as well. The absolute values of the gamma flux
are: 5.32×10 8 γ/cm 2 /s and 3.29×10 9 γ/cm 2 /s for 5 cm
and 15 cm radius respectively .
The collided spectrum at the detector position has been
analyzed in order to isolate the contribution due to
different zones of the torus (inboard and outboard zones).
The results obtained show that a large contribution to the
collided neutron spectrum is due to the particles scattered
only by the inboard side, which penetrate into the
collimator, thus reaching the detector position. These
results confirm the effectiveness of the port plug shielding
capability [3.11].
Three–dimensional neutronic analysis of the ITER
in–vessel coils
In the frame of the contract ITER/CT/09/4100001120
a complete neutronic analysis has been performed for the
design of the in–vessel coil systems by using the MCNP5
code in a full 3–D geometry. A detailed geometry of edge
localised mode (ELM) and vertical stabilizing (VS) coils
based on the last design specifications has been integrated
into the last version of 40° ITER Alite MCNP model
(fig. 3.31).
cm
800
400
0
VS upper
Upper ELM
Central
ELM
Lower
ELM
-400
VS lower
Figure 3.31 – Vertical section of Alite–4
MCNP model with In–vessel coils
-800
-800
-400
0
cm
400 800

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071
In a first stage, the blanket/manifold original description
was not modified: the coils cut across the interposing
structures. In a second stage, manifolds have been
simulated in front of poloidal coils located behind gaps
(fig. 3.32). Poloidal, radial and toroidal variations of all
relevant nuclear parameters are provided, as well as
detailed nuclear heating tables useful for thermal analysis.
In the original configuration (without front manifolds) the
total nuclear power deposited on ELM coils is ∼3 MW. The
peak nuclear parameters obtained on the conductor are:
nuclear heating 1.7 W/cm 3 and damage 0.4 dpa.
Concerning the insulator: maximum cumulative dose is
4210 MGy, dose rate 211 Gy/s and neutron fast fluence
3.45×10 20 n/cm 2 . For stainless steel components, the
He–production peak is 6.5 appm.
Nuclear parameters show a great spatial variation. Peak
values are found in limited zone close to the blanket gaps:
figure 3.33 shows the toroidal profile of the nuclear heating
in the upper toroidal ELM front components. The peak
corresponds to the poloidal gap. An increase of about 50%
is obtained in this zone with respect to the parts of the coils
far from the gap.
Figure 3.34 shows the radial profile of the nuclear heating
in CuCrZr in the toroidal coils of lower ELM coils and
connecting poloidal segments with and without a front
manifold. Without a front manifold, the ratio between
nuclear heating values in front and rear coils is about 0.3.
By comparing top and bottom toroidal results, the nuclear
heating in the bottom coils shielded by blanket modules is
about 50% of that in the top. With a front manifold, the
radial profile exhibits a steeper variation: the ratio between
rear and front coil nuclear heating values drops to 0.2. The
increase due to the manifold is about 60% due to the
presence of water and large void space. The increase in the
other quantities is lower than that in nuclear heating except
for Helium production in the zones behind the manifold.
Regarding the impact on the vacuum vessel reweldability,
the He–production value does not exceed that of the
original configuration (max He–production 0.6 appm).
With heterogeneous manifold located in front of the
poloidal coils, the He–production exceeds the limit but
only in zones where the rewelding is not foreseen.
Comparing the nuclear loads on CuCrZr and C10700 for
the conductor and of spinel and MgO for the insulator, no
significant variations are found. Hence, the neutronic loads
on these materials can be considered as non relevant issues
for selecting these components.
Activation analysis has been carried–out with FISPACT
2007 by using the neutron spectra calculated in 3–D with
MCNP5 and the safety scenario SA2. Total neutron fluxes
vary from 7.3×10 12 n/cm 2 /s in the more shielded segment
to 5.1×10 13 n/cm 2 /s on the nose of the connecting coils in
the gap between blanket modules. Figure 3.35 shows the
dose rate versus time after irradiation in the front part of
the connecting segment. At the shutdown, peak specific
Poloidal segment of ELM coils
Figure 3.32 – Radial section of MNCP model of ITER.
Poloidal ELM coils in the configuration a) without
front manifold and b) with manifolds
Nuclear heting (W/cm 3 )
1.6
1.2
0.8
0.4
0.0
SS support-front
SS envelope
MgO insulator
Water
CuCrZr
conductor
0 30 60 90 120
Toroidal position (cm)
Figure 3.33 – Nuclear heating toroidal profile of the
front coil and front support of upper ELM toroidal
top coils
Nuclear heating (W/cm 3 )
2
1
0
4
Tor top lower ELM
Tor bottom lower ELM
Pol no front manifold
Pol + front manifold
8 12 16 20
Distance from the VV (cm)
Figure 3.34 – Nuclear heating on CuCrZr radial
profile in lower ELM coils and poloidal connecting
segments with and without front manifold
Contact dose rate (sv/h)
10 4 CuCrZr
C10700
MgO
10 2
Spinel
SS
10
10 -2
10 -4
10 -6 Min Hour Days Month Year
10 -8 10 -4 10
10 2
Time after irradiation (years)
Figure 3.35 – Peak contact dose rate versus time
after irradiation in the most irradiated coil

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activity varies from 5.25×10 12 Bq/kg in stainless steel to 5.2×10 13 Bq/kg in C10700, maximum dose rate and
decay heat are 1.2×10 4 Sv/h and 9.6 W/kg in Spinel. After 50 years from the shutdown peak specific activity,
dose rate and decay heat are 7.6×10 9 Bq/kg, 102 mSv/h and 0.08 mW/kg respectively in CuCrZr. At medium
cooling times the activation of CuCrZr is higher than of that C10700 (the maximum ratio is 2) because of
Cobalt. Except at very short cooling times, the insulator activation is lower than that of conductor and SS
components and is dominated by the impurities. Copper resistivity increase at the shutdown varies in the range
33.6–276.7 pΩm (resistivity of pure copper is 17 nΩm, at T=20°C) due to transmutation in Ni, Co and Zn.
In the more shielded segments, the activation and transmutation are about one order of magnitude lower than
the peak. The impact of the front manifold on the activation of the poloidal components is within a factor 2
and depends on the material and cooling time. The complete inventory is provided to perform the safety
analysis and waste classification [3.12, 3.13].
Neutron and gamma spectra behind ITER blanket modules and in divertor
The installation of sensors behind ITER blanket modules and in the divertor region is under study. For this
reason, knowledge of the radiation field (i.e. neutron and secondary gamma spectra) is required. For this scope
3–D radiation transport simulations have been performed using MCNP5 code with FENDL2.1.
Z axis
6
4
2
-1
-4
Profile
Neutron and gamma fluxes and spectra have been calculated in
forty–five positions (fig. 3.36) assuming neutron production
from a 500 MW DT plasma. The positions of the detectors in
outboard regions have been varied toroidally to evaluate the
impact on the radiation field of the port plug and the shielding
capability of the blanket modules.
In the divertor region, the maximum total neutron flux is found
in the dome zone (5.5×10 13 n/cm 2 /s) and the minimum one is
found at the bottom of the outer vertical target. The gamma
flux varies between 5.7×10 12 γ/cm 2 /s and 3.6×10 13 γ/cm 2 /s.
Behind blanket modules, maximum fluxes are calculated in the
gaps between the central outboard modules (7.9×10 12 n/cm 2 /s
and 4.6×10 12 γ/cm 2 /s). The fluxes drop of one order of
magnitude at the top of the machine.
The effect of the difference between port plug and blanket
module is moderate. In fact, in the equatorial port zone the
toroidal variation is ≤12% on the neutron flux and ∼25% on the
gamma flux. The effect is higher in the upper port zone (within
a factor 2).
The poloidal behaviour of the neutron flux follows the neutron
R axis
current profile except in the regions close to the equatorial
Figure 3.36 – Positions of the detectors midplane. This is mainly due to the variability of blanket’s
shielding thickness. The radial profile of the flux is very
sensitive to blanket front attenuation: by moving the detectors
located close to the midplane 8 cm toward the plasmas, the neutron flux doubles as compared to the reference
position [3.14].
Neutronic analysis of ITER divertor rails
Nuclear heating for a 500 MW DT plasma, helium production and dpa at end–of–life have been calculated
in several positions of the divertor rails and surrounding components by using the MCNP5 code in a three
dimensional geometry. Maps in two poloidal inner and in five poloidal sections are provided.
Higher values of nuclear heating, He production and dpa are found in the inner zone, especially on the top
corner of inner cover plate because of the streaming through the gaps between cassettes, values are shown in
figure 3.37. In the most irradiated outboard zone the obtained values are lower by one order of magnitude as
compared to the inboard quantities.

technology programme (cont’d.)
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073
Nuclear heating
(kW/m 3 )
Bolt 63.1
Spring
washer
73.3
Insert 54.4
Bolt 27.1
Spring
washer
34.9
Insert 19.8
He Production at EOL
(appmax10 -2 )
Bolt 43.3
Spring
washer
48.3
Insert 6.9
Bolt 17.3
Spring
washer
21.9
Insert 4.5
Dpa at EOL (dpa x10 -2 )
Bolt 15.3
Spring
washer
17.0
Insert 16.1
Bolt 7.1
Spring
washer
8.6
Insert 6.5
Figure 3.37 – Nuclear heating, helium production and dpa in inner rail close to the cassette gap at end–of–life (EOL). Generic
scoring cells are spheres. Bolt cells are cylinders laying along radial direction
Nuclear responses in the bolts, inserts, spring washers
and shear keys have been calculated in some
representative positions. To estimate the expected values
in other zones, conversion factors taking into account the
different materials’ responses are provided as well [3.15].
Development of a high resolution compact neutron
spectrometry using diamond detectors
An EFDA contract was assigned to ENEA–Frascati to
7.5
develop a compact HRNS using diamond detectors. The 0 4 8 12 16
scope of the task is to demonstrate the capability of
Deposited energy (MeV)
artificial diamond detectors to operate as neutron
spectrometers with high resolution in the neutron energy Figure 3.38 – Experimental response functions
range 5.7–20 MeV. The diamond detector response
functions are necessary in order to characterize the diamond detector and to archieve the best possible energy
resolution and the widest neutron spectroscopy energy range. The response functions obtained for the
diamond detector will be used to unfold the line integrated pulse height spectra measured along the various
lines of sight of the ITER RNC by using the technique of a double reconstruction, in space and energy.
A single crystal diamond detector was exposed to quasi mono–energetic neutron fields in the energy range
from 5 MeV to 20.5 MeV produced by the Van de Graaff neutron generator of the EC–JRC–IRMM.
Response functions were measured for the first time for neutron energies above 14 MeV (fig. 3.38). The pulse
height spectra showed sharp peaks at specific energies. These peaks result from neutron induced charged
particle reactions occurring in carbon at different neutron energies. The centroid of the peaks due to the
12 C(n,α) 9 Be reaction permitted, by means of the reaction Q–value (–5.7 MeV), to point out an excellent
linearity of the detector response versus neutron energy and an extremely good intrinsic energy resolution of
about 56 keV at full width at half–maximum.
16.5
10 -4
10 -6
10 -8
20.5
Normalized
response
Neutron energy
(MeV)
Measurement of Ti profiles with the ITER radial neutron camera
The study of the capabilities of the ITER RNC as an ion temperature (Ti) profile diagnostic has been
completed (EFDA task WP08–DIAG–01–06). Two measurement approaches have been investigated under the
assumption of a) purely thermal plasma and b) use of liquid scintillator detectors. The first approach relies on
the exploitation of the spectrometric capabilities of liquid scintillator detectors through the unfolding of line
integrated-spectra. Results suggest that, under the above assumptions, the RNC can indeed work as a
multichannel neutron spectrometer [3.16] and provide Ti profile measurements with accuracy and time
resolution within the ITER requirements. An example of Ti reconstruction in an ITER plasma scenario (#4)
characterized by an internal transport barrier is given in figure 3.39: Ti (shown as a function of the magnetic
flux coordinate ψ) is reconstructed within 10% accuracy even in presence of a strong Ti gradient discontinuity.
The second approach is based on the use of the measurements of the integrated flux and other plasma
parameters. In this case it is not possible to match the measurement requirements, mainly because of the

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T(keV)
30
20
Δt=1 s
a)
Original
Reconstructed
%
20
10
Δt=1 s
Precision
Accuracy
b)
Figure 3.39 – Ion temperature
profile reconstruction using
unfolding of RNC line–
integrated pulse height spectra:
a) comparison between original
and reconstructed Ti profiles; b)
accuracy and precision in the
reconstruction
10
0
0
0.4
ψ
0.8
1
0
0
0.4
ψ
0.8
1
T(keV)
20
10
Original
Reconstructed
a)
%
20
10
Precision
Accuracy
b)
Figure 3.40 – Ion temperature
profile reconstruction using
RNC integrated flux measurements
and other plasma
parameters’ measurements: a)
comparison between original
and reconstructed Ti; b)
accuracy and precision in the
reconstruction
0
0
0.4
ψ
0.8
1
0
0
0.4 ψ 0.8 1
0.2
0.1
a)
strong variability of the neutron reactivity with temperature, and
because of the accuracy foreseen for density, effective charge and
impurity density profiles measurements. An example of the Ti
reconstruction is given in figure 3.40.
0.0
0.0 0.4 0.8
r/a
0.4
0.2
0.0
0.0 0.4 0.8
r/a
Figure 3.41 – a) Average reconstructed
fuel ratio profile (including error bars); b)
precision (dashed) and accuracy (solid) of
the reconstruction
b)
Measurement of fuel ratio profiles with the ITER radial neutron
camera
The study of the capabilities of the ITER RNC equipped with
liquid scintillator detectors as a diagnostic for the fuel ratio (n T
/n D
,
ratio of the tritium to deuterium density) is being investigated in the
frame of an EFDA task (WP10–DIA–01–03). To diagnose the
n T
/n D
profile, the RNC should be able to provide simultaneously
DD (2.5 MeV) & DT (14 MeV) neutron emissivity profiles
measurements and a measurement of the ion temperature profile;
the success of the measurement strongly relies on the background
due to scattered 14 MeV neutrons that, depending on its
magnitude, may preclude the measurement of the DD spectral
component. The following measurement procedure is proposed
and applied to ITER scenario 2: 1) unfolding of RNC
line–integrated spectra to recover separated DD and DT brightness
components. 2) Spatial inversion of DD and DT brightness signals
to determine separate DD and DT emissivites. 3) Determination of the ion temperature profile. Montecarlo
calculations (using MCNP) have been performed for a representative subset of the 45 RNC lines of sight, in
order to characterize the background due to 14 MeV scattered neutrons at the RNC detectors' position. The
calculations have been carried out by including the RNC in the latest MCNP 40° ITER model (Alite–4). The
results indicate that the RNC might measure flat n T
/n D
profiles with values between 0.01 and 0.1 (with 20%
accuracy and precision and 100 ms time resolution) up to r/a < ∼0.8. An example is shown in figure 3.41 for

technology programme (cont’d.)
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n T
/n D
=0.1. The maximum measurable n T
/n D
can be extended up to ∼0.2 for r/a < ∼0.4. The possibility to
measure non–flat n T
/n D
profiles has been also demonstrated but requires further investigation.
3.6 Materials
Development of a multi–scale methodology for composite structural modelling and validation of modelling
procedure by mechanical testing
In the field of computational material science the multiscale methodology plays an important role. It is based
on the hierarchical concept that there is a strong interconnection between phenomena that happen on
different length and time scales. In the field of composite materials, the approach is applied to the description
of the behaviour of the constituents, i.e. fibre and matrix.
In this work a constitutive model for a balanced plain weave fabric was developed. This model, starting from
geometrical and mechanical parameters of the single constituents (fibre and matrix) determines the effective
moduli of the representative unit cell. This model was implemented into the general purpose finite element
program Abaqus, thus building a specific user subroutine.
The last part of this job was to determine a material failure mechanism theory for the balanced plain weave
architecture that was implemented in the same specific user subroutine. The prediction of the failure at each
increment of the load was obtained by using a quadratic failure criterion, applied to the strains with stiffness
and strength reduction scheme to account for
damage within the yarns.
The subroutine is an augmentation for any
commercial finite element code, thus giving the
possibility to deal with any composite material
made with balanced plain weave fabric, provided
that the mechanical properties of the single
constituents and the specific failure mechanisms
are known.
In figure 3.42 an example problem is shown that
has been solved by using this standard user
subroutine. We have studied a square panel of side
600 mm and thickness 3.43 mm where all the
edges are clamped and the load is a uniform
pressure applied to the bottom surface.
Start matrix failure
in the yarn
Green= fill matrix failure
Yellow= warp matrix failure
Pressure (MPa)
Green= fill fiber failure
Yellow= warp matrix failure
1
Start fiber failure
in the yarn
a) Plain weave
Fill yarn
Fill yarn
No damage model
TSAI-WJ criterion
0
0 10 20 40 40
Displacement at the center (mm)
Figure 3.42 – Graphical visualization of the results
3.7 Safety and Environment, Power Plant Conceptual Studies and Socio
Economics
Several fusion safety issues have been faced throughout 2010. Some of them are still going on and final results
will be achieved in 2011. The tasks are described in the following.
Dust mobilization studies
For ITER fusion safety studies one of the most important and potentially dangerous deviations from the
normal operations is the loss of vacuum accident (LOVA) in the plasma chamber, where the presence of
hydrogen isotopes and dust coming in contact with air oxygen might produce conditions overcoming explosive
concentration limits. In this frame a literature research on dust mobilization during LOVA highlighted some
gaps in the experimental data base concerning mobilization in sub–atmospheric pressures. To overcome this
limitation a preliminary design was prepared for an appropriate channel facility operating under vacuum and
of reduced dimensions, which would allow the dust mobilization behaviour to be investigated in a simple flow
geometry and with restrained costs [3.17].

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Cell Reynolds number (log)
10 6 P=0.1×10 5
P=0.2×10 5
P=0.3×10 5
P=0.5×10 5
P=0.8×10 5
P=1×10 5
104
102
0
1 2
x
Figure 3.43 – Reynolds number along the
central axis for a horizontal cylinder, for
pressure range 0.1–1×10 5 Pa
Figure 3.44 – Set–up for the diagnostic
test of the dust mobilization
The work was done in the frame of the F4E Grant
F4E–2008–GRT–01–01 (ES–SF) and dealt with the design and
instrumentation of a channel flow facility and the proposal of an
experimental R&D program on dust mobilization.
The fluid–dynamical analysis showed that a cylindrical vessel,
vertical or horizontal with a high length–radius ratio (1 or 2 m
length/height vs 0.333 m radius) and restrained dimensions, such
that it can be hosted in a laboratory room, is suitable to carry out a
valuable study on the gas – dust particles interaction and dust
mobilization. The k–ε turbulence model has been adopted to
simulate the flow field inside the channel, and stationary conditions
have been extensively investigated. With the shape proposed and
under the boundary conditions imposed, the turbulence is
maintained in the main part of the channel (fig. 3.43).
In addition, the most suitable diagnostics to characterize the dust
concentration were explored, and some small–scale tests were
carried out to demonstrate their capability to match the goals of the
proposed experiment. As far as the safety issues are concerned, it is
necessary to monitor the dust during the mobilization phase, then to
trace the dust paths and to measure the dust velocities and
concentrations during a simulated accident. The dust diagnostics so
far adopted for dust detection inside the VV, during plasma burning,
are not completely suitable for the safety purpose.
A technique using a green laser to illuminate the particles, together
with a fast camera charge–coupled device (CCD), was chosen,
because it seemed to be the most promising one for the specific safety
problem. The main purpose of the experiment was to measure
particle concentrations by counting the number of particles inside a
known volume and pre–set time intervals. The optical setup is then
arranged in such a way to resolve individual particles. Images are recorded on a video–camera, stored in
picture formats and then analysed by an image processing procedure. A small scale test bench was set
up (fig. 3.44).
The proposed diagnostic system has the advantage to be flexible and adaptable to different particle sizes and
different concentration conditions.
Limits on the maximum dust concentrations measurable are imposed by the laser extinction on the beam.
Preliminary calculations and the experimental evidence lead to consider that densities higher by more than
one order of magnitude than those obtained in the existing test (3.7×10 7 SS316 particles/m 3 , for example)
would not limit the measurements. This will be one of the future scopes of the study. The type of dust is not
distinguished by the optical analysis proposed. A knowledge of the interactions among the different types of
dust during re-suspension for LOVA is complex and perhaps not feasible: the investigation of the different
particle velocities, depending on the specific weight of the materials, could be a possibility.
Validation of the PACTITER computer code and related fusion specific experiments in CORELE loop
The objective of this work is the verification and validation of the PACTITER code v3.3, to be used for
predicting activated corrosion products (ACPs) in ITER primary heat transfer system (PHTS) (EFDA Contract
07–1702/1548 task TW6–TSS–SEA5.6). The second part of this activity has dealt with the application of
PACTITER v3.3 to assess the ACP inventory of the ITER NBIs PHTS loop. The focus was given to the
impact of operation scenarios parameters (i.e. water chemistry, materials corrosion properties, etc.) and loop’s
piping architecture. The modelling of the NBIs PHTS loop has dealt with the description of the related
geometric, thermo–fluido–dynamic, material features. The neutron activation of the NBIs (1 Diagnostic and
2 Heating injectors) cooled regions was estimated in rough way as validated data were not available. The basic
criteria used to set up the model are: the total coolant inventory, the wet surface and the mass flow rate of the

technology programme (cont’d.)
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PACTITER model are the same as the real ones; inlet
coolant temperature of each injector component is 35°C.
The NBIs PHTS ACP inventory calculations were carried
out for a 7–week operation scenario split in 23 steps
alternating the various operational phases gathered in
table 3.I. Each operation day includes 14.4 h of mean up (or
operation) time (MUT) and 9.6 h of not–scheduled mean
down time (MDTNS). Every 11 operation days, 3 routine
maintenance days (MDTs) are foreseen.
The most important parameter governing the build–up of
the ACPs inventory is the Cu and Cu alloy corrosion rate.
The influence of chemical volume control system (CVCS)
flow rate and filter efficiency is relatively scarce, as no
appreciable reduction in the ACP mass was assessed by
variation of those parameters. The possibility to separate the
LV–active correction and compensation coils (ACCCs) from
the NBIs PHTS by a dedicated cooling loop was also
investigated. The impact of this choice would be remarkable
in terms of ACP mass reduction (factor ∼3.7; see fig. 3.45,
Run–8). That is explained by the large wet surface of
ACCCs (3798 m 2 which is ∼50% of the total loop wet
surface) made of Cu which is affected by a larger corrosion
and release rates as compared to stainless steel regions.
Another way to reduce the ACCCs wet surface of a factor 2
is by doubling the pancake piping diameter from 8 to
16 mm. That would cause a drop of ∼40% of the ACP mass
(see fig. 3.46, Run–5) [3.18].
One might argue that splitting the NBIs PHTS loop in two
parts would not reduce the overall ACP mass which would
be transferred to the dedicated ACCCs cooling loop. The
actual advantage is the reduction of ACPs radioactive
inventory (see fig. 3.46). The larger ACP mass inventory of
ACCCs would be contained in a dedicated loop, which will
be much less activated, considering their position far from
the plasma and from to the neutrons line of sight. On the
contrary, if contained in the NBIs PHTS this large ACP
inventory would be activated at higher level when
transported to loop’s regions where the neutron flux is larger.
Table 3.I – Main operation scenario data
Operational
phase
Time
(d)
H 2
(ppm)
T mh
*)
(°C)
Idle (MDT S ) 6.9 0.06 35
Conditioning
(MDT S )
3.6 2 55
Injecting (MUT) 5.15 2 61
Decay/dwell
(MUT)
Maintemance
(MDT NS )
18.0 0.06 45
15.35 0.06 35
*) T mh = means temperature of the primary fluid in the
under flux region
(g)
8000
4000
0
0
Run-3 ACCCs wet surface=3798 m 2
Run-5 ACCCs wet surface=1899 m 2
Run-8 ACCCs wet surface=0 m 2
20 40
Time (days)
Figure 3.45 – Impact of the ACCCs Ws on the ACP
inventory (Run-3: full wet surface, Run–5: ½ wet
surface, Run–8: no ACCCs)
Surface activity (GBq)
2000
1000
Run-3
Run-8
0
0 20 40
Time (days)
Figure 3.46 – ACP surface activity in NBI–PHTS
with ACCCs (Run–3) and without ACCCs (Run–8)
Safety analyses by hazard and operability studies
Several hazard and operability (HAZOP) studies have been performed for the ITER detritiation systems under
the frame of different ITER and F4E contracts/grant:
• Pre–conceptual design of the high tritiated water processing system to detritiate highly tritiated water
produced during normal and also abnormal situations of ITER operations (ITER
Contract/CT/09/4300000087) [3.19].
• Conceptual design of the tokamak complex detritiation system aimed at detritiating the gas effluents from
tokamak complex (ITER Contract/CT/09/4300000098) [3.20].
• Conceptual design of the water detritiation system aimed at storing and detritiating the aqueous effluents
produced by the detritiation system and from other sources – F4E Grant –2010–GRT–045 (PNS–VPT)
[3.21].
The HAZOP studies have been performed according to the “ITER Guide to Performing Hazard and

078
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2010
Operability Studies”. After a first preliminary phase dedicated to developing the background information
required for the analysis, such as to define the scope of the analysis, appoint a team leader with approval of
ITER Responsible Officer, assemble the team to perform the HAZOP Study with approval of IO, collect
relevant information and supporting documents, review the system and define the segments of the system to
analyze and, review the ITER “Consequences of interest and hazards table”, team of experts performed the
studies in dedicated meetings by applying the following procedure:
• Review the segments of the system to be analyzed and the “Consequences of interest and hazards table” in
order to get team consensus.
• Select process variables (flow, pressure, etc.) and define operation modes to investigate.
• For each segment, evaluate the potential consequence of the deviation of the process variables by applying
guidewords.
• Determine cause(s) leading to deviation, and identify whether no cause can be determine or the condition
of the cause.
• Evaluate potential consequences and the corresponding hazards resulting from each deviation condition,
and identify whether no consequence is identified or the consequence.
• Develop detection and controls to mitigate effect of the deviation.
• Determine the residual risk associated with the cause with the identified detections and controls, (e.g.:
estimation of the likelihood of the cause occurring and the severity of the cause with the identified detection
and controls).
• Identify the corresponding residual risk (minimal, low, medium or high) by using a risk matrix.
• When the residual risk is high or medium it will be noted as an action item to review whether additional
controls and detection are needed to mitigate the hazard.
• Repeat procedure for each item: deviation of a process parameter (no, less, more, etc.) for the specific
segment; for all process variables (flow, pressure, temperature, etc.) for the specific segment; for all segments
in a section; for all sections in a system.
• Document the identified detection and control for the system.
• Assign responsible individual and due date to address action items.
• Integrate action items into detection and controls for system when complete.
The HAZOP study has been concluded by the identification of limiting conditions of operations (LCOs) for
the overall events that might induce hazardous (incidental/accidental) conditions. LCOs have been
qualitatively identified in order to collect under common categories the detections and the controls most
significant from the safety point of view.
Most of the identified events inducing hazards for the environment and the workers have been classified with
acceptable level of residual risk, as “No additional detection and/or controls needed”, and “Detection and/or
controls reasonable”. Only few events have been judged with a residual risk such to require “Review additional
detection and/or controls”. These events have been highlighted with an indication about consequences, in
terms of the plant conditions and hazards, and additional detection and/or controls required.
In this frame, ENEA developed a software tool to generate a HAZOP spreadsheet database, in agreement with
the “ITER Guide to Performing Hazard and Operability Studies”.
3.8 Broader Approach
JT–60SA
Layout study of JT–60SA coils power supplies and switching network units. According to the Broader Approach
(BA) program development, ENEA and Commissariat à l'Energie Atomique (CEA) had updated and reviewed
the layouts of the new power supplies for JT–60SA PF coils, fast plasma position control coils, toroidal field
(TF) coils and switch network units.
JT–60SA will be located at JAEA NAKA Fusion Institute. Most of the existing building, infrastructure,
available for the JT–60U devices, will be re–used for JT–60SA.

technology programme (cont’d.)
progress report
2010
079
The layout design of the power supplies and switch network units includes the installation of new PS and the
re-use of the existing equipment and buildings.
Procurement of 4 switching network units for JT–60SA. In the frame of the BA agreement, Europe provides
for the procurement of the central solenoid SNU of JT–60SA. The technical specifications for the call for
tenders have been reviewed and modified to include the project progress; the system simulation model has
been integrated with thermal components in order to consider the heating/cooling behaviour of the static
components.
International Fusion Materials Irradiation Facility
The activities carried out in 2010 in the frame of the International Fusion Materials Irradiation Facility
(IFMIF) project are related to the following procurement arrangements.
Participation to the experimental activity of the ELITE loop
(Oarai, Japan) (LF 01 EU). The main components to be delivered
to Japan have been designed and the order for the procurement
of the instrumentation (Nano–ohmmeter, thermo–resistances,
datalogger for thermo–resistances, and NIs items) for the
R–meter, the E–Cubicle and the data acquistion systems (DACs)
have been issued. The order for the modification of the R–Meter
instrument itself, capable of fulfilling the Japanese safety rules,
was emitted as well.
During the second half of 2010 the procedure to issue the
contract for the final design and realization of the R–Meter box,
support structure and the assembly of the electrical cabinet for
the DACs of R–Meter and CASBA sensors was completed. In
figure 3.47 two views of the final assembly of the R–Meter, of the
magnet and of the heating system into the box are reported.
Moreover, the new design of the removable backplate (BP) as well
as the design of the frame were completed and their
thermo–mechanical analysis started. As the design of the BP had
been frozen, ENEA started the procedure for the procurement of
EUROFER.
Figure 3.47 – 3D model of R–meter, magnet,
the heating system and the box
In the figure 3.48 the 3D model of the removable BP is reported.
Lithium erosion corrosion activity (LF03 EU). During the
reporting period ENEA continued the commissioning of the
lithium loop LiFus3. A new test section was also prepared as
advised by the reviewer of the preliminary design review meeting.
The charging and draining procedures were implemented and
the performance of liquid level meters, thermocouples and
pressure transmitters checked.
Figure 3.48 – 3D model of the bayonet
blackplate prototype
During the month of November 2010 tests, the necessity to improve the control of the loop was underlined.
Thus, several thermocouples were added both in the air cooler and in the gas lines and a new pressure
transmitter was installed in the gas line over the test section. New Conax lithium level gauges have been
ordered and the procurement of a new mass flow–meter has been decided.
At the end of 2010 lithium circulation tests were performed and in December the lithium circulation by using
a procedure different from the standard one was obtained. Consequently, commissioning and testing of the
loop was stopped due to different technical and safety problems arisen during the lithium circulation tests. The
outcomes of this test suggested that the loop should be updated and a number of interventions planned.
Due to the limited time at the completion of the IFMIF project, a new test matrix was proposed and accepted
by F4E and Project Team.

080
progress report
2010
Figure 3.49 – Purification system of LiFus 3
loop
Figure 3.50 – Test rig after exposure
Figure 3.51 – Stereoscopic image of the test rig SEM analysis of the gasket
Implementation of the Li purification system (LF04 EU). LiFus 3 is provided a purification system consisting
of a cold trap, a nitrogen hot trap and a hydrogen trap, designed by ENEA in collaboration with the University
of Nottingham (fig. 3.49). This purification system has been updated in late 2009 to take into account the
advices of an expert panel. The installation and commissioning of this system was completed early in 2010.
Nevertheless, due to technical problems, still unsolved, of LiFus 3 loop the purification system has not been
tested yet. Together with the purification system, an online monitoring system for the measurement of the
nitrogen content in lithium was installed.
Testing of the purification system is now planned in February 2011.
Remote handling of the IFMIF target assembly system (LF 05 EU). During 2010, a series of technology and
remote handling (RH) tasks were performed. The technology tasks were aimed at validating the current BP
bayonet concept design. They included:
1) The qualification of the lithium sealing capability of the Helicoflex gasket (see fig. 3.50). This experimental work was
addressed at evaluating the compatibility with lithium of the outer Jacket material of the gasket (i.e. Soft
Iron). After 1800 hrs of exposure at 350°C, post analyses of the rigs, stereoscopic and scanning electron
microscopy (SEM) analyses (see fig. 3.51), were performed. The main outcomes of the tests highlighted the
suitability of the proposed sealing system. Further tests are planned in 2011, thus extending the exposure
time up to 6000 hrs.
2) Evaluation of the swelling phenomena and its impact on the BP coupling system (University of Palermo). This activity was
aimed at evaluating the thermo–mechanical issues potentially induced by neutron swelling in the threaded
connections of the IFMIF target BP. A parametric analysis was carried out for each kind of bolt by
considering a swelling volumetric strain ranging from 0.001% to 0.1%, and both the maximum equivalent
stress within the screw (see fig. 3.52) and the unscrewing torque (see fig. 3.53) were estimated. The results
obtained highlighted that screw dimensions seem to influence mainly the highest values of the unscrewing
torque which are reached at ε sw
∼0.04÷0.05%, regardless of the screw diameter. The maximum ε sw

technology programme (cont’d.)
progress report
2010
081
acceptable to avoid unscrewing torque to be higher than
80 Nm has been assessed for the bolts considered.
Further analyses will be carried out in 2011 to investigate the
potential influence of the screw length of engagement on the
bolt connection thermo–mechanical performances, in
presence of neutron swelling.
3) Bolting seizure and the pre–stress tests. This activity is addressed to
select suitable materials for bolts with the objective to prevent
their seizure or relaxation. The first test was performed early
in 2010, by using the test section shown in figure 3.54, and due
to the bad results (all the bolts broke after an exposure of 1000
hrs at 350°C) the experiment will be repeated in 2011.
4) Validation of the skate system concept. This task is focused at
evaluating the performance and the reliability of the skate
system. In 2010 the following activities were performed:
• Control and DAQ systems design was completed;
• A set of sensors (extensometers) to perform a preliminary
test on the BP already available in DRP were purchased
and delivered.
The setting up of the experiment will start early in 2011.
The RH tasks were focussed on the validation of the cleaning
procedure for the removal of sticked lithium metal from steel. In
detail:
• Cleaning procedure. In order to clean up steel specimens from
sticked liquid metals, at ENEA Brasimone a mixed low acidic
solution was developed. It consists of: acetic acid [1N]
(CH 3
COOH)+hydrogen peroxide (H 2
O 2
)+Ethanol
(CH 3
CH 2
OH) with the same ratio (1:1:1).
This cleaning solution has been used at room temperature, for
cleaning Li, PbLi, Pb, LBE and, according to weight
measurements and electronic microscopy, no steels' kind
(ferritic or martensitic or austenitic) have ever been attacked or
modified or corroded due to the treatment.
This cleaning solution can be used both by immersion or
spraying over a dirty surface.
• RH test of the cleaning procedure. To test the feasibility of the
cleaning operations a very simple tool (see fig. 3.55) was
manufactured and some trials were performed. Tests were
performed by using a prototype of the backplate frame and a
simulation of the TMs (a simple steel plate). The space
available to perform the operation was of about 5 cm, which
is the space left by the high flux test module (HFTM) once
removed.
Engineering design of the IFMIF target assembly system (ED 03).
The technical work performed in the reporting period was
dedicated to: 1) starting the design of the target assembly with
bayonet BP for IFMIF; 2) making a preliminary neutronic
calculation of the newly designed BP.
1) Design of the IFMIF target assembly. A preliminary 3D CATIA
model of the BP component – including the lateral skate
systems and the double reducer nozze – has been
accomplished (fig. 3.56). Starting from the Engineering
Validations and Environmental Engineering Design Activities
εsw=0.1%
Figure 3.52 – Von Mises stress distribution at
different swelling strain
T(Nm)
1200
800
400
0
0.0
M10
M12
M16
0.04 0.08
ε sw (%)
Figure 3.53 – Unscrewing torque vs swelling
volumetric strain
F82H
F82H
EUROFER
Inconel
Figure 3.54 – Test rig for the bolted coupling
system
Figure 3.55 – Cleaning tool
Figure 3.56 – Front view a) and rear view b)
ofthe newly designed IFMIF target

082
progress report
2010
BP
Lithium JET
Atom displacement, dpa/fpy 4.7–10 –1
He production, apppm/fpy 2
H production, apppm/fpy 9
Total heating W/cm 3 3.8–10 –1
HFTM
Atom displacement, dpa/fpy 54
He production, apppm/fpy 598
H production, apppm/fpy 2742
Total heating W/cm 3 23.08
Figure 3.57 – BP
geometrical layout of
MCNP5 code
Figure 3.58 – Main of the
outcomes of the neutronic
calculations on the BP
superimposed mesh tally
(EVEDA) lithium loop concept, the new design has been generated by enlarging the width of the lithium
flow channel from the 100 mm of the EVEDA loop target to the actual IFMIF dimensions (260 mm) and
modifying all other dimensions accordingly in order to fit the new channel size. The channel profile that has
been implemented is the variable–curvature profile (with no straight sections) formerly designed by ENEA.
As to the EVEDA design, the channel profile has now been entirely placed on the BP instead of distributing
it in between the BP and the interface frame. This solution is aimed at eliminating manufacturing problems
and alignment issues between the two components.
A first design optimization has been reached by eliminating all the unnecessary material in order to reduce
the weight as much as possible. This preliminary design has been assessed from the neutronic point of view.
Next step will be a thermo-mechanical assessment in order to evaluate the stress field in the componente,
thus optimizing the design features on this basis.
2) Neutronic calculations. In order to quantify the material damage due to neutron irradiation the following was
determined:
a) Heat deposition due to neutron and gamma interactions with nuclides.
b) Radiation damage (quantified by the number of dpa.
c) Gas production, e.g. total production of helium and hydrogen produced by neutron induced
nuclearreactions.
Preliminary neutron/gamma transport calculations have been performed for the BP via MCNP5 code,
version 1.4 (fig. 3.57).
The neutron induced cross section data files used in the calculations have been taken mainly from INPE–50
library, developed at INPE–Karlsruhe Institute of Technology (KIT), for neutron energies up to 50 MeV,
and LANL–150N, developed at Los Alamos National Laboratory, for neutron energies up to 150 MeV.
The BP profile was approximated in MCNP5 code through a combination of cylinders parallel to X
(horizontal) axis. Mapping on the BP through “superimposed mesh tally” feature of MCNP5 has been used
(fig. 3.58).
DEMO R&D in the Broader Approach activities
Development of CiC composite materials (task T1–2–5). In the framework of the Broader Approach activities
for the development of composite SiC materials, ENEA laboratories have completed the experimental plan
and the preliminary design of the test apparatus aimed at studying the erosion/corrosion of SiC composites
into PbLi.

technology programme (cont’d.)
progress report
2010
083
In particular, the manufacturing of the experimental apparatus consisting of a high temperature oven has
been continued according to an updated mechanical design which will permit operating up to 1200 °C with
a SiC/PbLi relative velocity of 1 m/s.
3.9 JET Fusion Technology
JET housekeeping wastes detritiating
A Pd–based membrane reactor for detritiating JET housekeeping wastes has been studied (JET task
JW10–FT–2.35) [3.22]. A finite element computer code permitted to simulate the behavior of the membrane
reactor by varying the main operating parameters [3.23,3.24]. The increase in the temperature affects very
slightly the detritiation capability while a reduction of the membrane thickness increases the decontamination
factor. On the basis of the reactor modeling, the design and the manufacturing of a Pd–Ag permeator tube
of diameter 10 mm, length 500 mm and wall thickness 0.150, has been carried out. Further, the design of the
membrane reactor has introduced an innovative module configuration [3.25], see figures 3.59 and 3.60. This
innovative design mainly consists of a special device applied to the closed end of the permeator tube. It is a
metallic spring coupled with a flexible wire with two functions:
• to apply a traction force to the permeator tube in order to avoid its contact with the inner walls of the
membrane module and to prevent deformations due to thermal and hydrogenation cycles,
• to ensure the electrical continuity between the closed end of the permeator tube and the outside of the
membrane module by allowing the heating of the tube via Joule effect.
In the reactor manufactured by ENEA, this component is composed by:
• an Inconel spring, able to guarantee the required mechanical performances at the working temperature
(300–400°C) and apply to the permeator tube a traction force sufficient to drive it along a straight direction
during its thermal/hydrogenation expansion,
• a copper wire, to ensure the electric current passage with a low electrical resistance.
Direct ohmic heating has the advantage to heat the membrane only, thus reducing the heating of the process
streams and saving power [3.26]. Furthermore, the traction force applied by the Inconel spring prevents
bending of the long (500 mm) Pd–Ag tube, thus ensuring long life to the membrane.
He
H 2 O
HTO
H
He
H 2 O
Pd-Ag membrane tube
H 2
HTO
2
HT
HT H 2 H 2 HT H 2
HT H 2 HT H 2
Bi-metallic
spring
Insulating electric
feedthrough
H 2
HT
Catalyst
Power
supply
Figure 3.59 – Scheme of the Pd–based membrane reactor for the JET housekeeping wastes detritiation
Figure 3.60 – Pd–based membrane reactor before assembling

084
progress report
2010
Arb. units
Counts/s
10 13
10 11
10 9
0
10 4
10 2
10 0
γ rays (NE213) 33519
33516
33517
33520
33522
33512
33661
a)
Time (s)
γ rays (GEM)
0 1
Time (s)
Figure 361 – Comparison of gamma-ray
signals in FTU plasma discharges: a) time
traces from a NE213 detector; b) time
traces from GEM pad #1
a)
1
33519
33516
33517
33520
33522
33512
33661
b)
-33012
-2577
GEM–based neutron detector for 2.5 and 14 MeV
The development of a gas electron multiplier (GEM)–based
neutron detector for simultaneous 2.5 MeV (DD) and 14 MeV (DT)
measurements has continued in the frame of an EFDA task on
Diagnostics (WP10–DIA–04–01–xx–02). The detector, consisting
of a proton recoil neutron converter and a triple GEM structure
based on a Ar/CO 2
/CF 4
– 45/15/40 gas mixture, is divided in two
sub–units (U DT
and U DD
) respectively measuring 14 MeV neutrons
only and 2.5 + 14 MeV neutrons [3.27]; the U DT
sub–unit is
provided with an aluminum layer (200 μm thick) for the rejection of
protons produced by 2.5 MeV neutrons. The detector has been
installed just outside the cryostat of the Frascati Tokamak Upgrade
(FTU) device on the equatorial plane between ports 11 and 12.
Preliminary acquisitions have been performed during plasma
discharges in the 2010 experimental campaign. Unfortunately, as
hydrogen gas (instead of deuterium) was used in these discharges
due operational constraints, it has been possible to study just the
response of the detector to gamma rays produced by runaway
electrons. Note that this detector is also sensitive to gamma rays
when a high high voltage (HV) setting is used. Results for seven
discharges are shown in figure 3.61 (HV=1180 V): in all cases the
GEM signals correctly reproduce the time behaviour and relative
intensity of the gamma signals from the reference gamma-ray
scintillator detector.
Further tests with the GEM neutron detector have been carried out
by using a 241 AmBe neutron source in the frame of an EFDA Joint
European Torus (JET) Fusion Technology task (JW9–FTFT–5.31).
The source (sketched in figure 3.62 as a cylinder) has been placed
diagonally at about 2 cm from the detector. Results are shown in
figure 3.62. At high HV both neutrons and gamma–rays are
detected by the two sub-units units. At lower HV the detected signal
is due to neutrons only and is mainly seen by the U DD
sub–unit as
the aluminum layer in the U DT
cuts the signal due to neutrons with
E

technology programme (cont’d.)
progress report
2010
085
Counts
10000
100
14 MeV neutrons
Q S /Q L
1
0.8
0.6
0.4
0.2
γ
14 MeV n from dt
2.5 MeV n from dd
300
200
100
0 0
0 400 800 1200 1600 2000
1
0 1000
Channels
2000
Figure 3.63 – Example of 14 MeV
monoenergetic neutron beam measurement
during the CNS detector characterization
campaign at PTB
Q T /Arb. units
Figure 3.64 – CNS capability of n/γ discrimination and
simultaneous 2.5 MeV and 14 MeV neutron
measurements: 98% of the recorded pulses represent
good data (2% of bad reconstructed pulses have now
been fixed using the last release of the pre-processing
routine)
characterization matrix is on–going. Moreover, data collected in the previous JET experimental campaign
have been analyzed in order to study the system’s performance during plasma discharges (see fig. 3.64): some
minor bugs in the data pre–processing routine have been found and fixed. The final installation and
commissioning at JET of the CNS diagnostic is foreseen in 2011–2012.
3.10 Cryogenics
Liquid helium service
The temperature of liquid helium at normal boiling point is 4.2 K (about –296°C). It is therefore used as a
typical refrigerant for cooling of experimental equipments at very low temperatures. Helium, however, is
rather expensive, so that recovering the helium evaporated during experiments may be beneficial, provided
that the overall helium consumption is large enough. The gas recovered and compressed in a suitable storage
system can indeed be liquefied again, subject to prior purification, thus allowing recycling the same amount
of helium many times.
A number of activities pursued by the Fusion Technical Unit at the Frascati Research Centre of ENEA require
considerable amounts of liquid helium every year. The experiments with the FTU need liquid helium to cool
down the magnets of the auxiliary heating facility electron cyclotron resonance heating (ECRH), or to freeze
solid deuterium pellets to be injected at high speeds into the fusion plasma, as well as to keep plasma
diagnostics at low temperature. The Superconductivity laboratory also uses sizeable amounts of liquid helium
for experiments with large superconducting magnets, as well as to test critical components at low temperatures.
A helium recovery facility and a helium liquefier are therefore operating in the Centre, which allows helium
to be recycled many times, thus saving a considerable amount of money. The helium liquefier is a Linde
TCF20, featuring a cold box equipped with an internal auto–purifier and two turbine expanders for gas
pre–cooling, a screw driven DS220 recycling compressor equipped with an oil removal system, a line drier, a
pressure control panel, two self–pressurizing dewars (1.000 l each) for liquid helium storage, two transfer and
decant lines, a 7 m 3 pure helium buffer tank and an analytical panel equipped with a purity monitor and a
moisture meter. An Allen Bradley Programmable Logic Controller automatically controls the plant, which was
supplied by Linde Cryogenics Ltd. (UK) on April 1999. The nominal liquid helium production rates, with or
without liquid nitrogen precooling, are of 60 and 30 l/h.
In 2010, nearly 14.300 litres of liquid helium have been delivered to the users, which is slightly less than the
amount of the previous year (about 19.500 litres). About 20% of the liquid evaporates (to the recovery system)
while decanting it from the storage vessel to transport dewars (in order to deliver the product to the users).
Moreover, the self–pressurizing storage dewar features an evaporation rate of about 25 liter/day, due to the
unavoidable heat losses; of course, the helium vaporized from the liquid storage is recovered too. As a
consequence, the overall liquid helium inventory managed in 2010 turns out to be roughly 23.200 litres. Nearly

086
progress report
2010
12.500 litres (corresponding to more than 230 hours of operation) have been produced by the Linde liquefier,
while the remaining 10.700 litres have been purchased to compensate for the system losses. The resulting
average recovery efficiency is estimated to be about 53.8%, less than that recorded in 2009 (when about 65.5%
was recovered). Such a low recovery efficiency is partially due to some demanding experiments carried out by
the Superconductivity laboratory, which may occasionally require considerable amounts of refrigerant to be
evaporated in a very short time, thus making both the capacity of the gas bag (5 m 3 ) and the throughput of
the recovery compressors (80 m 3 /hr), inadequate. Moreover, during about six months of 2010, the recovery
system was operating with only one of its two compressors, due to maintenance being performed in turn on
both machines, as it is necessary after 15.000 hours of operation. This resulted in a reduced recovery
throughput of only 40 m 3 /h and therefore in losses larger than usual.
Data from independent gas counters in the (separated) recovery lines of each user allowed, throughout 2010,
the different recovery efficiencies to be disentangled by FTU and Superconductivity laboratory. As a matter of
fact, by comparing the amounts of liquid helium delivered and of gas recovered by Superconductivity
laboratories and FTU it is possible to evaluate the percentage of total helium lost by each user.
References
[3.1] G. Ramogida et al., Final report on the analysis of the ITER divertor cassettes, Contract EFDA/07/1702–1596
(TW6–TVD–DIAGAN), ENEA Internal Report FUS–TN–DI–R–009/Rev.1, (October 2010)
[3.2] G. Ramogida et al., Final report on the preliminary electro–magnetic load analysis for the design of a blanket
manifold pipe concept for ITER, Contract EFDA/07/1702–1620 (TW6–TVB–MANEM), ENEA Internal Report FUS–
TN–BB–R–034/Rev.1, (December 2010)
[3.3] S. Tosti et al., Conceptual PFD of HTW processing, KIT Report TLK–CEW–0990PMT1–RD–0D02–01, 5 (May 2010)
[3.4] C. Rizzello et al., Fusion Eng. Des. 85, 58–63 (2010)
[3.5] M. D’Arienzo et al., Fusion Eng. Des. 85, 2288–2291 (2010)
[3.6] F. Borgognoni et al., Fusion Eng. Des. 85, 2171–2175 (2010)
[3.7] A. Santucci et al., A comparison study of highly tritiated water decomposition processes, Presented at the 9th
International Conference on Tritium Science and Technology – Tritium (Nara, Japan 2010)
[3.8] P. Rossi et al., Ultimate tensile strength testing campaign on ITER pre–compression ring mock–ups, Presented
at the 26th Symposium on Fusion Technology – SOFT (Porto 2010) and to appear in Fusion Eng. Des.
[3.9] F. Crescenzi et al., Mechanical characterization of glass fibre–epoxy composite material for ITER
pre–compression rings, Presented at the 26th Symposium on Fusion Technology – SOFT (Porto 2010) and to
appear in Fusion Eng. Des.
[3.10] L. Petrizzi, F. Moro, Final report on ITER diagnostic port integration: diagnostic port plug engineering and
integration, ENEA Internal Report FUS TN GE–VD–Q–001, Contract: FU06 CT 2006–00134 (EFDA/06–1432),
(February 2010)
[3.11] F. Moro et al., Neutronic calculations in support of the design of the ITER high resolution neutron spectrometer,
Presented at the 26th Symposium on Fusion Technology – SOFT (Porto 2010) and to appear in Fusion Eng.
Des.
[3.12] R. Villari, L. Petrizzi, G. Brolatti, Three–dimensional neutronic analysis of the ITER in–vessel coils, Final Report
of the Contract ITER/CT/09/4100001120, deliverable D2-2 (November 2010)
[3.13] R. Villari et al., Three–dimensional neutronic analysis of the ITER in–vessel coils, Presented at the 26th
Symposium on Fusion Technology – SOFT (Porto 2010) and to appear in Fusion Eng. Des.
[3.14] R. Villari, L. Petrizzi, Neutron and gamma spectra behind ITER blanket modules and in divertor, Final Report of
the Contract ITER/CT/09/4100001120, deliverable D1–1 (March 2010)
[3.15] R. Villari, L. Petrizzi, Neutronic analysis of ITER divertor rails, Final Report of the Contract
ITER/CT/09/4100001120, deliverable D2–1 (April 2010)
[3.16] D. Marocco, B. Esposito, F. Moro, Combined unfolding and spatial inversion of neutron camera measurements
for ion temperature profile determination in ITER, to appear in Nucl. Fusion
[3.17] V. Cocilovo, M.T. Porfiri, R. De Angelis, A channel facility for ITER safety relevant dust mobilization studies,
Presented at the 19th Topical Meeting on the Technology of Fusion Energy – TOFE (Las Vegas 2010)
[3.18] L. Di Pace et al., Application of PACTITER v3.3 to the ACPs assessment of ITER neutral beam injectors primary
heat transfer system, Presented at the 19th Topical Meeting on the Technology of Fusion Energy, – TOFE (Las
Vegas 2010)

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2010
087
[3.19] T. Pinna et al., HAZOP study for the HTW processing, ENEA Internal Report FUS–TN–SA–SE–R–208 (June 2010)
[3.20] T. Pinna, C. Rizzello, A. Santucci, HAZOP study for the TC–DS conceptual design, ENEA Internal Report
FUS–TN–SA–SE–R–209 (August 2010)
[3.21] T. Pinna, C. Rizzello, A. Santucci, HAZOP study for the preliminary conceptual design of water detritiation
system, ENEA Internal Report FUS–TN–SA–SE–R–214 (2010)
[3.22] S. Tosti et al., Processo per la detriziazione di soft housekeeping waste e impianto relativo, Domanda di brevetto
per invenzione industriale n. RM2010A000340 del 22.06.2010
[3.23] F. Borgognoni et al., Multi physic approach for Pd–based membrane reactor modeling for wet gas detritiation,
Presented at the 9th International Conference on Tritium Science and Technology – Tritium (Nara, Japan 2010)
[3.24] S. Tosti et al., Design of Pd–based membrane reactor for gas detritiation, to appear in Fusion Eng. Des.
10.1016/j.fusengdes.2010.11.021
[3.25] S. Tosti et al., Reattore a membrana per il trattamento di gas contenti trizio, Domanda di brevetto per
invenzione industriale n. RM2010A000330 del 16.06.2010
[3.26] S. Tosti, F. Borgognoni, A. Santucci, Electrical resistivity, strain and permeability of Pd–Ag membrane tubes, Int.
J. Hydrogen Energy 35, 7796–7802 (2010)
[3.27] B. Esposito et al., Nucl. Instrum. Meth. A 617, 155–157 (2010)
[3.28] F. Belli et al., Conceptual design, development and preliminary tests of a compact neutron spectrometer for
the JET experiment, Proceeding of the IEEE Nuclear Science Symposium Conference, Record, N13–170 (2009)

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2010
chapter 4
superconductivity
During 2010 the Superconductivity laboratory activities regarded manufacturing and testing of conductor
samples for ITER (the way in latin) and its satellite facilities, in addition to characterization and research on
conventional low–temperature superconducting (LTS) materials and high temperature superconductor (HTS)
coated conductors.
Concerning the cable–in–conduit conductor (CICC) production, ENEA was involved in manufacturing the
two poloidal field (PF) conductor samples PF1 and PF2 for ITER, and in the compaction of ITER centrals
solenoid (CS) empty tubes. Activities were also carried out for the feasibility verification of the superconducting
toroidal field (TF) magnet system of the Fusion Advanced Studies Torus (FAST). Besides these activities,
ENEA was involved in the heat exchanger design for the 30 kA current leads of ENEA CICC upgrade facility
and in the study on the effect of strand bending on the voltage-current characteristic of Nb 3
Sn CICCs.
R&D on LTS was devoted to transport and structural studies on NbTi and Nb 3
Sn strands. Direct inter–
filament resistivity measurements were extended from Nb 3
Sn to NbTi samples; with the aim to clarify how
the jacket affects the strain configuration in the complex system of multi–filamentary strands, ENEA carried
out high–resolution x–ray diffraction measurements at the high–energy scattering beam–line ID15B of the
European Synchrotron Radiation Facility (ESRF) in Grenoble. The ENEA 2–components pinning model for
NbTi has been updated in order to take into account the temperature non–scaling. Several commercial Nb 3
Sn
and NbTi samples have been characterized by means of inductive and transport measurements in the ENEA
superconductivity facilities.

superconductivity (cont’d.)
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089
4.1 Cable–in–Conduit Conductor
Manufacturing of ITER poloidal field conductor samples
ENEA is involved in the ITER program for the final qualification
of the conductors before the production phase. In this framework
ENEA was awarded two contracts, one from Iter Organization
(IO) and one from Fusion for Energy (F4E), for the manufacturing
of two PF conductor samples (PF1 and PF2,
ITER/CT/09/4300000048 Support on ITER PF conductor
sample design and preparation; F4E–2009–OPE–021 (MS–MG)
PF conductor sample). The activities started at the end of 2009
and went on for the first months of 2010, and have been split
between ENEA and Commissariat à l’Energie Atomique (CEA)
laboratories.
In particular, ENEA took care of jacketing and compacting the
cables supplied by IO and F4E (PF1 cable manufactured by the
Russian Domestic Agency and PF2 by Chinese Domestic Agency),
of preparing the cables and designing, manufacturing and
assembling the hairpin boxes. The hairpin box solution (fig. 4.1),
an ENEA concept which avoids the use of a resistive joint between
the two conductors, should diminish the possibility of current
redistribution effects close to the high field sample zone. CEA
prepared the upper terminations, assembled at ENEA
laboratories, and performed the sample insulation (fig. 4.2) and
instrumentation at CEA laboratories. As a result of these
contracts’ activities, two samples have been delivered to Centre de
Recherches en Physique des Plasmas (CRPP) and successfully
tested there in the SULTAN facility.
Figure 4.1 – Hairpin box
Figure 4.2 – PF sample, clamping and
insulation
Compaction of ITER CS empty tubes
To the purpose of obtaining a relevant stress–strain state to
perform bending trials in preparation of the winding operations,
IO placed a contract to ENEA to compact 10 empty tubes of
316LN steel. In this framework, ENEA took care of all the
relevant activities concerned with the compaction (fig. 4.3). In
particular, a deep inspection of the tubes external surface has been
performed together with a comprehensive dimensional survey of
Figure 4.3 – ITER CS during compaction

090
progress report
2010
each tube as to: i) total length; ii) wall thicknesses at head, middle and tail positions; iii) external width and
height at head, middle and tail positions; iv) inner diameter at head, middle and tail positions.
Once the compaction was completed, a verification of the final dimensions was conducted, thus establishing
the opportunity of estimating the thickness change and the elongation due to the compaction.
Feasibility verification of the superconducting proposal of the TF magnet system of FAST
FAST, the Italian proposal of a satellite facility to ITER, has been the object of an intense activity aimed at
investigating the feasibility of a superconducting solution for its magnet system. In particular, an analysis of
the TF magnets thermal behaviour has been carried out. The minimum temperature margin in the coil has
been calculated for the thermal load distribution on winding and cable jacket due to nuclear heating.
Moreover, only the room available in the resistive design has been considered and reference has been made to
one of the most severe scenarios envisaged for FAST. Indeed, one of the main critical aspects in the operation
of a superconducting TF magnet is the conductor heating due to direct energy deposition by neutrons and by
secondary gamma rays generated during plasma operation. The operating temperature and the relevant
temperature margin (i.e. the operating safety margin) of the magnet depend both on the heat loads and on the
capability of the coolant to remove it. In this framework, some analyses have been performed aimed at
verifying the validity of the assumptions based, where possible, on the design parameters of similar machines.
Finally, the need for a neutron shield has been preliminarily assessed by comparing the effect of different
nuclear heating thermal loads on the performance of the SC cable.
Heat exchanger design for the 30 kA current leads of ENEA CICC upgrade facility
ENEA is carrying out an upgrade of the existing CICC facility for the characterization of superconducting
long length wound conductors in forced He flow cooling. In this framework, a considerable effort has been
spent for the design of a pair of optimized gas–cooled current leads for a 30 kA insert. Indeed, several
geometrical and operational constraints have been considered due to the re–use of the pre–existing cryogenic
plant and background magnet. In particular, to limit the overall length of the leads and the amount of cooling
gas to be employed, a hybrid resistive/HTS configuration has been adopted. It consists of a copper bar, with
an adequate heat exchanger, from room temperature down to 40 K and an HTS part made of BSCCO from
40 K down to 5 K. Two helium inlets are foreseen for the lead
cooling: one at 5 K at the bottom edge, in parallel to the
background magnet refrigerant line, and one at 16 K in the
joint between the resistive and the HTS parts. To avoid the
formation of short paths, typical of the braided solution
adopted in the leads installed in the ENEA facility, a helical
configuration of the heat exchanger around the current
carrying copper bar has been conceived. Moreover, a reduction
Figure 4.4 – Temperature distribution in the
3D CFD analysis carried out in the exchanger
design
in the manufacture cost is expected as compared to other
solutions that rely on finned bars or zig–zag approaches.
Calculated temperature distribution is shown in figure 4.4.
Study on the effect of strand bending on the voltage–current characteristic of Nb 3
Sn CICC
The primary issue of Nb 3
Sn CICCs has been degradation, often observed during electromagnetic cycling, in
the current sharing temperature, n–index, and critical current, as to measured strands values, due to strain
effects. In the last years, work performed mostly relating to fusion magnet technology has led to a better
understanding of the parameters required to improve the constraints imposed by the brittle nature of the
Nb 3
Sn filaments, such as cables low void fraction and long twist pitch sequence. On the other hand,
experimental campaigns on bent Nb 3
Sn wires, pre-compressed into a stainless–steel jacket, have shown that
an appreciable decrease in the n–index values already occurs at the strand level, well below the irreversible
mechanical load regime for filament breakage.
This result is explained, with the support of simulation results, by taking into account the broadening of the
critical current distribution on the wires’ cross section due to the presence of the jacket and to bending strain.
Considering the impact of the CICCs’ layout on their overall performances, the effective resemblance between

superconductivity (cont’d.)
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091
pre–compressed bent wires and strands inside cables is emphasized, and an innovative interpretation of the
cabled conductors test results is given, from which a tool can be derived to predict their performances in terms
of the n–index versus critical current relation of constituting strands, characterized under simultaneous
pre–compression and bending strain.
Role of the cross section geometry in rectangular Nb 3
Sn CICC performances
Parameters such as cable twist pitch (TP) and void fraction (VF) have a strong impact on CICC performances.
A proper choice of their values in the direction of raising TP and lowering VF has been proven to considerably
enhance the transport properties, though increasing the AC losses, and to appreciably reduce the conductor
degradation with electromagnetic and thermal loading cycles. It has also been demonstrated that a further
route for the improvement of CICCs’ performance is represented by a suitable optimization of the conductor
shape as to the electromagnetic force distribution. In this sense, CIC conductors with high aspect ratio
rectangular geometry, if properly oriented, have shown a better response to high electromagnetic pressure, as
proved by experimental evidences.
The test results of a large rectangular Nb 3
Sn CICC have been interpreted, thus concluding that a high aspect
ratio rectangular cross section helps in preventing performance degradation in CICCs and offers, if properly
oriented as to the electromagnetic load, a better support to the strands, thus limiting the bending strain effects.
4.2 R&D on Superconducting Materials
Study on Nb 3
Sn strands
Direct measurement of transversal resistance in Nb 3
Sn strands. The first direct measurement of the interfilament
resistance vs. temperature by using a 4–probes method has been realized at ENEA Frascati and
applied to Nb 3
Sn samples as well as to NbTi ones, also including the effect of a background magnetic field
(0–3 T) at low temperatures.
This activity, started in 2009 and still going on, lead to the clarification of the role of copper in lowering the
measured resistance. This fact has been demonstrated, both in the NbTi and Nb 3
Sn strands, making use of
auxiliary samples with a reduced thickness of the outer stabilizing copper shell. The tranverse resistance R t
of
NbTi strands, which have a pure copper matrix, resulted to be an order of magnitude lower than that of the
Internal–Tin Nb 3
Sn samples and this is acceptable considering the presence of bronze in the strand
cross–section.
This higher resistivity in Nb 3
Sn samples in turn plays a role in determining the ac losses, thermal stability, and
sensitivity to the wire’s mechanical bending. A 2D FEM model of the wire cross section has been developed
which allows the per–unit length electrical conductance matrix between filament bundles at different
temperatures to be computed. The experimental and numerical results confirm that this matrix has a strong
temperature dependence. The transverse electrical parameters computed with the 2D model are then
implemented in a 3D lumped parameter non-linear electrical circuit that describes the entire wire sample
under test.
X–ray diffraction measurements. ENEA is carrying on high–resolution x–ray diffraction measurements at the
high–energy scattering beamline ID15B of the ESRF in Grenoble. The lattice parameters, averaged over
several Nb 3
Sn peaks along both axial and transverse orientations, have been computed for the bare and the
jacketed wires as function of the applied composite strain. Starting from a zero applied stress, the application
of the axial stress causes the lattice cell to relax its pre–compression in the axial direction, and to increase the
compression in the radial one. A cubic cell is again obtained at an applied tensile strain of 0.094% in the bare
strand and of 0.564% in the steel reinforced one.
A much larger strain range (up to about 1.1%) is allowed to take place in the stainless steel reinforced wire as
to the bare wire (up to about 0.5%), thus suggesting an important hint toward the improvement of the strain
tolerance in such systems. In fact, the application of the outer steel reinforcement is not accompanied by a loss
in performances, since the maximum in the critical current vs. axial strain curve is maintained in the jacketed
wire, as compared to the bare one.

092
progress report
2010
Critical current dependence on axial strain in stainless steel jacketed Nb 3
Sn wires. The critical current of an
internal tin Nb 3
Sn wire developed by Oxford Instruments Superconducting Technology for ITER has been
studied under axial strain at fields between 12 T and 19 T≅4.2 K. Simulating the situation in a cable in
conduit, where thermally induced compressive strain is important, the same wire was jacketed with AISI 316L
stainless steel. The reinforced wire shows an important increase of ε m
, the applied strain where the critical
current I c
reaches its maximum, from 0.25% to 0.57%. In addition the irreversibility limit, ε irr
, is improved
from 0.50% applied strain to >1.10%. It could also be shown that the I c
at zero intrinsic strain is almost
identical. This demonstrates that jacketing does not influence the physical parameters of the original wire.
Experimental data of the bare wire has been well fitted by different strain functions. However, it was not
possible to adequately model the data of the jacketed wire. There are indications that only models, which take
into account the multidimensional character of strain are able to describe the behaviour, but further
development is required.
Study on NbTi strands
Test results of the Chinese NbTi wire for the ITER poloidal field magnets: a validation of the 2–pinning
components model. A two–component model has been developed in the past years at ENEA in order to
describe the normalized bulk pinning force curves and the critical current density of NbTi strands over a wider
B–T range as compared to conventional single–component models. The model was previously successfully
applied to data collected on several NbTi commercial strands. For further validation, we have extensively
tested a strand recently produced by the Chinese Company Western Superconducting Technologies (fig. 4.5)
for the ITER PF magnets PF2 to PF5, and applied the model to these data. In order to take into account the
observed non–scaling with temperature of the reduced pinning force curves, the model has been updated by
including the observed difference in the temperature dependences of the two components, thus contributing
to the overall bulk pinning force.
Critical current density (A/mm2)
4
2
0
0
×103
8.5 K
4.2 K
Transport J c
4.2 K
5.0 K
5.5 K
6.0 K
6.2 K
6.5 K
7.0 K
7.5 K
8.0 K
4 8
Field induction (T)
Magn. J c
4.2 K
5.0 K
5.5 K
6.0 K
6.5 K
7.0 K
7.5 K
8.0 K
8.5 K
2-components
Fit
4.2 K
5.0 K
5.5 K
6.0 K
6.2 K
6.5 K
7.0 K
7.5 K
8.0 K
8.5 K
Figure 4.5 – Superconductor transport (large
symbols) and magnetization (small symbols)
current density as function of the magnetic
field, in the temperature range 4.2 K to 8.5 K
F p
= F p (1) + F p (2) = C 1·(1 t n ) 1·b 1·(1 b) 1 + C 2·(1 tn ) 2 ·b 2 ·(1 b) 2
With reduced field and temperature defined as follows:
b=B/B irr
(T) = B/B irr_0
(1-t n ); t = T / T c0
the importance of testing wires over very wide temperature
ranges is evidenced, and the good agreement between
experimental and fit results validate the proposed formulation. It
can be regarded as a reliable tool for the description of NbTi
performances, to be used in the design of superconducting
magnets. From the phenomenological point of view, it is shown
that, at low temperatures, the two pinning mechanisms
contribute to the bulk pinning force, thus resulting in pinning force peaking at a reduced field B/B irr
≅0.5. As
the temperature increases, the pinning force peak moves to lower fields, thus indicating that the low field
component pinning mechanism becomes dominant.
Quality control monitoring of NbTi strands for JT–60SA TF coils
ENEA was awarded a contract from F4E (F4E–BOA–004 (MS–MG)) for the performance of quality controls
in support of the evaluation process of industrial offers and for an in depth monitoring of the NbTi strand
production for JT–60SA TF conductor. The TF conductor is a CICC, and relies on twisted multifilament
NbTi–based composite strands with copper strands cabled and than jacketed to become a CICC. The
characterization of the superconducting strands in the temperature and magnetic field range of interest is in
fact a fundamental step in the design of a magnet for a tokamak reactor, based on them. During 2010 three
NbTi strands have been characterized at the ENEA Variable Temperature facility in terms of transport critical
current and n–value at various temperatures.

superconductivity (cont’d.)
progress report
2010
093
ITER NbTi strand benchmarking tests
Scope of this task was to test the performances and layout of
a NbTi strand intended for the ITER PF coils, as a
benchmark of different test facilities. Strand characterization
has been performed at ENEA in terms of strand layout
(diameter and twist pitch length), transport critical current
and n–value at liquid helium, hysteresis losses by
magnetization technique, critical current as function of
temperature, by magnetization technique (formally not
foreseen by the benchmarking activity), transport critical
current and n–value at variable temperature (formally not
foreseen by the benchmarking activity). Superconducting
critical current density (J c
) is shown in figure 4.6.
4.3 High Temperature Superconductors
Figure 4.6 – Superconductor Jc extracted from
magnetization data, carried out at different
temperatures (small dots), as compared to
transport data collected at the ENEA VTI facility
(large symbols). The data measured in the ENEA
liquid helium facility are also reported (green
square symbols) for comparison
0 4 8
Field induction (T)
The activity of the laboratory on High Temperature Superconductors (HTS) was focused on the development
and characterization of coated conductor (CC) tapes based on films of YBa 2
Cu 3
O 7–X
(YBCO) and their
application. In particular, the activities can be grouped as follows:
i) Coated conductor processing. Exploitation of the innovative low-fluorine chemical deposition for high current
YBCO films and development of a more robust and economic oxide buffer layer architecture suitable for
alternative copper based metallic substrates. These activities have been carried out in collaboration with
Technical University of Cluj (Romania), Physics Department of “Tor Vergata” Rome University, Institute
of Material Science of Consiglio Nazionale delle Ricerche (CNR) and Physics Department of Roma TRE
University;
ii) Conceptual design of the construction of toroidal field based on HTS–coils for a tokamak machine. This work was aimed
at the upgrading of the Istituto Superior Técnico Tokamak (ISTTOK) machine and has been carried out
in cooperation with Portuguese and Slovak EURATOM Associations.
Critical current density (A/mm2)
4
2
0
×103
Transport (VTI) Magnetization
3.5 K
4.2 K
4.2 K
5.0 K
5.0 K
6.0 K
6.0 K
7.0 K
7.0 K
8.0 K
7.5 K
8.5 K
4.2 K_LHe test facility
Study of Ni–Cu based alloy tapes for YBCO coated conductor application
Ni–Cu (Ni with 50 at% Cu) alloy based substrates are currently studied at ENEA in order to obtain substrates
with intermediate characteristics between Ni and Cu. To stabilize the microstructure of the binary Ni–Cu
alloy, 3 at% Co was added (Ni–Cu–Co). This alloy tape develops
a rather sharp cube structure, with a fraction around 97% of
cubic grains. The alloy is stronger than either the pure metals or
T c0 = 86.6 K
the binary Ni–Cu and shows a yield strength of around 120 MPa
at 0.2% offset. Ni–Cu–Co shows a reduced magnetism as
compared to the currently employed Ni–W alloy, since the Curie
temperature is around 155 K. This alloy substrate was successfully
used for the realization of high J c
YBCO films. The film is mainly
104
c–axis oriented with a minor fraction of a–axis grains. A J c
of
about 1.1 MA cm –2 was measured (fig. 4.7).
Another solution was the substitution of Co with W. A very strong
cube texture is obtained only for W content around 0.5 at%
(Ni–Cu–W), with a fraction of cubic grains around 99% without
secondary recrystallization. This alloy is as strong as Ni–Cu–Co
and is nonmagnetic at 77 K, since the Curie temperature is
22.5 K. A more suitable MgO–based buffer layer architecture was
studied. MgO film was epitaxially deposited by e–beam
evaporation in H 2
O atmosphere at temperature as low as 400°C.
A 10 nm– thick Pd seed layer was used to promote MgO epitaxy.
The x–ray θ–2θ spectrum shows that the MgO film is well
Critical current density (A cm-2)
10 6 10 nm
2 4 6
1000
100
0
Magnetic induction (T)
Figure 4.7 – Critical current density as a
function of the magnetic induction J c (B) for
a YBCO/CeO 2 /YSZ/CeO 2 sample grown on
Pd–buffered Ni–Cu–Co substrate. In the
inset, the cross section of the multilayer
architecture is shown

094
progress report
2010
Intensity (Arb. units)
1200
800
400
As grown
Annealed
42 43
(200) MgO
0
35 45 55
2θ (deg)
Figure 4.8 – X–ray diffraction θ–2θ scan of MgO
film grown on Pd–buffered Ni–Cu–W. The
spectra refer to the as–grown sample (full
dots) and after annealing at 800 °C in vacuum
(empty dots). In the inset, a detail of (200)MgO
peaks. The thickness of MgO, Pd and Ni–Cu–W
is 120 nm, 10 nm and 50 mm, respectively
Realtive–peak–to–peak
intensity (Arb. units)
2×10 3
a)
(200) Pd
b)
(200) Ni–Cu–W
Zr
Y
O
Ce
Ni
Pd
0 400 800
oriented (fig. 4.8). The MgO layer is adequate for the
deposition of a cap layer before YBCO deposition as tested by
the heat treatment at 800 °C in vacuum.
Oxidation behaviour of the Ni–W and CeO 2
interface: role of
Pd inter–layer
Considering that oxidation at the substrate interface could
influence the epitaxial growth and the mechanical stability of
the whole coating architecture, the role of the Pd interlayer at
the interface between NiW and CeO 2
/YSZ/CeO 2
buffer layer
structure has been studied by x–ray techniques and electron
Auger spectroscopy. Extended x–ray absorption fine structure
(EXAFS) analyses reveal that the inter–diffusion process
between the Pd layer and the substrate modifies the substrate
interface composition due to the formation of an ordered
Ni–Pd alloy even at temperatures as low as 600°C. At high
temperatures, the oxidation mechanism is dependent on the Pd
layer thickness, and competition between the NiO and the
NiWO 4
formation is observed. Oxidation also affects the CeO 2
interface with the substrate. Auger electron spectroscopic (AES)
analyses, reported in figure 4.9, reveal that the interface region
is more extended than that of samples in vacuum annealed,
and that outward migration of Ni and Pd in the CeO 2
layer
occurs. The CeO 2
layer contamination results to be reduced as
the Pd layer increases. The lower CeO 2
contamination and the
lower NiO formation could be associated to the good adhesion
obtained in coated conductor samples with a Pd interlayer.
4×10 3 400 800 1200 1600
Realtive–peak–to–peak
intensity (Arb. units)
0
0
8×10 4
4×10 4
Sputtering time (s)
c)
Zr
Y
O
Ce
Ni
Pd
0
0 400 800 1200 1600
Sputtering time (s)
Figure 4.9 – AES depth profiles for CeO 2 /YSZ
samples deposited on Ni–W substrate buffered
with a 50 nm Pd over layer annealed at 800 °C
a) in vacuum and b) in 10 mTorr of oxygen back
ground pressure. c) AES depth profiles for
CeO 2 /YSZ samples deposited on Ni–W
substrate buffered with a 200 nm Pd over layer
annealed at 800 °C in 10 mTorr of oxygen back
ground pressure. W signal is multiplied for 50
Low fluorine YBCO MOD
In this low fluorine method, only the Ba precursor is introduced
in the coating solution as a trifluoroacetate, while other
precursors are (Cu and Y)–acetate treated with an excess of
propionic acid. The reaction path for YBCO formation was
investigated by x–ray photoelectron spectroscopy (XPS) and
diffraction (XRD). From these analyses it resulted that the
YBCO formation occurs through a rather complex mechanism
involving hydrolysis of both Y and Ba fluorides and the
reaction with CuO. However, this path is probably hindered by
a competing reaction taking place at the same temperature
range as YBCO formation (around 700–800°C). In figure 4.10,
the evolution of the reaction path can be derived through
x–ray diffraction θ–2θ patterns. Within the 700–795°C
temperature range the YBCO phase coexists with both
Y 2
Cu 2
O and BaF 2
phases. At 795°C, the XRD spectrum
shows the presence of sharp and intense (00l) reflection of
YBCO together with other distinct peaks ascribable to the
presence of residual Y 2
Cu 2
O 5
phase, while BaF 2
phase
disappears.
Transport properties improvement in low fluorine YBCO MOD with artificial pinning sites
The possibility of enhancing the pinning efficiency by means of artificial pinning sites created by addition of
BaZrO 3
(BZO) has been investigated in YBa 2
Cu 3
O 7–x
(YBCO) films grown by metallorganic decomposition
(MOD) methods. YBCO and BZO coating solutions were prepared and subsequently mixed in molar ratios
corresponding to 5, 7.5, 10, 15 mol.% BZO. A marked increase in the J c
(0) value has been measured in MOD

superconductivity (cont’d.)
progress report
2010
095
Intensity (Arb. units)
(0.04)
(0.05)
26 30 34 38 42
2θ(Deg)
*
795°C+10
795°C
750°C
725°C
700°C
550°C
Pyrolysed
Figure 4.10 – XRD analyses for the samples
quenched at different temperatures. Continuous line
represent the Ba(O,F) reflection, while dashed line
are the peak ascribable to the Y 2 Cu 2 O 5 phase. (004)
and (005) are two reflection of the YBCO film, while
* is a substrate feature
Pinning force density (GN/m3)
10 11 a)
10 9
10 7
(MOD)YBCO
82 K
65 K
77 K
10 5 0.1 1 10
Magnetic induction (T)
10 K
30 K
50 K
Critical current density (A/m 2 )
10 5 T=77 K
10 3
10 1
10 -1
10 -3
10- 5
Correlated
Isotropic
γ=4
4 T
6 T
8 T
-20 20 60 100
Angle (degree)
b)
Figure 4.11 – a) Pinning force
densities of pure YBCO (red)
and 10 mol. % YBCO–BZO
(blue) measured as a function
of the applied field at several
temperatures; b) critical
current density as a function
of the applied field direction
measured for 10 mol.%
YBCO–BZO at 77 K at different
field field intensity. The green
line represent the isotropic
contribution while the black
arrow shows the correlated
contribution at 0° and 8 T
0.5
0
-0.5
0.4
0.2
0
-0.2
Figure 4.12 – Map of the magnetic flux density axial component B z
in the plane of the coil generated by the actual copper coil and by
one of the proposed version of superconducting coil having the
same internal bore. Copper coil characteristics: I op = 8 kA, coil is a
stack of 8 layer connected in series. HTS coil: I op =100 A, coil is
obtained as a stack of 2 layer with 320 turns each
0.8
0.4
0
-0.4
films, even though a clear dependence of J c
(0) values on the BZO content cannot be established. The results,
as shown in figure 4.11, revealed an improvement in the pinning efficiency if compared with pure YBCO
samples, as evidenced by the upward shift of the irreversibility field value (from 6.8 T to 8 T) and the increase
in the maximum pinning force density (from 4.5 to 11.5 GN/m 3 ) measured at 77 K. This result is consistent
with the presence of the BZO nano–particles acting as additional pinning sources.
Conceptual design of YBCO coil for the toroidal magnetic system of ISTTOK tokamak
The aim of this activity is to evaluate the possibility of operating ISTTOK tokamak with a HTS toroidal
magnetic field, thus allowing a continuous operation in steady state. The feasibility of a tokamak operating
with HTS is extremely relevant and ISTTOK is the ideal candidate for a meaningful test in this sense, due to
its small size, the possibility to operate in a steady–state inductive operation and therefore at lower cost. A finite
element analysis has been carried out to calculate the distribution of magnetic field components. It is found
that the operation at 77 K with 0.45 Tesla field on plasma axis would require 17 km of 12 mm wide tape
commercially available today. In Figure 4.12 the maps of the magnetic flux density axial component as
obtained from numerical analysis for a single actual copper a) and HTS coil b) are plotted.

096
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2010
chapter 5
inertialfusion
The research activity on the ABC facility is now restarting after a shut down of about one year. New
calibrations of the laser outputs are on going. The scientific program is being revised in the light of a possible
upgrade of the laser power or the pulse time shape. Digitization of the alignment and diagnostics sensor is
under implementation.
As a possible follow up of the ABC activity, the ABC team has designed a 200 J diode pumped laser. According
to this preliminary study, a prototype sector of a diode illuminating head has been designed and is presently
under realisation. The first tests will be devoted to checking the emission spectrum of the diodes under
operation, a dedicated fast time response spectrograph has been commissioned. A provisional cooling system
for a 25–30 diode brick has been designed and is now undergoing further investigation (fig. 5.1).
Figure 5.1 – The ABC target repository, positioning
manipulator and holder

progress report
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097
chapter 6
fusion&industry
In 2010, the construction of ITER has entered an important phase with the calls for tender for the largest
components of the reactor itself and the buildings on the European site in Cadarache (France). The activities
of the Italian Industry Liaison Office for ITER (ILO) have also decisively stepped up in strengthening the
involvement of the Italian industry in the construction of ITER.
Beside the information service announcing all calls for tender published by Fusion For Energy (F4E), several
initiatives were taken with the purpose of raising awareness and spreading information to potential
contractors, which were also assisted in understanding the technical content and the requirements of the above
mentioned calls.
• The ILO organised the Italian industry participation in the Fusion Technology Forum Exhibition in the
frame of the Symposium on Fusion Technology (SOFT 2010), held in Porto (Portugal) in September. The
Forum brought together hundreds of industries and research laboratories from all over Europe.
• A meeting was organized at ENEA Frascati Centre on February 26th to illustrate the European
procurement packages on ITER Remote Handling, with the participation of experts from ITER IO and
F4E. About forty participants from twenty companies attended the meeting.
• A seminar on ITER International Project for Fusion Energy was held during the CERN workshop
Opportunità per le Aziende, held in Turin on October 6th, 2010.
The close collaboration with the F4E Business Intelligence Group has been strengthened. A number of Italian
companies have been introduced in F4E, and to the ILOs of other European Countries in order to create
opportunities for the formation of consortia at a European level. The ILO also promoted the participation of
Italian companies in the F4E Information Meetings on Neutral Beam Test Facility in Padua (Italy, 18
November) and on Buildings for ITER (22 November), and in the First Monaco ITER International Fusion
Energy Days (23–25 November). All these initiatives have received a considerable interest and have been
attended by a growing number of participants.
The participation of Italian industries in F4E competitive tenders has been extremely successful in 2010: at
the end of the year a total of 25 industrial contracts have been awarded to Italian companies for a total value
of more than 500 millions € (on a total of 1000 millions € awarded by F4E since 2007). The major contracts
were related to the supply of critical components and greater technological content that make up the core of
the reactor (vacuum vessel, toroidal field coil winding packs, toroidal and poloidal field coil conductors,
divertor targets, power supplies (fig. 6.1).

098
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2010 fusion & industry (cont’d.)
Figure 6.1 – Researchers of the ENEA Fusion Technical
Unit visiting the Walter Tosto Spa in Chieti (Italy). Walter
Tosto with Ansaldo Nucleare and Mangiarotti has been
awarded the contract for the supply of the ITER vacuum
vessel. In the picture, together with ENEA
representatives, the company Business Development
Manager and some of the fifteen young engineers hired
by Walter Tosto for the realization of the vacuum
chamber of ITER. They all are enthusiast for the
opportunity they have given to work for a large
international project, highly technological and
innovative, and with strict quality management: a unique
experience for their professional development, and a
positive example in the current youth employment crisis
In the framework of Euratom – Fusion, the European Commission has established a forum dedicated to the
development of innovative fusion reactor technologies with the involvement of industry. The Fusion Industry
Innovation Forum will aim to involve the industry in elaborating the fusion energy development strategy, with
a view to its implementation in the time specified by the SET Plan. In particular, it will provide the tool to
develop a technology roadmap, produce the required industrial innovation, increase opportunities to transfer
know–how between research laboratories and industry, and develop the necessary human capacity. As a first
step in this direction, it proceeded with the consultation of industry representatives in various EU countries.
In this context, in March 2010, ENEA organized a meeting at its headquarter, with the participation of a
representative of the European Commission – Euratom and the major Italian industry players in the field of
fusion and energy in general, with the aim of gathering views and suggestions for the creation of such a forum.
Two Italian representatives are now members of the Fusion Industry Innovation Forum.

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099
chapter 7
Qualityassurance
Throughout 2010 ENEA – Unità Tecnica Fusione (UTFUS) has been carrying on the activities of
establishing, documenting and implementing a Quality Management System (QMS) in accordance with the
ISO 9001:2008 requirements.
The main steps undertaken towards the final implementation have been:
• Start–up of the certification process of
the QMS (pre–audit and stage 1 audit
performed)
• Definition of the certification scope of
the QMS as follows: “Design,
development and experimental tests
of components and systems for
magnetic thermonuclear fusion plants,
including construction of related test
prototypes. Advanced studies to
support magnetic thermonuclear
fusion”
• Determination of the main processes
needed for the QMS and their
application. They are: i) experimental
activities, ii) design activities, iii)
construction of plants or
components/prototypes usually
designed inside UTFUS
• Issue of the quality management
system documentation including: a)
the documented statements of a
Table 7.I – List of the prodcedure
Procedure
PGQ 01
PGQ 02
PGQ 03
PGQ 04
PGQ 05
PSO 01
PSO 02
PSO 03
PSO 04
PSO 05
PSO 06
PSO 07
PSO 08
PSO 09
PSO 10
quality policy and quality objectives; b) the Quality Manual; c) 15 documented procedures; some (5)
required by the ISO 9001:2008 and others (10) customised according to type of activities, complexity of
processes and their interactions; d) work instructions.
• Communication to the staff about establishment, documentation and implementation of the quality
management system, carried out through 8 seminars (involving 94% of the staff)
• Establishment and completion of internal audits
• Data collection of instrumentation and reference codes.
• Collection and analysis of data aiming to demonstrate the suitability and effectiveness of the QMS and to
evaluate where continual improvement can be pursued.
The list of 15 documented procedures is reported in the following table 7.I. The entire QMS documentation
is available to all staff at the ENEA UTFUS dedicated web page http://www.fusione.enea.it/
PROJECTS/sgq/index.html.en
Title
Management and control of documents
Internal audits
Management of non conformities
Management of preventive and corrective actions
Management and control of records
Management of Human Resources
Management of experimental activities
Management of design activities
Management of construction of prototypes or plants
Procurement/supplying and management of supplies
Management of measuring instruments
Management of special processes
Data analysis
Guideline for the preparation of quality plans
Supplying, storage and traceability of consumables

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chapter 8
publicationsandevents
8.1 Publications
Articles
A. BIERWAGE, L. CHEN, F. ZONCA: Pressure–gradient–induced Alfvén eigenmodes: I. Ideal MHD and finite ion Larmor radius effects,
Plasma Phys. Control. Fusion 52, 015004 (2010)
A. BIERWAGE, L. CHEN, F. ZONCA: Pressure–gradient–induced Alfvén eigenmodes: II. Kinetic excitation with ion temperature gradient,
Plasma Phys. Control. Fusion 52, 015005 (2010)
R. VILLARI, A. CUCCHIARO, B. ESPOSITO, D. MAROCCO, F. MORO, L. PETRIZZI, A. PIZZUTO, G. BROLATTI: Neutronic analysis of FAST,
IEEE Trans. Plasma Sci. 38, 3, 406–413 (2010)
A. CUCCHIARO, R. ALBANESE, G. AMBROSINO, G. BROLATTI, G. CALABRÒ, V. COCILOVO, A. COLETTI, R. COLETTI, P. COSTA, P. FROSI, F.
CRESCENZI, F. CRISANTI, G. GRANUCCI, G. MADDALUNO, V. PERICOLI–RIDOLFINI, A. PIZZUTO, C. RITA, G. RAMOGIDA: Conceptual design
of the FAST load assembly, Fusion Eng. Des. 85, 174–180 (2010)
C. RIZZELLO, F. BORGOGNONI, T. PINNA, S. TOSTI: Review of tritium confinement and atmosphere detritiation system in hot cells
complex, Fusion Eng. Des. 85, 58–63 (2010)
M. SORBA, D. MILANESIO, R. MAGGIORA. A.A. TUCCILLO: Optimization of the FAST ICRF antenna using TOPICA code, Fusion Eng. Des.
85, 161–168 (2010)
Y. HYROOKA, G. MAZZITELLI, S.V. MIRNOV, M. ONO, M. SHIMADA and F.L. TABARES: Conference report on the 1st International Workshop
on Li–applications to boundary control in fusion devices, Nucl. Fusion 50, 077001–077004 (2010)
B. ESPOSITO, F. MURTAS, R. VILLARI, M. ANGELONE, D. MAROCCO, M. PILLON, S. PUDDU: Design of a GEM–based detector for the
measurement of fast neutrons, Nucl. Instrum. Meth. Phys. Res. A 617, 155–157 (2010)
A. DELLA CORTE, V. CORATO, A. DI ZENOBIO, C. FIAMOZZI ZIGNANI, L. MUZZI, G.M. POLLI, L. RECCIA, S. TURTÙ, P. BRUZZONE, E.
SALPIETRO, A. VOSTENER: Successful performances of the EU–AltTF sample, a large size Nb 3 Sn cable-in-conduit conductor with
rectangular geometry, Supercond. Sci. Technol. 23, 045028–045033 (2010)
S. ALMAVIVA, M. MARINELLI, E. MILANI, G. PRESTOPINO, A. TUCCIARONE, C. VERONA, G. VERONA RINATI, M. ANGELONE, M. PILLON:
Improved performance in synthetic diamond neutron detectors: application to boron neutron capture therapy, Nucl. Instrum. Meth. Phys.
Res. A 612, 580–582 (2010)
D. PACELLA, G. PIZZICAROLI, D. MAZON, P. MALARD: A method for detecting layers or dust deposited on tokamak surfaces, Fusion Sci.
Technol. 57, 2, 142–151 (2010)
E. VISCA, A. PIZZUTO, S. ROCCELLA: Tecnologie per la costruzione di componenti ad alto flusso termico per reattori a fusione, Proc. 28th
UIT Heat Transfer Congress, Eds. M. Pilotelli and G.P. Beretta, Univesità di Brescia, p. 385–390 (2010)
Y. SADEGHI, G. RAMOGIDA, L. BONCAGNI, C. D’EPIFANIO, V. VITALE, F. CRISANTI, L. ZACCARIAN: Real–time reconstruction of plasma
equilibrium in FTU, IEEE Trans. Plasma Sci. 38,3, 352-358 (2010)
L. MUZZI, V. CORATO, G. DE MARZI, A. DI ZENOBIO, C. FIAMOZZI ZIGNANI, L.RECCIA, S. TURTÙ, A. DELLA CORTE, P. BARABASCHI, M.
PEYROT, P. BRUZZONE, B. STEPANOV: The JT–60SA toroidal field conductor reference sample: manufacturing and test results, IEEE
Trans. Appl. Supercond. 20,3, 442–446 (2010)
A. DI ZENOBIO, A. DELLA CORTE, L. MUZZI, A. BRAGAGNI, A. TANGUENZA, D. VALORI, A. BALDINI, D. BESSETTE, A. DEVRED, A. VOSTNER:
Conductor manufacturing of the ITER TF full–size performance samples, IEEE Trans. Appl. Supercond. 20,3, 1412–1415 (2010)

publications and events (cont’d.)
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101
ZHIYONG QIU, F. ZONCA, LIU CHEN: Nonlocal theory of energetic–particle–induced geodesic acoustic mode, Plasma Phys. Control.
Fusion 52, 095003 (13pp) (2010)
L. BONCAGNI, S. GALEANI, G.GRANUCCI, G. VARANO, V. VITALE, L. ZACCARIAN: Using dynamic input allocation for elongation control at
FTU, Fusion Eng. Des. 85, 3-4, 443–446 (2010)
A. PIZZUTO, F. GNESOTTO, M. LONTANO, R. ALBANESE, G. AMBROSINO, M.L. APICELLA, A. BRUSCHI, M. BARUZZO, G. CALABRO’, A. CARDINALI,
R. CESARIO, F. CRISANTI, V. COCILOVO, A. COLETTI, R. COLETTI, P. COSTA, S. BRIGUGLIO, P. FROSI, F. CRESCENZI, V. COCCORESE, A. CUCCHIARO,
C. DI TROIA, B. ESPOSITO, G. FOGACCIA, E. GIOVANNOZZI, G. GRANUCCI, G. MADDALUNO, R. MAGGIORA, M. MARINUCCI, D. MAROCCO, P.
MARTIN, G. MAZZITELLI, F. MIRIZZI, S. NOWAK, R. PACCAGNELLA, L. PANACCIONE, G. L. RAVERA, F. ORSITTO, V. PERICOLI RIDOLFINI, G.
RAMOGIDA, C. RITA, M. SCHNEIDER, A. TUCCILLO, R. ZAGÓRSKI, M. VALISA, R. VILLARI, G. VLAD, F. ZONCA: The Fusion Advanced Studies Torus
(FAST): a proposal for an ITER satellite facility in support of the development of fusion energy, Nucl. Fusion 50, 095005 (15 pp) (2010)
P. MICOZZI, F. ALLADIO, A. MANCUSO, F. ROGIER: Ideal MHD stability limits of the PROTO–SPHERA configuration, Nucl. Fusion 50,
095004 (10 pp) (2010)
A. BERTOCCHI, L. BONCAGNI, C. CENTIOLI, F. IANNONE, M. PANELLA, M. VELLUCCI, V. VITALE: Open source solutions in control and data
acquisition systems: FTU case studies, Fusion Eng. Des. 85, 3-4, 321–324 (2010)
B. GUILLERMINET, I. CAMPOS PLASENCIA, M. HAEFELE, F. IANNONE, A. JACKSON, G. MANDUCHI, M. PLOCIENNIK, E. SONNENDRUCKER,
P. STRAND, M. OWSIAK, EUROPEAN TASK FORCE ON INTEGRATED TOKAMAK MODELLING ACTIVITY: High performance computing tools
for the integrated tokamak modelling project, Fusion Eng. Des. 85, 3–4, 388–393 (2010)
V.F. PAIS, S. BALME, H.S. AKPANGNY, F. IANNONE, P. STRAND: Enabling remote access to projects in a large collaborative environment,
Fusion Eng. Des. 85, 3–4, 633–636 (2010)
A. BIANCALANI, LIU CHEN, F. PEGORARO, F. ZONCA: Continuous spectrum of shear Alfvén waves within magnetic islands, Phys. Rev.
Lett. 105,9, 095002(4) (2010)
C. CASTALDO, A. CARDINALI: Tritium minority heating with mode conversion of fast waves, Phys. of Plasmas 17, 072513 (5 pp) (2010)
X. WANG, F. ZONCA, L. CHEN: Theory and simulation of discrete kinetic beta induced Alfvén eigenmode in tokamak plasmas, Plasma
Phys. Control. Fusion 52, 115005 (12pp) (2010)
G.L. RAVERA, R. MAGGIORA, F. MIRIZZI, A.A. TUCCILLO: Performance of different load tolerant external matching unit for the FAST–ICRH
system, Proc. of the 40th European Microwave Conference, 1198–1201 (2010)
A. VANNOZZI, A. ANGRISANI ARMENIO, A. AUGIERI, G. CELENTANO, V. GALLUZZI, A. MANCINI, T. PETRISOR, A. RUFOLONNI, GY.
THALMAIER: Development and characterization of cube–texture Ni–Cu–Co substrates for YBCO coated conductors, Acta Materialia 58,
910–918 (2010)
G. CELENTANO, A. AUGIERI, A. MAURETTI, A.VANNOZZI, A. ANGRISANI ARMENIO, V. GALLOZZI, S. GAUDIO, A. MANCINI, A. RUFOLONI, I.
DAVOLI, C. DEL GAUDIO, F. NANNI: Electrical and mechanical characterization of coated conductors lap joint, IEEE Trans. Appl.
Supercond. 20, 3, 1549–1552 (2010)
V. CORATO, L. MUZZI, A. AUGIERI, U. BESI VETRELLA, C. FIAMOZZI ZIGNANI, A. RUFOLONI, A. DELLA CORTE: Measurement of the
transversal resistivity in superconducting strands with 4–probe technique, IEEE Trans. Appl. Supercond. 20, 3, 1630–1633 (2010)
A. AUGIERI, G. CELENTANO, V. GALLUZZI, A. MANCINI, A. RUFOLONI, A. VANNOZZI, A. ANGRISANI ARMENIO, T. PETRISOR, L. CIONTEA, S. RUBANOV,
E. SILVA, N. POMPEO: Pinning analyses on epitaxial YBa 2 Cu 3 O 7-‰ film with BaZrO 3 inclusions, J. Appl. Phys. 108, 063906,1–5 (2010)
M. ANGELONE, M. PILLON, M. MARINELLI, E. MILANI, G. PRESTOPINO, C. VERONA, G. VERONA-RINATI, I. COFFEY, A. MURARI, N. TARTONI

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AND JET–EFDA CONTRIBUTORS: Single crystal artificial diamond detectors for VUV and soft x–ray measurements on JET thermonuclear
fusion plasma, Nucl. Instrum. Meth. Phys. Res. A 623, 726–730 (2010)
R. CESARIO, L. AMICUCCI, C. CASTALDO, L. PANACCIONE, G. CALABRÒ, A. CARDINALI, C. CIANFARANI, M. MARINUCCI, C. MAZZOTTA,
V. PERICOLI, O. TUDISCO, FTU TEAM: Current drive at plasma densities required for thermonuclear reactors, Nature Commun. 1:
doi:10.1038/ncomms1052 (2010)
R. DE ANGELIS, M. BARUZZO, P. BURATTI, B. ALPER, L. BARRERA, A. BOTRUGNO, M. BRIX, L. FIGINI, A. FONSECA, C. GIROUD, N.
HAWKES, D. HOWELL, E. DE LA LUNA, F. ORSITTO, V. PERICOLI, E. RACHLEW, O. TUDISCO, and JET–EFDA CONTRIBUTORS: Localization
of MHD modes and consistency with q profiles in JET, Nucl. Instrum. Meth. Phys. Res. A 623, 734–737 (2010)
R. DE ANGELIS, L. DI MATTEO: Observations and analysis of FTU plasmas by video cameras, Nucl. Instrum. Meth. Phys. Res. A 623,
815–817 (2010)
M. GELFUSA, M. BROMBIN, P. GAUDIO, A. BOBOC, A. MURARI, F.P. ORSITTO, JET–EFDA CONTRIBUTORS: Modelling of the signal
processing electronics of JET interferometer–polarimeter, Nucl. Instrum. Meth. Phys. Res. A 623, 660–663 (2010)
J.R. MARTIN–SOLIS, R. SANCHEZ, B. ESPOSITO: Experimental observation of increased threshold electric field for runaway generation
due to synchroton radiation losses in the FTU tokamak, Phys. Rev. Lett. 105, 185002(1–4) (2010)
F.P. ORSITTO, A. BIBOC, P. GAUDIO, M. GELFUSA, E. GIOVANNOZZI, C. MAZZOTTA, A. MURARI, and JET–EFDA CONTRIBUTORS: Mutual interaction
of Faraday rotation and Cotton Mouton phase shift in JET polarimetric measurements, Rev. Sci. Instrum. 81, 10D533(1–4) (2010)
A. ROMANO, D. PACELLA, D. MAZON, F. MURTAS, P. MALARD, L. GABELLIERI, B. TILIA, V. PIERGOTTI, G. CORRADI: Characterization of a
2D soft–x ray tomography camera with discrimination in energy bands, Rev. Sci. Instrum. 81, 10E523(1–3) (2010)
E. GIOVANNOZZI, M. BEURSKENS, M. KEMPENAARS, R. PASQUALOTTO, A. RYDZY and JET–EFDA CONTRIBUTORS: Detection of dust on
JET with the high resolution Thomson scattering systems, Rev. Sci. Instrum. 81, 10E131(1–3) (2010)
G. MONDONICO, B. SEEBER, C. SENATORE, R. FLUKIGER, V. CORATO, G. DE MARZI, L. MUZZI: Improvement of electromechanical
properties of an ITE internal tin Nb 3 Sn wire, J. Appl. Phys. 108, 093906 (2010)
F. BORGOGNONI, D. DEMANGE, L. DORR, S. TOSTI, S. WELTE: Processing test of an upgraded mechanical design for PERMCAT reactor,
Fusion Eng. Des. 85, issue 10–12, 2171–2175 (2010)
C. FIAMOZZI ZIGNANI, V. CORATO, L. MUZZI, A. DELLA CORTE: Numerical simulation of Nb 3 Sn strands performances with bending strain
and comparison with experimental measurements, IEEE Trans. Appl. Supercond. 20, 3, 1432–1435 (2010)
O. TUDISCO, C. MAZZOTTA, A. BOTRUGNO, G. MAZZITELLI, M.L. APICELLA, G. APRUZZESE, D. FRIGIONE, L. GABELLIERI, A. ROMANO AND FTU
TEAM: Peaked density profiles and MHD activity on FTU in lithium dominated discharges, Fusion Eng. Des. 85,6, 902–909 (2010)
A. CARDINALI, V. FUSCO: Analytical and numerical studies of the cold electromagnetic LH wave equation in the mode converzion
regime, J. Physics: Conf. Ser. 260, 012007 (2010)
F. ZONCA, A. BIANCALANI, I. CHAVDAROVSKI, L. CHEN, C. DI TROIA, X. WNG: Kinetic structures of shear Alfvén and acoustic wave spectra
in burning plasmas, J. Physics: Conf. Ser. 260, 012022 (2010)
R. CESARIO, L. AMICUCCI, C. CASTALDO, D. DE ARCANGELIS, M. FERRARI, M. MARINUCCI, F. NAPOLI, E. POLLURA, L. PANACCIONE, A.A.
TUCCILLO, and FTU TEAM: Parametric instability and lower hybrid current drive at plasma densities required for thermonuclear reactors,
J. Physics: Conf. Ser. 260, 012008 (2010)
A. BIANCALANI, LIU CHEN, F. PEGORARO, F. ZONCA: Shear Alfvén wave continuous spectrum within magnetic island, Phys. of Plasmas
17, 122106(1–7) (2010)
V. PERICOLI RIDOLFINI, G. CALABRÒ, E. GIOVANNOZZI, L. PANACCIONE, A.A. TUCCILLO: LHCD efficiency and scattering by density
fluctuations at the plasma edge, Proc. of the 37th EPS Conference on Plasma Physics, ECA Vol. 34A, P5.176 (2010)
G. CALABRÒ, D.C. MCDONALD, M. BEURSKENS, C.F. MAGGI, I. DAY, E. DE LA LUNA, T. EICH, N. FEDORCZAK, O. FORD, E. GIOVANNOZZI, C.
GIROUD, P. GOHIL, M. LENNHOLM, P.J. LOMAS, J. LONNROTH, G.P. MADDISON, I. NUNES, V. PERICOLI RIDOLFINI, F. RYTER, G. SAIBENE, R.
SARTORI, W. STUDHOLME, E. SURREY, I. VOITSEKHOVITCH, K.D. ZASTROW and JET–EFDA CONTRIBUTORS: H–Mode threshold studies in
helium–4 JET plasmas, Proc. of the 37th EPS Conference on Plasma Physics, ECA Vol. 34A, O5.127 (2010), ISBN: 2–914771–62–2
V. PERICOLI–RIDOLFINI: Increased particle and energy transport induced by LH waves in the tokamak scrape–off layer plasma, Proc. of
the 37th EPS Conference on Plasma Physics, ECA Vol. 34A, P5.177 (2010), ISBN: 2-914771-62-2
M.L. APICELLA, G. APRUZZESE, G. MAZZITELLI, A.G. ALEKSEYEV, A. BOTRUGNO, P. BURATTI, R. DE ANGELIS, H. KROEGLER, V.B.
LAZAREV, S.V. MIRNOV, R. ZAGORSKI: Radiative layer with a liquid ithium limiter on FTU, Proc. of the 37th EPS Conference on Plasma
Physics, ECA Vol. 34A, P2.114 (2010), ISBN: 2–914771–62–2

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G. CALABRO’, E. GIOVANNOZZI, R. RAMOGIDA, F. CRISANTI, C. LABATE, M. MATTEI, P. MICOZZI, G. VLAD: Modelling of FAST equilibrium
configurations by a toroidal multipolar expansion code using Kepler workflows, Proc. of the 37th EPS Conference on Plasma Physics,
ECA Vol. 34A, P4.157 (2010), ISBN: 2–914771–62–2
A. BIERWAGE, L. CHEN, F. ZONCA: Pressure–gradient–induced Alfvén eigenmodes destabilized by ion temperature gradient, Proc. of the
37th EPS Conference on Plasma Physics, ECA Vol. 34A, P4.179 (2010), ISBN: 2–914771–62–2
A. JACCHIA, S. CIRANT, F. DE LUCA, P. BURATTI, E. LAZZARO, O. TUDISCO, C. MAZZOTTA, G. CALABRO’ G. RAMOGIDA, C. CIANFARANI,
D. MAROCCO, G. GROSSETTI, G. GRANUCCI, O. D’ARCANGELO, W. BIN and FTU and ECRH TEAM: Density response to modulated EC
heating in FTU tokamak, Proc. of the 37th EPS Conference on Plasma Physics, ECA Vol. 34A, P1.045 (2010), ISBN: 2–914771–62–2
A. BOTRUGNO, P. BURATTI, F. ZONCA: Comparison between BAE observations at FTU and theoretical models, Proc. of the 37th EPS
Conference on Plasma Physics, ECA Vol. 34A, P4.110 (2010), ISBN: 2–914771–62–2
M. GONICHE, Y. BARANOV, P. BONOLI, G. CALABRO’, A. CARDINALI, C. CASTALDO, R. CESARIO, J. DECKER, J. GARCIA, G. GIRUZZI, E.
JOFFRIN, A. HUBBARD, K. KIROV, X. LITAUDON, J. MAILLOUX, R. PARKER: Lower hybrid current drive for the steady state scenario, Plasma
Phys. Control. Fusion 52, 124031 (18pp) (2010)
B. BAIOCCHI, G. CALABRÒ, L. LAURO–TARONI, P. MANTICA, A. CARDINALI, G. CORRIGAN, F. CRISANTI, D. FARINA, L. FIGINI, G. GIRUZZI,
T. JOHNSON, M. MARINUCCI, V. PARAIL: Predictive modelling of H–mode and steady–state scenarios in FAST, Proc. of the 37th EPS
Conference on Plasma Physics, ECA Vol. 34A, P1.007 (2010), ISBN: 2–914771–62–2
M. BARUZZO, M. SCHNEIDER, V. BASIUK, A. CARDINALI, M. MARINUCCI, J.F. ARTAUD, O. ASUNTA, T. BOLZONELLA, F. CRISANTI, R.
DUMONT, G. GIRUZZI, F. IMBEAUX, P. MANTICA, M. VALISA, F. ZONCA: First NBI configuration study for FAST proposal, Proc. of the 37th
EPS Conference on Plasma Physics, ECA Vol. 34A, P5.142 (2010), ISBN: 2–914771–62–2
A. BIANCALANI, L. CHEN, F. PEGORARO, F. ZONCA: Continuous spectrum of shear Alfvén waves inside magnetic islands, Proc. of the
37th EPS Conference on Plasma Physics, ECA Vol. 34A, P4.109 (2010), ISBN: 2–914771–62–2
L. FRASSINETTI, M.N.A. BEURSKENS, D.C. MCDONALD, J. HOBIRK, D. WARZOSO, P. BURATTI, F. CRISANTI, C. CHALLIS, E. GIOVANNOZZI, E.
JOFFRIN, P. LOMAS, J. LONNROTH, C. MAGGI, I. NUNES, S. SAARELMA, G. SAIBENE and JET EFDA CONTRIBUTORS: Pedestal confinement in
hybrid versus baseline plasmas in JET, Proc. of the 37th EPS Conference on Plasma Physics, ECA Vol. 34A, P1.1031 (2010)
M. ANGELONE, M. PILLON, R. FACCINI, D. PINCI, W. BALDINI, R. CALABRESE: Silicon photo–multiplier radiation hardness tests with
a beam controlled neutron source, Nucl. Instrum. Meth. Phys. Res. A 623, 921–926 (2010)
P. BATISTONI, M. ANGELONE, P. CARCONI, U. FISCHER, K. FLEISCHER, K. KONDO, A. KLIX, I. KODELI, D. LEICHTLE, L. PETRIZZI, M. PILLON,
W. POHORECKI, M. SOMMER, A. TRKOV, R. VILLARI. Neutronics experiments on HCPB and HCLL TBM mock–ups in preparation of
nuclear measurements in ITER, Fusion Eng. Des. 85, 1675–1680 (2010)
C. NARDI, L. BETTINALI, M. LABANTI: Long term characterization on unidirectional fiberglass for ITER pre–compression rings, Fusion Eng.
Des. 85, 2241–2244 (2010)
E. VISCA, E. CACCIOTTI, S. LIBERA, A. MANCINI, A. PIZZUTO, S. ROCCELLA, B. RICCARDI, F. ESCOURBIAC, G.P. SANGUINETTI: Manufacturing
and testing of reference samples for ITER divertor acceptance criteria definition, Fusion Eng. Des. 85, 1986–1991 (2010)
T. PINNA, L.C. CADWALLADER, G. CAMBI, S. CIATTAGLIA, S. KNIPE, F. LEUTERER, A. MALIZIA, P. PETERSEN, M.T. PORFIRI, F. SAGOT, S.
SCALES, J. STOBER, J.C. VALLET, T. YAMANISHI: Operating experiences from existing fusion facilities in view of ITER safety and reliability,
Fusion Eng. Des. 85, 1410–1415 (2010)
T. PINNA, G. CAMBI, S. KNIPE and JET–EFDA CONTRIBUTORS: JET operating experience: global analysis of tritium plant failure, Fusion
Eng. Des. 85, 1396–1400 (2010)
G. CAMBI, D.G. CEPRAGA, L. DI PACE, F. DRUYTS, V. MASSAUT: The potential presence and minimisation of plutonium within the irradiated
beryllium in fusion power plants, Fusion Eng. Des. 85, 1139–1142 (2010)
D. LEICHTLE, U. FISCHER, I. KODELI, R.L. PEREL, A. KLIX, P. BATISTONI, R. VILLARI: Sensitivity and uncertainty analyses of the HCLL
mock–up experiment, Fusion Eng. Des. 85, 1724–1727 (2010)
S. NICOLLET, B. LACROIX, L. ZANI, P. HERTOUT, CH. PORTAFAIX, R. VILLARI: Development of an extended thermo-hydraulic simulation
tool for fusion magnet design study – Application to the initial versions of JT–60SA TF coils layout, Cryogenic 50,1, 18–27 (2010)
A. KLIX, P. BATISTONI, R. BOTTGER, D. LEBRUN-GRANDIE, U. FISCHER, J. HENNIGER, D. LEICHTLE, R. VILLARI: Measurement and analysis
of neutron flux spectra in a neutronics mock–up of the HCLL test blanket module, Fusion Eng. Des. 85, 1803–1806 (2010)
A. EKEDAHL, L. DELPECH, M. GONICHE, D. GUILHEM, J. HILLAIRET, M. PREYNAS, P.K. SHARMA, J. ACHARD, Y.S. BAE, X. BAI, C. BALORIN,
Y. BARANOV, V. BASIUK, A.BÉCOULET, J. BELO, G. BERGER-BY, S. BRÉMOND, C. CASTALDO, S. CECCUZZI, R. CESARIO, E. CORBEL, X.
COURTOIS, J. DECKER, E. DELMAS, X. DING, D. DOUAI, C. GOLETTO, J.P. GUNN, P. HERTOUT, G.T. HOANG, F. IMBEAUX, K.K. KIROV, X.

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LITAUDON, R. MAGNE, J. MAILLOUX, D. MAZON, F. MIRIZZI, P. MOLLARD, P. MOREAU, T. OOSAKO, V. PETRZILKA, Y. PEYSSON, S. POLI, M.
PROU, F. SAINT–LAURENT, F. SAMAILLE, B. SAOUTIC: Valdiation of the ITER–relevant passive–active–multijunction LHCD laucher on long
pulses in Tore Supra, Nucl. Fusion 50, 112002(5pp) (2010)
A. MILOVANOV: Self–organized criticality with a fishbone–like instability cycle, Europhys. Lett. 89, 60004 (2010)
G. MAZZITELLI, M.L. APICELLA, G. APRUZZESE, R. DE ANGELIS, D. FRIGIONE, E. GIOVANNOZZI, L. GABELLIERI, G. GRANUCCI, C.
MAZZOTTA, M. MARINUCCI, A. ROMANO, O. TUDISCO, FTU TEAM and ECRH TEAM: Review of FTU results with the liquid lithium limiter,
Fusion Eng. Des. 85, 6, 902–909 (2010)
F. CRISANTI, R. ALBANESE, F. ARTAUD, B. BAIOCCHI, M. BARUZZO, V. BASIUK, A. BIERWAGE, R. BILATO, T. BOLZONELLA, M. BRAMBILLA, S.
BRIGUGLIO, G. CALABRO’, A. CARDINALI, G. CORRIGAN, A. CUCCHIARO, C. DI TROIA, D. FARINA, L. FIGINI, G. FOGACCIA, G. GIRUZZI, G.
GRANUCCI, F. IMBEAUX, T. JOHNSON, L. LAURO TARONI, R. MAGGIORA, P. MANTICA, D. MILANESIO, V. PARAIL, V. PERICOLI–RIDOLFINI, A.
PIZZUTO, S. PODDA, G. RAMOGIDA, M. SANTINELLI, M. SCHNEIDER, A. TUCCILLO, M. VALISA, R. VILLARI, B. VIOLA, G. VLAD, X. WANG, F.
ZONCA: Scenario development for FAST in the view of ITER and DEMO, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 - Paper
FTP/2–4 (2010) online at: http://www-pub.iaea.org/mtcd/meetings/ cn180_papers.asp
A. CARDINALI, M. BARUZZO, C. DI TROIA, M. MARINUCCI, A. BIERWAGE, G. BREYIANNIS, S. BRIGUGLIO, G. FOGACCIA, G. VLAD, X. WANG,
F. ZONCA, V. BASIUK, R. BILATO, M. BRAMBILLA, F. IMBEAUX, S. PODDA, M. SCHNEIDER: Energetic particle physics in FAST H–mode
scenario with combined NNBI and ICRH, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 – Paper THW/P7-04 (2010) online
at: http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
P. MANTICA, B. BAIOCCHI, G. CALABRO’, L. LAURO TARONI,, O. ASUNTA, M. BARUZZO, A. CARDINALI, G. CORRIGAN, F. CRISANTI, D.
FARINA, L. FIGINI, G. GIRUZZI, F. IMBEAUX, T. JOHNSON, M. MARINUCCI, V. PARAIL, A. SALMI, M. SCHNEIDER, M. VALISA: Physics based
modelling of H–mode and advanced tokamak scenarios for FAST: analysis of the role of rotation in predicting core transport in future
machines, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper THC/P2-05 (2010) online at:
http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
M. TARDOCCHI, A. BRUSCHI, D. MAROCCO, N. NOCENTE, G. CALABRO’, A. CARDINALI, F. CRISANTI, B. ESPOSITO, L. FIGINI, G. GORINI,
G. GROSSETTI, G. GROSSO, M. LONTANO, S. NOVAK, F. ORSITTO, U. TARTARI, O TUDISCO: Production and diagnosis of energetic particle
in FAST, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXW/P7–26 (2010) online at: http://www–pub.iaea.org/mtcd/
meetings/cn180_papers.asp
G. MAZZITELLI, M.L. APICELLA, D. FRIGIONE, G. MADDALUNO, M. MARINUCCI, C. MAZZOTTA, V. PERICOLI–RIDOLFINI, M. ROMANELLI, G.
SZEPESI, O. TUDISCO AND FTU TEAM: FTU results with the liquid lithium limiter, Proc. of the 23rd IAEA Fusion Energy Conference, CN-
180 -Paper EXC/6–3 (2010) online at: http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
B. ESPOSITO, G. GRANUCCI, M. MARASCHEK, S. NOWAK, A. GUDE, V. IGOCHINE, R. MCDERMOTT, F. POLI, J. STOBER, W. SUTTROP, W.
TREUTTERER, H. ZOHM AND ASDEX UPGRADE TEAM: Avoidance of discruption at high ‚β n in ASDEX Upgrade with off–axis ECRH, Proc.
of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXW/10–2Ra (2010) online at: http://www-pub.iaea.org/mtcd/
meetings/cn180_papers.asp
R. DE ANGELIS, F. ORSITTO, M. BARUZZO, P. BURATTI, B. ALPER, L. BARRERA, A. BOTRUGNO, M. BRIX, K. CROMBE’, L. FIGINI, A. FONSECA, C.
GIROUD, N. HAWKES, D. HOWELL, E. DE LA LUNA, V. PERICOLI-RIDOLFINI, E. RACHLEW, O. TUDISCO AND JET–EFDA CONTRIBUTORS:
Determination of q profiles in JET by consistency of motional stark effect and MHD mode localization, Proc. of the 23rd IAEA Fusion Energy
Conference, CN-180 - Paper EXS/P2-03 (2010) online at: http://www-pub.iaea.org/mtcd/meetings/cn180_papers.asp
R. DE ANGELIS, F. ORSITTO, M. BARUZZO, P. BURATTI, B. ALPER, L. BARRERA, A. BOTRUGNO, M. BRIX, K. CROMBE’, L. FIGINI, A.
FONSECA, C. GIROUD, N. HAWKES, D. HOWELL, E. DE LA LUNA, V. PERICOLI-RIDOLFINI, E. RACHLEW, O. TUDISCO AND JET–EFDA
CONTRIBUTORS: Particle deposition, transport and fuelling in pellet injection experiments at JET, Proc. of the 23rd IAEA Fusion Energy
Conference, CN-180 - Paper EXC/P4-05 (2010) online at: http://www-pub.iaea.org/mtcd/meetings/cn180_papers.asp
P. BURATTI, M. BARUZZO, R.J. BUTTERY, C.D. CHALLIS, I.T. CHAPMAN, F. CRISANTI, L. FIGINI, M. GRYAZNEVICH, T.C. HENDER, D.F.
HOWELL, H. HAN, E. JOFFRIN, J. HOBIRK, F. IMBEAUX, O.J. KWON, X. LITAUDON,J. MAILLOUX AND JET–EFDA CONTRIBUTORS: Kink
instabilities in high–beta JET advanced scenarios, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXS/P5-02 (2010)
online at: http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
R. CESARIO, L. AMICUCCI, M.L. APICELLA, G. CALABRO’, A. CARDINALI, C. CASTALDO, C. CIANFARANI, D. FRIGIONE, M. MAZZITELLI, C.
MAZZOTTA, L. PANACCIONE, V. PERICOLI-RIDOLFINI, A.A. TUCCILLO, O. TUDISCO AND FTU TEAM: Lower hybrid current drive at densities
required for thermonuclear reactors, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXW/P7–02 (2010) online at:
http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
A.A. TUCCILLO ON BEHALF OF FTU TEAM AND A. ALEXEYEV, L. AMICUCCI, A. BIANCALANI, A. BIERWAGE, G. BREYANNIS, I.
CHAVDAROVSKI, L. CHEN, F. DE LUCA, Z.O. GUIMARAES–FILHO, A. JACCHIA, E. LAZZARO, F. PEGORRO, M. ROMANELLI, X. WANG:

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Overview of the FTU results, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper OV/4-2 (2010) online at:
http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
G. GRANUCCI, G. RAMPONI, G. CALABRO’, F. CRISANTI, G. RAMOGIDA, W. BIN, A. BOTRUGNO, P. BURATTI, O. D’ARCANGELO, D.
FRIGIONE, G. PUCELLA, A. ROMANO, O. TUDISCO, AND FTU TEAM: Plasma start–up results with EC assisted breakdown on FTU, Proc.
of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXW/P2–03 (2010) online at: http://www–pub.iaea.org/mtcd/
meetings/cn180_papers.asp
D.C. MCDONALD, G. CALABRO’, M. BEURSKENS, I. DAY, E, DE LA LUNA, T. EICH, N. FEDORCZAC, O. FORD, W. FUNDAMENSKI,
C. GIROUD, P. GOHIL, M. LENNHOLM, J. LONNROTH, P.J. LOMAS, G.P. MADDISON, C.F. MAGGI, I. NUNES, G. SAIBENE, R. SARTORI, W.
STUDHOLME, E. SURREY, I. VOITSEKOVITCH, K.D. ZATROW AND JET–EFDA CONTRIBUTORS: JET helium–4 ELMy H–mode studies, Proc.
of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXC/2–4rb (2010) online at: http://www–pub.iaea.org/mtcd/
meetings/cn180_papers.asp
M. BARUZZO, B. ALPER,T. BOLZONELLA, M. BRIX, P. BURATTI, C.D. CHALLIS, F. CRISANTI, E. DE LA LUNA, P.C. DE VRIES, C. GIROUD, N.C.
HAWKES, D.F. HOWELL, F. IMBEAUX, E. JOFFRIN, H.R. KOSLOWSKI, X. LITAUDON, J. MAILLOUX, A.C.C. SIPS, O. TUDISCO AND JET–EFDA
CONTRIBUTORS: Neoclassical tearing mode (NTM) magnetic spectrum and magnetic coupling in JET tokamak, Plasma Phys. Control.
Fusion 52, 075001 (18pp) (2010)
P. MAGET, H. LUTJENS, R. COELHO, B. ALPER, M. BRIX, P. BURATTI, R.J. BUTTERY, E. DE LA LUNA, N. HAWKES, G.T.A. HUYSMANS, J.
JENKINS, C.D. CHALLIS, C. GIROUD, X. LITAUDON, J. MAILLOUX, M. OTTAVIANI AND JET–EFDA CONTRIBUTORS: Modelling of (2,1) NTM
threshold in JET advanced scenarios, Nucl. Fusion 50, 045004 (16pp) (2010)
E. JOFFRIN, C.D. CHALLIS, J. CITRIN, J. GARCIA, J. HOBIRK, I. JENKINS, J. LONNROTH, D.C. MCDONALD, P. MAGET, P. MANTICA, M.
BEURSKENS, M. BRIX, P. BURATTI, F. CRISANTI, L. FRASSINETTI, C. GIROUD, F. IMBEAUX, M. PIOVESAN, F. RIMINI, G. SERGIENKO, A.C.C.
SIPS, T. TALA, I. VOITSEKOVITCH and JET–EFDA CONTRIBUTORS: High confinement hybrid scenario in JET and its significance for ITER,
Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXC/1_1 (2010) online at:
http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
J. MAILLOUX, X. LITAUDON, P.C. DE VRIES, J. GARCIA, I. JENKINS, B. ALPER, YU.BARANOV, M. BARUZZO, M. BEURSKENS, M. BRIX, P.
BURATTI, G. CALABRÒ, R. CESARIO, C.D. CHALLIS, K. CROMBE, O. FORD, D. FRIGIONE, C. GIROUD, M. GONICHE, N. HAWKES, D.
HOWELL, P. JACQUET, E. JOFFRIN, V. KIPTILY, K.K. KIROV, P. MAGET, D.C. MCDONALD, V. PERICOLI–RIDOLFINI, V. PLYUSIN, F. RIMINI, M.
SCHNEIDER, S. SHARAPOV, C. SOZZI, I. VOITSEKOVITCH, L. ZABEO and JET–EFDA CONTRIBUTORS: Towards a steasy-state scenario with
ITER dimensionless parameters in JET, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXC/1_4 (2010) online at:
http://www-pub.iaea.org/mtcd/meetings/cn180_papers.asp
P. MAGET, H. LUTJENS, B. ALPER, M. BARUZZO, M. BRIX, P. BURATTI, R.J. BUTTERY, C. CHALLIS, R. COELHO, E. DE LA LUNA, C. GIROUD,
N. HAWKES, G.T.A. HUYSMANS, I. JENKINS, X. LITAUDON, J. MAILLOUX, N. MELLET, D. MESHCHERIAKOV, M. OTTAVIANI, AND JET–EFDA
CONTRIBUTORS: Non linear MHD modelling of NTMs in JET advanced scenarios, Proc. of the 23rd IAEA Fusion Energy Conference, CN-
180 -Paper EXC/P5–09 (2010) online at: http://www-pub.iaea.org/mtcd/meetings/cn180_papers.asp
P. MANTICA, C. ANGIONI, B. BAIOCCHI, C. CHALLIS, J. CITRIN, G. COLYER, A.C.A. FIGUEIREDO, L. FRASSINETTI, E. JOFFRIN, T. JOHNSON, E.
LERCHE, A.G. PEETERS, A. SALMI, D. STRINTZI, T. TALA, M. TSALAS, D. VAN EESTER, P.C. DE VRIES, J. WEILAND, M. BARUZZO, M.N.A.
BEURSKENS, J.P.S. BIZARRO, P. BURATTI, F. CRISANTI, X. GARBET, C. GIROUD, N. HAWKES, J. HOBIRK, F. IMBEAUX, J. MAILLOUX, V. NAULIN,
C. SOZZI, G. STAEBLER, T.W. VERSLOOT AND JET–EFDA CONTRIBUTORS: A key to improved ion core confinement in the JET tokamak: ion
stiffness mitigation due to combined plasma rotation and low magnetic shear, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -
Paper EXC/9-2 (2010) online at: http://www-pub.iaea.org/mtcd/ meetings/cn180_papers.asp
P.A. POLITZER, C.D. CHALLIS, E. JOFFRIN, T.C. LUCE, M. BEURSKENS, P. BURATTI, F. CRISANTI, J.C. DEBOO, J.R. FERRON, C. GIROUD,
J. HOBIRK, C.T. HOLCOMB, A.W. HYATT, F. IMBEAUX, R.J. JAYAKUMAR, I. JENKINS, J.E. KINSEY, R.J. LA HAYE, D.C. MCDONALD, C.C.
PETTY, F. TURCO, M.R. WADE AND JET–EFDA CONTRIBUTORS: Understanding confinement in advaced inductive scenario plasmas -
dependence on gyroradius and rotation, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXC/P2–06 (2010) online at:
http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
M.P. GRAYAZNEVICH, Y.Q. LIU, T.C. HENDER, D.F. HOWELL, I.T. CHAPMAN, C.D. CHALLIS, S.P. PINCHES, E. JOFFRIN, H.R. KOSLOWSKI, P.
BURATTI, E. SOLANO AND JET–EFDA CONTRIBUTORS: Determination of plasma stability using resonant field amplification in JET, Proc.
of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper EXC/P5–06 (2010) online at: http://www-pub.iaea.org/mtcd/
meetings/cn180_papers.asp
M. ANGELONE, G. AIELLI, S. ALMAVIVA, R. CARDARELI, D. LATTANZI, M. MARINELLI, E. MILANI, M. PILLON, G. PRESTOPINO, R.
SANTONICO, C. VERONA, G. VERONA RINATI: Neutron spectroscopy by means of artificial diamond detectors using a remote read out
scheme, IEEE Trans. Nucl. Sci. 57, 6, 3655–3660 (2010)

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S. VENTICINQUE, R. AVERSA, B. DI MARTINO, R. DONINI, S. BRIGUGLIO, G. VLAD: Management of high performance scientific applications
using mobile agents based services, Scalable Computing: Practice and Experience 11, 2, 149–159 (2010)
C. CASTALDO, S. RATYNSKAIA, M. DE ANGELI, U. DE ANGELIS: On the feasibility of electro–optical detection of dust–impact ionization in
tokamaks, Plasma Phys. Control. Fusion 52, 105003 (6pp) (2010)
S. RATYNSKAIA, M. DE ANGELI, E. LAZZARO, C. MARMOLINO, U. DE ANGELIS, C. CASTALDO, A. CREMONA, L. LAGUARDIA, G. GERVASINI,
G. GROSSO: Plasma fluctuation spectra as a diagnostic tool for submicron dust, Phys. of Plasmas 17, 043703 (2010)
X. WANG, S. BRIGUGLIO, G. FOGACCIA, G. VLAD, C. DI TROIA, F. ZONCA, L. CHEN, A. BIERWAGE, H. ZHANG: Kinetic thermal ions effects
on Alfvénic fluctuations in tokamak plasmas, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper THW/2–4Ra (2010) online
at: http://www–pub.iaea.org/mtcd/meetings/cn180_papers.asp
L. CHEN, W. DENG, Z. LIN, D. SPONG, G.Y. SUN, X. WANG, X.Q.XU, H.S. ZHANG, W.L. ZHANG, A. BIERWAGE, S. BRIGUGLIO, I. HOLOD, G.
VLAD, Y. XIAO, F. ZONCA: Verification of gyrokinetic particle simulation of Alfvén eigenmodes excited by external antenna and by fast
ions, Proc. of the 23rd IAEA Fusion Energy Conference, CN-180 -Paper THW/P7–25 (2010) online at: http://www–pub.iaea.org/mtcd/
meetings/cn180_papers.asp
ZHIYONG QUI, F. ZONCA, L. CHEN: Kinetic theories of geodesic acoustic mode in toroidal plasmas, Proc. of the 23rd IAEA Fusion Energy
Conference, CN-180 - Paper THW/P8–01 (2010) online at: http://www–pub.iaea.org/mtcd/meetings/ cn180_papers.asp
A. MURARI, M. ANGELONE, G. BONHEURE, E. CECIL, T. CRACIUNESCU, D. DARROW, T. EDLINGTON, G. ERICSSON, M. GATU-JOHNSON,
G. GORINI, C. HELLESEN, V. KIPTILY, J. MLYNAR, C. PEREZ VON THUN, M. PILLON, S. POPOVICHEV, B. SYME, M. TARDOCCHI, V.L. ZOITA
AND EFDA–JET CONTRIBUTORS: New developments in the diagnostics for the fusion products on JET preparation for ITER (invited
paper), Rev. Sci. Instrum. 81, 10E136(1–8) (2010)
S. ALMAVIVA, M. MARINELLI, E. MILANI, G. PRESTOPINO, A. TUCCIARONE, C. VERONA, G. VERONA–RINATI, M. ANGELONE, M. PILLON, I.
DOLBNYA, K. SAWHNEY, N. TARTONI: Chemical vopor deposition diamond based multilayered radiation detector: physical analysis of
detection properties, J. Appl. Phys. 107, 014511(1–7) (2010)
S. ALMAVIVA, MARCO MARINELI, E. MILANI, G. PRESTOPINO, A. TUCCIARONE, C. VERONA, G. VERONA–RINATI, M. ANGELONE, M. PILLON:
Extreme UV single crystal diamond Schottky photodiode in planar and transverse configuration, Diamond & Rel. Mat. 19, 78–82 (2010)
S. SANDRI, A. CONIGLIO, A. DANIELE, M. D’ARIENZO, M. PILLON: Radiation shielding for the ITER neutral beam test facility, Proceedings
of the 12th Congress of the Radiation Protection Association (IRPA12), Proceedings Series STI/PUB/1460 Companion CD (2010)
K. SHINOHARA, T. KURKI-SUONIO, D. SPONG, O. ASUNTA, K. TANI, E. STRUMBERGER, S. BRIGUGLIO, S. GUNTER, T. KOSKELA, G.
KRAMER, S. PUTVINSKI, K. HAMAMATSU, ITPA TOPICAL GROUP ON ENERGETIC PARTICLES: 3D effect of ferromagnetic materials on
alpha particle power loads on first wall structures and equilibrium on ITER, Proc. of the 23rd IAEA Fusion Energy Conference, CN–180
– Paper ITR/P1–36 (2010) online at: http://www–pub.iaea.org/mtcd/meetings/ cn180_papers.asp
P. COSTA, C. NERI, M. SPAGNOLO, G. FALCINELLI, F. MICCHETTI: A software platform for the optimization of IVVS viewing performances,
Proceeding of the 2010 EnginSoft International Conference, p.1–19 (2010)
R. VITELLI, L. BONCAGNI, F. MECOCCI, S. PODDA, V. VITALE, L. ZACCARIAN: An anti–windup-based solution for the low current
nonlinearity compensation on the FTU horizontal position controller, Proceedings of the 49th IEEE Conference on Decision and Control,
Atlanta 2010, Article n. 5717837, pp. 2735–2740 (2011) - ISBN 978–142447745–6
G. VARANO, L. BONCAGNI, S. GALEANI, G. GRANUCCI, V. VITALE, L. ZACCARIAN: Results on plasma position and elongation regulation
at FTU using dynamic input allocation, Proceedings of the 49th IEEE Conference on Decision and Control, Atlanta 2010, Article no
5717468, pp 2753–2758 (2011) – ISBN 978–142447745–6
Articles in course of publication
R. CESARIO, L. AMICUCCI, C. CASTALDO, M. DEMPENAARS, J. JACHMICH, J. MAILLOUX, O. TUDISCO AND EFDA-JET CONTRIBUTORS:
Plasma edge density and lower hybrid current drive in JET (Joint European Torus), Plasma Phys. Control. Fusion
R. VITELLI, L. BONCAGNI, F. MECOCCI, S. PODDA, V. VITALE, L. ZACCARIAN: An anti–windup–based solution for the low current
nonlinearity compensation on the FTU horizontal position controller, Proc. 49th IEEE Conference on Decision and Control
G. VARANO, L. BONCAGNI, S. GALEANI, G. GRANUCCI, V. VITALE, L. ZACCARIAN: Plasma position and elongation regulation at FTU using
dynamic allocation, Proc. 49th IEEE Conference on Decision and Control
R. VITELLI, L. BONCAGNI, F. MECOCCI, S. PODDA, V. VITALE, L. ZACCARIAN: An anti–windup–based solution for the low current
nonlinearity compensation on the FTU horizontal position controller, IEEE Trans. Control Syst. Technol.

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2010
107
L. BONCAGNI, Y. SADEGHI, D. CARNEVALE, G. MAZZITELLI, A. NETO, D. PUCCI, F. SARTORI, S. SINIBALDI, V. VITALE, R. VITELLI, L.
ZACCARIAN, S. MONACO, G. ZAMBORLINI: First steps in the FTU migration towards a modular and distributed real–time control
architecture based on MARTe, IEEE Trans. Nucl. Sci.
L. BONCAGNI, A. BARBALACE, Y. SADEGHI, M. POMPEI, L. ZACCARIAN, F. SARTORI: Switched ethernet in synchronized distributed control
systems using RTnet, IEEE Trans. Nucl. Sci.
C. NERI, P. COSTA, M. FERRI DE COLLIBUS, M. FLOREAN, G. MUGNAINI, M. PILLON, F. POLLASTRONE, P. ROSSI: ITER in vessel viewing
system design and assessment activities, Fusion Eng. Des.
E. VISCA, E. CACCIOTTI, A. KOMAROV, S. LIBERA, N. LITUNOVSKY, A. MAKHANKOV, A. MANCINI, M. MEROLA, A. PIZZUTO, B. RICCARDI, S. ROCCELLA:
Manufacturing, testing and post–test examination of ITER divertor vertical target W small scale mock–ups, Fusion Eng. Des.
G. VARANO, L. BONCAGNI, S. GALEANI, G. GRANUCCI, V. VITALE, L. ZACCARIAN: Plasma position and elongation regulation at FTU using
dynamic allocation, IEEE Trans. Control Syst. Technol.
F. MIRIZZI, S. CECCUZZI, S. MESCHINO, J.F. ARTAUD, J.H. BELO, G. BERGER-BY, J.M. BERNARD, A. CARDINALI, C. CASTALDO, R. CESARIO, J.
DECKER, L. DELPECH, A. EKEDAHL, J. GARCIA, P. GARIBALDI, M. GONICHE, D. GUILHEM, G.T. HOANG, J. HUA, Q.Y. HUANG, J. HILLAIRET, F.
IMBEAUX, F. KAZARIAN, S.H. KIM, X. LITAUDON, R. MAGGIORA, R. MAGNE, L. MARFISI, D. MILANESIO, W. NAMKUNG, L. PAJEWSKI, L.
PANACCIONE, Y. PEYSSON, P.K. SHARMA, G. SCHETTINI, M. SCHNEIDER, A.A. TUCCILLO, O. TUDISCO, G. VECCHI, R. VILLARI, K. VULLIEZ:
Contribution to the design of the main transmission line for the ITER relevant LHCD system, Fusion Eng. Des.
S. GAUDIO, G. DE MARZI, G. CELENTANO, A. AUGIERI, V. GALLUZZI, A. MANCINI, A. RUFOLONI, A. VANNOZZI, A. DELLA CORTE, U.
GAMBARDELLA, A. SAGGESE, J.J. JIANG, J. WEISS, E. HELLSTROM, D. LARBALESTIER: The effect of doping on the magnetic properties
in Ba(Fe 1-x Cox) 2 As 2 polycristalline samples, IEEE Trans. Appl. Supercond.
V. CORATO, L. MUZZI, U. BESI VETRELLA, C. FIAMOZZI ZIGNANI, A. DELLA CORTE: Mapping of the transversal resistivity matrix measured
on NbTi and Nb 3 Sn superconducting strands, IEEE Trans. Appl. Supercond.
A. DI ZENOBIO, A. DELLA CORTE, L. MUZZI, G.M. POLLI, L. RECCIA,S. TURTU’, F. CRISANTI, A. CUCCHIARO, A. PIZZUTO, R. VILLARI: FAST:
feasibility analysis for a completely superconducting magnet system, IEEE Trans. Appl. Supercond.
C. FIAMOZZI ZIGNANI, L. MUZZI, S. TURTU’, A. DELLA CORTE: The effect of strand bending on the voltage-current characteristics of
Nb 3 Sn cable–in–conduit conductors, IEEE Trans. Appl. Supercond.
A. MANCINI, G. CELENTANO, A. VANNOZZI, V. GALLUZZI, A. RUFOLONI, A. AUGIERI, S. GAUDIO, A.A. ANGRISANI, I. COLANTONI, I. DAVOLI:
The oxidation behaviour at the Ni–W and CeO 2 interface with and without Pd over layer, IEEE Trans. Appl. Supercond.
L. MUZZI, G. DE MARZI, U. BESI VETRELLA, C. FIAMOZZI ZIGNANI, V. CORATO, A. RUFOLONI, A. DELLA CORTE: Test results of a NbTi wire
for the ITER PF 2–5 magnets: a validation of the 2–pinning components model, IEEE Trans. Appl. Supercond.
G.M. POLLI, L. AFFINITO, A. DELLA CORTE: Heat exchanger design for the 30 kA gas cooled current leads in the ENEA 12 T CICC facility,
IEEE Trans. Appl. Supercond.
L. RECCIA, S. TURTU’, G.M. POLLI, L. AFFINITO, F. MAIERNA, A. DELLA CORTE, P. DECOOL, A. TORRE, H. CLOEZ, A. VOSTNER, A. DEVRED, D.
BASSETTE, T. BOUTBOUL: Preparation of PF1/6 and PF2 conductor performance qualification sample, IEEE Trans. Appl. Supercond.
S. TURTU’, L. MUZZI, C. FIAMOZZI ZIGNANI, V. CORATO, A. DELLA CORTE, A. DI ZENOBIO, L. RECCIA: Role of the cross–section geometry
in rectangular Nb 3 Sn CICC performance, IEEE Trans. Appl. Supercond.
A. VANNOZZI, V. GALLUZZI, A. MANCINI, A. RUFOLONI, A. AUGIERI, A.A. ANGRISANI, L. CIONTEA, Y. THALMAIER, T. PETRISOR, G. CELENTANO: Study
of MgO–based buffer layer architecture for the development of Ni–Cu–based RABiTS YBCO coated conductor, IEEE Trans. Appl. Supercond.
A. COLETTI, O. BAULAIGUE, P. CARA, R. COLETTI, A. FERRO, E. GAIO, M. MATSUKAWA, L. NOVELLO, M. SANTINELLI, K. SHIMADA, F.
STARACE, T. TERAKADO, K. YAMAUCHI: JT–60SA power supply system, Fusion Eng. Des.
P. ROSSI, M. CAPOBIANCHI, F. CRESCENZI, A. MASSIMI, G. MUGNAINI, C. NARDI, A. PIZZUTO, L. BETTINALI, J. KNASTER, H. RAJAINMAKI,
D. EVANS: Ultimate tensile strength testing campaign on ITER pre–compression ring mock–ups, Fusion Eng. Des.
F. CRESCENZI, F. MARINI, C. NARDI, A. PIZZUTO, P. ROSSI, L. VERDINI, H. RAJAINMAKI, J. KNASTER, L. BETTINALI: Mechanical
characterization of glass fibre–epoxy composite material for ITER pre–compression rings, Fusion Eng. Des.
S. CECCUZZI, S. MESCHINO, F. MIRIZZI, J.F. ARTAUD, Y.S. BAE, J. BELO, G. BERGER-BY, J.M. BERNARD, PH. CARA, A. CARDINALI, C. CASTALDO, R.
CESARIO J. DECKER, L. DELPECH, A. EKEDAHL, J. GARCIA, P. GARIBALDI, M. GONICHE, D. GUILHEM, J. HILLAIRET, G.T. HOANG, Q.Y. HUANG, F.
IMBEAUX, H. JIA, S.H. KIM, Y. LAUSENAZ, R. MAGGIORA, R. MAGNE, L. MARFISI, D. MILANESIO, W. NAMKUNG, L. PAJEWSKI, L. PANACCIONE, Y.
PEYSSON, A. SAILLE, G. SCHETTINI, M. SCHNEIDER, P.K. SHARMA, A.A. TUCCILLO, O. TUDISCO, G. VECCHI, R. VILLARI, K. VULLIEZ, Y. WU, Q. ZENG:
Mode filters for oversized transmission lines, Fusion Eng. Des.

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R. VILLARI, L. PETRIZZI, G. BROLATTI, E. DALY, M. LOUGHLIN, A. MARTIN, F. MORO, E. POLUNOVSKY: Three–dimensional neutronic
analysis of the ITER in–vessel coils, Fusion Eng. Des.
S. TOSTI, C. RIZZELLO, F. BORGOGNONI, N. GHIRELLI, A. SANTUCCI: Design of Pd-based membrane reactor for gas detritiation, Fusion Eng. Des.
S. CECCUZZI, S. MESCHINO, F. MIRIZZI, G. SCHETTINI: Mode filters for oversized corrugated rectangular vaweguide, Proc. of the XVIII
Riunione Nazionale di Elettromagnetisco (RiNEm 2010) e 1^ Conferenza Nazionale della Commissione URSI B
F. MORO, B. ESPOSITO, D. MAROCCO, R. VILLARI, L. PETRIZZI, E. ANDERSSON SUNDEN, S. CONROY, G. ERICSSON, M. GATU JOHNSON, M. DAPENA:
Neutronic calculations in support of the design of the ITER high resolution neutron spectrometer, Fusion Eng. Des.
U. FISCHER, D. GROBE, F. MORO, P. PERESLAVTSEV, R. VILLARI, L. PETRIZZI, V. WEBER: Integral approach for neutronics analyses of the
European test blanket modules in ITER, Fusion Eng. Des.
M. DAPENA, L. PETRIZZI, F. MORO: Preliminary neutronic analyses of ITER high resolution Neutron spectrometer collimator, Fusion Eng. Des.
G. CALABRO’, V. COCILOVO, F. CRISANTI, A. CUCCHIARO, R. MAZZUCA, A. PIZZUTO, G. RAMOGIDA, C. RITA, Y. SADEGHI: Active toroidal
field ripple reduction system in FAST, Fusion Eng. Des.
A. CUCCHIARO, G. BROLATTI, G. CALABRO’, V. COCILOVO, P. FROSI, F. CRESCENZI, F. CRISANTI, G. MADDALUNO, V. PERICOLI-RIDOLFINI, A. PIZZUTO,
C. RITA, G. RAMOGIDA, S. ROCCELLA, P. ROSSI: Engineering evolution of the FAST machine, Fusion Eng. Des.
F. CRISANTI, A. CUCCHIARO, R. ALBANESE, G. ARTASERSE, M. BARUZZO, T. BOLZONELLA, G. BROLATTI, G. CALABRO’, F. CRESCENZI, R.
COLETTI, P. COSTA, A. DELLA CORTE, A. DI ZENOBIO, P. FROSI, L. LAURO TARONI, G. MADDALUNO, D. MARCUZZI, F. MAVIGLIA, L. MUZZI, V.
PERICOLI-RIDOLFINI, A. PIZZUTO, G. POLLI, G. RAMOGIDA, L. RECCIA, V. RIGATO, C. RITA, S. ROCCELLA, M. SANTINELLI, P. SONATO, S. TURTU’,
M. VALISA, B. VIOLA: FAST: a European ITER satellite experiment in the view of DEMO, Fusion Eng. Des.
G. POLLI, A. DELLA CORTE, A. DI ZENOBIO, L. MUZZI, L. RECCIA, S. TURTU’, G. BROLATTI, F. CRISANTI, A. CUCCHIARO, A. PIZZUTO, R. VILLARI: A
thermo–hydraulic analysis of the superconducting proposal for the TF magnet system of FAST, Fusion Eng. Des.
V. PERICOLI–RIDOLFINI, R. ZAGORSKI, B. VIOLA, G. CALABRO’, G. CORRIGAN, F. CRISANTI, G. MADDALUNO, L. LAURO TARONI:
Simulations of the SOL plasma for FAST, a proposed ITER satellite tokamak, Fusion Eng. Des.
F. MAVIGLIA, G, ARTASERSE, R. ALBANESE, G. CALABRO’, F. CRISANTI, A. PIRONTI, A. PIZZUTO, G. RAMOGIDA: Poloidal field circuits
sensitivity studies and shape control in FAST, Fusion Eng. Des.
M. BARUZZO, T. BOLZONELLA, G. CALABRO’, F. CRISANTI, A. CUCCHIARO, D. MARCUZZI, W. RIGATO, M. SCHNEIDER, P. SONATO, M.
VALISA, P. ZACCARIA, J.F. ARTAUD, M. BASIUK, A. CARDINALI, F. IMBEAUX, L. LAURO TARONI, M. MARINUCCI, P. MANTICA, F. ZONCA:
Requirements specification for the neutral beam injector on FAST, Fusion Eng. Des.
G. VARANO, L. BONCAGNI, S. GALEANI, G. GRANUCCI, V. VITALE, L. ZACCARIAN: Experimental results on elongation control using
dynamic input allocation at FTU, Fusion Eng. Des.
L. BONCAGNI, Y. SADEGHI, C. CENTIOLI, S. SINIBALDI, V. VITALE, L. ZACCARIAN, G. ZAMBORLINI: Progress in the migration towards the
real time framework MARTe at the FTU tokamak, Fusion Eng. Des.
G. MAZZITELLI, M.L. APICELLA, A. ALEXEYEV AND FTU TEAM: Heat loads on FTU liquid lithium limiter, Fusion Eng. Des.
M. RIVA, B. ESPOSITO, D. MAROCCO, F. BELLI, B. SYME AND JET–EFDA CONTRIBUTORS: The new digital electronics for the JET neutron
profile monitor: performances and first experimental results, Fusion Eng. Des.
R. BONIFETTO, P.K. DOMALAPALLY, G.M. POLLI, L. SAVOLDI RICHARD, S. TURTU, R. VILLARI, R. ZANINO: Computation of JT–60SA TF coil
temperature margin, Fusion Eng. Des.
H. FERNANDES, F. GOMORY, A. DELLA CORTE, G. CELENTANO, J. SOUC, C. SILVA, I. CARVALHO, R. GOMES, A. DI ZENOBIO, G. MESSINA:
Toroidal high temperature superconducting coils for ISTTOK, Fusion Eng. Des.
M. ANGELONE, M. PILLON, G. PRESTOPINO, M. MARINELLI, E. MILANI, C. VERONA , G. VERONA-RINATI, G. AIELLI, R. CARDARELLI, R.
SANTONICO, R.BEDOGNI, A. ESPOSITO: Thermal and fast neutron dosimety using artificial single crystal diamond detectors, Radiat. Meas.
M. ANGELONE, P. BATISTONI, F. MORO, M. PILLON, R.VILLARI, R. BEDOGNI, M. CHITI, A. GENTILE, A. ESPOSITO: Mixed n–γ fields
dosimetry at low doses by means of different solid state dosimetry, Radiat. Meas.
T. PINNA: Studies for the preparation of the preliminary safety reports for the european test blanket system, Fusion Eng. Des.
S. ROCCELLA, G. BURRASCA, E. CACCIOTTI, A. CASTILLO, A. MANCINI, A. PIZZUTO, A. TATI’, E. VISCA: Non–destructive methods for the
defect detection in the ITER high heat flux components, Fusion Eng. Des.
R. BEDOGNI, A. ESPOSITO, A. GENTILE, M. ANGELONE, M. PILLON: Comparing active and passive bonner sphere spectrometers in the
2.5 MeV quasi mono–energetic neutron field of the ENEA–Frascati neutron generator (FNG), Radiat. Meas.

publications and events (cont’d.)
progress report
2010
109
M. PILLON, M. ANGELONE, S. SANDRI: Measurements of the activation and decay heat produced in materials irradiated with D–T
neutrons. Comparison with easy-2007 code predictions, Fusion Sci. Technol.
F. BORGONONI, S. TOSTI, C. RIZZIELLO, M. VADRUCCI, N. GHIRELLI, K. LIGER: Multi physic approach for membrane reactor modelling for
wet gas detritiation, Fusion Sci. Technol.
V. COCILOVO, R. DE ANGELIS, M.T. PORFIRI: A channel facility for Iter safety relevant dust mobilization studies, Fusion Sci. Technol.
F. BORGOGNONI, S. TOSTI, M. VADRUCCI, A. SANTUCCI: Pure hydrogen production in a Pd–Ag multi–membranes reactor by methane
reforming, Int. J. Hydrogen Energy
A. VANNOZZI, G. CELENTANO: Development and characterization of nickel–copper–based substrates fo 2nd generation high–critical
temperature superconductor tapes, In “Copper Alloys: Preparation, Properties and Application”
P. COSTA, C. NERI, M. SPAGNOLO, F. MICCHETTI: A software platform for the optimization of IVVS viewing performances, Proc. of the
2010 EnginSoft International Conference - CAE Technologies for Industry and ANSYS Italian Conference
L. DI PACE, D. CARLONI, L. PERNA, S. PACI: Application of pactiter V3.3 code to the ACPS assessment of ITER neutral beam injectors
primary heat transfer system, Fusion Sci. Technol.
M. ANGELONE, P. BATISTONI, F. MORO, M. PILLON, R. VILLARI, M. LOUGHLIN: A Neutronics shielding mock–up experiment for reduction
of uncertainty on the prediction of the ITER–TFC nuclear heating, Fusion Sci. Technol.
E. VISCA, A. PIZZUTO, B. RICCARDI, S. ROCCELLA, G.P. SANGUINETTI: A reliable technology to manufacture the ITER inner vertical target,
Fusion Sci. Technol.
A.A. TUCCILLO ON BEHALF OF FTU TEAM AND A. ALEXEYEV, L. AMICUCCI, A. BIANCALANI, A. BIERWAGE, G. BREYANNIS, I.
CHAVDAROVSKI, L. CHEN, F. DE LUCA, Z.O. GUIMARAES-FILHO, A. JACCHIA, E. LAZZARO, F. PEGORARO, M. ROMANELLI, X. WANG:
Overview of the FTU results, Nucl. Fusion
V. PERICOLI–RIDOLFINI: The effect of the lower hybrid waves on turbulence and on transport of particles and energy in the FTU tokamak
scrape–off layer plasma, Plasma Phys. Control. Fusion
V. PERICOLI–RIDOLFINI, M.L. APICELLA, G. CALABRO’, C. CIANFARANI, E. GIOVANNOZZI, L. PANACCIONE: Lower hybrid current drive
efficiency in tokamaks and wave scatteringby density fluctuations at the plasma edge, Nucl. Fusion
P. BURATTI, M. BARUZZO, R.J. BUTTERY, C.D. CHALLIS, I.T. CHAPMAN, F. CRISANTI, L. FIGINI, M. GRYAZNEVICH, T.C. HENDER, D.F.
HOWELL, H. HAN, E. JOFFRIN, J. HOBIRK, F. IMBEAUX, O.J. KWON, X. LITAUDON,J. MAILLOUX AND JET–EFDA CONTRIBUTORS: Kink
instabilities in high–beta JET advanced scenarios, Nucl. Fusion
Contributions to conferences
O. TUDISCO, C. MAZZOTTA, A. BOTRUGNO, G. MAZZITELLI, M.L. APICELLA, G. APRUZZESE, D. FRIGIONE, L. GABELLIERI, A. ROMANO and
FTU TEAM: Peaked density profiles and MHD activity on FTU in lithium dominated discharges, 2nd International Symposium on Plasma
Surface Interaction Toki, Gifu (Japan), January 18-20, 2010
G. MAZZITELLI, M.L. APICELLA, G. APRUZZESE, R. DE ANGELIS, D. FRIGIONE, E. GIOVANNOZZI, L. GABELLIERI, G. GRANUCCI, C.
MAZZOTTA, M. MARINUCCI, A. ROMANO, O. TUDISCO, FTU TEAM AND ECRH TEAM: Review of FTU results with the liquid lithium limiter,
2nd International Symposium on Plasma Surface Interaction Toki, Gifu (Japan), January 18–20, 2010
L. GABELLIERI, D. PACELLA, D. MAZON. A. ROMANO: A new SXR tomography with energy discrimination capability, 6th Workshop on
Fusion Data Processing, Validation and Analysis Madrid (Spain), January 25-27, 2010
C. CASTALDO: Novel diagnostics for dust in space, laboratory and fusion plasmas, Plasma Diagnostics 2010 Pont à Mousson, Lorrain
(France), April 12-16, 2010
C. DI TROIA, S. BRIGUGLIO, G. CALABRÒ, A. CARDINALI, F. CRISANTI, G. FOGACCIA, M. MARINUCCI, G. VLAD, F. ZONCA: Fast ion transport
and confinement in the FAST conceptual design, U.S. Transport Task Force Workshop Annapolis, MD (USA), April 13–16, 2010
A.V. MILOVANOV, F. ZONCA: Aspects of complex behavior in magnetically confined plasma with intense energetic particlepopulation, 8th
Workshop on Complex Systems of Charged Particles and their Interaction with Electromagnetic Radiation Moscow (Russia), April 14–16, 2010
A. CARDINALI, F. SANTINI: Lower hybrid current drive at densities required for thermonuclear reactors, Int. Sherwood Fusion Theory
Conference Seattle, Washington (USA), April 19–21, 2010
F.P. ORSITTO, A. BIBOC, P. GAUDIO, M. GELFUSA, E. GIOVANNOZZI, C. MAZZOTTA, A. MURARI, AND JET EFDA CONTRIBUTORS: Mutual
interaction of Faraday rotation and Cotton Mouton phase shift in JET polarimetric measurements, 18th High Temperature Plasma
Diagnostics (HTPD-18) Wildwood, NJ (USA), May 16–20, 2010

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A. ROMANO, D. PACELLA, D. MAZON, F. MURTAS, P. MALARD, L. GABELLIERI, B. TILIA, V. PIERGOTTI, G. CORRADI: Characterization of a 2D soft–X
ray tomography camera with discrimination in energy bands, 18th High–Temperature Plasma Diagnostics (HTPD-18) Wildwood, NJ, (USA), May
16-20, 2010
E. GIOVANNOZZI, M. BEURSKENS, M. KEMPENAARS, R. PASQUALOTTO, A. RYDZY AND JET EFDA CONTRIBUTORS: Detection of dust on JET with
the high resolution Thomson scattering systems, 18th High–Temperature Plasma Diagnostics (HTPD–18) Wildwood, NJ, (USA), May 16–20, 2010
L. BONCAGNI, Y. SADEGHI, D. CARNEVALE, G. MAZZITELLI, A. NETO, D. PUCCI, F. SARTORI, S. SINIBALDI, V. VITALE, R. VITELLI, L.
ZACCARIAN, S. MONACO, G. ZAMBORLINI: First steps in the FTU migration towards a modular and distributed real–time
controlarchitecture based on MARTe, 17th Real Time Conference (RT10) Lisboa (Portugal), May 24–28, 2010
L. BONCAGNI, A. BARBALACE, Y. SADEGHI, M. POMPEI, L. ZACCARIAN, F. SARTORI: Switched ethernet in synchronized distributed control
systems using RTnet, 17th Real Time Conference (RT10) Lisboa (Portugal), May 24–28, 2010
A.V. MILOVANOV, F. ZONCA: Aspects of complex behavior in magnetically confined plasma with intense energetic particlepopulation,
Birkeland Workshop on Complex Natural Systems Tromso (Norway), May 27–31, 2010
E. VISCA, A. PIZZUTO, S. ROCCELLA: Tecnologie per la costruzione di componenti ad alto flusso termico per reattori a fusione, XXVIII
Congresso UIT sulla Trasmissione del Calore Brescia (Italy), June 21–23, 2010
V. PERICOLI RIDOLFINI, G. CALABRÒ, E. GIOVANNOZZI, L. PANACCIONE, A.A. TUCCILLO: LHCD efficiency and scattering by density
fluctuations at the plasma edge, 37th EPS Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
G. CALABRÒ, D.C. MCDONALD, M. BEURSKENS, C.F. MAGGI, I. DAY, E. DE LA LUNA, (INVITED PAPER) T. EICH, N. FEDORCZAK, O. FORD, E.
GIOVANNOZZI, C. GIROUD, P. GOHIL, M. LENNHOLM, P.J. LOMAS, J. LONNROTH, G.P. MADDISON, I. NUNES, V. PERICOLI–RIDOLFINI, F. RYTER,
G. SAIBENE, R. SARTORI, W. STUDHOLME, E. SURREY, I. VOITSEKHOVITCH, K.D. ZASTROW AND JET–EFDA CONTRIBUTORS: H–Mode
threshold studies in helium–4 JET plasmas, 37th EPS Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
V. PERICOLI–RIDOLFINI: Increased particle and energy transport induced by LH waves in the tokamak scrape–off layer plasma, 37th EPS
Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
M.L. APICELLA, G. APRUZZESE, G. MAZZITELLI, A.G. ALEKSEYEV, A. BOTRUGNO, P. BURATTI, R. DE ANGELIS, H. KROEGLER, V.B.
LAZAREV, S.V. MIRNOV, R. ZAGORSKI: Radiative layer with a liquid ithium limiter on FTU, 37th EPS Conference on Plasma Physics Dublin
(Ireland), June 21–25, 2010
G. CALABRO’, E. GIOVANNOZZI, R. RAMOGIDA, F. CRISANTI, C. LABATE, M. MATTEI, P. MICOZZI, G. VLAD: Modelling of FAST equilibrium
configurations by a toroidal multipolar expansion codeusing Kepler workflows, 37th EPS Conference on Plasma Physics Dublin (Ireland),
June 21–25, 2010
A. BIERWAGE, L. CHEN, F. ZONCA: Pressure–gradient–induced Alfvén eigenmodes destabilized by ion temperature gradient, 37th EPS
Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
A. JACCHIA, S. CIRANT, F. DE LUCA, P. BURATTI, E. LAZZARO, O. TUDISCO, C. MAZZOTTA, G. CALABRO’ G. RAMOGIDA, C. CIANFARANI,
D. MAROCCO, G. GROSSETTI, G. GRANUCCI, O. D’ARCANGELO, W. BIN AND FTU AND ECRH TEAM: Density response to modulated EC
heating in FTU tokamak, 37th EPS Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
A. BOTRUGNO, P. BURATTI, F. ZONCA: Comparison between BAE observations at FTU and theoretical models, 37th EPS Conference on
Plasma Physics Dublin (Ireland), June 21–25, 2010
M. GONICHE, Y. BARANOV, P. BONOLI, G. CALABRO’, A. CARDINALI, C. CASTALDO, R. CESARIO, J. DECKER, J. GARCIA, G. GIRUZZI, E.
JOFFRIN, A. HUBBARD, K. KIROV, X. LITAUDON, J. MAILLOUX, R. PARKER (Invited Paper): Lower hybrid current drive for the steady state
scenario, 37th EPS Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
B. BAIOCCHI, G. CALABRÒ, L. LAURO–TARONI, P. MANTICA, A. CARDINALI, G. CORRIGAN, F. CRISANTI, D. FARINA, L. FIGINI, G. GIRUZZI,
T. JOHNSON, M. MARINUCCI, V. PARAIL: Predictive modelling of H–mode and steady-state scenarios in FAST, 37th EPS Conference on
Plasma Physics Dublin (Ireland), June 21–25, 2010
M. BARUZZO, M. SCHNEIDER, V. BASIUK, A. CARDINALI, M. MARINUCCI, J.F. ARTAUD, O. ASUNTA, T. BOLZONELLA, F. CRISANTI, R.
DUMONT, G. GIRUZZI, F. IMBEAUX, P. MANTICA, M. VALISA, F. ZONCA: First NBI configuration study for FAST proposal, 37th EPS
Conference on Plasma Physics Dublin (Ireland), June 21-25, 2010
A. BIANCALANI, L. CHEN, F. PEGORARO, F. ZONCA: Continuous spectrum of shear Alfvén waves inside magnetic islands, 37th EPS
Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010
L. FRASSINETTI, M.N.A. BEURSKENS, D.C. MCDONALD, J. HOBIRK, D. WARZOSO, P. BURATTI, F. CRISANTI, C. CHALLIS, E. GIOVANNOZZI, E.
JOFFRIN, P. LOMAS, J. LONNROTH, C. MAGGI, I. NUNES, S. SAARELMA, G. SAIBENE AND JET–EFDA CONTRIBUTORS: Pedestal confinement in
hybrid versus baseline plasmas in JET, 37th EPS Conference on Plasma Physics Dublin (Ireland), June 21–25, 2010

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A.A. ANGRISANI, A. AUGIERI, L. CIONTEA, G. CONTINI, I. DAVOLI, V. GALLUZZI, A. MANCINI, A. RUFOLONI, T. PETRISOR, A. VANNOZZI, G.
CELENTANO: Analysis of YBCO phase formation in thin films grown using a metal propionate coatingsolution, Applied Superconductivity
Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
A. AUGIERI, V. GALLUZZI, A.A. ANGRISANI, A. MANCINI, A. RUFOLONI, A. VANNOZZI, G. CELENTANO, E. SILVA, N. POMPEO, T. PETRISOR,
L. CIONTEA: Pinning properties of BaZrO 3-Y Ba 2 Cu 3 O 7-x thin films deposited by PLD and MOD, Applied Superconductivity Conference
(ASC 2010) Washington D.C. (USA), August 1–6, 2010
G. MESSINA, G. CELENTANO, G. GIORGI, F. MAIERNA, S. RUECA, R. VIOLA, A. DELLA CORTE: Design and manufacturing of HTS coils for axial
flux electrical machine prototype, Applied Superconductivity Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
S. GAUDIO, G. DE MARZI, G. CELENTANO, A. AUGIERI, V. GALLUZZI, A. MANCINI, A. RUFOLONI, A. VANNOZZI, A. DELLA CORTE, U.
GAMBARDELLA, A. SAGGESE, J.J. JIANG, J. WEISS, E. HELLSTROM, D. LARBALESTIER: The effect of doping on the magnetic properties in
Ba(Fe 1-x Co x ) 2 As 2 polycristalline samples, Applied Superconductivity Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
V. CORATO, L. MUZZI, U. BESI VETRELLA, C. FIAMOZZI ZIGNANI, A. DELLA CORTE: Mapping of the transversal resistivity matrix measured on
NbTi and Nb 3 Sn superconductingstrands, Applied Superconductivity Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
G. DE MARZI, L. MUZZI, V. CORATO, C. FIAMOZZI ZIGNANI, A. DELLA CORTE, G. MONDONICO, B. SEEBER, C. SENATORE, R. FLUKIGER:
Improvement of electromechanical properties of stainless steel jacketed Nb 3 Sn wires, Applied Superconductivity Conference (ASC 2010)
Washington D.C. (USA), August 1–6, 2010
A. DI ZENOBIO, A. DELLA CORTE, L. MUZZI, G.M. POLLI, L. RECCIA,S. TURTU’, F. CRISANTI, A. CUCCHIARO, A. PIZZUTO, R. VILLARI: FAST:
feasibility analysis for a completely superconducting magnet system, Applied Superconductivity Conference (ASC 2010) Washington D.C.
(USA), August 1–6, 2010
C. FIAMOZZI ZIGNANI, L. MUZZI, S. TURTU’, A. DELLA CORTE: The effect of strand bending on the voltage–current characteristics of
Nb 3 Sn cable–in–conduit conductors, Applied Superconductivity Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
A. MANCINI, G. CELENTANO, A. VANNOZZI, V. GALLUZZI, A. RUFOLONI, A. AUGIERI, S. GAUDIO, A.A. ANGRISANI, I. COLANTONI, I. DAVOLI:
The oxidation behaviour at the Ni–W and CeO 2 interface with and without Pd over layer, Applied Superconductivity Conference (ASC
2010) Washington D.C. (USA) August 1–6, 2010
L. MUZZI, G. DE MARZI, U. BESI VETRELLA, C. FIAMOZZI ZIGNANI, V. CORATO, A. RUFOLONI, A. DELLA CORTE: Test results of a NbTi wire
for the ITER PF 2–5 magnets: a validation of the 2–pinning components model, Applied Superconductivity Conference (ASC 2010)
Washington D.C. (USA), August 1–6, 2010
G.M. POLLI, L. AFFINITO, A. DELLA CORTE: Heat exchanger design for the 30 kA gas cooled current leads in the ENEA 12 T CICC facility,
Applied Superconductivity Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
L. RECCIA, S. TURTU’, G.M. POLLI, L. AFFINITO, F. MAIERNA, A. DELLA CORTE, P. DECOOL, A. TORRE, H. CLOEZ, A. VOSTNER, A. DEVRED,
D. BASSETTE, T. BOUTBOUL: Preparation of PF1/6 and PF2 conductor performance qualification sample, Applied Superconductivity
Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
S. TURTU’, L. MUZZI, C. FIAMOZZI ZIGNANI, V. CORATO, A. DELLA CORTE, A. DI ZENOBIO, L. RECCIA: Role of the cross-section geometry in
rectangular Nb 3 Sn CICC performance, Applied Superconductivity Conference (ASC 2010) Washington D.C. (USA), August 1–6, 2010
A. VANNOZZI, V. GALLUZZI, A. MANCINI, A. RUFOLONI, A. AUGIERI, A.A. ANGRISANI, L. CIONTEA, Y. THALMAIER, T. PETRISOR, G.
CELENTANO: Study of MgO–based buffer layer architecture for the development of Ni–Cu–based RABiTS YBCO coated conductor,
APPLIED SUPERCONDUCTIVITY CONFERENCE (ASC 2010) WASHINGTON D.C. (USA), AUGUST 1–6, 2010
A. CARDINALI, V. FUSCO: Analytical and numerical studies of the cold electromagnetic LH wave equation in the modeconverzion regime, Joint
Varenna–Lausanne International Workshop on Theory of Fusion Plasmas Villa Monastero, Varenna (Italy), August 30 – September 3, 2010
F. ZONCA, A. BIANCALANI, I. CHAVDAROVSKI, L. CHEN, C. DI TROIA, X. WNG: Kinetic structures of shear Alfvén and acoustic wave spectra
in burning plasmas, Joint Varenna–Lausanne International Workshop on Theory of Fusion Plasmas Villa Monastero, Varenna (Italy)
August 30 – September 3, 2010
R. CESARIO, L. AMICUCCI, C. CASTALDO, D. DE ARCANGELIS, M. FERRARI, M. MARINUCCI, F. NAPOLI, E. POLLURA, L. PANACCIONE, A.A.
TUCCILLO, AND FTU TEAM: Parametric instability and lower hybrid current drive at plasma densities required forthermonuclear reactors, Joint
Varenna–Lausanne International Workshop on Theory of Fusion Plasmas Villa Monastero, Varenna (Italy) August 30 – September 3, 2010
M. ANGELONE, M. PILLON, G. PRESTOPINO, M. MARINELLI, E. MILANI, C. VERONA , G. VERONA–RINATI, G. AIELLI, R. CARDARELLI, R.
SANTONICO, R.BEDOGNI, A. ESPOSITO: Thermal and fast neutron dosimety using artificial single crystal diamond detectors, 16th International
Conference on Solid State Dosimetry (SSD–16) Sydney (Australia), September 19–24, 2010
F. MIRIZZI, J.F. ARTAUD, Y.S. BAE, J. BELO, G. BERGER–BY, J.M. BERNARD, PH. CARA, A. CARDINALI, C. CASTALDO, S. CECCUZZI, R. CESARIO,

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J. DECKER, L. DELPECH, A. EKEDAHL, J. GARCIA, P. GARIBALDI, M. GONICHE, D. GUILHEM, J. HILLAIRET, G.T. HOANG, Q.Y. HUANG, F. IMBEAUX,
H. JIA, S.H. KIM, Y. LAUSENAZ, R. MAGGIORA, R. MAGNE, L. MARFISI, S. MESCHINO, D. MILANESIO, W. NAMKUNG, L. PAJEWSKI, L. PANACCIONE,
Y. PEYSSON, A. SAILLE, G. SCHETTINI, M. SCHNEIDER, P.K. SHARMA, A.A. TUCCILLO, O. TUDISCO, G. VECCHI, R. VILLARI, K. VULLIEZ, Y. WU, Q.
ZENG: The revised LHCD system for ITER, Workshop on RF Heating Technology of Fusion Plasmas Como (Italy), September 13–15, 2010
S. CECCUZZI, F. MIRIZZI, S. MESCHINO, L. PAJEWSKI, G. SCHETTINI, J.F. ARTAUD, Y.S. BAE, J. BELO, G. BERGER–BY, J.M. BERNARD, PH.
CARA, A. CARDINALI, C. CASTALDO, R. CESARIO, J. DECKER, L. DELPECH, A. EKEDAHL, J. GARCIA, P. GARIBALDI, M. GONICHE, D.
GUILHEM, J. HILLAIRET, G.T. HOANG, Q.Y. HUANG, F. IMBEAUX, H. JIA, S.H. KIM, Y. LAUSENAZ, R. MAGGIORA, R. MAGNE, L. MARFISI, D.
MILANESIO, W. NAMKUNG, L. PANACCIONE, Y. PEYSSON, A. SAILLE, M. SCHNEIDER, P.K. SHARMA, A.A. TUCCILLO, O. TUDISCO, G. VECCHI,
R. VILLARI, K. VULLIEZ, Y. WU, Q. ZENG: Bends and mode filters in oversized rectangular waveguide for the ITER relevant LHCD system,
Workshop on RF Heating Technology of Fusion Plasmas Como (Italy) September 13–15, 2010
S. ALMAVIVA, L. CANEVE, F. COLAO, R. FANTONI, G. MADDALUNO: LIBS characterization of ITER–like tiles superficial layers, 6th International
Conference on Laser–induced Breakdown Spectroscopy (LIBS 2010) Memphis, TN (USA), September 13–17, 2010
M. ANGELONE, P. BATISTONI, F. MORO, M. PILLON, R.VILLARI, R. BEDOGNI, M. CHITI, A. GENTILE. A. ESPOSITO: Mixed n–γ fields
dosimetry at low doses by means of different solid state dosimetry, 16th International Conference on Solid State Dosimetry (SSD–16)
Sydney (Australia), September 19–24, 2010
R. BEDOGNI, A. ESPOSITO, A. GENTILE, M. ANGELONE, M. PILLON: Comparing active and passive bonner sphere spectrometers in the
2.5 MeV quasimono–energetic neutron field of the ENEA–Frascati neutron generator (FNG), 16th International Conference on Solid State
Dosimetry (SSD–16) Sydney (Australia), September 19–24, 2010
P. BURATTI, F. ALLADIO: Magnetic reconnection, current filaments and ejection events in laboratory plasmas, XCVI Congresso Nazionale
Società Italiana di Fisica (SIF) Bologna (Italia), September 20–24, 2010
A. CARDINALI: Teoria Hamiltoniana nello studio della propagazione dell’onda ibrida inferiore in plasmiconfinati magneticamente, XCVI
Congresso Nazionale Società Italiana di Fisica (SIF) Bologna (Italia), September 20–24, 2010
R. CESARIO: Generazione di corrente a densità di plasma richieste per reattori a fusione termonucleare (Invited Paper), XCVI Congresso
Nazionale Società Italiana di Fisica (SIF) Bologna (Italia), September 20–24, 2010
C. NERI, P. COSTA, M. FERRI DE COLLIBUS, M. FLOREAN, G. MUGNAINI, M. PILLON, F. POLLASTRONE, P. ROSSI: ITER in vessel viewing system
design and assessment activities, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
E. VISCA, E. CACCIOTTI, A. KOMAROV, S. LIBERA, N. LITUNOVSKY, A. MAKHANKOV, A. MANCINI, M. MEROLA, A. PIZZUTO, B. RICCARDI,
S. ROCCELLA: Manufacturing, testing and post–test examination of ITER divertor vertical target W smallscale mock–ups, 26th
Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 - October 1, 2010
F. MIRIZZI, S. CECCUZZI, S. MESCHINO, J.F. ARTAUD, J.H. BELO, G. BERGER-BY, J.M. BERNARD, A. CARDINALI, C. CASTALDO, R.
CESARIO, J. DECKER, L. DELPECH, A. EKEDAHL, J. GARCIA, P. GARIBALDI, M. GONICHE, D. GUILHEM, G.T. HOANG, J. HUA, Q.Y. HUANG,
J. HILLAIRET, F. IMBEAUX, F. KAZARIAN, S.H. KIM, X. LITAUDON, R. MAGGIORA, R. MAGNE, L. MARFISI, D. MILANESIO, W. NAMKUNG, L.
PAJEWSKI, L. PANACCIONE, Y. PEYSSON, P.K. SHARMA, G. SCHETTINI, M. SCHNEIDER, A.A. TUCCILLO, O. TUDISCO, G. VECCHI, R.
VILLARI, K. VULLIEZ: Contribution to the design of the main transmission line for the ITER relevant LHCD system, 26th Symposium on
Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
A. COLETTI, O. BAULAIGUE, P. CARA, R. COLETTI, A. FERRO, E. GAIO, M. MATSUKAWA, L. NOVELLO, M. SANTINELLI, K. SHIMADA, F. STARACE,
T. TERAKADO, K. YAMAUCHI: JT–60SA power supply system, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27
– October 1, 2010
P. ROSSI, M. CAPOBIANCHI, F. CRESCENZI, A. MASSIMI, G. MUGNAINI, C. NARDI, A. PIZZUTO, L. BETTINALI, J. KNASTER, H. RAJAINMAKI,
D. EVANS: Ultimate tensile strength testing campaign on ITER pre–compression ring mock–ups, 26th Symposium on Fusion Technology
(SOFT) Porto (Portugal), September 27 – October 1, 2010
F. CRESCENZI, F. MARINI, C. NARDI, A. PIZZUTO, P. ROSSI, L. VERDINI, H. RAJAINMAKI, J. KNASTER, L. BETTINALI: Mechanical
characterization of glass fibre–epoxy composite material for ITER pre–compression rings, 26th Symposium on Fusion Technology (SOFT)
Porto (Portugal), September 27 – October 1, 2010
S. CECCUZZI, S. MESCHINO, F. MIRIZZI, J.F. ARTAUD, Y.S. BAE, J. BELO, G. BERGER-BY, J.M. BERNARD, PH. CARA, A. CARDINALI, C.
CASTALDO, R. CESARIO J. DECKER, L. DELPECH, A. EKEDAHL, J. GARCIA, P. GARIBALDI, M. GONICHE, D. GUILHEM, J. HILLAIRET, G.T. HOANG,
Q.Y. HUANG, F. IMBEAUX, H. JIA, S.H. KIM, Y. LAUSENAZ, R. MAGGIORA, R. MAGNE, L. MARFISI, D. MILANESIO, W. NAMKUNG, L. PAJEWSKI,
L. PANACCIONE, Y. PEYSSON, A. SAILLE, G. SCHETTINI, M. SCHNEIDER, P.K. SHARMA, A.A. TUCCILLO, O. TUDISCO, G. VECCHI, R. VILLARI, K.
VULLIEZ, Y. WU, Q. ZENG: Mode filters for oversized transmission lines, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal),
September 27 – October 1, 2010
R. VILLARI, L. PETRIZZI, G. BROLATTI, E. DALY, M. LOUGHLIN, A. MARTIN, F. MORO, E. POLUNOVSKY: Three–dimensional neutronic

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analysis of the ITER in–vessel coils, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
S. TOSTI, C. RIZZELLO, F. BORGOGNONI, N. GHIRELLI, A. SANTUCCI: Design of Pd–based membrane reactor for gas detritiation, 26th
Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
F. MORO, B. ESPOSITO, D. MAROCCO, R. VILLARI, L. PETRIZZI, E. ANDERSSON SUNDEN, S. CONROY, G. ERICSSON, M. GATU JOHNSON, M.
DAPENA: Neutronic calculations in support of the design of the ITER high resolution neutron spectrometer, 26th Symposium on Fusion
Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
U. FISCHER, D. GROBE, F. MORO, P. PERESLAVTSEV, R. VILLARI, L. PETRIZZI, V. WEBER: Integral approach for neutronics analyses of the
European test blanket modules in ITER, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
M. DAPENA, L. PETRIZZI, F. MORO: Preliminary neutronic analyses of ITER high resolution neutron spectrometer collimator, 26th
Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
G. CALABRO’, V. COCILOVO, F. CRISANTI, A. CUCCHIARO, R. MAZZUCA, A. PIZZUTO, G. RAMOGIDA, C. RITA, Y. SADEGHI: Active toroidal field ripple
reduction system in FAST, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
A. CUCCHIARO, G. BROLATTI, G. CALABRO’, V. COCILOVO, P. FROSI, F. CRESCENZI, F. CRISANTI, G. MADDALUNO, V. PERICOLI-RIDOLFINI,
A. PIZZUTO, C. RITA, G. RAMOGIDA, S. ROCCELLA, P. ROSSI: Engineering evolution of the FAST machine, 26th Symposium on Fusion
Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
F. CRISANTI, A. CUCCHIARO, R. ALBANESE, G. ARTASERSE, M. BARUZZO, T. BOLZONELLA, G. BROLATTI, G. CALABRO’, F. CRESCENZI,
R. COLETTI, P. COSTA, A. DELLA CORTE, A. DI ZENOBIO, P. FROSI, L. LAURO TARONI, G. MADDALUNO, D. MARCUZZI, F. MAVIGLIA, L.
MUZZI, V. PERICOLI-RIDOLFINI, A. PIZZUTO, G. POLLI, G. RAMOGIDA, L. RECCIA, V. RIGATO, C. RITA, S. ROCCELLA, M. SANTINELLI, P.
SONATO, S. TURTU’, M. VALISA, B. VIOLA: FAST: a European ITER satellite experiment in the view of DEMO (Invited Paper), 26th
Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
G. POLLI, A. DELLA CORTE, A. DI ZENOBIO, L. MUZZI, L. RECCIA, S. TURTU’, G. BROLATTI, F. CRISANTI, A. CUCCHIARO, A. PIZZUTO, R.
VILLARI: A thermo–hydraulic analysis of the superconducting proposal for the TF magnet system of FAST, 26th Symposium on Fusion
Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
V. PERICOLI-RIDOLFINI, R. ZAGORSKI, B. VIOLA, G. CALABRO’, G. CORRIGAN, F. CRISANTI, G. MADDALUNO, L. LAURO TARONI: Simulations of the SOL
plasma for FAST, a proposed ITER satellite tokamak, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
F. MAVIGLIA, G, ARTASERSE, R. ALBANESE, G. CALABRO’, F. CRISANTI, A. PIRONTI, A. PIZZUTO, G. RAMOGIDA: Poloidal field circuits sensitivity
studies and shape control in FAST, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal) September 27 – October 1, 2010
M. BARUZZO, T. BOLZONELLA, G. CALABRO’, F. CRISANTI, A. CUCCHIARO, D. MARCUZZI, W. RIGATO, M. SCHNEIDER, P. SONATO, M.
VALISA, P. ZACCARIA, J.F. ARTAUD, M. BASIUK, A. CARDINALI, F. IMBEAUX, L. LAURO TARONI, M. MARINUCCI, P. MANTICA, F. ZONCA:
Requirements specification for the neutral beam injector on FAST, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal),
September 27 – October 1, 2010
G. VARANO, L. BONCAGNI, S. GALEANI, G. GRANUCCI, V. VITALE, L. ZACCARIAN: Experimental results on elongation control using
dynamic input allocation at FTU, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
L. BONCAGNI, Y. SADEGHI, C. CENTIOLI, S. SINIBALDI, V. VITALE, L. ZACCARIAN, G. ZAMBORLINI: Progress in the migration towards the real time
framework MARTe at the FTU tokamak, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
G. MAZZITELLI, M.L. APICELLA, A. ALEXEYEV AND FTU TEAM: Heat loads on FTU liquid lithium limiter, 26th Symposium on Fusion
Technology (SOFT) Porto (Portugal),ß September 27 – October 1, 2010
M. RIVA, B. ESPOSITO, D. MAROCCO, F. BELLI, B. SYME AND JET–EFDA CONTRIBUTORS: The new digital electronics for the JET neutron
profile monitor: performances and firstexperimental results, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September
27 – October 1, 2010
R. BONIFETTO, P.K. DOMALAPALLY, G.M. POLLI, L. SAVOLDI RICHARD, S. TURTU, R. VILLARI, R. ZANINO: Computation of JT–60SA TF coil
temperature margin, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
H. FERNANDES, F. GOMORY, A. DELLA CORTE, G. CELENTANO, J. SOUC, C. SILVA, I. CARVALHO, R. GOMES, A. DI ZENOBIO, G. MESSINA:
Toroidal high temperature superconducting coils for ISTTOK, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal),
September 27 – October 1, 2010
P. BATISTONI, M. ANGELONE, P. CARCONI, U. FISCHER, D. LEICHTLE, A. KLIX, I. CODELI, M. PILLON, W. POHORECKI, A. TRKOV, R. VILLARI:
Neutronics experiment on a HCLL breeder blanket mock–up, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal),
September 27 – October 1, 2010
M. ANGELONE: Neutronics shielding experiment on a ITER mock–up to reduce the uncertainties on thepredictions of the toroidal field

114
progress report
2010
coils nuclear heating, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
T. PINNA: Studies for the preparation of the preliminary safety reports for the european test blanket system, 26th Symposium on Fusion
Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
S. ROCCELLA, G. BURRASCA, E. CACCIOTTI, A. CASTILLO, A. MANCINI, A. PIZZUTO, A. TATI’, E. VISCA: Non–destructive methods for the defect
detection in the ITER high heat flux components, 26th Symposium on Fusion Technology (SOFT) Porto (Portugal), September 27 – October 1, 2010
G.L. RAVERA, R. MAGGIORA, F. MIRIZZI, A.A. TUCCILLO: Performance of different load tolerant external matching unit for the FAST–ICRH
system, 40th European Microwave Conference (EUMW2010) Paris (France), September 28–30, 2010
C. VERONA, M. MARINELLI, E. MILANI, G. PRESTOPINO, G. VERONA–RINATI, M. ANGELONE, M. PILLON: Effect of secondary electrons
emission on the spectral responsivity of single crystal diamondbased Extreme–UV detector, 21st European Conference on Diamond,
Diamond–like materials, Carbon Nanotubes, and Nitrides (DIAMOND 2010) Budapest (Hungary), September 5–9, 2010
S. CECCUZZI, S. MESCHINO, F. MIRIZZI, G. SCHETTINI: Mode filters for oversized corrugated rectangular vaweguide, XVIII Riunione Nazionale
di Elettromagnetismo (RiNEm 2010) e 1^ Conferenza Nazionale della Commissione URSI B, Benevento (Italy), September 6–10, 2010
O. TUDISCO: Electron temperature characteristic length and profile resiliency in FTU, 3rd EFDA Transport Topical Group Meeting and
15th EU–US Transport Task Force Workshop Cordoba (Spain), September 7–10, 2010
C. MAZZOTTA, M. ROMANELLI, M.L. APICELLA, M. MARINUCCI, G. MAZZITELLI, O. TUDISCO AND FTU TEAM: Liquid lithium limiter
discharges at FTU: density profiles and microstability analysis, 3rd EFDA Transport Topical Group Meeting and 15th EU–US Transport
Task Force Workshop Cordoba (Spain), September 7–10, 2010
A. CARDINALI, G. CALABRO’, F. CRISANTI, C. DI TROIA, L. LAURO–TARONI, M. MARINUCCI, B. BAIOCCHI, M. BARUZZO, A. BIERWAGE, G.
BREYIANNIS, S. BRIGUGLIO, G. FOGACCIA, P. MANTICA, G. VLAD, X. WANG, F. ZONCA, V. BASIUK, R. BILATO, M. BRAMBILLA, G. CORRIGAN, F.
IMBREAUX, T. JOHNSON, V. PARAIL, S. PODDA, M. SCHNEIDER: H–mode scenarios in FAST with combined NNBI and ICRH, 3rd EFDA Transport
Topical Group Meeting and 15th EU–US Transport Task Force Workshop Cordoba (Spain), September 7–10, 2010
V. COCILOVO, R. DE ANGELIS, M.T. PORFIRI: A channel facility for Iter safety relevant dust mobilization studies, 19th Topical Meeting on
Technology of Fusion Energy (TOFE) Las Vegas (USA), November 7–11, 2010
F. CRISANTI, R. ALBANESE, F. ARTAUD, B. BAIOCCHI, M. BARUZZO, V. BASIUK, A. BIERWAGE, R. BILATO, T. BOLZONELLA, M. BRAMBILLA,
S. BRIGUGLIO, G. CALABRO’, A. CARDINALI, G. CORRIGAN, A. CUCCHIARO, C. DI TROIA, D. FARINA, L. FIGINI, G. FOGACCIA, G. GIRUZZI,
G. GRANUCCI, F. IMBEAUX, T. JOHNSON, L. LAURO TARONI, R. MAGGIORA, P. MANTICA, D. MILANESIO, V. PARAIL, V. PERICOLI-RIDOLFINI,
A. PIZZUTO, S. PODDA, G. RAMOGIDA, M. SANTINELLI, M. SCHNEIDER, A. TUCCILLO, M. VALISA, R. VILLARI, B. VIOLA, G. VLAD, X. WANG,
F. ZONCA: Scenario development for FAST in the view of ITER and DEMO, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea),
October 11–16, 2010
A. CARDINALI, M. BARUZZO, C. DI TROIA, M. MARINUCCI, A. BIERWAGE, G. BREYIANNIS, S. BRIGUGLIO, G. FOGACCIA, G. VLAD, X. WANG,
F. ZONCA, V. BASIUK, R. BILATO, M. BRAMBILLA, F. IMBEAUX, S. PODDA, M. SCHNEIDER: Energetic particle physics in FAST H–mode scenario
with combined NNBI and ICRH, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
P. MANTICA, B. BAIOCCHI, G. CALABRO’, L. LAURO TARONI,, O. ASUNTA, M. BARUZZO, A. CARDINALI, G. CORRIGAN, F. CRISANTI, D.
FARINA, L. FIGINI, G. GIRUZZI, F. IMBEAUX, T. JOHNSON, M. MARINUCCI, V. PARAIL, A. SALMI, M. SCHNEIDER, M. VALISA: Physics based
modelling of H–mode and advanced tokamak scenarios for FAST: analysis ofthe role of rotation in predicting core transport in future
machines, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
M. TARDOCCHI, A. BRUSCHI, D. MAROCCO, N. NOCENTE, G. CALABRO’, A. CARDINALI, F. CRISANTI, B. ESPOSITO, L. FIGINI, G. GORINI,
G. GROSSETTI, G. GROSSO, M. LONTANO, S. NOVAK, F. ORSITTO, U. TARTARI, O TUDISCO: Production and diagnosis of energetic particle
in FAST, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
G. MAZZITELLI, M.L. APICELLA, D. FRIGIONE, G. MADDALUNO, M. MARINUCCI, C. MAZZOTTA, V. PERICOLI–RIDOLFINI, M. ROMANELLI, G.
SZEPESI, O. TUDISCO AND FTU TEAM: FTU results with the liquid lithium limiter, 23rd IAEA Fusion Energy Conference (FEC) Daejon
(Korea), October 11–16, 2010
B. ESPOSITO, G. GRANUCCI, M. MARASCHEK, S. NOWAK, A. GUDE, V. IGOCHINE, R. MCDERMOTT, F. POLI, J. STOBER, W. SUTTROP, W.
TREUTTERER, H. ZOHM AND ASDEX UPGRADE TEAM: Avoidance of discruption at high‚ β n in ASDEX Upgrade with off-axis ECRH, 23rd
IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
R. DE ANGELIS, F. ORSITTO, M. BARUZZO, P. BURATTI, B. ALPER, L. BARRERA, A. BOTRUGNO, M. BRIX, K. CROMBE’, L. FIGINI, A.
FONSECA, C. GIROUD, N. HAWKES, D. HOWELL, E. DE LA LUNA, V. PERICOLI-RIDOLFINI, E. RACHLEW, O. TUDISCO AND JET–EFDA
CONTRIBUTORS: Determination of q profiles in JET by consistency of motional stark effect and MHD modelocalization, 23rd IAEA Fusion
Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
R. DE ANGELIS, F. ORSITTO, M. BARUZZO, P. BURATTI, B. ALPER, L. BARRERA, A. BOTRUGNO, M. BRIX, K. CROMBE’, L. FIGINI, A.

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115
FONSECA, C. GIROUD, N. HAWKES, D. HOWELL, E. DE LA LUNA, V. PERICOLI–RIDOLFINI, E. RACHLEW, O. TUDISCO AND JET–EFDA
CONTRIBUTORS: Particle deposition, transport and fuelling in pellet injection experiments at JET, 23rd IAEA Fusion Energy Conference
(FEC) Daejon (Korea), October 11–16, 2010
P. BURATTI, M. BARUZZO, R.J. BUTTERY, C.D. CHALLIS, I.T. CHAPMAN, F. CRISANTI, L. FIGINI, M. GRYAZNEVICH, T.C. HENDER, D.F.
HOWELL, H. HAN, E. JOFFRIN, J. HOBIRK, F. IMBEAUX, O.J. KWON, X. LITAUDON,J. MAILLOUX AND JET-EFDA CONTRIBUTORS: Kink
instabilities in high–beta JET advanced scenarios, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
R. CESARIO, L. AMICUCCI, M.L. APICELLA, G. CALABRO’, A. CARDINALI, C. CASTALDO, C. CIANFARANI, D. FRIGIONE, M. MAZZITELLI, C.
MAZZOTTA, L. PANACCIONE, V. PERICOLI-RIDOLFINI, A.A. TUCCILLO, O. TUDISCO AND FTU TEAM: Lower hybrid current drive at densities
required for thermonuclear reactors, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
A.A. TUCCILLO ON BEHALF OF FTU TEAM AND A. ALEXEYEV, L. AMICUCCI, A. BIANCALANI, A. BIERWAGE, G. BREYANNIS, I.
CHAVDAROVSKI, L. CHEN, F. DE LUCA, Z.O. GUIMARAES-FILHO, A. JACCHIA, E. LAZZARO, F. PEGORRO, M. ROMANELLI, X. WANG:
Overview of the FTU results, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
G. GRANUCCI, G. RAMPONI, G. CALABRO’, F. CRISANTI, G. RAMOGIDA, W. BIN, A. BOTRUGNO, P. BURATTI, O. D’ARCANGELO, D.
FRIGIONE, G. PUCELLA, A. ROMANO, O. TUDISCO, AND FTU TEAM: Plasma start–up results with EC assisted breakdown on FTU, 23rd
IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
D.C. MCDONALD, G. CALABRO’, M. BEURSKENS, I. DAY, E, DE LA LUNA, T. EICH, N. FEDORCZAC, O. FORD, W. FUNDAMENSKI, C. GIROUD,
P. GOHIL, M. LENNHOLM, J. LONNROTH, P.J. LOMAS, G.P. MADDISON, C.F. MAGGI, I. NUNES, G. SAIBENE, R. SARTORI, W. STUDHOLME,
E. SURREY, I. VOITSEKOVITCH, K.D. ZATROW AND JET–EFDA CONTRIBUTORS: JET helium–4 ELMy H–mode studies, 23rd IAEA Fusion
Energy Conference (FEC) Daejon (Korea) October 11–16, 2010
E. JOFFRIN, C.D. CHALLIS, J. CITRIN, J. GARCIA, J. HOBIRK, I. JENKINS, J. LONNROTH, D.C. MCDONALD, P. MAGET, P. MANTICA, M.
BEURSKENS, M. BRIX, P. BURATTI, F. CRISANTI, L. FRASSINETTI, C. GIROUD, F. IMBEAUX, M. PIOVESAN, F. RIMINI, G. SERGIENKO, A.C.C.
SIPS, T. TALA, I. VOITSEKOVITCH AND JET–EFDA CONTRIBUTORS: High confinement hybrid scenario in JET and its significance for ITER,
23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea) October 11–16, 2010
J. MAILLOUX, X. LITAUDON, P.C. DE VRIES, J. GARCIA, I. JENKINS, B. ALPER, YU.BARANOV, M. BARUZZO, M. BEURSKENS, M. BRIX, P.
BURATTI, G. CALABRÒ, R. CESARIO, C.D. CHALLIS, K. CROMBE, O. FORD, D. FRIGIONE, C. GIROUD, M. GONICHE, N. HAWKES, D.
HOWELL, P. JACQUET, E. JOFFRIN, V. KIPTILY, K.K. KIROV, P. MAGET, D.C. MCDONALD, V. PERICOLI-RIDOLFINI, V. PLYUSIN, F. RIMINI, M.
SCHNEIDER, S. SHARAPOV, C. SOZZI, I. VOITSEKOVITCH, L. ZABEO AND JET–EFDA CONTRIBUTORS: Towards a steasy–state scenario
with ITER dimensionless parameters in JET, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
P. MAGET, H. LUTJENS, B. ALPER, M. BARUZZO, M. BRIX, P. BURATTI, R.J. BUTTERY, C. CHALLIS, R. COELHO, E. DE LA LUNA, C. GIROUD,
N. HAWKES, G.T.A. HUYSMANS, I. JENKINS, X. LITAUDON, J. MAILLOUX, N. MELLET, D. MESHCHERIAKOV, M. OTTAVIANI, AND JET–EFDA
CONTRIBUTORS: Non linear MHD modelling of NTMs in JET advanced scenarios, 23rd IAEA Fusion Energy Conference (FEC) Daejon
(Korea), October 11–16, 2010
P. MANTICA, C. ANGIONI, B. BAIOCCHI, C. CHALLIS, J. CITRIN, G. COLYER, A.C.A. FIGUEIREDO, L. FRASSINETTI, E. JOFFRIN, T. JOHNSON,
E. LERCHE, A.G. PEETERS, A. SALMI, D. STRINTZI, T. TALA, M. TSALAS, D. VAN EESTER, P.C. DE VRIES, J. WEILAND, M. BARUZZO, M.N.A.
BEURSKENS, J.P.S. BIZARRO, P. BURATTI, F. CRISANTI, X. GARBET, C. GIROUD, N. HAWKES, J. HOBIRK, F. IMBEAUX, J. MAILLOUX, V.
NAULIN, C. SOZZI, G. STAEBLER, T.W. VERSLOOT AND JET–EFDA CONTRIBUTORS: A key to improved ion core confinement in the JET
tokamak: ion stiffness mitigation due tocombined plasma rotation and low magnetic shear, 23rd IAEA Fusion Energy Conference (FEC)
Daejon (Korea), October 11–16, 2010
P.A. POLITZER, C.D. CHALLIS, E. JOFFRIN, T.C. LUCE, M. BEURSKENS, P. BURATTI, F. CRISANTI, J.C. DEBOO, J.R. FERRON, C. GIROUD,
J. HOBIRK, C.T. HOLCOMB, A.W. HYATT, F. IMBEAUX, R.J. JAYAKUMAR, I. JENKINS, J.E. KINSEY, R.J. LA HAYE, D.C. MCDONALD, C.C.
PETTY, F. TURCO, M.R. WADE AND JET–EFDA CONTRIBUTORS: Understanding confinement in advaced inductive scenario plasmas –
dependence on gyroradiusand rotation, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
M.P. GRAYAZNEVICH, Y.Q. LIU, T.C. HENDER, D.F. HOWELL, I.T. CHAPMAN, C.D. CHALLIS, S.P. PINCHES, E. JOFFRIN, H.R. KOSLOWSKI, P.
BURATTI, E. SOLANO AND JET–EFDA CONTRIBUTORS: Determination of plasma stability using resonant field amplification in JET, 23rd
IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
X. WANG, S. BRIGUGLIO, G. FOGACCIA, G. VLAD, C. DI TROIA, F. ZONCA, L. CHEN, A. BIERWAGE, H. ZHANG: Kinetic thermal ions effects
on Alfvénic fluctuations in tokamak plasmas, 23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
L. CHEN, W. DENG, Z. LIN, D. SPONG, G.Y. SUN, X. WANG, X.Q.XU, H.S. ZHANG, W.L. ZHANG, A. BIERWAGE, S. BRIGUGLIO, I. HOLOD, G.
VLAD, Y. XIAO, F. ZONCA: Verification of gyrokinetic particle simulation of Alfvén eigenmodes excited by external antenna and by fast ions,
23rd IAEA Fusion Energy Conference (FEC) Daejon (Korea), October 11–16, 2010
ZHIYONG QUI, F. ZONCA, L. CHEN: Kinetic theories of geodesic acoustic mode in toroidal plasmas, 23rd IAEA Fusion Energy Conference
(FEC) Daejon (Korea), October 11–16, 2010

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A. BIERWAGE, C. DI TROIA AND THE FRASCATI THEORY GROUP: Method for loading marker particles for arbitrary distribution functions
and application forsimulation of high–energy ion dynamics in tokamak plasma, Joint International Conference on Superconducting in
Nuclear Applications (SNA) Tokyo (Japan), October 17–21, 2010
P. COSTA, C. NERI, M. SPAGNOLO, F. MICCHETTI: A software platform for the optimization of IVVS viewing performances, 2010 EnginSoft
International Conference – CAE Technologies for Industry and ANSYS Italian Conference Brescia, Fiera Montechiari (Italy,) October 21–22, 2010
F. BORGONONI, S. TOSTI, C. RIZZIELLO, M. VADRUCCI, N. GHIRELLI, K. LIGER: Multi physic approach for membrane reactor modelling for
wet gas detritiation, 9th International Conference on Tritium Science and Technology (TRITIUM 2010) Nara (Japan), October 24–29, 2010
A. SANTUCCI, F. BORGONONI, C. RIZZIELLO, S. TOSTI: A comparison study of highly tritiated water decomposition processes, 9th
International Conference on Tritium Science and Technology (TRITIUM 2010) Nara (Japan), October 24–29, 2010
D. PACELLA, S. DABAGOV, F. MURTAS, L. GABELLIERI, A. ROMANO, D. HAMPAI, D. MAZON: Polycapillary optics for soft x-ray diagnostics
in magnetic fusion plasmas, 4th International Conference on Charged and Neutral Particles Channeling Phenomena (Channeling 2010)
Ferrara (Italy), October 3–8, 2010
M. ANGELONE, P. BATISTONI, F. MORO, M. PILLON, R. VILLARI, M. LOUGHLIN: A Neutronics shielding mock–up experiment for reduction
of uncertainty on the prediction ofthe ITER–TFC nuclear heating, IAEA–ITER Technical Meeting on Analysis of ITER Materials and
Technologies (MIIFED 2010) Montecarlo (Principato di Monaco), November 23–25, 2010
E. VISCA, A. PIZZUTO, B. RICCARDI, S. ROCCELLA, G.P. SANGUINETTI: A reliable technology to manufacture the ITER inner vertical target,
IAEA–ITER Technical Meeting on Analysis of ITER Materials and Technologies (MIIFED 2010) Montecarlo (Principato di Monaco),
November 23–25, 2010
F. ZONCA, LIU CHEN: Recent progress and future prospects in energetic particle physics in toroidal plasmas, 20th International Toki
Conference (ITC-20) on The Next Twenty Years in Plasma and Fusion Science Toki (Japan) ,November 7–10, 2010
M. PILLON, M. ANGELONE, S. SANDRI: Measurements of the activation and decay heat produced in materials irradiated with D-Tneutrons.
Comparison with easy–2007 code predictions, 19th Topical Meeting on Technology of Fusion Energy (TOFE) Las Vegas (USA), November
7–11, 2010
L. DI PACE, D. CARLONI, L. PERNA, S. PACI: Application of pactiter V3.3 code to the ACPS assessment of ITER neutral beam injectors primary
heat transfer system, 19th Topical Meeting on Technology of Fusion Energy (TOFE) Las Vegas (USA), November 7–11, 2010
R. VITELLI, L. BONCAGNI, F. MECOCCI, S. PODDA, V. VITALE, L. ZACCARIAN: An anti–windup–based solution for the low current
nonlinearity compensation on the FTU horizontal position controller, 49th IEEE Conference on Decision and Control Atlanta, GA (USA),
December 15–17, 2010
G. VARANO, L. BONCAGNI, S. GALEANI, G. GRANUCCI, V. VITALE, L. ZACCARIAN: Plasma position and elongation regulation at FTU using
dynamic allocation, 49th IEEE Conference on Decision and Control Atlanta, GA (USA), December, 15–17, 2010
E. BARBATO: The role of collisional absorption in LHCD, 6th IAEA Technical Meeting on Steady State Operation of Magnetic Fusion
Devices Vienna (Austria) December 6–10, 2010
8.2 Workshops and Seminars
Workshops
08–09/03/2010 Task EFDA TA TGS-01–04
16/03/2010 ENEA–CEA Meeting
12–14/04/2010 JT60–SA – 8th Technical Committe Meeting (TCM–8)
13–14/04/2010 1st Course on Membranes for Fusion Fuel Cycle
15–16/04/2010 Combined Network Meeting EURATOM Research Training – FUEL CYCLE EFDA Goal Oriented Training
– TRI–TOFFY
07/05/2010 Visita rappresentanti Ambasciata di Francia
29/06/2010 HAZOP Meeting
08/07/2010 50° Anniversario dell’Associazione EURATOM–ENEA sulla Fusione
1/10/2010 CEA “CFS2010” ENEA Meeting
14/12/2010 PP11 Steering Committee N°2

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117
Seminars
26/01/2010 E. Lazzaro —CNR–IFP, Milano (Italy) (in video-conferenza)
Uno sguardo sulla fisica dei dusty plasmas
27/01/2010 K. Czerski — Institute of Physics, University of Szczecin (Poland); Institut für Optik und Atomare Physik,
Technische Universität, Berlin (Germany)
Enhanced electron screening and nuclear mechanism of cold fusion
08/02/2010 F. Alladio, P. Micozzi, A. Mancuso — ENEA Frascati (Italy)
PROTO–SPHERA: Formare e sostenere un toro di plasma con un'iniezione di elicita' quasi-continua
09/02/2010 G. Sonnino — ULB, Bruxelles (Belgio)
Nonlinear transport and distributions functions in collisional tokamak–plasmas
10/03/2010 A. Jacchia – IFP-CNR, Milano (Italy)
Density response to modulated ECH heating in FTU tokamak
29/03/2010 A.V. Chechkin — Akhiezer Institute for Theoretical Physics National Science Center “Kharkov Institute of
Physics and Technology” and School of Chemistry, Tel Aviv University (Israel)
A few "PARADOXES" of Lévy flights
14/04/2010 Y. Kamada — JAEA, Naka (Japan)
JT–60SA plasma regimes and research plan
19/04/2010 A. Perona — Politecnico di Torino (Italy)
Electron response to inertial magnetic reconnection
21/04/2010 P. Martin and the RFX–mod team – Consorzio RFX, Associazione Euratom–ENEA sulla Fusione, Padova (Italy)
Overview of the RFX fusion science program
07/07/2010 R. Khan – National Tokamak Fusion Program, Islamabad (Pakistan)
Nonlinear simulation of MHD instabilities in tokamak plasmas
24/09/2010 A. Fasoli, I. Furno – CRPP (Svizzera)
Turbulence and transport in simple magnetized toroidal plasmas
29/10/2010 G. Montani – ENEA Frascati, ICRANet, Dip. Fisica, Università di Roma (Italy)
Plasma features in astrophysical accretion disks
23–25/11/2010 M.C.H. McKubre – Energy Research Center, SRI International, Menlo Park, California (USA)
Techniques applied to studies of anomalous effects in the palladium–deuterium system: impedance
spectroscopy; resistance ratio measurement; calorimetry 1,2,3
09/12/2010 Jiangang Li – Institute of Plasma Physics, Hefei (China)
Present MCF activities in ASIPP1
8.3 Patents
RM2010U000066 S. TOSTI, F. BORGOGNONI, F. MARINI, A. SANTUCCI, A. BASILE, L. BETTINALI (*)
Dispositivo compatto a membrana in lega di Pd per la produzione di idrogeno ultrapuro
RM2010A000330
RM2010A000340
S. TOSTI, N. GHIRELLI, F. BORGOGNONI, P. TRABUC, A. SANTUCCI, K. LIGER, F. MARINI
Reattore a membrana per il trattamento di gas contenenti trizio
N. GHIRELLI, S. TOSTI, P. TRABUC, F. BORGOGNONI, K. LIGER, A. SANTUCCI, X. LEFEBVRE
Processo per la detriziazione di soft housekeeping waste
RM2010A000380 M.S. SARTO, F. SARTO, A. LAMPASI, A. TAMBURRANO, M. D’AMORE (*)
Film sottile per schermi elettromagnetici trasparenti per risparmio energetico
(*) Not in Assocation framework

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2010
chapter 9
miscellaneous (*)
(*) Not in Association framework
9.1 The Fleischman&Pons Effect Through the Materials Science
Development
The state of the palladium metal has been identified on the basis of statistical data to play fundamental roles
in producing the Fleischman–Pons excess heat effect. The deuterium loading dynamics and its equilibrium
concentration are mostly controlled by the metallurgy; a minimum threshold loading (D/Pd ∼ 0.9) is necessary
to observe the excess. The crystallographic orientation is also
correlated with the phenomenon such that mainly oriented
120
samples gave the highest reproducibility. A specific cathode surface
Pout’ (mW)
morphology, identified by means of the power spectral density
80
function, represents an additional identified condition to observe
Pin (mW) the effect.
Power
40
0
0 50000 100000 150000
Time
Figure 9.1 – Excess of power evolution
Materials specimens respecting the characteristics described above
have been used to obtain a transportable reproducibility. Designed
materials giving excess power have been produced, but the
amplitude of the signals and full reproducibility are not yet
achieved. Other features of the material, such as the nature and
content of impurities and defects, seems to be crucial in obtaining
the required palladium characteristics.
Excess of power obtained by using designed materials are shown
in figure 9.1 a and 9.2
Figure 9.2 – Microscopy of the
“designed” Pd Electrode
9.2 Pd–Based Membranes
A wide development of Pd–based membranes and a study of
processes for ultra pure hydrogen production have been
performed [9.1]: particularly, a two-step process consisting of a
traditional reformer operating at high temperature (700–800 °C)
and a multi–tube membrane module performing the water gas
shift reaction (350–400 °C) has been tested, see figure 9.3. The membrane module consists of 19 Pd–Ag tubes
of diameter 10 mm, length about 270 mm and wall thickness 50–60 μm. Water–ethanol mixtures have been
treated via steam and oxidative reforming by attaining maximum values of hydrogen yield over 90%, which
correspond to about 3.5 NL min –1 of ultra pure hydrogen produced.
A compact Pd membrane module of reduced size and weight for mobile applications has also been designed
and manufactured [9.2]. A special configuration similar to the design of flat–and–frame heat exchanger has

miscellaneous (cont’d.)
progress report
2010
119
been proposed. The Pd–Ag composite membranes have
been produced in form of thin sheets supported over
stainless steel grids and welded to stainless steel frames. The
diffusion welding procedure has further been operated in
order to tightly join the Pd–Ag membranes to the stainless
steel frames. The resulting membrane module is very
compact: as an example, a permeator of surface area of
10 m 2 could be about 1000 × 120 × 180 mm in size.
Modelling and testing of membrane reactors using Pd–Ag
thin wall tubes have concerned the investigation of ethanol
steam and oxidative reforming [9.3–9.7] and water gas shift
reaction [9.8]. The tests have highlighted the complete
hydrogen selectivity of the thin wall membranes and their
capability of promoting the reaction conversion beyond the
thermodynamic equilibrium (shift effect of the
membrane) [9.9].
Figure 9.3 – Hydrogen production from ethanol
reforming: experimental apparatus
9.3 AGILE and LOFT
In 2010 the Astrorivelatore Gamma ad Immagini LEggero
(AGILE), the satellite for gamma and x–ray astronomy built
with the contribute of ENEA, made two interesting findings
(among other results): it discovered a possible new class of
celestial gamma-ray sources [9.10], and shed new light on the
nature of the terrestrial gamma ray flashes, an intriguing
phenomenon which might have consequences on the safety of
civil aviation [9.11].
The promising performances of a new x–ray detector, designed
in collaboration with ENEA Fusion Technology Unit [9.12],
based on silicon drift detectors, have led to the proposal to
“ESA call for a medium–size mission opportunity for a launch
in 2022” of a new satellite, called Large Observatory For x–ray
Timing (LOFT) [9.13], backed by 130 scientists from 16
countries. Thanks to an innovative design and the development
Figure 9.4 – Pictorial view of LOFT, with the six
petals of LAD deployed
of large monolithic silicon drift detectors, the large area detector (LAD) on board of LOFT would achieve an
effective area of ∼12 m 2 (more than one order of magnitude larger than the current spaceborne x–ray

120
progress report
2010
Figure 9.5 – LOFT, with the detectors folded, in the ogive of the
launcher
detectors) in the 2–30 keV range (fig. 9.4). The achievable time–resolution of
its x–ray observations of compact objects would constitute a relevant
diagnostic for strong–field gravity and the equation of state of ultra dense
matter. LOFT has been presented in December 2010; to date it has been
selected (with other 13, out of 47 proposals) for further consideration by the
Advisory Structure of the ESA Scientific Programme.
Figure 9.6 – Elecronic equipment for the RF interferometer for helicon
plasma hydrazine
Standard deviation (Deg)
20
10
5
3
1
0.5
0.2
0.1
0.05
0.02
10
implementation and is based on an ENEA electronic patent.
9.4 Digital Phase Detector for the RF
Interferometer for Mini Helicon Space Thruster
Within the HPH.com project a RF interferometer has been built for the mini
helicon thruster at the CISAS laboratory (Padua). This interferometer is
capable of detecting a phase variation of 0.3° within the band 0–15 kHz, and
0.06° within the band 0-100 Hz. This instrument has been used to detect the
density of the plasma generated by the thruster (which is a small cylindrical
plasma with 20 mm of diameter) with an error of 4×10 16 m –3 in the band
0–15 kHz, and of 1×10 16 m –3 in the band 0–100 Hz.
20 40 60 80
Attenuation (dB)
Figure 9.7 – Total error (including the linearity error)
vs attenuation
Commercial analog I/Q detectors for
phase shift measurements introduce an
error not less that 1°, so a dedicated phase
detector, with appropriate noise figures
(0.02°), has been developed, tested and
successfully integrated for the RF
interferometer. It is based on a digital
elaboration system that implements a
vector voltmeter working at the frequency
of 100 MHz and measures the complex
components of the signal under test as
compared
The system sends the elaborated data to a
PC, via an ethernet port at a speed of
31thousand measurements per second. On
the PC a GUI able to start receiving is
present, which saves the data and allows
for a preview of decimated data.
Furthermore, the system shows on a 4–line
LCD the value of the module and the
phase instantaneously measured.
The system calculates the quadrature
components, has a fully digital
The advantages of this system are high linearity, high measurement frequency and high noise immunity, see
figure 9.7.

miscellaneous (cont’d.)
progress report
2010
121
References
[9.1] S. Tosti, Int. J. Hydrogen Energy 35, 12650–12659 (2010)
[9.2] S. Tosti, et al., Dispositivo compatto a membrana metallica per la produzione di fluidi gassosi, Domanda di
brevetto per modello di utilità n. RM2010U000066 del 24.04.2010
[9.3] S. Tosti, et al., Asia–Pac. J. Chem. Eng. 5, 207–212 (2010)
[9.4] A. Iulianelli, et al., J. Hydrogen Energy 35, 3159–3164 (2010)
[9.5] A. Santucci, et al., Oxidative steam reforming of ethanol over a Pt/Al 2 O 3 catalyst in a Pd–based membrane
reactor, accepted for publication on Int. J. Hydrogen Energy
[9.6] S. Tosti, et al., L. M. Madeira, Catal. Today 156, 107–117 (2010)
[9.7] S. Tosti, F. Borgognoni, A. Santucci, Int. J. Hydrogen Energy 35, 11470-11477 (2010)
[9.8] S. Tosti, et al., Int. J. Hydrogen Energy 35, 12596–12608 (2010)
[9.9] A. Basile, et al., Adv. Sci. Technol. 72, 99–104 (2010)
[9.10] A. Pellizzoni, et al., Science 327, 663 (2010)
[9.11] M. Marisaldi, et al., J. Geophys. Res. 15, A00E13, (2010) doi:10.1029/2009JA014502
[9.12] ENEA - Unità Tecnica Fusione, Progress Report (2009)
[9.13] M. Feroci, et al., LOFT – a Large Observatory For x–ray Timing, Proceedings of SPIE Vol. 7732, Paper No.
7732–66 (2010); arXiv:1008.1009v1 [astro–ph.IM]

Organization Chart
The activities of the ENEA Research Group
are performed in the Nuclear Fusion Unit.
In 2010, a total of 149 professionals and 131
non professionals worked in fusion activities
Euratom–ENEA
Association
2010annual
fusion
activities

progress report
2010
123
UNITA’ TECNICA FUSIONE
Aldo Pizzuto
Servizio Diffusione Tecnologie
Paola Batistoni
Coordinamento Attività
Progetto FAST
Flavio Crisanti
Servizio Supporto Tecnico
Gestionale
Nicola Manganiello
Coordinamento Pianificazione
Logistica e Sicurezza
Giovanni Coccoluto
Laboratorio Fisica della Fusione a
Confinamento Magnetico
Angelo Tuccillo
Laboratorio Fusione Inerziale
Aldo Pizzuto a.i.
Laboratorio Gestione Grandi Impianti Sperimentali
Giuseppe Mazzitelli
Laboratorio Ingegneria Elettrica ed Elettronica
Giuseppe Mazzitelli a.i.
Laboratorio Tecnologie Nucleari della Fusione
Aldo Pizzuto a.i.
Laboratorio Superconduttività
Antonio della Corte
December, 2010

Abbreviation and Acronyms
A
ac
ACCC
ACP
ADC
AE
AES
AGILE
APD
Asdex–U
B
BA
BAE
BgB
BP
C
alternating current
active correction and compensation coils
activated corrosion product
analog–digital converter
Alfvén eigenmode
Auger electron spectroscopic
Astrorivelatore Gamma ad Immagini
LEggero
avalanche photodiode
Axially Symmetric Divertor EXperiment
Upgrade – Garching (Germany)
Broader Approach
beta–induced Alfvén eigenmode
Bohm–gyroBohm
backplate
Euratom–ENEA
Association
2010annual
fusion
activities
CB
cassette body
CC
coated conductors
CCD charge–coupled device
CD
current drive
CEA Commissariat à l’Energie Atomique –
Cadarache (France)
CFC
carbon/carbon fibre composite
CGM critical gradient model
CICC cable–in–conduit conductor
CNR
Consiglio Nazionale delle Ricerche (Italy)
(National Research Council)
CNS compact neutron spectrometer
CPOs Consistent Physical Objects
CRP
Coordinate Research Project
CRPP Centre de Recherches en Physique des
Plasmas – Villigen (Switzerland)
CS
central solenoid
CTS
collective Thomson scattering
CVCS chemical volume control system
CW
continuous wave
D
DACs
data acquistion systems

abbreviation and acronyms (cont’d.)
progress report
2010
125
DEMO
DPRW
DPSD
demonstration/prototype reactor
dynamic polarization random walk
digital pulse shape discrimination
E
EAE
EBBTF
EC
ECCD
ECE
ECH
ECRH
EFDA
EGAM
ELITE
ELM
EM
EMOBC
EO
EPs
EPMS
ESRF
ETG
ETS
EVEDA
EXAFS
ellipticity–induced Alfvén eigenmode
European Breeding Blanket Test Facility
electron cyclotron
electron cyclotron current drive
electron cyclotron emission
electron cyclotron heating
electron cyclotron resonance heating
European Fusion Development Agreement
energetic–particle–induced GAM
EVEDA LIthium Test Experiment
edge localized mode
electromagnetic
electromagnetic model of the building complex
electro–optical
energetic particles
energetic particle MODES
European Synchrotron Radiation Facility – Grenoble (France)
electron temperature gradient
European Transport Solver
Engineering Validations and Environmental Engineering
Design Activities
extended x–ray absorption fine structure
F
FAST
FEMs
FF
F4E
FLR
FMW
FNG
FOW
FPGA
FTU
FW
FWHM
Fusion Advanced Studies Torus
finite element models
For Fusion
Fusion For Energy
finite Larmor radius
fast magneto–sonic waves
Frascati Neutron Generator – ENEA (Italy)
finite orbit width
field programmable gate array
Frascati Tokamak Upgrade – ENEA (Italy)
first wall
full width at half maximum
G
GAM
geodesic acoustic mode

126
progress report
2010
GEM
GRS
GSEP
H
HAZOP
HEFUS
HFTM
HYMAGYC
HMGC
HPC–FF
HRNS
HRP
HRTS
HTS
HTW
HV
I
IAEA
IBW
IC
ICRANet
ICRF
ICRH
IELLLO
IFMIF
IFP
IFTS
ILO
ILW
IMP
INFN
gas electron multiplier
gamma ray spectroscopy
gyrokinetic simulation of energetic
particle turbulence and transport
hazard and operability
HElio for FUSion
high flux test module
new hybrid magnetohydrodynamic
gyrokinetic code
hybrid magnetohydrodynamic
gyrokinetic code
High Performance Computing For
Fusion
high resolution neutron
spectromete
hot radial pressing
high resolution Thomson
scattering
high–temperature superconductor
highly tritiated water
high voltage
International Atomic Energy Agency
– Vienna
ion Bernstein waves
ion–cyclotron
International Center for Relativistic
Astrophysics Network (Italy)
ion–cyclotron resonance frequency
ion cyclotron resonance heating
Integrated European Lead Lithium
LOop
International Fusion Materials
Irradiation Facility
Institute of Plasma Physics – CNR
Milan (Italy)
Institute for Fusion Theory and
Simulation, Zhejiang University –
Hangzhou (China)
Industrial Liaison Officer
ITER like wall
Integrated Modelling Project
National Institute of Nuclear
Physics (Italy)
IO
IOS
IPP
ISP
ISTTOK
ITB
ITER
ITG
ITG
ITM
ITPA
IVT
IVVS
J
JAEA
JET
JT–60SA
K
ITER Organization
Integrated Operation Scenario
Max–Planck–Institut für
Plasmaphysik – Garching
inner strike point
Istituto Superior Técnico Tokamak
internal transport barrier
“the way” in latin
ion turbulent gradient
ion temperature gradient
Integrated Tokamak Modelling
International Tokamak Physical
Activity
inner vertical target
in–vessel viewing system
Japan Atomic Energy Agency
Joint European Torus – Abingdon
(UK)
Japan Tokamak 60 Super
Advanced
KGAM kinetic GAM
KIT Karlsruhe Institute of Technology –
Karlsruhe (Germany)
KTH Royal Institute of Technology –
Stockholm (Sweden)
L
LAD
LCMS
LCOs
LCQ
LFS
LH
LHCD
LLL
LNF
LOFT
LOS
large area detector
last closed magnetic surface
limiting conditions for operations
linear current quench
low field side
lower hybrid
lower hybrid current drive
liquid lithium limiter
INFN - Frascati National Laboratory
(Italy)
Large Observatory For x–ray
Timing
line of sight

abbreviation and acronyms (cont’d.)
progress report
2010
127
LOVA
LTC
LTCC
LTS
M
MARTe
MCNP
MDT
MDTNS
MFG
MFP
MFRS
MHD
MiAE
MiEAE
MiKAE
MiUR
MOD
MSE
MTLs
MUT
N
NBI
NES
NICD
NMC
NNBI
NPM
NTM
O
OSP
loss of vacuum accident
Large Tokamak Cooperation
low temperature co–fired ceramic
low–temperature superconducting
multi–threated application real–time
executor
Monte Carlo n–Particle
routine maintenance days
not–scheduled mean down time
motor flywheel generator
magnetic fusion plasmas
magnetic field reduction system
magnetohydrodynamic
magnetic island induced Alfvén
eigenmode
magnetic island eccentricity
induced Alfvén eigenmode
magnetic island kinetic induced
Alfvén eigenmode
Ministero dell’Università e della
Ricerca
metal–organic deposition
motional Stark effect
main transmission lines
mean up time
neutral beam injection
neutron emission spectroscopy
non inductive current driven
new manifold concept
negative neutral beam injection
neutron profile monitor
neoclassical tearing modes
outer strike point
PERMCAT
PF
PFC
PFU
PHSC
PHTS
PI
PTB
PW
Q
QMS
R
rf
RH
RNC
RWM
S
SAW
Sci–DAC
SD
SEM
SiC
SNR
SOC
SOL
SSFPQL
ST
SXR
T
permeator catalyst
poloidal field
plasma–facing component
plasma–facing unit
Programmable High Speed
Controller
primary heat transfer system
parametric instability
Physikalisch–Technische
Bundesanstalt – Braunschweig
(Germany)
plasma wall
Quality Management System
radiofrequency
remote handling
radial neutron camera
resistive wall modes
shear Alfvén wave
Scientific Discovery through
Advanced Computing – USA
slowing down
scanning electron microscopy
silicon carbide
signal to noise ratio
self–organized criticality
scrape–off layer
steady state Fokker Planck quasi
linear
sawteeth
soft–x–ray
P
PAM
PBC
passive–active multijunction
pre–brazed casting
TAE
TAs
TBM
TDS
TEM
toroidal Alfvén eigenmode
task agreements
test blanket module
time–domain spectroscopy
trapped electron mode

128
progress report
2010
TF
TF
TFC
TFR
TFSC
TG
TORIC
TP
TRIEX
U
UCI
UT
UTFUS
UTS
V
VDE
VF
Task Force
toroidal field
toroidal field coils
toroidal field ripple
toroidal field superconductor coil
Topical Group
toroidal ion cyclotron
twist pitch
tritium extraction
University of California – Irvine
(Usa)
ultrasonic technique
ENEA - Unità Tecnica Fusione
ultimate tensile strength
vertical displacement events
void fraction
VHFS
VNA
VPCE
VPI
VS
VS
VUV
VV
W
WW
X
XHMGC
XPS
XRD
XSC
vertical high field side
vector network analyzer
vapor phase catalytic exchange
vacuum pressure impregnation
vertical system
vertical stabilizing
visible ultraviolet
vacuum vessel
wet winding
eXtended hybrid
magnetohydrodynamic gyrokinetic
code
x–ray photoelectron spectroscopy
x–ray diffraction
extreme shape control