1. magnetic confinement - ENEA - Fusione

ITALIAN AGENCY FOR NEW TECHNOLOGIES
ENERGY AND THE ENVIRONMENT
NUCLEAR FUSION DIVISION
2001 PROGRESS REPORT

Activities carried out by ENEA in the framework of the
EURATOM-ENEA Association on Fusion
(with minor exceptions as indicated in the list of contents)
This report was prepared by the Scientific Publications Office from contributions provided by the scientific and
technical staff of ENEA’s Fusion Division. Main collaborators in the preparation of this issue: Marisa Cecchini,
Lucilla Crescentini, Lucilla Ghezzi.
Editing:
Carolyn Kent
Cover: X-ray bright spot due to
particle accumulation inside a
magnetic island
Published by:
ENEA - Edizioni Scientifiche, Centro Ricerche Frascati, C.P . 65 - 00044 Frascati, Rome, Italy
Tel: +39(06)9400 5670 Fax: +39(06)9400 5015
e-mail: crescentini@frascati.enea.it

1. MAGNETIC CONFINEMENT 09
1.1 Tokamak Physics 09
1.1.1 Introduction 09
1.1.2 Experimental results 09
1.1.3 Downshifted and upshifted experiments with ECRH in LHCD plasmas 13
1.1.4 MHD behaviour in improved confinement regimes and new phenomena
by fast MHD analysis 14
1.1.5 Pellet injection 16
1.1.6 Boronisation: plasma results 17
1.1.7 Radiative improved mode in Ohmic plasmas 19
1.1.8 Fast x-ray imaging of the NSTX plasma by a micro-pattern gas
detector with a GEM amplifier 20
1.1.9 JET 22
1.2 FTU Facilities 27
1.2.1 FTU machine 27
1.2.2 Heating systems 30
1.2.3 Diagnostics 31
1.3 Plasma Theory 41
CONTENTS
1.3.1 Introduction 41
1.3.2 IBW-induced poloidal rotation on FTU 42
1.3.3 Generation of zonal flows by drift-Alfvén turbulence 43
1.3.4 Drift and drift-Alfvén wave structures near a minimum-q surface 44
1.3.5 Energetic particle mode destabilisation by ICR-heated fast
ions in reversed shear plasmas 45
1.3.6 Nonlinear dynamics of shear Alfvén modes and energetic ion
confinement in reversed shear tokamak equilibria 48
1.4 FT3 Conceptual Study 50
1.4.1 Introduction 50
1.4.2 Main objectives of the FT3 scientific programme 50
1.5 PROTO-SPHERA 52
1.5.1 Introduction 52
1.5.2 Mechanical engineering 53
2. IGNITOR PROGRAM* 59
2.1 Introduction 59
2.2 Physics 59
2.2.1 Advanced scenarios 59
(*) Not in association framework

2.3 Engineering of the Machine 59
2.3.1 EM analysis of vacuum vessel during plasma disruptions 59
2.3.2 Engineering models 60
2.3.3 Plasma-wall interaction and molybdenum contamination 61
2.3.4 Auxiliary plasma heating system: ICRH 61
3. FUSION TECHNOLOGY 65
3.1 Technology Programme 65
3.1.1 Introduction 65
3.2 First Wall and Divertor 65
3.2.1 Influence of manufacturing heat cycles on CuCrZr properties
(ITER Task DV4/04) 65
3.2.2 Manufacturing of small-scale W monoblock mockups by hot
radial pressing (ITER EFDA R&D Tasks) 66
3.2.3 Runaway electrons on ITER PFCs (EFDA Contract /00-520) 66
3.3 Vacuum Vessel and Shield 67
3.3.1 EM analyses of in-vessel components for ITER-FEAT 67
3.3.2 ITER-FEAT breeding blanket 68
3.4 Magnets 69
3.4.1 Installation and testing of ITER CS and TF model coils
(ITER Task M20) 69
3.4.2 Development of calculation codes for CIC conductors
(EFDA Task TWO-T400-1/01) 71
3.4.3 New diagnostics for a CIC conductor (EFDA Task TWO-T400-1/01) 71
3.4.4 Development of NbTi conductors for ITER PF coils
(ITER Task M50, EFDA Task TWO-T405/1 and TW1-TMC/SCABLE) 72
3.4.5 Test in SULTAN of the ENEA Nb3Sn magnet (ITER Task M20) 74
3.4.6 Chemical deposition of oxide buffer layers for YBCO-coated
metallic tapes 74
3.4.7 Development of Nb 3 Al strands for high-field applications 74
3.4.8 Feasibility study on eddy current testing of ITER coil case welds
(ITER Task TW1-TMS/MMTFRD) 75
3.5 Neutronics 75
3.5.1 3-D nuclear analysis for ITER-FEAT design 75
3.5.2 Experimental validation of shutdown dose rates for ITER 76
3.5.3 Design of the neutron cameras for ITER 78
3.5.4 Evaluation of neutron cross sections for fusion materials (EFF project) 79
3.5.5 Neutronics benchmark experiment on SiC (EFF project) 79
3.5.6 Experimental validation of neutron cross sections for fusion
materials (EAF project) 80

3.6 Remote Handling 81
3.6.1 IVROS articulated boom 81
3.6.2 Upgrade of DRP heavy manipulator/crane/trolley 82
3.6.3 Trials using ITER FDR 98 duct equipment in real remote conditions 82
3.6.4 Installation, commissioning and trials with the CEA/Cybernetix MAESTRO
radiation-hard servo-manipulator arm on DTP cassette toroidal mover 82
3.6.5 High-discharge electrical tests of multilink attachment pin concept
at CESI* 82
3.6.6 Final DRP trials using ITER FDR 98 cassette mockup with multilink
attachments 83
3.6.7 In-vessel viewing and ranging 83
3.7 Materials 83
3.7.1 Compatibility of SiC f /SiC composites with Pb-17Li 83
3.7.2 Microstructural investigation of radiation effects in RAFM
steel by SANS 84
3.7.3 Mechanical properties of RAFM steel-base material and joints 84
3.7.4 Low-cycle fatigue of RAFM steel in water with additives 85
3.7.5 Development of a low-activation brazing technique for SiC f /SiC
composites 86
3.7.6 Measurement of residual stresses using neutron diffraction techniques 87
3.7.7 SiC/SiC ceramic composites as PFC material 87
3.7.8 Mechanical characterisation of materials with miniaturised specimens 89
3.8 Liquid Metal Technology and Hydrogen Effects on Materials 89
3.8.1 Interaction between lead-lithium alloy and water in DEMO-relevant
conditions (EU Task TTBA-5) 89
3.8.2 Qualification of tritium permeation in Pb-17Li/gas 91
3.8.3 Transport parameters and solubility of hydrogen in Pb-17Li 92
3.8.4 Hydrogen permeability and embrittlement in EUROFER97 martensitic steel 92
3.8.5 Water detritiation systems (EU Task TTBA-D02) 93
3.8.6 Measurements of H/D diffusivity and solubility through tungsten
and tungsten alloys in the range 600-800°C (ITER Task 436) 94
3.8.7 Corrosion and mechanical tests on structural materials in flowing
Pb-17Li (EU Task TTMS-003-D13) 94
3.8.8 Interaction chemistry between Li 2 TiO 3 ceramic pebble bed and
EUROFER97 in He+0.1% H 2 purge gas at 600°C 95
3.8.9 Li 2 TiO 3 pebble reprocessing; recovery of 6 Li as Li 2 CO 3 96
3.9 Thermal-Fluidodynamics 97
3.9.1 Fatigue tests on six mockups of primary first-wall panel prototype
(EFDA Contracts 00/529 and 00/533) 97
3.9.2 HE-FUS3 experimental cassette of lithium-beryllium pebble beds 98
(*) Not in association framework

3.10 International Fusion Material Irradiation Facility (IFMIF) 98
3.10.1 Design and mockup tests of lithium jet target 98
3.10.2 System safety analysis and shielding calculations 99
3.10.3 Development of fast neutron diagnostics 100
3.11 Fuel Cycle 100
3.11.1 Tritium recovery from tritiated water 101
3.12 Safety and Environment, Power Plant Studies and Socio-Economics 102
3.12.1 Occupational radiation exposure assessment for ITER-FEAT 102
3.12.2 Validation of computer codes and models (EFDA Task SEA5) 102
3.12.3 Plant safety assessment for ITER-FEAT 104
3.12.4 Waste management 107
3.12.5 Power Plant Conceptual Study 108
3.12.6 European ITER site at Cadarache 110
4. MISCELLANEOUS 113
4.1 Development of CVD Diamond Detectors for Nuclear Radiation* 113
4.2 Light Response of a Pure Liquid Xenon Scintillator* 113
4.3 Partecipation in the agile Project: Collimator and Coded Mask of the
Superagile Detector* 113
4.4 Advanced Superconducting Materials and Devices* 114
4.4.1 Ni-W based architectures: preliminary results 114
4.4.2 Influence of the substrate on the YBCO-film transport properties 115
4.4.3 Inclined substrate deposition of CeO 2 films on randomly
oriented metallic substrate 115
4.4.4 MgB 2 film fabrication 116
4.5 Optical Metrology Survey 117
4.6 New Hydrogen Energy* 118
4.7 Cryogenic Testing of Diode Stacks for CERN* 119
4.8 Cryogenics* 119
4.8.1 Liquid helium service 119
4.8.2 Cryogenic technologies 119
5. INERTIAL CONFINEMENT 121
5.1 Introduction 123
5.2 Diagnostic Upgrading 123
(*) Not in association framework

5.3 Theory 123
5.3.1 Interaction of laser beams with multi-foil plastic structures 123
5.3.2 Code COBRAN implementation 128
5.3.3 DPSSL design activity 128
PUBLICATIONS, CONFERENCES AND REPORTS 129
Publications 131
Articles in Course of Publication 136
Contributions to Conferences 137
Reports 141
Conferences and Seminars 142
ORGANISATION CHART 145
ABBREVIATIONS AND ACRONYMS 147

ENEA’s activities in the field of controlled nuclear fusion form part
of its overall mandate to conduct research on energy sources that
have a low environmental impact and a high innovative content.
There are two promising approaches to controlled nuclear fusion -
magnetic confinement and inertial confinement. In the first approach,
suitably configured high magnetic fields are used to contain the reacting
plasma and limit energy loss. The second approach uses pulsed energy
sources (lasers or particle beams, generally known as drivers) to
compress the fuel and heat part of it to the critical temperature at which
fusion reactions are triggered. The results achieved over the past two
decades have convinced most researchers that it is now possible to
proceed with experiments that demonstrate so-called thermonuclear
ignition or, at least, a high ratio between the fusion energy generated and
the energy required to produce the reacting configuration. Towards this
end, large-scale projects are currently under design or already under
way.
PREFACE
The Italian programme addresses both concepts but focuses mainly on
magnetic confinement fusion. Italy accounts for 20% of the European
Fusion Programme and is second only to Germany and slightly ahead of
France. The EU through the European Fusion Programme leads the
world in the field of magnetic confinement and, indeed, has
demonstrated that a policy of close interaction and cooperation between
centres of excellence within a framework of a common EU strategy is
extremely gainful to all concerned.
The most important experimental activities of ENEA are carried out
on the Frascati Tokamak Upgrade (FTU) at the Frascati
laboratories. FTU is a high magnetic field tokamak device
dedicated to experiments on microwave plasma-heating. ENEA is also
involved in experimental work on the Joint European Torus (JET) and
collaborates in the design of the International Thermonuclear
Experimental Reactor (ITER). The integrated nature of this global
programme has generated extensive collaborations, in which the results
and goals are shared, as well as competition between the participants to
obtain the most significant tasks. ENEA has forged several national
collaborations which involve experimental work on FTU. The main

partner in this respect, the Institute of Plasma Physics of the National
Research Council (CNR) Milan, specialises in researching electron
cyclotron resonance plasma heating. Other partners include the
Reversed Field Experiment (RFX) Consortium (which operates the
reversed field pinch device), the interuniversity CREATE Consortium
and research teams from Turin Polytechnic and the Universities of
Catania and Rome. Outside Italy, ENEA has carried out joint experiments
on FTU with CEA, the John Hopkins University, the Lawrence Livermore
National Laboratory and various Russian institutes.
Technological R&D for magnetic fusion entails collaboration with
Italian industrial partners (Ansaldo, Belleli, Edison, Europa
Metalli, OECM, etc.) as well as universities in Italy and abroad
(Turin Polytechnic, Universities of La Sapienza, Tor Vergata, Bologna,
Dresden, Stanford) and research institutes (Commisariat à l’Energie
Atomique France, Forschungswentrum Karlsruhe Germany, Centre de
Recherches en Physique des Plasmas Switzerland).
Inertial confinement fusion (ICF) is studied at the ABC laser facility at
ENEA Frascati. The ABC, designed and built entirely at Frascati, is
used for preliminary experiments on laser-matter interaction and
studies on the resulting ablative acceleration. The ICF group’s experience
and expertise in theory and modelling have also gone to make ENEA
Frascati one of the top non-military ICF research centres.
PREFACE

1. MAGNETIC CONFINEMENT
9
1.1.1 Introduction
1.1 Tokamak Physics
The main goal of the Frascati Tokamak Upgrade (FTU) scientific programme is to
investigate transport, stability and radiofrequency physics issues at ITER-like
plasma densities and magnetic field values.
In 2001, the FTU lower hybrid system came close to delivering its maximum allowable
(~2.4 MW) power. Up to 2.2 MW were injected into the plasma, with a level of 2.0 MW
routinely achieved. During the 2001 experimental campaign it was, therefore, possible
to study both current-drive and internal transport-barrier formation in high-density
scenarios. In fact, collisional ion heating was observed at densities around 10 20 m -3 .
Electron cyclotron resonance heating was used to explore heat transport and was
combined with the injection of lower hybrid waves for synergy studies. Modulated
electron cyclotron heating was also applied for the transport studies, which focussed on
the issue of profile stiffness and the specific role of collisionality. As proposed at the 2001
Frascati workshop (see below) on the FTU programme, a campaign is being planned to
compare FTU results with those of other devices, such as ASDEX-Upgrade and Tore
Supra.
Steady pellet enhanced performance modes were extensively studied. Careful timing
of the pellet sequence allowed a high degree of reproducibility to be achieved for
these high-performance discharges. The experiments gave interesting information
about the role of sawteeth with respect to impurity accumulation. A compromise
seems possible, where the sawteeth are slowed down enough to achieve improved
confinement, but can still prevent impurity accumulation and also produce an
outward pinch that reduces central radiation. Results from preliminary experiments
performed with pellets plus lower hybrid indicated that good radiofrequency
coupling is possible at high field and density.
Installation of the new boronisation system allowed a significant decrease in
impurity contamination and radiated fraction. Consequently, it was possible to start
studies on the radiative improved mode in the last part of the 2001 campaign.
The new diagnostic for fast x-ray imaging, developed at Frascati, was installed on the
National Spherical Tokamak Experiment. The device is based on a micro-pattern gas
detector with a gas electron multiplier amplifier.
A workshop was held at Frascati on 22-23 November 2001 to discuss the mediumterm
FTU programme and to increase the participation of European laboratories in
the machine. The workshop was organised in plenary and parallel brainstorming
sessions, with each session chaired by an external participant. Forty people were
from ENEA Frascati and the Consiglio Nazionale di Ricerca Milan and about thirty
came from other labs (European, USA, Japanese and Russian). The discussions
focused on areas where FTU can provide unique results. Several interesting
proposals were made and are presently being considered for implementation.
The contribution of ENEA to the C4 Joint European Torus campaign in 2001 amounted to
about 1 ppy and was focussed on the activities of Task Forces S2 and H. Plasma
configurations with internal transport barriers of long duration (11s≈30τ E ≈τ R , with τ E the
energy confinement time and τ R the resistive diffusion time) were produced, thanks to the
current profile control now possible with the lower hybrid system.
1.1.2 Experimental results
Lower hybrid current drive and heating experiments at high density
The lower hybrid (LH) radiofrequency (rf) system in FTU (six gyrotrons,

10
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
f LH =8 GHz, two antennas) has achieved 2.2 MW, corresponding to a net power
density of 6.2 kW/cm 2 on the waveguide mouths, with an average reflection
coefficient of 10%. With this level of power, current drive (CD) is being studied in the
typical density range (i.e., at line-averaged density n _ e ≥1×1020 m -3 ) of a reactor
plasma.
In 2001, work was focussed on 1) studies on collisional coupling between electrons
and ions and on CD efficiency and 2) a way to establish and sustain internal
transport barriers (ITBs). For 1) the studies were performed at plasma current I p =0.5
MA and toroidal magnetic field B T ≤7.2 T so as to have
good LH-wave accessibility in the plasma core. For 2),
B T was in the useful range for the electron cyclotron
heating (ECH) power available in FTU (f ECH =140
GHz, P ECH up to 0.8 MW, B T =5.3 T for on-axis
electron cyclotron resonance heating). The discharges
run for the e - -i + coupling study exhibited complete
stabilisation of sawtooth activity, with more than 75%
of the current driven by the LH wave, an increase in
electron temperature of more than 2 keV and an
increase in the neutron yield of one order of
magnitude. Figure 1.1 shows the results for discharge
#20026 (I p =0.5 MA, B T = 6 T). The line-averaged
density at its maximum is 1.0×10 20 m -3 . The fraction
of driven current is estimated from V loop to be about
0.3 MA, with an increase from 1.8 to 3.8 keV in T e0 .
The neutron yield increases by a factor of 7,
corresponding to an increase from 1.2 to 1.55 keV in
T i . Even with full stabilisation of the sawtooth, m=1
activity persists. The launched N ⎟⎟
spectrum is peaked
at 1.82.
1.6
1.4
1.2
1
0.5
0
3.5
3
2.5
2
1.6
1.4
1.2
1
In a sawtooth-free plasma with weak or negative
central magnetic shear (WS-NCS) [1.1], the onset of
electron ITBs is generally indicated as a steep gradient in the electron temperature
(here q is the safety factor and the shear s is defined as s=rq’/q). To achieve and
sustain WS-NCS, it is necessary to break the “Ohmic” link between the electron
temperature and the current density profiles. One way to do this is to drive a
substantial fraction of the plasma current non-inductively and at the same time
produce a WS-NCS discharge.
MW keV keV V
1012s-1 1020m-3
0.5
0
2
1
0
n e0
V loop
T e0
T i0
Neutrons
P LH
Fig. 1.1 - Time evolution of
main plasma quantities in
a high-density LHCD
discharge. Top to bottom:
time traces of a) central
electron density; b) loop
voltage; c) central
electron temperature; d)
central ion temperature;
e) neutron rate; f)
coupled LH power.
0.45 0.5 0.55 0.6 0.65 0.7
Time (s)
Wide electron ITBs were obtained in FTU at a density of up to n e0 ~0.9 10 20 m -3
(n _ e ~0.6×1020 m -3 ) by combining ECH and lower hybrid current drive (LHCD) both
in full and in partial CD regimes. This shows that operations near the ITER density
and B T ranges do not prevent electron ITBs from setting in. The LH waves in FTU
control the current density profile j(r), driving a large part (sometimes all) of the
plasma current and heating the electrons, whereas the EC waves are used as a very
localised electron heating source at the resonance radius. The EC power is used
either to take advantage of the improved confinement by heating the plasma inside
the ITB or to enhance the peripheral LH power deposition and CD by setting the
resonance radius off axis.
[1.1] E. Barbato, Plasma
Phys. Control. Fusion, 43,
A287 (2001)
Two successful scenarios were developed and studied. In the first, LH waves
established full CD conditions and complete magnetohydrodynamic (MHD)
stabilisation, prior to EC-wave injection. The wave was launched during the current
flat-top, with the EC resonance located very close to the magnetic axis. In this way
ITBs were obtained at both low (n _ e =0.3×1020 m -3 ) and high (n _ e =0.6×1020 m -3 )

MW 1012 s-1 keV MA
1. MAGNETIC CONFINEMENT
11
1.1 Tokamak Physics
[1.2] G. Tresset et al., A
dimensionless criterion
for characterizing internal
transport barriers in
JET, accepted for
publication on Nucl. Fusion
Fig. 1.2 - Time evolution of
main plasma quantities in
an ITB discharge with
LHCD in the current
ramp-up phase and offaxis
ECH power (I p =0.5
MA, B T =5.5 T), compared
with an Ohmic shot. Top
to bottom: time traces of
a) plasma current; b) lineaveraged
density; c)
central electron temperature;
d) neutron yield; e)
LH and ECH power.
density. High central electron temperatures (T e0 >8 keV) were achieved at
n _ e =0.3×1020 m -3 , B T =5.3 T, I p =350 kA with P LH =0.6 MW and P ECH =0.35 MW. The
q profile was not measured, but transport simulations [1.1] showed a magnetic shear
reversal region at r/a≤0.35. At the border of this region, a large gradient developed,
L T
-1 =(dTe /dr)/T e =30 m -1 , corresponding to R/L T =28 (with R the major plasma
radius). The local ρ T * =ρ/L T value (where ρ is the ion Larmor radius for T e =T i ) was
≥ 0.03, which is double the ρ T * value considered in JET discharges as a threshold for
an ITB [1.2]. At higher density, n _ e =0.6×1020 m -3 (n _ e0 =0.9×1020 m -3 ), T e0 (central
electron temperature)=5.4 keV was achieved with P LH =1.7 MW, and P ECH =0.7 MW
at B T =5.3 T, I p =460 kA. The neutron flux also increased by a factor of 2 from the
Ohmic to the LHCD phase and reached a factor of 2.5 in the combined LH+ECH
phase. According to the transport analysis, the WS/NCS region extended up to half
radius because of a broad LH power deposition profile. At the border of this region
(r/a~0.5) L T
-1 =20 m -1 , corresponding to a local ρ T * , again larger than the JET
threshold value. The current was not fully driven by LHCD. The residual V loop was
≈0.4 V (fig. 1.2), corresponding to a residual Ohmic power P OH ≈ 0.2 MW. As usually
found with central ECH heating in conditions close to full LHCD, the hard x-ray
profile emission remained unchanged during the whole heating phase, which
indicates a stationary current density profile.
1020 m-3
0.5
0
0.5
0
5
0
0.2
0
2
1
I p
n e
T e0
Neutrons
0
0 0.1
P LH
PECH
0.2 0.3 0.4 0.5
Time (s)
In the second scenario,
both ECH and LHCD were
applied early on in the
discharge, during the
current ramp-up phase
(dI p /dt=2 MA/s), to take
advantage of any preexisting
WS-NCS associated
with initial non-relaxed
j-profiles. The ECH was
applied off axis, before
LHCD (P ECH ≈0.3 MW,
r dep /a=0.2), thereby
broadening the initial
temperature and possibly
triggering an off-axis
LHCD. Figure 1.2 shows
the time evolution of the
main plasma quantities.
The LH power was
injected in 100 ms in steps
of just over 0.3 MW up to
1.7 MW to compensate for the increasing electron density. In this way an electron ITB
(L T
-1 =20 m -1 , ρ T * >0.03) was sustained for more than 0.2 s (6-7 confinement times) well
inside the current flat-top. In this phase the driven current fraction was I LH /I p ~50%,
the central density increased up to n e0 ~0.8×10 20 m -3 , T e0 exceeded 11 keV, the ITB
footprint expanded from r/a~ 0.3 to ≈0.4 and T i0 went from 1 to 1.6 keV. The neutron
yield also increased during the main ITB phase and was three times larger than in a
reference Ohmic discharge. The ITB was terminated at t>0.34 s by an m=1 MHD
tearing mode related to a change in the current density profile after ECH switch-off.
The local transport analysis showed that, in the weak shear region, transport is
indeed reduced during the main heating. However, the low plasma volume still
involved implies that the global energy confinement time is generally in line with the
ITER89-P scaling, although it exceeds the ITERL-thermal scaling.

12
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
Energy transport and electron temperature profile stiffness with localised ECRH
Off-axis ECRH clearly reveals electron temperature profile stiffness in FTU [1.3],
particularly when absorption is located in the confinement region, i.e. outside the
sawtooth inversion radius (r/a > 0.2) but inside the radiation-dominated periphery
(r/a< 0.6). The typical marker of electron temperature profile stiffness, observed in
all similar experiments on ASDEX-U, D III-D, Tore Supra and TCV, is a step in the
radial dependence of the electron thermal diffusivity. The step is usually positioned
at the EC-wave absorption radius, particularly when the ECRH power density
greatly exceeds the Ohmic input. The step amplitude is just enough to keep the
temperature profile smooth. The gradient length L T =T e /∇T e of the profile hardly
changes from Ohmic heating to ECRH and is not influenced by ECRH intensity and
localisation.
Modulated ECH was applied to study electron temperature profile stiffness in FTU
plasmas during current ramp-up. Modulated ECH experiments at current flat-top on
ASDEX-UG [1.4] have shown that the heat wave propagates much faster outwards
than inwards, confirming the step-wise behaviour of thermal diffusivity at the EC
absorption radius. The experiments during current ramp-up were performed with
ECRH at a much lower power level than Ohmic heating in order to limit as much as
possible the impact of ECRH on profile shapes. In addition, target plasmas with very
different shapes were obtained through control of the breakdown and density buildup
phases. Figure 1.3 shows two typical targets, one with peaked temperature (and
current density) profiles, the other with flat-hollow profiles characterised by the
occurrence of typical double tearing modes. Heat wave propagation is much more
sensitive than power balance analysis to discontinuities in thermal conductivity. In
addition, by looking at the amplitude and phase radial distribution of electron
temperature oscillations, it can be excluded that the apparent drop in diffusivity is
due mostly to a heat pinch.
[1.3] S. Cirant et al., Proc.
14 th AIP Conf. on Radio
Frequency Power in
Plasmas (Oxnard 2001),
Vol. 595, p 338
[1.4] F. Ryter et al., proc.
28 th EPS Conf. on
Controlled Fusion and
Plasma Physics (Madeira
2001), Vol. 25A, p. 685
The experiments showed that in these conditions the low-high diffusivity transition
layer is not strictly positioned at the absorption radius and that it depends to some
extent on the profile shape. For a given position of the absorption layer (r/a≈0.25), in
Te (keV)
Te (keV)
2.5
2
1.5
1
0.5
0
3
2.5
2
1.5
1
0.05
#20144
#20146
a)
a)
b)
ρ ≈ 0.07
r ≈ r dep ≈ 0.28
P ECH
0
0.15 0.25 0.35 0.45
Time (s)
a)
200
100
0
200
100
PECH (kW)
PECH (kW)
Fig. 1.3 - Evolution in time of a) electron temperature on axis and at the deposition radius; b) temperature
profile for two discharges characterised by very different profile shapes. The heat wave is launched at the
EC wave absorption radius, which is well inside the flat region for shot #20144 and in the steep region in for
shot #20146.
Te (keV)
3.5
3
2.5
2
1.5
1
0.5
t = 0.10 ÷ 0.17 s
δt = 0.1 s
#20144
#20146
0
0.7 0.8 0.9 1 1.1 1.2 1.3
R(m)
b)

1. MAGNETIC CONFINEMENT
13
1.1 Tokamak Physics
Fig. 1.4 - Key elements
showing consistency
between critical gradient
modelling and experimental
data. The
effective gradient length
(open dots) saturates (at
≈10) when a critical value
derived from ETG
turbulence (x) is
exceeded (at r=5-7 cm).
Both steady-state and
transient thermal
diffusivity switch from
low to high values at
almost the same radial
position.
[1.5] A. Jacchia et al.,
Proc. 14 th AIP Conf. on
Radio Frequency Power in
Plasmas (Oxnard 2001),
Vol. 595, p.342
[1.6] G.T. Hoang et al.,
Phys. Rev. Letts 12,
125001 (2001)
R/LT,e
slow
low χ e,hp
#20145(0.140 − 0.160 s)
heat wave
fast
high χ e,hp
R/L T,e -experiment
R/L T,e,crit = 5 + 10 s/q (T.S.)
χ e (p.b.)
χe (m 2 /s)
the case of peaked discharges the
narrow EC deposition occurs mostly
in the high-diffusivity region, while
for flat-hollow discharges it is
located well inside the lowdiffusivity
central volume. The step
in diffusivity appears, therefore, to
depend on the gradient profile
shape, which is consistent with the
assumption that the maximum
temperature gradient length is
limited below a critical value.
In fact, the critical gradient length
model gives a good description of
most experimental findings on
profile stiffness in steady state [1.5].
The results of modulated ECH
experiments on current ramp-up can be consistently included within this
framework, as shown in figure 1.4. Firstly, the plasma column appears to be divided
in two regions, each with different confinement properties. Secondly, the radial
position of the step in both transient and steady-state electron thermal diffusivity
almost coincide. Thirdly, the low-high diffusivity transition layer is located where
the effective gradient length, which decreases with increasing radius, stabilises
around a critical value. All these features can be explained if it is assumed that
electron thermal transport is enhanced in the plasma region where 1/L T exceeds a
critical value 1/L T,c that depends on local plasma parameters.
Assuming as the critical gradient length L T,c the value of the actual L T at the
transition layer, data from different discharges can be correlated with the
corresponding local magnetic shear, as also observed on Tore Supra [1.6]. FTU data
show a dependence of L T,c on the s/q parameter very similar to Tore Supra, in spite
of the different electron heating methods (ECRH for FTU, fast wave in the ion
cyclotron frequency range for Tore Supra). This dependence is consistent with
theoretical predictions based on electron temperature gradient turbulence.
1.1.3 Downshifted and upshifted experiments with ECRH in LHCD
plasmas
To increase CD efficiency, the EC wave can be injected on a LHCD sustained plasma
by exploiting the suprathermal absorption mechanism. The presence of fast electrons
generated by LH waves allows the cyclotron resonant frequency to be shifted up or
down from the cold resonance, depending on the launched N ⎟⎟EC , the magnetic field
and the fast electron distribution.
In the downshifted configuration (B T in the range of 6.9 -7.2 T and the cold resonance
outside the plasma), up to 80% of EC power absorption is observed, with increments
in electron temperature (∆T~ 1 keV) and driven plasma current (up to ∆I p ~ 35 kA for
a plasma with I p =350 kA, =0.5×10 20 m -3 ). Electron cyclotron power absorption
results in a loop voltage drop and in an increase in fast electron energy. The EC
power absorption is closely related to the fast electron tail density that corresponds
to the absorbed LH power, and is in agreement with a linear model of suprathermal
interaction of the EC wave.
In preliminary experiments on the up-shift scheme, 700 kW of EC waves were
injected with 30° of toroidal angle in plasma with LHCD (1.5–2 MW), resulting in

14
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
partial CD. The central field was varied in the range 4.8-5.2 T, with I p =400–600 kA
and =0.5–0.8×10 20 m -3 . In line with theoretical predictions, the single-pass
absorption was well localised, occurring only on the low-field side where the wave
beam encounters (at r/a ≈0.5) the resonant fast electrons before it is fully absorbed
by the bulk resonant layer. The resulting EC current produced a local modification of
J(r), as observed from the reduced MHD activity and the widening of the fast
electron bremsstrahlung emission profile. The increase in the driven current
(∆I≥100kA) as calculated from the drop in loop voltage was larger than that
calculated from theory. The resulting increase in CD efficiency was above the error
bars and indicates a synergy process between the two waves.
1.1.4 MHD behaviour in improved confinement regimes and new
phenomena by fast MHD analysis
Discharges that exhibit improved confinement after pellet injection are characterised
by a change in the central MHD behaviour [1.7, 1.8]. The optimum condition is an
increase in the sawtooth period to values (20-100 ms, the typical pre-pellet value
being 5 ms) that are a significant fraction of the energy confinement time. The main
parameter controlling the post-pellet period is the pre-pellet central temperature
(fig. 1.5), for a wide range of plasma densities and plasma currents. This dependence
can be easily understood because pellet penetration is a strong function of electron
temperature. If pellet ablation is completed well outside the q=1 surface, the
sawtooth period barely changes. In the other extreme case, if part of the pellet is
ablated inside the q=1 surface, the sawtooth is completely suppressed. In the
intermediate case, the optimum condition is attained. Complete sawtooth
suppression can give transient confinement improvement, but in this case impurity
accumulation takes place and this can lead to central radiative collapse.
The temperature dependence was exploited to attain controlled access to pellet
enhanced performance. The pre-pellet temperature decreases with increasing
density, and in a first stage gas puffing was used for control. Another method that
proved more efficient at plasma currents I p >1 MA was based on pellet sequence
timing: a first pellet was used to cool the plasma, and the timing of the second pellet
was optimised to meet the optimum temperature in the subsequent re-heating phase.
In addition to sawtooth period modification, pellet injection produced MHD
phenomena of fundamental interest [1.9, 1.10]. In particular, macroscopic structures
with dominant m=1 poloidal mode number were observed to saturate at large
amplitudes and to survive across sawtooth collapses for times exceeding the resistive
diffusion period (fig. 1.6).
These structures were
recognised as m=1
120
magnetic islands with a
x
very strong soft-x-ray
100
emission from the o-point
region (fig. 1.7). The nonlinear
stability of these
The sawteeth are stabilised
80
islands seems to be due to
60
x
radiative cooling around
the o-point. In the absence
40
of sawtooth reconnection,
x
locking of the m=1 was
20
observed in some cases.
x
x
x x
x x
x
This phenomenon was due
x xx
x x x
x x x
0
to toroidal mode coupling. 1 1.5 2 2.5 3 3.5
T e (keV)
τst (ms)
x
x x x
I p < 1 MA
I p < 1 MA
[1.7] E. Giovannozzi et al.,
Proc. 28 th EPS Conf. on
Contr. Fusion and Plasma
Phys. (Madeira 2001), Vol.
25A, p. 69
[1.8] P. Buratti et al., Bull.
Am. Phys. Soc. 46, 156
(2001)
[1.9] E. Giovannozzi et al.,
Am. Phys. Soc. 46, 156
(2001)
[1.10] P. Buratti,
Turbolenza e strutture
non lineari coerenti in
FTU, invited oral
presentation, SIF,
LXXXVII Congresso
Nazionale (Milano 2001)
Fig. 1.5 - Sawtooth period
just after pellet injection
as a function of electron
temperature.

1. MAGNETIC CONFINEMENT
15
Fig. 1.6 - Time traces of
soft-x-ray emission showing
the coexistence of
m=1 oscillations and
sawteeth up to t=0.78 s.
Slowing-down and locking
occur afterwards. The
temperature trace has
very small oscillations,
showing that x-ray
modulation is due to
impurity trapping inside
the m=1 island. Oscillations
in the magnetic coil
signal are due to an m=2,
n=1 mode being forced by
the m=1 mode.
r (m)
0.15
0.1
0.05
0
−0.05
−0.1
−0.15
−0.15 −0.1−0.05
#18599 t = 0.812 s
r (m)
Fig. 1.7 - Reconstruction
of mode rotation by softx-ray
emissivity in the
poloidal section.
Te (keV) Soft-X
Soft-X
T/S
5
0
1
0.5
2.5
2
1.5
1
0.5
50
0
-50
SX@z = 0 cm
#18106 B tor = 7.1 T I p = 081 MA
SX@z = 10 cm
Magnetic coils
T e @ r = 0
0.7 0.75 0.8 0.9
Time (s)
0.85
1.1 Tokamak Physics
In fact, the n=1, m=1
seeded an n=1, m=2
island that was strongly
affected by wall braking.
A fast MHD data
acquisition system
allowing a 2-MHz
sampling rate has been
installed. Several new
phenomena have been
observed with this
system, such as fishbonelike
events during highpower
LH current drive,
mode locking and high
frequency modes.
Fishbone-like events occur above a power threshold
P LH >1.5 MW. The magnetic structure has a clear
(m,n)=(1,1) signature on the Mirnov coil diagnostic and
rotates in the electron diamagnetic direction. Hence,
these events are believed to be caused by the fast
electrons due to LH absorption.
Mode locking often preceded a major plasma disruption.
Locking of the (2,1) and (3,1) modes occurred where
1000
there was strong interaction with the mechanical
structures (poloidal ring structures supporting the
500 Mirnov coil system itself) in the vacuum vessel. The
location changed with the change in position of these
mechanical structures, suggesting that error fields alone
0
do not determine the lock position. The Mirnov coil
0.1
diagnostic system was removed at the end of 2001 to
reduce the incidence of disruptions provoked by mode
locking. At the same time, a new set of coils with a safer
mechanical structure was prepared for installation during the shutdown at the end
of 2001.
0 0.05 0.15
2000
1500
The MHD data were acquired with 250-kHz sampling rates over the whole discharge
for around 30 shots before the Mirnov coil diagnostic was removed. At the beginning
of these discharges, MHD activity started with very high m~20 modes cascading
down to m~10 in about 50 ms, with then a slower decay to m~4 in a time span of
200 ms. These MHD modes all rotated in the electron diamagnetic direction, while
the background broadband “noise” rotated in the ion diamagnetic direction at
apparently high (m,n) numbers. In some of these discharges, a large (2,1) mode
appeared in the current plateau and, after increasing to large amplitudes of ~1%,
locked until the end of the discharge. However, no disruption occurred over this
period of 1 s or more. During the growth and slowing down of the (2,1) mode,
another mode with (4→5,2) structure appeared at a much higher frequency (fig. 1.8).
Whereas the (2,1) mode started at around 5 kHz and slowed down to 0 kHz, the (4,2)
mode started at 42 kHz and spun up to 50 kHz. This all suggests that, during the
slowing down and locking of the (2,1) mode, changes occurred in the radial profile
of the radial electric field, which were perhaps due to plasma interaction with the
mechanical structures that act as limiters.

16
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
Table 1.I - FTU record discharges
Shot B I n T 0 Neutrons τ E H89P H97P n 0 T 0 τ e
[T] [MA] [10 20 m -3 ] [keV] [10 13 s -1 ] [ms] ( 10 19 m -3 keV/sec
11612 66 00.7 2.1 1.5 0.2 80 1.6 1.0 0.4
12744 77 00.8 3.0 1.3 0.5 90 1.6 1.2 0.9
18598 88 1.02 4.0 1.4 1.3 80-100 1.4-1.7 1.0-1.2 1.0
1.1.5 Pellet injection
The main performances
obtained with multiple
pellet injection are
summarised in table 1.I. At
8T, up to 5 pellets were
fired into a single discharge
at time intervals of 100 ms
so as to cover the entire
duration of the current flattop
[1.11]. The preliminary
results reported last year
have been confirmed and
high-performance pellet
discharges have become a
reliable and reproducible
scenario of FTU operation.
This has allowed a deeper
investigation into associated
transport phenomena,
with the inclusion of
particle and impurity
confinement.
f(kHz)
60
40
20
60
40
20
60
40
20
δB/B ||
Spectrum #20504; :4S310P Range: 10 ->1E-3%, Max: 0.100
0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4
m-number port 1, level:-100 Range: 4 ->-4 Max: 19 Min:-13
0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4
m-number port 1, level:-100 Range: 7 ->-7 Max: 34 Min:-33
0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4
δn e (m-3)
#18598
In all cases, improved confinement
is associated with the suppression
NGPS abl. code
or stabilisation of sawtooth activity: 2
post-pellet fast reheating combined
axis
with slow density decay increases
the plasma energy content, while
the Ohmic input power stays 1.5
around the pre-pellet level. Total
sawtooth suppression seems to
occur when the q=1 surface leaves
the plasma. Both suppression and
1
0.9 1 1.1 1.2
slowing down of sawtooth activity
R(m)
takes place only if the pellets
penetrate deeply enough. On a short time scale (~100µs), particles are rapidly
transported well beyond the pellet penetration point, as predicted by a neutral gas
and plasma shielding (NGPS) code and confirmed by fast ECE measurements under
an adiabatic assumption (fig 1.9). Asymmetries in the temperature profile during the
ablation process suggest the interaction of deposited matter with the m=1 magnetic
structures already present. After the end of the ablation, the central density decay
δne(1020 m-3)
2.5
Time (s)
0.05
10
10
10
10
10
4
2
0
−2
−4
−2
−3
−4
−5
−6
6
4
2
0
−2
−4
−6
Fig. 1.8 - MHD spectrograms
showing mode
amplitudes, toroidal (n)
number and poloidal (m)
number.
Fig. 1.9 - Particle
deposition derived from
fast ECE (nT=cost) at 50,
100 and 150 µs from
ablation start. Results of
NGPS simulation.
[1.11] D. Frigione et al.,
Nucl. Fusion 41, 1613
(2001)

1. MAGNETIC CONFINEMENT
17
1.1 Tokamak Physics
Fig. 1.10 - #12744.
Particle fluxes integrated
on the 0.16 m flux
surface: a) neoclassical
diffusion; b) neoclassical
diffusion minus Ware
pinch; c) experimental.
1.5
time is longer than a neoclassical
Particle flux at r = 0.16 m
prediction that includes the Ware
pinch. Figure 1.10 shows that across
the r=16 cm surface, where no particle
a)
1
source can be present after ablation,
the experimental particle losses are
b)
lower than the net computed
neoclassical flux. If the neoclassical
0.5
value is regarded as a lower limit for
c)
radial diffusivity, an anomalous
inward pinch is required to explain
the experimental observation.
0
0.60
0.65
0.70
The plasma impurity content plays an
Time (s)
important role in the evolution of the
discharge after strong pellet perturbation [1.12]. In some cases, hollow temperature
profiles are produced, usually leading to a major disruption when a further pellet is
injected. This happens when the density is raised close to a critical value determined
by the power balance between Ohmic heating and radiation losses from
molybdenum, at the plasma centre This criterion made it possible to prepare a
suitable target for the injection of a given number of pellets. Further investigation of
impurity control and the effect of additional heating is in progress. In 2002 another
injector is going to be installed in collaboration with Padua RFX in order to have also
a high-field-side pellet injection track. This should allow first-time studies on the
effect of the particle radial drift at high density and high field in view of an
extrapolation to ITER.
1021 s-1
1.1.6 Boronisation: plasma results
[1.12] D. Frigione et al.,
Proc. 28 th EPS Conf. on
Control. Fusion and Plasma
Phys, (Madeira 2001), Vol.
25A, p. 73
[1.13] M.L. Apicella et al.,
Proc. 27 th EPS Conf. on
Control. Fusion and Plasma
Physics (Budapest 2000),
Vol. 24B, p. 1573
The boronisation system for overcoming the problem of high Z eff values at low
electron density was tested for the first time in FTU in October 2001.
Regarding vacuum performance, the getter action of boron on low-Z impurities
causes a strong reduction (up to a factor of 2.5) in the overall degassing rate and in
the pressure limit, which is reduced by a factor of 1.7 after 1-2 days of operation
following a fresh boronisation. This condition lasts for a long time (≥ 300 discharges).
After boronisation, restart of operations as well as recovery from plasma disruptions
are immediate and are a strong indication of the reduction in light impurities.
Regarding plasma characteristics, there are two main results: for an Ohmic plasma
(n _ e ≤1.0×1020 m -3 ), the total radiated power typically drops from 70-90% to 35-40%,
and for I p =0.5 MA and n _ e =0.3−0.4×1020 m -3 , Z eff decreases from 6.0 to 2.2 (see
fig. 1.11). These results are due to the strong decrease in heavy-metal concentrations
(up to a factor of 5 for molybdenum) and to the getter action of boron on light
impurities (oxygen concentration in the plasma is reduced from 2.5% to 0.5% and the
carbon flux from the walls drops from 1.0x10 18 to 1.1x10 17 part/s/m 2 ).
However, an unfortunate consequence of boronisation with cold walls is that it is
difficult to control the plasma density during the first two days of operation (about
60 discharges). The wall can either pump or release a large amount of H, depending
on the saturation degree of the surfaces facing the plasma. This is similar to what has
been observed after strong titanisation [1.13]. These phenomena do not occur in
other tokamaks, which operate at high wall temperature (>150°C), because hydrogen
is pumped by the B film during a pulse and then released immediately after it.
Preliminary observations suggest that high recycling or high edge neutral density
could hinder ITB formation in FTU.

18
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
Another result of boronisation is
that hydrogen particles released
from the B film dilute the plasma.
The ratio of deuterium to hydrogen
+ deuterium fluxes, as measured by
the neutral particle analyser, can be
as low as 40% after a fresh
boronisation, despite pure D 2
puffing. The ratio then increases
slowly to 85% after about 200
discharges. The D-dilution, in turn,
reduces the plasma performance in
terms of neutron yield. The target
plasma used for pellet injection
(n _ e =1.7×1020 m -3 ) shows a much
lower radiated power than before
boronisation (P rad /P tot ≈35%
against 65% and Z eff ≈1 against
Z eff ≈1.4), but the neutron rate
decreases by up to a factor of 5. This
decrease is in good agreement with
simulations performed with the
EVITA transport code [1.14]. The
same code also shows that neither
the electron nor the ion transport
coefficients show any significant
difference after boronisation.
Zeff Prad/Pohm(%) ne(x10 19 m -3 )
The best plasma performance
0.0 0.5 1.0 1.5
following boronisation was
Time (s)
achieved only after about 100
discharges, when the boron had
been eroded by the limiter but was still present on the chamber walls. The metal
influx was lower than before boronisation because physical sputtering by oxygen
ions and atoms was strongly reduced, and it was possible to control the edge
temperature with D 2 gas puffing. For I p =0.5 MA and (n _ e =0.4×1020 m -3 ), low oxygen
(0.4%), molybdenum (0.1%) and iron (0.09%) concentrations were present in the
plasma with Z eff =3.0 and a total radiated power close to 65% of the input power.
During this phase, the reduction in Z eff allowed one of the best performances of FTU
to be reached in terms of the actual CD efficiency η CD . Full CD with η CD =0.2x10 20
Am -2 /W was obtained on a plasma target with I p =360 kA, B T =5.3 T, n _ e =0.4×1020
m –3 ) and P LH =1.5 MW, with only a small increase (from 1.5 to 2.2) in Z eff . With the
same plasma target, additional power P LH+EC =2.6 MW was coupled to the plasma
with Z eff =3.0, compared to 6.0 before boronisation, at lower power. Very good highdensity
plasmas were also obtained. With gas puffing, the density limit at I p =1.1 MA,
B T =7.2 T reached, with gas puffing only, n e =3×10 20 m -3 . By reducing the radiated
power, the boronisation technique has made it possible to study the so-called
radiative improved mode plasmas at higher densities than those of TEXTOR [1.15].
Neon is used as the injection gas until a fraction of 90% of the radiated power is
achieved, with a subsequent peaking of the density profile and an increase in the
neutron yield. In this case, no significant difference was found between the fresh and
old boronisation. The relative neutron rate production increases by a factor of 3-4 in
both cases, but the starting level after a fresh boronisation is about five times lower
due to H dilution. To overcome the problem of H dilution and hopefully to exceed
5
4
3
2
1
100
80
60
40
20
12
10
8
6
4
2
0
Fig. 1.11 - a) Line-averaged
density; b) ratio of
radiated to Ohmic power;
c) Z eff for two Ohmic
discharges at I p =0.5 MA:
(blue) before boronisation
and (red) after boronisation.
[1.14] V. Zanza.
http://efrw01.frascati.
enea.it/Software/Unix/F
TUcodici/evita
[1.15] B. Unterberg et al.,
J. Nucl. Mater. 266-269,
75 (1999)

1. MAGNETIC CONFINEMENT
19
the neutron production of the best FTU performances, deuterate diborane (B 2 D 6 ) is
going to be used as the working gas in the near future.
1.1.7 Radiative improved mode in Ohmic plasmas
1.1 Tokamak Physics
In many tokamaks, controlled injection of gas impurities (mainly noble gases such as
Ne, Ar and Kr) into the plasma has been found to lead, in certain conditions, to an
improved confinement regime, the so-called radiative improved (RI) mode. The
interest in this regime for FTU lies in the fact that it can be obtained with different
magnetic configurations (circular or elongated plasmas, limiter or divertor) and
different heating systems (neutral beam injection, ICRH and Ohmic). Moreover, it
couples good energy confinement with a large fraction of radiation losses (up to 90%
of total input power), thus alleviating the problems of plasma-wall interactions. Of
course the price to be paid is a larger Z eff and greater plasma dilution.
[1.16] M.Z. Tokar et al.,
Plasma Phys. Control.
Fusion 41, B317 (1999)
[1.17] M. Bessenrodt-
Weberpals et al., Plasma
Phys. Control. Fusion 34,
443 (1992)
The strategy to look for the RI mode in FTU is based on the interpretation given by
TEXTOR [1.16]: impurity injection attenuates the growth rate of the ion temperature
gradient (ITG) instability. This leads to a smaller particle outflow and hence to
peaking of the density profile. As a consequence, ITG turbulence is further
attenuated, or even quenched. In cases where ITG turbulence is the dominant heatloss
mechanism, an increase in energy confinement is achieved. In addition, it has
been found that in TEXTOR the energy confinement time increases with density.
An experimental campaign was started at FTU at the end of 2001 to explore the
possibility of a RI mode in Ohmically heated plasmas. The aim was to reproduce the
improved Ohmic confinement (IOC) regime of ASDEX. [1.17]. The plasma target was
chosen so as to clearly see this regime, if it really did exist in FTU. At magnetic field
B T =6 T, the plasma current was programmed to be at 0.8-0.9 MA to avoid the
insurgence of MARFEs. The operational density was set at 10 20 m -3 in order to be
well into the saturated Ohmic confinement (SOC) regime, where energy confinement
is independent of density. In FTU the critical density to access SOC is ~0.8 10 20 m -3 .
Deuterium gas puffing was interrupted at 0.45 s, just at the beginning of the current
flat–top, according to the experience on ASDEX. A neon puff (10-30 ms duration)
was injected at 0.6 s, just at the beginning of the current flat–top.
The standard FTU diagnostics was used to obtain the experimental results.
Fig. 1.12 - a) Lineaveraged
density; b) Z eff
from bremmstrahlung; c)
radiated power; d)
neutron yield for a
discharge without Ne
(red) and with Ne (violet)
puffing of 20 ms at 0.6 s.
1012 (s-1) 105 (W) Zeff 1019 (m3)
10
5
3.5
3
2.5
2
1.5
10
5
1
0.5
a)
b)
c)
d)
x
x
0
0 0.5
1
Time (s)
x
x
x
x
x
x
x
x
01
02
03
04
x
Figure 1.12a shows the
central line-averaged
density for a discharge
with a Ne puff of 20 ms
compared to a reference
discharge without Ne.
Two different phases
can be seen: First, there
is a slow increase in
density after Ne
injection, which cannot
be fully accounted for
by the electrons
contributed by Ne
ionisation. The Ne
concentration can be
roughly derived from
the variation in Z eff
(fig. 1.12b). Radiation

20
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
power losses increase and
reach 90% of the total
power (fig. 1.12c). At the
end of the pulse the
density abruptly increases
at a stronger rate, up to a
disruption. Neutron yield
(fig. 1.12d) increases by a
factor of ~2 after the Ne
puff.
Figure 1.13 shows the
total plasma energy,
calculated assuming
T i =T e , and the Ohmic
power for the two
discharges. Since at
equivalent Ohmic power
the thermal energy is
larger for the Ne-puffed
10-2 (s) 106 (W) 104 (J)
6
4
2
0
1.5
1
0.5
0
5
4
3
2
1
a)
b)
c)
0 0.5
1
Time (s)
shot, an increase in confinement time is derived. Density profiles are more peaked
after the Ne puff, while the electron temperature remains the same or even increases
a bit. To get more accurate values of these parameters requires simulation with a
transport code. The later, sudden density increase in the discharge with Ne is still
unexplained and is being analysed. A preliminary conclusion is that this regime has
all the signatures of a typical RI mode. A dedicated experimental campaign is to be
carried out next year to compare this improved regime with those observed in other
tokamaks.
x
x
x
x
x
x
x
x
x
x
x
01
02
03
Fig. 1.13 - a) Total
thermal energy (T i =T e );
b) Ohmic power; c)
energy confinement time
for a discharge without
Ne (red) and with Ne
(violet).
1.1.8 Fast x-ray imaging of the NSTX plasma by a micro-pattern gas
detector with a GEM amplifier
A new diagnostic device in the soft x-ray range has been developed at ENEA Frascati
for imaging of magnetic fusion plasmas. It is a pinhole camera with a micro-pattern
gas detector (MPGD) and a gas-electron multiplier (GEM) as the amplifying stage. A
readout board with 144 pixels (12×12) was designed and coupled to the GEM
detector, which has a 2.5×2.5 cm active area. The electron signal, corresponding to the
detected x-ray photon, is collected at the pixel and processed by a fast charge preamplifier
(LABEN 5231) and an amplifier (LABEN 5185). The data acquisition
system, carried out in VME standard by CAEN, is formed of discriminators and
counters for a total of 144 channels. The fast, low-noise electronics coupled to the
discriminators and asynchronous scalers ensure high-quality data that has only
statistical noise and single-photon counting at high rates of up to 10 7 ph/s×pixel and
high frame rates of up to 100 kHz.
The spatial resolution and imaging properties of the detector, fully illuminated by
very intense x-ray sources (laboratory tube and tokamak plasma) and under the
conditions of high counting rates and high gain, are reported in a previous work
[1.18].
[1.18] D. Pacella et al.,
Rev. Sci. Instrum. 72,2,
1372 (2001)
The system was successfully tested at FTU with a 1-D perpendicular view of the
plasma. It was then installed and used at the National Spherical Tokamak
Experiment (NSTX) with a full 2-D tangential view, in the framework of a
collaboration between ENEA, Princeton Plasma Physics Laboratory and John
Hopkins University.

1. MAGNETIC CONFINEMENT
21
1.1 Tokamak Physics
Detector parameters
The gas mixture used for the MPGD is 80% Ne and 20% dimethyl ether. For the
specific application, the operational voltages of the chamber are determined by two
requirements: an energy range of 3– 8 keV and very high counting rates. The lower
limit in energy is related to the present experimental set-up, i.e., a thick beryllium
window on the tokamak and air between the window and the detector. In future this
limit will be further lowered. The voltage differences are 600 V for the induction gap
and 480 V for the gem foil; the drift cathode is polarised to 3000 V. The printed circuit
board is grounded and all the applied voltages are negative.
The energy calibration was performed with a 10-kV x-ray tube. The gain of the
electronic amplifier of each pixel (144) was adjusted to reproduce the same spectrum,
with a precision of about 2%. Since each channel (144) behaves as an independent
spectrometer, this fine calibration is needed to get the same spectral response in
order to exploit the combination of imaging capability and energy discrimination,
which is one of the most powerful features of this system. The energy resolution of
the detector in this range of energy is about 20%, so the electronic discrimination of
the pulse amplitude can be sharp (with 20% uncertainty). The resolution can also be
changed dynamically during the shot.
Results on NSTX
Fig. 1.14 - 2-D x-ray image
superimposed on the
reconstruction of the
magnetic surfaces of
NTSX for shot #107332.
The 2-D image obtained by the detector was superimposed on the EFIT-code
reconstruction of the magnetic surfaces of NSTX for shot #107332 (fig. 1.14). The
detector view of the plasma for this shot is about 80 cm×80 cm. It is evident that the
spatial distribution of the photon counts, represented by different colours, is in good
agreement with the plasma magnetic reconstruction, despite the fact that the
tangential view in a spherical tokamak integrates over a large part of the plasma.
This 2-D image of the cross section
of the plasma core is very clear
because of the energy
discrimination capability of this
device. The effect of integration of
the plasma emissivity along the
line of sight, is indeed, strongly
reduced because the photons were
selected in the range 3-8 keV
(central electron temperature no
higher than 1 keV) and therefore
all the photons emitted outside the
central core are neglected.
The time histories of the camera
pixels were studied in different
plasma conditions and the results
compared with those of other
diagnostics. At the beginning of
the discharge (1-kHz sampling),
the minimum counts/pixel
needed to recognise the core
structure is about 20; the
maximum counts/pixel, when
H–mode appears and neutral
beam power reaches the
maximum, is about 5000. The

22
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
noise of the detector is no more
than 5 counts/pixel. Therefore the
signal-to-noise ratio at the highest
emissivity can be estimated to be
about 1000 and the dynamic range
of the system to be about 300. The
maximum counting rate per pixel,
before saturation, is 10 7
ph/s×pixel.
The capability to get very clear
images of the core can also be
observed during MHD instability.
The time history of a few central
pixels, in the presence of sawteeth,
exhibits strong oscillations, while
the central lines of sight of a
vertical and horizontal
perpendicular array of x-ray
diodes show just weak
modulations. This is related partly
to the effect of integration along
the line of sight and the energy
discrimination of the MPGD
system and partly to the tangential
view of the pinhole camera.
Fig. 1.15 - A high
acquisition frame rate
(50 kHz) showing the
capability of the
diagnostics and the
negligible influence of
statistical noise.
The time history of the whole plasma discharge can be acquired at a frame rate of
10 kHz. Higher acquisition rates, up to 1 MHz, can be set for shorter time intervals.
Increasing the rate reduces the counts per pixel and increases the statistical noise. A
reasonable limit is 100 kHz, where the statistic is still acceptable. Figure 1.15 shows
a frame with a 50-kHz acquisition rate. The plasma core exhibits the same shape,
despite the lower statistic, when compared with a frame acquired at 1 kHz, 2 ms
before the switch from 1 to 50 kHz (shot #107356).
In shot #107316, a phase of lack of confinement (normalised beta drops from 4 to 2)
lasting about 20 ms was observed with 10-kHz frame rates. The x-ray image of the
plasma cross section shows apparent poloidal rotations in the electron diamagnetic
direction, with a period of about 400 µs, and the loss of energy from the core is clearly
related to these rotations.
As the system is a pinhole camera, it has the flexibility and versatility of an optical
device. It is, therefore, easy to change the magnification of the plasma image or the
line of sight of the view to study off-centre plasma. The x-ray emission has been
filmed from many different views off centre and at different magnifications.
1.1.9 JET
ENEA Frascati contribution to JET Activity
The contributions (about 0.9 ppy) of the ENEA Frascati group to the activities of Task
Forces H, S2 and M in the one JET experimental campaign (C4) of 2001 were focussed
on the areas of data analysis (about 1.5 ppy) and scientific coordination of the C4
experiments reported below.
• Task Force H experiments aimed at maximising LHCD power in order to control

1. MAGNETIC CONFINEMENT
23
[1.19] A. A. Tuccillo et al.,
Proc. 14 th Topical Conf. on
Radio Frequency Power in
Plasmas (Oxnard 2001),
Vol. 595, p. 209
[1.20] V. Pericoli-Ridolfini
et al., Proc. 14 th Topical
Conf. on Radio Frequency
Power in Plasmas (Oxnard
2001), Vol. 595, p. 245
[1.21] V. Pericoli-Ridolfini
et al., Study and
optimisation of lower
hybrid waves coupling
in advanced scenario
plasmas in JET, to be
submitted to Plasma Phys
Control. Fusion
the q profile during the high phase of advanced-scenario plasmas.
• Task Force S2 experiment aiming at steady-state ITBs.
• Task Force M experiments on neoclassical tearing mode (NTM) stabilisation by
LHCD and some experiments on the error field.
The H and S2 activities were carried out in close collaboration with CEA and
UKAEA.
Task Force H
1.1 Tokamak Physics
The activity here consisted mainly in continuing LH coupling optimisation in
relevant scenarios. After the success of the previous campaigns in 2000
[1.19,1.20,1.21], efforts were concentrated on the utilisation of LH to actively control
the q profile. As expected, LH proved to be very effective, both in the pre-heat phase,
to optimise the target, and in the high power phase, to maintain the q profile, of
advanced-tokamak plasmas. An example of the capability of LH to model the q
profile is given in figure 1.16. During plasma current ramp-up in the pre-heat phase
of advanced scenario plasmas, the target q profile can be changed from weakly
reversed to deeply reversed by adjusting the power level. This tool has made it
possible to lower the power threshold of electron ITBs, and with the improvement in
LH wave coupling, more than 3 MW can be routinely coupled in H-mode in ITB
plasmas.
Task Force S2
[1.22] F. Crisanti, et al.,
The new LHCD capability strongly accelerated the progress of Task Force S2. Quasi
JET quasistationary
steady-state ITBs time limited only by technical constraints were achieved, and full
internal-transportbarrier
operation with
CD was obtained during the whole high-performance phase of 1.8-MA ITB
discharges [1.22]. Later on during C4, the ENEA collaboration was extended to
active control of the
experiments on feedback control of pressure and temperature profiles by using,
pressure profile, accepted
for publication on
respectively, neutral beam and ion cyclotron waves [1.22]. Figure 1.17 reports the
time evolution of a few quantities characterising one of the longest ITB discharges
Phys. Rev. Lett.
(q 95 ≈6.0, β p ≈1.1, β N ≈1.7, β T ≈1.%, B T =3.4 T and I p =2 MA). In this discharge, an
electron ITB is triggered at the beginning of LH coupling. The ion ITB forms
immediately after neutron beam injection (NBI) and ICRH. Both barriers disappear
when the additional power is switched off and are time limited only by JET
hardware constraints. The loop voltage close to zero and the internal magnetic
inductance practically
constant all over the highpower
phase indicate a
EFIT + MSE at 44.4S
B T = 2.6 T
5
51164
possible “freezing” of the
51465
CD profile due to LHCD.
51466
The electron barrier lasts
P LH = 2.2 MW
about 37 energy confinement
times and the ion
Fig. 1.16 - Control of q
4
51467
profile with LHCD
barrier about 27. The
preheating in the
duration of the ITB is,
optimised shear scenario.
however, comparable with
Weakly to deeply 3
the current resistive
reversed q profile as a
diffusion time.
function of LHCD power.
profile
2
2.0
P LH = 1.1 MW
2.5
3.0
R maj (m)
3.5 4.0
In-depth analyses aimed at
understanding turbulencereduction
mechanisms were
performed and reported at
several conferences, e.g. the
2001 EPS and APS. The

24
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
studies were focussed
mainly on the role played
by magnetic shear, under
the hypothesis that
turbulence is stabilised
when the ExB shearing rate
exceeds the linear growth
rate of ITG modes.
Task Force M
0
T io (keV)
12
Due to problems with the
8
plasma position feedback
(pickup from the generated 4
T eo (keV)
mode), the experiments on 0
Z eff (0)
LHCD stabilisation of the 6
NTM had just a couple of 4
n
useful shots. Only a slight
eo (1019m-3)
2
H
destabilising effect was
89 β N
observed in these 0.9
discharges, suggesting 0.6
li
Vs
interaction between LH 0.3
waves and the mode, with 0
inappropriate localisation
2 4 6 8 10 12 14
of the power. This is
Time (s)
encouraging in view of
continuing the experiments in the 2002 campaigns.
The aim of other Task Force M experiments was to study the behaviour and
threshold scaling for error-field-induced locked modes at high beta poloidal. In past
experiments on DIII-D, it was observed that the penetration threshold was lower at
high beta. Although the observations made at JET are still inconclusive, a
phenomenology has been observed that differs both from the classical penetration
observations and from the onset of neoclassical tearing modes. Lack of power made
it impossible to perform experiments far from the natural threshold for the onset of
2/1 neoclassical tearing modes.
Internal transport barrier analysis in JET
3
2
1
0
12
8
4
P LHCD (MW)
P NBI (MW)
I p (MA)
R nt (1015n/s)
P ICRH (MW)
Analysis of the ITB in JET discharges was continued during the C4 experimental
campaign (January-February 2001) and was focussed in particular on the effect of
magnetic shear on ITB formation. The IDL code previously developed [1.23] to
calculate the radial electric field (E r ) in the plasma, the E×B flow shearing rate (ω s )
and the linear growth rate of ITG modes (γ ηi ) in ITB discharges was upgraded. In
fact, γ ηi can be now calculated using an explicit dependence on the magnetic shear s
(as given either by gyrokinetic and gyrofluid codes or by theoretical predictions). The
results were applied to the analyses of ITB discharges from the C2 and C4
campaigns. It was found that by taking into account the dependence of γηi on s, it is
qualitatively possible to explain the radial location, the time of formation and the
time evolution of different kinds of transport barriers in terms of the E×B shear flow
suppression of ITG-driven electrostatic turbulence [1.24]. In addition, the carbon
poloidal velocity (which is another output of the above IDL code), calculated
according to the neoclassical theory, was compared with the results of the new JET
spectroscopic diagnostic system (currently being commissioned), which provides a
measurement of the impurity poloidal velocity. Reasonable agreement between the
data was found [1.25].
Fig. 1.17 - Shot #53521.
a) Plasma current - LHCD
power; b) NBI and ICRH
power - neutron yield; c)
central ion and electron
temperature; d) central
electron density and
effective Z-H 89 β N ; e)
V loop and l i .
[1.23] F. Crisanti et al.,
Nucl. Fusion 41, 883
(2001)
[1.24] B. Esposito et al.,
Proc. 28 th EPS Conf. on
Contr. Fusion and Plasma
Phys. (Madeira 2001), Vol.
25A, p. 553
[1.25] F. Sattin et al.,
Proc. 28 th EPS Conf. on
Control. Fusion and Plasma
Phys. (Madeira 2001), Vol.
25A, p. 373

1. MAGNETIC CONFINEMENT
25
[1.26] M.J. Mantsinen et
al., ICRF heating
scenarios in JET with
emphasis on 4 He plasmas
for the non-activated
phase of ITER, presented
at the 14 th Topical Conf.
on Radio Frequency Power
in Plasmas (Oxnard 2001),
Vol. 595
[1.27] V. G. Kiptily et al.,
Gamma-rays: measurements
and analysis at
JET, presented at the
6 th Inter. Conf. on
Advanced Diagnostics for
Magnetic and Inertial
Fusion (Varenna 2001)
[1.28] V. G. Kiptily et al.,
Gamma-rays diagnostics
of energetic ions in
JET, presented at the
7 th IAEA Technical
Committee Meeting on
Energetic Particles in
Magnetic Confinement
(Goteborg 2001) to be
published on Nuclear
Fusion
[1.29] J. Mailloux,et al.,
Progress in internal
transport barrier plasmas
with lower hybrid current
drive and heating in JET
accepted for publication
0
-1
I p (MA)
-2
n e (1020
1
m-2)
0.5
0
T
2 i (10 keV)
1.5
1
T e (eV)
8
6
4
Power (10 kW)
1.5
LH
1 #53429
0.5
40
45
Time (s)
Gamma diagnostics on JET
1.1 Tokamak Physics
Studies of fast-ion production during heating and the subsequent fast-ion behaviour
in magnetically confined plasma, and evaluations of the resulting bulk ion heating
efficiency are essentially important to fusion reactor development.
Gamma-ray emission from nuclear reactions between fast ions and the main plasma
impurities was observed during ICRH and NBI heating in the JET tokamak. Gammaray
energy spectra provided information on the energy distribution function of the
fast ions. The gamma-ray emission profiles obtained with the JET neutron profile
monitor supplied information on the spatial distribution of reaction sites.
In recent JET studies of the ITER-like ICRH scenarios ( 3 He)D and ( 3 He) 4 He,
gamma-ray measurements gave invaluable information on the fast-ion population:
a) first evidence for ICRF-induced pinch of 3 He-minority ions based on profile data;
b) variation in the fast 3 He tail temperature that depends on 3 He concentration and
c) experimental simulation of 3.5-MeV fusion-born alpha particles by diagnosing fast
4 He ions accelerated to the MeV range [1.26,1.27,1.28].
Effect of low magnetic shear induced by LHCD on high-performance ITBs in
JET
In JET a low/negative magnetic shear profile is maintained in a plasma target with
2.4-MA plasma current by using 2.2 MW of lower hybrid power combined with NBI
and ICRH. In this scenario, an ITB up to about 4 s is produced. The fraction of LH
driven current is about 25% of the total plasma current. During LH power
application, the layer with reversed shear q-profile can be maintained in a suitable
radial position to inhibit the onset of turbulence, which might otherwise force the ITB
to collapse. Lower hybrid power could be used to drive moderate amounts of noninductive
off-axis current and sustain high-performance ITBs at high plasma current.
The main plasma parameters and additional heating power time traces of two
discharges of JET [1.29] are compared in figure 1.18. In shot #53432, an increase in the
central ion and electron temperatures is observed in the time range from 45.4 s to
46.8 s. In shot #53429 the central temperatures show a prompt increase at the switchon
of LH power, during the main heating phase. The increase is maintained for
longer (3.2 s, from 45.8 s to 49.6 s). In both shots, the temperature rise is accompanied
by a peaking of the temperature profiles, which suggests
the formation of an ITB. The ITB collapse is accompanied
a) by increased plasma-edge interaction, which produces
an increase in Dα emission and the loss of LH antenna
power coupling. On the other hand, in shot #53432 the
b) ITB duration appears to be only 1.2 s (from t=45.3 s to
NBI
#53429
ICRH e)
#53429
c)
d)
Fig. 1.18 - Time traces of the main plasma parameters
of JET discharge #53429 compared with a similar
shot (#53432) without LHCD coupling in the main
heating phase. Plasma current: a) # 53429 red curve,
#53432 pink. Line integrated plasma density: b)
#53429 red curve, #53432 black. Central ion
temperature: frame c) #53429 dashed/red, #53432
continuous line/pink. Central electron temperature:
d) #53429 squares/red, #53432 rhombus/pink e)
NBI power, #53429 red, #53432 blue; ICRH power,
#53429 pink, #53432 yellow; LHCD power, #53429
green, #53432 black.

26
1. MAGNETIC CONFINEMENT
1.1 Tokamak Physics
t=46.5 s). No change in
plasma-edge activity is
observed in concomitance
with the ITB collapse.
1
The effect of LHCD on magnetic
shear might have helped to
sustain the ITB in shot #53429, as
0
suggested by the modelling
analysis performed by the
JETTO code [1.30] and LH ray 2-
D Fokker Planck ray tracing. As -1
a result [1.31], the power is fully
deposited off-axis within the
layer ρ≈0.4-0.7, with a maximum
at ρ≈0.5 (ρ is the square root of
the normalised toroidal flux).
The ITB is located in this layer.
The calculated fraction of LH
0.3 0.4 0.5
ρ
0.6 0.7 0.8
driven current is
I LHCD /I P ≈0.25, while the
noninductive current fraction is
(I LHCD +I boot +I NBI )/I P ≈0.65. 0.5
The q-profile of discharge
t = 47s
#53429 simulated by the
t = 46s
JETTO code shows a reversed 0
shape during the main
heating phase. Figure 1.19
reports the simulated -0.5
magnetic shear profiles at
0.3
0.4
different times during the
ρ
main heating phase. After the
LH power is switched on, the
0.5
0.6
s=0 layer of the magnetic shear profile moves outward and persists in the region
ρ>0.3. As anomalous transport is dominant in this region, the low/negative
magnetic shear could inhibit the growth of turbulent modes that cause the ITB
collapse. No change in the q profile is expected at the time of the ITB collapse in the
experiment. However, the collapse might be related to edge physics, as it
accompanies an increase in Dα emission.
Magnetic shear s
Magnetic shear s
For shot #53429, with LH power coupled during the main heating phase, modelling
suggests that the collapse can be produced by an inward movement of the s=0 layer
[(ρ

1. MAGNETIC CONFINEMENT
27
1.2.1 FTU machine
1.2 FTU Facilities
Summary of machine operation
The FTU machine operated throughout 2001, with only the one summer shutdown.
No major problems arose during the experimental campaigns, except for a vacuum
leak during startup in February and, consequently, the loss of two experimental
weeks. In fact, during the experimental phase, with full available power, strong
influxes on the manganese caused systematic plasma disruptions. This behaviour
originated from the heat load on the stainless steel structure of the MHD rings that
acted as a limiter inside the vacuum vessel, so the rings were removed during the
summer shutdown.
With the new boronisation system, the scientific objectives of reducing Z eff and
radiative losses, especially at low density, were reached. The first boronisation of the
vacuum vessel in October 2001 was performed with the machine at room
temperature. Afterwards, it was done with the machine cooled down to liquid
nitrogen temperature and baking the vessel so that only one experimental day was
lost instead of a week.
The remote handling tools for vacuum-vessel inspections are continuously updated.
It is now possible to visually examine the vessel through only one port. The new
remote arm covers half the torus in one direction and the other half in the opposite
direction. The images are digitalised and stored in a computer connected to Internet
and available to all users.
The scientific exploitation of FTU data is strictly related to data-access tools, so
particular attention was devoted to this aspect. In addition to the AFS and the
MDSplus servers, a new data layer that is compatible with the standard CORBA was
developed to allow users to access data from their own PC by running a local Matlab
or IDL code. This effort can also be considered as part of the general issue of remote
participation, which has been greatly emphasised since the establishment of the
multilateral European Fusion Development Agreement (EFDA) on which the
European fusion research program is based. Again in this framework, the assessment
of video conferencing tools, namely the Virtual Rooms Videoconferencing System
(VRVS), was completed. Two permanent conference rooms and one movable station,
equipped with PCs, were set up for Local Presentation, VRVS mbone, VNC sharing
presentation transmission, unidirectional WebCast and Yahoo chatting. A number of
personal VRVS-desktop set-ups are also available, but have not yet been used.
Remote meetings were organised between ENEA Frascati, CNR Milano, CNR
Padova and EFDA JET.
Fig. 1.21 - Sources of
downtime.
During 2001, 1909 shots were completed successfully out of a total of 2117 performed
in 91 experimental days. The average number of successful daily pulses was 20.95.
Table 1.II gives the main parameters for evaluating the efficiency of the experimental
sessions. Figure 1.21 reports the source of downtime in 2001. It is worth noting that
the time required to analyse the discharges is still the main source of downtime.
During the experimental
Analysis
27%
Diagnostic
Systems
8%
Others
9%
Machine
9%
Control System
17%
Power Supplies
15%
Radiofrequency
15%
campaigns the tokamak
power supplies and the
control and data acquisition
system operated at a very
high level of availability and
reliability. The percentage of
time lost because of controlsystem
problems was reduced
to 17%.

28
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
Table 1.II Machine efficiency in 2001
January February March April May June July September October November December
Total pulses 111 180 238 352 406 92 130 141 265 202
Successful pulses (sp) 103 159 214 312 375 80 116 138 239 173
I(sp) 0.93 0.88 0.90 0.89 0.92 0.87 0.89 0.98 0.90 0.86
Total
2117
1909
0.90
Potential experimental days 10 9.5 11 18 17 4 5 7 12.1 10
Real experimental days 5 8 10 14 17 4 5 7 12.1 9
I(ed) 0.50 0.84 0.91 0.78 1.00 1.00 1.00 1.00 1.00 0.90
104
91
0.88
Experimental minutes 1819 2879 3971 6023 7640 1803 2165 2561 4845 3867
Delay minutes 1323 1925 2402 2403 2801 621 640 1622 2672 1958
I(et) 0.58 0.60 0.62 0.71 0.73 0.74 0.77 0.61 0.64 0.66
A(sp/d) 20.60 19.88 21.40 22.29 22.06 20.00 23.20 19.71 19.75 19.22
A(p/d) 22.20 22.50 23.80 25.14 23.88 23.00 26.00 20.14 21.90 22.44
37573
18367
0.67
20.95
23.24
DELAY FOR SYSTEM (minutes)
January February March April May June July September October November December Total %
MACHINE 0 43 466 52 159 193 29 19 140 228 274 1603 8.7
POWER SUPPLIES 0 192 528 894 310 239 22 31 287 10 279 2792 15.2
RADIO FREQUENCY 0 81 59 280 697 738 188 112 247 221 53 2676 14.6
CONTROL SYSTEM (PROMETEO) 0 347 235 319 232 402 223 24 19 344 287 2432 13.2
DAS 0 0 0 172 44 92 7 4 52 63 54 488 2.7
FEEDBACK 0 20 0 0 22 0 0 17 27 24 10 120 0.7
NETWORK 0 143 208 92 0 0 0 0 0 461 84 988 5.4
DIAGNOSTIC SYSTEMS 0 213 219 85 179 276 17 166 89 133 150 1527 8.3
ANALYSIS 0 279 209 350 634 785 135 265 735 1009 734 5135 27.9
OTHERS 0 5 17 158 126 76 0 2 26 179 33 622 3.4
TOTALE 0 1323 1941 2402 2403 2801 621 640 1622 2672 1958 18383 100
Summary of machine maintenance
Maintenance of the FTU system was carried out according to schedule. Visual
inspection of the vacuum vessel revealed ten displaced tiles, most of which in the
upper and bottom rows. The rupture was investigated through specific laboratory
tests and it was found that the fragility of tungsten-zirconium-molybdenum (TZM)
can damage the supporting-screw thread. A new design tile-support structure was
developed and will be tested in 2002.
The new data storage system based on SAN architecture was released, and the whole
FTU experimental data archive is on line. A preliminary test of Opto22 technology in
slow acquisition, i.e., replacing the traditional PLC, was carried out.
Future activities
In 2002 the machine will operate up to July and then from mid-September to mid-
October if the injector for launching pellets from the high-field side is ready. New
diagnostics and the new passive-active multijunction (PAM) lower hybrid launcher
will be installed during the second shutdown in 2002. All four ECRH gyrotrons will
be in operation, so a total power of 1.6 MW should be available. A new density
feedback system based on VME architecture will be implemented.
Boronisation system
Direct current glow discharge deposition is used to coat the vacuum vessel walls
with a boron film. The deposition system is the same as that used for the vacuum
chamber conditioning, but the vessel is fuelled with a mixture of helium and
diborane (90% He and 10% B 2 H 6 ). The aim of boronisation is to reduce the effective
charge Z eff and the plasma radiation losses by introducing a low-Z element as firstwall
material.

1. MAGNETIC CONFINEMENT
29
1.2 FTU Facilities
Fig. 1.22 - Schematic of
the boronisation plant.
FTU
VG1-1
IE
AD
AD
IE
VG1-2
F2
F2
F2
F2
VG1-3
AD
IE
AD
IE
VG1-4
FTU
FTU HALL
VG1-M
GAS PIPELINE
ABOUT 30 m
CABINET
IE1
FC
Al camino
NEUTRAL GAS MODULE
V0
V1
Bottle He
D1
Al camino
S1
CR1
VG1-1 = valve
VG1-2 = valve
VG1-3 = valve
VG1-4 = valve
VG1-M = valve
S2
GAS ACTIVE MODULE
CR2 V3
Vacuum
Pump
V2
MHP
V5
V4
F1
VB
V6
MBP
D2
MMP
Bottle
He (90%)+B 2 H 6 (10%)
F2 = filter a 0.2 µm
IE = electric break
IE1= electric break
FC = flux controller
[1.32] W.R. Baker et al.,
Vuoto, XXVIII, N. 3-4,
(1999)
[1.33] W. Braker and A.L.
Mossman, Matheson gas
data handbook, (Ed.
Matheson, 1980) pp. 219-
223
[1.34] W. Stopford and
W.B. Bunn, Effects of
exposure to toxic gases,
(Ed. Matheson, 1988) pp.
21-27
The FTU boronisation system was designed and installed by the Italian company
RIVOIRA on the basis of the previous experience at RFX [1.32]. Diborane is both
toxic and explosive and hence requires particular safety measures [1.33,1.34]. To
facilitate operation and improve safety, the whole system is equipped with its own
remote control and protection system.
The diborane system consists of the following main elements (fig. 1.22):
• A special gas cabinet for toxic gas mixtures, with its own exhaust system
comprising a gas extractor and a chemical filter to decompose diborane in the case of
accidental release.
• A chemical filter placed at the inlet of the rotary pump to evacuate the gas
immission lines from diborane at the end of boronisation.
• Pneumatic valves inside the gas cabinet, which are all operated with compressed
nitrogen to avoid the risk of sparks or fires.
• A diborane bottle with a remotely controlled pneumatic valve.
• A gas flux controller to regulate the gas-feed rate.

30
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
• Gas lines with special connections to guarantee both high- and low-pressure
tightness.
• Four vertical ports to introduce the gas, which are located 90° toroidally apart
from each other;
• Two evacuation lines of the main vacuum system of FTU, placed at the two
opposite sides of the torus. Each line has a standard 2000 l/s turbomolecular pump,
a thermal decomposer for diborane (operation temperature 500°C) and a special
rotary pump modified to assure very good vacuum tightness and to dilute the
exhaust gas with nitrogen before it comes into contact with the air;
• Diborane detectors (sensitivity 1ppb).
Some interlocks were installed to ensure safe operation. They cause automatic
closure of the diborane pneumatic bottle valve and the gas flux should a fault occur.
The walls are maintained at 373 K during film deposition and subsequently cooled
to 77 K. Two electrodes, 180° toroidally apart, are inserted from two vertical ports up
to the centre of the vacuum chamber. The glow discharge conditions are 7×10 -3 mbar
total pressure with 1.7 mbar l s -1 of average flow rate, a voltage drop of +360V and
a 0.75-A driven current for each electrode, corresponding to 11 mA/cm 2 of total
current density on the vessel walls. According to the laboratory tests on silicon films,
three hours are necessary to reach the target film-thickness of 100 nm [1.35].
First-wall TZM tile refurbishment
[1.35] M.L. Apicella et al.,
J. Nucl. Mater. 212-215,
1541 (1994)
During the FTU experimental campaigns, some first-wall TZM tiles were damaged.
The main problem occurred at the location of the tile attachment where brittle
behaviour was observed. To improve the attachment strength, a new type of tile was
designed (fig. 1.23) according to tensile test results and to a detailed stress analysis
simulating the force history of the threaded region. A new set of tiles was ordered
and will be supplied in 2002.
Electromagnetic loads on toroidal limiter tiles
Following the failure of the TZM tiles, the loads on the toroidal limiter, due to eddy
currents during fast disruptions, were investigated by detailed electromagnetic (EM)
analysis of the components. The analysis input was derived from the most
dangerous event observed in the FTU machine, and the numerical simulation was
done with the time evolving MHD equilibrium code MAXFEA. Figure 1.24 reports
the time behaviour of the main
macroscopic plasma parameters, the
finite-element model (FEM) and the
main results. The conclusion is that
although the EM loads on the limiter
tiles could determine breakage of
tiles already damaged, the loads are
not the main reason for the tile
failure.
VOM MISES (MPa)
1.2.2 Heating systems
LHCD system
In 2001 the last two gyrotrons,
delivered by Thomson and tested on
dummy loads at full power, were put
in operation, so the LHCD system is
> 5.00e + 02
< 5.00e + 02
< 4.17e + 02
< 3.34e + 02
< 2.50e + 02
< 1.67e + 02
< 8.39e + 01
< 6.91e + 01
max = 9.80e + 02
min = 6.91e - 01
Fig. 1.23 - FEM simulation
of first-wall TZM tile.

1. MAGNETIC CONFINEMENT
31
1.2 FTU Facilities
1.8
1.4
1.1
0.7
a)
now complete (six gyrotrons and two coupling
structures). In this configuration, the system has
routinely operated with a total rf power of about 2.2
MW coupled to the plasma. The maximum electron
temperature achieved is about 12 keV. The system has
operated in synergy with the ECRH system, reaching,
in this case, an electron temperature of about 15 keV.
0.4
The new launchers
0.0
0.0
1.6
3.2
Fig.1.24 a
4.8
1 . 10-3
6.4 8.0
b)
At the end of 2001, the conventional multijunction
(MJ) grill was delivered to ENEA. The MJ with passive
waveguides (PAM) and the ancillary components
(dummy loads, short circuits, couplers, etc.) needed
for the complete rf test of both the launchers were still
under construction.
ECRH system
In 2001 the ECRH system operated with two gyrotrons
providing a total power of 800 kW at the plasma, at
nominal pulse length. To improve the power supply
system, a new voltage reference generator was
developed and successfully installed. The new system
is based on National Instruments hardware, has faster
control and a greater rejection of EM noise.
1.2.3 Diagnostics
Moments (Nm)
Moments (Nm)
Moments (Nm)
100
0
-100
40
0
-40
-100
-140
0.001
0.002
0.003
0.004
0.005
+ x x x x + + + + + + + + + +
x
+ + +
0.001 x 0.002 0.003
0.004
x x
x
x 0.005
x
x x x
x
x
x
x
x
x
Time (s)
0.006
Time (s)
0.006
40 x x
Time (s)
x
x
0 x x x + + + + + + + + + +
0.001 x 0.002 0.003 0.004 x x 0.006
x
x 0.005
-40
x x x
x
x x
-100
x
x
x
-140
x
x
x
x
+
x
+
x
+
Mx_1
My_1
Mz_1
Mx_2
My_2
Mz_2
Mx_3
My_3
Mz_3
Mx_4
My_4
Mz_4
Mx_5
My_5
Mz_5
Mx_6
My_6
Mz_6
Mx_7
My_7
Mz_7
Mx_8
My_8
Mz_8
Mx_9
My_9
Mz_9
Fig. 1.24 a) Current disruption at 1.6 MA. b) FEM used
for EM analysis of the toroidal limiter: tiles are
blanked out to expose the model. c) Radial (x),
vertical (y) and toroidal (z) torque components on
each tile, relative to tile centres, plotted vs. time for
the disruption event described in a).
c)
Development of active beam diagnostics
Measurements of the motional Stark effect (MSE) and
of charge-exchange spectroscopy will yield data on the
radial profiles of q, ion temperature and poloidal
velocity. Both measurements will use a fast-hydrogenatom
(40 keV) injector, which is being tested at the
Frascati laboratory.
The characteristics of the neutral beam injector, which
was previously used in the Canadian TdeV
laboratory, were measured in a reduced configuration,
i.e., by using a single-hole extraction grid rather than
the full 19-hole grid. A new bolometer, consisting of a
set of 12 Cu plates whose temperature was measured
by embedded thermocouples, was used to measure
the output power.
The beam parameters (gas feed, decelerating and
accelerating grid voltages and compressor coil
current) were optimised by maximising the output
power and emitted light. The optimum conditions in
this situation could not be reached because it was
impossible to operate both the source and the
neutraliser at the desired pressures. A directly heated
W filament that is simpler and easier to use has
substituted the original LaB6 cathode, which had to be
replaced frequently and had irregular behaviour. A

32
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
new control system using modules addressable via the Net was set up and tested. In
the next phase, the beam will be tested at full power before installation on FTU.
The MSE optics has been designed with the observed chord divided in two sections
to reduce the aperture of each section. The relay optics will include two mirrors on
each section and a system of lenses to image the plasma outside the port. All of the
glass components will be made out of a low Verdet-coefficient material. The first
periscope components will be placed close to the plasma, inside a temperaturecontrolled
tube to avoid the window freezing in the cold FTU environment. The
photoelastic modulator polarimeter was pre-assembled, and lock-in detection was
compared with a Fourier analysis deconvolution for the interesting signal frequency
acquired on fast analog-digital converters (ADCs). The Fourier analysis was
preferred as it is less sensitive to hardware noise.
Charge-exchange spectroscopy will operate on 12 vertical lines and f/2 collection
optics. A high-aperture high-resolution image spectrometer with an echelle grating
has been set up and is being equipped with a high-gain image intensifier.
Oblique ECE measurements
It is generally assumed that the bulk electron distribution function is well described
by a Maxwellian distribution function. While qualitative theoretical arguments
support this assumption, determination of the exact form of the bulk distribution
function, in the presence of additional heating and transport processes, is beyond
today’s computational capabilities. From the experimental point of view, the
hypothesis of a Maxwellian distribution is at the basis of the interpretation of
temperature measurements. In the case of Thomson scattering, this is so even if the
scattered spectrum, in principle, contains information about the form of the
(typically perpendicular) 1-D distribution function, which could be used to ascertain
the above hypothesis. In the case of electron cyclotron emission (ECE), the
temperature measurements give the perpendicular slope of the distribution function
averaged over a small region of phase space (for optically thick plasmas), whose
extension is determined by the temperature and density profiles and the specific
instrumental parameters of the diagnostic. Multiharmonic ECE spectra, measured by
Michelson interferometry perpendicularly to the magnetic field, can provide a coarse
scan at different perpendicular energies, for low parallel energy.
Oblique ECE spectra can provide a continuous scan of the distribution in parallel
energy by changing the observation angle, for roughly constant perpendicular
energy. These two types of ECE measurements are complementary, and only their
combination can give a 2-D scan of the electron distribution function in velocity
space [1.36].
Recently, both theoretical [1.37] and experimental [1.38] evidence has emerged that
points to the existence of a distortion of the bulk electron distribution function
during on-axis ECH on FTU. The analysis of this phenomenon motivates the present
study, in which the first measurements on FTU with oblique ECE are reported.
An oblique view of FTU plasmas is achieved through one of the transmission lines
of the ECH system [1.39]. The line, consisting partly of closed metallic oversized
waveguides and partly of quasi-optical sections, was opened and collimating optics
was installed to focus the radiation emitted by the plasma and transported through
the ECH waveguide up to the receiving antenna. The emitted radiation is analysed
by a 32-channel heterodyne radiometer (2 nd harmonic, X-mode). The radiometer,
originally designed for perpendicular measurements (φ=0˚) during ECH with high
spectral resolution (∆f~1 GHz), was moved temporarily to the ECH line for the
oblique measurements. The ECH launcher, which is used to collect the radiation
[1.36] V. Krivenski and V.
Tribaldos, Proc. 20 th EPS
Conf. on Contr. Fusion and
Plasma Physics (Lisboa
1993) Vol. 17C, 1045
[1.37] V. Krivenski, Proc.
11 th Joint Workshop on
ECE and ECRH, Fusion
Eng. &. Des. 53, 23 (2001)
[1.38] O. Tudisco et al.,
Proc. 26 th EPS Conf. on
Contr. Fusion and Plasma
Physics (Maastrict 1999)
Vol. 23J, p. 101
[1.39] C. Sozzi et al., Proc.
13 th Topical Conf.
Applications of RF power
to Plasmas (AIP Press,
1999) p. 462

1. MAGNETIC CONFINEMENT
33
1.2 FTU Facilities
emitted by the plasma, can be steered at different toroidal angles (φ=0˚, ±10˚, ±20˚,
±30˚, with respect to the normal direction) and can also perform a continuous
poloidal scan. Due to the narrow frequency range of the radiometer (247-287 GHz),
observation of the plasma centre during on-axis ECH is possible only for φ=0° and
±10°.
In order to minimise the amount of O-mode radiation reaching the radiometer for
oblique viewing, a polariser was installed in front of the antenna. The polariser is a
quarter-wave plate, whose optical axis can be rotated to compensate for the change
in polarisation of the X-mode when the observation angle is changed. For these
preliminary measurements, the polariser is not a perfect quarter wave device and
therefore conversion of the X-mode elliptical polarisation into linear is not complete.
It was estimated that, for measurements at φ=±10°, the polariser reduces the
contribution of the O-mode emission from 15% to 4%.
[1.40] P. Buratti and M.
Zerbini, Rev. Sci. Instrum.
66, 4208 (1995)
[1.41] O. Tudisco et al.,
Rev. Sci. Instrum. 67,
3108 (1996)
[1.42] P. Buratti et al.,
Phys. Rev. Lett. 82, 560
(1999)
Two other ECE systems are routinely used on FTU: an absolutely calibrated
Michelson interferometer that measures the ECE spectrum over five harmonics with
moderate temporal resolution (5 ms) [1.40] and a 12-channel grating polychromator
with a 10-ms time resolution [1.41]. Both spectrometers measure the emission with
the line of sight in the mid-plane, normal to the magnetic field.
The radiometer has better spatial resolution (∆r≈±1 cm, ∆z≈±1.8 cm) than the
interferometer (∆r≈±2.5 cm, ∆z≈±2 cm) and polychromator. This difference is
appreciable in the presence of peaked temperature profiles, like those discussed here.
Oblique ECE measurements were performed on the current ramp phase, with central
ECH and at low or reversed magnetic shear and moderate density [1.42]. The time
evolution of the radiation temperature profile for this type of discharge, as measured
by the Michelson interferometer, is shown in figure 1.25. Although only one gyrotron
(360 kW) was available in this experiment, very peaked radiation temperature
profiles were obtained, with maximum values of the central temperature up to
11 keV.
Analysis of the Michelson spectra measured in these conditions shows that, at the
3 rd , 4 th and downshifted 2 nd harmonics, the level of emission (corresponding to a
scan of the distribution function in perpendicular energy if the energy of the
electrons responsible for emission at these harmonics is taken into account) is much
lower than expected from the high value of the central temperature. This anomaly
occurs because the bulk distribution function is distorted due to strong localisation
of ECH energy, which provokes an increase in the 2 nd harmonic emission. This
effect is in good quantitative agreement with simulations [1.37].
Fig. 1.25 - Time evolution
of the radiation temperature
profile as
measured by the
Michelson interferometer.
The ECH pulse is
applied at t=0.10 s.
Tradiation (keV)
12
# 19462
0.134 s
10
0.129 s
8
6
4
2
0.119 s
0.109 s
0.104 s
0.1 s
0
0.8 0.9 1 1.1 1.2
R(m)
A scan of the distribution function in
parallel energy by ECE measurements at
different observation angles should also
reveal the existence of a non-Maxwellian
bulk [1.37], which is the goal of the
present experiment.
Figure 1.26 shows the computed angular
dependence of the emission spectra for
conditions similar to the experimental
ones and compares the effects of
Maxwellian and non-Maxwellian bulks.
If the perpendicular emission spectrum
is selected in the frequency range near
the 2 nd harmonic and the corresponding

34
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
temperature profile is obtained by
assuming a Maxwellian distribution,
and then this temperature profile is
used to compute the expected
emission spectra for different
observation angles, these spectra
overestimate the actual radiation
temperature, if the distribution
function is not Maxwellian.
The temperature profiles measured
during an ECH pulse by the Michelson
interferometer (for φ=0°) and by the
radiometer (for φ=10°) are compared in
figure 1.27. The radiometer was
calibrated using the Michelson
temperature in the Ohmic phase of the
discharge, when the temperature
profile is rather flat and the
distribution function nearly
Maxwellian. The peak of the
temperature profile measured with the
radiometer shows the characteristic
frequency upshift due to the Doppler
effect. The radiometer peak is also
thinner than the Michelson peak, due
to better instrumental resolution. For a
more direct interpretation of the
experiment, the perpendicular
emission should also be measured
with the radiometer. Unfortunately,
these spectra are not available because
stray radiation from the gyrotron
cannot be effectively filtered for φ=0°.
Therefore, the perpendicular spectra of
the radiometer were simulated, first
deriving the temperature profile over
this frequency range from the
Michelson spectra (assuming a
Maxwellian distribution) and then
computing the corresponding spectra
of the radiometer for φ=0° and 10°.
(This is the same procedure as that
followed to obtain the results of figure
1.26, but now using the experimental
Michelson spectrum.) In the
calculations, the nominal instrumental
resolution of the two diagnostics was
0
240 260 280 300 320
Frequency (GHz)
used. Figure 1.28 summarises the result. The perpendicular spectrum of the
radiometer is both higher and thinner than the corresponding spectrum of the
interferometer, due to the better instrumental resolution of the former. The computed
oblique spectrum (assuming a Maxwellian distribution) is higher than the measured
spectrum, in qualitative agreement with the discrepancy expected when the
distribution function has a non-Maxwellian bulk (fig. 1.25). Further experimental
investigation is needed to confirm this conclusion and to test the quantitative
agreement with the theory.
Tradiation (keV)
Tradiation (keV)
Tradiation (keV)
12
10
12
10
8
6
4
2
8
6
4
2
Maxw.
FP
Fig. 1.26
Michelson
# 19462
0° 10°
20°
Radiometer
Radiom.
0.134 s
0.124 s
0.1 s
0
220 240 260 280 300 320 340
Frequency (GHz)
14
12
10
8
6
4
2
Maxw.
Radiom.
Maxw.
Michelson
0°
10°
Exp.
Michelson
0
240 260 280 300 320
Frequency (GHz)
Exp.
Radiom.
Fig. 1.26 - Angular
dependence of emission
spectra computed for a
non-Maxwellian bulk (FP)
and for the temperature
profile which, assuming a
Maxwellian distribution,
gives an identical spectrum
over this frequency
range for φ=0°. (Instrumental
parameters of the
radiometer).
Fig. 1.27 - Measured ECE
spectra from perpendicular
(φ=0°, Michelson)
and oblique ECE (φ=10°,
radiometer) during
central ECH.
Fig. 1.28 - Measured
spectra at τ=0.134 s and
simulated spectra for the
radiometer instrumental
parameters (φ=0°, 10°),
computed with the temperature
profile obtained
over this frequency range
from the interferometer
and assuming a Maxwellian
distribution. Compare
with the angular dependence
in fig. 1.26.

1. MAGNETIC CONFINEMENT
35
[1.43] M. Talvard, G.
Giruzzi and W. D. Liu,
Proc. 19th EPS Conf. on
Contr. Fusion and Plasma
Physics (Innsbruck 1992)
Vol. 16C, 1103 (1992)
[1.44] S. Preische, P. C.
Efthimion and S. M. Kaye,
Rev. Sci. Instrum. 68,
409 (1997)
1.2 FTU Facilities
Oblique ECE measurements were performed for the first time at FTU. At present, this
is the only oblique ECE diagnostic installed in a fusion device, and the potential of
such a diagnostic appears to be still largely untapped [1.43-1.44]. High-field
operations on FTU and access to enhanced confinement regimes [1.42] allow this
diagnostic to have an excellent resolution in configuration and velocity space. This
capability can be exploited to determine the bulk form of the electron distribution
and the transition from the bulk to the tail of the distribution. This would allow a
detailed study of the kinetic processes involved in heating and current drive, in a
variety of transport mechanisms for which at present direct experimental evidence is
lacking.
New data acquisition system for the plasma-density laser interferometer
The two-colour interferometer was developed to get reliable density measurements
during pellet injection experiments. To calculate the density, the signal from the CO 2
detectors, the HgCdTe room-temperature photo resistor and HeNe photodiode has
to be acquired, amplified with the automatic gain control and then compared with
the signal taken from the Bragg cell driver.
A new acquisition system based on a PC with a 5-MHz ADC was installed to increase
the sampling rate of the old system. The system also processes, stores and sends the
signal data and density to the main pulse file (based on the UNIX system) so that
they are available to users through the usual display program (SHOX) (fig. 1.29).
Fig. 1.29 - Schematic of
the new PC-based
acquisition system.
The ADC VME module of
the old system was used
to acquire the phase
DAS
comparator outputs. It did
not have enough memory
to acquire the whole
discharge at the correct
sampling rate so, during
pellet injection or fast
machine vibrations, some
fringe jumps were
observed. The problem
was solved by acquiring
CH2
data with a National
Lab View CH1
Instruments NI6110E data
acquisition card that has
high performance, reliable
data acquisition capabilities and high speed. The new program is written in LabView
graphic language.
It is possible to acquire four analog inputs at 5 MS/s, with a 12-bit resolution. Data
are stored on the local hard disk and automatically processed to separate the density
and vibration contributions to the interferometer phase. The data are then
transferred via Ethernet to the main archive under “AFS” using a set of routines
(FTUWIN DLL) for the standard formatting of FTU archive files, via the LabView
program.
Turbulence measurements with correlation reflectometry
Turbulence physics is essential for understanding transport in tokamaks. It is
important to scale turbulence characteristics with the main discharge parameters; the
extension of turbulence measurements to high toroidal magnetic fields and densities
can be done in FTU by means of the recently installed correlation reflectometer. The

36
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
turbulence characteristics
were measured in
discharges with toroidal
magnetic fields of 5.2-8 T,
average densities of 0.35-
3.5x10 20 m -3 and plasma
currents 0.4-1.2 MA.
The heterodyne reflectometry
system is similar to
that of T-10 [1.45] and
operates in the 53–78 GHz
FTU
frequency band. There are two antenna arrays (fig. 1.30): the top array consists of
four antennas adjusted to operate in the O-mode and the bottom array has two
antennas to launch an extraordinary wave. It is, therefore, possible to probe the
plasma simultaneously with the O-mode (n e between 0.35 and 0.75×10 20 m -3 ) and
the extraordinary low-frequency (Xl) mode (n e between 1.4 and 2.9×10 20 m -3 ). One
of the antennas in the top array is used to launch the probing signal and two are used
to receive the incoming signal. The two receiving antennas are separated by 2.5° to
allow poloidal correlation measurements with the O-mode in the full frequency
band. As Q-band waveguides are used for the transmission lines, it is also possible
to work with the top antenna array in the Xl-mode from 60 to 78 GHz, therefore
extending the density range of poloidal correlation measurements to the maximum
critical density of 2.9×10 20 m -3 . A special processor controls the required frequency
variation during the discharge. The minimal sweep time for the whole band is 80 ms.
The amplitude (A) and the phase (ϕ) fluctuations of the reflected electric field vector
are decomposed by a quadrature detector into imaginary (U 1 =A×sinϕ) and real
(U 2 =A×cosϕ) parts. Thus, for each reflectometry channel two signals are recorded
with a sampling rate of up to 2 MHz during the whole discharge. For poloidal
correlation measurements, the signals of two poloidally separated antennas are
synchronously sampled using four ADCs. All signal processing is performed in
complex form. For each time interval of interest, the signals are divided into
subintervals and in each subinterval a complex fast Fourier transform is applied, to
provide the amplitude and the phase spectra for both signals. The absolute values of
the amplitude of the subspectra are then averaged to obtain two final amplitude
spectra (one for each channel). A signal equal to the normalised cross product of the
two complex spectra is also built; the signal is then averaged with the same
procedure as used for the original signals. The results are the cross-phase and
coherency spectra. Auto and cross-correlation functions of the two channels can also
be calculated over each subinterval by averaging over the whole time interval.
17.5°
10.8°
5°
4"O"
mode top
antennas
array
2"X" mode
bottom
antennas
array
Fig. 1.30 - FTU
reflectometer antennas.
The bottom pair are in
X–mode; the top four, in
O-mode.
[1.45] V.A. Vershkov, V.V.
Dreval, S.V. Soldatov,
Rev. Sci. Instr. 70, 1700
(1999)
[1.46] V.A. Vershkov, et
al., presented at the 28 th
EPS Conf. on Controlled
Fusion and Plasma Physics
(Madeira 2001), Vol.
25A, p. 1276
[1.47] G.D. Conway, et al.,
Phys. Rev. Lett. 84, 1463
(2000)
Figure 1.31 shows typical results of poloidal correlation measurements of a)
amplitude, b) cross-phase, c) coherency spectra, d) auto-correlation and e) crosscorrelation
functions. The results are similar to those of T-10 [1.46] and JET [1.47]. The
quasi-coherent fluctuations always rotate in the electron diamagnetic drift direction
with velocities of about 2×10 3 m s -1 . Their angular velocity is equal to or slightly less
than that of the m=2, n=1 mode at half radius and decreases significantly towards the
plasma centre. As per expectations, turbulence frequencies are rather high in some
regimes. The broadband fluctuations show frequencies of up to 1 MHz; the quasicoherent,
250 kHz. The estimated poloidal m numbers of the quasi-coherent
fluctuations can be as large as 100. However, in many discharges lower frequencies
and m numbers are also observed, suggesting a more complex dependence of the
wavelengths on the discharge parameters, rather than a simple scaling with the
toroidal magnetic field. The cross-phase slope and the cross-correlation function
show that the low-frequency component also rotates in the electron diamagnetic drift

1. MAGNETIC CONFINEMENT
37
[1.48] F. Alladio et al.,
Pressure Anisotropy in
Ohmic FTU Discharges,
presented at the 18 th
European Conf. on Control.
Fusion and Plasma Physics,
(Berlin 1991)
[1.49] V.A. Vershkov, et
al., Nucl. Fusion, Iokohama
special issue 2, IAEA, 39,
1775 (1999)
Fig. 1.31 - Turbulence
behaviour of a typical
FTU discharge: a)
spectrum, b) cross phase,
c) coherency, d) cross
correlation, e) autocorrelation.
The spectrum
consists of three distinct
parts: low frequency,
quasi-coherent and broad
band. The crosscorrelation
function
shows a fast and a slow
rotating structure.
Fig. 1.32 - Evolution of
turbulence behaviour
during a spontaneous
density peaking event.
Coherency Cross-phase Amplitude
π
a.u.
Amplitude
kHz ω, 104 s-1 a.u. Part in signal cm 1019 m-3
1.5
1.0
0.5
0.0
1
0.6
0.4
0.2
0.0
-300 -200 -100 0 100 200
0.4
0.2
0.0
1.0
0.5
0.0
-100
6
4
2
0
10
0
0.5
0.0
0.4
0.0
0.4
0.0 2
1
0
100
0
0
-1
a)
b)
c)
d)
e)
LF
g)
d)
d)
e)
-12,84 µs
(LF)
QC
-50
300 400
Time (ms)
Frequency kHz
0 50
Time lag, µs
i
Line averaged density
Reflection radius
Broad band
Low Frequency
Quasi-coherent
Angular velocity
Quasi-coherent
frequency
f)
-3.84µs (QC)
500
a)
b)
Cross
correlation
c) c)
Auto
correlation
300
100
1.2 FTU Facilities
direction but at a
much lower speed
(about 0.5×10 3 m s -1 ).
This differs from the
T-10 results, where
this component does
not move in the
plasma core.
Distinctive
turbulence behaviour
during a spontaneous
density peaking
“event” [1.48] is
shown in figure 1.32.
A strong decrease in
the quasi-coherent
and broadband
angular velocities
and a simultaneous
increase in the quasicoherent
amplitude
are observed at t=260
ms when the density
starts to rise.
Suddenly, at t=340
ms, this process
reverses and the
rotation of the high
frequencies starts to
increase, together
with a transient
disappearance of the
quasi-coherent component, which
appears again after a few ms at a
much higher frequency. The growth
of the low-frequency amplitude
begins about 10 ms earlier, while its
velocity remains constant. Such
complex turbulence behaviour is very
similar to that observed in T-10
during SOC to IOC transition [1.49]
and may be explained by the
formation at the periphery of a
transient velocity shear zone, which
travels towards the centre. It initially
flattens the gradients in the plasma
core and the quasi-coherent velocity
therefore decreases. The shear wave
arrives at the reflecting radius at
t=340 ms, as shown by a steep
velocity rise and by a transient
suppression of the quasi-coherent
component. Note that, in reality, the
low-frequency and quasi-coherent

38
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
components may rise simultaneously
if the non-locality of the measurement
is taken into account, since it
integrates the low frequency at some
outer radial zone. It is also important
that the low-frequency rotation
remain unchanged. It is worth noting
that the quasi-coherent rotation is
sensitive to gradients, whereas the
low-frequency is closer to that of the
bulk plasma. Figure 1.33 shows the
evolution of some turbulence
parameters in an experiment with the
injection of four deuterium pellets
[1.50]. The discharge has I p =1.2 MA
and B t =7.9 T. The probing Xl mode
was used with a cut-off density of
2.76×10 20 m -3 . The trace of the
reflected wave amplitude (fig. 1.33b)
shows that reflection appears just
transiently after the first pellet at
t=800 ms and becomes permanent
after injection of the second. This is
eV/cm 104 s-1 Part in signal cm a.u. 1019 m-3
ω rotation
pellets
Electron density
Reflection radius
Broad band
Low frequency
the first time that poloidal correlation measurements in the Xl mode have been
carried out in tokamaks, with the antennas located at the low-field side and the
poloidal angle 17.5° above the equatorial plane. Such geometry results in a
significant deviation of the incident wave from the perpendicular direction, due to
the dependence of the refractive index on the magnetic field and on the Shafranov
shift of the plasma column. This makes it impossible to measure reflections close to
the centre of a plasma with a flat density profile. Thus, reliable results can be
obtained only when the amplitude of the reflected wave is larger than some
minimum value, as shown in figure 1.33b. The most distinctive feature is the drop of
the quasi-coherent rotation velocity after the second pellet. An explanation for this
could be that the reflection layer gradually moves towards the central region where
the gradients are smaller due to density decay. At the same time, the low-frequency
component rotates more slowly and does not show any strong variation in time. This
may indicate again that this component is related to plasma rotation, whereas the
quasi-coherent one depends on gradients. The simultaneous suppression of both
components in the central region, as observed in these experiments, suggests a
physical relation between these turbulence features and also the presence of a good
confinement zone near the centre.
The fully non-inductive discharges with LHCD [1.51] make it possible to reveal the
fine structure of quasi-coherent turbulence. In figure 1.34 the usually smooth spectral
maxima of these fluctuations split into an envelope of five peaks with a 3.1-kHz
spacing in frequency. The depth of amplitude modulation shows that only the quasicoherent
fluctuations are 100% modulated whereas the broadband are not. This
supports the idea that the two components have different underlying physical
mechanisms. Poloidal correlation measurements during the same time slice give the
rotation velocity and show that the mean m number is equal to 48. Thus, the 3.1-kHz
frequency step corresponds to an m number increment of three instead of one. In
order to solve this problem, consider the reflection from a plasma region with a flat
current profile around the q=3 surface but with a rather steep density gradient. The
modes with m/n ≠3 will be far away from the reflection layer and will not be “seen”
by the reflectometer. Thus, only modes with an m increment of three will be
observed, in accordance with experimental data. Analysis of the bursting of the
4
2
20
10
0
20
10
0
0.5
0.0
0.2
0.0
1.0
0.5
0.0
m=2 HF
Signal amplitude
Reliability level
LF
100
0 dT/dr
700 800 900 1000 1100 1200
Time (ms)
Fig. 1.33 - Evolution of
turbulence behaviour
during a high-confinement
phase of a pellet-fuelled
discharge.
[1.50] V. Pericoli Ridolfini
et al. , Phys. Rev. Lett. ,
82, 93 (1999)
[1.51] S. Cirant et al.,
Mode coupling trigger of
tering mides in ECV
heated discharges in FTU,
presented at the 18 th
IAEA Conference,
(Sorrento 2000), paper
IAEA-CN-77/EX3/3

1. MAGNETIC CONFINEMENT
39
Coherency Cross-phase Amplitude
π a.u.
1.0
0.5
0.0
1
0
1.0
-1
0.5
0.0
-100 -50 0 50 100
Frequency kHz
Fig. 1.34 - Turbulence spectrum in a discharge with
full LH current drive. The quasi-coherent and lowfrequency
components show a peaked structure.
Fig. 1.35 - Comparison of
MHD and turbulence
properties in a discharge
with 2/1 island ECRH
stabilisation.
Fig. 1.36 - Temperature
time trace in discharges
with BT scan used for the
diagnostics comparison.
[1.52] D. Frigione et al. ,
Nucl. Fusion 41, 11, 16131
(2001)
Frequency
m=2, kHz
Amplitude
m=2, a.u.
ωrotation
104 s-1
4
3
2
1
0
1.5
1
0.5
0
5
4
3
4
2
0
4
2
0
2
1
0
a)
b)
c)
800
T ece
a)
b)
c)
850
T ece /T sc
T sc
B tor
Time (ms)
1.2 FTU Facilities
harmonics in time shows that they behave stochastically
and are not correlated. Therefore the conclusion in this
case is that each harmonic is just an independent
excitation of the helical mode with high values of m
numbers.
Important information was obtained in experiments
with m=2, n=1 stabilisation by ECR heating [1.52].
Figure 1.35 shows a comparison of frequency (fig. 1.35a)
and amplitude (fig. 1.35b) variation of the m=2, n=1
mode measured with magnetic probes and an O-mode
reflectometer. The good agreement is clearly seen. The
angular velocities of the m=2 mode and the quasicoherent
and low-frequency components are compared
in figure 1.35c. The rotation of the quasi-coherent
component is practically equal to or sometimes
transiently higher than that of the
900
Reflectometer
Magnetic probes
ECRH
Reflectometer
Magnetic probes
MHD
LF
HF
950
0 0.5 1 1.5 2
Time (s)
m=2 mode and does not change
after its stabilisation. The lowfrequency
component rotates
slightly more slowly and does not
vary after the m=2 stabilisation
either.
Comparison of ECE and
Thomson scattering temperature
measurements
After upgrading of the Thomson
scattering (TS) system and
realignment of the ECE
optics, with a consequent
re-calibration of the
Michelson interferometer,
the electron temperature
measurements performed
with the two diagnostics
were no longer in
agreement.
To understand which of
the two measurements
was correct, some toroidal
field ramps were
analysed to see if the
discrepancy depended on
the magnetic field or on
the temperature itself. In
the first case, the error
could be attributed to the
ECE optics as the emitted spectrum moves at different frequencies when the
magnetic field is changed, if there is an error in the calibration. In the second case,
the error can only be due to the TS system since the Michelson measurements are far
from saturation. In the following it is shown that the discrepancy depends strongly
on the magnetic field and, hence, is to be attributed to the ECE diagnostics.
The result of a single toroidal field ramp is shown in figure 1.36. The ECE

40
1. MAGNETIC CONFINEMENT
1.2 FTU Facilities
temperature (T ECE ) and
the TS temperature (T SC )
2
are in close agreement at
3.4 T, but during the
ramp the discrepancy
1.5
between the two
increases. The ratio of
(T ECE) / (T SC ) vs. the
1
magnetic field was
plotted for different B T
ramps (fig. 1.37); here, 0.5
the largest discrepancy
occurs between 5 T and
6 T. This behaviour is
0
reproducible, and the
4 6 8
same results were found
B T
after a new alignment of
the ECE collecting
1.8
system. Figure 1.38
shows the two calibration
campaigns in
1.6
different colours. The
data scattering is due to
the statistical noise on
1.4
the TS, connected with
fast temperature variations
during sawtooth
1.2
activity. The discrepancy
between the two
diagnostics is clearly
1
outside the data
scattering.
0.8
To check that there was
no implicit dependence
on electron temperature,
3 4 5 6
B T
7 8 9
the ratio T ECE /T SC was plotted for fixed magnetic 8
field vs. the temperature itself for a number of shots (fig.
1.39). The plots show that the discrepancy does not
depend on the electron temperature. In fact, the data
with the highest discrepancy (corresponding to 5-6 T)
6
have an intermediate temperature (light blue points in
fig. 1.39).
After the above analysis, the ECE data were recalibrated
on the TS temperature, using the smoothed
average curve of figure 1.38, for all the shots of the
analysis campaign.
Tece/Tsc
Tece/Tsc
Tece
4
2
Fig. 1.37 - ECE–TS temperature
ratio for
discharges with contiguous
B T scans vs B T .
Fig. 1.38 - TECE/TSC
ratio vs. the toroidal
magnetic field for the
two different calibration
campaigns.
Neutron diagnostics
The detector system based on the NE213 scintillators
used for the six-channel neutron camera was recalibrated
and is now routinely operating. An analysis
program was written in IDL programming language to
provide both neutron emission and ion temperature
0
0 2 4
Fig 1.39 - ECE temperature vs. TS temperature for
the shots of fig 1.34.
T sc

1. MAGNETIC CONFINEMENT
41
1.2 FTU Facilities
profiles from the neutron-camera data. Good agreement was found between the
experimental profiles and the results of transport analysis simulations.
[1.53] F.M. Poli, et al.,
Disruption generated
runaways in the FTU high
field tokamak, presented
at 43 rd APS Conference,
(Long Beach 2001)
Disruptions in FTU plasmas were analysed using data from the neutron detectors
(BF 3 chambers). A database was prepared to investigate the dependence of the
photoneutron production on the toroidal field (B T ) and then extended to the results
obtained on other devices (TS, JT-60U) to magnetic fields B T > 4 T. Results show that
the generation of runaways is due the Dreicer mechanism and that the increase in
runaway production observed at higher toroidal fields could be due to the effect of
the plasma current profile [1.53].
The studies on runaway electrons in FTU continued in collaboration with the
Universidad Carlos III (Madrid). The time evolution of the energy distribution
function of runaway electrons in several FTU discharges was evaluated and
compared with spectral measurements of γ-rays produced by runaway electrons
hitting the limiter/vessel structures.
Tests on organic scintillator neutron detectors (NE213, stilbene and anthracene) were
carried out in collaboration with the TRINITI Institute, Moscow to verify the detector
light outputs. The light-output spectra were acquired using a 137Cs gamma source
and a VME-based acquisition system. The results indicate that stilbene scintillators
have higher light output than NE213 (respectively 82% and 51%, compared to
anthracene).
1.3.1 Introduction
1.3 Plasma Theory
The plasma theory activities are directed along two major lines of research: direct
support to the FTU research program, in terms of modelling and interpretation of
experimental data; and investigation of more fundamental physics problems
regarding turbulent transport and fast-particle-induced collective effects. In what
follows, the highlights and results of the more basic physics research efforts are
reported.
The theoretical model of ion Bernstein wave (IBW)-induced poloidal rotation was
further developed in the framework of both fluid and kinetic models and reached a
remarkable reliability level in predicting the formation and control of internal
transport barriers (ITBs) for the FTU experiments. The peculiar features of IBWs
makes it possible to achieve significant results at moderate levels of injected power,
as low as a few hundred kWs (sec. 1.3.2).
To understand turbulent transport, it is crucial to assess the role of self-generated and
time-varying sheared flows (zonal flows) in (self)-regulating the turbulence level.
Section 1.3.3 addresses this issue with particular attention paid to the role of drift-
Alfvén instabilities. The effect of partial cancellation of the Reynolds vs. Maxwell
stress tensor is discussed.
For the plasma stability studies relevant to ITB formation, new, interesting
investigation tools are available within a novel and unified mathematical
formulation for analysing both “small but finite” and “vanishing” magnetic shear
near a minimum-q surface (sec. 1.3.4).
The confinement properties of fusion products and, in general, of fast ions may
deteriorate because of the onset of collective modes of the Alfvén branch. The
stability properties of the modes in reversed shear and “advanced” tokamak
equilibria are analysed in section 1.3.5. It is shown that energetic particle driven

42
1. MAGNETIC CONFINEMENT
1.3 Plasma Theory
(EPM) “gap” modes exist as well EPM “resonant” modes, which are generally
excited at different radial locations, depending on the plasma and fast ion
equilibrium profiles.
The dynamic complexity expected from linear stability analyses is reflected in the
richness of the nonlinear behaviour of modes and energetic particle transport.
Section 1.3.6 gives some highlights from numerical simulations of nonlinear Alfvén
mode dynamics and their effect on fast ion transport. The effectiveness of the
minimum-q surface in acting as an “insulating” layer is discussed, along with the
conditions for observing “avalanches”. The simulations were performed with the
hybrid MHD gyrokinetic code (HMGC) developed in Frascati.
1.3.2 IBW-induced poloidal rotation on FTU
IBW absorption near an ion cyclotron resonant layer in a tokamak plasma can
produce ion poloidal shear flows capable of improving plasma confinement [1.54,
1.55]. This effect is provided by IBWs more efficiently than by other waves because
their rf power flux is mainly carried by the kinetic contribution of the coherent
motion of the particles in the wave field. The theoretical model of IBW-induced
poloidal rotation was developed in the framework of both fluid and kinetic analyses
[1.56, 1.57, 1.58].
Sheared flow can be obtained by solving the compressible fluid momentum balance
equation, which in the case of IBWs is a reasonable starting point:
[1.54] M. Ono & PBX-M
Group, Proc. 15 th Inter.
Conf. on Plasma Physics
and Controlled Nuclear
Fusion Research, (Seville
1994), paper IAEA-CN-
60/A-3-I-7
[1.55] R. Cesario et al.,
Phys. Plasmas, 8, 4721
(2001)
[1.56] L.A. Berry, E.F.
Jaeger, D.B. Batchelor,
Phys. Rev. Lett. 82, 1871
(1999)
[1.57] E.F Jaeger, L.A.
Berry, D.B. Batchelor,
Phys. Plasmas 7, 3319
(2000)
[1.58] J.R. Mira, D.A.
D'Ippolito, Phys. Plasmas
7, 3600 (2000)
∂ vθ
∂t
rr
+ ∇ •( vv)
= µ neo
vθ
(1)
The brackets denote average over a
magnetic surface and over a time
longer than a wave period; µ neo is
the neoclassical viscosity
coefficient, which is mainly
produced by magnetic pumping. In
(1), the perturbed velocities are
proportional to the rf electric field
via the mobility tensor. In
performing the following
calculations, it was verified that the
power dissipated by magnetic
pumping was negligible compared
to the total power absorbed by the
ions.
Figure 1.40 shows the calculated
poloidal velocity and its spatial
gradient ∂ν θ /∂r, compared with the
expected threshold for turbulence
suppression in the FTU plasma, as
obtained using (1). n ||
-launched
spectra for 0-0, (n ||peak ≥2), 0-π/2
(n ||peak ≥2), and 0-π (n ||peak ≈5,
symmetric spectrum) were taken
into account, assuming the same
launched power (0.5 MW) for all the
νθ (km/s)
8
6
4
2
0
-2
0.30
4ΩH
a) poloidal velocity
∆φ=0
∆φ = π/2
∆φ = π
0.35 0.40
x
∆φ=0
∆φ=π/2
∆φ=π
0
0.30 0.35
b) shearing rate
threshold for
turbulence
suppression
Fig. 1.40
Fig. 1.40 - a) Poloidal velocity profile ponderomotively generated by
IBWs and b) its gradient, near the resonant layer. Three waveguide
phasings are considered: 0-0 (solid line); 0-π/2 (dashed line) and 0-π
(dotted line). The solid horizontal line in b) shows the shearing rate
threshold for turbulence suppression; the vertical dotted line, the
location of the hydrogen 4 th ion cyclotron harmonic. 0.5 MW launched
power is considered for all spectra. The rf coupled power is calculated
taking into account the rf power reflection coefficients: 20%, 40%,
60% for ∆φ=π, π/2, 0.
(dνθ /dr) (MHz)
6
5
4
3
2
1
4ΩH
x
0.40

1. MAGNETIC CONFINEMENT
43
1.3 Plasma Theory
spectra considered. As shown in the figure, the threshold for turbulence suppression
is exceeded near the resonant layer within a few millimetres. The shearing rate
exceeds the threshold by about a factor of 10 for ∆φ=0, 4 for ∆φ=π/2 and 2 for ∆φ=π.
It can be verified that a linear proportionality occurs between the coupled power and
the poloidal velocity. Therefore, the estimated IBW power threshold for turbulence
suppression in FTU is 0.05 MW for the ∆φ=0 spectrum, 0.1 MW for ∆φ=π/2, and 0.25
MW for ∆φ=π.
Turbulence suppression for the IBW-FTU experiment was studied by considering a
kinetic plasma model [1.57, 1.58]. As a result, the shearing rate in FTU is ≈4 MHz
with 0.3 MW of IBW power. This should correspond to a threshold of about 0.05 MW
for turbulence suppression, assuming a linear dependence between poloidal velocity
and IBW power. Therefore, the kinetic threshold is in the range of the threshold
estimated by the fluid model. In conclusion, both kinetic and fluid approaches
indicate that the shearing rate threshold for turbulence suppression can be
sufficiently exceeded by injecting a few hundred kW of IBW power in FTU.
Such power is within the range of the available rf power in the IBW experiment in
FTU (0.5 MW with one antenna, 1.5 MW with three). Improved confinement was
found during the experiment operating at low power (one antenna and one rf
generator at 0.4 MW) [1.55]. This result is consistent with turbulence suppression
caused by IBW-induced sheared flow.
1.3.3 Generation of zonal flows by drift-Alfvén turbulence
[1.59] L. Chen, Z. Lin and
R. White, Phys. Plasmas 7,
3129, (2000)
[1.60] L. Chen, Z. Lin, R.
White and F. Zonca, Nucl.
Fusion 41, 747, (2001)
The following work (collaboration with University of California at Irvine) is an
extension and application of a recently developed theory on identifying the major
nonlinear physics processes that may regulate drift-Alfvén turbulence, using a weak
turbulence approach [1.59, 1.60]. Based on the nonlinear gyrokinetic equation for
both electrons and ions [1.59, 1.60], an analytic theory is proposed for nonlinear zonal
dynamics described in terms of two axisymmetric potentials, δφ m,z and δA ||m,z ,
which spatially depend only on a (magnetic) flux coordinate.
In the long wavelength limit (spatial scales that are greater than the collisionless skin
depth), the zonal field (current) dynamic is passive and may be consistently
neglected. The role of zonal field dynamics becomes important at short wavelengths
typical of, e.g., electron temperature gradient (ETG) driven modes.
As in the electrostatic case [1.59, 1.60], zonal flows may be spontaneously excited by
drift-Alfvén turbulence, in the form of modulational instability of the radial envelope
of the mode [1.60]. From the analytic expression for the growth rate in the long
wavelength limit, the exact cancellation of the Reynolds and Maxwell stress tensors,
and - thus - of the zonal flow, depends on the considered drift-Alfvén branch and,
more specifically, on finite plasma compression and parallel electric field effects.
Plasma compression here is anything that makes the plasma response deviate from
the pure Alfvénic state ω 2 =k ||
2 vA
2 . Finite parallel electric field, meanwhile, is due to
charge separation effects that, in a toroidal plasma, are dominated by magnetic drifts.
Specialising to the cases of kinetic Alfvén waves (KAWs) and Alfvén ion temperature
gradient (AITG) driven modes [1.60], it can be concluded that zonal flows are
spontaneously excited by such modes only under certain conditions. For KAWs, the
zonal flow growth rate, above the excitation threshold in the pump mode amplitude,
scales linearly with the wave field, i.e., as the square root of the deposited power.
Compression and finite parallel electric field effects are weakening on the
modulational instability and result in a k ⊥ ρ Li scaling in the zonal flow growth rate.
Spontaneous zonal flow excitation in the propagating region of the wave requires

44
1. MAGNETIC CONFINEMENT
1.3 Plasma Theory
T e (3/4)T i , the zonal flow spontaneous
generation takes place in the wave cut-off region, i.e., with greatly reduced
effectiveness. In the case of AITG, similar conclusions may be drawn on the basis of
the mode dispersion relation [1.61]. Compression effects for AITG result in a ≈(1-
ω ∗pi /ω) scaling in the zonal flow growth rate. Thus, spontaneous excitation of zonal
flows by AITG is possible only for ω>ω ∗pi , which is typical of moderately unstable
AITG. For strong instability, detailed analyses still need to be carried out. On the
basis of the analytic dispersion relation, however, it is possible to conclude that -
sufficiently above threshold and for sufficiently low frequency - spontaneous
excitation of zonal flows is inhibited. If confirmed, this fact will have a strong impact
on the anomalous transport associated with AITG.
1.3.4 Drift and drift-Alfvén wave structures near a minimum-q
surface
Theoretical investigations of drift and drift-Alfvén mode structures near a minimumq
surface and eigenmode analyses that assume small but finite magnetic shear can be
discussed within a unified mathematical formulation. For the sake of clarity, the
problem is studied for the case where toroidal mode coupling can be consistently
neglected. The toroidal problem can be analysed in exactly the same fashion, but
with greater technical complexity.
It is part of common wisdom to assume that, within the usual ballooning formalism
[1.62], the translational invariance of radial mode structures breaks down for different
poloidal Fourier modes when magnetic shear vanishes. This is evidently true.
However, as shown in [1.63, 1.64], only the separation of spatial scales between
equilibrium quantities and wavelengths is really needed for the analysis of high-n
mode structures. This separation of scales is still valid for high-n modes (n being the
toroidal mode number) both near and at a minimum-q surface, where magnetic shear
vanishes by definition. Only this fairly general assumption is made in the following
treatment.
The strength of the formalism employing the separation of scales is based on the fact
that the fast radial scale and the spatial coordinate along the magnetic field can be
considered Fourier conjugate variables. This is obvious from the following identities:
[1.61] F. Zonca, et al.,
Phys. Plasmas 6, 1917,
(1999)
[1.62] J.W. Connor, R.J.
Hastie, and J.B. Taylor,
Phys. Rev. Lett. 40, 396
(1978)
[1.63] F. Zonca, Continuum
damping of toroidal
Alfvén eigenmodes in
finite-beta tokamak
equilibria, Ph.D. thesis.
Princeton University,
Plasma Physics Lab.,
Princeton N.J. (1993)
[1.64] F. Zonca and L.
Chen, Phys. Fluids B5,
3668 (1993)
'
∂ '
qR0k||; mn , = nq0( r −r0)
⇒i s ; q0
≠0
,
∂κr
''
nq
S
qR0k mn qR0k mn r 0
0 r r0 2 2
∂
'
||; , = ||; , ( ) + ( − ) ⇒ ΩAm
, - ; q
n
22 0 = 0
2 2 ∂κ
r
(2)
Here, the notation r 0 denotes the radial coordinate of the considered flux surface,
where q=q 0 , q’=q 0 ’ , etc. Meanwhile, κ r =(r 0 /m)κ r , s≡r 0 q 0 ’ /q 0 , Ω A,m =nq 0 -m,
S≡√r 0 q 0 ” /q 0 , and δφ m (κr), the Fourier Transform of the fluctuating field δΨ m (r), is
given by
∞
1 ⎛ m ⎞ m
δφm( κr) = ∫ exp i ⎜ κr( r−
r0) ⎟ δψm( r) d ( r−
r0)
2π
−∞ ⎝r0
⎠ r0
(3)
Here, s and S are, respectively, the usual and generalised magnetic shear definitions.
Note that for q 0
” < 0 the definition of S would change accordingly and that (2)
assumes Ω A,m =nq 0 -m=0, for q 0 ’ ≠0, as is always possible. From (2) and (3), it is
readily shown that

1. MAGNETIC CONFINEMENT
45
1.3 Plasma Theory
[1.65] L. Chen, Phys.
Plasmas 1, 1519, (1994)
[1.66] L.J. Zheng, L. Chen,
and R. A. Santoro, Phys.
Plasmas 7, 2469, (2000)
[1.67] F. Zonca and L.
Chen, Phys. Plasmas 7,
4600, (2000)
[1.68] H.L. Berk, et al.,
Alfvén cascades in JET
discharges with nonmonotonic
q-profile,
Inter. Fusion Theory
Conference, (Santa Fe
2001), paper 1C54
where Θ m =s 2 and Ω A,m =0 for s≠0, and Θ m =S 2 Ω A,m /n for s=0.
With this formalism, e.g., the vorticity equation for EPM resonant excitation by ion
cyclotron resonance frequency (ICRF) becomes
⎡ 2 2 2
⎤
⎢ ∂ Ω − ΩAm
, 1 Λm
Θ
+
− + m⎥
⎛
⎢ 2
2 2 2 ⎥ ⎜
⎣∂κ Θm
( 1+
κr
) ( 1+
κ ⎝
r ) ⎦
r
2
2 2 ∂
qR0k||; mn , Am , - m 2
∂κ
r
( ) = Ω Θ
Here, Ω≡ω/ω A , ω A =νA/qR 0 , and Λ m describes the fast ion response. The structure
of (3) is very general. The factor Θ m reflects the typical radial width of the mode
structure, ∆r. In fact,
2 2 2
m Ω − ΩAm
,
∆r
≈
r0
Θm
2 ⎞
1+
κr
δφm⎟
=0
⎠
(4)
(5)
(6)
It is then obvious that the largest value of Θ m between s 2 and S 2 Ω A,m /n determines
the radial mode width and the actual form of the mode structure and dispersion
relation. Since (∆ r /r 0 ) 2 ∝|1/m 2 Θ m | may be estimated, the transition from small but
finite shear to zero shear occurs for s 2

46
1. MAGNETIC CONFINEMENT
1.3 Plasma Theory
present parameters, i.e., at r/a=0.2, 2
0.6
s≡rq’/q=-0.58, q=4.4, α H ≡-R 0 q 2 (dβ/dr), 1.5
0.4
∆’=0.125, α H =0.515, β H =0.0072, 1
η H =0.395, ν H /ν A =0.43, ρ LH /a=0.019.
0.5
0.2
The fast ion tail distribution function is
0
Maxwelllian in energy, with a pitch
0
angle distribution highly peaked around
-0.5
-0.2
µB 0 /E=1, µ being the magnetic moment. -1
Results for the growth rate and the -1.5 -0.4
0 0.2 0.4 0.6 0.8 1
mode frequency of the EPM are shown
r/a
in figure 1.43. It is evident that the range
of unstable mode numbers corresponds well to the experimentally observed modes.
The reason why the mode can be considered a resonantly excited EPM and not a
toroidal Alfvén eigenmode (TAE) is given by the strength of the growth rate, which
is comparable with the gap width. A more articulated explanation of this
interpretation is provided in [1.66].
Consider modes localised near
0 1 2 3 4
r 0 (for the present parameters 0.05
0.6
r 0 /a=0.49), where q has a
minimum given by q 0 .
0.04
Consider also a given toroidal
mode number n and a poloidal
0.4
mode number m such that the 0.03
normalised parallel wave
vectors Ω A,m =nq 0 -m0. It is then
0.2
readily demonstrated that the
condition under which
0.01
continuum damping is
minimised is that with 0.00
0.0
–1/2

1. MAGNETIC CONFINEMENT
47
1.3 Plasma Theory
1
0.8
0.6
0.4
0.2
(m-1,n)
non-local
continuum damping
(m,n)
0
-2 -1.5 -1 -0.5 0 0.5 1 1.5
x
Fig. 1.44 - Radial structure Fig 1.44 of the shear
Alfvén continuous spectrum for the (m,
n) and (m-1, n) modes in the case
1≥Ω A,m +Ω A,m-1 >>-r 0 /R 0 . The value of
q 2 0 R 2 0 k 2
|| is shown vs. S≡√n ” 0 (r-r 0 ). The
frequencies of the (m, n) and (m-1, n)
modes are also shown as they are
expected from (8).
1
0.8
0.6
non-local
continuum damping
2
for ω d >ω B >>ω. Here, λ H =(k ⊥ ν ⊥ )/ω cH , ΘF 0H =(2ω∂/∂ν 2 +k×b•∇
/ω cH )F 0H , F 0H is the fast hydrogen tail distribution function, and
assuming deeply trapped banana orbits [1.70], for which the mode
drive is expected to be the strongest. For Ω A,m +Ω A,m-1 >>r 0 /R 0 ,
toroidal coupling between the (m,n) and (m-1,n) modes can be
neglected (see fig. 1.44). The two modes then satisfy the following
approximate dispersion relations, derived from a variational
principle [1.67]:
Sπ
⎛
2n
Λ
⎞
ΩAm
Ω
m
, + = −
/ / ⎜
1
52 12 2
n ⎝S
Ω ⎟
2
Am , ⎠
Sπ
⎛
2n
Λ
⎞
ΩAm
Ω
m
, − = −1
−1 −
/ / ⎜
1
52 12 2
n ⎝S
Ω ⎟
2
Am , −1
⎠
where Λ m-1 ≅>ω; thus, the fast ions are characterised by
negative compressibility, which causes the mode frequency to be
shifted upward, contrary to the general case for which fast ion
compression shifts the mode frequency downward [1.65-1.67]. For
this reason, only the (m,n) mode can exist just above the local
maximum of the Alfvén continuum at r 0 , but not the (m-1,n) mode
just below the local minimum of the continuum. The condition for
the existence of the (m,n)mode is Λ m /Ω A,m >S 2 /2n, which, as a
consequence of (8) is independent of n [1.68] and of the mode
frequency. This condition is a lower bound on the fast hydrogen tail
particle density, and for the present parameters
[S=1.54,–n H /(R 0 ∂ r n H )=0.13] it gives n H /n e >3.1%, consistent with
the experimental values (n H /n e =4%). Changing the fast-ion nonresonant
response, the (m-1,n) mode instead of the (m,n) can be
excited [1.70].
(9)
0.4
0.2
0
-2 -1.5 -1 -0.5 0 0.5 1 1.5
x
Fig. 1.45 - Same as fig. 1.44 but for
Fig 1.45
–r 0 /R 0 ≤Ω A,m +Ω A,m-1 ≤r 0 /R 0 .
[1.70] F. Zonca, et al.,
Energetic particle mode
stability in tokamaks with
hollow q-profiles, to be
published on Phys. of
Plasmas
2
Since Ω A,m +ΩA ,m-1 →0 + (which may occur when, as in the
experiment, q 0 drops), as in figure 1.45, toroidal coupling effects
become important. Two branches are still present as in (9), one of
which strongly continuum damped (odd mode [1.67]), and the
other (even mode [1.67]) satisfying the modified dispersion
relation:
which can be straightforwardly derived from within the theoretical
approach of [1.67] in a simple but still relevant limiting case, and where
ε 0 ≡2(r 0 /R 0 +∆’),Ω A,m ≅-1/2. Note that on the left-hand side of (10), the fourth root of
the quantity in parentheses – and not the square root as in the usual TAE case – is
due to the local minimum in the q-profile [1.67]. The existence condition for the
nearly undamped mode of (10) is exactly the same as that discussed for (9). Clearly,
in both cases the exponentially small continuum damping, due to the non-local
interaction with the (m,n)mode continuum, and other kinetic damping mechanisms
must be evaluated and compared with the resonant drive associated with fast ions
[1.67] before the existence of these modes is demonstrated on a rigorous basis. The
complete expressions of non-local continuum damping and fast ion resonant and non-
⎡
π
ε0
2 4 2 14
⎛
2 Λ
Ω −⎜
⎞⎤
/
S
⎛
n
⎞
Ω −14
/ ⎟ = m −1
⎣⎢ ⎝ ⎠⎦⎥ 32 / 12 / ⎜ 2
2 ⎝ Ω ⎟
n S Am , ⎠
(10)

48
1. MAGNETIC CONFINEMENT
1.3 Plasma Theory
resonant responses are derived and analysed in [1.70]. The same work
gives the complete toroidal dispersion relation, which generalises
(10).
When Ω A,m +Ω A,m-1

1. MAGNETIC CONFINEMENT
49
1.3 Plasma Theory
ω/ω A0
τ= 132.00
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/a
τ
240
r n
H
ω/ω A0
τ= 120.00
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/a
τ
192
r n
H
ω/ω A0
τ= 204.00
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/a
τ
240
r n
H
205
165
205
170
138
170
135
111
135
100
84
100
65
57
65
30
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/a
30
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/a
30
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/a
Fig. 1.47
Fig. 1.47 - (top) Power spectra of scalar potential fluctuations during the nonlinear saturated phase in the
plane (r/a, ω/ω A,0 ) with the Alfvén continuum structure superimposed (black curves) and (bottom) contour
plots of the energetic-ion line density r^n^H ≡(r/a)(ν H (r)/ν H,0 ) in the (r/a,τ) plane, with τ≡ω A,0 τ and ω A,0 the
Alfvén frequency at the plasma centre. Three cases are shown: (left) deeply hollow q profile and flat thermalplasma
density; (centre) deeply hollow q profile and decreasing thermal-plasma density and (right) moderately
hollow q profile and flat thermal-plasma density.
(fig. 1.47 bottom centre). A much more dramatic effect is obtained by acting on the q
profile. A case characterised by a moderately hollow q profile (q 0 ≈5, q min ≈3.6, q a ≈5)
and a flat thermal-plasma density profile (β H,0 =2.5%) clearly exhibits strong
degradation of energetic particle confinement. The reason for this is that, for a given
value of β ’ H , the mode radial width scales, near the q min surface, as 1/√nq", while
the typical orbit size is proportional to q min . Decreasing the hollowness of the q
profile, while taking q 0 and q a as fixed, yields lower q” and larger q min values and
makes both the mode and the orbit widths larger than in the deeply hollow q-profile
case. Moreover, the energetic-ion drive intensity (∝q 2 β H ’) scales as q 2 min . The mode
is then a more efficient scattering source for energetic-ion orbits. This fact and the fair
alignment of the frequency gap at different radial positions make the displaced
energetic ion source effective in destabilising an avalanche of outer poloidal
harmonics (fig. 1.47 right top and bottom).

50
1. MAGNETIC CONFINEMENT
1.4 FT3 Conceptual Study
1.4.1 Introduction
Table 1.III - FT3 parameters
The FT3 concept is a proposed
upgrade of FTU, which would
enable studies of sub-ignited
plasma conditions in deuterium
plasmas (Q equiv ≈1-5) with
particular reference to the
collective effects driven by the
fast ions produced by ICRH.
Therefore, FT3 would prepare
the operational scenarios of a
burning plasma experiment by investigating the approach to ignition in the presence
of the relevant dynamics of fast ions.
FT3 is similar to JET from the point of view of dimensionless parameters, but the
expected fusion performances are much higher because of the
higher magnetic field B (the triple product nTτ is proportional to
B at fixed dimensionless parameters). Indeed, the expected fastparticle
parameters in FT3 at maximum performance are closer to
those of a burning plasma experiment than the parameters
achievable on JET at maximum performance. Note also that FT3
has greater shaping flexibility at maximum plasma current than
JET.
The large range of magnetic field values achievable in FT3
includes the ITER magnetic field value, so the proposed device
would be a natural test bed for the development of ITER
diagnostics and of auxiliary heating systems such as ECRH.
Table 1.III reports the main engineering parameters.
B(T)/I(MA) 8/6
P aux (MW) (ICRH/ECRH/LH) 25 (20/3.2/6)
R(m) 1.3
a(m)/b(m) 0.48/0.9
κ/δ≅I=6MA 1.8/0.6
t flat-top (s) ≅8T 4
Three main operational scenarios are envisaged: single X-point
(fig. 1.48) at 8 T/6 MA for investigating H-mode and ITB
formation at high magnetic-field and density; limiter scenario at 8
T/7 MA to study enhanced L-mode regimes; single X–point at 5T/2.4 MA for longpulse
scenarios and advanced tokamak physics. H-mode plasmas are expected to
achieve an equivalent Q between Q = 1 and Q = 2, whereas the formation of an ITB
could allow an equivalent Q in the range Q=5.
Fig. 1.48 - FT3 single-null
equilibrium at B=8 T and
I= 6MA.
1.4.2 Main objectives of the FT3 scientific programme
• Investigation of fast-ion collective effects in the parameter range relevant for
burning plasmas. The fast-particle concentration achievable with 20-MW ICRH is
sufficient for studying the destabilisation of resonant collective modes, such as
fishbones and energetic particle modes (EPMs), which are in principle the most
dangerous fast-particle collective effects. Investigation of these effects in negative
magnetic shear discharges at B = 5 T will allow a better understanding of the role of
these instabilities in advanced scenarios. Note that these regimes are obtained on FT3
at a slowing down time/energy confinement time ratio comparable to that of a
burning plasma experiment.
• Test of H-mode threshold at high magnetic field. FT3 could prove the validity of
the most recent scaling law for the L-H threshold (fig. 1.49), which predicts a lower
threshold power on ITER than the IPB98 scaling. This would facilitate making a final
decision on the auxiliary heating systems of ITER. Note that JET data are consistent
with both expressions of the L-H threshold and cannot provide a definitive answer.

800
1. MAGNETIC CONFINEMENT
51
Fig. 1.49 - H-mode
threshold. Experimental
points compared with the
empirical scaling law. The
FT3 H-mode threshold
has a factor of two range
of variation. An
assessment of the power
threshold could allow an
earlier decision on the
auxiliary heating systems
of ITER.
10
1
0.1
C-Mod a
ASDEX
AUG
COMPASS
DIII-D
JET
JFT-2M
JJT-60U
PBXM
TCV
W=1
0.1
RMSE (%) = 27.8
1.0
0.054n s
0.40B T
0.05S0.84
Fig. 1.50 - FT3 load
assembly.
1.4 FT3 Conceptual Study
• Test of H-mode operation in near burning-plasma-experiment conditions.
H–mode data from high-field tokamaks and ICR-heated devices show very low edge
activity [e.g. the enhanced D α (EDA) mode in C-MOD]. On the contrary, H-mode
operation in neutral-beam-heated devices (the main source of heating in most
tokamaks) exhibits strong Type-I edge-localised mode activity that is not compatible
with divertor operation on ITER. FT3 could allow complete characterisation of these
regimes.
• Investigation of enhanced L-mode regimes. Recent FTU results show that quasistationary
pellet enhanced performance (PEP) modes can be achieved with
significant confinement improvement. The extrapolability of this regime to burning
plasma experiments requires the development of deep fuelling techniques, which
can be tested on FT3. The formation of radiative improved (RI) modes in highdensity
plasmas has not yet been demonstrated. Since the
radiated fraction increases with density, the achievement
of RI modes requires that a sufficient edge dilution be
reached before radiative collapse takes place. Both the PEP
and the RI mode are interesting for a burning plasma
experiment.
• Achievement of ITB at high magnetic field. Steady-state
conditions on the plasma current redistribution time scale
can be achieved on FT3 at B=5 T in less than 5 s. Fully noninductive
operation can be obtained at plasma currents of
1.6 MA with a bootstrap fraction of the order of 70%. Thus,
advanced tokamak scenarios could be investigated on FT3
at the ITER density and magnetic field values.
• Investigation of scrape-off-layer physics. Plasma
detachment from the divertor plates is expected at
10.0
densities well below the Greenwald limit in compact, high
magnetic field experiments. These conditions could be
studied in FT3 at values of the divertor similarity
parameter P/R of the order of those of JET and ASDEX-U,
with P being the power flowing in the scrape-off layer and R the major radius. An
open divertor configuration is foreseen for FT3 to be able to comply with the highly
localised heat fluxes expected in X-point plasmas without reducing the flexibility of
the equilibrium configuration.
Vacuum
Chamber
1400
FW
ICRH
antennas
Support
legs
5800
P10
Technical
structure
P11
P14
5000
- FTIII
C.R.E. Frascati ERG-FUS-TN-MC
Toroidal
Field Coils
Central Solenoid
External Poloidal
Field Coils
Cryostat
Meridian
cross-section
02/12/2000
FT3 is a cryogenic device, like
FTU. However, cooling is by He
gas at 30K, thus allowing a
substantial reduction in dissipated
power. The magnet is fed by the
grid (available power about 45
MW) and by the existing motor
flywheel generator (MFG1). The
toroidal magnet design is based on
the experience gained from the
previous FT and FTU devices and
from the IGNITOR project. The
poloidal field system should give
sufficient flexibility to the
magnetic configuration. The
central solenoid has been designed
so as to allow segmentation. The
FT3 load assembly is shown in
figure 1.50; the divertor

52
1. MAGNETIC CONFINEMENT
1.4 FT3 Conceptual Study
configuration, in figure 1.51. The vacuum
vessel is stainless steel to reduce activation.
Indeed, FT3 is expected to produce more than
10 17 neutrons per shot at maximum
performance.
The proposed upgrade can make full use of all
Target Plate
the buildings, power supply and auxiliary
heating systems as well as the diagnostics of
FTU. FT3 will be installed in the FTU hall. A
limited upgrade of the existing MFG3 from 200
to 330 MJ deliverable energy is necessary. The
LH heating and the CD system (8 GHz, 6 MW)
can be adapted to the FT3 port. The driven
current is in the range 0.7-1 MA at n=10 20 m -3 .
A limited upgrade of the ECH system (140 GHz
3.2 MW) is envisaged to allow a longer pulse length at a power capable of stabilising
neoclassical tearing modes. All the FTU diagnostics can be adapted, with minor
modifications, for use on FT3. New diagnostics have to be built for the X-point
measurements and the plasma current profile measurement by the motional Stark
effect.
Support Structure
Vacuum Vessel
Fig. 1.51 - FT3 open
divertor configuration.
The most important upgrade in additional heating capability is the ICRH system
(tunable in the range 70-90 MHz, 20 MW at the plasma), which will provide the fastparticle
component in the H minority scheme at B=5 T and in the He 3 minority
scheme at B=8 T.
The FT3 construction and assembly is expected to last five years. The total cost is
estimated to be about 100 MEURO.
1.5 PROTO-SPHERA
1.5.1 Introduction
Chandrasekhar-Kendall-Furth (CKF) magnetic configurations are a
novel approach to magnetic confinement fusion research. They are
simply connected axisymmetric plasma equilibria containing a
magnetic separatrix with ordinary X–points (B≠0). The magnetic
separatrix divides a main spherical torus, two secondary tori on the
top and bottom of the main torus and a spheromak discharge
surrounding the three tori. Two degenerate magnetic X-points (B=0)
are present on the symmetry axis at the edge of the configurations
(fig. 1.52).
Whereas CKF force-free fields cannot sustain any pressure gradient
(∇ → ∧Β → →
=0) and have a relaxation parameter µ=µ 0 j •Β → /B 2 constant
all over the plasma, unrelaxed ((∇ → ∧Β → ≠0, (∇ → ∧Β → ≠0) CKF equilibria
can be calculated with the boundary condition that µ=µ → 0 j •Β → /B 2
is constant only at the edge of the plasma. The surface-averaged
value =µ 0 < → j •Β → /B 2 > will decrease from the edge of the
surrounding spheromak to the axis of the main spherical torus.
ST
I ST
CKF < β > ST = 102%
S P
I e
Spherical
Torus
Secondary
Torus
Surrounding
Discharge
If the spheromak discharge is sustained by driving current along its
closed flux surfaces, magnetic helicity, flowing down the gradient, will be
injected into the main spherical torus through magnetic reconnections at the X-
points. The gradient of the pressure profile will presumably be concentrated in the
Fig. 1.52 - Unrelaxed CKF
configuration.

1. MAGNETIC CONFINEMENT
53
1.5 PROTO-SPHERA
b)
1.6
a)
p (a.u.)
µ (m-1)
1
0
20
0
3
c)
d)
IST/Ie
10
5
0
P
UNSTABLE
ST UNSTABLE
STABLE
1 1.5 2 2.5 3 qst
0
Fig. 1.54 - Ideal MHD stability plot for β=1
unrelaxed CKF configuration, expressed in terms
of the safety factor on the spherical torus
magnetic axis q 0
ST and of the ratio of currents
I ST /I e .
1.8
2.0
3.0
4.0
qst
95
q
0
0.3
R(m)
0
0
Fig. 1.53 - Unrelaxed CKF configuration, with
I ST /I e =3. a) Flux coordinates and profiles on
equatorial plane of b) pressure, c) and d) safety
factor q.
0.2
R(m)
same region where the gradient of has the largest
variation (see fig. 1.53).
Unrelaxed CKF equilibria with this kind of and
pressure profiles are stable to all ideal MHD
perturbations with low toroidal mode number (n=1,
2, 3), up to unity plasma beta values,
β=2µ 0 Vol / Vol ≈1 (fig. 1.54).
Unrelaxed CKF fusion reactors with the right helicity injection, β limit and energy
confinement will allow an unimpeded outflow of a part of the high-energy charged
fusion products. The charged fusion products will drift across the magnetic
separatrix to the degenerate magnetic X-points (B=0) on the top/bottom of the
configuration, easing direct energy conversion and the use of a burner as a space
thruster.
The high plasma β≈1 opens the possibility that plasma motions, i.e., radial electric
fields, can sustain the magnetic field of CKF configurations. In the case of a CKF
fusion reactor, the radial electric field can even be the natural result of losses of
charged fusion products. To begin an experimental study of unrelaxed CKF
configurations, a preliminary experiment is being proposed. The PROTO-SPHERA
experiment will study the properties of a CKF configuration where a hydrogen forcefree
screw pinch, fed by electrodes, replaces in part the surrounding spheromak
discharge, while poloidal field coils replace the secondary tori. PROTO-SPHERA,
with a longitudinal pinch current I e =60 kA, will produce a spherical torus of
diameter 2×R sph =70 cm, aspect ratio A=1.2-1.3 and toroidal current I ST =120-240 kA.
1.5.2 Mechanical engineering
PROTO-SHERA was designed in detail to define the load assembly (fig. 1.55). Table
1.IV gives the main engineering parameters of the machine. The plasma pulse
duration of 1 s and the inter-pulse time of 5 min are key data. The machine is
designed to operate at room temperature with a vacuum of ~1×10 -8 mbar. It can be
baked up to ~90°C.
Figure 1.55 shows the key components of the machine: electrodes (anode, cathode),
coils, support structure and divertor plates, together with the protection plates that
shield the coils from the hot electrodes.
The vacuum vessel is 2 m in diameter and 2.5 m high, with a thickness of 18 mm. It

54
1. MAGNETIC CONFINEMENT
1.5 PROTO-SPHERA
fl 2000
Insulation
Anode
Support
Structure
Type
coils B
Divertor
plates
Type
coils A
Fig. 1.55 - PROTO-
SPHERA load assembly.
Protection
plates
Cathode
Table 1.IV - Machine parameters
Spherical torus diameter
0.7 m
Longitudinal screw pinch current
60 kA
Toroidal spherical torus current
120-240 kA
Plasma pulse duration
1 s
Minimum time between two pulses (duty cycle)
5 min.
Maximum heat loads in divertor first-wall components ~2 MW/m 2
Maximum heat loads on rest of first wall
3 MW/m 2 , for 1 ms
Maximum current density on plasma-electrode interface 1 MA/m 2
has a flat 30-mm top, and bottom flanges for all services, while the diagnostic and
vacuum ports are in the main body.
The poloidal field coils are water-cooled vacuum-impregnated OFHC Cu to
accommodate the 5-min duty cycle of the machine. Type A coils (fig. 1.55) have
currents varying with time during plasma evolution and thin (1.5-mm Inconel)
casings; type B coils have constant currents in thick (10-mm stainless steel) casings.
The anode of the machine is made mainly from Cu, with replaceable WCu modules
to allow for the hotspots. The anode consists of six sectors with five modules per
sector. There are also 600 holes (10-mm-diam.) in total to accommodate the gas flow

1. MAGNETIC CONFINEMENT
55
1.5 PROTO SPHERA
required. The total energy deposited per shot is estimated to be
4 MJ. Although local hot spots around the anode holes of
~1000°C are expected, the main body of the anode reaches
much lower temperatures.
Figure 1.56 shows the cathode with the W spiral, supported via
Mo dispensers on the Cu cathode body. The cathode consists of
six sectors, with 24 dispensers per sector. Each dispenser holds
three spirals. The maximum temperature in a spiral is expected
to be ~2750°C, while the cathode main body reaches
temperatures much lower than 1000°C. The total energy
deposited in this component is ~ 8 MJ per shot.
Fig. 1.56 - View of cathode.
PROTO-SPHERA allows the 12 MJ deposited in the electrodes
in each shot to be safely accommodated within the 5-min
machine duty cycle.
Special attention was paid to the protection components (fig. 1.56), which are water
or inertially cooled to shield the coil insulation from the heat flux coming from
cathode.
The electromagnetic forces and stresses during plasma evolution and disruptions
were analysed. The Cu protection plates and the stainless-steel divertor plates were
cut to limit the EM loads.
The assembly has been designed to facilitate access and maximise the space for
diagnostics.

2. IGNITOR PROGRAM 59
2.1 Introduction
Pre-eminent among the significant events affecting the IGNITOR Project in 2001
were the conclusions of the scientific debate initiated by the US fusion community
when the U.S. withdrew from the ITER Project. These conclusions were summarised
in the final report presented by the Fusion Energy Science Advisory Committee
(FESAC) of the Department of Energy (DOE) [2.1]. The objective of the report was to
provide the basis for proposing a programme for the next burning plasma
experiment in the United States. The FESAC report clearly states that the only
magnetic configuration sufficiently developed at this time to serve as a burning
plasma experiment is the tokamak, and that all three burning plasma experimental
designs today under development worldwide (IGNITOR, ITER-FEAT and FIRE)
would deliver a large and significant advance in the understanding of burning plasma. This
is an outstanding acknowledgement, at high international level, of the relevance of
IGNITOR.
[2.1] FESAC Panel Report
on Burning Plasma Physics,
Sept. 2001
[2.2] G. Cenacchi, et al.,
Bull. Am. Phys. Soc. 46,
272 (2001)
[2.3] A. Airoldi,et al., Bull.
Am. Phys. Soc. 46, 271
(2001)
[2.4] G. Cenacchi, A.
Airoldi: Equilibrium configurations
for the
Ignitor experiment, IFP
report FP 01/1 (February
2001)
At ENEA, the design of the IGNITOR machine is sufficiently detailed to start the
procurement of systems and components, once the Italian Government assures the funds
for its construction and ENEA starts the licensing procedure. Meanwhile, the design work
on this very compact, heavily loaded machine is being constantly updated to incorporate
the latest progress in theoretical knowledge and experimental results in operating
tokamaks. Technical specifications were issued to define the activity to be carried out with
the support of industry and to update and revise the design of the main systems and
components, and the relative contract is presently under negotiation with Ansaldo. In
2001, the design activities concerned studies on the machine flexibility and advanced
scenarios, the effect of new disruption data, the development of simplified engineering
models for stress analysis of the nuclear core, plasma-wall interaction and impurity
production studies and the design of the auxiliary ion cyclotron resonance heating (ICRH)
system.
2.2.1 Advanced scenarios
2.2 Physics
The poloidal field system is formed of 13+13 coils, symmetrically located relative to
the machine equatorial plane and independently powered. The system is very
flexible and allows X-point configurations as well as the limiter configurations of the
reference scenario. Preliminary analyses, carried out with the equilibrium-transport
code JETTO, concerned the possibility of obtaining high confinement conditions (the
so-called “H-mode”) in the presence of double-null configurations around 10 MA
and with auxiliary heating. The threshold power required for the L- to H-mode
transition, evaluated according to the ITER scaling, is within the limits of the
auxiliary heating system already included in the machine design [2.2]. The flat-top
phase of the nominal IGNITOR scenario was analysed for situations in which a high
impurity content delays ignition and leads to the development of sawtooth-type
MHD instabilities [2.3].
The MHD equilibrium configurations supporting the 11-MA 13-T IGNITOR scenario
were carefully revised. The poloidal field coil currents required throughout the
plasma evolution, from start-up to ignition, were determined both for the reference
conditions and for the advanced scenario with the double null 10-MA configuration.
Equilibrium configurations were obtained for both cases [2.4].
2.3 Engineering of the Machine
2.3.1 EM analysis of vacuum vessel during plasma disruptions
The global forces induced on the IGNITOR vacuum vessel during plasma
disruptions were estimated more precisely on the basis of the JET and Alcator C-Mod

60
2. IGNITOR PROGRAM
2.3 Engineering of the Machine
disruption database. A disruption
simulation was performed, taking into
account the most dangerous EM
transient that could be predicted for the
machine. Figure 2.1 shows the
behaviour vs. time of the macroscopic
plasma-parameters, as well as the
resultant vertical force on the vessel
and the separate contributions to the
total force, due to halo and eddy
currents. The input excitation from this
simulation will be used for the 3–D EM
analysis of the plasma chamber.
I (MA),R(dm),q95
15
10
5
R centre
I plasma
q 95
Z centre
F z (eddy)
Fz(halo)
F z (eddy+halo)
0
-15
0 5 10 15 20 25 30 35
Time (ms) q 95 =2.0
0
-5
-10
Fig. 2.1 - Behaviour vs.
time of the macroscopic
plasma-parameters.
2.3.2 Engineering models
A flexible 2/3-D code for the structural analysis of the nuclear core (vacuum vessel,
toroidal and poloidal field coils, structural supports and bracing rings) is being
developed at ENEA. The model allows easy parametric analysis of different
machine operating conditions, optimisation of tolerances, and provides a helpful
instrument for the solution of possible non-conformities arising during component
manufacturing.
The first part, i.e., the models for the magnetic fields, magnetic forces and
temperature calculations of the IGNITOR poloidal coils, was completed. The EM
model computes forces and temperatures during normal operating conditions as
well as during dynamic events, such as disruptions and faults. The thermal model
includes the magneto-resistance effects, real geometry and the insulation of each
turn in each coil. As an example, the modelled components (plasma current
threads, plasma chamber, poloidal and press coils) are shown in figure 2.2; and the
results of a thermal calculation of the inner poloidal coils, at the end of the 12-MA
13-T scenario, in figure 2.3. No temperature produces any internal stress in the coils
that exceeds the allowable.
Fig. 2.2 - FEM models of plasma current threads, plasma chamber, poloidal and
press coils.
Fig. 2.3 - Temperatures in
central solenoid at the
end of the 12-MA 13-T
scenario.

2. IGNITOR PROGRAM 61
2.3 Engineering of the Machine
2.3.3 Plasma-wall interaction and molybdenum contamination
[2.5] C. Ferro, et al., Proc.
28 th EPS Conf. on
Controlled Fusion and
Plasma Physics, (Madeira
2001), Vol 25A, p. 2121
An investigation [2.5] was carried out by ENEA in cooperation with Plasma Surface
Engineering Inc., Canada to study plasma interaction with the wall and the
molybdenum material released into the plasma. At high plasma density, i.e., at high
edge collisionality and with a scrape-off layer that is nearly opaque for neutrals
(IGNITOR operational conditions), the limiter operation develops poloidal gradients
in the plasma temperature and density. Such behaviour, which allows a reduction in
the plasma temperature immediately adjacent to surfaces, is usually assumed to
occur only in the divertor devices. Since sputtering yields are strongly dependent on
particle energies, a very large reduction in sputtering rates can be obtained from a
modest reduction in plasma temperature. The Edge of IGNITOR (EDI) code (a 2-D
Monte Carlo code that models the hydrogenic neutral transport of IGNITOR)
confirmed the above characteristics for the IGNITOR boundary plasma. The code
has also shown that, even with 30 MW of power entering the boundary, the electron
temperature close to the limiter surface can be reduced and sputtering suppressed.
In addition, the large-area limiter is very effective in keeping the peak power density
at low levels, even in discharges with a low level of radiation. At moderate and high
density, the EDI code predicts the formation of an inner-wall MARFE, which will
spread the power, via radiation, to an even greater extent. This result is very similar
to experimental observations in other tokamaks, particularly Alcator C-Mod.
2.3.4 Auxiliary plasma heating system: ICRH
The characteristics, operational parameters and components of an auxiliary heating
system were defined. The system, equipped with 6 antennas, is capable of delivering
up to 18 MW of power, with an Ohmic efficiency of about 90%; the power levels
actually employed will depend on the operational scenario.
Antenna geometry. The antenna, housed in a port (“launcher”, fig. 2.4), consists of two
pairs of poloidally-directed straps. Each strap is fed by a coaxial line at one end and
short-circuited to the vacuum vessel at the other. The basic operation requires phase
reversal of the feed currents in each poloidal pair, to drive an equi-directed poloidal
current along the two straps. To achieve higher efficiency and to increase power
handling capabilities, two strap pairs are aligned toroidally in each port, with
variable relative phase. As the coaxial lines occupy the central part of the antenna,
slanted feeders are used to connect the poloidal sections of the straps to the feeding
points. The Faraday shield is made of a metal frame on which a single layer of metal
rods aligned with the static magnetic field lines is attached.
Feeding system. Each strap is independently fed and energized by a generator with a
rf driver and a power amplifier. Each feeding coaxial line is
matched to the power amplifier via a tuning and matching
system composed of two stubs and a line stretcher (“trombone”);
an additional decoupling system is planned between adjacent
straps. The rf drivers of the amplifiers are properly phased to
achieve the required phasing by means of a phase-locking
system. This system is also compatible with automatic matching
to the plasma, which may include variations in frequency and in
the stub and line-stretcher lengths.
Fig. 2.4 - ICRH antenna.
Power handling. The strap design shows a low fraction of reactive
electric energy storage in the operating frequency range.
Considering a conservative estimate of 35 kV for an overall
breakdown voltage in the system, each strap feeding a
subsystem can withstand a power of 1 MW in the entire
frequency range.

3. FUSION TECHNOLOGY 65
3.1.1 Introduction
3.1 Technology Programme
In 2001, ENEA contributed to the Next-Step and Long-Term Programmes, the Power
Plant Conceptual Studies and Underlying Technology, in the framework of the
European Fusion Technology Agreement (EFDA).
The activities covered almost all the R&D fields: vesssel in-vessel (blanket, first wall
and divertor, remote handling, fuel cycle); magnets (conductor development, coil
tests); safety (including site and socio-economic studies); physics integration
(neutron diagnostics); long-term activities (breeding materials; structural materials,
liquid metal technology, helium-cooled components, IFMIF).
Further to the fusion activities, some valuable applications were developed in the
field of nuclear detectors. Experimental campaigns were carried out on new
hydrogen energy and plasma-focus studies.
The technology activities were performed at the Frascati and Brasimone laboratories,
with valuable contributions from other ENEA laboratories. Significant industrial
collaborations were also established.
It is worth mentioning that the three patents granted in 2001 resulted from the R&D
activities.
3.2 First Wall and Divertor
3.2.1 Influence of manufacturing heat cycles on CuCrZr properties
(ITER Task DV4/04)
CuCrZr alloy is one of most suitable materials for heatsinks in the ITER plasmafacing
components (PFCs). The main problem of the alloy is that its thermal and
mechanical properties degrade as soon as 450°C is exceeded, e.g., during the
component manufacturing process or during operation in off-normal conditions.
A parametric study of the degradation was carried out to define the safe working
limits of CuCrZr and to choose the best thermal cycle for the component
manufacturing. The specific aim was to envisage the temperature and the influence
of the thermal treatment time on the mechanical and thermal properties of this ITER
grade alloy.
Hence, CuCrZr, in the solution annealing/water quench condition, was subjected to
different heat treatments at temperatures of 475, 500, 550, 600 and 700°C and hold
times of 5, 10, 20, 30, 40, 60, 120, 180, 240, 300 and 360 min. The reference ageing
treatment for CuCrZr is 3 h at 475°C, which corresponds to a hardness of 130 HV.
This is lower than the HV (159) of the as-received condition, due to the cold work
effect. Assuming a minimum acceptable value of 100 HV for the mechanical strength
of CuCrZr (which should guarantee a tensile strength in excess of 300 MPa at room
temperature), a hot isostatic pressing (HIP) temperature as high as 550°C would be
acceptable.
A previous study by the Joint Research Centre (JRC) Ispra, aimed at establishing the
minimum cooling rate required by the annealing temperature, had demonstrated
that successful ageing of CuCrZr can only be achieved if the cooling rate is at least 2
K/s from 970 to 870°C, and after that, faster than 1 K/s. Analysis of the results
showed that a HIP temperature of ~550ºC is probably the best compromise for a
reliable manufacturing process.
The results also confirmed that, starting from an ageing condition, a hold time of

66
3. FUSION TECHNOLOGY
3.2 First Wall and Divertor
600°C leaves the mechanical properties at acceptable values, although better results
can be expected from a manufacturing process that starts from an as-received
condition, with the temperature kept below or equal to 650°C.
3.2.2 Manufacturing of small-scale W monoblock mockups by hot
radial pressing (ITER EFDA R&D Tasks)
The aim of this activity is to develop an alternative technique for manufacturing the
ITER PFCs, which have a monoblock geometry (i.e. the vertical target).
The basic idea is to perform radial diffusion bonding between the cooling tube and
the tungsten tile, with the process parameters such that degradation of the thermalmechanical
properties remains limited.
The feasibility of joining Cu//Cu and Cu//W by diffusion bonding was studied,
and some small-scale W monoblock mockups were successfully manufactured by
placing them inside a special stainless steel container that does not deform during
HIP, and tested for thermal fatigue (20 MW/m 2 for 1000 cycles).
Following the good results obtained in the tests, a canister was then designed to
perform hot radial pressing (HRP) in a standard furnace in which only a section of
the canister (fig. 3.1) is heated and just the internal tube is pressurised up to the
bonding pressure. The main advantage of this technique compared to HIP is that
neither a high temperature/pressure furnace nor machining of the sheath is
required.
A dummy component was first tested using the following process parameters:
temperature 600°C and pressure 700 bar applied for 3 h. The tests confirmed the
capability of the canister to withstand the load conditions required by HRP.
3.2.3 Runaway electrons on ITER PFCs (EFDA Contract /00-520)
In 2001, the assessment of the thermal effects of runaway electrons (RAEs) on the
ITER-FEAT plasma-facing components was concluded [3.1, 3.2].
The integrated, versatile, multi-particle Monte Carlo code FLUKA was used to get
the energy deposited inside the PFCs by a 10- or 15-MeV RAE impinging on the firstwall
structures with an incidence angle of 1°. The geometrical model is a 3-D layered
structure divided into 24 unit regions centred on the cooling tubes. Starting from the
plasma, the model consists of armour, heatsink, cooling tube and coolant. Constant
conditions were assumed in the poloidal direction. Five different geometries were
investigated: 1) primary first wall armoured with Be (with and w/o protecting
carbon fibre composite (CFC) poloidal limiters); 2) two port limiter first-wall options;
3) Be flat tile; 4) CFC monoblock; 5) divertor baffle first wall armoured with W. The
deposited energy density, normalised to one electron, for the Be-armoured first wall,
and a 10-MeV RAE is shown in figure 3.2.
∅54
∅26
[3.1] G. Maddaluno, S.
Rollet, G. Maruccia,
Thermal effects of
runaway electrons on
ITER plasma facing
components, EFDA
Contract 00-520 - Final
Report - September 2001
[3.2] G. Maddaluno et al.,
Energy deposition and
thermal effects of
runaway electrons in
ITER-FEAT plasma
facing components, in
preparation
115
150
Fig. 3.1 - Cross section of
the canister.

3. FUSION TECHNOLOGY 67
3.2 First Wall and Divertor
5 mm
Fig. 3.3 - Temperature distribution for RAE=10 MeV for 0.1s.
Fig. 3.2 - Normalised
energy density (GeV/cm 3 )
deposited by 10-MeV
RAE.
On the basis of the FLUKA outputs, the temperature pattern inside
the first-wall structures was defined with the use of the finiteelement
heat-conduction code ANSYS. The RAE energy deposition
density was assumed to be 50 MJ/m2, and both 10- and 100-µs
deposition times were considered. The temperature pattern just after
the RAE energy deposition, for an electron energy of 10 MeV and
energy deposition time of 0.1 s, is shown in figure 3.3 for geometry
1). The amount of armour material exceeding the melting
temperature is shaded grey in the figure.
The analysis demonstrated that for all the options but the Be flat-tile port limiter, the
heatsink and the cooling tube beneath the armour are well protected for both the
RAE energies and both the energy deposition times. However, there is a high degree
of melting (ablation) of the W (Be) surface layers, which would eventually affect the
PFC lifetime. As for the primary first wall with CFC poloidal limiters, the limiters
suffer severe ablation, the heat loads being six times larger than those in toroidally
uniform structures. As much as 15 mm of carbon per pulse is removed from the
limiter heads.
3.3 Vacuum Vessel and Shield
3.3.1. EM analyses of in-vessel components for ITER-FEAT
[3.3] EFDA Contract 00-
544, Design of the plasma
facing component (PFC)
for the divertor of ITER-
FEAT (2000)
[3.4] EFDA Contract 00-
570, EM analyses of
shielding blanket for
ITER-FEAT design
options, during plasma
disruptions (2000)
In ITER, the electromagnetic (EM) loads driven by plasma disruptions are one of the
most problematic issues for the in-vessel engineering. Considerable effort has been
spent on design analysis and R&D to obtain in-vessel components capable of
withstanding the EM loads induced by plasma disruptions. During 2001, extensive,
very detailed EM analyses were performed in support of this issue. For the support
to be really effective, the analyses had to have competing objectives: very accurate
component modelling, precision in describing the EM transient and, due to the very
large number of cases to be treated, very short computing time. The objectives were
achieved with the use of the zooming procedure developed at ENEA, which made it
possible to run the number of cases needed to select, for each component, the design
option with the best performance [3.3, 3.4].
The EM loads induced by the ITER reference plasma disruptions were evaluated for
the following in-vessel components: divertor, ICRH assembly, equatorial port limiter

68
3. FUSION TECHNOLOGY
3.3 Vacuum Vessel and Shield
a)
10
1.8
I(MA) corrected
1.3
uncorrected
c)
0.9
R(m) correcteduncorrected
0.4
Z(m) corrected uncorrected
0.0
-0.5 10-2
0.0 1.3 2.6 3.8 5.1 6.4
b)
Moments (MNm)
1.5
1.0
5.0
+
0.0
0.88
-5.0
Mr corrected VDE
My corrected VDE
Mr original VDE
My original VDE
+
+ + + +
+
+
+
+
0.39 0.4 0.41
+
0.42+
0.43 0.44 0.45 0.46
+
+
+
d)
-1.0
Time (s)
Fig. 3.4 - FEM electromagnetic model of ITER
shielding blanket module.
assembly and shielding blanket modules. The aim was to get a precise assessment of
the loads and to investigate the effectiveness of the geometrical features of the
components in reducing them. The most critical in-vessel components were
indicated, and some design variants compatible with the main design philosophy
were suggested in order to reduce the loads. The input excitations from simulations
performed with time evolving MHD equilibrium codes were critically examined. It
was demonstrated that the behaviour of the very last phase of the current quench
could be numerical in origin. Accordingly, in agreement with most of the
experimental evidence from the main operational tokamaks, a modification of the
last current quench phase was suggested. Figure 3.4 shows the finite-element
method (FEM) model developed for shielding-blanket module #1, together with the
plot vs. time of the main EM loads on the component. The significant effect of the
modification, in spite of the very small correction to the input excitation, is clear from
the figure. The ITER Co-ordination Technical Activity (CTA) team agreed to consider
the correction, as it could lead to a more realistic estimate of the loads.
3.3.2 ITER-FEAT breeding blanket
A water-cooled breeding blanket with both breeder and multiplier in the form of
single-sized pebble beds was studied for ITER-FEAT (see fig. 3.5). Analyses showed
that the proposed solution is capable of keeping the minimum breeder temperature
at a sufficient level to enable continuous removal of the generated tritium, while the
maximum temperature in the multiplier (beryllium) does not lead to dangerous
chemical reactions with the water. Figure 3.6 reports the temperature distribution in
the module. A sensitivity analysis was performed to evaluate how the variation in
thermal conductivity of the multiplier affected the temperature distribution.

3. FUSION TECHNOLOGY 69
450
3.3 Vacuum Vessel and Shield
120
60
O30 /
O19 /
O10 /
15
20
150
Support plate
90
150
200
30
Column 2
Column 1
850
Cooling plates
(Thickness 5 mm)
First wall
Fig. 3.5 - Arrangement of
ITER-FEAT blanket module.
230
30
Fig. 3.6 - Blanket module:
general temperature distribution
from theoretical
conductivity of the beryllium
pebble bed.
[3.5] T. KATO et al., First test results for the ITER
central solenoid model coil, presented at the 21st Symp.
on Fusion Technology SOFT (Madrid 2000)
[3.6] Y. TAKAHASHI et al., Cryogenic Eng. 35, 7, 357
(2000)
[3.7] N. MARTOVETSKY et al., CSMC and CS insert test
results, presented at the 2000 Appl. Superconductivity
Conf. ASC (Virginia Beach 2000)
[3.8] N. MARTOVETSKY et al., First results on ITER CS
model coil and CS insert, presented at the 14th ANS
Topical Meeting on the Technology of Fusion Energy
(Park City 2000)
[3.9] H. TSUJI et al., Progress of the ITER central
solenoid model coil program, presented at the 18th IAEA
Fusion Energy Conf. (Sorrento 2000)
[3.10] N. Martovetsky et al., Test of the ITER central
solenoid model coil, CS insert and TF insert, I.P. at the
17thInt. Conf. on Magnet Technology (Geneva 2001)
[3.11] T. Ando et al., Pulsed operation test results in the
ITER-CS model coil and CS insert, presented at the 17th
Int. Conf. on Magnet Technology (Geneva 2001)
[3.12] E.P. Balsamo et al., Physica C 310, 258 (1998)
3.4 Magnets
3.4.1 Installation and testing of ITER CS
and TF model coils (ITER Task M20)
The central solenoid model coil (CSMC) was
extensively tested in static and pulsed regimes in the
first half of 2000. All the main goals of the testing
program were achieved and the results presented at
the major conferences [3.5-3.11].
The coil then became a large testing facility in which
samples of different types of conductors (40 – 90 m
long) have already been tested (CS insert in 2000,
toroidal field (TF) insert in 2001) or will be tested
(Nb-Al insert in 2002, poloidal field (PF) insert in
2003) in static or variable fields (up to 13 T).
The main contribution of ENEA to data analysis
concerned the study of AC losses, particularly a
quantitative determination of the continuous
decrease in the coupling losses in the conductor
during the test campaign. This phenomenon had
already been observed during the tests of an ITERrelevant
coil at the ENEA laboratories [3.12].
The toroidal field model coil (TFMC) (fig. 3.7) was
installed in the TOSKA facility at

70
3. FUSION TECHNOLOGY
3.4 Magnets
Fig. 3.7 - ITER toroidal field model coil prior to installation
in the TOSKA facility at FZK.
Forschungszeuntrum Karlsruhe (FZK) Germany in the first
half of 2001 [3.13, 3.14], and a first phase of tests (coil alone)
was carried out during the summer [3.15]. The rated current of
80 kA was successfully achieved; this is the highest current
level to date for a large superconducting magnet. All the other
measured parameters (e.g., joint resistance, current sharing
temperature, stress level, coil deformation, thermal-hydraulic
behaviour) were in fair agreement with the predicted values.
These results represent important milestones on the way to the
construction of the ITER reactor, as they demonstrate that
large high-field superconducting magnets with predictable
properties can be designed and constructed.
The Turin Polytechnic (POLITO) participated in these tasks
under an ITER-EFDA contract: The M&M code was used to
complete the development [3.16] of a multi-step heating
strategy for Tcs measurements on the TFMC, without the
Large Coil Test [3.17]. Five Tcs tests were performed at 80 kA,
one at 69 and two at 57, all ending with a quench and safe
dump of the coil current. The pressure drop in the heated
TFMC pancakes was analysed [3.18]. POLITO also performed
extensive analyses of and validation against data with the
MITHRANDIR code [3.19-3.22], contributed to the first steps
[[3.23, 3.24] in developing the THELMA code (coupled
thermal-hydraulic and EM description of a superconducting
cable-in conduit conductor) and, finally, participated in the
TFCI test campaign.
[3.13] R.K. Maix et al. Fusion Eng. Des. 58-59,
159 (2001)
[3.14] A. Ulbricht el al., Assembly in the test
facility, acceptance tests and first test
results of the ITER TF model coil, presented
at the 17th Int. Conf. on Magnet Technology
(Geneva 2001)
[3.15] D. Ciazynsky et al., Resistances of
electrical joints in the TF model coil of ITER,
presented at the 17th Int. Conf. on Magnet
Technology (Geneva 2001)
[3.16] L. Savoldi and R. Zanino, Extended
analysis of Tsc tests in DP1.2 using M&M,
presented at the 13th TFMC Test and Analysis
Meeting (Karlsruhe 2001)
[3.17] L. Savoldi et al., First measurement of
the current sharing temperature at 80 kA in
the ITER toroidal field model coil (TFMC),
presented at the 17th Int. Conf. on Magnet
Technology (Geneva 2001)
[3.18] R. Zanino and L. Savoldi, Pressure drop
analysis in DP1 @ 4 K, presented at the15th
TFMC Test and Analysis Meeting (Cadarache
2001)
[3.19] K. Hamada et al., Experimental results
of pressure drop measurements in ITER CS
model coil tests, presented at CEC (2001), to
appear in Adv. Cryo. Eng.
[3.20] R. Zanino et al., Pressure drop analysis
in the CS insert coil, presented at the
Cryogenic Engineering Conf. (Madison 2001)
[3.21] R . Zanino et al., Inductively driven
transients in the CS insert coil (I): heater
calibration and conductor stability tests and
analysis, presented at the Cryogenic
Engineering Conf. (Madison 2001)
[3.22] L. Savoldi, E. Salpietro and R. Zanino,
Inductively driven transients in the CS insert
coil (II): quench tests and analysis, presented
at the Cryogenic Engineering Conf. (Madison
2001)
[3.23] L. Savoldi and R. Zanino, Thermalhydraulic
module for the THELMA code,
presented at the Meeting on the Development
of Tools for the Analysis of AC current
Distribution in Superconducting Magnets
(Garching 2001)
[3.24] R. Zanino, L. Savoldi, P.L. Ribani,
Preliminary results on coupling between TH
and EM (conductor) modules for THELMA
code, presented at the Meeting on the
Development of Tools for the Analysis of AC
Current Distribution in Superconducting
Magnets (Frascati 2001)

3. FUSION TECHNOLOGY 71
3.4.2 Development of calculation codes for CIC conductors (EFDA
Task TWO-T400-1/01)
The development of a new calculation code, started in 2000, continued in 2001. The
code includes all the macroscopically relevant physical phenomena characterising a
forced flow cooled cable-in-conduit (CIC) conductor and is being developed by
various universities co-ordinated by ENEA. The code is structured as a set of four
separate software modules, each describing one conductor characteristic.
The modules deal with the EM and thermal-hydraulic behaviour of the conductor,
the EM behaviour of the joints and the mechanical behaviour of the cable. During
2001, all these modules were fully developed and the first tests started. The EM
section was first tested vs. the current distribution code (CUDI) developed by the
University of Twente. (This code is a simplified tool applied in the case of a short
single-stage cable.) Using 36 ad-hoc-fabricated insulated Cu wires cabled as 3x3x4, a
second code test was performed. The self- and mutual inductances of the conductor
were measured and compared successfully with those calculated by the code. Finally,
the code was used to calculate the expected value of the self- and mutual inductance
for a group of strands of the ENEA Stability Experiment Upgrade (SExUp) .
The thermal-hydraulic module was successfully tested vs. the already validated
MITHRANDIR code. Finally, the two electromagnetic and the thermal-hydraulic
sections were coupled together to check the coupling effectiveness.
The final tests of the results from the ENEA SEx are planned for 2002.
3.4 Magnets
3.4.3 New diagnostics for a CIC conductor (EFDA Task TWO-T400-
1/01)
The helium temperature in a CIC conductor is usually measured by means of
resistance sensors glued or soldered onto the external part of the conductor jacket.
This arrangement gives an indirect measurement of the helium temperature, a delay
in the temporal response during fast transients and a very light EM neutrality that
seriously affects the measurement accuracy. One way of overcoming these problems
is to use optical measurements, which can give fast, highly accurate, noise and
magnetic-field-insensitive helium temperature measurements.
This new approach is based on an optical fibre with Bragg gratings that can be photoimprinted
into the fibre. By illuminating the fibre with a broadband source of light,
a narrow band is reflected at the Bragg wavelength. Its dependence on temperature
comes from two effects: the index of refraction and the thermal expansion of glass.
The local temperature at different positions can be measured by imprinting along the
fibre various gratings with different pitches. In the cryogenic temperature range, the
heat expansion coefficient is not monotonous [3.25], and the feasibility of the
measurement still has to be demonstrated. Moreover, a strain effect competes with
the thermal effect by increasing the fibre length, although temperature/strain
discrimination has recently been successfully achieved [3.26].
[3.25] G. H. White et al.,
Phys. Chem. Glasses 6, 3
(1965)
[3.26] M.G. Xu et al.,
Electron. Lett. 30, 1085
(1994)
A cryogenic system was realised to test the optical fibre temperature measurements
by comparing them with measurements from a traditional sensor. Three different
fibres were tested (fig. 3.8): uncoated fibre (i.e., fibre whose external acrylate
protection coating has been removed) in order to have a reference value for the bare
fibre; fibre with its own acrylate coating; fibre whose acrylate coating has been
removed and the sensor coated with zinc by plunging the fibre into liquid Zn. As
shown in figure 3.8, the Zn-coated fibre shows a much larger heat expansion
coefficient than the others.

72
3. FUSION TECHNOLOGY
3.4 Magnets
Wavelength shift (nm)
0
-2
-4
-6
-8
uncoated fibre
acrylate coated fibre
Zn coated fibre
100 200 300 400 500 600
Time (s)
Fig. 3.8 - Optical fibre
temperature measurement.
3.4.4 Development of NbTi conductors for ITER PF coils (ITER
Task M50, EFDA Task TWO-T405/1 and TW1-TMC/SCABLE)
The joint R&D activity with CEA Cadarache on NbTi conductors for ITER continued
successfully. In accordance with the programme, two 108-strand cables made from
Alstom and Europa Metalli NbTi strands were jacketed by Europa Metalli and used
by CEA to manufacture two NbTi subsize joint samples (SSJS), under testing at the
JOSEFA facility, Cadarache.
A long subsize 36-strand (Ni-coated NbTi EM strands) CIC conductor manufactured
at Europa Metalli achieved a void fraction of 36.8%. To ascertain the role of the void
fraction in the coupling loss time constant (nτ), three additional short samples with
void fractions down to 31% were manufactured from the initial conductor. The nτ of
the samples was measured at the University of Twente, and the results showed that
the Ni coating makes the coupling loss time constant, i.e., the transverse resistivity of
the cable, rather insensitive to variations in the void fraction in the explored range.
A total conductor length of 120 m was delivered to the winding company, Ansaldo
CRIS, to manufacture the SExUp test magnet.
Characterisation of Europa Metalli NbTi strand
As already mentioned, two types of basic NbTi strands are used for the subsize and
full-size conductor samples to be tested at the Sultan facility: an internal Cu-Ni
barrier strand without any external coating, manufactured by Alstom, and a Nicoated
strand manufactured by Europa Metalli.
Complete electrical characterisation of the two strands was performed at the ENEA,
CEA and Twente University laboratories. The critical currents, AC losses and critical
temperatures of the EM strands were measured at ENEA. The strand characteristics
are reported in table 3.I.
Direct transport critical current measurements were
performed on a 1-m-long wire sample wound on a 43-mmdiam
Ti-Al-V sample holder, at liquid helium temperature,
with a transversal external field in the range 2-8 T. Critical
current values were determined by the “10 µV/m Electric
Field” criterion on a 20-cm voltage tap distance.
Magnetisation measurements were done on a 6-mm-diam
open-turn sample (9.87 mm 3 volume). The data were
obtained by a vibrating sample magnetometer system at
different temperatures (4.2, 5.0, 5.7 and 6.5 K), with
Table 3.I - Main characteristics of the EM
NbTi strand
Diameter
0.81mm
Cu:non-Cu 1.9
Twist pitch
8mm
N° of filaments 6534
Average filament diameter
6mm
Thickness of Ni coating
1mm
Guaranteed critical current at 6 T, 4.2 K > 380A

3. FUSION TECHNOLOGY 73
3.4 Magnets
Fig. 3.9 - NbTi critical
current curves I c (B,T).
Ic(A)
1200
1000
800
600
400
200
T = 4.2K
Ic fit TWENTE
Ic TWENTE
Ic magn
Ic ENEA
Ic fit CEA
0
1 2 3 4 5 6 7 8 9
B(T)
magnetic fields up to 10 T. The
system was periodically
calibrated using a nickel
sample with the same NbTi
strand shape. Magnetisation
cycles not only provide a
measurement of AC losses
during external magnetic field
cycles but also allow
determination of the critical
current; the critical current
density J c (B,T) is related to the
magnetisation cycle amplitude
∆M(T,B).
A basic difference between
transport and magnetisation Jc
measurement lies in the
presence of the transport current, which modifies the field penetrating the sample.
For comparison purposes, I c (T,B ext ) data have, therefore, to be referred to an
external applied field.
Critical temperatures were determined by the “half-of-full resistance” criterion of the
resistive transition, at 0 and 5 T, using a 3-cm-long wire sample. A 50-mA current was
applied, which resulted in very sharp transitions.
The transport I c (B) data measured on the same strand by Twente University and
CEA were compared with data obtained by ENEA. Figure 3.9 shows the NbTi critical
current curves I c (B,T) of CEA and Twente, together with ENEA’s experimental I c
data for both magnetisation and transport measurements. The agreement is quite
good in the field range from 4 to 8 T, while at very low field, the magnetisation Ic are
larger than those from the transport measurement.
Configuration and experimental programme of ENEA SExUp
One of the most interesting results obtained during the testing of the ITER model
coils was that the slope of the critical current curve of the CIC conductor seemed to
be reduced compared with that of the single Nb 3 Sn strands. Two different
hypotheses have been proposed: the first takes into account a possible uneven
current distribution across the cable; the second starts from possible damage caused
to the strand by Lorentz forces.
A NbTi superconducting magnet is scheduled to be tested in conditions as close as
possible to those foreseen for the ITER poloidal field coils. To investigate the effect of
uneven current distribution on CIC conductors, an innovative electrical joint was
added to this magnet. This configuration will make it possible to force a controllable,
measurable, uneven current distribution in the conductor and to evaluate the effect
of the current distribution on the magnet. A new set of very accurate flow meters will
be used to evaluate the helium flow during fast transients and its effect on stability.
[3.27] P. Bellucci et al.,
Stability dependence on
flow in a CICC, to be
published in Physica C
Dedicated voltage taps will be used to measure inter-strand resistivity as a function
of the number of charge/discharge cycles.
Interpretation of SExUp results
Analysis of the stability-experiment data addressed two main topics: magnet
stability dependence vs. helium flow [3.27] and AC loss evaluation.

74
3. FUSION TECHNOLOGY
3.4 Magnets
The first topic is mainly interesting for code validation. What is modelled in thermohydraulic
codes is a transition in cooling regime, the so-called well-cooled regime to
ill-cooled regime. Such a change should depend on flow speed and physically
corresponds to situations where, during a heat generation transient caused, for
example, by an EM external disturbance, all the cooling reservoir contained in the
helium can or cannot be absorbed by the strands. If the heat exchange is effective,
then there should be no dependence of the stability on helium speed.
In measuring the stability, no dependence on the helium flow was found, but the
results obtained from the simulation code were different. This means that probably
the transition zone between well/ill-cooled regimes is not well described in the code.
Due to the difficulties in measuring the helium flow, no final quantitative conclusion
can be drawn, but since the experiment is reproducible, other data can be acquired
to solve the uncertainties.
3.4.5 Test in SULTAN of the ENEA Nb 3
Sn magnet (ITER Task M20)
The ENEA 12-T CIC conductor Nb 3 Sn magnet was dismounted from the Pulsed
Field Facility (PuFF) at ENEA Frascati and modified and assembled in the
configuration for testing in the SULTAN facility at CCRP Villigen, Switzerland. The
magnet is now ready to be shipped as soon as the final test schedule is fixed.
3.4.6 Chemical deposition of oxide buffer layers for YBCO-coated
metallic tapes
The Sol-Gel approach was used to deposit buffer layers of Ca,Gd, Y oxides on
textured metallic substrates. The buffer layers have the double role of avoiding Ni
contamination of the YBCO film and inducing its growth in a well-textured structure.
The Sol-Gel approach is attractive as it is scalable to long-length industrial
manufacture.
After extensive development activities, the appropriate parameters for the chemical
deposition of the oxides were defined.
3.4.7 Development of Nb 3
Al strands for high-field applications
Nb 3 Al strands in a Nb matrix have been found to have significant critical
current density at high fields (B>15T). Their good performance is linked to the
formation of stoichiometric Nb3Al A-15 compound, achievable with the so-called
rapid-heating-quenching technique. A strand formed of a Nb matrix with embedded
Nb-Al unreacted filaments is heated up to 2000°C and then rapidly quenched
below 50°C.
A preliminary prototype of a rapid-heating-quenching apparatus was built and
tested. The experience gained will be used to design an updated version of the
system.
The relevant literature was investigated to check the progress in the experimental
data relative to improving the Nb 3 Al transport properties by addition of a third
element.
The technology implemented for chemical deposition of the buffer layers was
successful and produced layers with good microstructural properties, rugosity and
compactness. The next step will be to develop chemical deposition of YBCO thick
films on top of the metallic-tape + buffer-layer structure.

3. FUSION TECHNOLOGY 75
3.4 Magnets
3.4.8 Feasibility study on eddy current testing of ITER coil case
welds (ITER Task TW1-TMS/MMTFRD)
The feasibility study carried out through experimental tests was successfully
concluded. All the tests were performed in laboratory conditions on a series of
samples containing artificial and natural faults. Eddy current techniques were
successfully used to inspect tungsten inert gas (TIG) and submerged arc multipass
welding (SAW) on thick austenitic 316 LN. The VR-11 probe developed by ENEA
showed very high sensitivity compared to other commercial probes. Probe angle
configurations of 45° and 90° for lateral and central defects, respectively, were
assessed. Working frequencies of 15, 30 and 60 KHz were identified to better
distinguish between superficial and sub-superficial defects. Defects such as voids
and collages with only a few mm of extension can be detected with a good
probability.
With these results it was possible to specify the requirements for operating in field
conditions: probe frequency and data processing (fig. 3.10). Hence, the requirement
now is to validate passing from a prototype
system to an industrial testing system.
C2C30
30 kHz
In conclusion, the ITER coil case multipass
welding can now be inspected with the
eddy current technique proposed and
developed by ENEA. This technique is
easier, faster, less expensive and more
reliable than any other nondestructive
testing techniques. Moreover, at the
moment, it seems to be the only applicable
technique for thick-cast stainless-steel
multipass welds.
B/E-C frontal image
Fig. 3.10 - Reference block: central line (lateral passes
overlapping) inspection by VR-11 probe at 90°, lift-off 4 mm.
The most important target (the feasibility)
has been reach-ed, and the final goal (ITER
coil case real test) can be reached, too,
through subsequent engineering efforts.
[3.28] H. Iida, V.
Khripunov, L. Petrizzi,
Nuclear Analysis Report,
Nuclear Analysis Group,
ITER Garching JWS,
ITER report G73 DDD 01-
06-06 (2001)
[3.29] H. Iida et al.
“Nuclear Analysis of
ITER-FEAT” in preparation
[3.30] MCNP 4B, Monte
Carlo N-Particle Transport
System, Los Alamos
National Laboratory Ed.
by J. Briesmeister, LA-
12625-M, (1993)
3.5.1 3-D nuclear analysis for ITER-FEAT design
3.5 Neutronics
ENEA was strongly involved in the neutronics analysis for the ITER-FEAT (500-MW
fusion power) through support to the Nuclear Analysis Group (NAG) of the Joint
Central Team (JCT) in the nuclear analysis itself and in editing the NAG final report
[3.28, 3.29]. A fairly sophisticated nuclear analysis was performed by means of the
best-assessed nuclear data and codes and the most detailed models. A new 3-D basic
model for MCNP [3.30] was constructed according to a shared effort between the JCT
and the Home Teams (HT) of the ITER-EDA. The basic model is a 20° toroidal sector
with proper boundary conditions at both sides (fig. 3.11). The model includes
analysis of a) global and local nuclear heating for the design of each component; b)
global and local shielding optimisation for hands-on maintenance; c) radiation
conditions in materials sensitive to irradiation; d) activation of materials including
the cooling water.
Among the above nuclear responses, nuclear heating in the toroidal field coil (TFC)
inboard legs required very high accuracy, even at a very early stage of design

76
3. FUSION TECHNOLOGY
3.5 Neutronics
modification. The design limit of nuclear heating in the superconducting magnets is
~14 kW. The heating of the inboard legs has become a major contributor (~ 80%) in
the new machine, and detailed calculations
have shown that it can reach 24 kW if
uncertainties in the nuclear data (the Fusion
Evaluated Nuclear Data Libraries FENDL
[3.31]) and tools are considered. This is a
concern to be tackled by the designers. If the
limit has to be observed, additional shielding
(5-10 cm) will be required in the inboard part
of the machine. However, apart from the
heating on the magnet system, another issue
of equal relevance in the ITER design is the
dose rate outside the machine and around the
cryostat after shutdown. The dose-rate value
decides how long personnel have to wait
before accessing areas of the reactor for
repair/maintenance. The assigned limits are
100 µSv/hr in the cryostat, 12 days after
shutdown, and 10 µSv/hr in the bioshield, 1
day after shutdown. Dose-rate levels can be
derived with a simple 1-D model, and
adequate shielding can be provided, but
penetrations and ports bias the expected
attenuation. A novel method to calculate the
dose rate in ITER was proposed and applied
[3.32]. It is a flexible tool for treating decay
gamma transport in complex geometries and
was extensively used to analyse the shielding
problems related to the three major port
penetrations. Local shielding solutions were
proposed to keep the dose rate in the cryostat
below the assigned limit.
Fig. 3.11 - Poloidal section
of the 3-D ITER basic
model.
[3.31] H. Wienke, M.
Herman, FENDL/MG-2.0
and FENDL/MC-2.0 the
processed cross-section
libraries for neutron
photon transport calculations,
Report IAEA-NSD-
176, Rev. 1 (Vienna 1998)
[3.32] L. Petrizzi et al.,
Advanced Methodology
For Dose Rate Calculation
of ITER-FEAT in preparation
A complete nuclear analysis of the divertor components, high heat flux component
and the cassette was also performed. A map of the nuclear heating was calculated for
the thermal analyses. The maximum power density on the upper tungsten coating
ranges between 4-12 W/cm 3 . The overall nuclear heating on the component is about
58 MW. Special response functions were calculated for some critical issues, such as
reweldability of the manifolds, which are affected by helium production. Helium
production ranges between 0.35-7 appm for the fluence (0.3 MWy/m 2 ) foreseen for
ITER. The reweldability limit is about 3 appm for thin plate/tube welding and 1
appm for thick plate/tube welding. The present design incorporates the cassette
replacement scheme, so that the above reweldability limits are never exceeded. A
shielding analysis was done to check that the cassette reference design provides
sufficient shielding. Nuclear heating was calculated in a limited poloidal extension
of the TFC. (Just the part behind the divertor was described.) The nuclear power
deposited in the TFC is 380 W for the 18 coils. The model assumed closed ports in the
divertor regions, so streaming through them was not considered.
3.5.2 Experimental validation of shutdown dose rates for ITER
A shutdown dose-rate experiment was performed at the Frascati Neutron Generator
(FNG). The material assembly used was suitable for generating a neutron flux
spectrum similar to that expected for the outer vacuum vessel region of ITER. The

3. FUSION TECHNOLOGY 77
3.5 Neutronics
Fig. 3.12 - Measured dose
rate in the cavity centre.
[3.33] P. Batistoni et al.,
Experimental validation
of shutdown dose rate
experiment. Final report
of ITER Task T-426,
FUS-TN-SB-NE-R-002
(2001)
[3.34] P. Batistoni et al.,
Fusion Eng. Des. 58-59,
613 (2001)
[3.35] P. Batistoni et al.,
Benchmark experiment
for the validation of shut
down activation and dose
calculation in a fusion
device, presented at the
Int. Conf. on Nuclear
data for Science and
Technology (ND2001) and
accepted for publication
in J. Nucl. Sci. Technol.
Sv/h
10-3
10-4
10-5
10-6
Background inside the cavity
Measured dose rate (G-M)
Measured dose rate (TLD)
Background dose rate in the cavity
1 day 7 day 1 month
10-7
+ + +
10-5 10-4 10-3 10-2 10-1 100
Time after irradiation (years)
assembly was irradiated
long enough
to create a sufficiently
high level of activation
for monitoring by
dosimeters and other
radiation detectors
after shutdown. Provision
was made for
the cooling time
assumed necessary
before allowing
personnel access. The
objective of the
experiment was to
validate the present
dose rate calculations in a typical and complex shield geometry. It was started in 2000
and completed in 2001 in collaboration with the Technical University of Dresden
(TUD) and Forschungszentrum of Karlsruhe (FZK).
The mockup was irradiated with 14-MeV neutrons for three days at the FNG. The
resulting dose rate was measured for about four months of cooling time by two
independent experimental techniques (fig. 3.12). Other useful measurements, such as
the neutron spectrum, decay gamma-ray spectrum, dose-rate distribution and some
relevant activation reaction rates inside the mockup, were performed [3.33-3.35].
The experiment was then analysed with a rigorous, two-step method (R2S), i.e.,
using the neutron transport code MCNP-4C and the activation code FISPACT, and a
direct, one-step method (D1S), approximate but more straightforward, with an ad
hoc modified version of MCNP used in the nuclear analysis of ITER. The FENDL-2
nuclear data libraries (FENDL/MC-2 for the neutron flux calculation and
FENDL/A-2 for the activation calculation), which are the ITER reference libraries,
were used for both methods. The European libraries EFF/EAF-2001 and the Japanese
libraries JENDL-FF/JENDL-3.2(A) were used with R2S.
The analysis showed that the dose rate measurement inside the mockup is well
predicted by R2S and by all the nuclear data library packages: in the comparison in
figure 3.13, all the computed vs. experimental (C/E) values are close to unity within
the total uncertainty, with the exception of some under-estimations found at about 1
day of decay time.
Fig. 3.13 - C/E dose rate
in the cavity centre.
C/E
1.6
R2S/EFF/EAF2001
1.4
R2S/EEN-2
R2S/JENDL
D1S (FENDL-2/A)
1.2
1.0
8 . 10-1
6 . 10-1
4 . 10-1
10-4 10-3 10-2 10-1
Time after irradiation (years)
100
The approximate
D1S method with
FENDL-2 is also
in good agreement
with
measurements
and gives values
slightly but
systematically
lower than R2S.
This may be due
to the fact that
minor nuclides,
contributing to
the total dose
rate at the

78
3. FUSION TECHNOLOGY
3.5 Neutronics
percent level, are not considered in D1S. It was concluded that the shutdown dose
rate for the outer vessel region is well predicted within 25% by the R2S and D1S
methods with the FENDL-2 library.
3.5.3 Design of the neutron cameras for ITER
ITER will have two neutron cameras for measuring the neutron emission distribution.
This diagnostic system has to provide absolute neutron yield, fusion power, alphaparticle
birth profile and ion temperature, besides the neutron source profile.
The radial camera, located in a horizontal port, consists of a fan-shaped array of
flight tubes (totalling 12×3 ) viewing the plasma through a slot at the blanket/shield
level, intersecting at a common aperture (focal point) defined by a specialised
shielding plug, and penetrating the vacuum vessel, cryostat and biological shield
through stainless-steel windows. Each flight tube culminates in a set of neutron
detectors (both flux detectors and compact spectrometers) housed in a massive
shielded structure outside the biological shield. The geometry of the radial camera is
fixed by the port size; as a result, the plasma fraction covered is rather limited. The
vertical camera has a different configuration: the arrays of 15 chords viewing the
plasma downward are located at four different toroidal locations. Each array of
chords views the plasma through the first collimators in the upper radial port plug
and through the second collimators above the upper cryostat lid. Flight tubes are
placed in the vacuum vessel, above the plug of the upper radial port. The upper
collimators, the detectors and beam dumps are located between the cryostat lid and
the top bioshield and are housed in a massive shielded structure to prevent neutron
scattering and to limit the cryostat activation to allowable levels.
The measurement capability of the system was evaluated for relevant neutron source
profiles [3.36]. In particular, the chord integrals of the neutron emissivity and the
resulting fluxes at the detectors were calculated for both the radial and the vertical
camera, for the reference operation scenario (ELMy H-mode) and for the more
peaked neutron emissivity profiles. The results showed that the accuracy of the
absolute value of total neutron yield measured by the radial camera alone would not
be better than 20% due to the very limited plasma coverage. The combination of the
radial and vertical cameras will increase the accuracy of the absolute neutron yield
to better than 10%, as required. The minimum number of sightlines in the vertical
camera and the effectiveness of the most external sightlines were analysed, taking
into account the neutron backscattering from the first wall. As a result, it was found
that the most external channels of the vertical camera are still effective (although
considerable corrections have to be applied) in the case of the reference ELMy H-
mode operation scenario, which is characterised by a very flat neutron emissivity
profile. In the case of more peaked emissivity profiles, the most external channels
and the ones adjacent to them lose their effectiveness and can cause a significant
level of noise due to backscattering neutrons.
[3.36] P. Batistoni,
Design of the radial and
vertical neutron camera
for ITER, in preparation
The size of the collimator diameters was optimised in the range of variation in the
neutron production rate to improve the measurement capability. Flux monitors
suitable for the ITER camera requirements were identified. As for compact
spectrometers, a number of possible candidates exist; however, they require further
investigation and development before they can meet the ITER requirements for
energy and time resolution in neutron energy spectra measurements. ENEA is
investigating the capability of organic liquid scintillators (NE213) to provide an
effective energy resolution of about 2-3% at 2.45-MeV neutron energy and 1% at 14
MeV in tokamak conditions, i.e., proving neutron/gamma-ray and pulse-height
discrimination at high counting rates. In collaboration with PTB Braunschweig,
Germany the feasibility of the method is being investigated, and the capability of the

3. FUSION TECHNOLOGY 79
3.5 Neutronics
system to reach useful energy resolution will be tested during D-D and D-T
operations at JET.
3.5.4 Evaluation of neutron cross sections for fusion materials (EFF
project)
The correct design of a fusion reactor requires the availability of a complete nuclear
database extending up to 20 MeV in neutron energy. The ENEA Fusion and Applied
Physics Divisions participated in the European Fusion File (EFF) Project by updating
neutron cross-section data. In 2001, the carbon and oxygen cross sections were
newly evaluated on the basis of the latest experimental and theoretical findings. The
neutron capture cross sections were re-evaluated in the entire range of 10 -5 eV up to
20 MeV of incident neutron energy. Model calculations based on state-of-the-art
nuclear structure and nuclear reaction models were employed together with a global
analysis of the latest experimental information. Inverse photo-neutron reaction data
were utilised to cover the energy range above a few MeV. These data and the model
calculations were used for the transitions leading to excited states of residual nuclei.
The resulting nuclear cross-section data were compiled and organised into ENDF 6
nuclear data format. The files were integrated with complete existing data libraries
(including all the reaction channels other than capture). For 12 C, the JENDL 3.2 file
was chosen as a basis; for 16O, the JEF-2 file was selected. The data files were made
available to the community for testing. Preliminary tests were done with standard
format checking codes (FIZCON, PSYCHE, and CHECKR).
3.5.5 Neutronics benchmark experiment on SiC (EFF project)
[3.37] P. Batistoni et al.,
Measurements and
analysis of reaction rates
and of nuclear heating in
SiC, Final report of task
TTMN-002 (1)-001,
Report FUS- TEC- MA-
NE-R-2001
Fig. 3.14 - The SiC block in
front of the FNG target.
Silicon carbide (SiC) is one of the candidate structural materials for a fusion reactor
because it has excellent low-activation, low-decay-heat properties. To validate the
SiC neutron cross-section data in the EFF library, a benchmark experiment was
started in 2000 at FNG. A block of sintered SiC (457 mm x 457 mm, 711-mm thick, 470
kg total weight, 127 pieces) lent to ENEA by JAERI was used (fig. 3.14). The
experiment was completed in 2001 in collaboration with TUD, FZK and the Josef
Stefan Institute of Ljubljana [3.37].
Several nuclear quantities, including neutron and gamma-ray spectra, nuclear
heating and activation rates, were measured at different penetration depths inside
the block irradiated with 14-MeV neutrons (up to about 58 cm, corresponding to
about 10 mean free paths for 14-MeV neutrons). The measurements were compared
with the same quantities calculated using MCNP-4C and the deterministic 2-D code
DORT with EFF-2.4, the new evaluated cross sections for Si-28 included in EFF-3.0,
and the international FENDL-2 and Japanese JENDL-FF nuclear data libraries.
Comparison shows that the
European files and JENDL-
FF well reproduce the
measured quantities, within
the total uncertainty, while
FENDL-2 tends to
significantly under-estimate
the fast neutron flux, as
shown in figure 3.15 where
the C/E values are given for
the neutron flux in the
energy range E > 10 MeV.
The experiment was also
used to validate, through
deterministic and Monte

80
3. FUSION TECHNOLOGY
3.5 Neutronics
Carlo approaches, the numerical tools under development for sensitivity/uncertainty
analysis. Through the
1.3
Nb-93(n,2n) (E
analysis it was possible
1.2
n >10MeV)
to find the reason for
1.1
the underestimation
1.0
(mainly the low value
0.9
of the inelastic cross
0.8
MCNP-EFF-2.4
sections) in FENDL-2
0.7
MCNP-EFF-3.0
DORT/EFF-3.0
and also to check that
0.6
MCNP/FENDL-2
DORT/FENDL-2
the uncertainties in the
0.5
MCNP/JENDL-FF
Total error
cross sections were
0.4
compatible with the
0 10 20 30 40 50 60
experimental findings.
Penetration depth (cm)
C/E
Fig. 3.15 - C/E values for
the neutron flux in the
fusion peak energy range
E > 10 MeV, measured by
the activation technique
using the Nb-93(n,2n)
reaction.
3.5.6 Experimental validation of neutron cross sections for fusion
materials (EAF project)
Chromium is one of the candidate structural materials for a fusion reactor because of
its activation properties. In a relatively short time, pure chromium reaches the
conventional activity and dose-rate limits for disposal and maintenance. In the
framework of the activity for the validation of activation cross sections in the
European Activation File (EAF), a sample of chromium (manufactured by
PLANSEE) was irradiated by the 14-MeV FNG [3.38]. The induced activation was
measured by standard and very low background gamma spectroscopy detectors
(HPGe) located in the Gran Sasso (Italy) underground laboratories.
The sample was irradiated for about six hours at the maximum neutron flux
produced by FNG. The neutron flux and spectrum are well monitored by the
multifoil activation technique. The total neutron fluence at the sample was 4.87×10 12
n/cm 2 ± 3%. After irradiation, the activity was measured for several decay times
ranging from fifteen minutes to three months.
[3.38] M. Pillon, M.
Angelone, Final report
of Task TTMN-002 (5)-
002, Report FUS- TEC-
MA-NE-R-2001 (2001)
The calculations were carried out with the latest version of EAF (2001). The typical
material composition supplied by PLANSEE was input in the EASY code. These
maximum values of impurities were used for the C/E comparison. The
radionuclides not included in the EASY code were determined by measuring their
activity.
The results of the C/E comparison and uncertainty analysis are reported in table 3.II.
The radionuclides in bold are those produced by the impurities in the chromium
sample. The amount is given in Wppm, together with the total uncertainty.
The data in table 3.II show the good quality of the EAF-2001 cross-section data, since
most of the C/E values are, within the uncertainties, close to one. The only exception
is the radionuclides produced by the so-called sequential charge particle reaction
(SCPR), i.e., a two-step reaction like (n,p) → (p,n), which occurs with high-energy
neutrons. These reactions are treated by the EASY system, but in an approximate
way. Another radionuclide which shows a large discrepancy is the V-48 produced by
the reaction Cr-50(n,t)V-48. This is a threshold reaction with the energy threshold
close to the maximum FNG neutron energy, and it is most probable that there are
some errors in the cross-section values near the threshold energy. The experimental
results indicate that these values are overestimated. The high C/E value obtained for
Na24 may be due to the fact that the maximum impurity level for Al and Mg was
used.

3. FUSION TECHNOLOGY 81
Table 3.II - C/E comparison results
Nuclide Half life C/E Exp. err. Cal. err. Production pathways %
V-52 3.7 m 0.94 13.6 % 10.6%
Cr52(n,p)V52 97.0
Cr53(n,d)V52 3.0
CR-49 42 m 0.89 19.1 % 7.1% Cr50(n,2n)Cr49 100.0
MN-52 6 d 2.99 7.3 % Cr52[p,n]Mn52 100.0
V-48 16 d 3.45 43.4% 20.0% Cr50(n,t)V48 100.0
CR-51 28 d 1.03 8.5% 5.0% Cr52(n,2n)Cr51 100.0
MN-54 312 d 0.37 3.8%
MN-56 2.6 h 1.08 20.2% 3.2%
Cr54[p,n]Mn54 51.3
Fe54(n,p)Mn54 48.7
Fe56(n,p)Mn56 99.1
Fe57(n,d)Mn56 0.9
Mg24(n,p)Na24 17.8
Mg25(n,d)Na24 0.5
NA-24 15 h 3.13 12.0% 35.5% Al27(n,a)Na24 50.7
Mg24(n,p)Na24m(IT)Na24 8.0
Al27(n,a)Na24m(IT)Na24 22.7
CO-58 70.9 d 1.0 9.1% 57.3%
Ni58(n,p)Co58 95.1
Ni58(n,p)Co58m(IT)Co58 4.9
New radionuclides found Parent nuclide Error Production pathways %
Wppm
SC-46 83.8 d 0.5 53.1%
3.5 Neutronics
Ti46(n,p)Sc46 58.1
Ti47(n,d)Sc46 18.8
Ti46(n,p)Sc46m(IT)Sc46 15.1
Ti47(n,d)Sc46m(IT)Sc46 8.0
Y-88 107 d 5.1 14.3% Y89(n,2n)Y88 100.0
Activity(Bq/kg)
1015
1013
1011
109
107
105
103
101
10-1
10-6
+ V52
ILW/LLW Limit
IAEA Limit
Material and impurities + SCPR
Pure material
10-4
10-2
Fig. 3.16 - Comparison
between pure Cr sample
and Cr sample containing
impurities irradiated in a
first-wall spectrum for an
equivalent neutron flux of
1 MW/m 2 .
+ Cr51 + V49
Time after irradiation (years)
+ Fe55 + H3
+ + Ar39
Ni63 + C14
100 102 104
3.6.1 IVROS articulated boom
Comparison between the experimental data and the
EASY prediction for the chromium sample indicated that
the data libraries are adequate for a good estimation of
the neutron-induced activation of the sample.
Figure 3.16 compares the radiation induced in a pure
chromium sample and that induced in the sample with
the impurities plus the SCPR predicted by EASY. The
reactions that contribute to long lasting radioactivity are
N14(n,p)C14 and K39(n,p)Ar39 +Ca40(n,2p)Ar39.
3.6 Remote Handling
During 2001, some first-wall maintenance and inspection tasks were performed
according to the scheduled FTU shutdown. Two limiter sectors were replaced
remotely by means of the in-vessel remote operating system (IVROS) (fig. 3.17). A

82
3. FUSION TECHNOLOGY
3.6 Remote Handling
new setup of the multilink control software was developed and
tested. An experimental campaign to assess the inspection
procedure was successfully completed.
3.6.2 Upgrade of DRP heavy
manipulator/crane/trolley
Much of the remote handling equipment installed in the divertor
refurbishment platform (DRP) is for use with direct viewing, but
as the ITER hot cell is likely to be displaced from its control room,
direct viewing will not be possible. The first stage in upgrading
the existing handling equipment is to include position sensing on
all eleven axes of the heavy lifting and transport equipment. This
will assist the operator in reproducing key positions accurately
and paves the way for the next stage of the upgrading, which is
to include both automatic positioning of these axes, under the
control of a teach file, as well as the possibility to create a virtual
environment (i.e., a computer model) of the hot cell.
Fig. 3.17 - Toroidal limiter
sector replacement by
IVROS robotic arm.
3.6.3 Trials using ITER FDR 98 duct equipment in
real remote conditions
The remaining work on the divertor test platform (DTP) will
focus on gaining experience with the installed remote handling
equipment in real conditions. The Canadian duct vehicle was
used to carry out a series of representative trials, i.e.,
bolting/unbolting vacuum vessel door fixings, and
bolting/unbolting, and later installation and removal of, rail
sections. The trials were performed in the DTP control room,
without any access to or direct viewing of the handling
equipment.
3.6.4. Installation, commissioning and trials with
the CEA/Cybernetix MAESTRO radiation-hard
servo-manipulator arm on DTP cassette toroidal
mover
This was a long-planned task that underwent extensive delays
due to technical problems with the hydraulic arm. The
equipment was delivered and installed late in 2001, following a
series of interface control and planning meetings held at ENEA
Brasimone and at the CEA headquarters in Paris. The result was
a successful demonstration of the capability of MAESTRO (fig.
3.18) to operate in the confined space of the divertor region and
handle a heavy hydraulic tool to tighten the cassette locking
system following cassette installation.
3.6.5 High-discharge electrical tests of multilink attachment pin
concept at CESI
Although other EURATOM Associations had carried out a series of (mainly)
mechanical tests to establish the suitability of the multilink concept, its capability to
safely carry the thousands of amperes of halo current anticipated during operation
remained untested. Thus, two series of planned trials were performed at the Centro
Fig. 3.18 - MAESTRO
environment at Brasimone.
The slave arm is
between the yellow
cassette toroidal mover
and the blue cassette.

3. FUSION TECHNOLOGY 83
Elettrotecnico Sperimentale (CESI) Milan, with currents of 10, 20 and 30 kA for short
periods. The results (reported to EFDA) indicate that the multilink can safely pass
the current.
3.6.6 Final DRP trials using ITER FDR 98 cassette mockup with
multilink attachments
The latest multilink method for attaching the plasma-facing components to the
divertor cassette was commissioned in the DRP. The final series of trials utilising the
original 1998 FDR design cassette mockup was completed in early 2001 and reported
to EFDA. Further multilink trials await a major upgrade to the DRP environment to
include a new ITER FEAT cassette mockup and associated tooling and handling
equipment.
3.6.7 In-vessel viewing and ranging
3.6 Remote Handling
The 2001 activities, performed in collaboration with ENEA’s Applied Physics
Division, included the design, development, manufacturing and testing of the invessel
viewing probe (LIVVS; viewing accuracy ± 1 mm at 10 m) to be installed and
tested at JET and the in-vessel viewing & ranging probe (IVVS; ranging ± 0,3 mm at
5 m) for ITER. Both systems are based on the amplitude-modulated laser beam
technique.
A new mechanical design of LIVVS was done to meet the new JET EFDA
specifications and to overcome the scanning head oscillation problems. LIVVS is
scheduled for complete testing within July 2002, and a suitable testing period has to
be identified in the JET experimental program.
The IVVS components were all procured and the complete system is being
assembled: the overall probe dimensions (scanning head + launching and receiving
optics) are within 800×160×160 mm. The related optical and electronic parts were
successfully developed. Figure 3.19 shows an example of the IVVS viewing and
ranging performances. Complete testing of the probe should be completed within
2002. Possible applications for other viewing and ranging activities in the ITER
system (glove boxes,…) are now under study with the EFDA Close Support Unit.
Fig. 3.19 - IVVS viewing
and ranging performance:
a) coin photo; b) IVVS
view of coin; c) IVVS
ranging of coin. Note
submillimetric viewing &
ranging performance
compared with actual coin
dimensions.
3.7.1 Compatibility of SiC f
/SiC composites with Pb-17Li
3.7 Materials
The 2001 work was a continuation of the studies on the compatibility of SiC f /SiC
composite with liquid Pb17Li at about 550°C for exposure times of 100, 1000 and
6000 h in physical-chemical conditions representative of those of the TAURO blanket

84
3. FUSION TECHNOLOGY
3.7 Materials
(liquid metal velocity 0.5-1 m/s). The aim is to quantify
eventual degradation of the mechanical and elastic
properties of the composite because of corrosion, with the
use of nondestructive techniques including geometrical
dimensions, mass variation, longitudinal and torsional
dynamic moduli of elasticity (by the longitudinal and
torsional fundamental resonant frequency method).
The materials to be investigated include CERASEP N31, N41
and ENEA PIP composites.
The upgrading of the LIFUS2 facility at ENEA Brasimone to
increase the exposure temperature to 550°C was completed.
The exposure phase was completed for 100 and 1000 h. The
characterisation of the samples exposed for 100 h showed the
absence of erosion-corrosion phenomena but the presence of
consistent liquid infiltration (fig. 3.20). The characterisation of the samples exposed
for 1000 and 6000 h is ongoing.
3.7.2 Microstructural investigation of radiation effects in RAFM
steel by SANS
These activities are carried out in collaboration with FZK. Small-angle neutron
scattering (SANS) measurements were performed at the High Flux Reactor of ILL-
Grenoble. An automatic sample changer designed for handling highly activated
material (up to 1 Sv at 10 cm) was developed. OPTIFER (I and V) and F82H steels,
neutron irradiated at 250-450°C with 2.8 dpa, with and without post-irradiation
annealing at 525 and 700°C, were investigated. Non-irradiated oxide dispersion
strengthened (ODS) EUROFER97 samples (up to 0.5% Y 2 O 3 ) were also examined.
Under 2.8 dpa irradiation, the SANS cross section increases remarkably compared to
the non-irradiated samples, which reflects the presence of defects, such as He
bubbles. The ratio nuclear +
magnetic/nuclear scattering changes
significantly under irradiation, which
is a sign of changes in precipitate
composition. The difference between
the irradiated and reference
samples appears to be independent
of the orientation relative to
the applied magnetic field in the postirradiated
35.0
28.0
21.0
heat-treated specimens.
This can be attributed to the 14.0
growth of He bubbles. Figure 3.21
shows a comparison of oxide 7.0
particle distributions obtained by
transmission electron microscopy 0.0
(TEM) and SANS for the EUROFER97
0.0 7.0 14.0 21.0 28.0 35.0
ODS samples.
d (nm)
Rel. frequency
Fig. 3.20 - SEM
micrograph of sample
exposed for 100 h,
showing heavy Pb-17Li
infiltration.
Fig. 3.21 – Oxide particle
distributions in EURO-
FER97 ODS 0.3%.
Continuous line: data from
SANS measurements.
Dashed line: histogram
from TEM (Lindau et al.,
ICFRM 10 Proc.)
3.7.3 Mechanical properties of RAFM steel-base material and joints
The isothermal low cycle fatigue (LCF) programme without hold-time was
completed on both EUROFER97 and F82H mod. The final results confirmed the
behaviour found for the first set of tests. Both steels behaved like other hardened and
tempered martensitic alloys: no hardening was observed either for a strain-range of

3. FUSION TECHNOLOGY 85
3.7 Materials
1.5%. Also confirmed was the higher LCF resistance of F82H mod steel at 450°C and
moderate strain-range (0.4-0.75%). The few data obtained so far (at R σ =1.5) are not
sufficient to state whether a larger compressive strain (and related stress) has a
significant effect on the moderate variation of the number of cycles to failure
observed.
Results obtained by submitting EUROFER97 to continuous thermal cycling from 200
to 600°C (without hold time) showed that the behaviour of this alloy is similar to that
found for F82H steel tested in the same conditions. A series of thermal fatigue tests
was carried out with a different temperature range (thermal boundaries T min from
150 to 250°C, T max from 450 to 650°C). The results showed that EUROFER97 has a
slightly better resistance but a higher spread than F82H.
Structural investigation of EUROFER97 welded joints was done by means of x-ray
diffraction. Electron beam welded (EBW) joints on 8-mm- and 25-mm-thick plates
were made at ENEA Casaccia. The non-destructive examinations (dye penetrant and
ultrasonic) showed the presence of macroscopic cracks on the 25-mm-thick welded
plate. The flaws were discovered during strip sharing of the plate and were observed
all along the joint. Other EB welds were made, but the same serious drawback was
found. Examination revealed a crack (between 500 and 2000 µm wide) extending all
along the joint. This flaw seems to be a solidification or liquation crack. Owing to the
unsuitability of these welds for mechanical testing, the only activity related to EBW
concerned the study of post-welding heat treatment (PWHT). The as-welded
Vicker’s hardness ranges from 400 to 415 kg/mm 2 , which is a typical value for an
untempered martensitic structure, so the material is too brittle for structural
applications. PWHTs at 730 and 760°C (soaking time 1 h) decreased the hardness to
250-210 kg/mm 2 , so PWHT is a mandatory process. Further investigations on
EUROFER97 weldability are necessary.
Tensile and impact property testing of commercial as-received ODS PM 2000 and
ODS EUROFER97 was carried out. The ODS PM 2000 steel appears less resistant
than the ODS EUROFER97, as expected since PM 2000 is a ferritic, nontransformable
alloy. On the other hand, impact resistance seems slight higher (6 J
instead of ≈ 5 J), while the ductile-to-brittle transition temperature for both materials
is within 100-140°C. Thermal ageing was performed at 550°C for 1000 and 5000 h.
The tensile properties of aged specimens (testing temperature R.T., 450 and 650°C)
are fully comparable to those of un-aged material. The ageing temperature seems to
be too low to have any structural modification, so PM 2000 appears very stable at this
temperature.
[3.39] M. F. Maday,
Fusion Technol. 39, 2,
596 (2001)
[3.40] M.F. Maday,
Mechanisms governing
fracture of cyclically
loaded F82H mod. steel
in air and water at
240°C, presented at
ICFRM-10 (Baden Baden
2001), to appear in J.
Nucl. Mat.
3.7.4 Low-cycle fatigue of RAFM steel in water with additives
The objective of the experimental activity carried out in 2001 was to complete the
comparative study of the LCF behaviour of two different plates (31W-19 and 42W-
18) of F82H mod heat 9753 already undertaken in pure oxygen-free water [3.39]. The
tests done in the alkaline chemistry recommended for DEMO coolant to establish
eventual correlations between the observed fatigue performances and steel
microstructure were replicated. The nature of the underlying damaging mechanism
was clarified with the support of meaningful experimental indexes. Preliminary LCF
property evaluations of EUROFER97 were also carried out.
With the experimental conditions used [3.40], the fatigue lives and associated
fracture modes reported in figure 3.22 were obtained on specimens from 31W-19
(group I) and from 42W-18 (groups II and III). In water, a minor degree of fatigue
lifetime reduction with respect to air data and an associated minor tendency to
plastic deformation localisation were observed on F82H-31W-19, which included an

86
3. FUSION TECHNOLOGY
3.7 Materials
600
g
f
Air reference curve (groups I,II,III)
560
LiOH water/group I
LiOH water/group II
LiOH water/group III
520
d
Stress amplitude (MPa)
480
440
400
360
320
280
240
e
c
b
a
1 10 100 1000 10000 10000
Numbers of cycles to fracture (Nf)
Fig. 3.22 - F82H mod
specimen: number of
cycles to rupture and
associated macroscopic
fracture aspects after
LCF testing in air and
LiOH-dosed oxygen-free
water at 240°C.
extra population of dispersed and large Al-oxides. Specimens from 42W-18,
containing residual stresses from machining, exhibited the worst LCF performances.
The fractures were either brittle and frequency-dependant or cup-cone and cycledependant,
and involved lath boundaries, oxide/matrix interfaces or carbide/matrix
interfaces.
Based on thermodynamics, fractography and on environment-induced bulk effect
considerations, a hydrogen assisted cracking mechanism for steel fracture
enhancement in a water environment was strongly suggested.
In the light of the above data, the preferred fracture paths for F82H cracking
propagation and fatigue behaviour variability from plate-to-plate were explained
with the hydrogen decohesion theory, i.e., different and concurrent hydrogen trap
populations in the F82H microstructure and fracture behaviour kinetically favoured
at specific sites trigger the fracture event.
3.7.5 Development of a low-activation brazing technique for
SiC f
/SiC composites
The requirements of a fusion-relevant brazing technique are low neutron activation,
good compatibility with breeders, low brazing temperature to avoid fibre
degradation, good wettability with the composite, thermal expansion coefficient
similar to that of the composite and sufficient shear strength. Pure silicon has good
chemical compatibility and wettability with silicon carbide and has been used both
to infiltrate and to join samples. It also has a thermal expansion coefficient similar to
that of silicon carbide. On the other hand, the quite high melting temperature of
silicon, the limited strength exhibited in previous work and the neutron-induced
swelling of pure silicon make its use rather problematic in a fusion reactor
environment. The basic idea for the development of a new alloy and brazing
technique was to use a Si-16Ti (at%) eutectic (melting temperature 1330°C). Si16Ti
has a lower melting point and the titanium enhances the joint strength via the
formation of intermetallic compounds and/or carbide at the interface with SiC [3.41].
Several experiments were performed to obtain the eutectic “in situ” by melting Si-Ti
[3.41] B. Riccardi et al.,
“Low activation brazing
materials and techniques
for SiC f /SiC composites”
presented at ICFRM-10
(Baden Baden 2001), to
appear in J. Nucl. Mat.

3. FUSION TECHNOLOGY 87
3.7 Materials
powder mixtures at >1430°C, but the results were not
satisfactory because of incomplete melting of the mixtures,
which led to inhomogeneities and defects. Thus, before the
brazing operations, the eutectic was prepared by melting a Si-
Ti mixture in an argon plasma furnace and then re-melting it
in an electron beam to get a fine eutectic structure. Powders
were prepared by milling the small ingots obtained and were
then used for the brazing experiments. First monolithic and
then SiC f /SiC composites samples were brazed.
Fig. 3.23 - Si-Ti brazed
joint micrography.
The joining was performed in both vacuum and inert
atmosphere. The joints had a very interesting morphology
(fig. 3.23). In particular, the joint layer showed:
• the absence of discontinuities and defects at the interface as
a result of complete melting of the powders;
• a fine eutectic structure with morphology comparable to that of the starting
powder.
Fig. 3.24 - Calculated
stress distribution in the
sample.
[3.42] C.A. Nannetti, et
al., Development of 2D
and 3D Hi Nicalon
fibres/SiC matrix
composites manufactured
by a combined CVI-PIP
route, presented at
ICFRM-10 (Baden Baden
2001), to appear in J.
Nucl. Mat.
A shear test performed at room temperature by means of a modification of the ASTM
D905-89 standard method gave remarkable results: the samples manufactured with
monolithic SiC cracked at high shear stress level, not in the brazing layer or at the
interface, but in the SiC bulk; while the composite samples exhibited up to 80 MPa
shear strength.
3.7.6 Measurement of residual stresses using neutron diffraction
techniques
In the framework of Underlying Technology, samples of high-heat-flux
components were tested in the ILL High Flux Reactor (Grenoble) to verify the
relevance of the
Residual Stress (MPa)
400
300
200
100
0
-100
-200
-300
-400
-500
Glidcop
5 10 15 20 25
Tungsten
Position (mm)
Long in-plane
Short in-plane
Normal
Interface
3.7.7 SiC/SiC ceramic composites as PFC material
compliance layer in the
stress evolution of the first
sample tested. Preliminary
results show that the
strains in Glidcop vanish
at about 300°C, indicating
a possibility to evaluate
the null strain temperature,
if the lattice is not
deformed by the
compliance layer. Figure
3.24 shows the calculated
stress distribution.
The campaign to manufacture composites with superior properties (Underlying
Technology activity) continued during 2001. In particular, the objective was to
evaluate the effect of densification by chemical vapour infiltration (CVI) and
polymeric infiltration and pyrolysis (PIP) on the thermal/mechanical properties of
Tyranno SA/SiC matrix composites and to compare the results with those of similar
3-D fibre textures of Hi-Nicalon/SiC matrix composites densified by CVI-PIP [3.42].
The fibre volumetric percentage ranged from 35 to 40% and the thickness was about
4 mm.

88
3. FUSION TECHNOLOGY
3.7 Materials
The first step of the densification process was the deposition of a 0.10 mm pyrolithic
carbon by chemical vapour deposition (CVD). Afterwards a first layer of SiC was
provided by CVD for 20 h for a thickness of 0.20 mm. Then, infiltration with polymer
and SiC alpha particles was performed followed by pyrolysis at 1300°C. Finally, an
additional six cycles of PIP were performed without adding any SiC powders. For
comparison, 2-D and 3-D panels were also manufactured without any powder
addition in the first PIP cycle, but the number of PIP cycles was increased to 14. The
micrographs (fig. 3.25) of the 2-D and 3-D composites (manufactured by SiC powder
injection) show that good intrabundle infiltration and a fine, well distributed
porosity was reached. Table 3.III Shows the main properties of the composites
manufactured.
The use of powder during the first PIP cycle seems to reduce the porosity for both
the 2-D and 3-D composites compared to the standard PIP technology, even with a
high number of cycles. Powder injection and advanced SiC fibres marginally increase
thermal conductivity/diffusivity. The 3-D textures show a higher thermal
conductivity than the 2-D, but a high fraction of fibres across the thickness is not a
big advantage without a high crystallinity matrix. A sufficient bending strength was
measured for all the composites produced by PIP+powders, but the 2-D composites
show a better bending strength, probably due to the higher fibre content. In addition,
the mechanical properties of Tyranno SA/SiC composites are generally lower than
the properties of Nicalon CG or Hi Nicalon fibre/SiC composites.
a) b)
Fig. 3.25 - ENEA 2-D (a)
and 3-D (b) composite
micrographs.
Table 3.III - Main properties of the manufactured composites
Panel sample
Fibre density Porosity Th.diffusivity Th.conductivity MOR
% (g/cm 3 ) % (cm 2 /s) [W/(mK)] (MPa)
2-D-A 40.5 2.36 15 0.027 4.2 391
(no powders 13PIP)
2-D-B 43.6 2.54 11.5 0.045 7.5 496
2-D-C 40.6 2.51 13.4 n.a. -- 511
3-D-A 34.9 2.48 12 0.07 11.4 409
(no powders 13 PIP)
3-D-B 34.9 2.56 1 0.6 0.065 10.9 411
3-D-C 36.5 2.62 8.4 n.a. -- 506
3-D-D 37.2 2.6 10.6 n.a. -- 499

3. FUSION TECHNOLOGY 89
3.7.8 Mechanical characterisation of materials with miniaturised
specimens
Development of the portable flat-top indenter for mechanical characterisation
(FIMEC) continued in 2001. Following the achievement of the demonstrative
apparatus for in situ testing, which is based on the use of a 0.7-mm-diam flat
indenter, a portable prototype was designed. Work was also started on developing a
numerical methodology as a comprehensive tool for interpreting the load
penetration curves of different materials.
Two options of the portable FIMEC apparatus were analysed: The first is based
on the same stepping motor and load displacement detection features as used in
the fixed apparatus; the second is more compact and has a different layout, with
an encoder, a small motor and a kinematic chain which drives the indenter tip.
Both solutions are provided with a fixing tool suitable for cylindrical and flat
geometry.
3.8.1 Interaction between lead-lithium alloy and water in DEMOrelevant
conditions (EU Task TTBA-5)
Large water leaks
3.7 Materials
3.8 Liquid Metal Technology and
Hydrogen Effects on Materials
The interaction between molten lead-lithium alloy (in a eutectic composition) and
pressurised water is studied to predict the behaviour of a water-cooled lithium-lead
(WCLL) blanket module in the case of a cooling-tube rupture.
In 2001, three tests (#3,4,5) were conducted at the LIFUS 5 apparatus at ENEA
Brasimone in thermal-hydraulic conditions similar to those foreseen for the WCLL
DEMO blanket.
In test #3, water was injected into the reaction tank at a pressure of 155 bar with
different values of sub-cooling and different free volumes in the expansion vessel.
The initial liquid metal temperature was fixed at 330°C.
In test #4, the water temperature was fixed at 325°C (corresponding to a water subcooling
of about 20°C), the free volume of the expansion vessel was 5 l and the
duration of water injection was 6 s. After about 2 s from the beginning of the test, the
rupture disk D1 (on the line connecting the expansion vessel to the dump tank)
failed, with subsequent depressurisation of the system. Because of the large amount
of injected water, a significant heat effect was also found; a maximum temperature of
683°C was detected in the upper part of the reaction vessel, with an increase of 353°C
over the initial value.
In test #5, water was injected at 265°C, the compressibility of the system was reduced
by decreasing the free volume in the expansion vessel from 5.0 to 4.0 l and the time
of water injection was increased to 12 s. During the experiment, 3.28 kg of water were
injected into the reaction vessel. A maximum temperature of 525°C was detected in
the upper part of the reaction vessel, about 10 s from the beginning of the test, with
an increase of 195°C over the initial value.
Comparing tests #3 and 5, it appears that the free volume in the expansion vessel is
much more significant than the water enthalpy in determining the pressurisation
evolution of the blanket module. This consideration is confirmed by comparing tests
#4 and 5.

90
3. FUSION TECHNOLOGY
3.8 Liquid Metal Technology and
Hydrogen Effects on Materials
Pressure (bar)
160
140
120
100
80
PT1(5)
60
PT2(5)
PT1(4)
40
PT2(4)
PT1(3)
20
PT2(3)
0
0 200 400 600 800 1000 1200 1400 1600 1800 2000
Time (ms)
Fig. 3.26 - Comparison of pressure evolution in tests
#3, 4 and 5.
Temperature (C)
650
600
550
500
450
400
350
300
0 5000 10000 15000 20000 25000
Time (ms)
TC1 test n. 3
TC1 test n. 4
TC1 test n. 5
Fig. 3.27 - Comparison of temperature evolution in
tests #3, 4 and 5.
The behaviour of the pressure and of the temperature vs. time is reported in figures
3.26 and 3.27. Note that the temperature evolutions have similar behaviour until the
failure of the rupture disk.
Small water leaks
The interaction between pressurised water and liquid Pb-17Li, as a consequence of
coolant micro-leaks inside a WCLL blanket module, is a relevant issue and needs to
be studied because of the potential consequences. This issue was evaluated through
an extensive experimental campaign on the RELA loops, which was concluded with
the last test (no.9) performed on RELA III during 2001.
The main operating conditions (lithium-lead velocity 5 mm/s and temperature
330°C in the test section, water-circuit pressure155 bar) were chosen taking into
account the thermal-hydraulic parameters foreseen for the DEMO WCLL blanket
module.
In the last test, the water injection was stopped three hours from the beginning of the
injection phase because of the total interruption of the Pb-17Li flow rate in the circuit.
A total of 1051 grams of water were injected, and 30.5 mol of hydrogen were
recovered by the Ar stream. The molar ratio between the hydrogen and the injected
water was 0.52. The evolution with time of the hydrogen concentration and water
leak rate is shown in figure 3.28. No differences were found in hydrogen recovery
passing from sweeping to bubbling argon in the storage/re-circulation vessel.
Because of the low
linear velocity of the
liquid metal inside the
test section of the
RELA loops (the same
as foreseen in the
ITER and DEMO
WCLL blanket
module) and the high
melting point of some
reaction products of
the chemical reaction
between lithium and
water, a solid shell is
Water leak rate (g/s)
0.25
0.2
0.15
Water leak rate
0.1
0.05
H 2 concentration
0
4000 6000 8000 10000
Time (s)
6
5
4
3
2
1
0
H2 (%)
Fig. 3.28 - Hydrogen
concentration and water
leak rate during test #9
on Rela III.

3. FUSION TECHNOLOGY 91
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formed around the microcrack, enveloping the part of the cooling pipes close to it.
The dimension of this solid aggregation depends on the amount of water injected as
well as on the tube area density. This phenomenon has two main consequences: a
reduction in the liquid metal flow-rate, which, at most, can completely stop, and a
reduction in the heat exchange coefficient, on the liquid metal side, which can cause
hot thermal spots.
In any case, depending on the thermal hydraulic conditions in the module, a part of
the solid reaction products can be removed from the module through the return line
of the circuit. As a consequence, to avoid total or partial obstruction of the pipe
because of plugging, particular care has to be taken in designing the return line to
avoid long horizontal parts as much as possible.
No corrosive effects on the microcrack were found, and the variation with time of the
water micro-leak is to be ascribed only to the formation of solid reaction products
just at the mouth. Consequently, the probability of a “self” stopped microleak is not
negligible.
The amount of hydrogen produced from the chemical reaction between lithium and
water is about 0.3-0.5 mol H 2 per mole of injected water; this data scattering is
probably due to the trapping of un-reacted water inside the solid aggregation
developed around the microcrack. This confirms that, for this particular kind of
interaction, the main solid reaction product is lithium hydroxide (LiOH), while only
the external surface of the solid aggregation contains lithium oxide (Li 2 O).
The hydrogen concentration is more or less in phase with the variation with time of
the water microleak. This means that, particularly for systems of limited dimension
(such as in the WCLL TBM), the hydrogen concentration as detected on the cover gas
could be used, in principle, as a water microleak detector.
3.8.2 Qualification of tritium permeation in Pb-17Li/gas
Aluminium-rich coatings (which form Al 2 O 3 at their surface) produced by CVD and
hot dipping (HD) processes have been selected as the reference solution for the
tritium permeation barriers (TPBs) of the DEMO WCLL blanket.
Fig. 3.29 - Arrenhius plot
of CVD-coated specimen
permeabilities (gas
phase).
The internal surface of the specimen is initially in contact with the vacuum (10-5 Pa),
and the external surface is exposed to hydrogen gas with a nominal purity of
99.9999%. The hydrogen permeates in the sample and causes a pressure rise in the
inner volume. The pressure rise can be converted into the amount of gas in moles
permeating through the unit area of the sample per second. The procedure can be
repeated for different temperatures and
gas flows.
T(K)
Φ (mol m-1s-1Pa-1/2)
1 . 10-11
1 . 10-12
1.3
750
1.4
700
1.5
650
1.6
1000/T (1/K)
600
1.7
550
CVD1 T increase
CVD2 T increase
CVD1 T increase
CVD2 T increase
Reference specimen
Disk shaped sample
1.8
The permeation reduction factor (PRF)
of the CVD-coated specimens was very
poor, probably because of an incorrect
coating procedure. The specimens were
tested only in the gas phase, and the
results are depicted in terms of
permeabilities in the Arrenhius plot of
figure 3.29. Only one HD specimen of
the two tested gave acceptable results in
the gas phase, while the PRF was
significantly lower in the liquid metal
than in the gas the phase (fig. 3.30).

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A second experimental campaign on CVD- and HD-coated
specimens was started at the end of 2001.
3.8.3 Transport parameters and solubility of
hydrogen in Pb-17Li
1 . 10-11
Knowledge of the hydrogen-isotope mass transfer parameters
1
in liquid metal is fundamental for the design of some tritium
. 10-12
processing systems, particularly the devices that extract tritium
from Pb-17Li, which are based on the technology of gas-liquid 1 . 10-13
contact equipment.
1
From previous experiments and theoretical considerations,
. 10-14
1.3 1.4
diffusion in the gas phase is considered to be faster than
transport phenomena through the bulk of the liquid, the liquid
transition layer and the gas-liquid interface. However, identification of the
controlling mechanism in the overall desorption kinetics is complex because it
depends on several parameters, such as hydrodynamic conditions of the liquid-gas
system, gas composition and liquid metal impurity content.
The results achieved in the past by different techniques often disagree with each
other and do not provide a clear understanding of the controlling transport step. The
LEDI device at ENEA Brasimone is based on hydrogen/deuterium permeation
through a thin layer of Pb-17Li, stagnant over a metallic membrane. This system
seems to be more flexible as it is possible to vary the thickness of the liquid metal as
well as the surface conditions, which can strongly affect the kinetics of the whole
transport mechanism. First results with the LEDI device were obtained in the last
part of 2001. Their analysis demonstrated that steady state was not perfectly reached
due to some problems in maintaining high-vacuum conditions during long
experiments. The device is now under modification to improve the accuracy of the
next experiments.
SOLE is an upgrade of LEDI and should provide direct measurement of the
solubility of the hydrogen isotope in Li17Pb83 in the range 300-500°C. The design of
SOLE was based on accurate theoretical modelling performed in co-operation with
the Moscow Engineering Physics Institute (MEPHI). The solubility is determined by
the amount of gas absorbed into the bulk of the liquid metal when the system is at
the steady state.
3.8.4 Hydrogen permeability and embrittlement in EUROFER97
martensitic steel
Hydrogen/deuterium permeation experiments performed in the past on
EUROFER97 showed a non-negligible decrease in permeability with respect to other
fusion-oriented martensitic steels. In 2001, experimental activities were focused on
determining the hydrogen/deuterium transport parameters through aged
EUROFER97 in the temperature range 423-723 K, by a time-dependant permeation
technique, with a hydrogen or deuterium upstream pressure of about 75000 Pa. On
the basis of experimental results, permeability, lattice diffusivity and Sieverts
constant K s,l for deuterium in EUROFER97 are being processed.
Mechanical tests were also done on hydrogen-charged specimens at room
temperature to determine the threshold concentration of hydrogen for hydrogen
embrittlement. Low strain-rate tensile tests were conducted on notched and smooth
Φ (mol m-1s-1Pa-1/2)
1 . 10-10
500 450 400 350
1.5
1.6
1000/T (1/K)
1.7
300
Disk
Reference T increase
Reference T decrease
HD T increase
HD T decrease
HD T increase 2
HD gas phase
1.8
Fig. 3.30 - Arrenhius plot
of HD-coated specimen
permeabilities (gas and
liquid metal phases).

3. FUSION TECHNOLOGY 93
3.8 Liquid Metal Technology and
Hydrogen Effects on Materials
Fig. 3.31 - Area reduction
as a function of hydrogen
content at room and high
temperature.
Area reduction /Area red. virgin mat. (%)
120
100
80
60
40
20
EUROFER steel
T =20°C
T =100°C
T =200°C
0
0 1 2 3 4 5 6 7 8 9
cylindrical
specimens that had
been previously
electrochemically
charged with
hydrogen (contents
of up to 3 wppm) at
high temperature
(90°). The experimental
activities
were performed in
collaboration with
the University of
Pisa.
Hydrogen content (wppm)
As expected, the
hydrogen concentration necessary to have a marked decrease in the area reduction
coefficient was found to be quite high compared to that determined at room
temperature (fig. 3.31).
The experimental activity on hydrogen embrittlement will be completed during first
months of 2002.
3.8.5 Water detritiation systems (EU Task TTBA-D02)
The aim is to assess a design to simplify the WCLL blanket concept by eliminating
the TPBs on the double walled tubes of the primary cooling system and recovering a
significant part of the bred tritium directly through the water detritiation system
(WDS).
From previous studies, it was found that this approach to tritium management
strategy is feasible from a techno-economic point of view only if a steady-state
tritium concentration of several Ci/kg is allowed in the primary cooling loops. In
other words, the tritium specific activity in the primary cooling system must be a
good deal higher than that foreseen in the reference design (1Ci/kg).
A detailed safety analysis on the consequences of a relatively high tritium specific
activity in the primary coolant was, therefore, performed in collaboration with the
University of Bologna. The environmental tritium release was determined for an exvessel
loss of coolant accident (LOCA) in normal operation. The tritium specific
activity considered corresponded to the “economical optimum” for a water
detritiation system, based on electrolysis, distillation columns + vapour phase
catalytic exchange and combined electrolysis catalytic exchange (CECE), in all cases
with a tritium permeation rate (TPR) of 10 g/day from the breeder into the coolant.
Such a TPR corresponds to a PRF of 10 for the tritium permeation barriers. This value
is achievable, in principle, only by using double walled EUROFER97 tubes with
copper as brazing material. A water leak rate of 2 kg/h from the primary cooling
circuit was assumed, with 1.9 kg/h towards the steam generator vault and the
remaining 0.1 kg/h into the secondary circuit through the steam generators.
For an ex-vessel LOCA, even in the worst case, which corresponds to the highest
enthalpy content of the cooling water, the environmental tritium release was
determined to be much lower than the limit of 5 g of tritium in HTO form; this is the
maximum acceptable value according to the ITER Guidelines for Environmental
Tritium Release.

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3. FUSION TECHNOLOGY
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As for the tritium environmental release in normal operation (chronic release), a
tritium specific activity of 9 Ci/kg was found to be acceptable, which is lower than
the limit of 0.37 PBq/y (as
recommended by the
DEMO Safety Working
Group). Of course, in this
case a fundamental role is
played by the tightness of
the steam generator (0.1
kg/h was the assumed
water leakage toward the
secondary circuit through
the steam generator),
which is a crucial issue.
Figure 3.32 reports the
chronic tritium release vs.
TPR for different WDS
technologies.
3.8.6 Measurements of H/D diffusivity and solubility through
tungsten and tungsten alloys in the range 600-800°C (ITER
Task 436)
The evaluation of hydrogen transport and solubility parameters in tungsten, which
is one of the candidate
materials for the ITER first
wall was continued from
Temperature (°C)
2000 (see Progress Report
838
727
1
2000).
. 10-12
W-H 2 gas
The experiments in 2001
were performed with a
hydrogen driving
pressure of 105 Pa. The
preliminary results in
terms of permeability are
in good agreement with
the past results and with
the literature data (see fig.
3.33). Further experiments
will be carried out during
2002.
Environmental tritium release (PBq/y)
Φ (mol m-1s-1Pa-1/2)
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0 10 20 30 40
Tritium permeation rate (g/day)
1 . 10-13
1 . 10-14
tritium activity in the coolant at the
optimum economical point
Linear Fit PERI2
PERI2
Frauenfelder
Serra
Perujo
0.95
1
1.05
1000/T (1/K)
ELECTROLYSIS
DC+VPCE
CECE
ANNUAL LIMIT
1.1
1.15
Fig. 3.32 - Chronic
environmental tritium
release vs. TPR for different
WDS technologies.
Fig. 3.33 - Arrenhius plot
of tungsten permeability
(experimental) compared
with literature data.
3.8.7 Corrosion and mechanical tests on structural materials in
flowing Pb-17Li (EU Task TTMS-003-D13)
This activity concerns corrosion rate evaluations and tensile tests on specimens of
EUROFER97 steel pre-exposed in the LIFUS II loop to flowing Pb-17Li with a rate of
0.6 l/h, at 480°C for up to 5000 h. Corrosion and tensile samples were extracted from
the LIFUS II loop test section every 1500 h. Post-test weight-change measurements as
well as metallurgical analysis were performed on the corrosion specimens to
estimate the corrosion rate and to evaluate its mechanism. Tensile tests were done at
480°C on the corroded tensile specimens to assess the mechanical properties of
exposed steel.
The weight change measurements showed that the steel has a linear trend of weight

3. FUSION TECHNOLOGY 95
3.8 Liquid Metal Technology and
Hydrogen Effects on Materials
Fig. 3.34 - SEM-BSE
micrograph of the cross
section of the 4500-h
tested sample.
Fig. 3.35 - Tensile properties
obtained at 480°C
on EUROFER97 steels
exposed to Pb–17Li.
Tensile strength (N/mm2)
Pb-Li
500
400
300
200
100
Corroded
layer
20 µm
RM
RP 0.2
Z%
80
60
40
20
0
0 1000 2000 3000 4000 5000 0
Reduction in area (%)
loss with increasing exposure time.
The corrosion rate of the steel,
evaluated in the given experimental
conditions, was 40 mm/y. This rate
seems to be in agreement with the
general trend of corrosion rates of
ferritic-martensitic steels exposed to
flowing Pb-17Li. The corrosion
mechanism could be explained by
considering that EUROFER97 steel
suffers typical corrosive attack due
to dissolution of the steel elements
in the liquid metal. Figure 3.34
shows the scanning electron
microscopy (SEM) micrograph
obtained with the back-scattered
electron (BSE) detector on the cross
section of the 4500-h tested samples.
The figures show that a layer about
1 mm thick seems to be detached
from the surface of the steel. At the
interface between the layer and the
bulk material, voids can be seen,
and energy-dispersion x-ray
analysis showed that the quasidetached
layer was Cr depleted.
Exposure time in flowing Pb-17Li (h)
With regard to the tensile
properties, figure 3.35 reports the average yield strength, maximum strength and the
area reduction obtained on sets of five specimens for each point are plotted vs.
exposure time in flowing Pb-17Li. The plot shows that the exposure of the steel to the
liquid metal did not affect its mechanical properties. The observed tensile behaviour
confirms the results obtained in the past, that the mechanical properties of the
ferritic-martensitic steels are unaffected by flowing liquid Pb-17Li.
Simultaneously to the experimental campaign carried out on EUROFER97, LIFUS II
was used for experiments on the effects of corrosion on tensile properties of
SiC f /SiC. The first tests were conducted at 550°C for an exposure time of 100 h. The
results are presently under evaluation.
The activity should continue during 2002 through experimental campaigns with a
longer exposure time.
3.8.8 Interaction chemistry between Li 2
TiO 3
ceramic pebble bed
and EUROFER97 in He + 0.1% H 2
purge gas at 600°C
The helium-cooled pebble bed (HCPB) blanket is one of the two concepts in the
European Blanket Programme under development for testing in a next-step fusion
machine. For this particular task, the interaction chemistry between EUROFER97
and the Li 2 TiO 3 ceramic breeder pebble bed in flowing He + 0.1% H 2 was studied.
Experimental data were obtained, for up to 200 h of exposure time, by using a
thermo-balance with a controlled flow gas of 80 scmc/min and a water content of
less than 20 ppm at the start. For an exposure time of between 500 and 2000 h, a
flanged alumina tube was set in a horizontal oven with a controlled flow gas of 80
scmc/min, in case the water content was verified to be less than 50 ppm at the start.

96
3. FUSION TECHNOLOGY
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Weight change data were calculated and reported in a parabolic diagram. Two
different parabolic rate constants that had similar values for the two different
systems were obtained, indicating a self-limiting mechanism due to a diffusioncontrolled
step of the reacting species. In this case, the system studied can basically
be regarded as the oxidation of metals due to possibility of the presence of lithium.
Moreover, the chromium content in the alloy is sufficient to promote the formation
of a chromium oxide layer able to protect the metal from further oxidation. In any
case, the two kinetics are basically identical due to the water-production rate, which
is determined by temperature, and both the parabolic rate constants fall within the
range reported in the literature.
Oxidation proceeds with time from
areas under the pebbles in contact with
the metal surface (fig. 3.36), up to
almost the entire surface after 2000 h,
when an oxide layer not exceeding 2
µm anywhere is observed. Under the
pebbles, the metal surface shows cracks
with a maximum depth of about 4 µm
(fig 3.37). It is not clear whether the
mechanism proceeds via the oxygen
that enters the metal or whether the
chromium diffuses outward or, more
probably, both.
X-ray diffraction analysis (XRD)
showed the absence of lithium
compounds and the presence of only
oxides (Cr,Fe) 2 O 3 and (Fe,Cr) 3 O 4 , with
iron and chromium present as minor
elements in the first and second case,
respectively.
Fig. 3.36 – Scattered
electron image at 20X of
EUROFER97 specimen
exposed for 1000 h in
Li 2 TiO 3 pebble bed.
Fig. 3.37 – Backscattered
cross-section
image at 2500X.
Considering the thermodynamic conditions and the oxygen chemical potential, the
temperature of 600°C seems to be borderline for (Cr,Fe) 2 O 3 .
3.8.9 Li 2
TiO 3
pebble reprocessing; recovery of 6 Li as Li 2
CO 3
Lithium titanate is one of the most promising candidates for tritium breeding. The
temperature of tritium release from polycrystalline Li 2 TiO 3 ceramic pellets and
pebbles was found to be lower than from many other Li ceramics. This material also
shows good chemical stability in air and has acceptable mechanical strength.
A process for obtaining Li 2 CO 3 from Li 2 TiO 3 sintered pebbles by wet chemistry was
developed. This is considered useful in view of the recovery of the 6Li isotope from
lithium titanate breeder burned to its end of life in a fusion reactor. The process was
optimised with respect to the chemical attack of titanate by using an aqueous HNO 3
solution. The subsequent precipitation of lithium carbonate by Na 2 CO 3 produced a
powder with chemical and morphological characteristics suitable for its reexploitation
in the fabrication of Li 2 TiO 3 pebbles. Reprocessing was also planned to
adjust the 6 Li concentration to the desired value by using 6 Li-enriched LiOH*H 2 O
and to obtain its homogeneous distribution in the powder batch.
A specific procedure was used to add a number of small carbonate batches (each one
obtained from 40 g of starting pebbles) in order to produce a batch of about 400 g of

3. FUSION TECHNOLOGY 97
3.8 Liquid Metal Technology and
Hydrogen Effects on Materials
lithium carbonate (named LC-RT518). The XRD pattern of the final LC-RT518 was in
accordance with the ASTM JCPDS-831454 standards for monoclinic Li 2 CO 3 .
Table 3.IV - Comparison of ENEA lithium
carbonate and CEA carbonate sample
Li 2 CO 3 main CEA ENEA
characteristics sample LC-RT518
Bed density (g/cm 3 ) 0.56 0.45
True density (g/cm 3 ) 2.07 2.06
evaluated closed por. % 2.1 2.2
Spec. Surface area (m 2 /g) 0.90 0.97
Equival. Spher. Dia. (µm) 3.2 3.0
Some further characterisations
were performed on the final LC-
518 carbonate batch, such as true
density by He picnometry, and
specific surface area by nitrogen
adsorption through the classical
three-point Brunauer-Emitt-Teller
(BET) method. The results,
reported in table 3.IV, were
compared with parallel
measurements on a reference
carbonate sample from CEA.
[3.43] P. Lorenzetto,
Technical specification
for the thermal fatigue
tests of Be protected
EDA mock-ups, EFDA
/00-529, 19-11-2001
Fig. 3.38 -Mockup frame
in EDA-BETA.
Fig. 3.39 - CFC electric
radiative resistor.
3.9 Thermal-Fluidodynamics
3.9.1 Fatigue tests on six mockups of primary first-wall panel
prototype (EFDA Contracts 00/529 and 00/533)
The thermal fatigue testing of the first-wall components, i.e., six primary first-wall
mockups (EDA) and two panels, has been committed to ENEA [3.43]. The objective
is to perform thermal fatigue on beryllium armoured first-wall test sections. The
experimental campaigns will be performed at ENEA Brasimone at the CEF 1-2
thermal-hydraulic facility. For each test campaign, pairs of mockups or panels, to be
tested in parallel, are assembled inside two special vacuum vessels called EDA-BETA
(fig. 3.38) and THESIS. Special CFC electric radiative resistors (fig. 3.39) placed
between each pair of facing mockups provide heating at a nominal heat flux of 0.8
MW/m 2 with a period of 300 s. During the dwell phase, the fatigue stresses are
magnified by inlet cooling water
temperature from 120 to 20 °C.
During the tests, the thermalhydraulic
parameters (coolant,
heater and material
temperatures, water flow rate
and pressure) are also measured.
In 2001, the EDA-BEDA vacuum
chamber was appropriately
modified to provide the mockup
cooling and the electrical
feedthroughs. Preliminary
testing of the modified EDA-
BETA facility was also
performed. Six mockups were
delivered to ENEA Brasimone.
They are armoured with Be
grade S65C tiles with different
dimensions and thickness (5 or
10 mm). The heatsink is made of
DS-Cu (Glidcop-Al25) joined to a
stainless steel (AISI316 L) back
plate, both provided with water
cooling channels. At the end of

98
3. FUSION TECHNOLOGY
3.9 Thermal-Fluidodynamics
2001, the first experimental campaign started on four of the EDA mockups mounted
inside EDA-BETA [3.44, 3.45].
3.9.2 HE-FUS3 experimental cassette of lithium-beryllium pebble
beds
During 2001, a new thermal test campaign was started on the HELICHETTA solidbreeder
mockup. The objective of the tests on a single prismatic cell filled with the
reference breeder Li 4 SiO 4 and Li 2 TiO 3 pebble beds was to determine the influence
of the filling factor on the thermal-mechanical parameters and the behaviour of the
ceramic bed after mechanical pre-cycling and application of the spring-system lateral
load. From July 2001 to the end of the year, 60 tests were carried out in air on the
HEFUS-3 facility at ENEA Brasimone on both the reference materials. The measured
Li 4 SiO 4 and Li 2 TiO 3 pebble packing factors were, respectively, 0.65 and 0.64. The
results of the first HELICHETTA test campaigns are:
i) the displacement of the beds is as large as 0.2 mm towards the cooling plates and
ranges from 0.5 to 1 mm for L i 2TiO 3 and up to 1.5 mm for Li 4 SiO 4 towards the
sliding plug;
ii) the washer springs and sliding plug systems prevent larger stresses on the
containment structure;
iii) the pebble bed thermal conductivities in air show good agreement with previous
FZK experiments;
iv) the pebble thermal mechanical hysteresis, well evident during the cyclic ramp
up/down tests, affects the thermal-mechanical bed behaviour.
The tender for construction of both HELICA and HEXCALIBER was launched in
December 2001 and their fabrication will be finished, respectively, by June 2002 and
December 2002 [3.46, 3.47].
3.10 International Fusion Material
Irradiation Facility (IFMIF)
3.10.1 Design and mockup tests of lithium jet target
One of the main tasks of the lithium target design is to guarantee the jet stability
against overheating by the powerful deuteron beam. This is achieved with the use
of a curved plate (backplate) on which lithium flows. A computer code (RIGEL)
developed ad hoc by ENEA Bologna was used to determine the best working
conditions (table 3.V) for the new IFMIF design parameters (Reduced Cost Design).
[3.44] G. Dell’Orco et al.,
Status of the Contracts
EFDA 00/529 and
00/533 for the thermal
fatigue tests of Be
protected EDA – PFW
mock-ups, ENEA-EFDA
Meeting (Brasimone
2001)
[3.45] G. Dell’Orco et al.,
Report for the Task
T216+, subtask E1, on the
thermal fatigue tests of
Be protected first wall
mock-ups, ENEA Internal
Report, SB-G-R-0051
(2001)
[3.46] G. Dell’Orco et al.,
TAZZA mock-up pebble
beds - Experimental and
theoretical investigations,
presented at the
10th Int. Workshop on
Ceramic Breeder Blanket
Interactions (CCBI-10)
(Karlsruhe 2001)
[3.47] G. Dell’Orco et al.,
Progress on pebble bed
experimental activity for
the HE-FUS3 mock-ups,
presented at the 10th
Int. Workshop on
Ceramic Breeder Blanket
Interactions (CCBI-10)
(Karlsruhe 2001)
Tab 3.V - IFMIF target input data and Li jet stability results
Main input data
r
Main results
Beam footprint 5×20 [cm×cm] Reynolds number 694417 [-]
Jet thickness 0.025 [m] Max. pressure 12493 [Pa]
Jet width 0.26 [m] Max. surface temperature 297 [°C]
Jet velocity 15 [m/s] Max. temperature in the 441 [°C]
bulk
Backplate curvature 0.25 [m] Min. free surface boiling 35 [°C]
radius
margin
Inlet temperature 250 [°C] Min. bulk boiling margin 403 [°C]
Jet power deposition 10 [MW] Free surface evaporation 16 [g/year]

3. FUSION TECHNOLOGY 99
3.10 International Fusion Material
Irradiation Facility (IFMIF)
Fig. 3.40 - Boiling margins
for two values of
curvature radius R.
800
The results show that, for a 10-
MW beam power, the Li
700
surface temperature at the jet
outlet increases by about
600
50°C, and the corresponding
boiling margin is 35°C. That
500
is, the jet is thermally stable
against up to a 70% increase in
400 R = 250 cm
R = 474 cm
beam power. The jet bulk
stability is even better.
300
0 5 10 15 20 25
Furthermore, calculations
demonstrate that the boiling
Distance from lithium jet free surface (mm)
margins in the bulk do not
change appreciably by changing the curvature radius, as shown in figure 3.40.
Temperature (°C)
During 2001, ENEA started a co-operation with JAERI to monitor the lithium
experimental facility at OSAKA University against the risk of cavitation. The
cavitation noises will be detected by specific instruments and analysed by ENEA
CASBA equipment mounted on the existing pipe lines.
The experiments consist of a water jet mockup simulating the lithium jet. The jet will
flow through a double nozzle on a curved target back plate. The nozzle was designed
by means of the correlations derived from the Shima model. The design of the water
mockup assures some flexibility in selecting the target curvatures (250-450 mm), the
precision of the joint between the nozzle and the target back plate (± 0.1 mm, for a
total of five intermediate positions) and its surface roughness (1-10 mm).
Fig. 3.41 - General view of
the IFMIF target
assembly.
It has been established that a replaceable backplate is the optimal solution for the
IFMIF target design. This solution has been further developed into the so-called
bayonet concept in which the backplate is supported in a
sliding holder integrated with the target assembly
structure. This target concept facilitates removal and
installation operations when the backplate has to be
replaced, and they can be carried out without removal of
Backplate the test assembly. Figure 3.41 shows a general view of the
3-D model of the target assembly. The DRP facility at
ENEA Brasimone is suitable for the feasibility trials of the
remote handling operations needed to replace the IFMIF
back-plate.
Heavy manipulator
Support frame
Besides the design activities, a review of on-line methods
for impurity control and measurements in liquid Li was
performed in collaboration with the University of
Nottingham, UK to measure the corrosion rate on the
reference structural materials in well-defined testing
conditions.
3.10.2 System safety analysis and shielding calculations
Analyses of the safety of the lithium target and loop as well as shielding calculations
and a safety review of the whole IFMIF plant are in progress.
The failure mode and effect analysis (FMEA) approach was used to assess the
hazards related to the lithium-target operation. The main conclusions are that the
target of the reduced-cost plant fulfils the safety requirements with a negligible

100
3. FUSION TECHNOLOGY
3.10 International Fusion Material
Irradiation Facility (IFMIF)
environmental impact. Calculations with the RELAP5 Mod 3.2 code were performed
for the thermal transient analysis of the lithium loop, both in operational and in
accident conditions. The results show that IFMIF has good thermal-hydraulic
stability and a good agreement with the specified conditions for the facility at 10
MW. The study also pointed out that the hazards related to the accelerator are
confined within the plant boundaries, and concern mostly operator exposure to the
radiation induced by accelerator operation.
The IFMIF conceptual design was reviewed with regard to occupational radiation
exposure. After careful assessment and analysis of the doses associated with the
TRIUMF 520 MeV accelerator, it was concluded that the collective worker doses
could be relatively high.
Several key issues were identified and discussed to find possible solutions.
International radiation protection practices as well as the as-low-as-reasonably
achievable (ALARA) requirements were examined. The ALARA optimisation
process identifies the design improvements that could be made at a reasonable cost.
Reasonable cost is judged on the basis of the financial expenditure required to
achieve a unit reduction in dose. At an estimated facility dose of 1800 p-mSv/a (12
mSv/a is the average worker dose), IFMIF is way above current operating
experience.
With regard to the shielding and activation calculations in 2001, new computational
tools and data libraries were developed to define and establish design criteria and to
assess the radiological protection related to IFMIF beam-on and beam-off operational
phases.
A new intermediate energy coupled 256-neutron and 49 gamma-ray multi-group
cross-section library Vitenea-IEF, containing data for 37 materials/isotopes, was
produced. It was obtained by processing the evaluated files of the ENDF/B-VI
release 6 and the FZK nuclear data via the Njoy-Smiler-Ampx code systems. A new
file, containing the group-wise neutron and gamma fluence to ambient dose
equivalent conversion factors was developed for the dose-rate calculations.
Application tests to check the Vitenea-IEF library via the Scale-ENEA shielding
analysis sequence are in progress on the IFMIF configurations. A new activation code
package (Anita-IEAF) based on the Anita-2000 code (NEA-1638, RSICC CCC-606)
was developed to handle the numerous reaction channels for neutron energies over
20 MeV. The upgrade of the Anita code to let it manage the many reaction channels
now available in the activation library is in progress. Preliminary application tests of
the Anita-IEAF activation library are being performed for activation analysis of the
IFMIF test cell materials.
3.10.3 Development of fast neutron diagnostics
Work continued on the development of the IFMIF miniaturised (1.5 mm diameter)
fast on-line neutron monitors. Three prototype miniaturised fission chambers were
produced by CEA Cadarache in 2001: two chambers with fissile materials (237Np
and 238U) and one without fissile coating. These detectors, successfully tested at the
MINERVE thermal reactor at CEA, will be irradiated during 2002 at the cyclotronbased
Fast Neutron Facility (FNF) of the Nuclear Physics Institute, Rez (Czech
Republic). To produce an IFMIF-like neutron spectrum, the p(35 MeV)+D2O reaction
will be used as a source of high-energy neutrons. Although the technology for the
construction of the 1.5-mm-diam fission chambers foreseen for IFMIF is already
available, it was decided to build prototype fission chambers with larger diameter (8
mm) so that a sufficient signal level can be detected with the low neutron flux
(~3×10 12 n/s/sterad) provided by the cyclotron.

3. FUSION TECHNOLOGY 101
3.10 International Fusion Material
Irradiation Facility (IFMIF)
[3.48] S. Tosti et al.,
Testing of a catalytic
membrane reactor (CMR)
for decomposition of
tritiated water from
breeder blanket purge
gas in a closed loop pilot
plant, ENEA Internal
Report FUS TN BB-TS-
R-003 (2001)
[3.49] S. Tosti et al., Fus.
Eng. Des. 49–50, 953
(2000)
[3.50] S. Tosti et al,
Method of bonding thin
foils made of metal alloys
selectively permeable to
hydrogen, particularly
providing membrane
devices, and apparatus
for carrying out the
same, European Patent EP
1184125 A1 (2001)
[3.51] S. Tosti et al., Pd-
Ag membrane reactors
for water gas shift
reaction, to be published
in Chem. Eng. J.
[3.52] A. Basile et al,
Sep. Purif. Technol. 25,
549 (2001)
[3.53] S. Tosti et al.,
Characterization of thin
wall Pd-Ag rolled
membranes, to be
published in Int. J.
Hydrogen Energy Science
Fig. 3.42 - Measured
conversion values for the
water gas-shift reaction.
Two data acquisition systems for the tests at the cyclotron were developed at ENEA
Frascati: a) pulse mode based on a PCI card and LabVIEW software, with inputs for
8 channels and counting frequency up to 10 MHz; b) current mode based on CAMAC
modules and LabVIEW software. The neutron code SAND-II (for analysis of the
neutron activation data taken in the irradiation experiment) was implemented.
3.11.1 Tritium recovery from tritiated water
3.11 Fuel Cycle
The activities carried out concern the development, production, and testing of rolled
Pd-ceramic membranes and membrane reactors for the water gas-shift reaction
[3.48].
A pilot plant for testing membranes and membrane reactors in operating conditions
relevant to a “closed loop process” [3.49] for tritium recovery from tritiated water
was assembled.
The Pd-Ag rolled sheets were used to fabricate permeator tubes by a new welding
procedure. This new technique, an alternative to the inert gas tungsten arc welding
previously used, avoids the formation of thermal stressed zones along the permeator
tubes [3.50]. The rolled membranes were tested at 135-360°C with a hydrogen
transmembrane pressure in the range of 130-180 kPa and hydrogen flow rates up to
1.02×10 -4 mol s -1 . Complete hydrogen selectivity and good chemical and physical
stability were observed in long-term tests. The membrane reactors were tested at 325-
330°C, with a feed pressure of 100 kPa with reference to the water gas-shift reaction
CO+H 2 O⇔H 2 O+CO 2 .
High reaction conversion values in the range 95-99% (above the equilibrium value,
about 80%) for the water gas-shift reaction were measured, and the effects of the flow
rate and excess water in the feed stream were evaluated [3.51, 3.52]. The excess water
in the feed flow rate produces an increase in the reaction yield, according to the
theoretical analyses. Figure 3.42 reports three cases:
• equimolar feed ratio (CO=H 2 O);
• excess water (H 2 O=0.6, CO=0.4);
• excess water and the presence of a reaction product (H 2 O=0.3, CO=0.5, CO 2 =0.5).
The tests on the membrane reactors have demonstrated the applicability of
membrane technologies for the decomposition of tritiated water from breeder
blanket purge gas as well as for hydrogenation or dehydrogenation processes
involving the use or
Reaction conversion (%)
99
98
97
96
95
0
CO=H 2 O=0.5
CO=0.4 H 2 O=0.6
CO=0.2 H 2 O=0.3 CO 2 =0.5
4 . 10-5 8 . 10-5 1.10-4
CO feed flow rate (mol/s)
production of extremely
pure hydrogen.
In addition, the effect of
contaminants on the
interaction of hydrogen
gas with palladium and
the modification of the
membrane surface were
studied by characterising
the Pd-Ag rolled
membranes in long-term
tests [3.53].

102
3. FUSION TECHNOLOGY
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
3.12.1 Occupational radiation exposure assessment for ITER-FEAT
The first occupational radiation exposure (ORE) analysis for ITER-FEAT was
completed during 2001 [3.54]. The collective dose results provided in the previous
analysis were reviewed, and the final assessment was completed regarding hands-on
activities for maintaining, inspecting and/or replacing the following items:
blanket/limiter, electron cyclotron heating system, ion cyclotron heating system,
cryopumps, divertor cassettes, three loops of the tokamak cooling water system
(TCWS). Airborne tritium was also considered, together with the other radiological
sources already included in the previous study (neutron activation due to plasma
burning, activated corrosion products on the inner surface of the TCWS pipes and
components).
The collective dose results are: for the hands-on assistance activities at the tokamak
ports, 197 person-mSv/y; for the five loops of the TCWS (first-wall blanket, divertor
and neutron-beam injector), 105 person-mSv/y, about 21 person-mSv/y per loop
[3.55 (Vol VI)].
3.12.2 Validation of computer codes and models (EFDA Task SEA5)
Several simulation codes are used for the ITER safety analyses. The codes for treating
thermal-hydraulic phenomena, aerosol and neutron transport, materials activation
and the generation of activated corrosion products are validated by comparing them
with experimental results and codes already validated, by means of benchmarks.
Thermal-hydraulic phenomena
To validate the thermal-hydraulic computer codes, calculations were performed
[3.56] against a set of five experimental cases of loss of coolant in volumes at subatmospheric
pressure in the Inlet of Coolant Events (ICE) facility at the JAERI
laboratories in Japan. A further check was performed [3.57], [3.58] against a second
set of four test cases, to verify the behaviour of the codes in condensation and
evaporation phases. The codes involved in the campaign were the ISAS system
(linking ATHENA, for thermal-hydraulic transients, and INTRA, for containment
simulations) and the fast running code CONSEN. The comparison between blindtest,
experimental and post-test results was examined in detail to point out the main
differences.
The peak of pressure in the plasma chamber (figs. 3.43 and 3.44) and vacuum vessel,
the most important outcome of the accident analyses, is quite well matched in the
post-test calculations by both ISAS and CONSEN. In all the cases, they showed a
maximum deviation from the experiments within the range of 5%.
The overall results indicate the high accuracy of the ISAS tool in coupling the
different codes for accident analyses. CONSEN proved to be flexible and also
suitable in two-phase flow
conditions. Both the simulation
tools can be considered adequate
for the thermal-hydraulic
applications requested.
CONSEN [3.59] was also
implemented for validation
against the results of the French
EVITA facility (built by CEA,
Cadarache), with the same
satisfactory results.
Pressure (kPa)
700
600
500
400
300
200
100
Exper
Post
Pre
0
0 50 100 150 200
Time (s)
[3.55] ITER Generic Site
Safety Report, ITER
JCT (2001)
[3.56] M.T. Porfiri, P.
Meloni, ISAS post test
calculation for inlet of
coolant experiments,
FUS-TN-SA-SE-R-006
(2001)
[3.54] S. Sandri, EU
Task TW0-SEA2:
Personnel Safety ITER
Task No. G81TD10FE
(D453), FUS-TN-SA-
SE-R-013 (2001)
[3.57] M.T. Porfiri, P.
Meloni ISAS calculations
for inlet of coolant
(ICE) experiments in
2001: pre and post-tests
2a, 2b, 6 and 7, FUS-TN-
SA-SE-R-031 (2001)
[3.58] G. Caruso, M.T.
Porfiri,“CONSEN validation
against ICE –
Experimental campaign
2001 - Post-test calculations
for the cases 2a,
2b, 6 and 7-Post-test
calculations for the
cases 2a, 2b, 6 and 7,
FUS-TN-SA-SE-R-032
(2001)
[3.59] G. Caruso, M.T.
Porfiri, “CONSEN validation
against EVITA -
Experimental campaign
2001 - Post-test calculations”,
FUS-TN-SA-SE-
R-033 (2001)
Fig. 3.43 – Plasma
chamber pressure for ICE
case 6 (CONSEN).

3. FUSION TECHNOLOGY 103
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
800
Neutron transport and materials activation
Pressure (kPa)
600
400
200
ISAS pre-test
Experiment
ISAS post-test
The code package updating, completed in November
2000, was released to the OECD-NEA Data Bank.
Validation of the ANITA-2000 code package continued
[3.60, 3.61] by comparing calculations with the
experiments performed at the Fusion Neutronics Source
(FNS), JAERI, Tokai, Japan.
0
0 20 40 60 80 100
Time (s)
Fig. 3.44 – Plasma chamber pressure for ICE case
7 (ISAS).
The material samples were irradiated by a 14-MeV
neutron flux in two series lasting 5 min and 7 h,
respectively. The neutron energy spectrum and neutron
source intensity of the experimental irradiation, as well as
the sample compositions, were provided by JAERI. A 175
energy-level neutron flux distribution was considered.
[3.60] D.G. Cepraga, G.
Cambi, M. Frisoni,
ANITA-2000 activation
code package. Part II :
code validation, ENEA
FUS-TN-SA-SE-R-020
(2001)
[3.61] D.G. Cepraga et al.,
Decay heat estimate for
fusion relevant materials
based on EAF-99 and
FENDL/A-2 libraries in
comparison with FNS-
Jaeri experiments, EFF-
DOC-797, EFF/EAF
Monitoring Meeting,
NEA-OECD (Paris 2001)
The European Activation File EAF99, the Fusion Evaluated Nuclear Data Library
FENDL/A-2) and the decay data library for fusion applications FENDL/D-2 were
used.
Tables 3.VI and 3.VII summarise the experiment-calculation comparison, by range of
discrepancies, for the 5-m and the 7-h irradiation scenarios, respectively, and for both
activation libraries.
As a general conclusion, it can be observed that, for the experimental irradiation
scenario analysed, EAF99 generally provides a better agreement with the experiment
than FENDL/A-2.
Table 3.VI - Summary of calculation-experiment comparison (C-E)/E for
samples irradiated for 5 min.
(C-E)/E EAF99 FENDL/A-2
50 % B4C, BaCO 3 , Bi, Cr, Na 2 CO 3 , B4C, BaCO 3 , Bi, CaO, Cr,
SiO 2 , Y2O 3 Na 2 CO 3 , SiO 2 , Ta, Y2O 3
Table 3.VII - Summary of calculation-experiment comparison (C-E)/E for
samples irradiated for 7 h
(C-E)/E EAF99 FENDL/A-2
< 10% Co,Mn, Nb, NiCr, Ni, Re, S, SrCO 3 , Co, Mn, Nb, NiCr, Ni, Re, S, SrCO 3 ,
SS-304, Ti, Zr, Inconel, SS-316 SS-304, Ti, Zr, Inconel, SS-316
10 to 50
%
BaCO 3 , CaO, Fe, Mo, Na 2 CO 3 ,
SnO2, Ta, V,Y 2 O 3 , Cu
BaCO 3 , CaO, Fe, Mo, Na 2 CO 3 ,
SnO 2 , V, Y 2 O 3 , Cu
> 50 % Al, Bi, Cr, K 2 CO 3 , Pb Al, Bi, Cr, K 2 CO 3 , Pb, Ta

104
3. FUSION TECHNOLOGY
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
Activated corrosion products
The experimental corrosion tests carried out by CEA with the CORELE2 facility were
simulated by ENEA to validate the PACTITER code. The aim of the corrosion tests
was to determine the 316L(N)-IG stainless-steel release rates in the thermalhydraulics
and chemistry conditions envisaged for the ITER TCWS. The tubes for the
corrosion tests in the CORELE2 loop were irradiated in the OSIRIS reactor. Pre-test
calculations were carried out before the corrosion tests. The model also simulated the
irradiation of the tube by assuming a decay period equal to 110 days. The four
corrosion tests, assumed to last 10 days each, were simulated at temperatures of 100
and 150°C and coolant velocities of 1 and 5 m/s. The pre-test calculations gave good
results [3.62]. A simulation with the ANITA activation code confirmed that the results
were satisfactory.
Preliminary results of the experiments were made available in October 2001. A
problem in the PACTITER code related to element solubility at low temperature was
solved. The four corrosion tests were simulated using the modified PACTITER
version: for test 2 (150 °C and 5 m/s), a stainless-steel release rate of 35
mg/dm 2 /month was obtained. The release rates computed increased in conditions
of higher temperature and coolant velocity [3.63].
3.12.3 Plant safety assessment for ITER-FEAT
The following activities were performed in contribution to the ITER Generic Site
Safety Report (GSSR) [3.55].
Activation calculation support for safety analysis
The radiation transport and activation calculations were carried out as support to the
safety analyses. The activation calculation results [3.64] include the specific activity,
decay heat, contact dose, clearance index, list of isotopes at shutdown and dominant
isotopes vs. cooling time up to 1x10 6 years, for each material and all the ITER radial
zones. The nuclear heating of each zone was also obtained.
Uncertainty analyses related to the activation characteristics of relevant ITER invessel
components/materials were also carried out [3.65, 3.55 (Vol. III), 3.55 (Vol. V)].
The uncertainties were obtained with the use of the Fispact-99 code and based on the
nuclear activation data libraries EAF-99. Both cross-section and decay data errors
were taken into account. Calculations were performed for:
• First-wall heatsink outboard (Zone 39 – FWCUO): Cu and SS316;
• Back of blanket outboard (Zone 55 – BLBKO): SS316;
• Vacuum vessel front-wall outboard (Zone 56 – VVFWO): SS316.
[3.62] L. Di Pace, D.G.
Cepraga, CORELE 2 Pretest
calculations, ENEA
FUS-TN-SA-SE-R-011
(2001)
[3.63] L. Di Pace, D. G.
Cepraga, CORELE 2 Posttest
PACTITER calculations,
in preparation
[3.64] G. Cambi et al.,
“Anita-2000 activation
code package: clearance
assessment of ITER
activated materials”, in
preparation
[3.65] D.G. Cepraga et al.,
“Radiation transport and
activation calculation for
ITER GSSR: analysis
updates”, ENEA FUS TN
SA-SE-R 05A (2001)
[3.66] D.G. Cepraga et al.
“Impact of special
impurities on the ITER
outboard vacuum vessel
activation”, ENEA FUS
TN SA-SE-R 05B (2001)
Table 3.VIII shows the results related to zone 39 – FWCUO.
In addition, the impact of 1 to 10 ppm each of actinide and rare-earth impurities
(dysprosium, holmium, thorium and uranium) on clearance of the outboard
vacuum-vessel steel, plasma side and rear side was assessed [3.66].
Table 3.VIII – Uncertainty in first-wall outboard activities at plasma shutdown (SA1 operational scenario)
FW heatsink outboard (Cu)
FW heatsink outboard (SS316)
Isotope Activity [GBq/m 3 ] Uncertainty [%] isotope Activity [GBq/m 3 ] Uncertainty [%]
Total 2.93×10 09 3.4 Total 5.34×10 08 4.71

3. FUSION TECHNOLOGY 105
[3.67] G. Cambi, M.T.
Porfiri, P. Meloni, NAUA
model for accident
analyses for ITER GSSR,
ENEA FUS-TN-SA-SE-
R-007B (2001)
[3.68] M.T. Porfiri,
G.Cambi, P. Meloni,
Accident Safety
Analyses, for ITER GSSR
– Pump seizure in divertor
HTS, large divertor exvessel
coolant leak, ENEA
FUS-TN-SA-SE-R-007A
(2001)
[3.69] M.T. Porfiri,
G.Cambi, P. Meloni,
Additional accident
safety analysis for ITER
GSSR – Large DV exvessel
coolant leak, ENEA
FUS-TN-SA-SE-R-008A
(2001)
[3.70] M.T. Porfiri,
G.Cambi, P. Meloni,
Parametric accident
safety analysis for ITER
GSSR - A) Divertor exvessel
coolant leak during
baking, ENEA FUS-TN-
SA-SE-R-008B (2001)
[3.71] T. Pinna, C.
Rizzello, Failure mode
and effect analysis for
tritium systems of ITER
FEAT, ENEA FUS-TN-
SA-SE-R-017 (2001)
[3.72] T. Pinna, L.
Burgazzi, Failure mode
and effect analysis for
neutral beam injectors of
ITER FEAT, ENEA FUS-
TN-SA-SE-R-018 (2001)
[3.73] T. Pinna, L.
Burgazzi, Failure mode
and effect analysis on
ITER FEAT: last results
to be introduced on
GSSR Vol.X., neutral
beam Injectors, Magnets,
ENEA FUS-TN-SA-
SE-R-016 (2001)
Fig. 3.45 – Ex-vessel DV
coolant leak.
Deterministic accident analyses
The design basis accidents relating to pump seizure and large divertor (DV) coolant
leak were assessed for the ITER-FEAT design. The NAUA nodalizations [3.67] were
updated, and the model for the wet aerosol deposition was implemented. The
accident analyses for the GSSR were iterated twice to optimise safety system
operation and evaluate critical parameters such as containment pressure. Different
boundary conditions were set for the first [3.68] and second [3.69] analyses:
a) The set point for the bleed lines opening towards the drain tank and the
suppression tank (110 kPa in the first assessment and 80 kPa in the second), to
maintain the vacuum vessel in low pressure conditions as long as possible.
b) The position of an ex-vessel break in the DV cooling loop, to maximise the vault
pressurisation for fluid discharge.
For pump seizure, the leakage from the vacuum vessel results in the following
environmental releases: about 0.18 mg of tritium and about 88 µg of tungsten dust in
the first accident analysis. In the second iteration, about 0.12 mg of tritium and about
250 µg of tungsten dust were calculated. The combined releases remain four orders
of magnitude below the accident release guidelines.
For the first analysis of the ex-vessel DV leak, the total environmental release of
tritium was 66 mg-T. The releases of dust and activated corrosion products (ACPs)
totalled 24 and 270 mg, respectively. In the second iteration, the total environmental
release of tritium was 685 mg-T, and the total releases of dust and ACPs were 20 and
600 mg, respectively. Even considering the worst accident conditions, which are
more severe in this second scenario, the combined releases are about a factor of 3
below the accident release guidelines.
An ex-vessel DV coolant leak was assessed in baking conditions [3.70] to estimate the
maximum vault pressurisation due to high coolant temperature (240 °C). Figure 3.45
shows the trend of the pressurisation in this case. The design pressure of 200 kPa is
never reached [3.55 (Vol. VII)].
Probabilistic accident analysis
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
A component-level FMEA was used to identify postulated initiating events (PIEs)
and possible accident consequences for the tritium systems [3.71], neutral beam
injectors [3.72] and magnet systems [3.73] of the ITER-FEAT reactor. The tokamak
and exhaust processing (TEP) system, storage and delivery system (SDS) and isotope
separation systems (ISS) were assessed for the fuel cycle.
(Pa)
2
1.6
1.2
1
0.8
0.4
0
TCWS vault pressure
0 10 1000 100000
Time (s)
The FMEA table is structured so
that the following data are
reported for each component: all
the possible failure modes that
could occur in the different
operating states; the related
accident frequencies and category
classification; causes of failure and
the preventive actions; consequences
and the preventive/
mitigating actions; the PIEs as
identified by safety-relevant
elementary failure modes and
pertinent comments.

106
3. FUSION TECHNOLOGY
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
Component failures that could induce a PIE were listed, and the total frequency for
each PIE was determined. The accident sequences following a PIE were qualitatively
defined. All elementary failures without safety-relevant consequences were
classified as a Not Safety Relevant (N/S) PIE.
The overall work is reported in [3.55 (Vol. X)].
Corrosion product modelling and inventories
The divertor/limiter (DV/LIM) cooling loop was modelled with regard to geometry
and thermal-hydraulics. The PACTITER code was utilised to determine the ACP
inventory. The results of the neutron transport and activation analyses were
incorporated in the PACTITER input. To represent the ITER operation, two scenarios,
SA1_acp and M-DRG1, were simulated. The SA1 fluence was 0.5 MWy/m 2 for a
total of 320 days of burn at an average n-wall load of 0.57 MW/m 2 . The M-DRG1
activation scenario (192 days of plasma burn in about 20 years, including two 6-day
campaigns at 25% duty cycle) fluence was 0.3 MWy/m 2 .
The operative phases considered in the two scenarios were burn, dwell, cold and hot
standby, baking and decontamination.
The ACP deposit mass was 8882 g in the SA1 scenario and 7808 g in the M-DRG1
scenario. A large reduction in the ACP inventory in terms of mass and activity was
recorded, compared to the previous ACP assessment for the ITER-FDR DIV loop
(1998 design).
The Cu-alloy release rate strongly depends on the different operating phases. It is
enhanced during baking because of the greater solubility of Cu at 240°C, unlike
stainless steels, where a lower release rate occurs at 240°C. During burn phases, the
release rate increases for all materials because of the temperature gradient in the
loop. The Cu-alloy release rates are of the order of 10 µm/y for baking and burn
periods, while they are extremely low during the other periods (0.01-0.1 µm/y)
[3.74].
Accident sequence analysis for tritium systems
The accident initiators identified by means of the probabilistic assessment were
studied to determine possible accident scenarios. Success and/or failure of the
systems implementing the required safety functions were considered [3.75].
[3.74] L. Di Pace,
Activated Corrosion
Products E valuation for
the ITER TCVS DIV/LIM
Loop, FUS-TN-SA-SE-R-
014 (2001)
[3.75] C. Rizzello, T.
Pinna, Accident sequences
analysis related
to ITER FEAT fuel cycle
systems, ENEA FUS-TN-
SA-SE-R-004 (2001)
The principle of the “defence in depth” is implemented in tritium systems by the use
of successive barriers preventing the release of radioactive materials to the
environment. The rationale is to limit the dilution of tritium within each barrier
system, so that it can be recovered before penetrating to the next barrier.
The environments are all equipped with devices capable of automatically isolating
the ventilation systems and switching on the detritiation systems in the case of
accidental release of tritium.
The effectiveness of the fuel cycle system isolation and of the containment barriers
was assessed by evaluating the amount of tritium released from process equipment,
for the reference accidents.
As an example, for the break of a tritium process line inside the glove box (GB)
containment of the tokamak exhaust processing system, the bounding accident
conditions were found for the failure of a fuelling surge tank, which results in 14.5 g
of tritium lost into the GB. Although the GB ventilation is isolated and the

3. FUSION TECHNOLOGY 107
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
detritiation system activated, until the GB tritium decontamination is over, a small
amount of tritium can escape from the GB towards the operating area, mainly due to
permeation through the gloves and to the environment through the room ventilation
system. For such a sequence, the maximum tritium dispersion to the environment
was estimated to be 0.16 mg, well below the design release limit of 1 g.
The “ultimate safety margins” of the plant were also assessed to confirm that a
further degradation of systems would not lead to cliff-edge effects. Included were
some possible aggravating factors, i.e., lack of isolation of the
broken/malfunctioning system/component, consequential failure of nearby
systems, or H 2 -air reactions in the form of a fire or explosion inside operating areas.
The work demonstrates that tritium releases to the environment are below the design
limits for the overall accident conditions and that the no-evacuation goal of ITER-
FEAT is attained even if the ultimate safety margins are challenged.
[3.76] A. Natalizio, L. Di
Pace, Review of the
current methods for the
management of tritiated
waste, ENEA FUS-TN-
SA-SE-R-001 (2001)
[3.77] M. Zucchetti,
Clearance of activated
materials: the ‘de
minimis’ problem, ENEA
FUS-TN-SA-SE-R-021
(2001)
[3.78] M. Zucchetti, R.
Forrest, L. Di Pace,
Clearance of activated
materials: optimisation
of ex-vessel material
composition, ENEA FUS-
TN-SA-SE-R-022 (2001)
[3.79] D. G Cepraga, G.
Cambi, M. Frisoni,
Neutronic calculations
for SEAFP-2 plant
models 2 and 3, ENEA
FUS-TN-SA-SE-R-024
(2001)
3.12.4 Waste management
Three different studies were performed for this issue.
Review of current methods of tritiated waste management [3.76]
The Canadian radioactive waste management experience was reviewed because of
its potential relevance for fusion reactor studies. In fact, tritium is the element
expected to be common to both fusion and CANDU reactor waste.
The operation and final decommissioning of a fusion power reactor will generate
significant quantities of solid and liquid waste. Whereas some of the waste will be
similar to CANDU reactor waste (e.g., ion-exchange resin, tritiated and activated
cooling system components, tritiated steel, etc.), other waste will be unique to fusion
(e.g., activated and tritiated beryllium, tungsten and silicon carbide).
Clearance of activated materials: the “de minimis” problem [3.77] and
optimisation of ex-vessel material composition [3.78]
The minimisation of active waste from operation and decommissioning of a fusion
power plant has to be one of the main objectives for fusion waste management
studies. Clearance is one of the ways to achieve this goal.
The problem of defining operative clearance levels was discussed in the study. Two
main options were proposed: clearance for non-active disposal or free-release
recycling. The limits proposed were derived from clearance levels for radionuclides
in solid materials (IAEA-TECDOC-855, 1996) and adopted for the “disposal as nonactive
waste” option. The limits indicated by the European Commission Radiation
Protection 89 were taken into account.
Furthermore, a possible optimisation of ex-vessel component composition for the
SEAFP Plant Models 2 and 3 was studied in order to enable their clearance,
considering both options cited above. The study pointed out that the clearance goal
could be reached for many materials without any optimisation. The ex-vessel
components that do not reach the clearance levels are inboard winding pack
(magnets), inboard insulator and the inboard vessel itself. This is valid for both plant
models, even though some differences exist between the two.
Neutronic calculations for SEAFP-2 Plant Models 2 and 3 [3.79]
The activation calculations for assessment of the composition optimisation of fusion
reactor ex-vessel material components were performed by the FISPACT code

108
3. FUSION TECHNOLOGY
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
[TR-022]. The neutron flux spectra for the ex-vessel zones were provided for the
activation calculation with the FISPACT activation code, in FISPACT format. To get
this data, it was necessary to carry out a neutronic (neutron transport) calculation.
The input data (geometry, material data, irradiation characteristics) were defined in
the framework of the SEAFP project.
The neutron (and gamma) flux distributions were obtained [3.80] by means of the
Bonami-XSDNRPM Sn coupled n-g 1-D discrete ordinates transport calculation
sequence from the SCALE-4.4 computer code system. The Vitamin-ENEA Master
Library (174n-38g groups), based on ENDF/B-VI data, was used for the transport
calculation.
3.12.5 Power Plant Conceptual Study
In the framework of studies for future fusion power reactors, three issues were
pursued: occupational radiation exposure (ORE), radioactive waste and recycling
and accident-sequence analysis.
Occupational radiation exposure
In Stage II of the Power Plant Conceptual Study (PPCS), the validity of the public
dose target and the ensuing environmental release targets, as set out in the General
Design Requirements Document (GDRD), were confirmed [3.80].
A centralised refurbishing facility servicing several power plants offers good
economics but requires the transport of activated and contaminated reactor
components on public right-of-ways. Alternatively, a co-located refurbishing facility
eliminates the need for public transport, but offers poor economics. It is envisaged
that the first few fusion power reactors could be constructed on a site adjacent to the
refurbishing facility.
[3.80] A. Natalizio, L. Di
Pace, Assessment of
GDRD safety requirements
in normal
operations (effluents and
occupational doses),
ENEA FUS-TN-SA-SE-
R-023 (2001)
The key issues related to ORE include the selection of an appropriate station dose
target, a longer plasma-facing-component (PFC) life, short special maintenance
outages, low-speed pellet injectors and a tubeless blanket.
The station dose target, fixed in GDRD Stage I at 700 p-mSv/a per GWe, could be
modified to 700 p-mSv/a per station unit, irrespective of the net electrical power
output.
The design life of PFCs determines the reactor cycle – the time between in-vesselcomponent
replacement outages. The reactor cycle can be increased in two ways: by
increasing the performance of the PFCs from the GDRD value of 2 FPY (full power
year) and/or reducing the neutron wall loading. The PFC performance can only be
increased by utilising high-performance materials. The neutron wall loading,
however, can be reduced by changing the tokamak design parameters, i.e., making
the tokamak bigger. Physics parameters and magnet technology permitting, the net
result of a bigger tokamak would be a higher capital cost, but the net result of the
increased reactor cycle could be lower annualized, unit electricity cost and worker
doses.
There is a strong correlation between plant unavailability and station dose in nuclear
plants – the higher the unavailability, the higher the dose. The special maintenance
outage in fusion plants adds to plant unavailability and therefore increases station
dose. The GDRD special maintenance outage of six months could potentially
increase normal station dose by an estimated 300-2,000 p-mSv/a. Such a potentially
large dose underscores the importance of reducing the outage duration.

3. FUSION TECHNOLOGY 109
The cooling tubes inside the breeding blanket will be subjected to neutron-induced
sputtering, which is the primary contributor to cooling system maintenance doses.
The total cooling system dose was estimated to be of the order of 290 p-mSv/a.
Therefore, a self-cooled liquid-metal blanket, for example, which eliminates the need
for cooling tubes, would significantly reduce cooling system maintenance doses.
Waste management
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
[3.81] A. Natalizio, L. Di
Pace, Assessment of
clearance and recycling
from the policy point of
view, in preparation
[3.82] A. Natalizio, L. Di
Pace, Tritium transport
and proliferation issues,
in preparation
Assessment of clearance and recycling from the policy viewpoint [3.81]. A valid
analogy exists between in-vessel components that need to be replaced on a regular
basis and used fission reactor fuel, even if there are significant differences in the type
of waste and radiotoxicity. The fission power industry has followed two basic
strategies for used fuel disposal: the once-through fuel cycle and the closed-fuel
cycle, which includes reprocessing of the used fuel. These two strategies are also
available to the future fusion power industry. The key question is to determine the
technical and economic feasibility of used, in-vessel component (IVC) refurbishment
and reprocessing. Seven scenarios were developed and studied to identify the factors
that are important in the development of a suitable fusion-power waste management
strategy dealing with the interim storage, refurbishment/reprocessing, and final
disposal of waste. The scenarios studied range from doing very little
refurbishment/reprocessing to doing the maximum refurbishing/reprocessing that
is practical, both on-site and off-site. The key criterion in evaluating the various
scenarios was the environmental acceptability of the fusion power plant. More
specifically, the aim was to identify scenarios that would eliminate or reduce the
shipment of radioactive materials to and from centralised fuel reprocessing facilities.
The following conclusions were drawn from the study:
1. Future fusion power plants should be constructed as multi-unit plants with
adjacent in-vessel component refurbishing facilities.
2. Tritium recovery from blanket modules is expected to reduce the cost of IVC
refurbishment.
3. Reprocessing the breeder and neutron multiplier material may not be economical
unless the reprocessing unit cost is significantly lower than a few hundred Euro/kg.
4. Without IVC refurbishing, the operational IVC waste volume will far exceed the
decommissioning waste volume.
Tritium transport and proliferation issues [3.82]. The objective was to analyse the
tritium transport and proliferation issues that could arise from the transport of
considerable quantities of tritium to and from future fusion power plants. These
potential issues were addressed in the context of a mature fusion power industry.
Two aspects were considered: the public safety of tritium shipments (i.e., the
potential for radiation exposure in the event of a shipping accident); and the
proliferation aspects of tritium shipment (i.e., the potential for the hijacking of
tritium shipments by terrorist organisations). Based on simple analyses it was
demonstrated that there will be a need to ship significant quantities of tritium, to and
from a future fusion power plant, on an annual basis. This could be of the order of
10 kg per year.
Tritium has been safely shipped in licensed shipping containers for many years and
without incidents that could constitute a public risk. Current shipping containers in
Canada are designed and licensed to transport 50 g of tritium, but larger containers
have also been considered in past studies, for example for ITER. Considering the
large quantities of tritium to be shipped to and from a future fusion power plant, a

110
3. FUSION TECHNOLOGY
3.12 Safety and Environment, Power
Plant Studies and Socio-Economics
container of 1000-g capacity would not be considered unreasonable to limit the
number of shipments and hence the risk of possible hijacking by terrorists.
Tritium (and other fusion materials) is excluded from the Non-Proliferation Treaty;
nevertheless, tritium can be shipped across national borders only if there exists a
nuclear co-operation agreement between the countries involved.
Accident sequences analysis
Accident analyses have to verify that the future reactors concepts do no represent a
safety concern. A list of the data necessary for the accident analyses in the PPCS was
prepared [3.83]. The main scope of the work was to have a common database for the
analysts working on accident analyses, to avoid incongruent final results due to the
different assumptions and parameters used.
[3.83] M.T. Porfiri,
Required data for the
accident analyses in
power plant conceptual
study assessment, ENEA
FUS-TN-SA-SE-R-027
(2001)
3.12.6 European ITER site at Cadarache
In 2001, EFDA charged the European ITER Site Study Group (EISSG) with the tasks
of verifying whether the ITER Site Requirements and Assumptions could be met at
Cadarache and studying and estimating any necessary adaptation works. ENEA’s
Fusion Division acted as overall co-ordinator of the work and also provided
substantial support in the following fields:
Safety and Licensing. The scope of the task was to verify whether the ALARA
approach had been implemented in the ORE assessments for ITER and that it was
compatible with French regulations.
Socio-Economy. The Socio-Economic Research on Fusion (SERF) already performed
for the ITER Generic Site was reviewed for adaptation to Cadarache. The problem of
public awareness and acceptance of fusion was investigated.
Electrical Power Supply. The effect of the ITER electrical load on the Cadarache local
network was studied. The ITER power supply design modifications were identified
and evaluated, and improvements to the ITER design were suggested. ENEA was
also responsible for co-ordinating the work of other laboratories (CEA, RFX) and of
European industries (e.g., European Fusion Engineering & Technology).

4. MISCELLANEOUS 113
4.1 Development of CVD Diamond
Detectors for Nuclear Radiation
Diamond detectors are of particular interest as neutron detectors in fusion
environments since they present higher radiation resistance than silicon detectors. In
collaboration with the Faculty of Engineering of Tor Vergata University in Rome,
diamond films produced with the chemical vapour deposition (CVD) method are
being developed and their characteristics analysed for nuclear detection.
During 2001, several new samples of CVD diamond were grown and tested with
nuclear particles (alpha particles and electrons) to investigate important parameters
such as grain dimensions, film purity and lattice properties [4.1].
[4.1] M. Marinelli et al.,
Phys. Rev. B, 64,
195205–1 (2001)
[4.2.] R. Bernabei et al.,
Eur. Phys. J. Direct, C11,
1 (2001)
[4.3] G. Barbiellini et al.,
CP587, GAMMA 2001:
G a m m a - R a y
Astrophysics, pag 774
(2001)
The film quality was studied as a function of the growing parameters (methane
concentration, substrate temperature, chamber volume, film thickness) to establish
whether or not it was possible to optimise the quality of diamond films. Alpha
particles of different energy were used to study the film properties. Hence, by
considering the penetration depth at each energy, it was possible to define the grain
size and the collection length. The present quality of CVD films grown at Tor Vergata
University in collaboration with ENEA represents the state-of-the-art for these
materials in terms of charge collection efficiency (70%). The work pointed out that,
to improve detection efficiency, it is necessary to improve film purity.
4.2 Light Response of a Pure
Liquid Xenon Scintillator
The search for dark matter is one of the most stimulating fields in fundamental
physics. To establish in which form the so-called “missing mass of universe” does
exist represents a result of paramount importance to understanding the structure of
the universe. Several experiments are being carried out world-wide to detect
particles that are candidates as “dark matter”. One of these experiments, actually
named Dark Matter (DAMA), is being conducted by the Italian Institute for Nuclear
Physics (INFN) at the “Gran Sasso” underground laboratory and is devoted to
searching for the so-called Weak Interacting Massive Particle (WIMP). The DAMA
experiment makes use of different detectors (liquid xenon and NaI) and is based
upon the search of the recoil spectra produced by WIMPs when interacting with the
detectors. It is very important to calibrate the detectors in the energy range where
the recoil spectra are expected. In particular, the so-called quenching of the
scintillator has to be known to properly measure the recoil spectrum. Kinematics
calculations show that the quenching factor can be well reproduced if high-energy
neutrons are used.
Based on this theoretical finding, the liquid xenon detector was calibrated at the
Frascati Neutron Generator (FNG) with the use of 2.5-MeV neutrons. Results of the
calibration and details of the method are reported in [4.2]. The main output of the
calibration campaign was the ratio of the measured amount of light from the xenon
recoil nucleus to the amount of light from an electron of the same kinetic energy
(quenching). Results substantially in agreement with the previous determination
were obtained.
4.3 Partecipation in the AGILE Project:
Collimator and Coded Mask of the
SuperAGILE Detector
AGILE (Astrorivelatore Gamma ad Immagini Leggero) [4.3] is the first mission of the
Small Missions Programme of the Italian Space Agency (ASI). Its main goal is to
monitor the gamma-ray sky in the energy range 30-50 GeV, with a large field of view
(~3 sr), good sensitivity, good angular resolution and good timing. The satellite is

114
4. MISCELLANEOUS
4.3 Partecipation in the AGILE Project:
Collimator and Coded Mask of the
SuperAGILE Detector
presently under construction and is
scheduled for launch in 2004 in an
equatorial orbit, for a lifetime of two
years. SuperAGILE is the x-ray
monitor added on top of the gammaray
tracker. It will have a large field of
view, providing hard x-ray imaging
thanks to its division into four
mutually orthogonal detectors, each
one coupled to a 1-D coded mask
through a collimator. SuperAGILE
will enable the study of a large variety
of cosmic x-ray sources, including gamma-ray bursts, persistent and transient
galactic x-ray sources, as well as many of the brightest extragalactic sources [4.4].
Prediction of the expected x-ray background is of paramount importance in the
design of the sensitivity and performance of a scientific spacecraft [4.5]. ENEA
contributed to the design of the SuperAGILE masks and collimators through a
Monte Carlo study that has optimised the noise-to-signal ratio and the detector
response.
The technical realisation of the coded mask implied the development of
manufacturing technologies not present in Italy. The task was performed by the firm
of Vaiarelli Milan, a subcontractor of Laben and Oerlikon/Contraves. Figure 4.1
shows one of the alpaca-mask samples made during research to determine the
correct etching flux.
4.4 Advanced Superconducting
Materials and Devices
The experimental activities were focused on transport-property optimisation of
YBCO thick films on metallic substrates. The structural, morphological and transport
properties were characterised as a function of film thickness, and different metallic
substrates, buffer layer architectures and deposition techniques were realised. Two
approaches were followed: the RABiTS [4.6] technique and the inclined substrate
deposition (ISD) technique [4.7]. Nickel-tungsten alloys [4.8] were developed to
obtain a suitable RABiTS substrate for the YBCO-coated conductor fabrication [4.9],
and ISD-CeO 2 film deposition was studied to provide a biaxially textured buffer
layer on polycrystalline metallic substrates [4.10]. The discovery of
superconductivity in MgB 2 compound [4.11] aroused great interest in this new
material; hence, an activity was started to study MgB 2 thin-film deposition on singlecrystal
substrates.
4.4.1 Ni-W based architectures: preliminary results
During 2001 the research on metallic substrates for YBCO-film deposition led to the
fabrication of a new Ni 1-x -W x alloy. Substrates with different tungsten atomic
percent concentration (x) were analysed in terms of magnetic properties, revealing a
Curie temperature linear decrease with x, leading to x=5 for temperatures around
350°C.
The structural and morphological properties of Ni 95 W 5 (indicated as Ni-W)
substrates were also studied. The results were better than in the case of other
Fig. 4.1 - Alpaca-mask
sample made during the
research work.
[4.4] G. Barbiellini,
CP587, GAMMA 2001:
Gamma-Ray Astrophysics,
pag 729 (2001)
[4.5] I. Lapshov et al,
CP587, GAMMA 2001:
Gamma-Ray Astrophysics,
pag 769 (2001)
[4.6] A. Goyal et al., Appl
Phys. Lett. 69, 1795
(1996)
[4.7] K. Hasegawa, et al.,
Appl. Supercond. 4 (10-
11), 487 (1996)
[4.8] E.Varesi et al.
Biaxial texturing of Ni
alloy substrates for YBCO
coated conductors, to be
published in Physica C
[4.9] E. Varesi et al.,
Pulsed laser deposition of
high critical current
density YBa 2 Cu 3 O 7 -
y / C e O 2 / N i - W
architecture for coated
conductors applications,
in preparation
[4.10] A. Mancini et al.,
Inclined substrate deposited
CeO 2 films by
electron beam evaporation
on randomly oriented
metallic substrate, in
preparation
[4.11] J. Nagamatsu et al.,
Nature 410, 63 (2001)

4. MISCELLANEOUS 115
4.4 Advanced Superconducting
Materials and Devices
Fig. 4.3 - Polar figures achieved on the (113)YBCO, (111)CeO 2 and
(111)Ni-W peaks. A sharp texture with circular symmetry poles can be
seen.
Ni–based alloys, such as as Ni 89 V 11 , Ni 88 Cr 12 (Ni-V, Ni-Cr), and of pure
Ni as well.
The Ni-W substrates were thermo-mechanically treated with a standard
process that had previously been experimented on different Ni alloys.
Structural analysis of the as-treated Ni-W substrates showed a well-defined
cubic texture with sharp poles (typical w and j-scan FWHM of about 5.5°
and 7°, respectively) and a reduction in the presence of twinned grains and
spurious orientation grain contribution compared to Ni-V and Ni-Cr alloys
(figs. 4.2 and 4.3). Scanning electron microscopy (SEM) investigation of the
substrate surface morphology revealed a lower prominence of grain
boundaries (fig. 4.2).
Fig. 4.2 - Comparison
between surface
morphology of (a) Ni-V,
(b) Ni-Cr and (c) Ni-W.
For Ni-W, the grain
boundaries are less
prominent and there is a
lower percentage of twins
and spurious grains.
4.4.2 Influence of the substrate on YBCO-film transport properties
The YBCO films deposited on Ni-W buffered CeO 2 substrates showed J c values
greater than 1 MA/cm 2 at 77 K in the case of thinner films (250 nm) (fig. 4.4). On
the other hand, 1-mm-thick, 4-cm-long samples exhibited J c values of 6×10 5 A/cm 2
and 1.4×10 6 A/cm 2 at 77 and 65 K, respectively. The critical current I c achieved on a
strip of 3.5 mm with the length scaled to 1 cm was 57 A at 77 K and 140 A at 65 K.
Another remarkable advantage of the Ni-W substrate is the possibility to deposit
single buffer layer architecture, since the substrate exhibits enhanced oxidation
resistance with respect to pure Ni and other Ni alloys such as Ni-V.
Fig. 4.4 - J c dependence
on external magnetic field
for a 220-nm-thick YBCO
film deposited on
CeO 2 /Ni-W architecture.
Critical current density (A/cm2)
106
105
104
103
0
77 K
10
1
0.1
1 2 3 4 5 6
Magnetic Field (tesla)
Critical current (A)
4.4.3 Inclined substrate
deposition of CeO 2
films
on randomly oriented
metallic substrate
CeO 2 film growth by ISD on
randomly oriented metallic
substrate was studied to develop
a biaxially aligned buffer layer
for YBa 2 Cu 3 O 7-δ (YBCO) coated
conductors. CeO 2 films were
deposited by electron beam
evaporation. The deposition
system was equipped with a

116
4. MISCELLANEOUS
4.4 Advanced Superconducting
Materials and Devices
freely rotating sample holder to incline the substrate normal with respect to the
vapour incidence direction of an angle α.
At normal incidence (α=0°), the CeO 2 film showed a fibre texture, with the [111]
direction normal to the substrate and no in-plane texture. On inclining the substrate,
a biaxial textured growth of CeO 2 films is induced (fig. 4.5). The [00l] axis is tilted
relative to the substrate normal n of an angle γ in the opposite direction with respect
to the incidence of the vapour flux. The (200) and (020) poles, at the same χ angle, are
well defined, indicating a high degree of texture. The φ-scan FWHM values decrease
rapidly on increasing α from 15° to 45°, and become almost constant between 45° and
75° (fig. 4.6a). The film thickness has an important influence on the degree of texture:
films more than 1-µm thick show a sharp texture, as can be seen from the φ–scan
FWHM value reported in figure 4.6b.
Qualitatively, the texturing of ISD-CeO 2 film can be explained by considering the
anisotropic growth rate of crystal planes and the anisotropic diffusion along different
crystal directions. In general, films preferentially grow with the fast growing plane
perpendicular to the vapour flux. Typically in fcc materials, such as CeO 2 , the fast
growing plane is the close-packed {111} plane. In the inclined configuration, the same
growth mechanism is present and the {111} planes are the top planes of the column.
These planes are not exactly orthogonal to the vapour incidence direction because of
directional diffusion, which is due to the momentum conservation of the adsorbed
atom parallel to the film surface. Directional diffusion is also responsible for CeO 2
in-plane alignment. The grains that present the crystallographic direction with the
highest diffusion rate aligned with the directional diffusion can grow more than the
other grains, due to a higher mass transport effect. In ISD-CeO 2 films, this direction
coincides with the direction along the {111} planes. The bigger grains mask the
other grains, hence promoting a selection of film orientation; texture improves with
film thickness.
Fig. 4.5 - X-ray (002)
CeO 2 pole figures for
1.5-mm-thick film
deposited at α=55° and
Tsub=200°C in vacuum.
The arrow indicates the γ
angle; the ×, the
deposition direction.
φ-scan FWHM (degrees)
40
30
20
10
10 20 30 40 50 60 70 80
α (degrees)
FWHM
γ
a) b)
90
FWHM 90
40
γ
60
γ (degrees)
φ-scan FWHM (degrees)
30
20
10
0 0,5 1 1,5 2
thickness (µm)
60
γ (degrees)
Fig. 4.6 - φ-scan FWHM
and γ angle values for
CeO 2 film deposited at
Tsub=200°C in vacuum a)
vs. α value (1.5-µm-thick
samples) and b) vs. film
thickness (samples
deposited at α=55°).
4.4.4 MgB 2
film fabrication
Different fabrication techniques were used for MgB 2 films on single crystal
substrates. Two main methods were followed: as grown, performed by pulsed laser
deposition (PLD) and in situ annealing, performed both by PLD and by electron
beam evaporation.
In the as-grown method, the MgB 2 film is deposited directly on the single-crystal
substrate, which is heated at a certain deposition temperature in an inert gas
atmosphere.

4. MISCELLANEOUS 117
4.4 Advanced Superconducting
Materials and Devices
Fig. 4.7 - Resistance vs. temperature dependence
for an in situ annealed B/Mg/B trilayer precursor
structure annealed at 630°C for 10 min. The
transition region is magnified in the inset.
Fig. 4.8 - Surface morphology of a B/Mg/B trilayer
annealed at 630°C for 10 min, showing polygonalshaped
grains with a mean grain dimension of about
50 nm.
The in-situ annealing procedure consists of a two-step process. It is performed
without vacuum breaking in a deposition chamber in which a precursor structure is
deposited and then heated to the annealing temperature, at which it is maintained in
inert gas pressure for a certain time. The film is then cooled to room temperature.
Due to high Mg volatility and oxidation, it was difficult to obtain the correct Mg:B
stoichiometry in the deposited films. The highest temperatures at the beginning and
end of the superconducting transition T Conse t and T C0 were, respectively, 33 K and
31.5 K (fig. 4.7), obtained for an in situ annealed sample in which the precursor
structure was a multilayer B/Mg/B architecture deposited by the e-beam technique.
Morphological analysis showed a polygonal-shaped-grain film surface, with a mean
roughness value of about 50 nm (fig. 4.8).
4.5 Optical Metrology Survey
Fig. 4.9 - Alignment of
Elettra Synchrotron
Trieste.
In August 2001, the Fusion Technology Division carried out a 3-D survey of the
storage ring of the Elettra Synchrotron at Trieste, using the Division’s own optical
metrology system (Leica laser tracker). Within the two-weeks’ time window, more
than 2300 measurements
were taken on the 24
bending magnets and 360
quadrupoles and sextupoles
placed in the 260
m circumference storage
ring. The overall root
mean square error of the
oriented network was
less than 100 µm, so it
was possible to
successfully align one
fourth of the synchrotron
(October 2001). This
performance completely
fulfilled the very
demanding survey

118
4. MISCELLANEOUS
4.5 Optical Metrology Survey
specifications in terms of accuracy, time required, and usability of the measurements.
Figure 4.9 shows the metrology apparatus.
4.6 New Hydrogen Energy
During 2001, the activities on the so-called “new hydrogen energy” were based on a
critical revision of the results obtained the previous year. In particular, the objectives
were:
• Experimental demonstration of the Cohn-Aharonov effect, as suggested by G.
Preparata in 1993.
• Design of a compact and easy-to-handle electrolytic cell capable of showing
reproducible excess heat for some hours.
• Detection of 4 He as nuclear ashes generated by the phenomenon.
Past studies have shown that the cold fusion phenomenon, i.e., the capability of
deuterated palladium to produce energy, starts only when a minimum concentration
of deuterium inside the palladium lattice is reached. Studies at ENEA Frascati
demonstrated that the intrinsic characteristics of the material have a deep influence
on the possibility to reach the threshold. However, a new effect has recently been
proposed, which is that a voltage drop along the palladium sample could strongly
affect the chemical potential of deuterons inside the metal and decrease it
dramatically. To verify this effect experimentally, thin films with a high electrical
resistance are required in order to prevent Joule heating due to the current flow
through the cathode itself. The solution is represented by a very thin strip (50 µm
wide and 2 µm thick) deposited on a substrate in a geometry that is suitable for a
length of about one meter This choice of geometry was a crucial feature because of
the intrinsic mechanical fragility of such a structure. However, after studies on how
to optimise the deposition technique, highly stable films were obtained that were
capable of undergoing several load-deload cycles without mechanical break. This
was fundamental for obtaining tangible proof of the Cohn-Aharonov effect and
easily reaching the threshold in deuterium concentration. Extra power was
measured clearly each time such a threshold was
exceeded.
The average amount of power released in each
experiment was about 20 mWatt (standard
deviation 1.6 mWatt); but if related to the sample
dimensions (10 -4 cm 3 ), this value corresponds to
200 Watt/cm 3 . Further development is required
to increase such a yield. Figure 4.10 shows a
device hosting six electrolytic cells.
The nature of the heat excess produced during
deuteride formation is an outstanding problem.
The most common idea is that 4 He atom
formation from the D+D reaction releases about
24 MeV to the lattice and is responsible for the
excess heat measured. The most common
method to detect 4 He atoms eventually present
in the gases evolving from an electrolytic cell is
based on high-resolution mass spectroscopy,
which allows the resolution of 4 He and D 2 masses. Due to the high level of 4 He
contamination in normal air (5 ppm), an ultrahigh vacuum system was assembled in
which all gases except the nobles are pumped out from the analytic chamber by
means of a getter alloy pump. This feature allows a static mode of operation, in
Fig. 4.10 - Multiple-cell
prototype.

4. MISCELLANEOUS 119
4.6 New Hydrogen Energy
which gases evolving from the cell are accumulated and analysed in real time to
measure the amount of 4 He atoms contained in each sample and to correlate it with
the excess heat measured. The feasibility of a static measurement of the possible
presence of 4 He, intended as nuclear ash of the D+D nuclear reaction, was
demonstrated and extensive and accurate calibration of the quadrupole mass
spectrometer was carried out.
The activities concerning the contract awarded to ENEA and OCEM in June 2000 for
the manufacture and testing of by-pass diodes for the quench protection of the dipole
and quadrupole magnets of the Large Hadron Collider (LHC) at CERN were
successfully pursued.
During 2001, about 320 diode-stacks for dipole magnets and 100 diode-stacks for
quadrupole magnets were tested, thus maintaining the proper testing rate to be able
to complete the measurement campaign on all the stacks (1250 for the dipoles and
400 for the quadrupoles) within December 2004, as requested by CERN.
A paper was presented at the 2001 CEC/ICMC conference, describing the
experimental set-up, the data acquisition system developed at ENEA for the diodestack
testing and the results obtained.
4.8.1 Liquid helium service
The new helium liquefier, installed at the end of April 1999, operated regularly
during 2001. The facility consists of a TCF20 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 1,000 l dewars 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. The plant, supplied by Linde Cryogenics Ltd. (UK), is
automatically controlled by an Allen Bradley Programmable Logic Controller. The
liquid helium production rates, with or without liquid nitrogen precooling, are better
than the nominal rates of 60 and 30 l/h.
The liquid helium consumption underwent a consistent increase during the year
because of a new activity started at the Frascati Superconductivity Section
concerning the testing of diodes for the superconducting magnets of CERN. The
overall liquid helium production was about 38,000 l. Nearly 50,000 l of liquid helium
were delivered to the users, and about 12,000 l were acquired to reintegrate the
helium inventory, with a recovery efficiency of about 76%.
4.8.2 Cryogenic technologies
4.7 Cryogenic Testing of
Diode Stacks for CERN
4.8 Cryogenics
The development of a low-noise single shot 3 He/ 4 He dilution refrigerator based on
the use of cryosorption pumps was pursued during the first eight months of 2001, in
the context of a two year co-operative agreement between ENEA and the Physics
Department at the University of Rome 3. The objective of the activity was to
demonstrate the feasibility of a dilution refrigerator capable of achieving
temperatures below 100 mK with an operating cycle of at least 10 h. Cryosorption
pumps eliminate vibrations and mechanical noise, which is one of the main
requirements for many space- and earth-based applications.

120
4. MISCELLANEOUS
4.8 Cryogenics
The prototype consists of three refrigerating stages; the first can be cooled down to
about 1.5 K, by pumping on a liquid 4 He bath, which allows liquefaction of the 3He
and the 3 He/ 4 He mixture stored in the second and third stages, respectively.
Pumping on liquid 3 He in the second stage makes it possible to achieve
temperatures below 300 mK, thus ensuring a good phase separation of the gas
mixture in the mixing chamber. The final cool-down is realised by pumping on the
3 He rich phase.
Preliminary experiments gave encouraging results, with a minimum recorded
temperature at the mixing chamber of about 127 mK. The temperatures of the second
and third stages, during a typical dilution test, are as follows. The mixing chamber
temperature reaches a first plateau at 175 mK, lasting about 3 h, then it achieves its
minimum (135 mK in this run). The existence of the first plateau in not yet well
understood. The overall duration of the dilution process (T

5. INERTIAL CONFINEMENT 123
5.1 Introduction
For the reference period we shall report on (i) the preparation of a new experimental
campaign on laser-foam interaction that implied the assembling and testing on a new
diagnostic line, (ii) the theoretical activity for the preparation of the new experiment
and for the implementation of a new package in the code COBRAN for the treatment
of the energy deposition of the nuclear products (charged particles and neutrons),
and (iv) the design of the diode pumped amplifier.
5.2 Diagnostic Upgrading
The diagnostic package shown in figure 5.1 was assembled for measurements of light
transmission through the target during the laser irradiation.
After the transmitted light conversion to 2ω the target is imaged on a camera and on
the photodiode phd2ω by the lenses 2 and 3. The photodiode phdω is used to register
the waveform of the incident laser beam. A mask was placed on the phd2ω image to
select the probed area where transmission will be measured (typically smaller than
the laser focal area, see figure 5.2).
5.3 Theory
5.3.1 Interaction of laser beams with multi-foil plastic structures
In the following we report on the 2D simulations performed with the lagrangian
code COBRAN to study the evolution of structured plastic targets irradiated by laser
beams1. The method used was to start with the simplest material assemblies to begin
a computational study of the interaction of laser beams with large pore foams.
We began with simulations relative to the irradiation of single thin foils, to frame the
Polarizers
Target
Lens 1 Lens 2
Polarizers
Beam A
phd
Array of 256
lenses
Camera
/4 plates
stop
SHG
Infrared absorber
Lens 3
Beam
splitter
phd2
Filter
Dump
Target images
Fig. 5.1 - Package for the measurement of the transmitted light and for target
imaging in transmitted light. The photodiode phdω is used to register the
waveform of the incident laser beam. After the transmitted light is converted
to 2 ω, the target is imaged on a camera and on the photodiode phd2ω by the
lenses 2 and 3. A mask is placed on the phd2ω image to select the probed area
where transmission is measured (typically smaller than the laser focal area). The
transmission coefficient is deduced by normalization with shots without target
(anything else unchanged) and taking in to account the dependence on intensity
of the conversion to 2ω. Since the bandwidth of the waveforms registering
system is 6 GHz, the method allows time-resolved measurements within the
laser waveforms.

124
5. INERTIAL CONFINEMENT
5.3 Theory
a
Target
Laser spot
Opaque mask
400 µm
Table 5.I - Single foil simulations
b
5.2 - a) Relative positioning
and sizes of target
and laser spot. The spot
image was taken without
target and combined with
the typical cross section
of a foam target. The
relative positioning is that
adopted in the experiments.
b) Masking of the
laser spot. Although completely
opaque, the mask
is represented as partially
transmitting to show the
relative hole -spot
positions and sizes.
Case number Focalization Foil transparency Transit time
(cm) (tb ns) (ts, ns)
1 -0.015 0.79 0.62
2 -0.030 1.1 0.76
3 -0.045 1.37 0,85
main physical parameters (typical velocities, burn through time, etc…). Then a
structure composed by 3 parallel layers was considered. The material was assumed
to be CH and the foil thickness d=0.5 µm. In the multi-foil simulations the spacing
between them was s=75 µm (that is an average density of 6.7 mg).
The material was irradiated by 1.054 µm radiation, focused along the negative
direction of the z-axis (the optical axis) according to a F/1 geometry. The focal spot
was set at different positions along the z-axis for different cases. The pulse of laser
power, triangular as time waveform, started at t=0, achieved the maximum at t=0.7
ns and was set to zero at t=2 ns. The total energy used was 40 J. The solid material
was set on the positive portion of the z axis, starting at z=0.
In the single foil simulations the d=0.5 µm foil was set between z=0 and z=0.5 µm,
and irradiated by focusing with F/1 optics behind the target at z=-0.015 cm, or at
z=–0.03 cm or at z=-0.045. Some of the findings are reported in table 5.I.
The quantity t b in table 5.I represents the time when the foil becomes transparent to
the laser light (due to ablation and transverse expansion), whereas t s represents the
time when the accelerated matter is displaced by a distance, towards negative z,
s=75 µm, the “pores” size. For the cases listed in table 5.I, in spite of the twodimensional
effects, t b >t s . This means that, in a multi-layer structure, matter will be
accumulated from the irradiated layer upon the following one, so that the light will
become faced with an even thicker foil (and so on). In other words a sort of
snowplough process occurs, as mentioned in previous papers in which the
structured nature of the material was not considered. In the following we report
some results for case 3. In figure 5.3 the fraction of absorbed light is represented
(abscissa is time in ns). The absorption drops sharply near the transparency time t b .
In the following some quantities are represented just before t b . In figure 5.4 the

5. INERTIAL CONFINEMENT 125
5.3 Theory
density and laser rays are displayed. About 10 ps later the light is transmitted.
In the picture at left in figure 5.4 the density r is represented as function of the space
coordinates. The map in the center is the density as function of the calculation grid
indexes. At right is shown a detail of rays propagation, with different colors for
different incidence angle.
The situation near t s
is represented in figure 5.5. Rays are refracted in a sort of ring
as seen in the map at top of figure 5.5. In the same figure are also represented the
quantities Z k T e
(where Z is the average ion charge and T e
the electronic
temperature) and the average ion kinetic energy in the flow (that is 1/2 m i
n 2 , where
n is the flow velocity and m i
the ionic mass). Both are measured in °K. Kinetic energy
prevails only in a thin, dense layer destined to splash on the next one in a multi foil
target. From this follows that the flow energy is partially dissipated in a shock wave
driven in the next layer.
The last map shows the distribution of the flow velocity (U z is the component along
z, the positive axis pointing towards the laser).
The process of layer collision was produced in simulations for the interaction of a
three layer target with a light beam focused in such a way to initially reproduce, at the
first exposed surface, the conditions at surface of case 3.
Fig. 5.3 - Fraction of
absorbed light as function
of time for a single foil. A
fast drop in absorption
occurs when the target
becomes transparent.
At the time 0.65 ns the irradiated foil
impinges on the second foil. In figure
5.6 is shown the map of the velocity
along z at this time. Typical negative
velocity is about 2 (in units of 10 7
cm/s). In figure 5.7 is shown the
density map at the same time. The
propagation of the shock wave in the
second foil is clearly seen by the
representation of density in terms of
grid indexes.
-6 -5 -4 -3 -2 -1 0
Logr(g/cc)
Fig. 5.4 – Left: density (r) map 10 ps before light starts to be transmitted
through the target. Center: the figure is a representation of the density as
function of the grid coordinates. Details of ray propagation are shown in the
figure at right.

126
5. INERTIAL CONFINEMENT
5.3 Theory
-5 -4 -3 -2 -1 0
Log[r(g/cc)]
5 5.5 6 6.5 7 7.5 8 8.5
Log[ZTe(°K)]
5 5.5 6 6.5 7 7.5 8 8.5
Log[kin(°K)]
-2 0 2 4 6 8 10
Log[Uz(cm/s/10 -7 )]
Fig. 5.5 - Maps of some relevant quantities at the time ts corresponding to the vacuum closure aftera a flight
of 75 µm. (Te the electronic temperature, kin the ionic kinetic energy associated to the flow, U z the velocity
along the z axis).
Time=0.65 ns
-2 0 2 4 6 8 10
Log[Uz(cm/s/10 -7 )]
Fig. 5.6 - Evolution of a
three foil target. The
focusing conditions on
the first layer from the
right the same as in case
3 of table 5.I. It is shown
the map of the velocity
along z when the first
foil collides with the
second (at t=0.65 ns).
Typical negative velocity
is about 2 (in units of 10 7
cm/s).

5. INERTIAL CONFINEMENT 127
Time=0.65 ns
5.3 Theory
-5 -4 -3 -2 -1 0
Log[r(g/cc)]
Time=0.9 ns
Fig. 5.7 - Density maps at
different times and ray
tracing at 0.9 ns when the
second layer impinges on
the third.
-5 -4 -3 -2 -1 0
Log[r(g/cc)]
At 0.9 ns the second foil, and a part of the first, start to impinge on the third. To be
noted that the impinging time of the first foil on the second foil was 0.65 ns, whereas
the collision with the third follows after 0.25 additional nanoseconds. At 0.9 ns a
substantial transverse flow can be seen. Part of the first foil material flows
transversally between the unperturbed remnants of the first and second foil. Due to
this expansion the light succeeds in penetrating near the second foil surface. The ray
trajectories become quite complex and a remarkable transverse light excursion is
noted. In the same figure 5.7 it is represented the detail of the propagation for 10 rays
and, separately, the path of two most external rays.
The computation has been advanced up to t=0.97781ns. At this time the
computational grid becomes severely distort. At any rate many of the most

128
5. INERTIAL CONFINEMENT
5.3 Theory
interesting and peculiar features contain several grid points and occur where the grid
maintains a reasonable shape. For this reason we believe these features to be
meaningful.
Conclusions could be the following. In the framework of the physical model
included in our code (probably adequate for the modest power densities here
considered), the succession of the events agrees with the model based on the
formation of a cavity surrounded by a dense matter of increasing mass. The ablation
pressure pushes this layer. Most of the cavity energy content is thermal, the kinetic
one being associated to the dense layer flow. Light may wander in a somewhat
erratic way in the cavity.
5.3.2 Code COBRAN implementation
Cobran is a 3D (2 space + time), 3 temperature lagrangian code including "real
matter" equation of state and opacity coefficients. The package for driving energy
deposition includes light/heavy charged particles, ray-traced laser light and raytraced
x-rays. During the reference year the code was implemented by a more
complete package for reaction products energy deposition. Now the thermonuclear
reaction treatment includes finite range charged particles diffusion and non-thermal
nuclear reactions and the diffusion and energy deposition of neutrons is treated by a
Montecarlo code.
5.3.3 DPSSL Design activity
The block diagram of the diode pumped ABCD was completed and the design of the
sub-amplifiers was frozen. A solution for an efficient energy transfer from the diode
array was found and studied a by a ray-tracing Montecarlo code (fig. 5.8). The
simulations have shown that the quality of the pumping on the active material is
quite good (fig. 5.9).
Fig. 5.9 - Montecarlo simulations show
that about 97.3% of rays hit the useful
slab area. Reflection losses on the
involved optics are modest.
Fig. 5.8 - Montecarlo simulations
for energy transfer from a diode
array to an active element.

Projects
FUSION - DIVISION DIRECTORATE
FRASCATI
M.Samuelli
Assoc. Directors: F. De Marco
G.B. Righetti
G. Valli
Plasma Physics Application
L. Rapezzi
Scientific Secretariat
F. De Marco (acting)
Intense Neutron Source
B. Riccardi
Administration & Control
N. Manganiello
Radiofrequency
G.B. Righetti (acting)
JET/NET Personnel
M. Samuelli (acting)
Conceptual Reactor Studies
A. Pizzuto (acting)
Research Center Brasimone
D. Cassarini
NET/ITER
A. Pizzuto
Electrical Engineering
A. Coletti
Deputy Director Experimental Engineering
G. Benamati
Deputy Director Fusion Technology
A. Pizzuto
Inertial Confinement Fusion
A. Caruso
Deputy Dir. Magnetic Confinement Fusion Physics
F. Romanelli

PUBLICATIONS, CONFERENCES AND REPORTS 131
Publications
01/002 M. MARINELLI, E. MILANI, A. PAOLETTI, A. TUCCIARONE, G. VERONA
RINATI, M. ANGELONE, M. PILLON
Systematic study of the normal and pumped state of high efficiency diamond particle
detectors grown by chemical vapor deposition
J. Appl. Phys. 89, 2, 1430 (2001)
01/004 D. PACELLA, G. PIZZICAROLI, L. GABELLIERI, M. LEIGHEB, R. BELLAZINI,
A. BREZ, G. GARIANO, L. LATRONICO, N. LUMB, G. SPANDRE, M.M. MASSAI,
S. REALE
Ultrafast soft x ray 2D plasma imaging system based on gas electron multiplier detector
with pixel read-out
Rev. Sci. Instrum. 72, 2, 1372 (2001)
01/005 P. SARDAIN, C. GIRARD, J. ANDERSSON, M.T. PORFIRI, R. KURIHARA,
X. MASSON, G. MIGNOT, T. PINNA, L. TOPILSKI
Modelling of two-phase flow under accidental conditions fusion codes benchmark
Fusion Eng. Des. 54, 555-561 (2001)
01/007 A. NATALIZIO, L. DI PACE, T. PINNA
Assessment of occupational radiation exposure for two fusion power plant designs
Fusion Eng. Des. 54, 375-385 (2001)
01/008 V. VIOLANTE, P. TRIPODI, C. LOMBARDI
Le conoscenze attuali sulla fusione nucleare fredda
La termotecnica, marzo 2001, pp. 67-72
01/015 H. FREIESLEBEN, D. RICHTER, K. SEIDEL, S. UNHOLZER, Y. CHEN,
U. FISCHER, M. ANGELONE, P. BATISTONI, M. PILLON
Experimental validation on shut-down dose rates measurement of dose rates, decay rays and
neutron flux
Dresden Report TUD-IKTP/01-01
01/016 A. LA BARBERA, B. RICCARDI, A. DONATO, C.A. NANNETTI,
L. F. MORESCHI
Stability of SiC/SiC fibre composites exposed to Li 4 SiO 4 and Li 2 TiO 3 in fusion relevant
conditions
J. Nucl. Mater. 294, 223-231 (2001)
01/017 B. DI MARTINO, S. BRIGUGLIO, G. VLAD, P. SGUAZZERO
Parallel PIC plasma simulation through particle decomposition techniques
Parallel Computing 27, 295-314 (2001)

132
PUBLICATIONS, CONFERENCES AND REPORTS
01/018 G. VLAD, S. BRIGUGLIO, G. FOGACCIA, B. DI MARTINO
Gridless finite-size-particle plasma simulation
Comp. Phys. Commun 134, 58-77 (2001)
01/019 V. BOFFA, G. CELENTANO, L. CIONTEA, F. FABBRI, V. GALLUZZI,
U. GAMBARDELLA, G. GRIMALDI, A. MANCINI, T. PETRISOR
Influence of film thickness on the critical current of YBa 2 Cu 3 O 7-x thick films on Ni-V
biaxially textured substrates
IEEE Trans. on Appl. Superconductivity, 11, NO 1, 3158-3161 (2001)
01/020 G. GRIMALDI, V. BOFFA, G. CELENTANO, F. FABBRI, U. GAMBARDELLA,
S. PACE, T. PETRISOR
Critical current hysteresis in low angle Y-BaCu-O bicrystals
IEEE Trans. Appl. Supercond. 11, NO 1, 3776-3779 (2001)
01/021 L. BOTTURA, M. CIOTTI, P. GISLON, M. SPADONI, P. BELLUCCI, L. MUZZI,
S. TURTU' A. CATITTI, S. CHIARELLI, A. DELLA CORTE, E. DI FERDINANDO
Stability in a long length NbTi CICC
IEEE Trans. Appl. Supercond. 11, NO 1, 1542-1545 (2001)
01/022 M.PILLON, M. ANGELONE, R.A. FORREST
A new detector to measure gamma and beta decay power from radionuclides
Nucl. Instrum. Methods Phys. Res. A 461, 582-583 (2001)
01/023 S.E. SEGRE, V. ZANZA
Evolution of polarization for radiation crossing a plasma layer of quasi-transverse
propagation and the interpretation of radioastronomical measurements
Astrophys. J. 554, 408-415 (2001)
01/024 F. CRISANTI, B. ESPOSITO, C. GORMEZANO, A. TUCCILLO, L. BERTALOT,
C. GIROUD, C. GOWERS, R. PRENTICE, K.D. ZASTROW, M. ZERBINI
Analysis of ExB flow shearing rate in JET ITB discharges
Nucl. Fusion 41,7, 883-889 (2001)
01/025 V. VIOLANTE, A. TORRE, G. SELVAGGI, G.H. MILEY
Three dimensional analysis of the lattice confinement effect on ion dynamics in condensed
matter and lattice effect on the D-D nuclear reactor channel
Fusion Techn. 39, 266-281 (2001)
01/027 C. FAZIO, G. BENAMATI, C. MARTINI, G. PALOMBARINI
Compatibility tests on steels in molten lead and lead-bismuth
J. Nucl. Mater. 296, 243-248 (2001)

PUBLICATIONS, CONFERENCES AND REPORTS 133
01/028 F. BARBIER, G. BENAMATI, C. FAZIO, A. RUSANOV
Compatibility tests of steels in flowing liquid lead-bismuth
J. Nucl. Mater. 295, 149-156 (2001)
01/030 M. ANGELONE, P. BATISTONI, M. PILLON
Effect of the encapsulating material on the peak3/peak5 response ratio of TLD-300
irradiated with neutrons of variuos energy
Rad. Phys. Chem. 61, 415-416 (2001)
01/031 R. BERNABEI, P. BELLI, R. CERULLI, F. MONTECCHIA, A. INCICCHITTI,
D. PROSPERI, C.J. DAI, M. ANGELONE, P. BATISTONI, M. PILLON
Light response of a pure liquid Xenon scintillator irradiated with 2.5 MeV neutrons
EPJ direct C11, 1-8 (2001)
01/033 A. SESTERO
The Ignition Brachystochrone
J. Plasma Phys. 65, 1, 59-72 (2001)
01/034 A.R. RAFFRAY, R. JONES, G. AIELLO, M. BILLONE, L. GIANCARLI,
H. GOLFIER, A. HASEGAWA, Y. KATOH, A. KOHYAMA, S. NISHIO, B. RICCARDI,
M.S. TILLACK
Design and material issues for high performance SiC f /SiC-based fusion power cores
Fusion Eng. Des. 55, 55-95 (2001)
01/038 M. ANGELONE, T. BUBBA, A. ESPOSITO
Measurement of the mass attenuation coefficient for elemental materials in the range
6

134
PUBLICATIONS, CONFERENCES AND REPORTS
01/046 A. BASILE, G. CHIAPPETTA, S. TOSTI, V. VIOLANTE
Experimental and simulation of both Pd and Pd/Ag for a water gas shift membrane reactor
Sep. Puri. Technol. 25, 549-571 (2001)
01/047 R. CESARIO, A. CARDINALI, C. CASTALDO, M. LEIGHEB, M. MARINUCCI,
V. PERICOLI-RIDOLFINI, F. ZONCA, G. APRUZZESE, M. BORRA, R. DE ANGELIS,
E. GIOVANNOZZI, L. GABELLIERI, H. KROEGLER, G. MAZZITELLI, P. MICOZZI,
L. PANACCIONE, P. PAPITTO, S. PODDA, G. RAVERA, B. ANGELINI, M.L. APICELLA,
E. BARBATO, L. BERTALOT, A. BERTOCCHI, G. BUCETI, S. CASCINO, C. CENTIOLI,
P. CHUILON, S. CIATTAGLIA, V. COCILOVO, F. CRISANTI, F. DE MARCO,
B. ESPOSITO, G. GATTI, C. GOMERZANO, M. GROLLI, F. IANNONE, G. MADDALUNO,
G. MONARI, F. ORSITTO, D. PACELLA, M. PANELLA, L. PIERONI, G.B. RIGHETTI,
F. ROMANELLI, E. STERNINI, N. TARTONI, P. TREVISANUTTO, A.A. TUCCILLO,
O. TUDISCO, V. VITALE, G. VLAD, M. ZERBINI
Reduction of the electron thermal conductivity produced by ion Bernstein waves on the
Frascati Tokamak Upgrade tokamak
Phys. Plasma 8,11, 4721 (2001)
01/049 P. BURATTI AND JET TEAM
High beta plasmas and internal barrier dynamics in JET discharges with optimised shear
Nucl. Fusion 41, 12, 1809 (2001)
01/053 F. DE MARCO
Prospettive della fusione nucleare
Il Nuovo Saggiatore, 17, 5-6, 61-65 (2001)
01/063 G. CELENTANO, A. CAPRICCIOLI, A. CUCCHIARO, M. GASPAROTTO,
A. BIANCHI, G. FERRARI, B. PARODI, G.P. SANGUINETTI, F. VIVALDI, S. ORLANDI,
B. COPPI
Engineering evolution of the ignitor machine
Fusion Eng. Des. 58-59, 815-820 (2001)
01/066 A. CARUSO, C. STRANGIO
Studies on nonconventional high-gain target design for ICF
Laser Part. Beams 19, 295-308 (2001)
01/069 R.A. FORREST, M. PILLON, U. VON MÖLLENDORFF, K. SEIDEL
Validation of EASY-2001 using integral measurements
UKAEA Report FUS 467 (2001)
01/071 S.E. SEGRE
New formalism for the analysis of polarization evolution for radiation in a weakly
nonuniform, fully anisotropic medium: a magnetized plasma
J. Opt. Soc. Am. A 18, 10, 2601 (2001)

PUBLICATIONS, CONFERENCES AND REPORTS 135
01/074 A. NATALIZIO, T. PINNA, L. DI PACE
Impact of plant incidents on worker radiation exposure for the SEAFT design
Fusion Eng. Des. 58-59, 1065-1069 (2001)
01/076 R.K. MAIX, H. FILLUNGER, F. HURD, E. SALPIETRO, N. MITCHELL,
P. LIBEYRE, P. DECOOL, A. ULBRICHT, G. ZHAN, A. DELLA CORTE, M. RICCI,
D. BRESSON, A. BOURQUARD, F. BAUDET, B. SCHELLONG, E. THEISEN, N. VALLE
Completion of the ITER toroidal field model coil (TFMC)
Fusion Eng. Des. 58-59, 159-164 (2001)
01/077 T. KATO, H. TSUJI, T. ANDO, Y TAKAHASHI, H. NAKAJIMA, M. SUGIMOTO,
T. ISONO, N. KOIZUMI, K. KAWANO, M. OSHIKIRI, K. HAMADA, Y. NUNOYA,
K. MATSUI, T. SHINBA, Y. TSUCHIYA, G. NISHIJIMA, H. KUBO, E. HARA,
H. HANAWA, K. IMAHASHI, K. OOTSU, Y. UNO, T. OOCHI, J. OKAYAMA,
T. KAWASAKI, M. KAWABE, S. SEKI, K. TAKANO, Y. TAKAYA, F. TAJIRI,
A. TSUTSUMI, T. NAKANURA, H. HANAWA, H. WAKABAYASHI, K. NISHII,
N. HOSOGANE, M. MATSUKAWA, Y. MIURA, T. TERAKADO, J. OKANO,
K. SHIMADA, M. YAMASHITA, K. ARAI, T. ISHIGOUOKA, A. NINOMIYA, K. OKUNO,
D. BESSETE, H. TAKIGAMI, N. MARTOVETSKY, P. MICHAEL, M. TAKAYASU, M. RICCI,
R. ZANINO, L. SAVOLDI, G. ZAHAN, A. MARTINED, R. MAIX
First test results for the ITER central solenoid model coil
Fusion Eng. Des. 56-57, 59-70 (2001
01/078 J.L. DUCHATEAU, H. FILLUNGER, S. FINK, R. HELLER, P. HERTOUT,
P. LIBEYRE, R. MAIX, C. MARINUCCI, A. MARTINEZ, R. MEYDER, S. NICOLLET,
S. RAFF, M. RICCI, L. SAVOLDI, A. ULBRICHT, F. WUECHNER, G. ZAHN, R. ZANINO
Test program preparations of the ITER toroidal field model coil (TFMC)
Fusion Eng. Des. 58-59, 147-151 (2001)
01/079 B. DI MARTINO, S. BRIGUGLIO, G. VLAD, G. FOGACCIA
Workload decomposition strategies for shared memory parallel systems with openMP
Scientific Programming 9, 109-122 (2001)
01/081 F. SCARAMUZZI
Dieci anni di fusione fredda: una testimonianza diretta
Bimestrale dell’ENEA, Anno 47, 5, 21 (2001)
01/082 M. MARINELLI, E. MILANI, A. PAOLETTI, A. TUCCIARONE,
G. VERONA-RINATI, M. ANGELONE, M. PILLON
Pulse-shape analysis of high efficiency chemical vapor deposition diamond particle
detectors in the normal and pumped state: trapping and detrapping effects
Phys. Rev. B64, 195205-1/195205-8 (2001)
01/085 O.N. JARVIS, P. VAN BELLE, M.A. HONE, G.J. SADLER, G.A.H. WHITFIELD
F.E. CECIL, D.S. DARROW, B. ESPOSITO
Measurements of escaping fast particles using a thin-foil charge collector
Fusion Technol. 39, 84 (2001)

136
PUBLICATIONS, CONFERENCES AND REPORTS
Articles in Course of Publications
L. RAPEZZI, M. PILLON, M. RAPISARDA, M. SAMUELLI, M. ANGELONE, E. ROSSI,
F. MEZZETTI
Development of a mobile and repetitive Plasma Focus
Plasma Sources Science and Technology
F. ROMANELLI
Transport and boundary physics: summary review
Fusion Technol.
G. CELENTANO, T. PETRISOR, V. BOFFA, L. CIONTEA, F. FABBRI, V. GALLUZZI,
U. GAMBARDELLA, A. MANCINI , A. RUFOLONI, E. VARESI
Epitexial oxidation of Ni-V biaxially textured tapes
Physica C
B. DI MARTINO, S. BRIGUGLIO, M. CELINO, G. FOGACCIA, G. VLAD, V. ROSATO,
M. BRISCOLINI
Development of large scale high performance applications with a parallelizing compiler
Int. J. of Computer Research: Special issued on Industrial Applications of Parallel Computing
B. DI MARTINO, S. BRIGUGLIO, M. CELINO, G. FOGACCIA, G. VLAD, V. ROSATO,
M. BRISCOLINI
Experiences on parallelizing compilation for development and porting of large scale
applications on distributed memory parallel systems
Advances in Computation: Theory and Practice
M. SHOUCRI, A. CARDINALI, J.P. MATTE, R. SPIGLER
Numerical study of plasma-wall transition using an Eulerian Vlasov code
European Phys. J. D
V. KRIVENSKI, G. BRACCO, P. BURATTI, G. GIRUZZI, O. TUDISCO, S. CIRANT,
F. CRISANTI
Distortion of the electron distribution bulk during electron cyclotron heating on FTU
Phys. Rev.
S. BRIGUGLIO, B. DI MARTINO, G. VLAD
Workload decomposition strategies for hierarchical distributed-shared memory parallel
systems and their implementation with integration of high level parallel languages
Concurrency and Computation practice & Experience
S. BRIGUGLIO, G. VLAD, F. ZONCA, G. FOGACCIA
Nonlinear saturation of shear Alfvén modes and energetic ion transports in tokamak
equilibria with hollow-q profile
Phys. Lett. A

PUBLICATIONS, CONFERENCES AND REPORTS 137
Contributions to Conference
B. RICCARDI, C.A. NANNETTI, T. PETRISOR, M. SACCHETTI
Low activation brazing materials and techniques for SiCf/SiC composites
ICFRM-10 International Conference on Fusion Reactor Materials
Baden Baden (Germany)) October 14-19, 2001
L. DI PACE, A. NATALIZIO
Waste management aspects of fusion power plant
The 8th International Conference on Environmental Management
Bruges, Belgium September 30-October 4, 2001
M. CIOTTI, A. DI ZENOBIO, P. GISLON, L. MUZZI, M. SPADONI, S. TURTÙ
Loss calculations in a CICC solenoid exposed to rapidly changing magnetic fields
EUCAS 2001
Copenaghen (Denmark) August 26-30, 2001
E. BALSAMO, P. BELLUCCI, A. CATITTI, M. CIOTTI, A. DELLA CORTE, P. GISLON,
L. MUZZI, G. PASOTTI, M. RICCI, M. SPADONI
An experiment for the study of the current distribution effect on stability with different conductors
EUCAS 2001
Copenaghen (Denmark) August 26-30, 2001
G. GELENTANO, V. BOFFA, L. CIONTEA, F. FABBRI, V. GALLUZZI,
U. GAMBARDELLA, A. MANCINI, T. PETRISOR, R. ROGAI, A. RUFOLONI, E. VARESI
High J C YBCO coated conductors on non-magnetic metallic substrate using YSZ-based buffer layer
architecture
EUCAS 2001
Copenaghen (Denmark) August 26-30, 2001
E. VARESI, V. BOFFA, G. CELENTANO, L. CIONTEA, F. FABBRI, V. GALLUZZI,
U. GAMBARDELLA, A. MANCINI, T. PETRISOR, A. RUFOLONI, A. VANNOZZI
Biaxial texturin of Ni Alloy substrates for YBCO coated conductors
EUCAS 2001
Copenaghen (Denmark) August 26-30, 2001
N. MARTOVETSKY, P. MICHAEL, J. MINERVINI, A. RADOVINSKY, M. TAKAYASU,
C.Y. GUNG, R. THOME, T. ANDO, T. ISONO, T. KATO, H. NAKAJIMA, G. NISHIJIMA,
Y. NUNOYA, M. SUGIMOTO, Y. TAKAHASHI, H. TSUJI, D. BESSETTE, K. OKUNO,
N. MITCHELL, M. RICCI, R. ZANINO, L. SAVOLDI, K. ARAI
Test of the ITER central solenoid model coil and CS insert
17th Int. Conf. on Magnet Technology
Geneva (Switzerland) September, 24-28, 2001

138
PUBLICATIONS, CONFERENCES AND REPORTS
H. FILLUNGER, F. HURD, R.K. MAIX, E. SALPIETRO, D. CIAZYNSKY, J.L.
DUCHATEAU, P. LIBEYRE, A. MARTINEZ, E. BOBROV, W. HERZ, M. SÜER, A.
ULBRICHT, F. WÜCHNER, G. ZAHN, A. DELLA CORTE, M. RICCI, E.
THEISEN, G. KRAFT, A. BOURQUARD, F. BEAUDET, B. SCHELLONG, R.
ZANINO, L. SAVOLDI
Assembly in the test facility, acceptance and first test results of the ITER TF model coil
17th Int. Conf. on Magnet Technology
Geneva (Switzerland) September, 24-28, 2001
D. CIAZYNSKI, M. RICCI, J.L. DUCHATEAU, A. ULBRICHT, F. WUECHNER, G. ZAHN,
H. FILLUNGER, R. MAIX
Resistances of electrical joints in the TF model coil of ITER: comparison of first test results with
samples results
17th Int. Conf. on Magnet Technology
Geneva (Switzerland) September, 24-28, 2001
R. CESARIO, A. CARDINALI, C. CASTALDO, M. LEIGHEB, M. MARINUCCI,
V. PERICOLI-RIDOLFINI, F. ZONCA AND THE FTU GROUP
Transport analysis results of the ion Bernstein wave experiment on the FTU tokamak
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 18-22, 2001
O.TUDISCO, F. CRISANTI, P.LOMAS, E. JOFFRIN, F. RIMINI, A.BECOULET, L.
BERTALOT, T. BOLZONELLA, G.BRACCO, C. GIROUD, S.CORTES, B. ESPOSITO, N.
HAWKES, S. POPOVICHEV, E.RACHLEW, M. RIVA, AND CONTRIBUTORS TO THE
EFDA-JET WORKPROGRAMME
Effect of internal flux shaping in JET transport barrier
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 18-22, 2001
O. TUDISCO, E. DE LA LUNA, V. KRIVENSKI, G. GIRUZZI, P. AMADEO, A. BRUSCHI,
F. GANDINI, G. GRANUCCI, V. MUZZINI , A. SIMONETTO, FTU AND ECRH GROUP
Oblique ECE measurements during strong ECH at 140 GHz in FTU
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 18-22, 2001
M. ROMANELLI, F. ROMANELLI, F. ZONCA
On the optimal choice of the dimensionless parameters of burning plasma physics experiments
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 18-22, 2001

PUBLICATIONS, CONFERENCES AND REPORTS 139
D. FRIGIONE, P. BURATTI, M. MARINUCCI, E. GIOVANNOZZI, F. POLI, M.
ROMANELLI, M.L. APICELLA, P. AMADEO, G. BRACCO, B. ESPOSITO, L.
GARZOTTI, C. GORMEZANO, G. MONARI , D. PACELLA, L. PANACCIONE, L.
PIERONI, O. TUDISCO AND FTU TEAM
High field, high performance operation in FTU with multiple pellet injection
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 18-22, 2001
E. GIOVANNOZZI, P. BURATTI, D. FRIGIONE, L. PANACCIONE, O. TUDISCO,
P. SMEULDERS AND FTU TEAM
Sawtooth and M=1 mode behaviour in FTU pellet enhanced discharges
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 18-22, 2001
E. BARBATO
ECHR studies: internal transport barriers and MHD stabilisation
28th EPS- Conference on Controlled Fusion and Plasma Physics
Madeira (Portugal) June 17-22, 2001
A. DELLA CORTE, A. GHARIB, D. HAGEDORN, S. TURT, G.L. BASILE, A. CATITTI,
S. CHIARELLI, E. DI FERDINANDO, G. TADDIA, M. TALLI, L. VERDINI, R. VIOLA
Cryogenic testing of by-pass diode stacks for the superconducting magnets of the large hadron collider
at CERN
CEC-ICMC 2001 Conference
Madison, Visconsin (USA) July, 2001
F. ALLADIO, A. MANCUSO, P. MICOZZI, F. ROGIER
Chandrasekhar-kendall-Furth configurations for magnetic confinement
4th Symp. on Current Trends in International Fusion Research
Ottawa (Canada) 2001
A. CARDINALI
Modeling of current drive in the high harmonic fast wave experiments
14th Topical Conference on Radio Frequency Power in Plasmas
Oxnard, California (USA) May 7-9, 2001

140
PUBLICATIONS, CONFERENCES AND REPORTS
A.A. TUCCILLO, Y. BARANOV, E. BARBATO, PH. BIBET, C. CASTALDO, R. CESARIO,
W. COCILOVO, F. CRISANTI, R. DE ANGELIS, A.C. EKEDAHL, A. FIGUEIREDO, M.
GRAHAM, G. GRANUCCI, D. HARTMANN, J. HEIKKINEN, T. HELLSTEN, F.
IMBEAUX, T.T.H. JONES, T. JOHNSON, K.V. KIROV, P. LAMALLE, M. LAXABACK, F.
LEUTERER, X. LITAUDON, P. MAGET, J. MAILLOUX, M.J. MANTSINEN, M.L.
MAYORAL, F. MEO, I. MONAKHOV, F. NGUYEN, J-M.NOTERDAEME, V.
PERICOLI-RIDOLFINI, S. PODDA, L. PANACCIONE, E. RIGHI, F. RIMINI, Y. SARAZIN,
A. SIBLEY, A. STAEBLER, T. TALA, D. VAN EESTER AND EFDA-JET
WORK-PROGRAMME CONTRIBUTORS
Recent heating and current drive result on JET
14th Topical Conference on Radio Frequency Power in Plasmas
Oxnard, California (USA) May 7-9, 2001
V. PERICOLI RIDOLFINI, S. PODDA, J. MAILLOUX, Y. SARAZIN, Y. BARANOV,
S. BERNABEI, R. CESARIO, V. COCILOVO, A. EKEDAHL, K. ERENTS, G. GRANUCCI,
F. IMBEAUX, F. LEUTERER, F. MIRIZZI, G. MATTHEWS, L. PANACCIONE, F. RIMINI,
A.A. TUCCILLO AND EFDA-JET CONTRIBUTORS
LHCD coupling during H-mode and ITB in JET plasmas
14th Topical Conference on Radio Frequency Power in Plasmas
Oxnard, California (USA) May 7-9, 2001
V. PERICOLI RIDOLFINI, E. BARBATO, A. BRUSCHI, R. DUMONT, F. GANDINI,
G. GIRUZZI, C. GORMEZANO, G. GRANUCCI, L. PANACCIONE, Y. PEYSSON, S. PODDA,
A.N. SAVELIEV, FTU TEAM, ECH TEAM
Combined LH and ECH experiments in the FTU Tokamak
14th Topical Conference on Radio Frequency Power in Plasmas
Oxnard, California (USA) May 7-9, 2001
F. MIRIZZI , P. PAPITTO
The very high power radiofrequency additional heating systems for the FTU tokamak of the Fusion
Division of ENEA in Frascati
Int. Seminar on “Heating by Internal Sources” HIS-01
Padova (Italy) September 12-14, 2001
F. DE MARCO
A look to the future: Fusion
Int. Conference “E. Fermi and Nuclear Energy,
Pisa (Italy) Ottobre 15-16, 2001
P. BATISTONI
Research in the field of neutronics and of nuclear data for fusion
International Conference on “Nuclear Energy in Central Europe 2001”
Portoro (Slovenia) September 10-13, 2001

PUBLICATIONS, CONFERENCES AND REPORTS 141
S. TOSTI, G. CHIAPPETTA, C. RIZZELLO, A. BASILE, V. VIOLANTE
Pd-Ag Membrane reactors for water gas shift
17th North American Catalysis Society Meeting
Toronto, Ontario (Canada) June 3-8, 2001
C.NERI, L. BARTOLINI, A. COLETTI, M.FERRI DE COLLIBUS, G.FORNETTI,
S. LUPINI, F. POLLASTRONE, L. SEMERARO, C. TALARICO
Advanced digital processing for amplitude and range determination in optical RADAR systems
2001 IEEE Real Time Conference
Valencia (Spain) June 4-8, 2001
Reports
RT/ERG/FUS/2001/01 S.E. SEGRE
Exact analytic expressions for the evolution of polarization for radiation propagating in a
plasma with nonuniformly sheared magnetic field
RT/ERG/FUS/2001/08 C. LO SURDO
A glorious, yet almost forgotten, mathematical theory, and some possibly new applications
of it to physics
RT/ERG/FUS/2001/09 M. RAPISARDA, M. SAMUELLI
A portable neutron source for landmines detection
RT/ERG/FUS/2001/13 S.E. SEGRE
Comparison of two alternative approaches for the analysis of polarization evolution of em
waves in a nonuniform, fully anisotropic medium: a magnetized plasma
RT/ERG/FUS/2001/14 F. ALLADIO, A. MANCUSO, P. MICOZZI, L. PIERONI, C.
ALESSADRINI, G. APRUZZESE, L. BETTINALI, G. BRACCO, P. BURATTI, A. COLETTI, P.
COSTA, C. CRESCENZI, A. CUCCHIARO, R. DE ANGELIS, T. FORTUNATO, D.
FRIGIONE, M. GASPAROTTO, G. GATTI, R. GIOVAGNOLI, L.A. GROSSO, G.
MADDALUNO, G. MAFFIA, S. MANTOVANI, G. MONARI, C. NARDI, S.
PAPASTERGIOU, M. PILLON, A. PIZZUTO, M. ROCCELLA, M. SANTINELLI, L.
SEMERARO, A. SIBIO, B. TILIA, O. TUDISCO, L. ZANNELLI, V. ZANZA
Proto-Sphera

142
PUBLICATIONS, CONFERENCES AND REPORTS
The Nuclear Fusion Department promotes the dissemination of information on plasma physics
and fusion technology, both nationally and internationally
Conferences organised at ENEA Frascati in 2001
Frascati, 22-23/11/01:
International Workshop on FTU Program
Seminars organised and held at Frascati in 2001
12-02-2001 A. DE NINNO - ENEA - Frascati, Italy
Il Progetto Nuova Energia da Idrogeno: Nuovi Elementi di Discussione
19-02-2001 P. HAGELSTEIN - MIT - Cambridge, USA
Basic Theory for Lattice-Nuclear Coupling and Anomalous in Metal Deuterides
26-03-2001 C. LO SURDO -ENEA -Frascati, Italy
Una Gloriosa ma Quasi Dimenticata Teoria Matematica, e Certe sue Applicazioni alla Fisica
(anche del Plasma) Presuntivamente Nuove
3-04-2001 G. BORRELLI - ENEA - Anguillara, Italy
Fusione Termonucleare e Opinione Pubblica: L'Esperienza di Porto Torres
6-04-2001 M. ULRICKSON - Albuquerque, USA
Lithium Conditioning and Liquid Walls in Tokamak
23-04-2001 A. COLETTI - ENEA - Frascati, Italy
FT3: Risultati dello Studio Concettuale
23-04-2001 A. PIZZUTO - ENEA - Frascati, Italy
FT3: Risultati dello Studio Concettuale
23-04-2001 G.B. RIGHETTI - ENEA - Frascati, Italy
FT3: Risultati dello Studio Concettuale
23-04-2001 F. ROMANELLI - ENEA - Frascati, Italy
FT3: Risultati dello Studio Concettuale
21-05-2001 C. TSALLIS - Centro Brasileiro de Pesquisas - Rio de Janeiro, Brazil
Thermostatistically Speaking, What Anomalous Diffusion, Turbulence, High Energy Physics and
Hydra Viridissima Have in Common
28-05-2001 C.S. PITCHER - MIT - Cambridge, USA
Modelling of the Ignitor Edge Plasma
4-06-2001 A. MAAS -CEA - Cadarache, France
ITER in Cadarache
Conferences and Seminars
11-06-2001 YU. KRAVTSOV - Russian Ac. of Sciences - Moscow, Russia
Complex Rays: From Intellectual Toy to Effective Instrument of Wave Theory

PUBLICATIONS, CONFERENCES AND REPORTS 143
2-07-2001 F. Rogier - ONERA - Toulouse, France
Modello Numerico dei Propulsori ad Effetto Hall
3-07-2001 A. SYKES - UKAEA - Abingdon, U.K.
The Spherical Tokamak Programme at Culham
3-07-2001 G.M. VOSS - UKAEA - Abingdon, U.K.
Spherical Tokamak Power Plant Studies
3-07-2001 H.R. WILSON - UKAEA - Abingdon, U.K.
Theory and Modelling for the Spherical Tokamak
16-07-2001 M. SHOUCRI - C.C.F.M. - Varennes, Montreal, Canada
Numerical Simulation of Plasma-Wall Transition and Plasma Detachment Using an Eulerian
Vlasov Code
23-07-2001 F. ZONCA - ENEA- Frascati, Italy
Role of Resonant Vs. Non-Resonant Wave-Particle Interactions in Electromagnetic Turbulence
17-09-2001 M. FLEISCHMANN - Salisbury -U.K.
Unfinished business
10-12-2001 J. HOW - CEA - Cadarache, France
The Technical Infrastructure for Remote Participation in the European Fusion Programme
10-12-2001 V. SCHMIDT - CNR - Padova, Italy
The Technical Infrastructure for Remote Participation in the European Fusion Programme

ABBREVIATIONS AND ACRONYMS
149
AC
ACP
AGILE
AITG
ALARA
Alcator C-Mod
ASDEX-U
ASI
alternating current
activated corrosion product
Astrorivelatore Gamma ad Immagini LEggero
Alfvén ion-temperature gradient
as-low-as-reasonably achievable
Tokamak at Massachusetts Institute of Technology, Boston, USA
Axisymmetric Divertor Experiment Upgrade. Tokamak at Garching, Germany
(Association EURATOM-IPP)
Agenzia Spaziale Italiana
BET
BSE
Brunauer-Emitt-Teller
backscattered electron
CEA
CECE
CERN
CESI
CFC
CFK
CIC
CNR
CRPP
CSMC
CTA
CVD
CVI
Commissariat à l’Energie Atomique - France
combined electrolysis catalytic exchange
Organisation Europeénne pour la Recherche Nucléaire- Geneva
Centro Elettrotechnico Sperimentale, Milan
carbon fibre composite
Chandrasekhar-Kendall-Furth
cable in conduit
Consiglio Nazionale delle Ricerche - Italy
Centre de Recherches en Physique des Plasmas, Villigen, Switzerland
central solenoid model coil
Co-ordination Technical Activity
chemical vapour deposition
chemical vapour infiltration
DARMA
DC
DIII-D
DOE
DRP
DTP
DV
Dark Matter (Experiment)
direct current
Doublet III - D-shape. Tokamak at General Atomics San Diego, USA
Department of Energy - U.S.A.
Divertor Refurbishment Platform - ENEA - Brasimone
Divertor Test Platform - ENEA - Brasimone
divertor
EAF
EBW
European Activation File
electron beam welding

150
ABBREVIATIONS AND ACRONYMS
EC
ECE
ECH
ECRH
EDA
EDI
EFDA
EFF
EFTP
EISSG
EM
EPM
ETG
EU
electron cyclotron
electron cyclotron emission
electron cyclotron heating
electron cyclotron resonance heating
Engineering Design Activities
Edge of IGNITOR (code)
European Fusion Development Agreement
European Fusion File
European Fusion Technology Programme
European ITER Site Study Group
electromagnetic
energetic particle mode
electron temperature gradient
European Union
FDR
FEAT
FEM
FESAC
FMEA
FFMEA
FIMEC
FNF
FNG
FNS
FTU
FWHM
FZJ
FZK
Final Design Report
Fusion Energy Advanced Tokamak
finite-element method/model
Fusion Energy Science Advisory Committee
failure mode and effect analaysis
functional failure mode and effect analysis
flat-top indentor for mechanical characterization
fast neutron facility
Frascati Neutron Generator - ENEA - Frascati
Fusion Neutronics Source - JAERI - Japan
Frascati Tokamak Upgrade - ENEA - Frascati
full width at half maximum
Forschungszeuntrum - Jülich - Germany
Forschungszeuntrum - Karlsruhe - Germany
GAE
GB
GDRD
GEM
GSSR
global Alfvén eigenmode
glove box
General Design Requirement Document
gas-electron multiplier
Generic-Site Safety Report
HCPB
HD
helium-cooled pebble bed
hot dipping

ABBREVIATIONS AND ACRONYMS
151
HELICA
HIP
HMGC
HRP
HT
HE-FUS3 Lithium Cassette
hot isostatic pressing
hybrid MHD gyrokinetic code
hot radial pressing
home team
IBW
ICE
ICF
ICRF
ICRH
IEA
IFMIF
INFN
IOC
IR
ISAS
ISD
ISS
ITB
ITER
ITG
IVC
IVROS
IVVS
ion Bernstein wave
Inlet of Coolant Events
inertial confinement fusion
ion cyclotron resonance frequency
ion cyclotron resonance heating
International Energy Agency
International Fusion Materials Irradiation Facility
Istituto Nazionale di Fisica Nucleare - Italy
improved Ohmic confinement
infrared
Integrated Safety Analysis Code System
inclined substrate deposition
isotope separation system
internal transport barrier
International Thermonuclear Experimental Reactor
ion-temperature gradient
in-vessel component
in-vessel remote operating system
in-vessel viewing system
JAERI
JCT
JET
JHU
JRC
JT-60U
Japan Atomic Energy Research Institute - Japan
Joint Central Team
Joint European Torus. Largest EU tokamak, Abingdon U.K. (UKAEA).
John Hopkins University - Maryland - U.S.A.
Joint Research Centre - Ispra - Italy
JAERI Tokamak 60 Upgrade, Naka, Japan
KAW
kinetic Alfvén wave
LCF
LH
low-cycle fatigue
lower hybrid

152
ABBREVIATIONS AND ACRONYMS
LHC
LHCD
LIM
LIVVS
LOCA
Large Hadran Collider (CERN)
lower hybrid current drive
limiter
laser in-vessel viewing system
loss-of-coolant accident
MARFE
MEPHI
MHD
MPGD
MSE
multifaceted asymmetric radiation from the edge
Moscow Engineering Physics Institute
magnetohydrodynamic
micro-pattern gas detector
motional Stark effect
NAG
NBI
NGPS
NSTX
NTM
Nuclear Analaysis Group
neutral beam injection
neutral gas and plasma shielding (code)
National Spherical Tokamak Experiment
neoclassical tearing mode
ODS
ORE
oxide dispersion strengthened
occupational radiation exposure
PAM
PEP
PFC
PIE
PIP
PLC
PLD
POLITO
PPCS
PRF
PTB
PuFF
PWHT
passive-active multijunction
pellet enhanced performance
plasma-facing component
postulated initiating event
polymer infiltration and pyrolysis
programmable logic controller
pulsed-laser deposition
Politecnico di Torino, Italy
power plant conceptual studies
permeation reduction factor
Physikalisch-Technische Braunschweig, Germany
Pulsed Field Facility
post-welding heat treatment
RAE
runaway electron

ABBREVIATIONS AND ACRONYMS
153
rf
RFX
RH
RI
radiofrequency
Reversed Field Pinch Experiment, Padua, Italy (Association EURATOM-
ENEA)
remote handling
radiative improved
SANS
SAW
SDS
SEM
SERF
Sex
SexUp
SOC
ssjs
SULTAN
small-angle neutron scattering
submerged arc welding
storage and delivery
scanning electron microscopy
socio economics research on fusion
Stability Experiment
Stability Experiment Upgrade
saturated Ohmic confinement
subsize joint samples
Superconductor Test Facility, Villigen, Switzerland (Association EURATOM-
Swiss Confederation)
T-10 Large Russian tokamak, Kurchatev Institute, Moscow
TAE
TCV
TCWS
TdeV
TEM
TEP
TEXTOR
TFC
TFMC
TIG
Tore-Supra
TPR
TUD
TZM
toroidal Alfvén eigenmode
Tokamak à Configuration Variable, Lausanne, Switzerland (Association
EURATOM-Swiss Confederation)
tokamak cooling water system
Tokamak de Varenne. Large Canadian tokamak
transmission electron microscopy
tokamak and exhaust processing
Torus Experiment for Technology Oriented Research. Tokamak at Jülich,
Germany (Association EURATOM-FZJ)
toroidal field coil
toroidal field model coil
tungsten inert gas (welding)
Large tokamak at Cadarache, France (Association EURATOM-CEA)
tritium permeation rate
Technical University of Dresden
tungsten-zirconium-molybdenum
UKAEA
United Kingdom Atomic Energy Agency
VRVS
Virtual Rooms Videoconferencing System

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ABBREVIATIONS AND ACRONYMS
WDS
WIMP
WS-NCS
XRD
water detritiation system
weak interacting massive particle
weak or negative central magnetic shear
x-ray diffraction