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Please note that terms and conditions apply.Beryllium plasma-facing components for the ITER-like wall project at JET
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Beryllium plasma-facing components for the ITER-Like Wall
Project at JET
M.J. Rubel1, V. Bailescu2, J.P. Coad3, T. Hirai4, J. Likonen5, J. Linke4, C.P.
Lungu6, G.F. Matthews3, L. Pedrick3, V. Riccardo3, P. Sundelin1, E. Villedieu3
and JET-EFDA Contributors*
1Alfvén Laboratory, Royal Institute of Technology, Association Euratom-VR, Sweden
2Nuclear Fuel Plant, Pitesti, Romania
3Culham Science Centre, Euratom-UKAEA Fusion Association, Abingdon, UK
4IEF-2, Forschungszentrum Jülich, Association Euratom-FZJ, Jülich, Germany
5VTT, Association Euratom-Tekes, 02044 VTT, Finland
6NILPRP, Association Euratom-MEdC, Bucharest, Romania
*See Appendix to the paper: M.L. Watkins et al, 21st IAEA Fusion Energy
Conference, Chengdu, China 2006 (IAEA Vienna)
[anonimizat]
Abstract . ITER-Like Wall Project has been launched at the JET tokamak in order to study a
tokamak operation with beryllium components on the main chamber wall and tungsten in the
divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion
environment, a broad program has been undertaken to develop plasma-facing components and
assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.
1. Introduction
The Joint European Torus (JET) is the largest pr esent-day tokamak, i.e. a magnetic controlled
thermonuclear fusion device for energy research. It s main scientific mission is to develop plasma
operation scenarios for a reactor-class machine such as ITER [1]. Equally important is to test the
performance of plasma-facing components (PFC). At present, most components are made of carbon
fibre composites (CFC). JET is fully compatible with operation using deuterium-tritium mixture and
beryllium PFC, which are key features for the next-s tep fusion device. To achieve further progress in
controlled fusion, the ITER-Like Wall (ILW) Project at JET is under way in order to explore tokamak
operation and plasma-wall interactio n processes with a full metal wa ll: beryllium (Be) in the main
chamber and tungsten (W) in the divertor [2-4]. The main driving forces for a large scale test of the
metal wall are: (i) expected reduced retention of hy drogen isotopes in operation with a metal wall in
comparison to carbon PFC; (ii) good plasma performance and gettering of oxygen impurities by beryllium; (iii) low erosion of tungsten at low io n temperature in the divertor [5]. Experimental
campaigns with the fully modified PFC structure are planned to begin in year 2010.
The aim of this paper is to overview scientific an d technical issues related to the development of Be
components of two major categories: (i) bulk limiter tiles including so-called markers designed for
studies of beryllium erosion from the wall and (ii) Be-coated inconel plates for the inner wall cladding.
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Journal of Physics: Conference Series 100(2008) 062028 doi:10.1088/1742-6596/100/6/062028
c/circlecopyrt2008 IOP Publishing Ltd
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2. Bulk beryllium components
Images in figure 1 show the present structure of the JET in-vessel components (a) and the distribution of materials to be implemented for ILW (b). Beryl lium tiles are to be located in the main chamber
wall. These are the inner wall guard limiter and the outer poloidal limiters, lower hybrid launcher
frame, upper dump plates and other protection tiles (antenna private limiters, mushroom tiles, saddle
coil protection tiles). Dump plates and mushroom-shaped limiters protect the upper part of the vessel.
The size of the main limiter tiles (approx. 10×30 x6 cm) has imposed the search for engineering
solutions to ensure proper performance of the lim iters. Figure 2 provides details of a wide poloidal
limiter tile assembly consisting of seven bulk Be segments (Brush Wellman Inc. grade S65J, hipped structural Be) installed on a vacuum cast Inconel-625 carrier. The segmented construction reduces eddy currents, whereas the castellation is to improve thermal durability under heat loads. The
optimized surface profile and lack of plasma-faci ng bolt holes ensure better power handling.

Figure 1. View inside the JET vessel: (a) present structure of wall components with CFC limiter and
divertor tiles and Be ICRH Faraday screens; (b) pl anned distribution of beryllium and tungsten for the
ITER-Like Wall operation.

Figure 2. Structure of a carrier and a segmented Be tile (a); assembly of wide poloidal limiter.
3. Beryllium marker tiles
An important goal of the ILW Project is to assess th e erosion of beryllium components in order to give
best-possible predictions for ITER. To facilitate such studies, so-calle d marker tiles are being
developed. They will be placed in several toroidal and poloidal locatio ns in the vessel. A marker is a
regular beryllium tile coated first with a high-Z metal film acting as an interlayer and then with a Be
layer of density similar to that of bulk beryllium. To ensure good adherence and thermo-mechanical
(best match of linear thermal expansion coefficients) and physical properties of the marker coatings
nickel (2-3 μm) was selected as an interlayer material to separate the bulk Be tile from a 7-10 μm thick
CFC
ICRH
Faraday
screen
a b
carrier segmented tile
IVC-17/ICSS-13 and ICN+T2007 IOP Publishing
Journal of Physics: Conference Series 100(2008) 062028 doi:10.1088/1742-6596/100/6/062028
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beryllium coating. The films are obtained by the thermionic vacuum arc (TVA) method [6] which
allows production of high-density layers. Fo r measurements of erosion greater that 10 μ m, there will
be precise notches (10, 20 μ m deep) on the tile surface. A series of marker coupons were produced and
examined by several material analysis techniques before and after high-heat flux (HHF) testing with
an electron beam in the JUDITH facility. HHF screening tests allowe d the determination of power and
energy density limits deposited onto the surface until the damage to a marker occurred. A cyclic test
served to assess the thermal fatigue under repetitive power loads. Not coated Be blocks were tested for
comparison. The major results may be summarised by the following: (i) the markers survived without
noticeable damage power loads of 4.5 MW m
-2 for 10 s (energy density 45 MJ m-2) and fifty repetitive
pulses performed at 3.5 MW m-2 each lasting 10 s, i.e. corresponding to the total energy deposition of
1750 MJ m-2; (ii) in both cases the surface temperature measured with an infrared camera was around
600 oC; (iii) the damage to th e Be coating occurred at power loads of 5 MW m-2 for 10 s.
Plots in figure 3 show depth profiles obtained by secondary ion mass spectrometry (SIMS) for two
marker coupons: (a) unexposed to heat loads and (b) after HHF test carried out for 10 s at power density of 4 MW m
-2, i.e. total energy density of 40 MJ m-2. Both profiles are quite similar (Be coating
thickness ~9.5 μm) thus indicating that the applied power loads neither damage the coating nor cause
intermixing of Be and Ni. There are some impurity sp ecies (Al, Si, Fe) but th eir content is below 1 %
as determined by ion beam analysis, energy and wa velength dispersive X-ray spectroscopy. Figure 4
shows a metallographic cross-section of the HHF tested coupon. A clear separation of beryllium and nickel proves the durability of the coatings.

Figure 3. SIMS depth profiles for two markers: (a) “as produced”; (b) HHF tested at 40 MJ m
-2.

Figure 4. Metallographic cross-sec tion of a marker coupon
after cyclic test (50 pulses) at 3.5 MW m
-2. Depth (μm)02468 1 0 1 2 1 4Intensity (s-1)
100101102103104105Be
C
Al
Si
Ca
Fe
Ni
Depth (μm)02468 1 0 1 2 1 4Intensity (s-1)
100101102103104105Be
C
Al
Si
Ca
Fe
Ni
a b
Ni
BeO
C
Ni
BeO
C
Ni
BeO
CEpoxy resin
Be coating
Bulk Be Ni interlayer
IVC-17/ICSS-13 and ICN+T2007 IOP Publishing
Journal of Physics: Conference Series 100(2008) 062028 doi:10.1088/1742-6596/100/6/062028
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4. Beryllium coatings on Inconel
The inner wall cladding and the dump plate tile carriers will be made of cast Inconel. These tiles are in the shadow of bulk Be tiles, but to minimize the risk of high-Z impurity (Ni, Cr, Fe) influx, the
Inconel tiles will be protected by 8 μm thick evaporated Be coatings.
During regular plasma operation in JET, the esti mated power load to the cladding is 0.5-0.7 MW
m
-2 for 10 s corresponding to energy deposition of 5-7 MJ m-2. To check the adherence and thermo-
mechanical properties of the Be layer, a number of test coupons were exposed to high power loads in
JUDITH [7]. The screening test was ca rried out in the range from 0.4 MW m-2 to 2.6 MW m-2 in
pulses lasting of up to 11 s. In the cyclic test fift y consecutive 10 s pulses were performed at the power
of 1 MW m-2, i.e. 10 MJ m-2 per pulse. Figure 5 shows the layer st ructure before (a) and after the test
at the power load of 1.8 MW m-2 for 11 s corresponding to the energy load of 20 MJ m-2 (b). In both
cases the coating topography is nearly identical. It proves that no damage (e.g. melting or exfoliation)
is caused by energy loads exceeding at least thr ee times the level characteristic for a regular plasma
operation. As assessed, the coating on Inconel would melt at energy loads exceeding 30 MJ m-2 [7].

Figure 5. Beryllium coatings on Inconel: (a) “as pr oduced”; (b) after HHF test of 20 MJ m-2.
5. Concluding remarks
The best efforts have been taken to develop and te st the performance of beryllium components being
prepared for the installation in the ILW operation of JET. Power handling capabilities and purity have
been of primary interest. The results of material anal ysis before and after HHF testing indicate that the
coatings on Inconel and marker limiters should withstand conditions of the regular JET operation
without melting, exfoliation or phase transformation. This is particularly important in case of the
marker tiles for long-term Be erosion studies in the main chamber. However, local melting of Be tiles (with and without markers) cannot be excluded in ca se of events resulting in deposition of excessive
power loads. In this case the extent of erosion w ill be assessed by mechanical methods. The scientific
and technical program has led to th e selection of methods for a large-scale manufacturing of protective
coatings on the inner wall cladding and marker tiles. The thickness of markers, prior to their installation in JET, will be determined by means of ion beam analysis methods.
References
[1] Ikeda K et al “ITER Physics Basis” 2007 Nucl. Fusion 47 (6)
[2] Matthews G F et al 2007 Phys. Scr. T 128 137
[3] Maier H et al 2007 Nucl. Fusion 47 222
[4] Matthews G F et al 2007 These Proceedings
[5] Federici G et al 2001 Nucl. Fusion 41 1967
[6] Lungu C P et al 2007 Phys. Scr. T 128 157
[7] Hirai T et al 2007 Phys. Scr. T 128 166 a b
50 μm 50 μm
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Journal of Physics: Conference Series 100(2008) 062028 doi:10.1088/1742-6596/100/6/062028
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