Overview of comprehensive characterisation of erosion zones on plasma facing components

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1 Fusion Engineering and Design 81 (2006) Overview of comprehensive characterisation of erosion zones on plasma facing components M.J. Rubel a,, E. Fortuna b, A. Kreter c, E. Wessel d, V. Philipps c, K.J. Kurzydłowski b a Alfvén Laboratory, Royal Institute of Technology, Association EURATOM VR, Teknikringen 31, S Stockholm, Sweden b Faculty of Materials Science and Engineering, Warsaw University of Technology, PL Warsaw, Poland c Institute for Plasma Physics, Forschungszentrum Jülich, Association EURATOM-FZJ, Trilateral Euregio Cluster (TEC), D Jülich, Germany d Institute for Materials and Processes in Energy Systems, Forschungszentrum Jülich, Association EURATOM-FZJ, D Jülich, Germany Received 29 January 2005; received in revised form 26 July 2005; accepted 26 July 2005 Available online 27 December 2005 Abstract Morphology of carbon plasma facing components retrieved from the TEXTOR tokamak after long operation periods and exposure to total particle doses exceeding m 2 was determined. Emphasis was on the composition and structure of the erosion zones. Tiles from two limiters the main toroidal belt pump ALT-II and auxiliary inner bumper were examined using high-resolution microscopy, surface profilometry, ion beam analysis techniques and energy dispersive X-ray spectroscopy. The essence of results regarding the net-erosion zones is following: (i) microstructure of surfaces is significantly smoother than on a non-exposed graphite, whereas carbon fibre composites show similar appearance prior to the exposure and after; (ii) deuterium retention is m 2 ; (iii) the presence of plasma impurity atoms (e.g. metals) is detected predominantly in small cavities acting as local shadowed areas on the surface. The results are discussed in terms of processes of material erosion/re-deposition and tokamak operation conditions influencing the morphology of wall components Elsevier B.V. All rights reserved. PACS: Hf Keywords: Graphite; Erosion; Plasma facing components; Fuel inventory; TEXTOR 1. Introduction Corresponding author. Tel.: ; fax: addresses: marek.rubel@alfvenlab.kth.se, rubel@kth.se (M.J. Rubel). Lifetime of plasma facing components (PFC) and the level of tritium inventory will be decisive for efficient and safe operation of a future reactor-class fusion device. Erosion of carbon is critical for that assessment /$ see front matter 2005 Elsevier B.V. All rights reserved. doi: /j.fusengdes

2 212 M.J. Rubel et al. / Fusion Engineering and Design 81 (2006) [1]. Therefore, processes associated with the erosion have been in focus of plasma material interaction studies in order to determine: (i) the rate of hydrocarbon formation under various operation scenarios [2], (ii) material migration in a tokamak [3,4] and (iii) the rate of carbon re-deposition together with fuel species, i.e. co-deposition [5,6]. The latter process is decisive for the in-vessel accumulation of vast quantities of hydrogen isotopes and, in case of a regular deuterium tritium operation, for an unacceptable level of tritium content [7]. Carbon-based components are considered for the divertor target plates in ITER. This implies that all aspects of material behaviour in present-day devices operated with carbon PFC must be studied in detail to determine from where and how much of material is eroded. Quantitative study of eroded thickness presents serious challenges. It has been shown that conclusive assessment can be approached only by using special instrumented tiles either with marker layers and/or precisely machined depth identers [6,8,9]. Until now, in PFC morphology studies efforts have been strongly concentrated on re-deposition areas (structure of co-deposits and long-term fuel retention) [5,6,8 14], whereas much less attention has been paid to properties and appearance of erosion zones. They have been considered less interesting because of low fuel retention and small content of co-deposited (coimplanted) plasma impurity atoms in such regions. Recent laboratory studies performed in the PISCES- B plasma wall interaction simulator [15] and also in RF plasma facility and ion-beam experiments [16] have shown significant changes in surface topography of graphite subjected to large doses (around m 2 ) of deuterium ions. The formation of strongly cratered surfaces and grass-like structures leading to a considerable increase of surface area has been observed [15,16]. No such drastic effects have been reported for real tokamak components. However, the verification of this issue was needed because it could be assumed that a significant change in surface topography of target plates in ITER would be detrimental in many ways to their performance: decreased lifetime and high hydrocarbon production resulting eventually in the increased fuel retention. A dedicated programme initiated in the framework of the EU Task Force on plasma wall interactions has been directed towards: (i) comprehensive characterisation of material modification following the exposure to the total particle (ions and charge-exchange neutrals) fluence exceeding m 2 and (ii) detailed comparison with the morphology of re-deposition zones. Studies of PFC after long operation periods were carried out. This paper is focused on the morphology of erosiondominated areas. We first address the fuel inventory in target plates and this is followed by characterisation of erosion zones. 2. Experimental: materials, exposure conditions and surface analysis methods Graphite and CFC target plates retrieved from the TEXTOR tokamak were examined. These were tiles of the toroidal belt pump limiter (ALT-II, Advanced Limiter Test II) and auxiliary inner bumper limiter. The toroidal view inside the TEXTOR vessel and the arrangement of components has been shown in [5]. ALT-II is the main limiter defining the minor radius at a = 46 cm. Limiter is composed of eight blades with 28 tiles each, the total area is 3.4 m 2. The corner tiles intercepting the greatest loads are made of felt type CFC (Toyo Tanso, CX-2002U) whereas all other tiles are made of isotropic graphite (Toyo Tanso, IG-430U). The inner bumper (Ringsdorf EK98 graphite) protecting the liner at the high-field side is about 3 cm deep in the scrape-off layer (SOL) at r = 49 cm. This difference in limiters position is decisive for particle fluxes to target plates in the two locations. ALT-II plates from three campaigns were studied. The duty time was from to s of plasma operation under various scenarios. The particle flux to the ALT-II was in the range (3 5) m 2 s 1 corresponding to the total fluence of (5 7.5) m 2. Bumper limiters tiles were facing the plasma for a similar period of time, but the dose was approximately five to six times smaller (around m 2 ) because of the short flux decay lengths in the SOL, λ = mm, as proven with several types of edge probes [17 20]. On many instances the extent of net erosion and net re-deposition zones on PFC can be determined by analysing interference fringes characteristic for thin films (less than 1 m) [21]. In case of flaking layers, the deposit thickness is measured using microscopy [6,22]. However, the most reliable distinction of zones is based on mapping by means of ion beam analysis (IBA) methods allowing for the determination of concentration and

3 M.J. Rubel et al. / Fusion Engineering and Design 81 (2006) depth profiles of species, i.e. deuterium and plasma impurity atoms on PFC [23,24]. Using accelerator-based IBA techniques the deuterium content was determined by means of the nuclear reaction analysis (NRA) with a 3 He + analysing beam of energy 0.7, 1.5 and 1.8 MeV: 3 He(d,p) 4 He. For these energies, the information depth in carbon materials is around 1.25, 4.5 and 6.2 m, respectively. NRA technique was also applied to quantify the content of boron [ 11 B (p, ) 8 Be] originating from the regular boronisation of the TEXTOR wall. With a proton beam of the energy of 0.65 MeV the analysis depth is around 6 7 m. Heavier impurities in the surface region were traced with Rutherford backsattering spectroscopy (RBS) and with energy dispersive X-ray spectroscopy (EDX). High-resolution scanning electron microscopy (SEM) gave insight into the topography of following materials: (i) non-exposed, i.e. asdelivered ALT-II (graphite and CFC) and bumper plates; (ii) sand-blasted ALT-II plate (graphite), i.e. a tile already used in TEXTOR and then cleaned ex situ for the re-installation; (iii) ALT-II CFC and graphite tiles exposed to the dose of about m 2 ;(iv) a bumper tile exposed to approximately m 2. Surface roughness was measured with the Dektak stylus profiler for all these plates. The lateral resolution of this technique is about 5 m, whereas the vertical resolution is in the nm range. 3. Results and discussion 3.1. Fuel inventory Plots in Fig. 1 show the deuterium distribution along the poloidal direction on two ALT-II tiles. The D contents are fairly similar in both cases. Very similar deposition pattern has been recorded for all other examined belt limiter tiles. There is a sharp transition from the erosion-dominated zone (about 2/3 of the tile area) to the deposition zone (remaining part of the tile). The results are for the layer accessible for NRA with a 1.5 MeV beam, i.e. around 4.5 m. The erosion zone contains little fuel (up to D atoms m 2 ). This amount is greater than can be expected for a direct implantation of low energy ions [25]. NRA depth profiling has detected the majority of deuterium in a nearsurface layer nm thick [23]. The total amount Fig. 1. Poloidal distribution of deuterium on two different ALT-II tiles. of deuterium in the whole erosion zone is assessed to be D atoms per tile corresponding to D atoms in the entire erosion-dominated area of ALT-II. Deuterium content sharply increases in the deposition zone reaching a fairly constant level of about Dm 2 (in a 4.5 m layer). This region has been characterised previously in detail: (i) the flaking layer thickness is up to m, (ii) the concentration ratio C D /C C 0.1 is quite typical for co-deposits at TEXTOR [22,26]. Sharp transition and distinct differences in the fuel content between the erosion and deposition have been reported for PFC at JET: limiters [27] and divertor components [14]. The same behaviour has also been observed on the inner bumper tiles at TEXTOR [28] but the layer in the deposition zone was relatively thin: 4 6 m Surface morphology of erosion zones and non-exposed tiles SEM images in Fig. 2 show typical overview (at small magnification) of surface topography features on the graphite plates (Fig. 2a c): the original nonexposed, a sand-blasted, in the erosion zone facing the plasma for s. Surface features of the original and exposed CFC are shown in Fig. 2d and e, respectively. A brief glance allows for a conclusion that the eroded of graphite and CFC plates have smoother appearance than the non-exposed or sand-blasted surfaces. Plots in Fig. 3a d are roughness profiles measured over a distance of 3 mm (note much larger scale than on SEM images) on surfaces of non-exposed and plasma-treated

4 214 M.J. Rubel et al. / Fusion Engineering and Design 81 (2006) Fig. 2. Surface topography of ALT-II plates: (a) non-exposed as-delivered graphite, (b) sand-blasted graphite, (c) erosion zone on the graphite plate facing the plasma for s, (d) non-exposed CFC, and (e) erosion zone of the CFC corner tile of ALT. plates. Roughness of graphite decreases after the exposure, whereas no significant change is observed on CFC. For a non-exposed bumper tile (not shown in the figure) the maximum roughness (hill-to-valley) was over 30 m, i.e. more than for graphite ALT-II plates. Small-scale surface features are shown in Fig. 4a h. The non-exposed surfaces (Fig. 4a, b, e and f) have distinct and fairly deep pits. Some material on edges of these cavities is loosely bound to the surface. This objects resemble small flakes and fine dust grains. If such particles are not removed by hovering the surface before the plasma operation, they can be a significant source of dust entering the plasma especially during the start-up phase of a tokamak discharge, as documented in [29]. On the contrary, the erosion zone is very smooth and free of dust (Fig. 4c, d, g and h). Images reveal a complete lack of cratered or grass-like structures reported in laboratory experiments [15,16]. Instead, there is a gentle and regular wavy structure. One notices, however, some remaining pits (cavities) as shown in Fig. 4c and g; this issue will be discussed later. Given the fact that flux densities onto the target in PISCES-B [30] and in TEXTOR are of the same order, the difference between the erosion effects in these experiments is probably attributed to the ion impact angle and the overall composition of particle fluxes. In the plasma wall interaction simulator or an ion-beam experiment the ion impact occurs at the right or somewhat oblique angle. In the plasma edge of a tokamak ions are tied and gyrate around the magnetic field lines which intersect the limiter surface at a very shallow

5 M.J. Rubel et al. / Fusion Engineering and Design 81 (2006) angle [31,32]. For the ALT-II plates the angle is below 1 [33] causing possibly a distortion of the electric sheath. Sheath properties under such conditions (below 5 ) are not fully understood and described. The issue certainly needs more theoretical efforts to be solved conclusively, but it is not a subject of this work. In our opinion, the shallow ion impact angle is very important but probably not the only role-plying factor responsible for topographical changes. The other factors are related to differences in parameters of the near-target plasma in a tokamak and a plasma wall interaction simulator: density, velocity distribution of particles and composition of the plasma. The TEXTOR plasma contains a few per cent of impurity species (at various ionisation degrees) with a Maxwellian velocity distribution: carbon, boron and metals eroded from the wall [34]. There is also a significant fraction of charge-exchange neutrals contributing by 30 50% to the total particle flux [35]. According to spectroscopy data from PISCES, the deuterium plasma in front of the target contains very small amount of impurities [36]. Surface temperature of target plates has also been considered as a factor influencing the topography. Eventually, it can be excluded because the ALT-II temperature in quiescent operation does not exceed 600 C [33,37]. Rapid thermal excursions leading to material ablation (over 2200 C for carbon) have not been observed. In conclusion, we suggest that the observed structure in the erosion-dominated zone on graphite targets at TEXTOR is a result of milling caused by a mixture of deuterium and other species impinging on the surface with a broad energy distribution. This includes also ions impacting at a shallow angle. The highly ordered wavy structures are the remains of hills on the original surface (compare to Fig. 2a) after being ion milled Material migration and localised re-deposition in erosion zones Fig. 3. Examples of the Dektak stylus profiler scans along the ALT- II limiter surface: (a) non-exposed graphite tile, (b) graphite tile exposed for s, (c) non-exposed CFC corner tile, and (d) exposed CFC corner tile. Note different height scales for the graphite and CFC tiles. Erosion of the ALT-II blades is the major source of carbon impurities in TEXTOR. The total erosion amounts to 22gCh 1 as determined in meticulously planned experiments with instrumented tiles followed by surface measurements [6]. Taking into account that erosion rate, the total surface area of the erosion zone (2.25 m 2 ), atomic density of graphite ( cm 3 ) and the plasma operation time of s one estimates

6 216 M.J. Rubel et al. / Fusion Engineering and Design 81 (2006) Fig. 4. High-resolution SEM images of: (a) and (b) non-exposed graphite tile of ALT-II, (c) and (d) erosion zone on a graphite plate exposed to the plasma for s, (e) and (f) non-exposed CFC tile, (g) and (h) erosion zone on a CFC plate exposed for s. the total eroded layer thickness of m. Removal of this thickness is then enough to smooth the original tile surface but deeper pits still remain on the eroded plates, as noted in Figs. 2c and e and 4 c and g. EDX examination of a bumper limiter tile proves significant concentration of metal atoms in the pit as shown in Fig. 5. Nickel and other metals (Cr, Fe, Mo) originate from the TEXTOR Inconel liner. Point analyses inside and outside the pit, have clearly indicated: (i) detection of only carbon on the smooth part of the tile, (ii) highly localised re-deposition of impurity atoms inside the cavity. Backscattered electron image in Fig. 6 recorded for the exposed CFC tile proves that the result discussed above is not an isolated example of impurities

7 M.J. Rubel et al. / Fusion Engineering and Design 81 (2006) plates of ALT-II. It allows for the statement that impurity presence in the erosion-dominated zone is limited only to the pits whose walls create a trap for atoms once deposited there. This proves inhomogeneous distribution of material and explains the presence of heavy species in that region. Plasma impurities (metals, B, Si, etc.) have been observed with standard IBA techniques, but the lateral has not allowed for micro-mapping. Nonuniform distribution of boron (not traced with EDX) has been observed previously [38], but the association between the boron location and topographical structures has not been determined in detail. One may also expect that more deuterium is retained in cavities and other shadow-creating surface imperfections than on flat parts of the erosion-dominated area. This hypothesis will be verified in the future using NRA microprobe. 4. Concluding remarks Fig. 5. A pit in the erosion zone of the bumper limiter tile and X-ray mapping of plasma impurity species (nickel) in that region. re-deposition. It is a general feature over that plasmaexposed surface. Accumulation of metals (Ni, Cr, Fe, Mo) and Si in all cavities and their edges has been confirmed by EDX on all examined graphite and CFC Fig. 6. Backscattered electron image of the exposed CFC corner tile. Mass contrast indicates heavier atoms (light) accumulated on the edges and inside pits in the carbon matrix (dark field). This work brings two important contributions improving the understanding of material erosion, migration and its re-deposition in tokamaks. The main result, obtained with microscopy methods, clearly proves that the surface in the erosion zone exposed to a high dose of high-flux plasma is neither damaged nor converted into a cratered structure. On the contrary, on micro-scale the surfaces are much smoother than those of virgin non-exposed or sand-blasted plates. Smooth surfaces in erosion zones have also been seen when inspecting limiter and divertor CFC tiles from JET. The second important point is the proof of nonuniform distribution of minority atoms in the erosion zone. Small cavities still remaining on the eroded surface can be considered as local shadowed areas in which accumulation of transported material takes place [6,39]. These detailed surface studies have explained why, how and where the migrating impurity species can be re-deposited even in the net-erosion part of PFC. The study was carried out with targets exposed to a tokamak plasma. Though the temperature regime has been different from that expected at the strike points in a next-step device, the angle of particle incidence with the target will be similar. Therefore, overall results could indicate that carbon PFC in the net-erosion region of a future device might show similarly smooth appearance. This research has also identified several important issues for future studies using experimental and

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