Fuel inventory and deposition in castellated structures in JET-ILW

PAPER Fuel inventory and deposition in castellated structures in JET-ILW To cite this article: M. Rubel et al 2017 Nucl. Fusion 57 066027 View the article online for updates and enhancements. Related content Fuel Inventory and Co-Deposition in Grooves and Gaps of Divertor and Limiter Structures - Overview of the JET results in support to ITER - Be ITER-like wall at the JET tokamak under plasma - Recent citations Comparison of Hydrogen Isotope Retention in Divertor Tiles of JET with the ITER-Like Wall Following Campaigns in 2011–2012 and 2015–2016 Yasuhisa Oya et al - Fuel inventory and material migration of JET main chamber plasma facing components compared over three operational periods A Widdowson et al - Fuel inventory and impurity deposition in castellated tungsten tiles in KSTAR: experiment and modelling Sunwoo Moon et al - This content was downloaded from IP address 87.218.223.151 on 07/08/2020 at 12:43 1 © EURATOM 2017 Printed in the UK 1. Introduction All plasma-facing components (PFC) in ITER will be castellated because this structure is considered to be the best solution to ensure thermo-mechanical durability and integrity of materials under high heat flux loads [1]. However, in the environment containing also low-Z elements (e.g. beryllium and some impurities of carbon), material eroded by various processes is transported and co-deposited together with fuel species in areas shadowed from the direct plasma line-of-sight. As a consequence, re-deposition and material mixing may occur in grooves and on the side surfaces of the PFC, e.g. on surfaces located in the gaps separating the tiles. The issue may be quite Nuclear Fusion Fuel inventory and deposition in castellated structures in JET-ILW M.Rubel1,b, P.Petersson1, Y.Zhou1, J.P.Coad2, C.Lungu3, I.Jepu3, C.Porosnicu3, D.Matveev4, A.Kirschner4, S.Brezinsek4, A.Widdowson2, E.Alves5 and JET Contributorsa EUROfusion Consortium, JET, Culham Science Centre (CSC), Abingdon, OX14 3DB, United Kingdom 1 Department of Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm, Sweden 2 CCFE, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom 3 Institute of Atomic Physics, NILPRP, 077125 Bucharest-Magurele, Romania 4 Institute of Energy and Climate Research (IEK-4), Forschungszentrum Jülich, D-52425 Jülich, Germany 5 Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa, Portugal E-mail: [email protected] Received 17 December 2016, revised 1 March 2017 Accepted for publication 22 March 2017 Published 26 April 2017 Abstract Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in the divertor. This has allowed for a large scale test of castellated plasma-facing components (PFC). Procedures for sectioning the limiters into single blocks of castellation have been developed. This facilitated morphology studies of morphology of surfaces inside the grooves for limiters after experimental campaigns 2011–2012 and 2013–2014. The deposition in the 0.4–0.5 mm wide grooves of the castellation is ‘shallow’. It reaches 1–2 mm into the 12 mm deep gap. Deuterium concentrations are small (mostly below 1 × 1018 cm−2). The estimated total amount of deuterium in all the castellated limiters does not exceed the inventory of the plasma-facing surfaces (PFS) of the limiters. There are only traces of Ni, Cr and Fe deposited in the castellation gaps. The same applies to the carbon content. Also low deposition of D, Be and C has been measured on the sides of the bulk tungsten lamellae pieces. Modelling clearly reflects: (a) a sharp decrease in the measured deposition profiles and (b) an increase in deposition with the gap width. Both experimental and modelling data give a strong indication and information to ITER that narrow gaps in the castellated PFC are essential. X-ray diffraction on PFS has clearly shown two distinct composition patterns: Be with an admixture of Be–W intermetallic compounds (e.g. Be22W) in the deposition zone, whilst only pure Be has been detected in the erosion zone. The lack of compound formation in the erosion zone indicates that no distinct changes in the thermo-mechanical properties of the Be PFC might be expected. Keywords: JET, ITER-like wall, beryllium limiters, castellation, deposition, fuel inventory (Some figuresmay appear in colour only in the online journal) M. Rubel etal Printed in the UK 066027 NUFUAU © EURATOM 2017 57 Nucl. Fusion NF 10.1088/1741-4326/aa6864 Paper 6 Nuclear Fusion IOP International Atomic Energy Agency a See the author list of ‘Overview of the JET results in support to ITER’ by X. Litaudon etal to be published in Nuclear Fusion Special issue: overview and summary reports from the 26th Fusion Energy Conf. (Kyoto, Japan, 17–22 October 2016). b Author to whom any correspondence should be addressed. 2017 1741-4326 1741-4326/17/066027+7$33.00 https://doi.org/10.1088/1741-4326/aa6864 Nucl. Fusion 57 (2017) 066027 (7pp) M. Rubel etal 2 serious if significant amounts of fuel were accumulated in the gaps. The composition of deposits and their pattern in the gaps is crucial for the overall assessment of fuel inventory as the area of surfaces in the castellation in ITER will be approximately three times larger than the area of plasma-facing surfaces (PFS). Based on the earlier experience from JET, co-deposits formed in remote areas (e.g. inner divertor corner and water-cooled louvers) [2–5] would be very difficult to remove by any of the cleaning methods known today [6]. This, in turn, may lead to significant long-term fuel retention. In addition, very limited access to in-vessel comp onents of ITER calls for detailed studies of deposition in the gaps and grooves of the PFC from present-day devices. All aspects related to the fuel content and the surface structure in gaps must be taken into account. Deposition in the gaps was noticed already in the 1980s [7–10] and some quantitative results for TFTR and DIII-D have been published [10]. In TFTR 15% of the retained fuel was found in gaps, 44% on the PFS and 41% on the outer wall, whereas in DIII-D accumulation in gaps between the tiles reached 40% of the retained fuel [9], but no information has been given about the gap geometry, especially the gap width. Large-scale tests of castellated structures have been performed in the past at JET when operating in 1989– 1992 with beryllium belt limiters [11] and in 1994–1995 with Mk-I divertors [12–14]. The latter were composed of arrays of water-cooled small tiles made of carbon-fibre composites (Mk-I-CFC) and, at a later stage, were replaced with castellated beryllium (Mk-I-Be). The tiles were 35 × 72 mm in the PFS and 55 mm high. The gaps between the tiles were 6 and 10 mm wide, while the width of the castellation grooves in the Be tiles were 0.5 mm wide. In Mk-I-CFC the retention on surfaces between the tiles exceeded at least by a factor of two [13, 14] the amount of fuel on the PFS [12]. In the case of Mk-I-Be, the fuel content in the gaps separating the tiles was 30 times bigger than inside the castellation, where the deposition was shallow and had a steep gradient, i.e. the decay length of 1.5 mm. For technical reasons the examination inside the castellation was limited to only one groove. Broader studies were performed for several grooves of castellated Be limiters. In all cases, in both the divertor and limiter tiles the deposition was dominated by carbon and the profiles were steep. Dedicated studies with various castellated probes (test limiter or divertor tiles) have been performed in several tokamaks [15–21]; some of them have been summarized in [22]. This has been followed by modelling which included both the development of advanced codes, e.g. 3D-GAPS [23–26] and particle-in-cell calculations [27, 28]. The efforts were focused on the reproduction of the results obtained in experiments comprising a limited number of discharges performed under a defined set of operation conditions. Currently a large-scale test of a castellated PFC is being carried out at the JET tokamak which has been in operation since the year 2011 with the metal ITER-like wall (JET-ILW): beryllium limiters and tungsten divertor tiles [29–34]. This work is focused on the morphology of the castellated beryllium structures in JET-ILW: upper dump plates (UDP), Wide and Narrow Outer Poloidal (WOPL and NOPL, respectively) and inner wall guard limiters (IWGL) after two experimental campaigns in 2011–2012 (ILW-1) and in 2013–2014 (ILW-2). Each of those campaigns lasted for about 19.5 h of plasma discharges including approx. 13 h of limiter and 6.5 h of x-point operation. There were, however, differences in the total input energy (ohmic, neutral beams, ion cyclotron, lower hybrid): 150.6 GJ in ILW-1 and 200.5 GJ in ILW-2. The emphasis in the study was on: (i) material mixing and compound formation on the PFS; (ii) fuel inventory and deposition inside the grooves of the castellation and on the side surfaces of the tiles, i.e. in areas located in the gaps; (iii) modelling of material transport and deposition in the gaps. 2. Experimental approach and modelling The images in figures1(a)–(c) show a segmented structure of a WOPL and the appearance of two types of the castellated limiters in JET, while the data collected in table 1 provide detailed information on the number of the various limiters and their respective castellation blocks. As inferred from these data, the total number of castellated blocks is nearly 170 000, the length of surfaces inside the castellation (both sides of the groove) is 7325 m, with an area of nearly 88 m2. Therefore, the area is distinctly greater than the area of the PFS of the limiters: 24.5 m2. The area of the side surfaces (in the gaps between tiles) is approximately 12.4 m2, while the total length along both sides of all tiles is about 425 m. The condition sine qua non for studies of castellated structures was the sectioning of the tiles in order to expose the surfaces located inside the grooves. The limiter blocks were sectioned into smaller specimens: single castellation pieces, i.e. typically 12 × 12 × 12 mm. Cutting was carried out at the Institute of Atomic Physics (Bucharest, Romania) under strict control of the tile temperature (below 60 °C) in order to avoid the release of hydrogen isotopes, because thermal desorption of fuel species from several samples was planned. In general the cutting was done approximately 0.5 mm above the bottom of the castellated groove, but in a few cases cutting was performed approx. 0.5 mm below the bottom of castellation. The latter samples were then split to expose the all of the surfaces located in the groove. This was to check the material transport to the very bottom of the groove. The analyses described in sections3.1 and 3.2 were performed by means of a number of complementary techniques: (a) x-ray diffraction (XRD) in order to determine the phase composition of limiter surfaces; (b) micro ion beam analysis (µ-IBA) techniques (lateral resolution of 8–10 µm) such as nuclear reaction analysis (µ-NRA) to determine the content of deuterium and particle-induced x-ray emission (µ-PIXE) to quantify the content of metals inside the castellation: tungsten and Inconel constituents, i.e. Ni, Cr, Fe. From the top of the castellated groove several consecutive regions, 1.8 × 1.8 mm, were scanned. Results from a given region were averaged to obtain a line scan and this procedure was repeated for all analysed regions. It is stressed that a complete analysis of one sample takes between 14 and 44 h. Therefore, only a limited number of specimens could be probed. The analysis was preceded by a very thorough selection of specimens to be Nucl. Fusion 57 (2017) 066027 M. Rubel etal 3 examined in order to get the best possible overview of deposition and retention. Poloidal and toroidal gaps from all major types of limiters have been studied with ion beams: IWGL, OPL and UDP. Specimens from other components (e.g. the protection screen of the ion cyclotron antennae and saddle coils) were not available for studies. For comparison also the side surfaces of the bulk tungsten lamellae from the JET-ILW divertor (Tile 5) were analysed. Modelling of Be re-deposition and respective D influx within the poloidal gaps of a relevant depth of 12 mm and three different widths (0.5, 1.0 and 2.0 mm) was performed with the 3D-GAPS code [23, 24]. 3. Results and discussion 3.1. Intermetallic compounds on PFS It has been shown in earlier studies of the limiter and divertor tiles from JET-ILW that—as expected—the surfaces contain a mix co-deposited species: fuel atoms, beryllium, carbon, nitrogen and metals [35–39]. However, no information on the chemical form has been given, while the point should not be neglected in the environment containing both Be and W which are known to form binary mixed materials of low melting point [40]. Herewith, the results of the first-ever studies are presented. A whole limiter tile after exposure in JET-ILW is shown in figure 2(a) while plots in figure 2(b) show the diffractograms recorded for a reference Be target and in the deposition and erosion zones of the limiter. The results clearly prove that some Be–W intermetallic compounds (Be2W, Be12W, Be22W) were formed in the deposition zone. The erosion zone contains only metallic Be; the diffractogram has the same features as that for the reference Be sample, thus Figure 1. Structure of segmented inner wall guard limiter with castellated Be blocks (a); lower hybrid antenna protection (b); UDP (c). Table 1. Detailed data on castellated beryllium limiters in JET-ILW. Limiter type Number of tiles Number of castellated blocks Length along castellation (cm) Inner wall guard 217 43 000 191 134 Wide outer poloidal 225 50 013 216 056 Narrow outer poloidal 57 8126 38 426 UDP 448 40 448 186 624 Saddle coil 232 26 208 91 123 Lower hybrid protection 4 1408 9000 Totals 1183 168 203 732 363 Figure 2. Castellated beryllium limiter tile from JET-ILW (a); x-ray diffractograms recorded for the initial limiter surface, erosion and deposition zones (b). Nucl. Fusion 57 (2017) 066027 M. Rubel etal 4 proving the absence of the binary Be–W materials. The latter result is perceived as a positive one: no compound formation in the erosion zone indicates that no changes in the thermomechanical properties of the Be PFC could be expected. 3.2. Deposition in the castellation and gaps between tiles The image in figure3(a) shows the appearance of the surfaces in the castellated grooves, while in figures3(b) and (c) there are side surfaces of limiter tiles, i.e. the surface located in the gap between tiles. In all cases one perceives deposition belts, marked with red arrows, in the top part of the tiles, i.e. at the very entrance to the groove. It is noticed, however, that the extent of deposition (i.e. width of the belts) in the castellation is significantly narrower than that on the side surfaces of the tiles. For the castellated grooves µ-IBA was performed on more than 70 specimens from the top to the bottom of the groove. The plots in figure4 show the deposition profiles of the deuterium and metals (the plot for metals is magnified by a factor of 1000) in a toroidal gap of the IWGL (first ILW campaign 2011–2012). The deposition width of the deuterium is approximately 1 mm and this is characteristic for the profiles recorded in narrow gaps. The profile has a fine structure: (i) low D content at the very entrance to the gap, (ii) increase in the concentration with a maximum reached at about 0.5 mm and then a sharp decrease. The plots in figure5 show a comparison of depositions in two perpendicular gaps, toroidal and poloidal, on the same specimen from the OPL. The differences are insignificant both in the shape of the profiles and the deuterium content. Also the content of the nickel from the Inconel eroded from the wall is very small: well below 1 × 1015 cm−2, while the level of tungsten does not exceed 2 × 1013 cm−2. Most of the profiles measured after the two campaigns are qualitatively and quantitatively very similar to those presented in figures4 and 5. There are a few different cases: (i) flat profiles with a very small D content, below 1 × 1017 cm−2; (ii) very narrow profiles, less than 0.5 mm, peaked at 1 × 1019 cm−2. The latter is presented in figure6. It shows a deposition from the top to the very bottom of the castellated groove. The measurement was performed in order to verify the hypothesis on the deposition at the bottom which may act as the ultimate trap for neutrals which either entered the gap at the right angle or reached the bottom as a consequence of multiple reflections. Indeed, a small and narrow peak is detected at the bottom, but the deuterium content is very low. It is, therefore, not decisive for the overall inventory. It should also be stressed that the amount of carbon is very low thus confirming a small amount of carbon impurities in JET-ILW. As already mentioned in section2, the selection of samples for µ-IBA examination was very thorough to get the best possible overview of the deposition and retention in the castellation grooves of tiles from various regions of JET-ILW. A summary of the results is given in table2, where one also finds a comparison of the total retention on PFS and in the grooves. The estimated total retention in the castellation is in the range of 0.7 × 1022 to 14.6 × 1022 D atoms. The upper value is on the same level as that found for PFS. It should be stressed that these quantities are low as they correspond in total to about 1 g of deuterium retained in the limiters on the main chamber wall. The plot in figure 7 shows deuterium deposition on the side surface, i.e. in the 2.0–2.5 mm wide gap between the two tiles of the IWGL. One may notice some important features. The profile is broad; it extends over 20 mm into the gap. The D concentration reaches its maximum approximately 5 mm deep in the gap and then it sharply decreases. The maximum is 1.4 × 1018 cm−2 and the average is about 0.5 × 1018 cm−2, i.e. it is not greater than that measured in the narrow grooves castellated. Taking into account the length of the gaps along all tiles and the width of the deposition belts on the tile Figure 3. Deposition in the castellated groove (a); deposits of different width on the side surfaces of the tiles, in the gaps between the tiles (b) and (c). Figure 4. Deposition profiles of deuterium and metals inside the castellated groove of a beryllium limiter. Nucl. Fusion 57 (2017) 066027 M. Rubel etal 5 surfaces, from 3.5 to 30 mm wide, one may estimate the total fuel content in the gaps in the range 0.6–3.0 × 1022 D atoms, corresponding to 0.02–0.1 g of deuterium. These amounts are low in general and they are lower than assessed for the castellated grooves, but it must be stressed that the length of all gaps between the tiles (425 m) is over 17 times smaller than the lengths in the castellation. Therefore, all data give a strong indication that tiles or their segments should be separated by narrow grooves or gaps to reduce fuel retention. 3.3. Modelling The drawings in figures8(a) and (b) show the geometry of the gaps for which modelling was performed; ion flux to the gaps is marked in red, while the results are plotted in figures8(c) and (d ). The influx of D and Be ions into the poloidal gaps and respective Be re-deposition there were modelled with the 3D-GAPS impurity transport code [23, 24]. In the simulations, a poloidal gap of a relevant depth of 12 mm is assumed. The effect of the gap width on the re-deposition is addressed. Since a detailed simulation of the entire campaign with different operation scenarios and varying plasma parameters and fluxes is not feasible, the simplest case was simulated assuming a projected along magnetic field lines plasma flux and an additional uniform isotropic neutral flux to the gap aperture. Such a setup allows for a linear scaling of the particle influx into the gaps with plasma flux, impurity content, gap width and magnetic field Figure 5. Deposition profiles of deuterium and metals in two perpendicular gaps of the outer poloidal limiter. Figure 6. Deposition profiles of deuterium and metals along the entire depth of the castellated groove. Table 2. Comparison of deuterium retention on the PFS and in the castellation and on the side surfaces of the limiters. Plasma-facing side Inside castellation Side surfaces Surface are (m2) 24.5 87.9 12.4 Deuterium content (1022) 12.4 [10] WOPL, UDP, IWGL 0.7–14.6 All grooves 0.6–3.0 All gaps between tiles Figure 7. Deposition profile of deuterium on the surface located in the gap between the two limiter tiles. Nucl. Fusion 57 (2017) 066027 M. Rubel etal 6 angle. A parallel plasma flux of 2 × 1017 D+ cm−2 s−1 with the ion impact energy of 100 eV was assumed in the simulations. Field lines inclination was set to 4.5 degrees with respect to the wall surface. The neutral D flux isotropically entering the gap aperture was set to 2 × 1016 D0 cm−2 s−1 with the impact energy of 10 eV. In both fluxes a Be-fraction of 1% was assumed. In such a setup, the total neutral influx appears to be 25% higher than the total ion influx due to the ratio between the plasmawetted area (0.04 mm for 0.5 mm gap width) and the gap width. This ratio remains the same for all gap widths since both, the gap aperture and the projected plasma-wetted area, increase proportionally. Reflected particles were set to follow the cosine angular distribution, and the reflection coefficient was fixed to 0.2 for D and chosen according to Eckstein [41] for Be. Energy reflection coefficients of 0.1 for D and 0.3 for Be were imposed. Sputtering of bulk Be by incoming ions was taken into account. No D recycling was considered, thus the maximal D deposition (influx) was simulated. This gross D amount cannot and should not be compared with much lower measured D content in deposited layers. Recycling during plasma operation and outgassing of dynamically retained D during the long storage time until actual measurement reduce the D content by orders of magnitude. The resulting Be re-deposition and D influx profiles along the gap side from the top (near plasma) to the bottom of the gap are shown in figures8(a) and (b). The simulation results are shown for the plasma-shadowed side of the gap and for the gap bottom. It has to be noted that absolute comparison with experimental results is not possible since Be re-deposition on Be cannot be measured while the D content in re-deposited layers is not directly assessable in simulations, as it was explained above. Nevertheless, an empirical scaling of D content in codeposits with Be exists [42] that may be used to estimate the total D content based on the re-deposition of beryllium. The main message, however, is that the simulations show a qualitatively similar picture as measurements with a fairly steep decay of deposition inside the gaps and a relatively small deposition near the gap bottom. Profiles for broader gaps are distinctly less sharp and the quantities deposited are significantly increased: the deposition increases proportionally with the gaps width. This is because of the increased particle fluxes reaching the gap. In summary, the modelling fully reflects the main features of the measured deposition profiles both for the grooves of the castellation (0.5 mm) and the broader gaps (approx. 2.0–2.5 mm) between the tiles. 4. Summary and concluding remarks This work has brought several important results which have provided insight into deposition phenomena in the largescale castellated beryllium structures. For the first time ever a comprehensive analyses have been carried out for materials retrieved from all important locations in the machine. Results on the fuel retention on surfaces in the gaps are summarised by a number of points. Very shallow deuterium deposition is measured in the castellation: 0.5–1.5 mm deep into the groove. No qualitative and quantitative differences are identified between the results from the two JET-ILW campaigns. This occurs despite the fact that approximately 30% more deuterium was measured on the PFS after the first campaign in comparison to ILW-2 [43]. The latter was finished with approximately 300 discharges performed with hydrogen fuel. It is an important indicator that fuel once stored in narrow gaps is difficult to remove by any means. Small quantities of D are found in the castellation both in the erosion and deposition zones. No difference is observed between the poloidal and toroidal gaps. This result could be expected after long operation periods, i.e. full experimental campaigns when different operation scenarios were tested and a large number of disruptions also occurred. No dust accumulation was detected inside the castellation. It Figure 8. Modelling of deposition in the castellation: geometry of the gaps (a) and (b); simulated deposition profiles of deuterium (c) and beryllium (d ) in the gaps of various widths. Nucl. Fusion 57 (2017) 066027 M. Rubel etal 7 was also found that the preparation of castellated tiles before their installation in the JET wall (machining and cleaning of the grooves) was thoroughly performed: only traces of copper residue (below 3 × 1014 cm−2) could be detected. These comprehensive studies of the deposition in the castellated grooves of the beryllium limiters from JET-ILW indicate that the total deuterium content can be estimated in the range from 0.7 × 1022 to 14.2 × 1022 in the castellation and 0.6 × 1022 to 3.0 × 1022 in the gaps between the tiles. The upper value is on the same level as the retention determined by Heinola [38] on the PFS of the limiters: WOPL, IWGL and the upper dump plate. It is also worth mentioning that only small retention was measured also on the surfaces in the gaps between the W lamellae in the divertor. Most important is that the overall retention is significantly lower than that measured after the campaigns in JET-C. All measurements consistently show small retention and steep deposition profiles on the surfaces inside the castellation. These results actually could be expected. The statement is based on earlier data for metallic castellated structures used in the presence of carbon walls: (i) in short-term probes or test limiters exposed in TEXTOR [16, 20], (ii) other machines [22] and, especially, in the beryllium divertor and limiters in JET-C [11, 14]. The following D concentrations were measured with a standard size ion beam (spot around 1 mm in diameter) for the divertor: 6 × 1017 cm−2 at the entrance to the 0.6 mm wide castellated gap, 2 × 1019 cm−2 in 6 mm wide gaps separating the tiles [14]. In the 1 mm wide castellation of the belt limiters the amounts of D at the level of 7 × 1017 cm−2 were determined [11]. The decay length for the castellated grooves was around 1.5 mm in both cases. The tendency indicating the increase in deposition and retention is shown here, but the absolute D content must be taken with care, because measurements were performed a long time after the end of operation: 11 and 15 years for the divertor and limiters, respectively. Short-term experiments (e.g. in TEXTOR) could be model led, but modelling of results after entire experimental campaigns is difficult because of a variety of operation scenarios. However, modelling with the 3D-GAPS code has successfully reproduced steep profiles in narrow gaps (0.5 mm). The calcul ations also show a very significant increase in deposition (and inventory) with the increase in the gap width, e.g. by a factor exceeding 10 when the width of the castellation is increased from 0.5 mm to 2 mm. In conclusion, the experimental and modelled results give a clear indication for ITER regarding the need for a very careful design of the tiles with a particular emphasis on the tile shaping and small width of the castellation grooves. Acknowledgments This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 under grant agreement no. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. This work has been supported by the Swedish Research Council (VR), Grant 2015–04844. References [1] LoarteA. etal 2007 ITER physics basis, chapter 4 Nucl. Fusion 47S203 [2] CoadJ.P. etal 2001 J. Nucl. Mater. 290–3224 [3] CoadJ.P. etal 2003 J. Nucl. Mater. 313–6419 [4] RubelM. etal 2003 J. Nucl. Mater. 313–6321 [5] CoadJ.P. etal 2006 Nucl. Fusion 46350 [6] CounsellG.C. etal 2006 Plasma Phys. Control. Fusion 48B189 [7] WamplerW.R. etal 1988 J. Vac. Sci. Technol. A 62111 [8] HsuW.L. etal 1989 J. Vac. Sci. Technol. A 71065 [9] WalshD.S. etal 1992 J. Vac. Sci. Technol. A 101174 [10] SkinnerC.H. etal 1999 Nucl. Fusion 39271 [11] RubelM., CoadJ.P. and HoleD.E. 2009 J. Nucl. Mater. 386–8729 [12] CoadJ.P., RubelM. and WuC.H. 1997 J. Nucl. Mater. 241–3408 [13] RubelM. etal 2004 Phys. Scr. T 111112 [14] RubelM., CoadJ.P. and PittsR.A. 2007 J. Nucl. Mater. 367–701432 [15] LitnovskyA. etal 2005 J. Nucl. Mater. 337–9917 [16] LitnovskyA. etal 2007 J. Nucl. Mater. 367–701481 [17] KreterA. etal 2013 J. Nucl. Mater. 438S746 [18] KriegerK. etal 2007 J. Nucl. Mater. 363–5870 [19] SergienkoG. etal 2007 Phys. Scr. T 12881 [20] RubelM. etal 2008 Fusion. Eng. Des. 831049 [21] PsodaM. etal 2009 J. Nucl. Mater. 386–8740 [22] LitnovskyA. etal 2011 J. Nucl. Mater. 415S289 [23] MatveevD. etal 2010 Plasma Phys. Control. Fusion 52075007 [24] MatveevD. 2012 Computer simulations of material deposition and fuel retention in remote areas and castellated structures of fusion machines PhD Thesis Ghent University [25] MatveevD. etal 2013 J. Nucl. Mater. 438S775 [26] MatveevD. etal 2014 Phys. Scr. T 159014063 [27] KommM. etal 2011 Plasma Phys. Control. Fusion 53115004 [28] KommM. etal 2013 Plasma Phys. Control. Fusion 55025006 [29] MatthewsG.F. etal 2007 Phys. Scr. T128137 [30] MatthewsG.F. etal 2011 Phys. Scr. T145014001 [31] MaierH. etal 2007 Nucl. Fusion 47222 [32] MatthewsG.F. etal 2013 J. Nucl. Mater. 438S2 [33] BrezinsekS. etal 2015 Nucl. Fusion 55063021 [34] RubelM. etal 2013 J. Nucl. Mater. 438S1204 [35] WiddowsonA. etal 2014 Phys. Scr. T159014010 [36] Baron-WiechecA. etal 2015 J. Nucl. Mater. 463157 [37] PeterssonP. etal 2015 J. Nucl. Mater. 463814 [38] HeinolaK. etal 2016 Phys. Scr. T167014075 [39] MayerM. etal 2016 Phys. Scr. T167014051 [40] LinsmeierC. etal 2007 J. Nucl. Mater. 363–51129 [41] EcksteinW. 2002 Calculated sputtering, reflection and range values Report IPP 9/132 Max-Planck-Institut für Plasmaphysik [42] De TemmermanG. etal 2008 Nucl. Fusion 48075008 [43] WiddowsonA. etal 2017 Nucl. Mater. Energy (https://doi. org/10.1016/j.nme.2016.12.008) Nucl. Fusion 57 (2017) 066027