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We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases. 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 and 2019-2020 under Grant Agreement No. 633053. Peer Reviewed Article signat per 497 autors/es: Thomas Sunn Pedersen1,2,∗ , I. Abramovic3, P. Agostinetti4, M. Agredano Torres1, S. Äkäslompolo1, J. Alcuson Belloso1, P. Aleynikov1, K. Aleynikova1, M. Alhashimi1, A. Ali1, N. Allen5, A. Alonso6, G. Anda7, T. Andreeva1, C. Angioni8, A. Arkhipov8, A. Arnold1, W. Asad8, E. Ascasibar6, M.-H. Aumeunier9, K. Avramidis10, E. Aymerich11, S.-G. Baek3, J. Bähner1, A. Baillod12, M. Balden1, M. Balden8, J. Baldzuhn1, S. Ballinger3, M. Banduch1, S. Bannmann1, A. Banon Navarro8, A. Bañon Navarro ´ 1, T. Barbui13, C. Beidler1, C. Belafdil9, A. Bencze7, A. Benndorf1, M. Beurskens1, C. Biedermann1, O. Biletskyi14, B. Blackwell15, M. Blatzheim1, T. Bluhm1, D. Böckenhoff1, G. Bongiovi16, M. Borchardt1, D. Borodin17, J. Boscary8, H. Bosch1,18, T. Bosmann19, B. Böswirth8, L. Böttger1, A. Bottino8, S. Bozhenkov1, R. Brakel1, C. Brandt1, T. Bräuer1, H. Braune1, S. Brezinsek17, K. Brunner1, S. Buller1, R. Burhenn1, R. Bussiahn1, B. Buttenschön1, A. Buzás7, V. Bykov1, I. Calvo6, K. Camacho Mata1, I. Caminal20, B. Cannas11, A. Cappa6, A. Carls1, F. Carovani1, M. Carr21, D. Carralero6, B. Carvalho22, J. Casas20, D. Castano-Bardawil17, F. Castejon6, N. Chaudhary1, I. Chelis23, A. Chomiczewska24, J.W. Coenen13,17, M. Cole1, F. Cordella25, Y. Corre9, K. Crombe26, G. Cseh7, B. Csillag7, H. Damm1, C. Day10, M. de Baar27, E. De la Cal6, S. Degenkolbe1, A. Demby13, S. Denk3, C. Dhard1, A. Di Siena8,28, A. Dinklage12, T. Dittmar17, M. Dreval14, M. Drevlak1, P. Drewelow1, P. Drews17, D. Dunai7, E. Edlund3, F. Effenberg29, G. Ehrke1, M. Endler1, D.A. Ennis5, F.J. Escoto6, T. Estrada6, E. Fable8, N. Fahrenkamp1, A. Fanni11, J. Faustin1, J. Fellinger1, Y. Feng1, W. Figacz4, E. Flom13, O. Ford1, T. Fornal24, H. Frerichs13, S. Freundt1, G. Fuchert1, M. Fukuyama30, F. Füllenbach1, G. Gantenbein10, Y. Gao1, K. Garcia13, J.M. García Regaña6, I. García-Cortés6, J. Gaspar31, D.A. Gates29, J. Geiger1, B. Geiger13, L. Giudicotti32, A. González6, A. Goriaev26,33, D. Gradic1, M. Grahl1, J.P. Graves12, J. Green13, E. Grelier9, H. Greuner8, S. Groß1, H. Grote1, M. Groth34, M. Gruca24, O. Grulke1,35, M. Grün1, J. Guerrero Arnaiz1, S. Günter8, V. Haak1, M. Haas1, P. Hacker1, A. Hakola36, A. Hallenbert1, K. Hammond29, X. Han17,37, S.K. Hansen3, J.H. Harris38, H. Hartfuß1, D. Hartmann1, D. Hathiramani1, R. Hatzky8, J. Hawke39, S. Hegedus7, B. Hein8, B. Heinemann8, P. Helander12, S. Henneberg1, U. Hergenhahn8,40, C. Hidalgo6, F. Hindenlang8, M. Hirsch1, U. Höfel1, K.P. Hollfeld17, A. Holtz1, D. Hopf8, D. Höschen17, M. Houry9, J. Howard19, X. Huang41, M. Hubeny17, S. Hudson29, K. Ida9, Y. Igitkhanov10, V. Igochine8, S. Illy10, C. Ionita-Schrittwieser42, M. Isobe39, M. Jabłczynska ´ 24, S. Jablonski24, B. Jagielski1, M. Jakubowski1, A. Jansen van Vuuren1, J. Jelonnek10, F. Jenko8, F. Jenko8, T. Jensen35, H. Jenzsch1, P. Junghanns8, J. Kaczmarczyk24, J. Kallmeyer1, U. Kamionka1, M. Kandler8, S. Kasilov43, Y. Kazakov26, D. Kennedy1, A. Kharwandikar1, M. Khokhlov1, C. Kiefer8, C. Killer1, A. Kirschner17, R. Kleiber1, T. Klinger12, S. Klose1, J. Knauer1, A. Knieps17, F. Köchl44, G. Kocsis7, Ya.I. Kolesnichenko45, A. Könies1, R. König1, J. Kontula34, P. Kornejew1, J. Koschinsky, M.M. Kozulia14, A. Krämer-Flecken17, R. Krampitz1, M. Krause1, N. Krawczyk24, T. Kremeyer1, L. Krier10, D.M. Kriete5, M. Krychowiak1, I. Ksiazek46, M. Kubkowska24, M. Kuczynski1, G. Kühner1, A. Kumar15, T. Kurki-Suonio34, S. Kwak1, M. Landreman47, P.T. Lang8, A. Langenberg1, H.P. Laqua12, H. Laqua1, R. Laube1, S. Lazerson1, M. Lewerentz1, C. Li17, Y. Liang17, Ch. Linsmeier17, J. Lion1, A. Litnovsky17,48, S. Liu37, J. Lobsien1, J. Loizu12, J. Lore38, A. Lorenz1, U. Losada6, F. Louche26, R. Lunsford29, V. Lutsenko45, M. Machielsen12, F. Mackel8, J. Maisano-Brown3, O. Maj8, D. Makowski49, G. Manduchi50, E. Maragkoudakis6, O. Marchuk17, S. Marsen1, E. Martines4, J. Martinez-Fernandez6, M. Marushchenko1, S. Masuzaki41, D. Maurer5, M. Mayer8, K.J. McCarthy6, O. Mccormack4, P. McNeely1, H. Meister8, B. Mendelevitch8, S. Mendes1, A. Merlo1, A. Messian26, A. Mielczarek49, O. Mishchenko1, B. Missal1, R. Mitteau9, V.E. Moiseenko14, A. Mollen1, V. Moncada9, T. Mönnich1, T. Morisaki41, D. Moseev1, G. Motojima41, S. Mulas6, M. Mulsow1, M. Nagel1, D. Naujoks1, V. Naulin35, T. Neelis19, H. Neilson29, R. Neu8, O. Neubauer17, U. Neuner1, D. Nicolai17, S.K. Nielsen35, H. Niemann1, T. Nishiza1, T. Nishizawa1, T. Nishizawa8, C. Nührenberg1, R. Ochoukov8, J. Oelmann17, G. Offermanns17 K. Ogawa41, S. Okamura41, J. Ölmanns17, J. Ongena26, J. Oosterbeek1, M. Otte1, N. Pablant29, N. Panadero Alvarez6, N. Panadero Alvarez6, A. Pandey1, E. Pasch1, R. Pavlichenko14, A. Pavone1, E. Pawelec46, G. Pechstein1, G. Pelka24, V. Perseo1, B. Peterson41, D. Pilopp1, S. Pingel1, F. Pisano11, B. Plöckl8, G. Plunk1, P. Pölöskei1, B. Pompe2, A. Popov51, M. Porkolab3, J. Proll19, M.J. Pueschel19,27, M.-E. Puiatti52, A. Puig Sitjes1, F. Purps1, K. Rahbarnia1, M. Rasinski ´ 17, J. Rasmussen35, A. Reiman29, F. Reimold1, M. Reisner8, D. Reiter17, M. Richou9, R. Riedl8, J. Riemann1, K. Riße1, G. Roberg-Clark1, V. Rohde8, J. Romazanov17, D. Rondeshagen1, P. Rong1, L. Rudischhauser1, T. Rummel1, K. Rummel1, A. Runov1, N. Rust1, L. Ryc24, P. Salembier20, M. Salewski35, E. Sanchez6, S. Satake41, G. Satheeswaran17, J. Schacht1, E. Scharff1, F. Schauer8, J. Schilling1, G. Schlisio1, K. Schmid8, J. Schmitt5, O. Schmitz13, W. Schneider1, M. Schneider1, P. Schneider8, R. Schrittwieser42, T. Schröder1, M. Schröder1, R. Schroeder1, B. Schweer26, D. Schwörer1, E. Scott1, E. Scott8, B. Shanahan1, G. Sias11, P. Sichta29, M. Singer1, P. Sinha29, S. Sipliä34, C. Slaby1, M. Sleczka53, H. Smith1, J. Smoniewski54, E. Sonnendrücker8, M. Spolaore4, A. Spring1, R. Stadler8, D. Stanczak24, T. Stange1, I. Stepanov26, L. Stephey13, J. Stober8, U. Stroth8,55, E. Strumberger8, C. Suzuki41, Y. Suzuki41, J. Svensson1, T. Szabolics7, T. Szepesi7, M. Szücs7, F.L. Tabares6, N. Tamura41, A. Tancetti35, C. Tantos10, J. Terry3, H. Thienpondt6, H. Thomsen1, M. Thumm10, J.M. Travere9, P. Traverso5, J. Tretter8, E. Trier8, H. Trimino Mora1, T. Tsujimura41, Y. Turkin1, A. Tykhyi45, B. Unterberg17, P. van Eeten1, B.Ph. van Milligen6, M. van Schoor26, L. Vano1, S. Varoutis10, M. Vecsei7, L. Vela56, J.L. Velasco6, M. Vervier17, N. Vianello50, H. Viebke1, R. Vilbrandt1, G. Vogel8, N. Vogt1, C. Volkhausen1, A. von Stechow1, F. Wagner1, E. Wang17, H. Wang57, F. Warmer1, T. Wauters26, L. Wegener1, T. Wegner1, G. Weir1, U. Wenzel1, A. White3, F. Wilde1, F. Wilms1, T. Windisch1, M. Winkler1, A. Winter1, V. Winters1, R. Wolf118, A.M. Wright29, G.A. Wurden39, P. Xanthopoulos1, S. Xu17, H. Yamada58, H. Yamaguchi41, M. Yokoyama41, M. Yoshinuma41, Q. Yu8, M. Zamanov14, M. Zanini1, M. Zarnstorff29, D. Zhang1, S. Zhou17, J. Zhu1, C. Zhu29, M. Zilker8, A. Zocco1, H. Zohm8, S. Zoletnik7 and L. Zsuga7 // 1 Max Planck Institute for Plasma Physics, Garching and Greifswald, Germany: 2 University of Greifswald, Greifswald, Germany; 3 Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139, United States of America; 4 Consorzio RFX, Corso Stati Uniti, 4-35127 Padova, Italy; 5 Auburn University, Auburn, AL 36849, United States of America; 6 CIEMAT, Avenida Complutense, 40, 28040 Madrid, Spain; 7 Center for Energy Research, Konkoly-Thegeut 29-33, 1121 Budapest, Hungary; 8 Max-Planck-Institute for Plasma Physics, Boltzmannstraße 2, 85748 Garching bei München, Germany; 9 CEA Cadarache, 13115 Saint-Paul-lez-Durance, France; 10 Karlsruhe Institute of Technology, Kaiserstr. 12, 76131 Karlsruhe, Germany; 11 University of Cagliari, Via Universita, 40, 09124 Cagliari, Italy; 12 École Polytechnique Fédérale de Lausanne, Swiss Plasma Center, CH-1015 Lausanne, Switzerland; 13 University of Wisconsin–Madison, Engineering Drive, Madison, WI 53706, United States of America; 14 Institute of Plasma Physics, National Science Center ‘Kharkiv Institute of Physics and Technology’, Kharkiv, Ukraine; 15 The Australian National University, Acton ACT 2601, Canberra, Australia; 16 Department of Engineering, University of Palermo, Viale delle Scienze, Edificio 6, Palermo, 90128, Italy; 17 Forschungszentrum Jülich GmbH, Institut für Energie-und Klimaforschung—Plasmaphysik, 52425 Jülich, Germany; 18 Technical University of Berlin, Strasse des 17. Juni 135, 10623 Berlin, Germany; 19 Eindhoven University of Technology, 5600 MB Eindhoven, Netherlands; 20 Universitat Politècnica de Catalunya. BarcelonaTech, C. Jordi Girona, 31, 08034 Barcelona, Spain; 21 Culham Center for Fusion Energy, Abingdon OX14 3EB, United Kingdom; 22 Instituto de Plasmas e Fusao Nuclear, Av. Rovisco Pais, 1049-001 Lisboa, Portugal; 23 Department of Physics, National and Kapodistrian University of Athens, 15784 Athens, Greece; 24 Institute of Plasma Physics and Laser Microfusion, 23 Hery Str., 01-497 Warsaw, Poland; 25 ENEA—Centro Ricerche Frascati, Via Enrico Fermi, 45, 00044 Frascati RM, Italy; 26 Laboratory for Plasma Physics, LPP-ERM/KMS, TEC Partner, B-1000 Brussels, Belgium; 27 Dutch Institute for Fundamental Energy Research, PO Box 6336, 5600 HH Eindhoven, Netherlands; 28 University of Texas, Austin, TX, United States of America; 29 Princeton Plasma Physics Laboratory, Princeton, NJ 08543, United States of America; 30 Kyushu University, 744 Motooka Nishi-ku, Fukuoka 819-0395, Japan; 31 Aix-Marseille University, Jardin du Pharo, 58 Boulevard Charles Livon, 13007, Marseille, France; 32 Department of Physics and Astronomy, Padova University, Via Marzolo 8, 35131 Padova, Italy; 33 Department of Applied Physics, Ghent University, Sint-Pietersnieuwstraat 41 B4, 9000 Ghent, Belgium; 34 Aalto University, 02150 Espoo, Finland; 35 Department of Physics, Technical University of Denmark, Anker Engelunds Vej, 2800 Kgs Lyngby, Denmark; 36 VTT Technical Research Center of Finland Ltd., PO Box 1000, FI-02044 VTT, Finland; 37 Institute of Plasma Physics, Chinese Academy of Sciences, 230031 Hefei, Anhui, China; 38 Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37830, United States of America; 39 Los Alamos National Laboratory, NM 87545, United States of America; 40 Fritz-Haber-Institut der Max-Planck-Gesellschaft, 14195 Berlin, Germany; 41 National Institute for Fusion Science, National Institutes of Natural Sciences, 322-6 Oroshi-cho, Toki, Gifu Prefecture 509-5292, Japan; 42 Institute for Ion Physics and Applied Physics, University of Innsbruck, Innsbruck, Austria; 43 Graz University of Technology, Rechbauerstraße 12, 8010 GRAZ, Austria; 44 Austrian Academy of Science, Doktor-Ignaz-Seipel-Platz 2, 1010 Wien, Austria; 45 Institute for Nuclear Research, prospekt Nauky 47, Kyiv 03028, Ukraine; 46 University of Opole, plac Kopernika 11a, 45-001 Opole, Poland; 47 University of Maryland, Paint Branch Drive, College Park, MA 20742, United States of America; 48 National Research Nuclear University MEPhI, 115409 Moscow, Russian Federation; 49 Department of Microelectronics and Computer Science, Lodz University of Technology, Wolczanska 221/223, 90-924 Lodz, Poland; 50 Consiglio Nazionale delle Ricerche, Piazzale Aldo Moro, 7, 00185 Roma, Italy; 51 Ioffe Physical-Technical Institute of the Russian Academy of Sciences, 26 Politekhnicheskaya, St Petersburg 194021, Russian Federation; 52 Istituto di Fisica del Plasma Piero Caldirola, Via Roberto Cozzi, 53, 20125 Milano, Italy; 53 University of Szczecin, 70-453, aleja Papieza Jana Pawła II 22A, Szczecin, Poland; 54 Lawrence University, 711 E Boldt Way, Appleton, WI 54911, United States of America; 55 Physik-Department E28, Technische Universität München, 85747 Garching, Germany; 56 Universidad Carlos III de Madrid, Av. de la Universidad, 30 Madrid, Spain; 57 Yale University, New Haven, CT 06520, United States of America; 58 University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chhiab 277-0882, Japan Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant::7.a - Per a 2030, augmentar la cooperació internacional per tal de facilitar l’accés a la investigació i a les tecnolo­gies energètiques no contaminants, incloses les fonts d’energia renovables, l’eficiència energètica i les tecnologies de combustibles fòssils avançades i menys contaminants, i promoure la inversió en infraestructures energètiques i tecnologies d’energia no contaminant

A micro-structuring of the tungsten plasma-facing surface can strongly reduce near surface thermal stresses induced by ELM heat fluxes. This approach has been confirmed by numerical simulations with the help of ANSYS software. For experimental tests, two 10 × 10 mm2 samples of micro-structured tungsten were manufactured. These consisted of 2000 and 5000 vertically packed tungsten fibres with dimensions of Ø240 µm × 2.4 mm and Ø150 µm × 2.4 mm, respectively. The 1.2 mm bottom parts of the fibres are embedded in a copper matrix. The top parts of the fibres have gaps about of 10 µm so they are not touching each others. The top of all tungsten fibres was electro-polished. A Nd:YAG laser with a pulse duration 1 ms and a pulse repetition frequency of 25 Hz was used to simulate up to 105 ELM-like heat pulses. No damage on either of the micro-structured tungsten samples were observed. Neon plasma erosion rate and fuel retention of the micro-structured tungsten samples were almost identical to bulk tungsten samples. Keywords: Micro-structured tungsten, PFC, PFM, High heat load, Thermal cycling, Retention, Erosion, Emissivity

Surface morphology development and sputtering behavior of Cr, as a test material, have been explored under He plasma exposure at a low incident ion energy of ∼80 eV in multiple linear plasma devices: PISCES-A, PSI-2, and NAGDIS-II. From comparison of the experiments in these devices, deposition of a small amount of heavier impurities (Mo in NAGDIS-II and Ta in PISCES-A) onto Cr is found to result in the formation of cone structures on the Cr surface due to preferential sputtering, resulting in a significant reduction (up to ∼10 times) in the sputtering yield of Cr due to line-of-sight redeposition onto the neighboring cones. The heavier impurities are thought to originate from a sample holding cap/cover, which can be sputtered by a trace amount of intrinsic impurities (C, O, etc) as well as by Cr ionized in the plasma. It can be concluded from the Cr experiments, as well as additional Be data collected in PISCES-B, that heavier impurity deposition plays a major role in the cone structure formation, while other mechanisms (e.g. surface irregularity and oxide) also exist. Keywords: Impurity deposition, Sputtering, Nanostructure

The roughness of metallic surfaces has a vital impact on the erosion of plasma-facing materials. Roughness determines the effective sputtering yield Yeff of the facing material. The angular/energy distribution of sputtered particles, and the spatial erosion and deposition distribution. The model for simulation the effect of the surface roughness was earlier implemented into the 3D Monte-Carlo code ERO2.0 and validated using results of ion beam experiments and experiments in the linear plasma device PSI-2. In the present study the developed ERO2.0 surface morphology model was applied to the JET-ILW tungsten (W) divertor consisting of smooth bulk W and W-coated CFC components. Influence of the surface roughness on the W erosion as well as on the transport of sputtered material in conditions of inclined magnetic field was investigated. Simulation results are in a good agreement with existing experimental findings.

Long pulse discharges in the Large Helical Device have often been interrupted by large amounts of dust particle emission from the divertor region caused by the exfoliation of carbon-rich mixed material deposition layers. The plasma wall interaction code ERO2.0 has provided the simulation results of the three-dimensional distribution of the carbon flux density in the divertor region which is quite reasonable with the observed distribution of the carbon-rich deposition layers. The code has also succeeded in reproducing the reduction of the carbon deposition layers on dome plates by changing the target plate configuration in the divertor region. The ERO2.0 simulations have also successfully explained dust particle emission from the inboard side near the equatorial plane for the new target plate configuration at the termination of a long pulse discharge. These simulation results prove that the ERO2.0 code is applicable to predicting the possible position from where the dust particles are released, and to designing an optimized divertor configuration for performing stable long pulse discharges with controlled dust particle emission.

The surface morphology of plasma-facing components (PFCs) and its evolution during plasma irradiation has been shown to have a significant effect on the erosion and subsequent transport of sputtered particles in plasma. This in turn can influence the resulting lifetime of PFCs. A model for treatment of the effect of surface roughness on the erosion of PFCs has recently been incorporated into the three-dimensional Monte Carlo code ERO2.0. First simulations have confirmed a significant influence of the assumed surface roughness (for both regular and stochastic numerically constructed samples) on both the effective sputtering yields Y-eff and the effective angular distributions of sputtered particles. In this study, a series of experiments at the linear plasma device PSI-2 are conducted to test the effect of surface roughness on the sputtering parameters. Graphite samples prepared with a 100 nm molybdenum layer with various surface roughness characteristic sizes (R-a = 110 nm, 280 nm, 600 nm and R-a < 20 nm) were exposed to a helium plasma in the PSI-2 linear plasma device at a magnetic field B.=.0.1 T. These PSI-2 experiments were simulated using ERO2.0 with a surface morphology model. Simulations are able to reproduce the experimentally observed significant suppression of erosion for higher R-a values.

Sputtering by ions with low near-threshold energies is investigated. Experiments and simulations are conducted for tungsten sputtering by low-energy, 85-200 eV Ar atoms. The angular distributions of sputtered particles are measured. A new method for molecular dynamics simulation of sputtering taking into account random crystallographic surface orientation is developed, and applied for the case under consideration. The simulations approximate experimental results well. At low energies the distributions acquire "butterfly-like" shape with lower sputtering yields for close to normal angles comparing to the cosine distribution. The energy distributions of sputtered particles were simulated. The Thompson distribution remains valid down to near-threshold 85 eV case. (C) 2017 Elsevier B. V. All rights reserved.

Surface morphology and its evolution during the plasma irradiation is known to have a large influence on the erosion and resulting lifetime of plasma-facing components as well as tritium retention. For instance, surface roughness can affect physical sputtering, re-deposition, as well as angular distributions of the sputtered species. In this study the effect of surface roughness is implemented into the 3D Monte-Carlo code ERO2.0. First modelling results for molybdenum (Mo) irradiated with deuterium (D) in the conditions foreseen for the planned experiments at the linear plasma device PSI-2 are presented. Using the constructed examples of surfaces with various (regular and fractal) roughness types it is shown that the effective sputtering yield decreases for rough surfaces in comparison to smooth ones. The angular distribution of particles escaping from the rough surface collimates with the increase of the surface structure's aspect ratio. Moreover, the modelling predicts flattening of the surface during the plasma irradiation due to the preferable re-deposition in the “valleys” and sputtering of the peak tops.

The present study addresses the uncertainties that affect the recently performed predictions of beryllium (Be) erosion and migration in ITER using the Monte-Carlo code ERO2.0. The focus of the study is a D–T baseline discharge with fusion power gain Q = 10 , scrape-off layer (SOL) input power P SOL = 100 MW , toroidal plasma current I p = 15 MA , and central toroidal field B t = 5.3 T . The parameter studies used to investigate uncertainties include variations of the radial extrapolation of plasma parameters in the far-SOL (scan A), the assumptions on impact angle distributions (scan B) and the anomalous transport of eroded Be (scan C). Variations by factors ∼ 3 , ∼ 18 and ∼ 2 for scans A, B and C, respectively, are found.

The three-dimensional Monte-Carlo impurity transport and plasma surface interaction code ERO2.0 is applied to a full-torus model for the Large Helical Device (LHD). In order to find an optimum experimental condition for effective real-time wall conditioning (boronization) using an Impurity Powder Dropper (IPD), the toroidal and poloidal distribution of the boron flux density on the divertor components and the vacuum vessel are surveyed in various experimental conditions. The source profile of the neutral boron atoms originated from boron powders supplied from the IPD is calculated using the DUSTT code in background plasmas provided by the EMC3-EIRENE code. The simulations using ERO2.0 predict that higher plasma density operation is inappropriate for the effective wall conditioning because of the toroidally localized boron flux density in a closed helical divertor region. The ERO2.0 simulations have successfully revealed an optimum experimental condition for the wall conditioning with the toroidally uniform boron flux density in the closed helical divertor region.

After vertical displacement events (VDEs) in the JET-ILW tokamak, debris is produced in liquid form and ejected from the vessels beryllium upper dump plates (UDPs). Most of the droplets splash near the UDPs outboard tile 8 as documented by an in-vessel photographic survey. Moreover, the systematic collection of 'dust' from the main chamber and the divertor region provides evidence of small solid spherical Be particles and flat splats (with fairly circular, 'pancake-like', or elongated shapes) related to melting events (with a diameter of few mu m prior the deposition). The migration of metallic Be is studied with the numerical code DUSTTRACK, capable of describing the full droplet dynamics in a non-stationary tokamak plasma configuration. The description of the JET plasma during the VDE proposed here allows for a sufficiently consistent reconstruction of the transient conditions of the background plasma, relying on actual physical measurements. Due to the lack of experimental data related to the initial properties of the ejecta, the statistically relevant sample of 4x10(5) particles was considered in the DUSTTRACK simulations. A reasonable fit of the initial parameters for the ejected droplets, as well as the time-dependent plasma parameters for the VDE resulted in good qualitative agreement with the post-mortem experimental findings from dust collectors and with the Be deposition pattern near UDPs on the low-field side.

We present experimental data on the rate coefficient for electron impact ionisation of W atoms at electron temperatures between 5 and 10 eV. Our results undercut different theoretical data sets based on the Born approximation and the distorted wave approximation, respectively, by more than a factor of 3. Better agreement in absolute values is found with a more recent work by Goswami et al. The present data match well with earlier measurements by Steinbrink obtained with a different method.

Results of a new dedicated experiment addressing the problem of impurity deposition at the bottom in gaps are presented along with modelling. A test limiter with an isolated gap was exposed to the scrape-off layer plasma in TEXTOR. The exposure was accompanied by injection of 13 C-marked methane in the vicinity of the gap. Deposition at the bottom of the gap was monitored in situ with Quartz-Microbalance diagnostics. The 13 C deposition efficiency of about 2.6 × 10 −5 was measured. Post mortem analysis of resulting deposited layers performed with SIMS and EPMA techniques yields about a factor 2 smaller value corresponding to approximately 10% contribution of the gap bottom to the total 13 C deposition in the gap. This measured contribution is effectively much smaller than observed earlier in TEXTOR, taking the difference in geometry into account, and is in reasonable agreement with modelling performed with ERO and 3D-GAPS codes.

To quantify tungsten (W) sputtering measurements by spectroscopy, two experiments have been carried out in the tokamak TEXTOR. The erosion of a W limiter in the plasma edge was studied spectrocopically under different plasma conditions. Conversion of the photon fluxes of neutral W lines was performed [1] with the aid of inverse photon efficiencies, measured in situ in a second experiment with WF6 injection to realise a controllable W source. Penetration depths, particle velocities and line ratios of W I were determined and compared between the two sources of W. The velocities of injected and sputtered W differ by a factor of 3 as the analysis of the penetration depth shows, while the line ratios for different W I lines are comparable. A comparison of the e-folding length with the model code GKU was performed which reveals a deviation of at least a factor of 6. Inverse photon efficiencies at Te = 40 eV were determined to be about 44 for W I (400.88 nm) and about 63 for W I (429.46 nm).

On the TEXTOR tokamak various experiments aimed at investigation of tungsten erosion and transport are performed. In one experiment a spherical W/C twin limiter positioned close to the last-closed flux surface in the near scrape-off layer was exposed to a number of comparable plasma discharges with stepwise variations of edge plasma parameters. Spatial distribution of tungsten and carbon light emission was recorded with two dimensional CCD cameras and spectrometer systems with high spectral and spatial resolution. Penetration depths, tungsten sputtering fluxes and erosion yields were measured. Comparison between experimental data and the results of modelling with the 3D Monte-Carlo code ERO is performed. The main objective of this study was to test the adequacy of the existing atomic data for neutral tungsten. The modelled penetration depths of the light emission of tungsten are a factor of 2–3 smaller than in experiment, which may indicate the overestimation of ionization rates.

Use of new 3-strap ICRF antennas with all-tungsten (W) limiters in ASDEX Upgrade results in a reduction of the W sources at the antenna limiters and of the W content in the confined plasma by at least a factor of 2 compared to the W-limiter 2-strap antennas used in the past. The reduction is observed with a broad range of plasma shapes. In multiple locations of antenna frame, the limiter W source has a minimum when RF image currents are decreased by cancellation of the RF current contributions of the central and the outer straps. In JET with ITER-like wall, ITER-like antenna produces about 20% less of main chamber radiation and of W content compared to the old A2 antennas. However the effect of the A2 antennas on W content is scattered depending on which antennas are powered. Experiments in JET with trace nitrogen (N-2) injection show that a presence of active ICRF antenna close to the midplane injection valve has little effect on the core N content, both in dipole and in -90 degrees phasing. This indicates that the effect of ICRF on impurity transport across the scape-off-layer is small in JET compared to the dominant effect on impurity sources leading to increased impurity levels during ICRF operation. (C) 2016 The Authors. Published by Elsevier Ltd. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2016.10.026

At FOM Rijnhuizen, linear plasma generators are used to investigate plasma-material interactions under high-density (≤1021 m -3), low-temperature (≤5 eV) plasma bombardment. Research into carbon-based materials has been focused on chemical erosion by hydrogen plasmas. Results from plasma exposure to high-flux (>1023 H +/m2 s) and low-temperature hydrogen plasma indicate silicon carbide has a lower relative rate of gross erosion than other carbon-based materials (e.g. graphite, diamond, carbon-fiber composites) by a factor of 7-10. Hydrogenic retention is the focus of research on tungsten and molybdenum. For target temperatures of 700-1600 K, the temperature dependence of hydrogenic retention is the dominant factor. Damage to the surface by heavy ion irradiation has shown to enhance retention by a factor of 2.5-4.1. Thermal stressing of W via. e-beam thermal cycling also enhances hydrogenic retention by a factor of 2.1 ± 0.2, likely due to the introduction of thermal defects, which act as trapping sites for implanted hydrogenic isotopes.

The mitigation of carbon chemical erosion observed in PISCES-B due to the presence of a minor beryllium concentration in plasma can be determined to a large extent by formation of carbide (Be2C) on the plasma-facing surface. To study this, dedicated beryllium seeding experiments in the linear plasma simulator PISCES-B have been simulated using the Monte-Carlo code ERO. The spectroscopy is used to benchmark the simulations against observations. A temporal dependent evolution of the beryllium carbide formation on the carbon surface of the target is introduced on the basis of characteristic time scales. The influence of carbide formation on the effective mitigation time for chemical erosion is investigated.

A molecular dynamics technique was used to study the sputtering characteristics of co-deposited carbon (C) and to investigate subsequent effects of introducing beryllium (Be) impurities. Amorphous C layers containing hydrogen (H) with atomic ratios of H/(C+H)

Castellation will be used in divertor and first wall components to provide thermo-mechanical stability of ITER. Radioactive fuel may be stored in the gaps of castellated structures representing a safety issue for ITER. Tungsten castellated structures with different shapes were exposed in TEXTOR to investigate the impact of cell shaping on impurity transport and fuel deposition in the gaps. After exposure a significant intermixing of tungsten was detected in carbon deposits in the gaps reaching 70 at.% of W in the deposited layer. This will provide difficulties in cleaning the gaps in ITER. Poloidal gaps of shaped cells contained a factor or 3 less deuterium than those of rectangular cells, the carbon deposition exhibited only marginal advantages of a new geometry. Poloidal and toroidal gaps contained comparable amount of C and D. Significant deposition at the bottom of gaps was measured which could only partly be reproduced by modeling.

13 CH 4 was injected through an aperture in a roof-like test limiter covered by a graphite, tungsten or molybdenum plate during Ohmic discharges. The evolution of the deposition pattern was studied by in situ reflectometry during each plasma pulse. Amounts of locally deposited 13 C and co-deposited background 12 C in each shot were determined post-mortem . The 13 C deposition efficiency is in a range of 0.05–0.2% and grows with the increasing amount of 12 C. For the same 12 C content the efficiency is of 0.11% for tungsten, 0.14% for molybdenum and 0.17% for graphite.

21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI), Natl Inst Fus Sci, Kanazawa, JAPAN, MAY 26-30, 2014; International audience; This paper aims at understanding the main processes responsible for the erosion/deposition pattern observed on the surface of the Toroidal Pumped Limiter of Tore Supra, using the 3D local impurity transport code ERO. The influence of the plasma impurity content, CX-flux and surface temperature on the global carbon balance and erosion/deposition pattern is discussed. Main results are (1) that considering medium-range transport of C ions is mandatory for reproducing the main characteristics of the global C balance and erosion/deposition pattern, (2) that impurities and CX-atoms increase the erosion by a factor 42 (without changing the net/gross erosion ratio), and (3) that chemical erosion is governed by the re-erosion of deposits, which depends strongly on the surface temperature. (C) 2014 Elsevier B.V. All rights reserved.

Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented.

The Monte Carlo code ERO simulates the three-dimensional transport of impurity particles in a background plasma using the test particle approximation. It treats the interaction of the plasma ions, atoms and of the impurities with the plasma facing components (PFC). ERO modelling for ITER estimates the lifetime of the inner target (erosion of 0.5 cm) to about 7000 shots with a shot length of 400s. The maximum allowed tritium retention due to eroded carbon not redeposited locally on the target plates would be already reached after 380-440 shots. Strike point sweeping might increase the target life time by a factor of about 4 while tritium retention seems not to be influenced. Apart from transient events especially material mixing effects of the elements foreseen for ITER as PFC (carbon, beryllium and tungsten) introduce uncertainties in the prediction. The injection of hydrocarbon ( 13 CH 4 ) through spherical shaped limiters (one made of tungsten, one of graphite) in TEXTOR reveal that carbon has a reduced deposition efficiency on tungsten compared to graphite. This effect has not been taken into account until now in the modelling by the ERO code. The surface model of the code has therefore been substituted by a customized version of the Monte Carlo Code SDTrimSP which will help to clarify the effects of material mixing. A first successful test simulation shows that the depth distributed concentration of PFC influences sputtering yields.

(CH4)-C-13 was injected through graphite and tungsten spherical limiters in reproducible TEXTOR discharges. These materials were chosen, as they represent the actual compromise for the plasma facing components in the ITER divertor. C-13 was used in order to distinguish injected and intrinsic carbon in the layer deposited on the limiter surface. Shot-by-shot video recordings show a continuous growth of the deposit near the injection hole. A pronounced difference in the C-13 deposition pattern on the graphite and tungsten limiters was observed. Post-mortem surface analysis showed that the ratios of the locally deposited to the injected amount of carbon are 4% for graphite and 0.3% for tungsten. The margins of the carbon layer deposited on tungsten are significantly steeper in comparison with the graphite limiter case. The large difference in the C-13 deposition efficiency can be explained by direct reflection of carbon from tungsten and the enhanced sputtering of carbon on the tungsten substrate. Nucleation is suggested to be an important mechanism for carbon deposition on tungsten. Monte Carlo impurity transport calculations by the ERO code reproduce reasonably the experimental results for the graphite limiter.

The prediction of the first wall deterioration and possible plasma contamination by impurities is a high priority task for ITER. 3D Monte-Carlo code ERO is a tool for modeling of eroded impurity transport and spectroscopy in plasma devices useful for experiment interpretation. Chromium (Cr) is a fusion-relevant reactor wall element (e.g. component of RAFM steels expected for use in DEMO). Linear plasma devices including PSI-2 are effective tools for investigations of plasma-surface interaction effects, allowing continuous plasma operation and good control over irradiation parameters. Experiments on Cr sputtering were conducted at PSI-2. In these experiments the Cr erosion was measured by three techniques: mass loss of the sample, quartz micro-balance of deposited impurities at a distance from it and optical emission spectroscopy. Experiments were modeled with the 3D Monte-Carlo code ERO, previously validated by application to similar experiments with tungsten (W). The simulations are demonstrated to reproduce the main experimental outcomes proving the quality of the sputtering data used. A significant focuses of the paper is the usage and validation of atomic data (resent metastable-resolved dataset from ADAS) for interpretation of Cr spectroscopy. Initial population of quasi-metastable state was fitted by matching the modeling with the experimental line intensity profiles.

13th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems, Peking Univ, Fus Simulat Ctr, Beijing, PEOPLES R CHINA, SEP 17-20, 2013; International audience; The effective sputtering yield of Be (Y-Be(tot)) was determined in situ by emission spectroscopy of low ionizing Be as function of the deuteron impact energy (E-in = 25-175 eV) and Be surface temperature (T-surf = 200 degrees C-520 degrees C) in limiter discharges carried out in the JET tokamak. Be self sputtering dominates the erosion at high impact energies (E-in > 150 eV) and causes Y-Be(tot) far beyond 1. Y-Be(tot) drops to low values, below 4.5%, at the accessible lowest impact energy (E-in similar or equal to 25 eV) achievable in limiter configuration. At medium impact energies, E-in = 75 eV, two contributors to the measured Y-Be(tot) of 9% were identified:two third of the eroded Be originates from bare physical sputtering (Y-Be(phys)) and one third from chemical assisted physical sputtering (Y-Be(chem)). The later mechanism has been clearly identified by the appearance of BeD A-X emission and quantified in cause of a temperature dependence at which the BeD practically vanishes at highest observed Be limiter temperatures. The recorded T-surf dependence, obtained in a series of 34 identical discharges with ratch-up of T-surf by plasma impact and inertial cooling after the discharge, revealed that the reduction of BeD is correlated with an increase of D-2 emission. The release mechanism of deuterium in the Be interaction layer is exchanged under otherwise constant recycling flux conditions at the limiter. The reduction of Y-Be(chem) with T-surf is also correlated to the reduction of the Be content in the core plasma providing information on the total source strength and Be screening. The chemical assisted physical sputtering, always present at the nominal limiter pre-heating temperature of T-surf = 200 degrees C, is associated with an additional sputtering channel with respect to ordinary physical sputtering which is surface temperature independent. These JET experiments in limiter configuration are used to benchmark the ERO code and verify ITER first wall erosion prediction. The ERO code overestimates the observed Be sputtering in JET by a factor of about 2.5 which can be transferred to ITER predictions and prolong the expected lifetime of first wall elements.

An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s(-1) to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas.

A new simplified analytical expression for the electromagnetic field in the Debye sheath in the presence of an oblique magnetic field including surface biasing effect is suggested. It is in good agreement with the numerical solution of the integral equation for the potential distribution in the Debye sheath. The energy and angular impact distributions and corresponding surface sputtering yields were analyzed in the presence of an oblique magnetic field and surface biasing. The analytical expression was used to estimate a) the effective sputtering yield of the W target with a varying negative voltage against plasma in PSI-2 linear device and b) erosion of the JET outer wall Be limiter near the ICRH antenna enhanced during RF emission.

Pioneering experiments to study local erosion and deposition processes have been carried out in TEXTOR by injecting 13C marked hydrocarbons (CH4 and C2H4) as well as silane (SiD4) and tungsten-hexafluoride (WF6) through test limiters exposed to the edge plasma. The influence of various limiter materials (C, W, Mo) and surface roughness, different geometries (spherical or roof-like) and local plasma parameters has been studied. Depending on these conditions the local deposition efficiency of injected species varies between 0.1% and 9% – the largest deposition has been found for 13CH4 injection through unpolished, spherical C test limiter and ohmic plasma conditions. The most striking result is that ERO modelling cannot reproduce these low deposition efficiencies using the common assumptions on sticking probabilities and physical and chemical re-erosion yields. As an explanation large re-erosion due to background plasma and possibly low “effective sticking” of returning species is applied. This has been interpreted as enhanced re-erosion of re-deposits under simultaneous impact of high ion fluxes from plasma background.

Previous studies (Carpentier et al 2011 J. Nucl. Mater. 415 S165–S169) carried out with the LIM code of the ITER first wall (FW) on beryllium (Be) erosion, re-deposition and tritium retention by co-deposition under steady-state burning plasma conditions have shown that, depending on input plasma parameter assumptions and sputtering yields, the erosion lifetime and fuel retention on some parts of the FW can be a serious concern. The importance of the issue is such that a benchmark of this previous work is sought and has been provided by the ERO code (Pitts et al 2011 J. Nucl. Mater. 415 S957–S964) simulations described in this paper. Provided that inputs to the codes are carefully matched, excellent agreement is found between the erosion/deposition profiles from both codes for a given ITER-shaped FW panel. Issues regarding the difficult problem of the correct treatment of Be sputtering are discussed in relation to the simulations. The possible influence of intrinsic Be impurity is investigated.

Methane (CH4) was injected into the high density (n(e) similar to 10(20) m(-3)), low temperature (T-e similar to 1 eV) hydrogen plasma in Pilot-PSI to determine the CH A-X photon efficiency in this unexplored plasma regime. The effects of particle transport and particle reflection on the emission of directly excited CH under these plasma conditions were assessed with the 3D Monte Carlo code ERO. The simulations of the inverse photon efficiency showed a difference of similar to 20% between full hydrocarbon sticking or no sticking (reflection). In addition it predicts that particle transport may lead to more than a factor of 10 increase. The measured inverse photon efficiency is however constant at 100 +/- 30 for 0.1 < T-e < 1.0 eV. The constancy is consistent with dissociative recombination of CH4+, CH3+ and CH2+ to produce excited CH instead of direct excitation. These results form a framework for in situ carbon erosion measurements in future fusion reactors such as ITER.

The simulations for linear devices using the 3D Monte-Carlo code ERO have a number of particularities of both physical and technical nature. Many physical effects introduced and tested for these simulations (elastic collisions, population of metastable states, etc.) can be relevant for tokamaks as well. The current status of these activities is summarized. As an example some simulations reproducing the experimental observations at PISCES-B and Pilot-PSI are presented. Main deviations and ideas for their origins are discussed. (c) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

We have studied the chemical erosion of different carbon composites in Pilot-PSI at ITER-relevant hydrogen plasma fluxes (∼1024 m -2 s-1) and low electron temperatures (Te∼1 eV). Optical emission spectroscopy on the CH A-X band was used to characterize the chemical sputtering. Fine grain graphite (R 6650, SGL Carbon Group), ITER-reference carbon fiber composite material (SNECMA NB31 and NB41; Dunlop 3D), nano- and micro-crystalline diamond coatings on molybdenum and SiC (Silit® SKD Reaction-Bonded, Saint-Gobain Ceramics) were compared. The chemical sputtering was similar for the different composites under comparable plasma conditions, except for SiC, which produced a ten times lower rate. The CH emission was constant at electron temperatures Te>1 eV and ion fluxes ranging between 1023 and 1024 m- 2 s-1, but decreased at lower temperatures. This decrease is possibly due to changes in the excitation of CH and not due to a change in the chemical erosion rate.

The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8??m is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

Measurements of Si I- and Si II-emission lines were performed in the front of a test limiter under two different conditions: (i) silicon is sputtered from a silicon layer deposited on a graphite test limiter, (ii) silane is puffed into the plasma through a gas inlet in the limiter head, setting Si-atoms free from the dissociation of SiD4 molecules. It was shown that the relative intensities of the emission lines within the multiplets are in good agreement with theoretical data calculated in the SL-coupling scheme. The values for ‘ionization per photon’ were calculated for the electron temperature range below 100 eV using the GKU kinetic code and compared with measured ones. The measured and theoretical values for the Si II-emission lines are in reasonable agreement within a deviation of 30%. It was found that experimentally measured values for Si I lines match the calculated ones only when a LTE-population of 3p 2 SL terms ( 3 P, 1 D and 1 S) is assumed.

Estimations of the ITER first wall (FW) lifetime, previously made using the three-dimensional Monte-Carlo ERO code (Borodin et al 2011 Phys. Scr. T145 014008), depend strongly on the assumptions of the physical sputtering yield for beryllium (Be). It is of importance to validate the respective model and data at existing devices including the JET ITER-like wall (ILW) as most ITER-relevant experiments. Applying the same sputtering input data in ERO as those used before in the ITER-predictions, the ERO simulations for the Be light intensity (using up to date atomic data from ADAS and measured plasma conditions) reveal a factor of 2 overestimation in the assumed yield even if the low estimate assuming 50% D surface content is used. This result indicates the preference of this assumption for plasma-wetted areas. It points to a possible necessity to correct (reduce) the respective estimates for the Be sputtering yield and, accordingly, re-visit the ITER FW lifetime predictions.

Transport of hydrocarbon impurities in a high-density (>10 20 m−3), low-temperature (

A Monte-Carlo code called SURO was developed to study the influence of surface roughness on the impurity deposition characteristics in fusion experiments. SURO uses the test particle approach to describe the impact of background plasma and the deposition of impurity particles on a sinusoidal surface. The local impact angle and dynamic change of surface roughness as well as surface concentrations of different species due to erosion and deposition are taken into account. Coupled with the three-dimensional Monte-Carlo code ERO, SURO was used to study the impact of surface roughness on 13C deposition in 13CH4 injection experiments in TEXTOR. The simulations showed that the amount of net deposited 13C species increased with surface roughness. Parameter studies with varying 12C and 13C fluxes were performed to gain insights into impurity deposition characteristics on the rough surface. Calculations of the exposure time needed for surface smoothing for TEXTOR and ITER were also carried out for different scenarios.

To analyse the impurity transport in plasma-shadowed, remote areas, methane CD4 has been injected into the far scrape-off layer of TEXTOR through a cylinder equipped with a quartz micro balance (QMB). CD4 transport including break-up and resulting deposition on the QMB (shot-resolved) and on the cylinder top surface (shot-integrated) has been modelled with the codes ERO and 3D-GAPS. The modelling shows good agreement with the observations if reflection coefficients based on molecular dynamics simulations are used. In contrast to plasma-wetted areas, no enhanced erosion has to be applied.

Tungsten is foreseen as the plasma-facing component material for baffles, the dome and strike-point area in the ITER divertor. Quantification of the W source, which is connected with the components lifetime and W plasma concentration, is one of the outstanding issues in the qualification process. A dedicated experiment in TEXTOR with the exposure of a W/C twin limiter to the near scrape-off layer plasma has been carried out in order to address the W sputtering and local material mixing in the electron temperature range between Te=30 and 85?eV, achieved with deuterium fueling in four steps. The Te range is comparable to the baffle region and the strike-point area during non-detached transient phases of the ITER divertor plasma. Quantification of the W sputtering yield and the impinging impurity fluxes was performed with the aid of optical spectroscopy, in particular by observation of WI and WII lines. As no inverse photon efficiencies in the plasma parameter range of the twin limiter experiment exist, we performed in a second experiment for the first time a calibration of WI and WII photon efficiencies with local injection of WF6 through a gas inlet into the TEXTOR edge plasma. The in situ determined effective inverse photon efficiency of about 85 for the most prominent WI line at 400.9?nm, which is in good agreement with GKU modelling for the covered Te range and 650 for the WII line at 434.8?nm, has been applied to the corresponding photon fluxes in the twin limiter experiment. The W sputtering yield decreases from 5.2 to 0.5%, thus by about one order of magnitude, with a reduction of Te from 85?eV down to 30?eV and a simultaneous increase of the impinging deuterium ion flux by 50% occurs. A lower limit for the prompt redeposition has been estimated at 50% by analyzing the WI to WII flux ratio. Local measurement of OII (441.6?nm) and CII (426.7?nm) provided impurity flux ratios of 0.6% for O and 5.2% for C related to the deuterium recycling, respectively ion flux. Both flux ratios remain constant for all phases of the discharge with plasma edge cooling. W erosion is predominantly caused by sputtering of higher ionization stages of O and C impinging on the W limiter half and not by the fuel species itself. Plasma cooling below the physical sputtering threshold could not be achieved without impurity seeding.

The Monte-Carlo neutral transport code 3D-GAPS is described. The code models impurity transport and deposition in remote areas, such as gaps between cells of castellated plasma-facing surfaces. A step-by-step investigation of the interplay of different processes that may influence the deposition inside gaps, namely particle reflection, elastic neutral collisions, different particle sources, chemical erosion and plasma penetration into gaps, is presented. Examples of modelling results in application to the TEXTOR experiment with a castellated test limiter are provided. It is shown that only with the assumption of the presence of species with different reflection probabilities, do simulated carbon deposition profiles agree with experimental observations for side surfaces of the gaps. These species can be attributed to different particle sources, e. g. carbon atoms and hydrocarbon radicals. Background carbon ions and atoms have low and moderate values of the reflection coefficient (R = 0.9). Deposition at the bottom of the gaps cannot be adequately reproduced unless extreme assumptions on particle sources and reflection properties are imposed. Elastic neutral collisions and ionization of neutrals escaping the gaps have no significant influence on the results. Nevertheless, particle-in-cell simulations of plasma penetration into gaps are essential for estimating the incoming ion flux and leading to a better quantitative agreement with experimental observations.

The 3D Monte Carlo code ERO has been used to simulate dedicated TEXTOR experiments in which methane 13CH4 and ethene 13C2H4 were injected into the plasma through graphite and tungsten spherical limiters to investigate local carbon transport and deposition. Spectroscopy was used to follow the observable hydrocarbon break-up products, and the local 13C deposition efficiency was measured by post-mortem surface analysis. The experimental observations showing the dependence of the 13C deposition on substrate and gas have been modelled. The main uncertain input parameters for the modelling, as 12C concentration in the background plasma, effective sticking coefficient for hydrocarbons and enhanced erosion of redeposited carbon, have been defined by benchmarking between modelling and experiment. With the same set of parameters, the modelled 13C deposition efficiencies are in good agreement with the experimental ones (0.8–2.1%). Also, the observed distribution of 13C deposition on the limiter surfaces is well reproduced by modelling. The higher amount of 13C deposition on graphite compared with tungsten limiter is mainly due to the higher reflection of carbon on tungsten and the higher physical sputtering of carbon film on top of the tungsten compared with graphite. The larger amount of 13C deposition for 13C2H4 instead of 13CH4 injection can be explained by more 13C returning to the limiter surface from 13C2H4 injection than 13CH4 injection, which is mainly determined by the different masses and reaction rate coefficients.

ERO modelling of long-term tritium (T) retention has been done for the divertor of ITER with graphite target plates assuming a certain beryllium influx into the divertor, eroded from the main chamber. The divertor beryllium (Be) influx relative to the deuterium ion flux has been fixed at 0.1% for the outer divertor and 1.0% for the inner divertor. In addition to the original B2-Eirene plasma background, the influence of variations of temperature and density in the divertor has been studied. Moreover, assumptions for enhanced erosion of redeposited carbon and effective sticking for hydrocarbons have been analysed. With graphite target plates, long-term tritium retention is dominated by T co-deposition in deposits. Within the studied parameter range, the modelling yields 200?500 possible ITER discharges without cleaning before reaching the safety limit of 700?g of in-vessel retained tritium. Surface temperature-dependent tritium amounts in carbon and beryllium deposits have been taken into account.

Reaction rate coefficients for higher hydrocarbon molecules (ethane family) have been implemented into the 3D Monte Carlo code ERO, which is used for modelling of hydrocarbon transport and light emission of spectroscopically observable radicals in the edge plasma of fusion experiments. For benchmarking of the ERO modelling, dedicated experiments were carried out at TEXTOR in which different hydrocarbon molecules (CD4, C2D4 and C2D6) were injected into the plasma near the last closed flux surface. The light emission of the hydrocarbon break-up products CD, C2 and C+ was observed. The modelled penetration depths of the light emission from the CD Ger?-band, C2 Swan band and C?II are in good agreement with the experiments. The effective inverse photon efficiencies (D/XB values) for CD and C2, which include the dissociation chains of the initial molecules, are more dependent on local electron temperature than on electron density. The calculated D/XB values are in fair agreement with the experimental ones.

The 3D Monte-Carlo code ERO, which calculates erosion processes, impurity transport and deposition, has been coupled to the Monte-Carlo code SDTrimSP to simulate material mixing processes in wall components more precisely. SDTrimSP calculates the transport of ions in solids by means of the binary collision approximation. It keeps track of the depth dependent material concentration caused by implantation of projectiles in the solid. Modelling with the coupled code ERO-SDTrimSP is compared with dedicated TEXTOR experiments, in which the formation of mixed surface layers has been studied. In these experiments, methane 13CH4 was injected through graphite and tungsten spherical limiters during plasma exposure and the local redeposition probability was measured post mortem by surface analysis. A significant difference in the carbon 13C deposition efficiency, i.e. the ratio of the locally deposited to the injected amount of 13C, between graphite and tungsten was found, 4% for graphite and 0.3% for tungsten. Modelling of these experiments with ERO-SDTrimSP reproduces the clear substrate dependence with about 2% deposition efficiency on graphite and less than 0.5% on tungsten in good agreement with the experiment. The reason for the substrate dependence is partly explained by the higher physical sputtering yield of a thin carbon film on top of a tungsten substrate compared with a graphite substrate. Surface roughness of the materials has been identified to be another important parameter for the interpretation of the results.

Wall conditioning is essential in tokamak and stellarator research to achieve plasma performance and reproducibility. This paper presents an overview of recent conditioning results, both from experiments in present devices and modelling, in view of devices with superconducting coils, with focus on W7-X, JT-60SA and ITER. In these devices, the coils stay energised throughout an experimental day or week which demands for new conditioning techniques that work in presence of the nominal field, in addition to the proven conditioning methods such as baking, glow discharge conditioning (GDC) and low-Z wall coating through GDC-plasma, which do not work under such condition. The discussed techniques are RF conditioning without plasma current, both in the ion cyclotron and electron cyclotron range of frequencies, and diverted conditioning plasmas with nested magnetic flux surfaces. Similarities and differences between tokamaks and stellarators are highlighted. Finally a conditional tritium recovery strategy for ITER is proposed based on Ion Cyclotron Wall Conditioning and L-mode plasma results from JET, equipped with an ITER-like wall (beryllium main chamber wall and tungsten divertor).