Plasma–surface interaction in the stellarator W7-X: conclusions drawn from operation with graphite plasma-facing components

Brezinsek, S.; Dhard, C. P.; Jakubowski, M.; König, R.; Masuzaki, S.; Mayer, M.; Naujoks, D.; Romazanov, J.; Schmid, K.; Schmitz, O.; De-you, Zhao; Balden, M.; Brakel, R.; Butterschoen, Birger; Dittmar, T.; Drews, P.; Effenberg, F.; Elgeti, S.; Ford, O.; Fortuna-Zaleśna, E.; Fuchert, G.; Gao, Yu; Goriaev, A.; Hakola, A.; Kremeyer, T.; Krychowiak, M.; Liang, Yun-Feng; Linsmeier, Ch.; Lunsford, R.; Motojima, G.; Neu, R.; Neubauer, O.; Oelmann, J.; Petersson, P.; Rasiński, M.; Rubel, M.; Sereda, S.; Sergienko, G.; Pedersen, T. S.; Vuoriheimo, Tomi; Wang, Erhui; Wauters, T.; Winters, V.; Zhao, Mingzhong; Yi, Rongxing

doi:10.1088/1741-4326/ac3508

Cited by 19 publications

(28 citation statements)

References 66 publications

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“…This decrease of the low-Z impurity concentration significantly extended the operational window of W7-X towards higher plasma densities: the line-integrated electron density increased from 4×10 19 m −2 to more than 1×10 20 m −2 [11]. This decrease of impurities was also associated with a substantial decrease of the observed carbon erosion rates at the TDU strike line position by a factor of more than five [10,13]. Nevertheless, despite the observed decrease of the carbon erosion rate due to decreased oxygen content, the carbon erosion at the TDU was still not small, and during OP 1.2b in total 20±6 g carbon were eroded at the TDU [13].…”

Section: Introductionmentioning

confidence: 82%

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Erosion of tungsten marker layers in W7-X

et al. 2021

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In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s−1 averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 1012 W-atoms/cm2s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 1017–1018 B + C atoms/cm2 (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

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Section: Introductionmentioning

confidence: 82%

“…Erosion of carbon by oxygen subsequently also resulted in high carbon levels in the plasma. The oxygen originated from outgassing of water from carbon components [10]. During the whole operational phase OP 1.2a in total 48±14 g carbon were eroded from the TDU [9].…”

Section: Introductionmentioning

confidence: 99%

Erosion of tungsten marker layers in W7-X

et al. 2021

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“…0.9 g) were injected (via gas injection nozzles built in the TDU) during 30 consecutive plasma discharges in standard magnetic configuration with identical plasma conditions. Simulations have been performed using 3D codes WallDYN-3D [17] and ERO2.0 [23] to model the erosion and deposition patterns showing good agreement with the initial measurements. Detailed measurements on a number of TDU targets removed after OP1.2b, are under progress, therefore in the present C-balance investigations, the 13 C experiments are not included.…”

Section: Carbon Balancementioning

confidence: 90%

Plasma-wall interaction studies in W7-X: main results from the recent divertor operations

Dhard

Brezinsek²,

Mayer

et al. 2021

Phys. Scr.

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Wendelstein 7-X (W7-X) is an optimized stellarator with a 3-dimensional five-fold modular geometry. The plasma-wall-interaction (PWI) investigations in the complex 3D geometry of W7-X were carried out by in situ spectroscopic observations, exhaust gas analysis and post-mortem measurements on a large number of plasma-facing components extracted after campaigns. The investigations showed that the divertor strike line areas on the divertor targets appeared to be the major source of carbon impurities. After multistep erosion and deposition events, carbon was found to be deposited largely at the first wall components, with thick deposits of >1 μm on some baffle tiles, moderate deposits on toroidal closure tiles and thin deposits at the heat shield tiles and the outer wall panels. Some amount of the eroded carbon was pumped out via the vacuum pumps as volatile hydrocarbons and carbon oxides (CO, CO2) formed due to the chemical processes. Boron was introduced by three boronizations and one boron powder injection experiment. Thin boron-dominated layers were found on the inner heat shield and the outer wall panels, some boron was also found at the test divertor unit and in redeposited layers together with carbon. Local erosion/deposition and global migration processes were studied using field-line transport simulations, analytical estimations, 3D-WallDYN and ERO2.0 modeling in standard magnetic field configuration.

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“…PFMC in most tokamaks [8, 14-21, 27, 50-55] and stellarators [56][57][58] because of excellent power handling capabilities.…”

Section: Controlled Fusion and Plasma-wall Interactions: Impact On Ma...mentioning

confidence: 99%

Accelerator techniques and nuclear data needs for ion beam analysis of wall materials in controlled fusion devices

et al. 2023

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A brief overview of ion beam analysis methods and procedures in studies of materials exposed to fusion plasmas in controlled fusion devices with magnetic confinement is presented. The role of accelerator techniques in the examination and testing of materials for fusion applications is emphasised. Quantitative results are based on robust nuclear data sets, i.e. stopping powers and reaction cross-sections. Therefore, the work has three major strands: (i) assessment of fuel inventory and modification of wall materials by erosion and deposition processes; (ii) equipment development to perform cutting-edge research; (iii) determination of nuclear data for selected ion-target combinations. Advantages and limitations of methods are addressed. A note is also given on research facilities with capabilities of handling radioactive and beryllium-contaminated materials.

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Plasma–surface interaction in the stellarator W7-X: conclusions drawn from operation with graphite plasma-facing components

Cited by 19 publications

References 66 publications

Erosion of tungsten marker layers in W7-X

Erosion of tungsten marker layers in W7-X

Plasma-wall interaction studies in W7-X: main results from the recent divertor operations

Accelerator techniques and nuclear data needs for ion beam analysis of wall materials in controlled fusion devices

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