Plasma-surface interaction issues of an all-metal ITER

J. N. Brooks, Jean Paul Allain, R. P. Doerner, A. Hassanein, R. Nygren, T. D. Rognlien, D. G. Whyte

Research output: Contribution to journalArticle

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Abstract

We assess key plasma-surface interaction issues of an all-metal plasma facing component (PFC) system for ITER, in particular a tungsten divertor, and a beryllium or tungsten first wall. Such a system eliminates problems with carbon divertor erosion and T/C codeposition, and for an all-tungsten system would better extrapolate to post-ITER devices. The issues studied are sputtering, transport and formation of mixed surface layers, tritium codeposition, plasma contamination, edge-localized mode (ELM) response and He-on-W irradiation effects. Code package OMEGA computes PFC sputtering erosion/redeposition in an ITER full power D-T plasma with convective edge transport. The HEIGHTS package analyses plasma transient response. PISCES and other data are used with code results to assess PFC performance. Predicted outer-wall sputter erosion rates are acceptable for Be (0.3 nm s-1) or bare (stainless steel/Fe) wall (0.05 nm s-1) for the low duty factor ITER, and are very low (0.002 nm s-1) for W. T/Be codeposition in redeposited wall material could be significant (∼2 gT/400 s-ITER pulse). Core plasma contamination from wall sputtering appears acceptable for Be (∼2%) and negligible for W (or Fe). A W divertor has negligible sputter erosion, plasma contamination and T/W codeposition. Be can grow at/near the strike point region of a W divertor, but for the predicted maximum surface temperature of ∼800 °C, deleterious Be/W alloy formation as well as major He/W surface degradation will probably be avoided. ELMs are a serious challenge to the divertor, but this is true for all materials. We identify acceptable ELM parameters for W. We conclude that an all-metal PFC system is likely a much better choice for ITER D-T operation than a system using C. We discuss critical R&D needs, testing requirements, and suggest employing a 350-400 °C baking capability for T/Be reduction and using a deposited tungsten first wall test section.

Original languageEnglish (US)
Article number035007
JournalNuclear Fusion
Volume49
Issue number3
DOIs
StatePublished - Apr 8 2009

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surface reactions
metals
erosion
tungsten
contamination
sputtering
baking
transient response
tritium
beryllium
surface temperature
stainless steels
surface layers
degradation
requirements
irradiation
carbon
pulses

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Condensed Matter Physics

Cite this

Brooks, J. N., Allain, J. P., Doerner, R. P., Hassanein, A., Nygren, R., Rognlien, T. D., & Whyte, D. G. (2009). Plasma-surface interaction issues of an all-metal ITER. Nuclear Fusion, 49(3), [035007]. https://doi.org/10.1088/0029-5515/49/3/035007
Brooks, J. N. ; Allain, Jean Paul ; Doerner, R. P. ; Hassanein, A. ; Nygren, R. ; Rognlien, T. D. ; Whyte, D. G. / Plasma-surface interaction issues of an all-metal ITER. In: Nuclear Fusion. 2009 ; Vol. 49, No. 3.
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Brooks, JN, Allain, JP, Doerner, RP, Hassanein, A, Nygren, R, Rognlien, TD & Whyte, DG 2009, 'Plasma-surface interaction issues of an all-metal ITER', Nuclear Fusion, vol. 49, no. 3, 035007. https://doi.org/10.1088/0029-5515/49/3/035007

Plasma-surface interaction issues of an all-metal ITER. / Brooks, J. N.; Allain, Jean Paul; Doerner, R. P.; Hassanein, A.; Nygren, R.; Rognlien, T. D.; Whyte, D. G.

In: Nuclear Fusion, Vol. 49, No. 3, 035007, 08.04.2009.

Research output: Contribution to journalArticle

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AU - Brooks, J. N.

AU - Allain, Jean Paul

AU - Doerner, R. P.

AU - Hassanein, A.

AU - Nygren, R.

AU - Rognlien, T. D.

AU - Whyte, D. G.

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AB - We assess key plasma-surface interaction issues of an all-metal plasma facing component (PFC) system for ITER, in particular a tungsten divertor, and a beryllium or tungsten first wall. Such a system eliminates problems with carbon divertor erosion and T/C codeposition, and for an all-tungsten system would better extrapolate to post-ITER devices. The issues studied are sputtering, transport and formation of mixed surface layers, tritium codeposition, plasma contamination, edge-localized mode (ELM) response and He-on-W irradiation effects. Code package OMEGA computes PFC sputtering erosion/redeposition in an ITER full power D-T plasma with convective edge transport. The HEIGHTS package analyses plasma transient response. PISCES and other data are used with code results to assess PFC performance. Predicted outer-wall sputter erosion rates are acceptable for Be (0.3 nm s-1) or bare (stainless steel/Fe) wall (0.05 nm s-1) for the low duty factor ITER, and are very low (0.002 nm s-1) for W. T/Be codeposition in redeposited wall material could be significant (∼2 gT/400 s-ITER pulse). Core plasma contamination from wall sputtering appears acceptable for Be (∼2%) and negligible for W (or Fe). A W divertor has negligible sputter erosion, plasma contamination and T/W codeposition. Be can grow at/near the strike point region of a W divertor, but for the predicted maximum surface temperature of ∼800 °C, deleterious Be/W alloy formation as well as major He/W surface degradation will probably be avoided. ELMs are a serious challenge to the divertor, but this is true for all materials. We identify acceptable ELM parameters for W. We conclude that an all-metal PFC system is likely a much better choice for ITER D-T operation than a system using C. We discuss critical R&D needs, testing requirements, and suggest employing a 350-400 °C baking capability for T/Be reduction and using a deposited tungsten first wall test section.

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Brooks JN, Allain JP, Doerner RP, Hassanein A, Nygren R, Rognlien TD et al. Plasma-surface interaction issues of an all-metal ITER. Nuclear Fusion. 2009 Apr 8;49(3). 035007. https://doi.org/10.1088/0029-5515/49/3/035007