Plasma–surface interaction issues of an all-metal ITER

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.