Plasma-surface interaction issues of an all-metal ITER

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.