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Abstract:Plasma-wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strikepoint tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor / Be first wall and all-W or all-C.One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q=10 ITER discharge [ 1 ] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4±3 higher, this margin has been adopted as uncertainty of the scaling.With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated:• It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ.• Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated.• For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms.• For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account.Finally the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.

As part of the ITER Design Review, the physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

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The work of the ITPA SOL/divertor group is reviewed and implications for ITER discussed. Studies of near SOL gradients have revealed a connection to underlying turbulence models. Analysis of a multi-machine database shows that parallel conduction gradients near the separatrix scale as major radius. New SOL measurements have implicated low-field side transport as driving parallel flows to the inboard side. The high-n nature of ELMs has been elucidated and new measurements have determined that they carry ~10-20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. Analysis of ELM measurements imply that the ELM continuously loses energy as it travels across the SOL-larger gaps should reduce surface loads. The predicted divertor power loads for ITER disruptions has been reduced as a result of finding that the divertor footprint broadens during the thermal quench and that the plasma can lose up to 80% of its thermal energy before the thermal quench (not true for VDEs or ITBs). On the other hand predictions of power loading to surfaces outside the divertor have increased. Disruption mitigation through massive gas puffing has been successful at reducing divertor heat loads but estimates of the effect on the main chamber walls indicate 10s of kG of Be could be melted/mitigation. Estimates of ITER tritium retention have reduced the amount retained/discharge although the uncertainties are large and tritium cleanup may be necessary every few days to weeks. Long-pulse studies have shown that the fraction of injected gas that can be recovered after a discharge decreases with discharge length. The retention rate on the sides of tiles appears to ~ 1-3% of the ion flux to the front surface for C tiles and ~100x less for Mo tiles. T removal techniques are being developed based on surface heating and surface ablation although ITER mixed materials will make T removal more difficult. The use of mixed materials gives rise to a number of potential processes-e.g. reduction of surface melting temperatures (formation of alloys) and reduction of chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.