Plasma-Wall Interactions in ITER

Abstract This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100–150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW / m 2 . Such regimes promote relatively high pressure (> 1 Pa ) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m 3 / s . An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.