PLASMA-WALL INTERACTIONS IN ITER

This paper reviews the status of the design of the divertor and first- wall/shield, the main in-vessel components for ITER. Under nominal ign ited conditions, 300 MW of alpha power will be produced and must be re moved from the divertor and first-wall. Additional power from auxiliar y sources up to the level of 100 MW must also be removed in the case o f driven bums. In the ignited case, about 100 MW will be radiated to t he first wall as bremsstrahlung. Allowing the remaining power to be co nducted to the divertor target plates would result in excessive heat f luxes. 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 geomet ry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densitie s on the target plates can be less than or equal to 5 MW/m(2). Such re gimes promote relatively high pressure(>1 Pa) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping spee d to pump helium is less than or equal to 50 m(3)/s. An important phys ics question is the quality of core confinement in these attractive di vertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performanc e 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 allow ables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties a re major factors determining component lifetime. Considering the poten tial for impurity contamination and minimizing tritium inventory as we ll as thermomechanical performance, the present material selection cal ls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux co uld be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the fin al selection for the first assembly will be made at the end of the EDA .