Investigation of plasma wall interactions between tungsten plasma facing components and helium plasmas in the WEST tokamak

ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating similar to 2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E (inc) > 20 eV, helium fluence >10(24) He/m(2), surface temperature T (surf) > 700 degrees C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.