Abstract
Due to a number of technological properties and, first of all, to a low atomic number, beryllium will be used as plasma facing material in ITER. Tritium control, including both the permeation through and inventory in the beryllium, is of great importance for the safety of the device. Experimental data have shown that, under ITER-like plasma conditions, the plasma facing surfaces of the beryllium develop high porosity (bubbles) and become saturated with bubbles, leading to a strong uptake of tritium and deuterium ions almost independent of the incident flux. At fluxes typical of ITER, surface erosion of beryllium should be also taken into account. A computational model has been used with the computer code TMAP4 to reproduce the available experimental data concerning hydrogen ion implantation in beryllium. The results described in this paper refer to the first wall of the European Helium Cooled Pebble Bed Blanket (HCPB) Test Blanket Module (TBM-I).
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