Progress of plasma surface interaction study on low activation materials

Abstract

Ferritic steel, vanadium alloy and SiC/SiC composite are candidate low activation materials for blanket components and first walls in fusion demonstration reactors. Several issues on these materials as the first wall have been investigated so far. Amount of deuterium retained in mechanically polished ferritic steel, F82H, after deuterium ion irradiation, was observed to be several times smaller than that of stainless steel, 316L SS. Physical sputtering yield of the ferritic steel due to deuterium ion was comparable to that of 316L SS. These results suggest that the property of the ferritic steel as the first wall material is superior to that of 316L SS, with respect to fuel hydrogen retention and in-vessel tritium inventory. Since first walls of blanket modules are exposed to both fuel hydrogen and helium, the helium is also trapped in the walls. Helium retention of V–4Cr–4Ti was investigated using helium ion irradiation apparatus. The amount of helium retained was comparable to those of other plasma facing materials. One of the major concerns in use of SiC/SiC composite for blanket is permeation of helium gas coolant into fusion plasma. Helium gas permeability of the SiC/SiC composite after heat cycles was measured using a vacuum device consisting two chambers. The increase in the permeability was not observed when the heating rate was suitably adjusted. Therefore, the blanket module may be made using only SiC/SiC composite if a vacuum pumping for the inside of blanket module is attached.