Chemical Behavior of Molten Salt Tritium Breeding Material for Fusion Reactors

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A molten salt mixture of LiF and BeF_2 (denoted as Flibe) has been proposed to be a candidate material for tritium breeding and cooling medium in fusion reactor blanket systems. In this study, its chemical properties such as tritium release property and compatibility with structural materials under blanket-simulated conditions were investigated.In-situ tritium release experiments were carried out under neutron irradiation at the temperatures of 500-700 C using the fast neutron source reactor "YAYOI", University of Tokyo in order to measure the released chemical species of tritium, diffusion coefficient of tritium in molten Flibe and tritium release rate. Based on these parameters, a model on tritium release from Flibe was constructed. In addition, the tritium permeation rate through structural wall facing molten Flibe was measured. By using, all the date obtained in the experiments, an effective tritium recovery system with low tritium permeation was proposed.A compatibility test of structural materials with molten Flibe was also carried out. Specimens of some candidate materials including Mo, W, V, Fe and their alloys as well as SiC/SiC_f composite were immersed in molten Flibe at 550 C for 1 - 30 days, and they were examined by XRD, RBS, SEM, XPS, etc. on the composition and the crystallographic change of the surface. The corrosion rate was evaluated and the mechanism was discussed in comparison with thermodynamic calculation.