Chemical Behavior of Molten Salt Tritium Breeding Material for Fusion Reactors

LiF和BEF_2的熔盐混合物(记为Flibe)已被建议作为聚变反应堆包层系统中氚增殖和冷却介质的候选材料。利用东京大学快中子源反应堆Yayoi在500~700℃的中子辐照下进行了原位氚释放实验，测量了释放出的氚化学形态、氚在熔体中的扩散系数和氚释放率。在此基础上，建立了飞碟的氚释放模型。此外，还测量了氚在面向熔融飞板的结构壁上的渗透率。利用实验获得的所有数据，提出了一种有效的低氚渗透率的氚回收系统，并进行了结构材料与熔体的配伍性试验。将钼、钨、钒、铁及其合金和碳化硅/碳化硅复合材料等候选材料的试件在550℃熔体中浸泡1-30天，用X射线衍射仪、X射线衍射仪、扫描电子显微镜、X射线光电子能谱等手段对其表面成分和晶体结构变化进行了分析。结合热力学计算，对腐蚀速率进行了评价，并对腐蚀机理进行了探讨。

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.