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Chemical Behavior of Molten Salt Tritium Breeding Material for Fusion Reactors

聚变反应堆熔盐氚增殖材料的化学行为

基本信息

批准号：
09480107
负责人：
TERAI Takayuki
金额：
$ 7.23万
依托单位：
The University of Tokyo
依托单位国家：
日本
项目类别：
Grant-in-Aid for Scientific Research (B).
财政年份：
1997
资助国家：
日本
起止时间：
1997 至 1999
项目状态：
已结题

来源：
https://kaken.nii.ac.jp/en/grant/KAKENHI-PROJECT-09480107/
关键词：
fusion reactor tritium breeding material molten salt tritium release permeation corrosion structural material blanket 液体ブランケットトリチウム増殖材料溶融塩放出 LiF BeF_2 熱力学

项目摘要

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.

LiF和BeF_2的熔盐混合物（Flibe）被认为是聚变堆包层系统中氚增殖和冷却介质的候选材料。本研究以日本东京大学的快中子源反应堆“YAYOI”为研究对象，在500-700 ℃的中子辐照下，对该材料的氚释放特性、与结构材料的相容性等化学性质进行了研究，氚在熔融Flibe中的扩散系数和氚释放率。在此基础上，建立了Flibe氚释放模型。此外，氚渗透率通过结构壁面对熔融Flibe进行了测量。利用实验获得的数据，提出了一种低氚渗透的氚回收系统，并进行了结构材料与熔融Flibe的相容性试验。将Mo、W、V、Fe及其合金和SiC/SiC_f复合材料试样在550 ℃的熔融Flibe中浸泡1 - 30天，用XRD、RBS、SEM、XPS等方法对试样表面的成分和晶相变化进行了研究。通过热力学计算，对腐蚀速率进行了评价，并对腐蚀机理进行了探讨。