Chemical Behavior of Tritium Breeding Material for Fusion Reactor

熔融锂铅（17 Li-Pb）合金是聚变堆包层系统中最重要的氚增殖候选材料之一。本文通过堆内和堆外实验研究了其化学行为，特别是氚在合金中的行为。这里得到的结果总结如下。1.在Ar气氛下，用元素试剂在手套箱中制备了17 Li-Pb合金熔体样品，通过XRD、SEM、化学分析和DSC等手段对样品的晶体结构、热力学物相、杂质浓度和熔点进行了表征。2.中子辐照17 Li-Pb产生的氚以HT和T_2等氢分子形式释放出来。在高温中子辐照下，测量了氚在熔融~（17）Li-Pb中的扩散系数，在国际上尚属首次。3.用堆内实验测量了氚从熔融17 Li-Pb到吹扫气体的传质系数，随吹扫气体中氢分压的增大而增大。这一结果表明，氚的释放是由氚在熔融合金液膜中的扩散控制的。4.从氚安全性和氚经济性的角度考虑，氚在面向熔融合金的结构材料中的渗透系数是一个非常重要的参数。它是针对铁和304型不锈钢测量的，并且它受到面向结构材料后表面的吹扫气体中的氢分压的强烈影响。这一结果表明，在结构材料表面形成的氧化物层可以抑制氚的渗透，在不锈钢表面形成的稳定的氧化物层如Cr_2O_3和FeCr_2O_4可以作为氚渗透的阻挡层。

Molten lithium-lead (17Li-Pb) alloy is one of the most important candidate materials for tritium breeding in a fusion reactor blanket system. In this research, its chemical behavior, in particular, tritium behavior in the alloy was investigated by in-pile and out-of pile experiments. The results obtained here are summarized as follows. 1. Molten 17Li-Pb alloy samples were prepared in a glove box of an Ar atmosphere from reagents of the elements, and they were confirmed to have a good quality in crystal structure, thermodynamic phase, impurity concentration and melting point by XRD, SEM, chemical analysis and DSC. 2. The tritium generated in 17Li-Pb by neutron irradiation was released in the molecular forms of hydrogen such as HT and T_2. It is for the first time in the world that the diffusion coefficient of tritium in molten 17Li-Pb was measured under neutron irradiation at elevated temperatures. 3. The mass-transfer coefficient of tritium from molten 17Li-Pb to purge gas was measured by in-pile experiment, and it increased with the partial pressure of hyderogen in the purge gas. This results indicates that tritium release is controlled by the tritium diffusion in the liquid film of the molten alloy. 4. The permeation coefficient of tritium through structural material facing the molten alloy is a quite important parameter from the viewpoints of tritium safety and tritium economy. It was measured for iron and stainless steel type 304, and it was influenced strongly by the partial pressure of hydrogen in the purge gas facing the rear surface of the structural materials. This result sugggests that oxide layer formed on the surface of the structural materials can surpress tritium permeation, and that a layer of stable oxides such as Cr_2O_3 and FeCr_2O_4 formed on the surface of stainless steel can be utilized as a tritium permeation barrier.