Chemical Behavior of Tritium Breeding Material for Fusion Reactor

熔融锂铅（17Li-Pb）合金是核聚变堆包层系统中氚增殖最重要的候选材料之一。在本研究中，通过桩内和桩外实验研究了其化学行为，特别是合金中的氚行为。所得结果总结如下：1. 在Ar气氛手套箱中制备了17Li-Pb合金的熔融样品，并通过XRD、SEM、化学分析和DSC等手段证实了样品在晶体结构、热力学相、杂质浓度和熔点等方面具有良好的质量。2. 中子辐照17Li-Pb中生成的氚以HT和T_2等氢的分子形式释放。在高温中子辐照下，首次测量了氚在熔融17Li-Pb中的扩散系数。3. 通过桩内实验测量了氚从17Li-Pb熔液向吹扫气体的传质系数，该传质系数随吹扫气体中氢气分压的增大而增大。结果表明，氚的释放受合金液膜中氚扩散的控制。4. 从氚安全性和氚经济性的角度来看，氚通过面向合金熔液的结构材料的渗透系数是一个非常重要的参数。对铁和304型不锈钢进行了测量，结果表明，结构材料后表面的吹扫气体中氢气的分压对其影响较大。这说明在结构材料表面形成的氧化层可以抑制氚的渗透，而在不锈钢表面形成的Cr_2O_3和FeCr_2O_4等稳定的氧化物层可以作为氚的渗透屏障。

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.