DEMO及未来聚变堆设计中多采用低活化钢作为包层结构材料，钨作为面向等离子体第一壁材料，即“钨＋低活化钢”的复合结构。在装置中，来自边界的高密度氘氚（氢同位素）等离子体将通过“等离子体驱动渗透”穿透第一壁进入包层模块，造成冷却剂污染以及氚损失；同时，第一壁长期处于复杂的服役环境中，表面将产生辐照损伤等缺陷，这些物理过程将影响氢同位素渗透行为，然而其机理尚缺乏系统研究。本项目围绕第一壁“钨＋低活化钢”结构，拟采用不同的工艺方法（磁控溅射，化学气相沉积、真空等离子体喷涂等）在低活化钢基体上制备具有特征微观结构的钨涂层并利用实验室直线等离子体装置开展氢同位素渗透实验，研究钨涂层微观结构及表面辐照损伤对氢渗透的影响机理，测定氢同位素在材料中的输运参数并结合计算模拟对未来聚变堆中氚渗透行为进行预测。项目成果可为CFETR和DEMO堆的氚循环计算以及核安全分析提供支持。

In a number of DEMO and future fusion power reactor studies, reduced activation ferritic/martensitic steels (RAFMs) are selected as the structure material and tungsten (W) is chosen as the plasma-facing material for the first wall, i.e., “W＋RAFMs” multi-layer structure. In a fusion reactor, energetic hydrogen isotopes (deuterium and tritium) from the edge plasma will permeate through the first wall into blanket modules by a phenomenon called “plasma-driven permeation (PDP)”, which will lead to the contamination of the coolant and the loss of tritium. Meanwhile, the first wall works under severe service conditions and its surface will be irradiation-damaged inevitably. These processes will influence hydrogen isotopes permeation through the first wall, however the effects have not been systematically investigated and the relevant mechanisms are unclear. In this project, aiming at the multi-layer “W + RAFMs” first wall, several kinds of techniques including magnetron sputtering, chemical vapor deposition and vacuum sputtering deposition etc. will be applied to form W coatings on RAFMs, and then PDP experiments will be carried out using a laboratory-scale linear plasma device. Effects of microstructure and irradiation damage of tungsten coatings on hydrogen isotopes PDP will be studied. Transport parameters of hydrogen isotopes in wall materials will be measured. Then the tritium permeation behavior throughout the first wall of fusion reactors will be predicated by code simulation. This research will provide data support to the calculation of tritium circulation and nuclear safety analysis for CFETR and DEMO.