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，钨作为第一壁面向等离子体材料。在热壁运行模式下，边界等离子体中的氢同位素（氘、氚）将通过“等离子体驱动渗透”穿透第一壁进入包层模块中，造成冷却剂污染、氚损失等一系列问题。第一壁长期处于聚变堆复杂工况下，表面难免受到损伤，同时，等离子体中的杂质可能在壁表面形成再沉积层，这些物理过程对氢同位素在第一壁中的渗透行为起关键作用，但其机理尚缺乏充分研究。本项目针对钨材料以及“钨+低活化钢”结构，利用实验室直线等离子体装置以及EAST托卡马克材料实验平台，研究表面条件对氢同位素在第一壁中的渗透行为的影响机理，测定氢同位素在壁材料中的渗透参数；结合计算模拟，进而实现对聚变堆第一壁中氚渗透量的评估。项目成果将为CFETR和DEMO的内部部件设计以及核安全分析提供支持。

In the design of DEMO and future fusion power reactors, reduced activation steels are usually selected as the structure material for the blanket and tungsten is chosen as the plasma facing material for the first wall. The first wall will be operated at elevated temperatures, and as a result, hydrogen isotopes (deuterium and tritium) from edge plasma will penetrate through the first wall and into the blanket modules by a phenomenon call “plasma-driven permeation, PDP”. New issues such as coolant contamination and tritium loss will be raised. The first wall will work under extremely harsh conditions in the reactor and its surface will be damaged inevitably. Meanwhile, impurities from plasma may be re-deposited on the wall surface. Although it has been known that these processes play a key role in hydrogen isotopes permeation through the first wall, its effects have not yet been sufficiently investigated. Using a laboratory linear plasma facility and a material evaluation system on the EAST tokamak, the surface condition effects on hydrogen isotopes permeation through tungsten and “tungsten + reduced activation steels” structures will be studied. The permeation parameters for hydrogen isotopes in wall materials will be experimentally measured as well. Then the tritium permeation flux through the first wall of fusion reactors can be evaluated by code simulation. This research will provide data support to the in-vessel components design and nuclear safety analysis for CFETR and DEMO.