多场耦合环境下聚变堆第一壁材料W-ZrC服役行为的基础研究

The plasma facing materials (PFMs) of fusion reactor will face multi-fields coupling interaction including high thermal field, stress field and strong irradiation field, their comprehensive servicing performance is closely related to the safe operation and using life of fusion facilities. After 7 years of exploration by focusing on the brittlement, thermal load resistance and irradiation resistance of W based PFM, we successfully fabricated bulk W-0.5%wtZrC alloy with high performance: ductility-brittle transition temperature is lower than 100oC, critical energy density for surface cracking induced by transient thermal load at room temperature is about 4.4MJ/m2; the plasma sputtering corrosion resistance and the hydrogen retention are better than the similar W based materials, which make W-0.5wt%ZrC be a promising PFM candidate. However, whether it can be truly a PFM, it is urgent to evaluate the servicing performance under the approximate working conditions. However, currently, real multi-fields coupling experimental conditions are lacked, therefore this project will rely on the EAST, HL-2A/M and domestic related experiment platform to evaluate the servicing performance and investigate its evolution law of W-0.5wt%ZrC under multi-fields coupling conditions. This programme will be realized by adopting cross experiments with first applying thermal field and then irradiation field as well as irradiation field→high thermal field, combining the multiscale simulation with coexistence of thermal field and irradiation field as well as the assessment experiments in EAST or HL-2A/M. The purpose is to build the servicing database of W-0.5wt%ZrC material under approximately working condition and to reveal the mechanisms of the microstructure damage and performance failure, finally to provide scientific basis for choosing PFMs for China Fusion Engineering Thermal Reactor.

聚变堆第一壁部件材料面临高热场、热应力场及强辐照场等多场耦合服役环境，其综合服役性能优劣关系到聚变装置能否安全运行及其寿命。针对W基第一壁材料脆性、抗热冲击及抗等离子刻蚀等性能问题，我们历经7年探索，成功研制出高性能大块体W-0.5%wtZrC材料：韧脆转变温度≤100度；室温热冲击开裂阈值4.4 MJ/m2；抗等离子刻蚀及氢滞留等性能优于ITER-W，有望用做未来聚变堆第一壁材料。它能否真正被选用，亟需近工况多场服役条件下的考核。为此本项目以W-0.5%wtZrC材料为研究对象，依托EAST和HL-2A/M及国内相关研究平台，通过先热负荷场后辐照场、先辐照场后热负荷场的交叉实验，结合多场同时作用下的多尺度模拟，以及EAST/HL-2A/M装置中的考核实验，开展多场耦合作用下的材料服役性能评价及其演化规律研究，揭示材料组织损伤机理及性能失效机制，为中国聚变工程实验堆第一壁选材提供科学依据。

聚变堆第一壁部件材料面临高热场、热应力场及强辐照场等多场耦合服役环境，其综合服役性能优劣关系到聚变装置能否安全运行及其寿命。本项目以课题组前期开发的高性能大块体W-0.5%wtZrC材料为研究对象，依托EAST和HL-2A/M及国内相关研究平台，通过先热负荷场后辐照场、先辐照场后热负荷场的交叉实验，结合多场同时作用下的多尺度模拟，以及EAST/HL-2A/M装置中的考核实验，开展多场耦合作用下的材料服役性能评价及其演化规律研究，揭示材料组织损伤机理及性能失效机制。.经过五年的努力，我们建立了W-ZrC合金板材常规基础力学性能和微结构基础数据库；搭建了电子束热负荷冲击平台，并初步建立了热负荷（场）作用后W-0.5wt%ZrC力学性能和微结构数据库；建立了离子辐照与热冲击冲击下W-ZrC合金的力学及微结构演化数据库及协同增强效应；借助国际合作初步完成了W-ZrC板材抗中子辐照性能评价；研制了W-0.5wt%ZrC/Cu复合部件；揭示了近工况环境下W-ZrC合金微结构损伤与力学性能退化的之间的定量/定性关系。开展了W-0.5wt%ZrC材料中界面及其氢效应计算模拟研究，获得了点缺陷及其团簇的结构、能量学和动力学性质，以及缺陷与晶界基本作用参数，以及晶界处间隙偏聚诱导的晶界结构相变、空位沿晶运动抑制的空位发射、纳米空洞在晶界处的形态与能量学性质，揭示了不同相界面/晶界与氢、离位缺陷相互作用图像。此外，发展了钨与W-ZrC合金中累积离位损伤与氢滞留模拟方法，建立了钨中纳米孔洞俘获氢的定量预测模型，解决了长期以来无法准确描述和预测氢在孔洞中的结构与能量的基本问题,找到了有效描述孔洞内表面氢吸附位的方法，得到了孔洞芯区氢分子析出的条件，并建立了空洞内氢吸附和氢泡形成的定量预测模型，它可定量描述纳米孔洞中氢的构型、占据能级、成核、起泡过程以及热脱附行为，为评估钨材料在中子辐照下氢滞留量提供了关键的物理过程。开展的工作为多场耦合条件下中国聚变工程实验堆的第一壁选材提供了科学依据。