超临界水冷堆燃料包壳和堆内构件候选材料的辐照损伤研究

项目名称： 超临界水冷堆燃料包壳和堆内构件候选材料的辐照损伤研究

项目编号： No.U1532134

项目类型： 联合基金项目

立项/批准年度： 2016

项目学科： 数理科学和化学

项目作者： 郭立平

作者单位： 武汉大学

项目金额： 66万元

中文摘要： 超临界水冷堆是第四代核能系统国际论坛选定研发的六种堆型之一。目前包壳材料和堆内构件的选材研究是超临界水冷堆的关键问题之一。中国也研发了自己的百万千瓦级超临界水冷堆的设计方案CSR1000，HR3C不锈钢已经被选定为CSR1000燃料包壳和堆内构件的主要候选材料。然而，关于HR3C不锈钢耐辐照性能的研究很少。目前急需堆外辐照数据，为下一阶段的堆内中子辐照方案提供参考。基于这一需求，本项目计划系统研究重离子辐照对HR3C不锈钢微观结构的影响，包括位错环产生、辐照肿胀、辐照偏析等，揭示其物理机制。并采用纳米压痕测试和微观结构分析相结合的方法研究辐照的硬化规律。这些研究将会为我国超临界水冷堆包壳材料和堆内构件的选材研究提供参考数据和技术支持。

中文关键词： 超临界水冷堆；燃料包壳材料；堆内构件；辐照损伤；重离子辐照

英文摘要： The Supercritical Water-cooled Reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). The material selection for fuel cladding and reactor internals still remains one of the key issues of SCWRs. HR3C stainless steels have been selected as the main candidate materials for fuel claddings and reactor internals of CSR1000, which is the Chinese concept design of a 1000MW supercritical water-cooled reactor. However, few studies have been carried out on the irradiation resistance of HR3C stainless steels. At present, outside-reactor irradiation data are in urgent need, which would provide basic references for the design of in-pile neutron irradiations which will be performed at next development stage of CSR1000. In this project, the effect of heavy-ion irradiation on the microstructure of HR3C stainless steels will be investigated, including the produce of dislocation loops, void swelling, radiation-induced segregation (RIS). Irradiation hardening will also be investigated with nano-indentation and microstructure analysis. The project will provide experimental data and technical support for the material selection for fuel cladding and reactor internals of SCWR.

英文关键词： Supercritical Water-cooled Reactor;Fuel cladding materials;Reactor internals;Irradiation damage;Heavy-ion irradiation