压力容器用奥氏体不锈钢316H的高温辐照行为研究OA

To address the lack of high-temperature irradiation performance data for the austenitic stainless steel 316H,the material used for the reactor pressure vessel of a fourth-generation high-temperature gas-cooled microreactor,at the design operating temperature of 550℃,a research plan involving irradiation damage tests and aging tests was proposed.The SRIM software was used to simulate the trajectories and irradiation damage defects of energetic ions(He2+)incident on the material.Combined with high-temperature irradiation tests under different cumulative irradiation doses(0.005,0.05,0.5 dpa)and an irradiation aging test at 650℃for 100 h,the evolution laws of the material's high-temperature damage behavior and aging behavior with microstructure were established.The results show that at the design temperature of 550℃,316H exhibits good irradiation resistance.The irradiation damage layer is approximately several micrometers thick,and the damage mainly manifests as microvoids and dislocations.After aging at 650℃for 100 h,carbides precipitate along grain boundaries,with C and Mo elements enriching at the grain boundaries and Cr element depleting near the grain boundaries,presenting a certain risk of failure.This study provides a reference for subsequent engineering implementation of in-reactor irradiation tests.