摘要
Abstract
Sodium-cooled fast reactors represent a critical technological pathway for closed nuclear fuel cycles and the sustainable nuclear energy development.Owing to their well-balanced comprehensive performance,austenitic stainless steels have been adopted as key structural materials for fuel cladding in fast reactors.However,during service,they are subjected to extreme environments,such as high temperatures and intense neutron irradiation,which lead to irradiation-induced hardening,swelling,and embrittlement.Among these,irradiation swelling is recognized as the most critical issue limiting cladding lifetime.The degradation of macroscopic properties originates from the evolution of irradiation-induced microstructures.This article systematically reviews research progress on the microstructural changes in austenitic stainless steel claddings for fast reactors following neutron irradiation.It focused on elucidating the evolution mechanisms of dislocation loops and dislocation networks,the nucleation and growth of cavities(voids and bubbles)and the resulting irradiation swelling behavior;as well as the formation conditions,compositional characteristics,and evolution of thermal-aged precipitates and irradiation-induced precipitates(such as the γ′ phase,G-phase,and phosphides).The role of radiation-induced segregation(RIS)as a central driver of microchemical redistribution was analyzed in depth.The influence of the aforementioned microstructural evolution on the macroscopic properties of the materials was also discussed.Finally,the review summarizes existing research achievements,points out key challenges in current studies,such as the precise characterization of vacancy-type dislocation loops,formation mechanisms of γ′ and G-phases,and controlling mechanisms of stable phosphides—and suggested future research directions.These include employing advanced characterization techniques,such as in-situ transmission electron microscopy and atom probe tomography,to reveal underlying micro-mechanisms.This review provides an important reference for a deeper understanding of irradiation damage mechanisms in fast reactor cladding materials and for the design of a new generation of radiation-resistant materials.关键词
钠冷快堆/奥氏体不锈钢/包壳材料/辐照损伤/微观结构演化/析出相/辐照肿胀/辐照诱导偏析Key words
Sodium-cooled fast reactor/Austenitic stainless steel/Cladding material/Irradiation damage/Microstructure evolution/Precipitates/Irradiation swelling/Irradiation-induced segregation分类
能源科技
