摘要
总结了核反应堆中应用非常广泛的奥氏体不锈钢的不同辐照损伤行为,包括辐照诱导显微结构变化、辐照诱导偏析、辐照诱导析出、辐照诱导应力腐蚀断裂等;从试验方法和奥氏体不锈钢种类等方面提出了核反应堆用奥氏体不锈钢辐照损伤的研究方向。
Different irradiation damage behavior,including irradiation induced change of microstructure,irradiation induced segregation,irradiation induced precipitation and irradiation induced stress corrosion cracking of austenitic stainless steels commonly used in nuclear reactors is summarized.The research direction of irradiation damage of austenitic stainless steels for nuclear reactors is proposed from the aspects of test methods and category of austenitic stainless steel.
作者
郝予琛
赵美玲
罗来马
HAO Yuchen;ZHAO Meiling;LUO Laima(School of Mechanical Engineering;School of Materials Science and Engineering, Hefei University of Technology, Hefei 230009, China)
出处
《机械工程材料》
CAS
CSCD
北大核心
2018年第7期1-5,共5页
Materials For Mechanical Engineering
基金
国际热核聚变实验堆(ITER)计划专项项目(2014GB121001B)
国家自然科学基金面上项目(51574101)
关键词
奥氏体不锈钢
核反应堆
辐照损伤
austenitic stainless steel
nuclear reactor
irradiation damage
