超临界水冷堆用候选奥氏体耐热不锈钢热时效组织研究  

Study on the Thermal Aged Microstructure of Candidate Austenitic Heat-resistant Stainless Steel for Supercritical Water-cooled Reactor

作　　者：李俊[1] 李绍宏[1] 熊茹[2] 杨红梅 李萌蘖 Li Jun;Li Shaohong;Xiong Ru;Yang Hongmei;Li Mengnie(School of Materials Science and Engineering,Kunming University of Science and Technology,Kunming,650093,China;Key Laboratory of Nuclear Reactor System Design Technology,Nuclear Power Institute of China,Chengdu,610213,China)

机构地区：[1]昆明理工大学材料科学与工程学院,昆明650093 [2]中国核动力研究设计院反应堆燃料及材料重点实验室,成都610213

出　　处：《核动力工程》2023年第6期148-154,共7页Nuclear Power Engineering

基　　金：国家重点研发计划(2018YFE0116200)。

摘　　要：为了研究热时效过程中超临界水冷堆(SCWR)用候选包壳材料含铝奥氏体耐热钢(AFA)热时效组织和冲击性能的变化,对铝含量为2.5%的AFA钢在650℃进行了500~3000 h热时效处理。利用场发射扫描电镜对析出相及冲击断口进行观察,利用透射电镜对热时效试验钢中析出相的类型和结构进行研究。结果表明:试验钢的冲击韧性随时效时间延长而逐渐降低,试验钢断裂由韧窝断裂逐渐向韧窝断裂和解理断裂的混合断裂方式过渡。热时效过程中Laves相在晶界上析出以及γ'-Ni3Al相大量析出并粗化是AFA钢冲击韧性随时效时间延长而降低的主要原因。In order to study the change of thermal aged microstructure and impact properties of alumina-forming austenitic stainless(AFA)steel,a candidate cladding material for Supercritical water-cooled reactor(SCWR),the AFA steel with 2.5%aluminum content was subjected to thermal aging treatment at 650℃for 500~3000 h.The precipitated phases and the impact fracture were observed by field emission scanning electron microscopy.The types and crystal structures of the precipitated phases were studied by transmission electron microscopy.The results show that the impact toughness of the test steel decreases gradually with the extension of aging time,and the fracture of the test steel gradually transits from dimple fracture to mixed fracture mode of dimple fracture and cleavage fracture.The precipitation of Laves phase at grain boundaries and the precipitation and coarsening ofγ'-Ni3Al phase during thermal aging are the main reasons for the decrease of impact toughness of AFA steel with the extension of aging time.

关 键 词：超临界水冷堆(SCWR) 燃料包壳 AFA钢 冲击性能 

分 类 号：TL334[核科学技术—核技术及应用]