反应堆压力容器用Ni-Cr-Mo-V钢焊缝热老化脆化行为研究  

Thermal Ageing Embrittlement Behavior of Ni-Cr-Mo-V Steel Weld Metal for Reactor Pressure Vessel

作　　者：王成龙[1] 佟振峰[1] 张长义[1] 杨兴旺[2] 宁广胜[1] 杨文[1] WANG Chenglong;TONG Zhenfeng;ZHANG Changyi;YANG Xingwang;NING Guangsheng;YANG Wen(China Institute of Atomic Energy,P.O.Box 275 51,Beijing 102413,China;Jiangsu Nuclear Power Corporation,Lianyungang 222042,China)

机构地区：[1]中国原子能科学研究院反应堆工程技术研究部,北京102413 [2]江苏核电有限公司,江苏连云港222042

出　　处：《原子能科学技术》2018年第7期1243-1249,共7页Atomic Energy Science and Technology

基　　金：国家重点研发计划资助项目(2017YFB0702200)

摘　　要：对反应堆压力容器用Ni-Cr-Mo-V钢焊缝温度监督样品的热老化脆化行为进行了研究。焊缝属于压力容器的薄弱环节,服役时间最高达120 430h(服役温度归一化到300℃)。3批次的焊缝监督样品冲击实验表明,焊缝材料在热老化过程中发生了脆化。通过研究发现,金相组织和显微维氏硬度在热老化期间未发生明显的变化,表明在热老化过程中不存在硬化脆化机制。断口分析及扫描俄歇纳米探针研究表明,晶界发生了P的偏析,弱化了晶界结合力,因此,反应堆压力容器用Ni-Cr-Mo-V钢焊缝在热老化过程中发生了由杂质元素P偏析引起的非硬化脆化。The study of thermal ageing embrittlement of temperature sets of reactor pressure vessel surveillance Ni-Cr-Mo-V steel weld metal was conducted.Weld metal is the weakness of the reactor pressure vessel and its service time gets to 120 430 h（the operating temperature is normalized to 300℃）.The impact test of the 3 batches of weld surveillance sample indicates that the weld metal is embrittled during thermal ageing.Through the study,significant change is not found in the microstructure and microVickers hardness during thermal ageing,indicating that there is no hardening embrittlement mechanism in the thermal ageing process.The study of impact fracture and Auger electron spectroscopy indicates that the impurity element P segregates to thegrain boundaries and lowers their cohesion during the long-term thermal ageing.Therefore,the reactor pressure vessel Ni-Cr-Mo-V steel weld metal occurs non-hardening embrittlement which is probably caused by the intergranular segregation of impurity element P during the long-term thermal ageing.

关 键 词：反应堆压力容器 Ni-Cr-Mo-V钢焊缝 杂质元素偏析 非硬化脆化 

分 类 号：TG407[金属学及工艺—焊接]