反应堆压力容器承压热冲击分析研究  

Pressurized Thermal Shock Analysis for Reactor Pressure Vessels

作　　者：郑斌[1] 臧峰刚[1] 孙英学[1] 

机构地区：[1]中国核动力研究设计院核反应堆系统设计技术重点实验室,成都610041

出　　处：《核动力工程》2012年第1期1-3,13,共4页Nuclear Power Engineering

摘　　要：依据RCC-M规范和美国NRC 10CFR50.61,对存在假想裂纹的反应堆压力容器堆芯带区进行承压热冲击分析研究。计算核电厂寿期末的基准温度,并采用承压热冲击筛选准则进行评定;计算了承压热冲击瞬态作用下裂纹尖端的应力强度因子,并按RCC-M规范进行评定。According to RCC-M rule and NRC 10CFR50.61,the PTS analysis was performed for the reactor pressure vessel beltline with assumed defect.The end of life reference temperature TR of nuclear power plant was calculated and assessed with PTS screening criteria.The stress intensity factors were calcu-lated under PTS transients and assessed with RCC-M Rule.

关 键 词：反应堆 压力容器 断裂 承压热冲击 应力强度因子 

分 类 号：TL351.6[核科学技术—核技术及应用]