摘要
本文通过对我国自主知识产权三代非能动压水堆(国和一号)与国内成熟运行的核电机组(CPR1000)在无运行人员干预和有运行人员干预情况下的SGTR事故演变过程进行对比,提出了二者在反应堆冷却剂系统(RCS)降温降压手段、蒸汽发生器状态管理、主泵状态、放射性后果、破损蒸汽发生器传热管(SG)降压方式等方面的不同,以及二者放射性释放可能性的差异。这种对比分析便于运行人员在事故中采取更有针对性的干预措施,以使干预效果更加有效。
Through the comparison of the SGTR accident evolution process of China's independent intellectual property third-generation passive pressurized water reactor(Guo He One)with that of a mature domestic nuclear power unit(CPR1000)without or with operator intervention,presents the differences between the two in terms of reactor coolant system(RCS)cooling and pressure reduction means,steam generator state management,main pump state,radioactive consequences,damaged steam generator heat pipe(SG)pressure reduction means,and the differences in the possibility of radioactive release.This comparative analysis is convenient for operators to take more targeted intervention measures to make the intervention effect more effective.
作者
韩凯
Han Kai(State Nuclear Power Demonstration Plant Co.,Ltd.,Rongcheng 264312,China)
出处
《核安全》
2020年第3期19-25,共7页
Nuclear Safety
关键词
SGTR
降温
降压
事故过程
主泵
SGTR
cooling
pressure reduction
accident process
main pump
