基于衰变热不确定性的压水堆IVR能力边际研究  

Study on IVR Capability Margin of Pressurized Water Reactor Based on Decay Heat Uncertainty

作　　者：宋建 余红星[1] 邓坚[1] 向清安 朱大欢[1] 许幼幼 罗跃建 Song Jian;Yu Hongxing;Deng Jian;Xiang Qingan;Zhu Dahuan;Xu Youyou;Luo Yuejian(Science and Technology on Reactor System Design Technology Laboratory,Nuclear Power Institute of China,Chengdu,610213,China)

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610213

出　　处：《核动力工程》2021年第6期161-166,共6页Nuclear Power Engineering

基　　金：国家研发计划课题(2019YFB1900703)。

摘　　要：为确定衰变热对高功率压水堆熔融物堆内滞留(IVR)能力边际的影响,采用显著性水平评价与抽样失效率相结合的评价方式,对IVR能力边际进行评价。利用熔融物堆内滞留分析工具CISER开展抽样计算,获得不同核电厂电功率水平、不同衰变热分布参数条件下的下封头壁面热流密度峰值与当地临界热流密度(CHF)的比值,对热流密度比分别开展显著性水平估算与失效率计算,根据小于局部CHF的下封头熔穿准则,判定IVR措施是否有效,以获得IVR能力边际。研究结果表明,若不对下封头内外传热构成进行任何优化措施,电功率超过1400 MW压水堆电厂不推荐单独使用IVR作为严重事故条件缓解措施。In order to determine the influence of decay heat on the IVR(In-Vessel Retention)capability margin of HPWR,an evaluation method combining significance level evaluation and sampling failure rate was used to evaluate IVR capability margin.Using CISER for IVR to carry out the sampling calculation,the ratio of the peak heat flux on the lower head wall to the local critical heat flux(CHF)under different power levels and decay heat distribution parameters of nuclear power plants was obtained to carry out the significance level estimation and failure rate calculation of the heat flux ratio to judge whether the IVR measure is effective based on the penetration criterion of lower head(less than local CHF)to obtain the IVR capability margin.The results show that IVR alone is not recommended as a serious accident mitigation measure for PWR plants with an electrical power of more than 1400 MW without any optimization of the heat transfer composition inside and outside the lower head.

关 键 词：衰变热 临界热流密度(CHF) 熔融物堆内滞留(IVR) 抽样失效率 

分 类 号：TL331[核科学技术—核技术及应用] TL364.4