压水堆核电厂稳压器波动管热分层分析关键技术探讨  

Discussion on the key technology for thermal stratification analysis of pressurizer surge tube in pressurized water reactor plant

作　　者：陈明亚[1,2] 孙兴悦 刘晗 余伟炜 史芳杰[1,2] 彭群家 赵万祥[1,2] CHEN Mingya;SUN Xingyue;LIU Han;YU Weiwei;SHI Fangjie;PENG Qunjia;ZHAO Wanxiang(Suzhou Nuclear Power Research Institute,Suzhou 215004,China;National Nuclear Power Plant Safety and Reliability Engineering Research Center,Suzhou 215004,China;Tianjin University,Tianjin 300072,China;EDF R&D China Center,Beijing 100005,China)

机构地区：[1]苏州热工研究院有限公司,江苏苏州215004 [2]国家核电厂安全及可靠性工程技术研究中心,江苏苏州215004 [3]天津大学,天津300072 [4]法国电力公司中国研发中心,北京100005

出　　处：《压力容器》2023年第9期55-61,共7页Pressure Vessel Technology

基　　金：国家重点研发计划项目(2020YFB1901500,2021YFB3702602)。

摘　　要：压水堆核电厂稳压器波动管(以下简称“波动管”)存在冷热流体分层的现象,影响核电厂的安全运行。针对波动管热分层运行工况存在不确定性的问题,鲜有基于核电厂真实监测数据的分析研究;对于存在热分层的实际运行瞬态,尚缺乏有效的基于设计瞬态参数的包络方法;同时,对于疲劳损伤较为显著的情况,当前基于疲劳裂纹萌生准则的评定方法存在难以满足长寿期安全运行需求的问题。针对上述技术现状,通过调研国内外学者在波动管热分层研究方面的工作,对有限元数值仿真中的网格划分、材料性能设定、边界条件选择、热分层流动仿真和结构应力响应分析技术等内容进行了探讨。同时,对国内某大型压水堆核电厂真实的运行监测数据进行了分析,梳理了基于设计瞬态信息的疲劳损伤包络分析准则和采用疲劳裂纹扩展的损伤容限分析方法。The pressurizer surge tube of the pressurized water reactor(hereinafter referred to as“surge line”)has the phenomenon of cold and hot fluid stratification,which affects the safe operation of nuclear power plant.At the same time,there are some uncertainties in the thermal stratification condition,and there are few analysis contents based on the real monitoring data of the nuclear power plant.There is still a lack of effective envelope method based on the design transient information with thermal stratification.For the case of significant fatigue damage,the current evaluation method based on fatigue crack initiation criterion is difficult to meet the needs of long-term safe operation of nuclear power plants.In view of above technical situation,by investigating the work of the domestic and foreign scholars in the study of thermal stratification of surge line,the mesh division,material information setting,boundary condition selection,thermal stratification flow simulation and structural stress response analysis technology in finite element numerical simulation were discussed.At the same time,the real operation monitoring data of a large pressurized water reactor nuclear power plant in China was analyzed.The damage envelope analysis criterion based on design transient information and the damage tolerance analysis method based on fatigue crack propagation were also studied.

关 键 词：波动管 热分层 疲劳 瞬态包络 损伤容限 

分 类 号：TM623.91[电气工程—电力系统及自动化] TM44