非能动先进压水堆核电厂SGTR事故概率安全评价  

Probabilistic safety assessment for SGTR in advanced passive nuclear power plant

作　　者：潘亚兰[1,2] 栾秀春 王喆[1] 左嘉旭[1] 宋维[1] 

机构地区：[1]环境保护部核与辐射安全中心北京100082 [2]哈尔滨工程大学核安全与仿真技术国防重点学科实验室哈尔滨150001

出　　处：《核技术》2016年第8期73-78,共6页Nuclear Techniques

基　　金：国家科技重大专项(No.2013ZX06002001-004)资助~~

摘　　要：蒸汽发生器传热管破裂(Steam Generator Tube Rupture,SGTR)事故是核电厂的重要事故之一,并具有其自身的特点。该事故的研究和评价对核电站安全具有较大意义。选取典型非能动先进压水堆核电厂AP1000的SGTR事故进行一级概率安全评价(Probabilistic Safety Assessment,PSA),采用事件树分析方法得到电厂事件发生后系统、设备和人员不同响应所产生的事故序列,然后建立相关系统的故障树模型进行可靠性分析。借助Risk Spectrum软件,计算SGTR事故导致AP1000核电厂的堆芯损伤频率(Core Damage Probability,CDF),并进行堆芯损伤的最小割集分析及重要度和敏感性分析。通过一系列分析得到导致堆芯损伤的重要基本事件,从而找到系统存在的薄弱环节。Background： The Steam Generator Tube Rupture （SGTR） accident, which has its own characteristics, is one of the important accidents in nuclear power plants, and it is significant to the safety of the nuclear power station. Purpose： In this paper, AP1000 reactor is selected as the typical advanced passive nuclear power plant to analyze the core damage consequence caused by SGTR accident, so as to find out the weak links existing in the system. Methods： The Probabilistic Safety Assessment （PSA） method in level one has been used to analyze the SGTR accident in API000. After the power plant accident occurs, systems, equipment and personnel respond differently, event tree analysis method is used to obtain sequence, and the systems related to this accident are analyzed by fault tree models. Results： By the Risk Spectrum software, the total Core Damage Probability （CDF） has been calculated, and the minimal cut sets, the importance measures and the sensibility of the core damage have also been analyzed respectively. Conclusion： According to a series of analysis results, the most important basic events resulting in the core damage can be obtained, and the weak link of the system can be found, which has a certain theoretical support for the further strengthening of accident prevention and mitigation of the core damage caused by the accident.

关 键 词：能动先进压水堆核电厂 蒸汽发生器传热管破裂 堆芯损伤频率 概率安全评价 

分 类 号：TL364.5[核科学技术—核技术及应用]