非能动先进压水堆核电厂SGTR事故概率安全评价

潘亚兰; 栾秀春; 王喆; 左嘉旭; 宋维

doi:10.11889/j.0253-3219.2016.hjs.39.080605

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非能动先进压水堆核电厂SGTR事故概率安全评价

核能科学与工程 | 更新时间：2021-05-24

- 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
- 核技术 2016年39卷第8期文章编号：080605
- 作者机构：
  
  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
- 作者简介：
  
  [ "潘亚兰, 女, 1990年出生, 2016年毕业于哈尔滨工程大学, 现为硕士研究生, 研究方向为核能与核技术工程", "PAN Yalan, female, born in 1990, graduated from Harbin Engineering University in 2016, master student, major in nuclear energy and nuclear technology engineering" ]
  宋维, E-mail:sv98@163.com SONG Wei, E-mail: sv98@163.com
- 基金信息：
  
  国家科技重大专项(No.2013ZX06002001-004)
- DOI：10.11889/j.0253-3219.2016.hjs.39.080605
  中图分类号： TL364+.5
- 收稿日期：2016-04-10，
  
  修回日期：2016-05-21，
  
  纸质出版日期：2016-08
- 稿件说明：
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潘亚兰, 栾秀春, 王喆, 等. 非能动先进压水堆核电厂SGTR事故概率安全评价[J]. 核技术, 2016,39(8):080605

Yalan PAN, Xiuchun LUAN, Zhe WANG, et al. Probabilistic safety assessment for SGTR in advanced passive nuclear power plant[J]. Nuclear techniques, 2016, 39(8): 080605
潘亚兰, 栾秀春, 王喆, 等. 非能动先进压水堆核电厂SGTR事故概率安全评价[J]. 核技术, 2016,39(8):080605 DOI： 10.11889/j.0253-3219.2016.hjs.39.080605.

Yalan PAN, Xiuchun LUAN, Zhe WANG, et al. Probabilistic safety assessment for SGTR in advanced passive nuclear power plant[J]. Nuclear techniques, 2016, 39(8): 080605 DOI： 10.11889/j.0253-3219.2016.hjs.39.080605.

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。

Abstract

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 AP1000. 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

Keywords

references

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