轻水反应堆RBI-Part1
CRTD-Vol. 20-2
基于风险的检测的导则制订
轻水反应堆(LWR)核电站部件
PREPARED BY
The Research Task Force on Risk-Based Inspection Guidelines
APPOTNTED BY
The Codes and Standards Research Planning Committee of the ASME Center for Research and
Technology Development
For
The ASME Council on Codes and Standards
The United States Nuclear Regulatory Commission
The National Board of Boiler and Pressure Vessel Inspectors
The Pressure Vessel Research Committee - Welding Research Council
American Nuclear Insurers
The Hartford Steam Boiler Inspection and Insurance Company
Industrial Risk Insurers
The American Petroleum Institute
The National Rural Electric Cooperative Association
The United States Department of Energy
Oil Insurance Limited
Edison Electric Institute
REVIEWED AND EDITED BY
Steering Commit tee on Risk-Based Inspection Guidelines and an Independent Peer Review
Committee
NRC GRANT NO. NRC-04-89- 102
THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS
United Engineering Center 345 East 47th Street Now York, N.Y. 100 17
DISCLAIMER
This report was prepared as an account of work sponsored through the American Society of Mechanical Engineers (the Society) Center for Research and Technology Development by the United States Nuclear Regulatory Commission, the National Board of Boiler and Pressure Vessel Inspectors, the Pressure Vessel Research Committee - Welding Research Council, the American Nuclear Insurers, The Hartford Steam and Boiler- Inspection and Insurance Company, the Industrial Risk Insurers, the American Petroleum Institute, the National Rural Electric Cooperative association, The United States Department of Energy, Oil Insurance: Limited, and Edison Electric Institute (collectively referred to herein as the Sponsors).
Neither the Society. nor the Sponsors, nor Westinghouse Electric Corporation, the University of Maryland, Rolls Royce and Associates Ltd., Battelle Pacific Northwest Laboratories, Failure Analysis Associates, Inc., Factory Mutual Research Corporation. Idaho National Engineering Laboratory, and the McDonnell Aircraft Company (collectively referred to herein as the Sponsorees), nor any financial contributors or others involved in the preparation or review of this report. nor any of their respective employees, members, or persons acting on their behalf, makes my warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe upon privately owned rights.
Reference herein to any specific commercial product, process or service by trade name. trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the Society, the Sponsors, the Sponsorees, or any financial contributors or others involved in the preparation or review of this report, or any agency thereof. The views and opinions of the authors, contributors, and reviewers of the report expressed herein do not necessarily reflect those of the Society, the Sponsors, the Sponsorees, or any financial contributors or others involved in the preparation or review of this report, or any agency thereof,
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advanced in papers… or printed in its publications ( 7.1.8)
ISBN Nn. 0.791 8-0658-8
Library of Congress
Catalog Number 92-5.1327
Copyright: I- 1992 by
THE AMERICAN SOCIETY OF MECHANTCAL ENGINEERS
All Rights Reserved
Printed in U.S.A.
Although this document represents the work of the research task forcc mcmbers, this study would not bo possible without the contributioris of a large number of leaders in their respective fields from academia, government, and indust1y.
The steering committee members have carefully guided the project. and the independent peer review members teamed with the steering committee to diligently review and edit this document. The valuable and gencrous cnntrihution of these members, who are identified in this document, is most appreciated. The rosua~chta sk force acknowledges with appreciation thc contributions of Truong Vo of Battelle Pacific Northwest Laboratories, who attended most of the meetings and provided results from scvcral of the pilot studies cited herein. He has been invited to be an honorary member of the task force because of his significant contribution. Comments by Dr. Lee Abramson of the U.S+ Nuclear Regulatory Commission regarding the elicitation of expert opinion and his provision of an equation format for the risk-based inspection methodology were much appreciated.
Charles H. Boyd, Barney L. Silverblatt, Richard E. Schwirian, and Barry D. Sloane of Westinghouse Electric Corporation's Nuclear and Advanced Technology Division are noted for their contribution to the inspection prioritization example for reactor internals subcomponents. Dr. Robert Perduc and Bruce A. Bishop of Westinghouse Electric Corporation's Science and Technology Center and Nuclear and Advanced Technology Division. respectively, are cited for their valuable assistance in preparing the decision analysis example for choosing an inspection strategy.
Finally, the Westinghouse Energy Center Word Processing staff and the ASME Technical Publishing Department members are acknowledged for their dedicated and diligent efforts in compiling, editing, and publishing this document.
基于风险的检测导则研究指导委员会
Raymond J. Art, Assistant Director, ASME Center for Research and Technology
Dcvclopment , Washington, D.C.
Robert J. Bosnak, ASME Council on Codes and Standards, ASME Codes and
Standards Research Planning Committee, ASME Board on Research and
Technology Development, Deputy Director - Division of Engineering, Office
of Nuclear Regulatory Research, U.S. Nuclear Regulatory Cornmission.
Washington, D.C.
Dr. Spencer J. Bush, Consultant, Past Chairman - ASME Section XI, Richland,
Washington
John Blackburn, American Petroleum Institute, Washington, D.C.
Ray Davies, Dc t Norse Veritas Industrial Services, Inc.
Thdore A. Meyer, Manager, Westinghouse Electric Corporation, Pittsburgh,
Pennsylvania
Evangelos Michalopoulos, P.E., Senior Engineer. The Hartford Steam Boiler
Inspection and Insurance Company, Hartford, Connecticut
Dr. Joseph Muscara, Senior Metallurgical Engineer, U. S + Nuclear Regulatory
Commission. Washington, D.C.
Michael E. G. Schmidt, P.E., Research Consultant, Industrial Risk Insurers,
Hartford, Connecticut
Ernest W. Thrmkmorton, Virginia Power, Glen Allen, Virginia
William G. Wendland, P.E., Manager - Engineering Projects, American Nuclear
Insuruers, Farmington, Connecticut
viii
独立同业互查委员会
Dr. Vicki Bier, Professor, University of Wisconsin, Madison, Wisconsin
John D. Boardman, Inservice Inspection Engineer, Southern California
Edison, San Clemente. California
Michael Beford, Robin L. Dyle, and Dennis M. Swann, Engineers, Inspection and Testing Services, Southern Nuclear Operating Company,
Birmingham, Alabama
T. N. (Bud) Epps, Chairman ASME BPVC Section XI Long Range Planning Committee, T.K.S. International Inc., Birmingham, Alabama
CRTD-Vol. 20-2
基于风险的检测的导则制订
轻水反应堆(LWR)核电站部件
执行摘要
本文件是题为“基于风险的检测的导则制订”系列的第二部分,正由美国机械工程师学会(ASME)的多学科特别小组制订。第一刊第一卷——一般性文件(ASME 1991)说明了一个总体的基于风险的程序,该程序可被用来制订任何工业设施或结构系统的检验指南。该系列的后续各卷给出了阐明特定工业的结构完整性问题的一般方法论的具体应用。
本文件(第二卷第一部分)即是首个这种应用。文件是针对轻水反应堆(LWR)核电站部件的检验。
第一卷中推荐的以基于风险的检测的一般性程序提供了一个以节省成本的方式配置检验资源的总体框架,并有利于将检查运用到最需要的地方。该一般性方法论已对在核部件中的应用作了进一步地规定和扩充。 该程序包括以下五个部分:
(a) 系统的定义;
(b) 定性风险评估;
(c) 定量风险分析,包括失效模式、影响与危害性分析(FMECA)以选择待检验的部件并对其进行评级;
(d) 采用分摊法选择单个部件的目标失效概率以使因所有部件的失效而产生的影响保持在总风险目标值以下;
(e) 确定一个使用具有结构可靠性/风险评估(SRRA)方法的风险决策分析法将部件失效概率保持在目标值以下的费用低的检验计划。
对于程序的头两部分,已经为LWR核电站作了充分说明,定性风险方法也已暗含在现行的检验计划之中。说明制订核部件检验计划的定量风险分析的使用是本第二卷所报告的研究工作的主焦点。尤其是:
●现在已经为许多核电站而产生的风险概率评估信息的使用已并入该方法论中,从而可以提高与部件压力边界失效相关的风险的量化程度。
●已根据这些定量风险估计为规定通过检验活动保持的目标部件失效概率值推荐了一个程序。
●已经阐述了确定一个检验计划必须具有的特性以便在考虑经济因素的同时符合目标失效概率的方法。
已经进行了基于风险的检测方法论的初步试验,包括萨里-1(Surry-1)发电厂的一个主要研究。在整个第二卷第一部分介绍该程序的各个部分的结果。先导性研究证明以下:
●整个程序中可计算并使用定量风险值,说明该方法论切实可行;
●使用风险决策分析和SRRA方法可以结合安全和经济因素,从而选择最佳的部件检验策略;
●还需要合理的资源投入;
●结果合理,并符合常识性质量评定。
尽管基于风险的检测的方法可用于整个轻水堆核电站,今后努力的主要目标将是全面推荐采用ASME锅炉压力容器规范(BPVC)第XI卷。为了BPVC第XI卷相关组的考虑,第二卷第二部分计划为核部件推荐一个基于风险的检测计划,包括一个坚实的技术基础。
今后的工作应:
●完成萨里-1核电站各部件失效风险;
●进一步确定整个电站上结果能够普及化的程度;
●论证用于所选部件的检验策略的制订将以一种节约成本的方式使由于元件失效引起的风险保持在目标风险值以下。
尽管完成此项工作需要相当长的时间,但基于风险的检测的一些关键好处如下: ●运用其程序和方法时获得的真知卓见;
●在参与保持商用核电站的安全可靠运行的 众多专业和组织之间的交流得到增强。
目录
致谢
执行摘要
基于风险的检验导则特别研究小组 基于风险的检验导则研究指导委员会 独立同业互查委员会
基于风险的检测—导则制订:出版物清单 出版物清单—ASME研究和技术开发中心 1 引言
1.1 与一般的基于风险的检测方法论的关系
1.2 目标与范围
2 基于风险的检测的全过程
2.1 综述
2.2 系统定义
2.3 定性风险评估
2.4 定量风险分析
2.5 故障模式影响及危害性分析(FMECA)方法论
2.6 目标失效概率的选择
2.7 考虑安全与经济因素的检验计划的制订 3 定义
4 摘要与建议
4.1 方法论概述
4.2 进一步制订状况与计划
4.3 结论
5 参考文献
5.1 正文参考文献
5.2 附录参考文献
插图
1-1 一个加压水冷反应堆系统的基本元件
1-2 直接循环沸水反应堆系统
2-1 轻水堆核电站部件的基于风险的检测程序
2-2 核电站安全系统的典型分类
2-3 瑞典在确定检验时间间隔时采用的方法
2-4 技术信息与核电站部件的基于风险的检测的FMECA的整合
2-5 使用专家判断估计失效概率的程序
2-6辅助给水(AFW)系统部件的失效频率估计
2-7 反应堆压力容器的失效频率估计
2-8 轻水堆核电站系统与部件的基于风险的评级技术方法与信息
2-9 六个代表性加压水冷却反应堆(PWR)电站的基于风险的评估
2-10 萨里-1电站所选系统的基于风险的评估
2-1 1 两个代表性PWR电站上所选系统的基于风险的评估
2-1 2 大海湾-1(Grand Gulf-1)核电站所选系统的基于风险的评估
2- 13 奥科尼-3(Oconee-3)的应急给水(EFW)管段
2- 14 萨里-1部件堆芯损坏频率的单个影响
2- 15 萨里-1部件的累积风险影响
2- 16 基于堆芯损坏风险的反应堆内部子部件的风险评估
2-17 基于经济风险的反应堆内部子部件的风险评估
2- 18 基于堆芯损坏频率的萨里-1部件的风险评估
2- 19 考虑安全与经济因素后对检查方案的改进
2-20 用于选择备用检验策略的决策树构架
2-21 核电站系统和部件使用寿命期间的在役检验(ISI)的作用
2-22 用于选择备用检验策略的决策树构架
2-23 用于选择备用检验策略的决策树构架
2-24 超声检验时一个裂纹探测不到的概率与其深度的关系
2-25 使用ASME锅炉和压力容器规范要求进行超声波检验时不能探测到容器缺陷的概率 2-26 使用特殊的程序进行超声波检验时不能探测到容器缺陷的概率
2-27 用于选择检验策略的决策树架构
2-28 对于一个疲劳裂纹扩展损伤机理评价管道可靠性的 SRRA过程步骤示意图
2-29 一个范例中对于各种裂纹深度分布和检验计划一根双头管破裂(DEPB)的累积条件概
率
2-30 适用于管道应力腐蚀裂纹SRRA评估的一个扩展的“PRAISE”模型的各种部件的示意
图
2-31 加压水冷反应堆容器环带区的辐射脆化SRRA评估的各种部件的示意图
2-32 利用循环试验检测10英寸不锈钢管中晶粒间应力腐蚀裂纹(IGSCC)
2-33 沸水反应堆(BWR)管道中的晶间应力腐蚀裂纹
2-34 根据电力研究院(EPRI)/ 沸水反应堆(BWR)业主集团计划预测晶间应力腐蚀裂纹扩展 2-35 用于选择合适行动的决策树架构
2-36 以压力容器为例的失败概率直方图
2-37 以压力容器为例的置信度—在役检验曲线图
2-38 确定BWR管道中晶粒间应力腐蚀裂纹的位置的改进措施的SRRA评估结果 2-39 制订一个用于低压安全检验系统的检验计划的影响图
2-40 制订一个用于低压安全注射系统的检验计划的方案决策树
2-41 三种检验策略的风险调整成本—风险承受力
2-42 “裂纹出现”的信息价值
表格
2-1 轻水堆核电站部件典型失效原因
2-2 加压水冷反应堆容器部件的潜在退化机理矩阵
2-3 加压水冷反应堆冷却系统管道的潜在退化机理矩阵
2-4 失效模式、影响与危害性分析(FMECA)工作表式样
2-5 应急给水(EFW)系统失效概率
2-6奥科尼-3 应急给水(EFW)系统管段的重要性
2-7 奥科尼-3 应急给水(EFW)系统管段的重要性 与ASME锅炉压力容器第XI卷分类和在役检验要求的对比
2-8 萨里-1电站中所选系统的重要部件
2-9 萨里-1电站所选系统部件与ASME锅炉压力容器第XI卷分类和在役检验要求相比的重
要性
2- 10 轻水堆核电站部件检验可靠性研究的状况
2-11 保证沸水反应堆主系统中每年泄漏的一个规定的概率所要求的在役检验间隔 2-12 保证沸水反应堆主系统中一个规定的可靠性提高的因素所要求的在役检验间隔 2-13 焊缝检验备选抽样方案
2-14 典型压力容器检验结果
2- 15 低压安全注入管线概率估计
2- 16 低压安全注入管线成本因素
2- 17 使用风险决策分析对一个低压安全注射系统的检验策略进行的典型评估
附件
A. ASME锅炉压力容器第XI卷中风险因素的使用
B.专家判断的正式采用
C. 反应堆内件的风险评估
D. 将在役检验与风险管理相联系的方程式
E. 风险状态的量化
图
B-1 专家裁断的流程
D-1 预计检测时间T—失效概率WP的变化D
D-2检验的频次(f)和检测概率(pD)之间的协调
E-1 图解性单一效用函数
表格
C-1 FMECA:基于风险的检测-第XI卷--反应堆内件
C-2FMECA:基于风险的检测-第XI卷--反应堆内件