WOG核电厂风险指引型管道在役检查方法应用研究

为促进概率安全分析技术在核电厂管道在役检查领域的更好应用,本文介绍西屋用户集团(WOG)开发的核电厂管道风险指引型在役检查(RI-ISI)优化方法,并重点从管段失效可能性分析、后果判断、风险重要度划分等三方面分析对比了该方法与EPRI型RI-ISI方法的不同。此外,以国内某M310核电机组为例,本文基于国家安全局牵头开发的标准电厂分析风险(SPAR)模型,在国内当前技术条件基础上使用简化WOG方法完成该核电厂辅助给水系统管道环焊缝的RI-ISI优化分析。计算表明,使用WOG方法开展RI-ISI后,受检焊缝数量减少55%,而相应导致的内部事件一级概率安全分析风险增量则基本为零,可以满足NNSA-0147和NNSA-0153等技术文件中推荐的风险准则。总的结论为,使用WOG方法开展核电厂管道RI-ISI优化是可行的。     

核电厂风险指引型管道在役检查应用研究

简要介绍了风险指引型管道在役检查(RI-ISI)的发展现状和方法,给出大亚湾核电站RI-ISI试点研究的结果.并阐述了在国内核电厂实施RI-ISI存在的难点及建议.     

风险指引型在役检查优化申请的独立审核计算

本文详细介绍了环境保护部核与辐射安全中心针对风险指引型在役检查(RI-ISI)优化申请所开展的独立审核计算。在大亚湾核电厂RI-ISI优化申请的审评工作中,采用国家核安全局标准概率安全分析(PSA)监管模型,计算管段失效后果和拟实施变更后的风险增量,对申请者管段失效后果分析结果进行核算,并独立评价该申请是否满足风险可接受准则。实现了核安全监管部门对PSA应用试点项目的独立审核计算,为核安全决策提供进一步的支持,提高核安全监管的独立性、科学性和有效性。     

关于EPRI型风险指引管道在役检查优化方法的探讨及改进建议

当前国内核电厂普遍采用EPRI型方法开展风险指引管道在役检查优化,其需要完成管段失效可能性分析、管段失效后果分析及风险增量计算等工作,对此,本文开展探讨研究并论述其中可能存在的问题。此外,通过对风险指引型分级方法理念及WOG风险指引型管道在役检查优化方法的简要介绍和探讨,本文提出不断提高管段失效可能性计算水平的要求以及结合使用风险减少因子(Risk Reduction Worth,简称RRW)和风险增加因子(Risk Achievement Worth,简称RAW)完成管段失效后果分析的改进建议,以在我国当前技术水平条件下,找出一套能够恰当评价核电厂风险变化的在役检查优化方法。     

大亚湾核电站风险指引型在役检查分析

详细描述大亚湾核电站风险指引型在役检查分析的方法及实施流程,并应用该方法对大亚湾核电站1号机组余热排出系统现行在役检查进行优化。优化后,余热排出系统的检查位置和检查数量发生变化。风险指引型在役检查方法能在保持和改善公众健康和安全的同时,尽可能地减少检查数量进而减少在役检查人员的职业照射。然而,风险指引型在役检查的关键不是简单地减少检查数量或改变检查方法,更重要的是关注风险重要的管段,检查基于"降质机理"(基于原因),能够提高在役检查的效果。文章最后对风险指引型在役检查方法实施过程中一些关键问题进行讨论,为方法的应用提出若干建议。     

M310型核电厂在役检查的改进和优化

核电厂在役检查是核电厂机械设备安全运行的重要保证手段之一。本文结合国内核电厂在役检查的经验反馈,对RSE-M规范中某些技术条款的合理性进行探讨,介绍风险指引型在役检查方法的应用,并为我国M310型压水堆核电厂在役检查的优化提供建议。     

核电厂射线探伤安全监督经验与改进

射线探伤作为一种常用的工业无损检测技术,在使用过程中具有电离辐射风险,如操作不当将造成人员超剂量照射事故。核电厂在设备管道、压力容器的役前和在役检查中大量采用射线探伤技术,所使用的设备以移动式探伤设备为主,设备存储、运输、作业过程中均存在风险。某核电厂在总结历史事件经验的基础上,对射线探伤辐射风险因素进行了归纳与总结,并对射线探伤风险管理与安全监督进行了改进,应用至今取得了良好的效果,保障了作业人员及公众的辐射安全。     

含缺陷承压管道裂纹扩展计算及程序开发

通过对某核电厂承压设备在役检查发现缺陷后的处理计算,阐述了基于线弹性断裂力学机理的裂纹扩展的计算分析和评定方法,计算方法中考虑了缺陷深度和长度的耦合作用。并在此基础上,运用Visual C++语言,开发裂纹扩展计算程序。程序内全面考虑了反应堆冷却剂系统(RCS)一次侧设计瞬态、反应堆冷却剂系统二次侧设计瞬态的交变载荷及缺陷初始形状、方位分类等问题,适用范围广泛。本文的计算方法及评定方法为电厂管道在役检修和能否继续运行提供重要依据。     

核电厂商业运行前管道支吊架检查与调整

详细描述在核电厂调试期间对管道支吊架进行冷、热态检查的依据、内容和方法;对检查出有异常支吊架的维修、调整策略及机组临界前热态工况下管道支吊架的验证方式加以说明。提出了支吊架的安装、改造和在役检查中的注意事项,力求通过对管道支吊架的检查与调整,使管道支吊架处于正常的工作状态,保证系统管道安全可靠地运行。     

含内表面裂纹核级管道的初始塑性失效分析

我国在役和新建的大部分核电厂在主管道上应用了破前漏技术,针对该技术ASME采用净截面屈服准则对完全塑性断裂进行缺陷评定,大量研究表明,净截面屈服准则高估了结构的承载能力。本文采用有限元方法模拟了含内表面裂纹的核级管道在内压作用下的变形过程,并利用裂纹前沿J积分随内压变化的曲线特征确定了含裂纹管道的初始塑性失效载荷。随后,将初始失效载荷的计算值与ASME规范定义的理论值相比较,结果表明理论解高估了结构的承载能力。最后,评价了ASME-BPVC-XI规范中A级使用限制对应的允许薄膜应力的适用性。     

Updating of risk-informed ISI programme for Ignalina RBMK-1500 nuclear power plant in Lithuania: Results and challenges

The paper presents a risk-informed in-service inspection (RI-ISI) pilot study project of 300 mm piping at Ignalina nuclear power plant (INPP) RBMK-1500 reactor, located in Lithuania. The RI-ISI study investigates optimal 300 mm piping ISI strategies with respect to risk and required resources. In total 1240 stainless steel welds were analyzed, assuming inter-granual stress corrosion cracking (IGSCC) to be the main damage mechanism. Pipe break frequency was estimated by probabilistic fracture mechanics methods and combined with safety barriers, provided by probabilistic safety assessment (PSA) study.After 3 years of operation, updating of RI-ISI was performed by taking into account new statistical data on pipe defects. Comparison with previous RI-ISI program was performed. The paper includes discussion on uncertainties in the study and robustness of RI-ISI programs.     

Risk-informed inservice inspection program

The Nuclear Regulatory Commission (NRC) has developed draft guidance for power reactor licenses on acceptable methods for using probabilistic risk assessment (PRA) information and insights in support of plant-specific applications to change the current licensing basis (CLB) for inservice inspection (ISI) of piping. This process is also known as risk-informed inservice inspection programs (RI-ISI). The risk-informed inservice inspection process for operating nuclear power plants provides an alternative method for selecting and categorizing piping components that are inspected for the purposes of meeting the requirements of ASME Section XI. A RI-ISI approach will incorporate probabilistic techniques to help define the scope, type and frequency of inservice inspection. The risk-informed process may support a decrease in the number of inspection and inspection intervals but will also identify areas where increased resources should be allocated to enhance safety. The approach discussed in this paper follows the method developed by NRC staff.     

Case studies of risk-informed inservice inspection of nuclear piping systems

In the present paper, a probabilistic failure analysis is used to find failure probabilities of piping segments, and a probabilistic risk assessment model is employed to obtain risks to a nuclear power plant should these failures occur. The multiplication of the piping failure probability and the consequence for that particular failure results in the risk contribution of the pipe. The degrees of risk for different piping segments can then be ranked, and their results can be used as the basis for planning a risk-informed inservice inspection program. Numerical studies are offered with special emphases on: (1) the status and experience with RI-ISI applications in Taiwan; (2) the comparison of risk-rankings performed with three different methods developed in the US; (3) aspects of the probabilistic fracture mechanics calculation including the flaw size distributions and stress corrosion cracking model. The results indicate the proposed method can indeed be adopted for planning a cost effective inservice inspection program.     

Risk-informed inservice inspection

The operational readiness and functional integrity of certain safety-related piping and associated structural elements such as piping supports are vital to the safety of operations in nuclear power plants. Inservice inspection (ISI) is one of the mechanisms used by the power plant owners to ensure piping integrity. Previously, the type and frequency of ISI have been based on the collective best judgment of the NRC and industry in a consensus code and rulemaking process. The ASME code-based ISI requirements and practices have not explicitly taken into consideration unique aspects of piping functions, piping degradation mechanisms, weld integrity, fabrication details, and the extent of the contribution to overall plant risk. Due to the general nature of the ASME code ISI requirements and non-reliance on quantification of risk estimates, current ISI requirements may unnecessarily emphasize inspection of less safety-significant piping segments, and thereby unnecessarily expose plant personnel to radiation exposure. Nuclear power plant owners are currently interested in optimizing inspection and testing by applying resources in more safety-significant areas. They are also interested in maintaining system availability and reducing overall maintenance costs which do not have any adverse effects on safety. The NRC has confirmed its intent of using probabilistic, as an adjunct to deterministic, techniques, to help define the scope, type, and frequency of ISI. The development of risk-informed inservice inspection programs (RI-ISI) has the potential to optimize the use of NRC and industry resources and to continue assuring adequate protection of the public's health and safety. Currently there are two methodologies being proposed by the industry for the implementation of the RI-ISI programs. One is being developed jointly by the ASME Research and Westinghouse Owners Group (WOG) and the other by the Electric Power Research Institute (EPRI). Both methodologies will be implemented for pilot plant applications. Based on discussions with the interested licensees, the NRC staff has tentatively accepted Surry, ANO-2, and Fitzpatrick as the RI-ISI pilot plants. The Surry pilot application is based on the WOG methodology, whilst ANO-2 and Fitzpatrick are based on the EPRI methodology.     

压水堆核电厂核岛机械设备在役检查规则研究

压水堆核电厂核岛机械设备在役检查规则研究是修订和编制我国相关核电在役检查标准的基础和前提。本文简介了在役检查规则研究目标、方法、主要内容和结果以及在役检查规则制定依据,简述了规则研究相关主要问题的处理方法和结果,对比分析了依据研究结果编制的NB/T 20312标准与EJ/T 1041标准在役检查规则的主要不同点,给出了准确理解和正确应用NB/T 20312标准有关在役检查规则的提示和说明,为有效应用该标准在役检查规则提供重要参考。     

A study of risk-based non-destructive in-service inspection

Based on the concept of probabilistic fracture mechanics, non-destructive inspection policies are studied in this paper. Special attention is paid to the reduction of structural failure probability through the employment of in-service inspection (ISI). Analytical formulation is derived and numerical examples are presented. In addition to the conclusions drawn from numerical experience, the present study indicates how appropriate ISI schedule can be planned by considering the risk of structural failure in situ.     

核电厂围板螺栓超声检测技术研究

围板螺栓是核电厂堆内构件的关键连接部件,长期服役下可能产生辐照应力腐蚀裂纹（IASCC）等缺陷,有必要对其结构完整性进行无损检测。分析围板螺栓的结构特点和在役检查工况,开发针对外六角头结构螺栓的组合晶片超声检测方案,介绍探头设计选型原则和缺陷评定技术,确保良好的声场有效覆盖以及检测出螺栓不同部位的裂纹缺陷。通过对含缺陷试块的试验验证了超声检查工艺的可行性,结果表明该技术能够有效检测30%螺杆横截面当量的裂纹缺陷,信噪比可达12 dB以上,满足在役检查要求。      

核电厂BOSS焊缝在线堆焊修复相控阵超声检验技术

核电厂反应堆换料水池与乏燃料水池冷却和处理系统（PTR）及设备循环冷却系统（RRI）中含有大量管座接头（BOSS）焊缝,其安全性和可靠性直接影响所存储核燃料的安全状态,对其进行缺陷排查和在线修复是核电厂在役检查监督的重点和难点。本文针对BOSS焊缝在线堆焊修复的特殊要求和检验难点以及射线检验的局限性,设计了一套专用的相控阵超声探头和检验工艺,试验验证结果满足堆焊修复要求,并制订了核电厂BOSS焊缝堆焊修复无损检验的方法和在役检查监督的策略。     

核电厂一回路管座焊缝超声波检测仿真研究

小尺寸支管接头（BOSS）焊缝作为核电厂一回路压力边界的薄弱环节,对其有效监控是核电厂日常和在役大修的重点和难点。采用仿真技术、工艺试验和现场应用验证等方法,设计并验证了BOSS焊缝的超声波相控阵检测工艺,解决了核电厂日常和在役大修中BOSS焊缝的监督难点。并得到类似超声波相控阵检测工艺的设计和验证方法。