关于核电厂可靠性保证大纲的核安全审查探讨

核电厂可靠性保证大纲对于提升设备的可靠性、可用率、可维修性和经济性具有重要作用。在设计阶段,对于风险重要的SSCs进行分析、归类,确定合理可行的可靠性参数指标,并在每一个阶段制定严格的质量控制措施,整体SSCs的累积可靠性会大幅提高,从而电厂的安全性和经济性会不断改善。通过分析当前国内外的可靠性法规要求,结合最新的研究成果和技术见解,整理提出了设计可靠性保证大纲的构成要素,并就大纲的审查问题进行了分析探讨,提出了一些建议措施供监管部门和设计单位参考。      

核电厂运行阶段可靠性工作计划概述

为了保障可靠性保证大纲的有效实施,核电厂需编制相应的可靠性工作计划.文章开展了国内外可靠性工作计划相关标准的调研,结合国内核电厂的应用实际,将核电厂运行阶段可靠性工作计划划分为性能监测、性能评估、问题优先级、问题分析和纠正性措施建议、纠正措施的实施及反馈五个工作要素,并阐述了各要素的基本内容.可靠性工作计划的顺利实施,...     

核工业可靠性工作展望

1 核工业可靠性工作简况 核工业的可靠性工作开始于80年代初,通过培训和学习掌握了一些可靠性技术。80年代中期以来,随着军用核动力、核燃料分离技术和秦山核电厂研究工作的进展,可靠性的定量要求作为重要技术指标列入研制任务书,可靠性工作日益受到重视,各级人员的可靠性意识有了很大提高。我们相继宣贯了《军工产品质量管理条例》和GJB450《装备研制和生产的可靠性通用大纲》等一系列法规和标准;结合核工业发展的需要,制定了一些核工业的标准,如GB7163《核反应堆保护系统的可靠性分析要求》、GB/T9225《核反应堆保护系统可靠性分析一般原则》、GB11931《保证所收集的核电厂可靠性数据质量的导则》和GB11932《核电厂可靠性数据交换导则》等;中核总还召开了多次可靠性研讨会和可靠性论文发布会,从而大大推动了核工业可靠性工程的开展。     

压水堆蒸汽发生器传热管完整性要求

１ 相关标准１．１　美国联邦法规１０ＣＦＲ５０要求符合５０．２１（ｂ）或 ５０．２２的运行许可证持有者必须制定和执行一个保证大纲,该大纲要保证压水堆蒸汽发生器传热管的安全功能。首要的安全功能是由于蒸汽发生器传热管是反应堆冷却剂压力边界（ＲＣＰＢ）的主要组成部分,必须要保持反应堆冷却剂的总量和压力。其次,蒸汽发生器传热管作为一、二回路之间的热交换导热体,还保证了反应堆的停堆能力。第三,蒸汽发生器传热管隔离了一回路系统里的放射性介质,避免它们进入二回路系统和释放到环境中去。 制定传热管完整性大纲是为了…     

AP1000非安全级系统的监管要求(RTNSS)探析

根据核行业经验和福岛核事故经验反馈,美国核管会在2014年将用于审查美国核电厂安全分析报告的《标准审查大纲》新增了19.3章"非能动先进轻水堆非安全级系统的监管要求"。其中"增强的设计标准"从纵深防御的角度全面提高了非能动先进轻水堆事故72h后和地震后所使用的重要非安全级系统的可靠性和可用性要求。非能动先进轻水堆AP1000设计与标准审查大纲的一致性评估是核安全监管当局的审查重点,也是核电厂设计的重要工作之一。首先介绍了非安全级系统监管要求的演变历程和实施步骤,其次评估了AP1000设计与《标准审查大纲》19.3章要求的一致性,并进一步从可用性、抗震能力、飓风、内部灾害以及水淹防护等多个因素重点分析AP1000设计能否满足"增强的设计标准"要求。最后针对AP1000无法满足《标准审查大纲》19.3章的情况给出具体解决方案的建议。     

海阳核电厂RCM的应用研究

简要介绍海阳核电厂以可靠性为中心的维修(RCM)的应用现状,分析其后续可能的应用领域,包括设备可靠性分级的优化、性能监测及预测性维修大纲的开发、预防性维修大纲的优化、风险重要设备的管理、备件储备的优化等。通过应用RCM能帮助用户制定完善的设备维修管理策略,提高设备可靠性和可用率,降低维修成本。     

管理部门审查的法规要求和操作实践

1 引言 核安全法规《核电厂质量保证安全规定》[HAF0400（91）]要求所有质量保证大纲“必须规定,参与实施大纲的单位的管理部门要对其负责的那部分质量保证大纲的状况和适用性定期进行审查。当发现大纲有问题时,必须采取纠正措施。”安全导则《核电厂质量保证大纲的制定》（HAF0401）把上述要求具体表述为“必须定期地或在情况需要时进行管理部门的审查。应由本单位关键岗位上的管理人员参加这些审查。当需要时,还应邀请参加较低层次分大纲各单位的适当人员参加。审查必须包括由其他单位对质量保证大纲实施情况的监控,验证是否制定了计划并适时完成了有关的工作,以保证计划和工作的有效性,并验证是否已真正达到质量要求。”     

民用核安全设备无损检验中的质量保证要素

在总结和分析国内有关无损检验单位质量保证的良好实践及经验的基础上,根据核安全质量保证法规的原则和要求,提出了核安全设备无损检验质量保证大纲应包括的主要要素,可作为相关无损检验单位在制定质量保证大纲时的参考.     

核电厂火灾探测和报警系统预防性维修大纲优化

通过对预防性维修大纲编制和优化原则介绍,以核电厂火灾探测报警系统预防性维修大纲的优化为实例,说明了根据现场实际,遵循合理原则,采取合适方式对预防性维修大纲进行优化,提高关键设备可靠性,节约运行成本,达到保障核电厂安全稳定运行的目的。     

秦山核电二期工程反应堆及反应堆冷却剂系统设计质量保证

为确保核电厂的安全,在秦山核电二期工程的设计过程中,按照核安全法规HAF0400及其相关导则规定的原则和要求,制订并实施了设计质量保证大纲.这一质保大纲为设计规定了各种控制、验证措施,使所有影响设计质量的活动都在受控状态下进行并达到了期望的设计质量.     

大亚湾核电站18个月换料设计质量保证大纲的建立与实施

质量保证体系标准ISO9001-94呼国家核安全法规HAF003（91）对质量体系的要求有所差异，这是核行业质量认证组织面临的主要问题。如何才能同时满足两套标准建立体系，中国核动力研究设计院（NPIC）在大亚湾核电厂18个月换料设计论证项目上作了尝试，并建立、实施和不断完善了质量保证体系，制订并实施了专用的质量保证大纲，采取了一系列的控制、验证措施，向法马通公司（FAMATOME）和大亚湾核电厂提供满意的核设计服务产品。     

核设备制造中的项目管理

核电站项目管理首先应完善管理机构。核设备质保大纲和项目管理的组织机构由核电容器部、制造部、质量保证部等5个部门组成。总经理对核承压设备制造质量负全面责任。分管核电工作的副总经理统一负责公司核承压设备制造有关的质量、技术和进度,组织管理部门进行审查。质量管理部部长由总经理授予足够的权力,包括不受经费和进度的约束,以建立和不断完善质量保证大纲,并验证有关部门和人员的执行情况。制造进度计划的编制应该保证实施的可行性、工作的连续性和执行计划的灵活性。对计划应该进行全过程的跟踪,检查计划的实施情况并及时反馈。     

关于质量保证的分级管理

介绍了核电厂（核岛、常规岛、BOP）物项、服务和工艺质量保证分级管理方面的相关法规、导则和技术文件中的要求,目的,对相关单位的要求,分级原则和级别划分。对各级质量保证要求的差别的考虑方法,以及国内质量保证分级管理现状和建议。     

Graded approach to establish QA requirements for type B and fissile material transportation packagings

Abstract

The essence of the graded approach is the establishment of applicable quality assurance (QA) requirements to an extent consistent with the importance to safety of an item, component, system or activity. The genesis of the graded approach is a study conducted by the US Nuclear Regulatory Commission (NRC) for the US Congress in 1987 to assess the effectiveness of QA activities. That study demonstrated the need to improve the application of QA requirements for the nuclear industry in general. The conclusion of the study indicated that a graded approach for establishing QA requirements is the most viable method to satisfy federal safety standards that result in protecting public health and safety. The application of QA requirements for type B and fissile material transportation packagings is not based solely on importance to safety or safety related considerations. The operability of items, components, systems and activities is considered to be equally important. The nuclear industry, along with regulatory agencies, recognises the significance of operability considerations, as well as the evaluation of each item, component, system or activity for safety related considerations. The graded approach for QA requirements for type B and fissile material transportation packagings is based on Title 10, Part 71 of the US Code of Federal Regulations (CFR), ‘Packaging and transportation of radioactive material.’ Guidance for implementation of the QA requirements specified in §71 is provided in NRC Regulatory Guide 7·10, ‘Establishing quality assurance programmes for packaging used in transport of radioactive material,’ and ASME NQA-1, ‘Quality assurance requirements for nuclear facility applications’. The graded approach for QA requirements is based on criteria for containment, shielding and subcriticality specified in 10 CFR Part 71.     

Approach for fusion reactor safety and fusion safety work in Russia

Main directions of work on experimental fusion reactors safety assurance in Russia are given. Work on safety includes: the elaboration of the main criteria and principles of safety assurance, the development of the first priority standards in safety on the basis of the fission experience and international safety documents requirements, fusion reactor safety analysis, and work to provide a base for the standards development and for the safety analysis activity. The results of some work on fusion safety are presented. They include: assessments of safety and reliability of Liquid Metal Cooling System draft design, evaluations of the buildings and equipment response on external dynamic influences, and analysis of radiological situation in th environment as a result of tritium-containing dust release.     

Quality assurance for objects utilizing nuclear power

The basic aspects of quality assurance for purposes of safety of objects utilizing atomic energy are examined. The formulation of quality assurance requirements of a regulatory agency is analyzed, the experience with enterprises in nuclear power and the atomic industry is examined, and current trends in quality assurance, which are reflected in IAEA manuals and ISO series 9000 standards, are analyzed. The relationship between the quality system and the quality assurance program for objects utilizing atomic energy is discussed. 1 figure, 1 table, 7 references. National Science Center YARB of the Federal Nuclear and Radiation Safety Agency of Russia. Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 230–235, September, 1999.     

碘吸附器用吸附介质的国内外标准差异对比与分析

国家能源局标准NB/T20039.15—2012《核空气和气体处理规范 通风、空调与空气净化 第15部分:吸附介质》规定了用于核设施空气和气体净化系统中吸附介质的性能、设计、验收试验和质量保证等内容的最低要求,是保证核设施安全运行的重要文件。本文对该行业标准与国外标准各版本的差异进行了比较,并对差异的原因进行了分析,也对相关于此类吸附介质的导则或标准的立场进行了综合说明,便于专业技术人员充分了解核设施对于该类吸附介质的相关要求及其背景。     

核工程土建及安装质量的保证及控制

从如何建立核工程建造的质量保证体系的方法入手,分别介绍了质量保证体系的建立及运转、承建单位的QA和QC控制、监理公司的质量监督和控制、建设单位的质量保证监查和监督等工程经验.     

Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles

As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.