浅谈核电厂核安全级设备焊接质量保证

通过对核安全级设备焊接质量保证体系重要性进行分析,从核安全监管的角度总结和梳理焊接质量保证体系建立的基本要求,并提出了保持质量保证体系有效运行的主要措施和方法,阐述实践和培训是体系提升和改进的主要方式,最终目的保证设备焊接质量受控。     

核电厂及设备的寿期和剩余寿期预测分析方法的研究

截至目前,全世界已有450多座在运核电厂,其中相当一部分核电厂是70年代建造的,这些核电厂已达到或接近预定寿期,在这种情况下,不对其设备进行寿期预测、评估而继续运行,对安全是极为不利的。而对于还没有到预定寿期的核电厂,为了安全管理的需要,其设备的寿期及剩余寿期也应预测。同时,也应对设备运行及核电厂运行的经济性如何这些人们普遍关心的问题进行预测。该文的目的是给出核电厂及其设备寿期与剩余寿期的预测方法。其方法是:通过对核电厂运行的经济性分析,给出预测核电厂寿期的方法;通过损坏计算法和概率统计法及设备运行费用经济性分析的方法给出核电厂的设备寿期与剩余寿期的预测方法;结果是:正确地给出了核电厂及其设备寿期与剩余寿期的预测方法;结论是:该方法适用于核电厂及设备,也适用于其他行业如化工行业的设施及设备。     

核安全重要设备和系统的定期试验以及再鉴定试验

1前言 WWER型压水堆核电站利用低富集度铀核裂变释放大量热能产生的蒸汽推动汽轮发电机组发电。核燃料在核裂变过程中除了释放巨大的能量以外，还伴随着大量放射性物质的生成，为阻止大量放射性裂变产物释放到周围环境中，核电站采用了纵深防御设计概念，并贯彻予核电站安全有关的全部活动过程。     

高性能核安全级控制显示操作系统软件设计技术研究

核安全级控制显示操作设备(Nuclear safety classified control and information display,SCID)是核电站数字化仪控系统的关键设备,实现运行操作人员与数字化核安全级控制保护系统人机交互功能.本文对我国SCID的需求进行了分析,得出研制高性能SCID的必要性;提出了基...     

核安全级水泵鉴定试验系统模块化设计研究

作为压水堆核电机组中的水力能动设备,泵类产品的可靠性对整个核电站的安全起着及其关键的作用。核一级、二级泵的研发中,在样机鉴定阶段要进行各种苛刻的鉴定试验,以保证设计符合系统要求。然而目前国内在关于核级泵试验方面的研究较少,本文针对目前核电站采用的核二级泵鉴定试验要求,采用模块化设计的思想,对满足其要求的试验系统设计进行了研究,就试验台架的设计思路及问题进行了讨论,并提出合理的综合试验系统设计方案。     

基于质量特性的数字化核安全级仪控平台系统测试需求分析方法研究

在数字化核安全级仪控平台系统测试中,测试需求分析是明确测试特征的重要活动。基于核安全级仪控平台的特点,在系统测试需求分析过程中引入质量特性的概念,提出了适用于核安全级仪控平台的质量特性分类,并建立了基于质量特性及度量的平台系统测试需求分析方法。通过在某核安全级仪控产品平台中的应用,对方法的有效性进行了说明和验证,表明所提出的测试需求分析方法能较为全面地识别关键测试特征,有效提高平台系统测试分析的完整性。     

关于核级设备的静动态抗震试验的思考

核级设备必须通过抗震鉴定,抗震试验是能动设备的主要鉴定方法.美国ASME QME-1-2002提出可采用静力法.本工作论述静力法的使用要点、前提条件和使用限制,并提出静力法适用范围的建议.     

基于退化失效模型的旋转机械寿命预测方法

退化失效模型与传统可靠性预测的根本区别在于,不论在统计推断还是寿命分布拟合过程中,可以充分利用退化数据提供的更多过程和寿命信息,能较准确地进行具有耗损特性的机械产品的寿命预测.针对旋转机械运行过程中强度破损失效模式,本文利用正态随机过程模型描述其退化失效过程,进行了旋转机械的寿命预测方法研究.通过分析加速寿命方程与退化失效模型的关系,考虑到加速寿命试验方法以"应力换时间"的有效性,进行了旋转机械加速寿命试验.通过对试验结果进行最佳线性无偏估计,得到强度退化失效模型的退化轨迹;在解决了退化失效方程奇异性的基础上,进行了旋转机械的寿命预测,得到点估计与区间估计的可靠寿命预测结果.     

进口民用核安全设备开箱检查的实施及若干问题的探讨

介绍了进口民用核安全设备安全检验的第二阶段—开箱检查的目的、内容、流程及要求,明确了开箱申报材料的审查重点,介绍了安全检验机构在核设施现场外实施开箱检查的监督检查(以下简称开箱见证)步骤,探讨了对开箱检查工作、监造、装运前检验、监装及验收的理解以及出具开箱见证报告等典型问题,并提出了一些加强进口核安全设备开箱检查工作的建议。     

核电厂安全级设备鉴定及技术发展的探讨

核电厂安全级设备鉴定是保障电厂安全功能完整的一项重要措施,也是核安全文化的重要体现。但是,目前设备鉴定存在着诸多理解偏差或缺失。本文结合核安全设计理念,阐述了设备鉴定的基本流程和方法以及具体的实施原则,指出了设备鉴定技术的改进和发展方向,从而为建立一个整体的设备鉴定体系提供参考和依据。     

核电站仪控系统数字化开发仿真测试技术研究

在核电站应用数字化仪表与控制 (I&C)取代模拟 I&C 系统,已成为必然的发展趋势。本文分析了核电站全范围模拟机的蒸汽发生器数学模型,研制开发独立的核电站蒸汽发生器实时仿真系统,并与控制系统形成能够相互作用的闭环系统,用于数字化仪控系统改造提供仿真对象及进一步控制方案研究。在仿真过程中,除了仿真模型之外,其他的硬件和软件由真实的控制系统构成。不但为核电站仪表与控制 (I&C)系统数字化开发提供理论分析,也为今后现场调试工作创造有利条件。     

压水堆核电厂仪表控制与计算机化的发展概况

阐述了当今世界压水堆核电站（ＰＷＲ）仪表与控制及自动化设备的发展概况;描述了模拟仪表与控制所存在的缺陷和问题;重点论述了当代先进核电站数字化仪表控制、保护系统与先进主控制室的性能和对计算机化仪表控制提出和要求。     

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.     

核动力装置运行过程可靠性研究现状与发展

核动力装置运行过程的可靠性研究是保障装置安全，提高装置效能的重要基础，本文从运行过程可靠性问题的背景出发，研究了可靠性分析方法的适应性，并论述了问题研究的技术途径。     

Recommendations on legislative and regulatory framework and regulatory body of nuclear security in China

This paper describes the definition of nuclear security that has been changing from the cold war age to the post-911 period, and clarifies the close relationship and yet a clear distinction between nuclear security, nuclear safety and nuclear safeguard. Based on analyses of the current state of nuclear security activities in China as well as the requirements and the law infrastructure, a legislative and regulatory framework of nuclear security and the mandate of a regulatory body in China are recommended.     

Developing architecture for upgrading I&C systems of an operating nuclear power plant using a quality attribute-driven design method

This paper presents the architecture for upgrading the instrumentation and control (I&C) systems of a Korean standard nuclear power plant (KSNP) as an operating nuclear power plant. This paper uses the analysis results of KSNP's I&C systems performed in a previous study. This paper proposes a Preparation–Decision–Design–Assessment (PDDA) process that focuses on quality oriented development, as a cyclical process to develop the architecture. The PDDA was motivated from the practice of architecture-based development used in software engineering fields. In the preparation step of the PDDA, the architecture of digital-based I&C systems was setup for an architectural goal. Single failure criterion and determinism were setup for architectural drivers. In the decision step, defense-in-depth, diversity, redundancy, and independence were determined as architectural tactics to satisfy the single failure criterion, and sequential execution was determined as a tactic to satisfy the determinism. After determining the tactics, the primitive digital-based I&C architecture was determined. In the design step, 17 systems were selected from the KSNP's I&C systems for the upgrade and functionally grouped based on the primitive architecture. The overall architecture was developed to show the deployment of the systems. The detailed architecture of the safety systems was developed by applying a 2-out-of-3 voting logic, and the detailed architecture of the non-safety systems was developed by hot-standby redundancy. While developing the detailed architecture, three ways of signal transmission were determined with proper rationales: hardwire, datalink, and network. In the assessment step, the required network performance, considering the worst-case of data transmission was calculated: the datalink was required by 120 kbps, the safety network by 5 Mbps, and the non-safety network by 60 Mbps. The architecture covered 17 systems out of 22 KSNP's I&C systems. The architecture is implementable with the equipment developed in South Korea. The architecture can be used as a model to upgrade the existing I&C systems in a planned, large-scale, and one-shot manner. A more detailed architecture down to software level will be developed in the future.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.     

Advanced qualification process of ANAV NPP integral dynamic models for supporting plant operation and control

Extensive technical literature exists aimed at establishing the requirements needed to qualify a Nuclear Power Plant model. Most of this literature is focused on qualifying a model for licensing uses. Less documentation is available nowadays on the requirements needed when an Integral Plant Model is used for supporting plant operation and control of an actual commercial facility, while fulfilling its goals of safety and competitiveness. For the last 15 years the Technical University of Catalonia (UPC) has been working in this field along with Asociación Nuclear Ascó–Vandellòs (ANAV), which is a utility that presently runs three operating PWRs. The paper develops an advanced qualification process (AQP) of plant models for operation support, introduces the concept of plant configuration and explains how this activity complements other usual validation tasks.