Safety functions and component classification for the HTR-10

In the design and construction of the HTR-10, the standards and criteria of design and manufacture for structures, systems and components must be defined. This paper refers to the relative nuclear safety codes to formulate the principles of safety classification and the relative requirements of design and manufacture, according to the safety philosophy and feature of the HTR-10, and the requirements for safety functions of structures, systems and components. We can find practical use and application meaning of this work in the design and manufacture of the HTR-10. It will be used to ensure the safety and reliability of the HTR-10.     

HTR-10安全功能与设备分级

参照有关的核安全法规,结合我国设计、建造各类研究堆的经验,根据HTR-10的安全特性和对构筑物、系统和部件安全功能的要求,制定了HTR-10的构筑物、系统和部件等物项的安全分级原则和相应的设计、制造要求及验证措施等,对HTR-10的设计和建造具有实际的指导和应用价值,确保了HTR-10的安全与可靠。     

基于风险指引安全分级的维修规则实施方案

近年来,美国核电厂的业绩始终保持世界领先水平,维修规则的实施起了很大的作用.本文研究了美国核电厂实施维修规则的法规要求以及实施方法,结合我国正在研究中的风险指引安全分级及其处理方法,提出了适用于我国的核电厂维修规则实施方案.     

集散控制系统在核电站保护系统中的应用

在核电站保护系统中应用基于数字化技术的安全级DCS已替代模拟技术成为主流,因为核电站的安全性要求很高,所以DCS技术的可靠性至关重要。为了确保应用DCS技术设计的可靠性,通过分析相关法规、标准的要求,总结了安全级DCS设计须进行的质量鉴定项及软件验证和确认过程,并结合当前主要应用的几个安全级DCS产品的技术特点,提出了安全级DCS设计开发中应考虑的控制器、智能IO模块、优选驱动模块以及通信等关键技术的要点。     

核动力厂物项安全分级中的剂量准则研究

核动力厂的设计中通过对物项进行安全分级，来确保物项的设计、制造、建造等满足适当的要求，达到与其执行的功能相符的可靠性。本文简要介绍了根据安全重要性对核动力厂物项进行安全分级的方法以及应考虑的因素。针对物项安全分级中应考虑的未能执行某一安全功能的后果，使用核动力厂不同工况下对公众和工作人员的剂量准则来划分“高、中、低”后果。通过研究提出放射性“高、中、低”后果定量化的建议，以使得该方法在用于核动力厂物项安全分级时更具有可操作性。     

核电站安全分级对DCS系统设计影响分析

根据核电站设计总体要求,特别是对仪控系统可用性和可靠性的要求,通过分析核电站中系统、设备及其功能的安全分级,解析现代数字仪化控系统( DCS)的技术特点.结合实际在建核电站中不同DCS总体技术方案设计实施过程中的差异,从满足核电站安全运行以及安全评审相关法规标准的需求出发,阐明核电站中不同安全分级的系统和设备对DCS总...     

Design procedure for the primary circuit of the Konvoi plants

The loading data for the design and fatigue analyses of the primary system components are derived from the plant operating transients as specified in the Operational Manual, the accident analyses and from the external event requirements. The methods used, applicable codes and standards, and QA procedures produce a design concept that contains significant safety margins. Measurements conducted in pre-Konvoi plants and during commissioning of the Konvoi plants deliver additional verifications of the analyses.     

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

浅谈我国核电厂重要厂用水系统换热器隔离阀门与放射性监测仪表配置优化

对于运行核电厂来说,重要厂用水系统与质量和安全密切相关.核电厂重要厂用水系统用于导出设备冷却水系统所传输的热量,将其输送到海水中,因此是核岛的最终热阱.本文描述了当前我国大部分核电厂重要厂用水系统换热器隔离阀门与放射性监测仪表的配置现状,分析了包括美国、法国以及国际原子能机构对于重要厂用水系统设计要求的相同点与不同点,...     

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。      

Seismic analysis and design of electrical cable trays and support systems

Most cable trays in nuclear power plants are classified as seismic category I components. Current safety requirements dictate that all such components be adequately designed in order to remain functional during and after the most severe possible earthquake, so that a safe and orderly plant shut-down can be ensured. The design aspects of electrical cable trays and support systems are discussed from the seismic and structural standpoint. The effects of the inherent flexibility of commonly used cable trays is considered. A procedure for the selection of trays and the design of their support structure is recommended. The dynamic characteristics of typical trays are determined analytically and also from test results. The advantages of a rigid support system are discussed. Two procedures are presented for the design of the support systems, namely, static loading and spectral analysis. The interaction between trays and supports is discussed.     

Development of aseismic design of piping, vessels and equipment in nuclear facilities

This report involves the development of aseismic design procedures of piping, vessels and equipment in Japan. These mechanical structures show their various characteristics of vibration. Pressure boundaries, a containment vessel and safety systems belong to such structures. The vital components of nuclear power plants are classified to “A” class according to the classification for the aseismic design in Japan. All components in “A” class are required to be based on dynamic earthquake-resistant design, of which level is decided in consideration of local seismisity.

For dynamic design purposes, the following processes are the most important: 1. estimating eigenfrequencies and modes of the system; 2. estimating its damping characteristics; 3. estimating the behavior of the system during strong earthquakes; 4. deciding the design criteria, especially the allowable stresses to earthquake loadings.     

福清核电厂1、2号机组抗震裕量评价

地震是核电厂主要外部灾害之一,地震风险评估对于核电厂的安全评价具有重要的价值。抗震裕量评价(SMA)是开展核电厂地震灾害风险分析的重要方法之一,其目的是为了判断核电厂的抗震设计能力相对于设计基准地震的抗震裕量,找出核电厂的抗震薄弱环节,提高核电厂的抗震能力。本文针对福建福清核电厂1、2号机组进行抗震裕量评价,分析表明电力支持系统和一回路辅助管道的抗震能力相对薄弱,是导致核电厂抗震能力薄弱的主要原因,电力支持系统和一回路辅助管道需进一步提高其抗震能力,且核电厂需考虑编制地震应急规程。     

基于NuPAC的核电厂反应堆保护系统关键特性分析

为确保核电厂反应堆保护系统满足核安全要求和用户需要,对基于NuPAC平台的反应堆保护系统的关键特性进行了分析。归纳出基于NuPAC平台的反应堆保护系统的14个关键特性,这些关键特性不仅覆盖了法规和标准的重要的安全要求,如单一故障准则、独立性、完整性、质量、多样性、可靠性、安保性、可操作性、可维护性及系统性能等,而且覆盖了重要的用户需求,如可兼容性、设计裕量、可持续性、灵活性和经济性等。分析得到的关键特性为下一步反应堆保护系统的需求分析提供了良好的基础和指导。     

Introduction to the proceedings of the international seminar on structural reliability of mechanical components and subassemblies of Nuclear Power Plants

Although the integrity and safety of many mechanical components and subassemblies of nuclear power plants are demonstrated by the appropriate design codes and supplementary requirements, such procedures seldom provide guidance as to “how safe” the structures are. By combining the technologies of solid mechanics and probabilistic structural reliability methods, engineers are finding many and varied opportunities to demonstrate margins in terms of probabilities of failure.With reference to the large mechanical system components typical of nuclear power plants, reliability assessments are receiving more emphasis in recent years as evidenced by the increased attention to such reliability-related techniques as failure mode and effects analyses, fault tree analyses, common cause failure analyses, and single point failure analyses. These techniques are ordinarily applied at the outset in a qualitative manner, tracing the casual sequences of potential component failure. The results of these analyses serve as the foundation for more sophisticated probabilistic structural reliability analyses which have the objective of calculating the probability of failure (that is, unreliability) of the system or component in question.     

国产先进压水堆核电厂通用调试导则的设计方法

通用调试导则作为核电厂调试的基础性技术指导文件之一,其作用是针对核电厂中同类型设备、部件或某种给定类型试验给出通用试验方法。基于国内外核电厂调试相关法规和标准的要求,分析国产先进压水堆核电厂的设计特点和调试工作的实际需求,制定了一种国产先进压水堆核电厂通用调试导则文件的设计方法。通过核电厂主要设备和功能梳理、导则试验项目筛选、标准化分析等关键步骤,同时结合国产二代压水堆核电厂调试经验,设计了一套具有自主知识产权、标准化和规范化的国产先进压水堆核电厂通用调试导则文件体系,并在此基础上确定了一种新的文件分类和编码形式,降低文件被错误使用和引用的风险,一定程度上减轻了调试人员的工作负担,同时满足文件的使用、管理和归档要求。     

浮动核电站设计中的安全准则研究

浮动核电站作为船海工程与核电工程的结合,属于核能工程的新领域,国内尚缺少相应的安全设计准则。结合海洋核动力平台示范工程实际设计需求,基于对陆上压水堆核电厂、海上移动式平台、核动力舰船规范的分析,从浮动核电站总体设计、平台设计以及核安全3个层面分别提出了相应的安全设计准则。研究表明,浮动核电站的安全设计应围绕3项基本安全功能进行;平台设计应考虑布置、结构、辅助系统、电力、通信、消防6个因素;核安全设计应充分考虑其孤岛运行和海洋应用场景对核动力装置系统设备设计、运行的制约影响。      

基于分离变量的地震PSA相关性分析方法研究

为解决现有地震概率安全评价（PSA）相关性分析简化假设存在的问题,建立更准确反映核电厂构筑物、系统和部件（SSC）地震相关性的分析方法,对基于分离变量的易损度相关性分析开展了研究。结合易损度模型对分析方法进行了理论推导,并对方法的实施过程进行了介绍。利用该方法对不同条件下SSC的联合失效开展案例分析,得到了联合失效的易损度曲线和失效频率分析结果,并与现有相关性简化假设得到的结果进行了对比。研究结果表明,基于分离变量的地震PSA相关性分析方法能弥补现有方法的不足,支持核电厂地震PSA开发和应用。     

国际原子能机构SSG-30安全分级方法应用解析

安全分级的目的是确保物项的设计、制造、建造、调试和运行采用恰当的要求,使物项在所有预期的运行工况下有适宜的质量,进而确保安全功能的实现。国际原子能机构（IAEA）2014年颁布的核电厂构筑物、系统和部件（SSC）安全分级导则（SSG-30）,其安全分级原则涵盖核电厂5个纵深防御层次,从设计预防措施和安全功能分类两个维度识别安全相关物项的重要性,考虑核电厂运行工况状态和放射性与运行限值的要求,进而确定物项的安全级别和相关的规范要求。      

Method for Fault Diagnosis of Digital Control Systems in Nuclear Power Plant

This paper presents a method for localizing faulty components of control systems by replaceable parts such as print boards and cables, in a large scale plant like a nuclear power plant. Most of today's control systems form a distributed configuration including many digital controllers interconnected by data communication networks. Usually, to localize the faulty components in nuclear plant control systems, suspected faulty components are narrowed down by executing manual tests to examine whether the objects are normal or abnormal based on design documents and personnel know-how, besides the use of self-diagnosis functions built into the control systems. In the present method, procedures of various tests including the know-how and checking of self-diagnosis functions are provided as knowledge of tests. The test to be executed is determined by considering failure probabilities of objects, and easiness and effectiveness of testing. Then, the suspects are narrowed down sequentially based on the test result. In checking feasibility of this diagnosis method for a simulated control system, intended faults are satisfactorily localized. This method is confirmed to be practicable for diagnosis of large scale digital control systems.