核电厂异常重要性判定(SDP)方法及其应用

在核电厂的日常安全管理过程中,核安全管理人员会遇到大量的安全事项,正确、快速和有效地处理这些事件和异常是保证核电厂安全运行的关键。目前核电厂和核安全监管机构都应用分级分类管理的方式来处理这些核安全相关事项,这样做可以使得安全重要度高的事项能够得到足够的关注,保证核电厂的总体安全水平。这种分级分类管理方式的重要的一环是能够正确地确定安全事项的安全重要程度。随着以概率安全分析(PSA)为代表的风险指引型安全管理方法的广泛应用,核安全管理人员可以利用风险重要程度来确定安全相关事项的重要程度。本文主要讲述了目前广泛使用的核电厂异常重要性判定方法(SDP)在开发及核安全管理中的应用,以及其对未来我国核安全管理带来的影响。     

核电厂火灾异常重要性判定方法简介及适用性分析

对火灾的预防是核电厂安全工作的重要一环,但近年来,国内核电厂与火灾相关的事件和异常时有发生。本文主要介绍了美国核管会(NRC)率先开发使用的核电厂火灾异常重要性判定(火灾SDP)方法,并对该方法在我国核安全监管工作中的适用性进行了分析。     

核电厂RCV主泵断轴事件的重要性判定

本文介绍了美国核管会的异常重要性判定(SDP)方法及实施过程,并以此为基础开发了功率工况SDP方法。应用功率工况SDP方法对国内某核电厂发生的化学和容积控制系统(RCV)主泵断轴事件进行了分析,证明了SDP方法的应用对于提高我国的核安全监管效率和核电厂的安全管理水平具有重要的意义。     

功率工况异常重要性判定方法在核安全监管中的应用

介绍了国家核安全局开发的功率工况下核电厂异常重要性判定方法（SDP）的基本原理及方法,并使用该SDP对国内某核电厂发生的汽动辅助给水泵（ASG004PO）不可用事件进行了重要度和敏感性分析,结果表明该汽动辅助给水泵的再循环流量试验周期偏长。本文针对此问题给出了优化建议是将ASG004PO再循环流量试验的周期优化为小于34 d。      

国内核电厂安全壳地坑滤网设计改进考虑

在冷却剂丧失事故（LOCA）工况下核电厂安全壳地坑滤网堵塞问题已是核能界广泛关注的核安全问题,国内核安全监管部门和核电厂营运单位正积极推动该问题的解决。本文介绍了国内核电厂安全壳地坑滤网设计改进工作的进展情况,从审评人员的角度说明了对解决该问题所持的态度及相应的监管要求,并阐述对国内相关核电厂逐步开展该项工作的总体设想。     

标准监管PSA模型开发与应用

本文介绍了NRC标准化电厂风险模型(SPAR)的发展历程以及在风险指引型安全监管活动中的应用情况和发展趋势。基于核电厂性能和风险指引型的安全监管理念,开发国内核电厂的标准监管PSA模型,本文重点介绍了模型开发的技术路线、分析范围、开发周期与计划,并对该模型拟在风险指引型的监管活动中的应用前景。标准监管PSA模型是作为核安全监管当局作为独立评价和验算核电厂PSA应用的重要工具之一,是保证国家核安全局对核电厂的安全监管活动的独立性和技术权威性的重要手段。     

风险指引型核动力厂建造异常重要性判定方法的应用研究

文章介绍了一种风险指引型的核动力厂建造异常重要性判定方法（cSDP）的基本原理及应用流程，该方法建立了建造质量和反应堆风险之间的联系，能够对核动力厂建造阶段发生的异常进行重要性的量化评价。采用该方法对AP1000核电厂建造阶段发生的非能动堆芯冷却系统调试试验失败和核岛基础建造缺陷两起案例进行了实际应用，结果表明，该方法能够快速、准确地筛选出重要的建造异常，有利于实现监管资源的合理分配，提高核动力厂建造监管水平。     

核电厂应急运行规程与严重事故管理指南接口分析

福岛核事故后,核工业界及核安全监管当局对严重事故更加重视,严重事故管理指南（SAMG）的制订已经成为国内核安全监管要求.核电厂制定了应急运行规程（EOP）用以防止核电厂事故升级为严重事故,在SAMG研制时,如何从EOP合理地过渡到SAMG成为必须解决的问题.本文详细分析了EOP与SAMG的接口准则和影响因素,并结合国内核电厂SAMG研制现状,对EOP与SAMG接口方案进行了分析和建议,可为其他核电厂SAMG的研制工作提供参考.     

应用于核电厂老化PSA的SSC筛选方法研究

核电厂运行许可证延续必须考虑其延寿期内的核安全问题,确保核电机组在延期运行期间的核安全水平不低于原设计寿期内的核安全水平。可应用PSA技术对许可证延续期间的核电厂建立老化PSA模型,从而评估SSC老化对核电厂整体安全的影响,验证其仍可满足原设计标准。基于此提出了应用于核电厂老化PSA的SSC筛选分析方法,通过考虑趋势分析,老化失效模式与影响分析,风险重要度分析,在三种分析方法基础上建立核电厂SSC筛选的决策矩阵,为选择易老化且安全重要的部件建立了可行的方法。该项工作也为核电厂在许可证延续阶段的风险指引型管理奠定技术基础。     

基于FPGA的核安全级仪控系统设计与验证

为解决基于微处理器技术的核电厂安全级数字化仪控系统（DCS）中软件共因故障（CCF）的问题,通过多样性手段避免当未能紧急停堆的预计瞬态（ATWS）发生或反应堆保护系统（RPS）因CCF导致丧失安全功能的风险,本文设计了一种基于现场可编程逻辑门阵列（FPGA）技术的核安全级DCS系统平台,并以核电厂中RPS为实例测试验证平台的功能性能。结果表明:基于FPGA的核安全级DCS系统平台在可用性、适用性和可靠性等方面都满足核电厂安全级数字化仪控系统的要求。      

多机组核电厂总体风险的一级PSA方法研究

福岛核事故发生后,多机组核电厂的总体风险受到越来越多的关注,但国内外缺乏评价多机组核电厂总体风险的方法或导则。本文结合有关法规对核电厂的总体安全要求,探索将单机组的一级概率安全评价（PSA）方法拓展为多机组的风险评价方法。以双机组核电厂为例,讨论了多机组厂址PSA定量化的一些问题,提出了机组间相关性的一些见解,并阐明了数学原理。本文讨论的方法对研究多机组厂址PSA方法具有重要价值。     

Safety assessment of nuclear power plant under volcanic phenomena – Background and technical basis of the revision of JEAG4625

Japan is one of the countries with abundant active volcanoes and has a long history of developing countermeasures to mitigate volcanic disasters. In the field of nuclear energy, it is also necessary to assess safety against volcanic hazards, and in 2009, a voluntary guideline was published as JEAG4625 in order to set up requirements of site assessments and basic designs of nuclear power plants (NPPs). This guideline has been revised to satisfy the requirements for examining the necessity of considering volcanic phenomena and concrete countermeasures in detailed designs of NPPs. This paper focuses on the background and technical basis of this voluntary guideline and shows the basic policy to ensure safety of NPPs and the requirements to prevent nuclear hazards due to volcanic phenomena based on the defense in depth concept.     

Development of a standard communication protocol for an emergency situation management in nuclear power plants

Correct communication between main control room (MCR) operators is an important factor in the management of emergency situations in nuclear power plants (NPPs). For this reason, a standard communication protocol for the management of emergency situations in NPPs has been developed, with the basic direction of enhancing the safety of NPPs and the standardization of communication protocols. To validate the newly developed standard communication protocol, validation experiments with 10 licensed NPP MCR operator teams was performed. From the validation experiments, it was found that the use of the standard communication protocol required more time, but it can contribute to the enhancement of the safety of NPPs by an operators’ better grasp of the safety-related parameters and a more efficient and clearer communication between NPP operators, while imposing little additional workloads on the NPP MCR operators. The standard communication protocol is expected to be used to train existing NPP MCR operators without much aversion, as well as new operators.     

Safeguard assessment for life extension in nuclear power plants (NPPs) using a production function

Life extension is investigated as a safeguard assessment for the stability on the operation of the nuclear power plants (NPPs). The Cobb-Douglas function, one of the production functions, is modified for the nuclear safeguard in NPPs, which was developed for the life quality of the social and natural objects. Nuclear Safeguard Estimator Function (NSEF) is developed for the application in NPPs. The cases of NPPs are compared with each other in the aspect of the secure performance. The results are obtained by the standard productivity comparisons with the designed power operations. The range of secure life extension is between 1.008 and 5.353 in 2000 MW_e and the range is between 0.302 and 0.994 in 600 MW_e. So, the successfulness of the power operation increases about 5 times higher than that of the interested power in this study, which means that the safeguard assessment has been performed in the life extension of the NPPs. The technology assessment (TA) is suggested for the safe operation which is an advanced method comparing conventional probabilistic safety assessment (PSA).     

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.     

Seismic protection technology for nuclear power plants: a systematic review

Seismic protection systems (SPS) have been developed and used successfully in conventional structures, but their applications in nuclear power plants (NPPs) are scarce. However, valuable research has been conducted worldwide to include SPS in nuclear engineering design. This study aims to provide a state-of-the-art review of SPS in nuclear engineering and to answer four significant research questions: (1) why are SPS not adopted in the nuclear industry and what issues have prevented their deployment? (2) what types of SPS are being considered in nuclear engineering research? (3) what are the strategies for location of SPS within NPPs? and (4) how may SPS provide improved structural performance and safety of NPPs under seismic actions? This review is conducted following the procedures of systematic reviews, where possible.

The issues concerning the use of SPS in NPPs are identified: cost, safety, licensing and scarcity of applications. NPPs demand full structural integrity and reactor's safe shutdown during earthquake actions. Therefore, horizontal isolation may be insufficient in active seismic zones and isolation in the vertical direction may be required. Based on the results in this review, it is likely that next generation reactors in seismic zones will include state-of-the-art SPS to achieve full standardised design.     

Transient identification in nuclear power plants: A review

A transient is defined as an event when a plant proceeds from a normal state to an abnormal state. In nuclear power plants (NPPs), recognizing the types of transients during early stages, for taking appropriate actions, is critical. Furthermore, classification of a novel transient as “don't know”, if it is not included within NPPs collected knowledge, is necessary. To fulfill these requirements, transient identification techniques as a method to recognize and to classify abnormal conditions are extensively used. The studies revealed that model-based methods are not suitable candidates for transient identification in NPPs. Hitherto, data-driven methods, especially artificial neural networks (ANN), and other soft computing techniques such as fuzzy logic, genetic algorithm (GA), particle swarm optimization (PSO), quantum evolutionary algorithm (QEA), expert systems are mostly investigated. Furthermore, other methods such as hidden Markov model (HMM), and support vector machines (SVM) are considered for transient identification in NPPs. By these modern techniques, NPPs safety, due to accidents recognition by symptoms rather than events, is improved. Transient identification is expected to become increasingly important as the next generation reactors being designed to operate for extended fuel cycles with less operators' oversight. In this paper, recent studies related to the advanced techniques for transient identification in NPPs are presented and their differences are illustrated.     

Risk managed technical specifications

Application of probabilistic risk assessment (PRA) technology has become an essential component in the decision-making processes associated with the operation and regulation of commercial nuclear power plants (NPPs). As PRA technology has matured, it increasingly has been utilized to provide risk insights in the support of both operational and regulatory decision-making. This paper describes the next significant application of PRA technology to risk inform NPP operation. This Risk Managed Technical Specification (RMTS) application utilizes the results of the plant PRA to determine risk-informed technical specification (TS) allowed out of service times (AOTs). The RMTS process utilizes the PRA results to specify appropriate configuration specific TS AOTs and ensures the risk of events that could result in core damage or large early release are maintained below acceptable levels. In addition, RMTS requires development of integrated risk management actions to actively mitigate risks associated with the inoperability of TS structures, systems and components (SSCs). RMTS has been approved for implementation at commercial NPPs in the United States with the South Texas Project Electric Generating Station (STPEGS) serving as the initial application. In this paper we describe the programmatic requirements necessary to implement RMTS and provide several examples illustrating its application; thus demonstrating the applicability of RMTS to manage nuclear safety risk while simultaneously enhancing operational flexibility.     

Development of CFD fire models for deterministic analyses of the cable issues in the nuclear power plant

Additional fire barriers of electrical cables are required for the nuclear power plants (NPPs) in Taiwan due to the separation requirements of Appendix R to 10 CFR Part 50. The risk-informed fire analysis (RIFA) may provide a viable method to resolve these fire barrier issues. However, it is necessary to perform the fire scenario analyses so that RIFA can quantitatively determine the risk related to the fire barrier wrap. The CFD fire models are then proposed in this paper to help the RIFA in resolving these issues. Three typical fire scenarios are selected to assess the present CFD models. Compared with the experimental data and other model’s simulations, the present calculated results show reasonable agreements, rendering that present CFD fire models can provide the quantitative information for RIFA analyses to release the cable wrap requirements for NPPs.