火灾PSA人员可靠性分析中的Scoping方法研究

核电厂火灾情景下的人员可靠性分析(HRA)是核电厂火灾概率安全评价(PSA)中的重要内容。本文介绍了一种新的火灾HRA定量筛选分析方法 Scoping方法,并将其应用于某非能动核电厂火灾PSA建模中,将所得出的结果与运用NUREG/CR-6850筛选法得出的结果进行了比较。结果表明,Scoping方法一般具有更少的保守性,能合理地模化火灾情景下的人员失误,具有较好的工程应用价值。     

核电厂传统人员可靠性分析方法中引入班组因素的研究

在核电厂等大型复杂系统中,人员干预行为通常以班组的协作来完成,而目前核电厂概率安全评价（PSA）采用的以人的失误率预测技术（THERP）和人的认知可靠性（HCR）方法为代表的人员可靠性分析（HRA）方法主要关注对个人绩效的影响,它们在评估核电厂主控室班组绩效时存在一定局限。本文定义一种新的绩效形成因子“班组绩效形成因子（TPSF）”,并将其合理地引入THERP和HCR方法的定量化体系中,使它们可在一定程度上体现班组环境对人员绩效的影响。文章提出了TPSF等级的评价方法及将其引入THERP和HCR方法的定性实施框架。结果证明,合理地将班组因素引入传统HRA方法能改进它们对班组环境下人员绩效模化的合理性。     

二级PSA中人员可靠性分析方法研究

在概率安全分析(PSA)中,人员可靠性分析(HRA)是必不可少的组成部分。国内在一级PSA中的HRA做了大量的研究工作,已有良好的基础和工程实践,但由于核电厂严重事故下人员响应的复杂性,有关二级PSA的HRA还处于摸索阶段。通过研究二级PSA中人员响应特点,调研国内外在二级PSA中采用的HRA方法,最后以我国某三代压水堆核电厂严重事故下一回路快速卸压为例,采用THERP、HCR+THERP以及SPAR-H三种方法,分别进行了HRA,并给出相应的结论和建议。     

核电厂火灾概率安全评价中详细火灾情景分析方法探讨

火灾概率安全评价(PSA)是评估核电厂风险并发现薄弱环节的有效工具,详细的火灾情景分析是其中一个重要环节。在火灾概率安全评价的火灾隔间定量筛选的过程中,火灾隔间的分析通常较为保守,为使结果更贴近核电厂实际,有必要对风险重要的火灾隔间进行详细的火灾情景分析。通过确定特定的火灾情景,分析火灾的发展蔓延并评估火灾情景的发生频率,从而为最终的火灾风险定量化提供基础。本文探讨了详细火灾情景分析在火灾概率安全评价中的应用,并以单一火灾隔间为例阐述分析方法,为核电厂火灾概率安全评价工作提供支持和参考。     

HTR-10常规岛热力系统中特殊子系统的运行可靠性分析

在利用事件树和故障树相结合的方法对核电站进行概率安全评价（PSA）分析中,其备用系统和可修复系统的运行可靠性分析是一个难点。本文利用故障树分析的技术和可靠性工程数学的分析方法,对10MW高温气冷堆（HTR－10）常规中的上述子系统的运行可靠性进行了深入的分析和研究,可供PSA分析人员参考。     

核电厂地震PSA中HRA方法研究

在调研外部事件人员可靠性分析（HRA）的发展历史及方法缺陷的基础上，研究了当前国内外外部事件HRA技术标准，并以某核电厂地震概率安全评价（PSA）中的人因事件为研究对象开展方法研究，说明了地震场景HRA与内部事件HRA的不同点，明确地震人员响应的流程，阐述地震场景下HRA方法，使其得到的分析结果更加接近外部事件的实际情况，并可为外部事件PSA的开发提供技术依据。      

秦山第三核电厂人因可靠性分析

人因可靠性分析（HRA）是概率安全评价（PSA）的重要组成部分。秦山第三核电厂（简称秦山三核）初版HRA由加拿大原子能公司（AECL）完成,其采用的HRA方法为简化的ASEP HRA。为获得更符合秦山三核运行状态实际的HRA结论,本工作对秦山三核重新进行了HRA分析,并增加了事件间的相关性分析。在对国际HRA方法比较研究的基础上,秦山三核HRA采用了规范化的THERP+HCR分析方法。新分析所得数据与AECL数据比较分析结果表明,新分析与AECL的分析判断基本一致,但在合理性和准确性方面较原分析有明显提高,分析结论更符合秦山三核实际。     

火灾人员可靠性分析中的定性分析

福岛事故后,国家安全监管部门对核电厂火灾概率安全分析中人员可靠性分析提出了新的要求。火灾情景下,合理评估人误概率,并根据评估结果对电厂火灾后的管理和响应提出合理化建议,对电厂安全具有重要意义。NUREG-1921导则是专门的和最新的火灾HRA导则,首次明确提出定性分析在整个火灾人员可靠性分析活动中的重要性。基于导则的学习、消化和吸收,并结合实际工作经验,本文首先阐述了火灾人员可靠性分析的基本框架,然后分别从信息收集、操作的可行性评估、绩效形成因子等三个方面阐述了火灾人员可靠性分析中定性分析的主要内容及特征,并通过了一个工程实例阐述了如何开展定性分析,以期更好指导其在实际工程项目中应用。     

概率安全评价在核电厂设计和运行期间的技术特征及应用

概率安全评价（以下简称PSA）作为核电工程项目中的一项风险评估方法和应用技术,其模型开发的详细程度随着核电工程项目的设计认证、建造调试和装料运行依次提升,对电站的安全性、可靠性以及经济性方面作出巨大贡献。为了更好地完成从设计阶段到运行阶段PSA模型的升级完善,更好地发挥PSA在核电厂整个寿期内的作用,通过分析梳理PSA技术在核电厂设计和运行阶段所支持的应用,凝练PSA的技术特征并总结模型完善的要点,促进PSA模型转化更加平顺,对更好地支持PSA技术在核电厂中的应用有重要意义。此外,本文根据分析对PSA建模提出了建议,对未来PSA模型动态化、模块化趋势进行了展望。     

IDHEAS方法应用研究

人员可靠性分析（HRA）作为核电厂概率安全评价（PSA）中的重要组成要素,一直是影响PSA分析质量和风险见解的关键内容。目前业界中已有的HRA方法众多,不同的HRA方法各有优缺点且存在基础数据过老的问题,为此,美国核管理委员会联合HRA领域权威专家开发了一种综合性的HRA方法--人员失误事件综合分析系统,简称IDHEAS方法。本文对IDHEAS方法进行了系统性的研究,对相关实施流程和要点进行归纳,并运用IDHEAS方法进行了实例分析。理论研究和实例分析表明,IDHEAS方法在工程应用上具备可操作性,能较好弥补其他HRA方法的局限性。同时,IDHEAS方法亦存在对时间参数不敏感、部分分析内容依赖于分析人员经验等特点。     

Application of thermal comfort theory in probabilistic safety assessment of a nuclear power plant

Human factor errors in probabilistic safety assessment(PSA) of a nuclear power plant(NPP) can be prevented using thermal comfort analysis.In this paper,the THERP+HCR model is modified by using PMV (Predicted Mean Vote) and PPD(Predicted Percentage Dissatisfied) index system,so as to obtain the operator cognitive reliability,and to reflect and analyze human perception,thermal comfort status,and cognitive ability in a specific NPP environment.The mechanism of human factors in the PSA is analyzed by operators of skill,rule and knowledge types.The THERP+HCR model modified by thermal comfort theory can reflect the conditions in actual environment,and optimize reliability analysis of human factors.Improving human thermal comfort for different types of operators reduces adverse factors due to human errors,and provides a safe and optimum decision-making for NPPs.     

An approach to the construction of a one top fire event PSA model

An internal fire event probabilistic safety assessment (PSA) model has been generally quantified by modifications of a pre-developed internal events PSA model. New accident sequence logics not covered in the internal events PSA model are separately developed to incorporate them into the fire PSA model. Previous studies on the changes of the one top internal event PSA model for the one top fire event PSA model have been limited to the equipment failures affected by the fire. In addition, they assumed that the probabilities of basic events associated with equipment or cables impacted by the fire are one. However, the probabilities of spurious operation events and human failure events affected by the fire might not be estimated as one. In this study, new modification rules were proposed for the construction of a one top PSA model for fire events by using a one top internal event PSA model. The proposed new modification rules can be applied to all the fire damage events for the fire-induced equipment failure events and spurious operation events, human error events impacted by a fire, regardless of whether they are estimated as one or not. Applications of the proposed modification rules to the compartment and scenario-level fires for the hypothetical plants were performed for demonstrating their appropriateness to the changes of the one top internal event PSA model to the one top fire event PSA model. In addition, quantification procedure with the one top fire event PSA model was presented and discussed.     

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.     

百万千瓦级核电厂停堆工况下的概率安全评价

对百万千瓦级核电厂停堆运行事故进行内部事件1级概率安全评价(PSA),根据不同的停堆进程分别建立停堆PSA模型,分析经历余热排出系统(RRA)低运行区间(LOI-RRA)水位对电厂风险水平构成的影响;同时采用事故系列先兆标准电厂风险分析模型人员可靠性分析(SPAR-H)方法进行人员可靠性分析,评价其定量化结果的适用性。分析结果表明,停堆工况下的电厂风险不可忽视,在停堆工况下的事故规程有待完善之处,冷停堆工况下由LOI-RRA水位导致堆芯损坏频率明显增加,人因失误是造成停堆高风险的关键因素。     

Study on construction of loss of component cooling water system initiating event fault tree for fire event PSA

ABSTRACT

In this study, the construction of the loss of component cooling water system (LOCCWS) initiating event (IE) fault tree (FT) for an actual fire event probabilistic safety assessment (PSA) model of the Korean reference nuclear power plant considering only IE initiators was validated. The quantification results of the LOCCWS accident sequences obtained using an LOCCWS IE FT model with only initiators are similar to that with initiators and enabling events. This confirmed that the LOCCWS IE FT for an actual fire event PSA model could be constructed by considering only IE initiators. In addition, the same LOCCWS accident sequences were quantified assuming that fire triggering only the LOCCWS IE leads to reactor shutdown. Compared with the quantification result obtained based on the assumption that any fire included in the fire event PSA leads to reactor shutdown, the core damage frequency (CDF) can be reduced. Thus, it can be concluded that there is a possibility of underestimation of CDF when the LOCCWS IE FT model with only initiators is used and the assumption that fire triggering only the LOCCWS IE results in reactor shutdown is employed for the quantification of LOCCWS accident sequences.     

Insights from fire PSA for enhancing NPP safety

This paper presents the findings of an effort to gain insights from fire probabilistic safety assessment (FPSA) conducted in nuclear power plants. Using probabilistic models, the fire PSA takes into account the possibility of a fire at specific plant locations and its propagation, detection and suppression of the fire; and also helps to assess the effect of the fire on safety-related cables and equipment. The results of FPSA contributed to design modifications in plant to enhance the safety and thereby reduce its contribution to core damage frequency. It also highlights the sources of uncertainty while conducting and suggesting values of risk parameter in FPSA study.     

核电厂火灾PSA中火灾场景频率分析方法

核电厂的火灾场景频率分析是火灾概率安全分析的核心内容。本文根据美国NRC和EPRI的《核动力设施火灾概率风险评价方法》,介绍了核电厂内部火灾概率安全评价中,火灾场景频率的分析方法及参数不确定性的处理方法。以福清一期核电厂某房间的电气柜火灾为例,进行了定量分析计算,计算结果表明,在计算中考虑热释放速率参数不确定性的传递可以有效降低计算结果的保守性。     

Concept for comprehensive Hazards PSA and Fire PSA application

This contribution presents results of recent research and development activities in the field of Hazards PSA (HPSA). The reactor accidents at Fukushima Dai-ichi in March 2011 gave reason and indications for checking the risk assessment approach for internal and external hazards as currently described in the German PSA Guideline and its supplementary technical documents. A standardized approach for performing a comprehensive HPSA has been developed emphasizing the complete consideration of all potential failure dependencies induced by hazards. The systematic extension of the given plant model of Level 1 PSA is the real crux of the new HPSA approach. The extension is carried out for each hazard H using the corresponding hazard equipment list (H-EL) and the corresponding hazard dependency list (H-DL). Parts of the approach have already been tested.In the paper a successful application for the plant internal hazard fire is presented. A German licensee plans a system modification of the spent fuel pool cooling, therefore a Level 1 PSA has been carried out to compare the fuel damage frequencies for the existing and the modified version. It is outlined how the systematic (and partly automatic) extension of the fault trees is performed using a so-called Fire Equipment List (F-EL). The F-EL contains a compartment assignment for all relevant components and cables. The probability of a compartment failure by fire must be determined for any compartment mapped. This is the conditional probability that the components and cables within the compartment are inoperable due to the fire.