概率安全分析软件RiskA与RiskSpectrum的故障树计算引擎计算对比分析研究

本文采用完整的秦山第三核电厂PSA模型,分别用中科院FDS团队自主研发的概率安全分析软件RiskA的计算引擎RiskAT与瑞典斯堪伯奥公司开发的RiskSpectrum的计算引擎RSAT进行了计算,结果表明二者定性和定量计算结果一致,在计算性能方面,RiskA的计算速度快于RiskSpectrum。     

不确定性分析在概率安全评价中的应用

分析了概论安全评价（PSA）中存在的完整性，模型假设条件及输入数据的不确定性和它们的来源。针对输入参数的不确定性，阐述了Risk Spectrum软件关于不确定性分析的原理，方法和误差因子的选取。对输入参数的不确定性进行定量计算后，得到13个初因和各工况的堆芯损坏频率的均值。介绍了表征不确定性的概率密度函数和累计密度函数曲线。     

概率安全评价软件RiskA中的非逻辑处理方法

非单调关联系统广泛存在于实际工程应用中,传统针对单调系统的处理方法不适合于这类系统的处理。因此,如何处理针对非单调关联系统所建立的模型成为概率安全评价软件研发面临的问题之一。本工作在调研一些国际流行概率安全评价软件非逻辑处理方法的基础上,探讨了非逻辑求解难点,基于RiskA的数据结构,设计并实现了非逻辑处理模块,并通过例题验证了RiskA软件非逻辑处理模块的正确性和可靠性。     

经验贝叶斯方法在核电站概率安全评价中的应用

文章阐述在应用贝叶斯估计时选择共轭先验分布的指导思想。详细论述了当运行失效和需求失效的先验分布为Lognormal分布时，利用重点式转换原则，将Lognormal分布转换为Gamma分布和Beta分布的方法。该方法计算简便，易于工程应用。基于该方法，研究开发了适于核电站可靠性参数计算分析的软件。     

核电厂地震概率安全评价中的电气设备易损度计算

随着福岛事故的发生,核电厂外部事件概率安全评价工作的重要性被各国核安全当局所认同。而地震,作为核电厂最为主要的外部事件,其对应的概率安全评价工作便更为人们所重视。易损度计算是完成地震概率安全评价的关键技术环节,其结果将被使用作概率安全评价事故序列模型的输入条件。因此,易损度计算的准确性和正确性对地震概率安全评价工作最终结论的影响也就不言而喻了。本文首先总体性介绍了设备易损度计算的基础数学模型,随后详细描述了核电厂地震概率安全评价中电气设备易损度计算的操作步骤,并重点探讨了电气设备功能失效模式下对试验反应谱和要求反应谱的处理简化技巧,最后通过具体算例阐述了电气设备易损度计算过程中的注意事项和简化技巧应用。     

法国的核电站概率安全评价技术

法马通公司发现,概率安全评价技术(PSA)使其现有核电站得到了很大的改进,PSA技术将用作今后反应堆设计的一种手段,和确保进一步提高其运行性能的方法。法马通公司使用PSA技术进行核电站设计,差不多有20年了。最近,该公司还参加了比利时的多伊尔3号堆和蒂昂热2号堆的PSA工作。希望下一代核电站进一步有改进的重要方面是,降低堆芯损坏概率。PSA可用来分析     

概率安全评价在设计中的应用

概率安全评价(PSA)已经成为核电站设计过程的一项重要的工具。     

核电站风险概率评价

本文说明核电站风险概率评价(PRA)的目的、时机、所需资料与人员以及研究对象。简述了用功能事件树、系统事件树确定事件列,用故障树推导系统模型的方法。最后给出一些事故发生频率的统计数据。     

核电站严重事故后果概率安全评价方法研究

核电站严重事故后果概率安全评价(PSA)是采用概率论的方法对核电站放射性后果进行分析,并定量给出放射性物质对核电站周围公众的健康效应影响。以国内某压水堆核电站为参考厂址,建立合适的场外后果分析模型。采用分层抽样方法对参考厂址1a的气象数据进行抽样,源项和释放特征等数据取自二级PSA的研究结果。利用事故后果评价程序对核电站严重事故后果进行计算,并用概率论方法对结果进行评估。通过计算将各事故和事故谱的场外个人剂量表示为CCDF曲线和总频率-剂量曲线,再用概率论方法得到不同距离处个人剂量超过指定剂量的条件概率;也可用此方法对确定烟羽应急计划区的安全准则中所描述的"大多数严重事故序列"进行量化。     

概率安全评价在核电厂安全设计中的应用研究

概率安全评价(PSA)已在核电厂安全设计中起到了不可或缺的作用,随着新电厂设计中安全性和经济性矛盾的不断凸显,PSA的应用会进一步得到加强。本文从核安全设计的基本逻辑出发,从知裕量的设计、减轻不必要的安全负担、支持厂址相关的分析及闭环管理等方面提出了在核安全设计中更加深入应用PSA的若干建议,这些建议可供核电厂安全设计和评审参考。     

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.     

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.     

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components.This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.     

Seismic probabilistic safety analysis of unit 1 of the Loviisa nuclear power plant

The seismic probabilistic safety assessment consists of five phases. In the seismic hazard analysis the seismicity of the plant site is quantified. In the second phase, the structural response of plant buildings is evaluated. On the basis of structural response, the seismic fragilities of selected plant components are developed. In the following phase, the plant logic in the form of fault trees and event trees is established. In the last step, quantification of the core damage risk on the basis of the above information is carried out. For the median value of the annual core damage frequency, a value of 4.4 × 10⁻⁷ was determined.     

Success criteria analysis in support of probabilistic risk assessment for nuclear power plants: application on SGTR accident

Success criteria analysis (SCA) bridges the gap between deterministic and probabilistic approaches for risk assessment of complex systems.To develop a risk model,SCA evaluates systems behaviour in response to postulated accidents using deterministic approach to provide required information for the probabilistic model.A systematic framework is proposed in this article for extracting the front line systems success criteria.In this regard,available approaches are critically reviewed and technical challenges are discussed.Application of the proposed methodology is demonstrated on a typical Westinghouse-type nuclear power plant.Steam generator tube rupture is selected as the postulated accident.The methodology is comprehensive and general;therefore,it can be implemented on the other types of plants and complex systems.     

Success criteria analysis in support of probabilistic risk assessment for nuclear power plants: application on SGTR accident

Success criteria analysis (SCA) bridges the gap between deterministic and probabilistic approaches for risk assessment of complex systems.To develop a risk model,SCA evaluates systems behaviour in response to postulated accidents using deterministic approach to provide required information for the probabilistic model.A systematic framework is proposed in this article for extracting the front line systems success criteria.In this regard,available approaches are critically reviewed and technical challenges are discussed.Application of the proposed methodology is demonstrated on a typical Westinghouse-type nuclear power plant.Steam generator tube rupture is selected as the postulated accident.The methodology is comprehensive and general;therefore,it can be implemented on the other types of plants and complex systems.     

Thermal-hydraulic calculations using MARS code applied to low power and shutdown probabilistic safety assessment in a PWR

The methods developed for full-power probabilistic safety assessment, including thermal-hydraulic methods, have been widely applied to low power and shutdown conditions. Experience from current low power and shutdown probabilistic safety assessments, however, indicates that the thermal-hydraulic methods developed for full-power probabilistic safety assessments are not always reliable when applied to low power and shutdown conditions and consequently may yield misleading and inaccurate risk insights. To increase the usefulness of the low power and shutdown risk insights, the current methods and tools used for thermal-hydraulic calculations should be examined to ascertain whether they function effectively for low power and shutdown conditions. In this study, a platform for relatively detailed thermal-hydraulic calculations applied to low power and shutdown conditions in a pressurized water reactor was developed based on the best estimate thermal-hydraulic analysis code, MARS2.1. To confirm the applicability of the MARS platform to low power and shutdown conditions, many thermal-hydraulic analyses were performed for the selected topic, i.e. the loss of shutdown cooling events for various plant operating states at the Korean standard nuclear power plant. The platform developed in this study can deal effectively with low power and shutdown conditions, as well as assist the accident sequence analysis in low power and shutdown probabilistic safety assessments by providing fundamental data. Consequently, the resulting analyses may yield more realistic and accurate low power and shutdown risk insights.     

非能动先进压水堆核电厂SGTR事故概率安全评价

蒸汽发生器传热管破裂(Steam Generator Tube Rupture,SGTR)事故是核电厂的重要事故之一,并具有其自身的特点。该事故的研究和评价对核电站安全具有较大意义。选取典型非能动先进压水堆核电厂AP1000的SGTR事故进行一级概率安全评价(Probabilistic Safety Assessment,PSA),采用事件树分析方法得到电厂事件发生后系统、设备和人员不同响应所产生的事故序列,然后建立相关系统的故障树模型进行可靠性分析。借助Risk Spectrum软件,计算SGTR事故导致AP1000核电厂的堆芯损伤频率(Core Damage Probability,CDF),并进行堆芯损伤的最小割集分析及重要度和敏感性分析。通过一系列分析得到导致堆芯损伤的重要基本事件,从而找到系统存在的薄弱环节。     

Quantile estimate of the uncertainties of probabilistic safety analysis for objects of the nuclear power industry

Methods for performing comparative analysis of nuclear power plant safety and estimating the residual risk, which are based on analysis of the 95% quantiles of the resulting distributions of the probability density of events which are important for safety, are formulated using an approximate calibration method of quantile estimates of the uncertainties and for the example of the results of a probability analysis of the safety of nuclear power plants in the USA which are presented in the NUREG-1150 report. The basic assumptions of the methods which make it possible to estimate the stationary risk are presented. __________ Translated from Atomnaya énergiya, Vol. 101, No. 3, pp. 167–176, September, 2006.