核电厂主控室火灾场景分析及风险评价

主控室是核电厂火灾概率安全评价的主要关注对象之一。本文对典型核电厂的主控室火灾场景进行分析并对由其导致的反应堆堆芯损坏频率进行计算评价,主要使用事件树方法演绎火灾场景,通过火灾模拟计算确定火灾场景的危害,最后在电厂内部事件一级PSA模型的基础上建立火灾风险评价模型,完成主控室火灾风险定量化。火灾演绎分析结果获得了4个火灾场景,分别能够导致不同的电厂始发事件,并对相关的操纵员动作产生较大影响。风险定量化结果表明:主控室火灾导致的堆芯损坏频率为1.953×10~(-9)/堆年。     

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。     

核电厂电气设备间火灾概率安全分析研究

电气设备间是核电厂火灾风险评价的重要内容之一。本文对典型核电厂的电气设备间进行火灾风险分析,使用事件树方法演绎分析火灾场景并确定火灾场景的危害,最后在核电厂电气设备间火灾序列演绎分析的基础上建立火灾风险评价模型,完成电气设备间火灾风险定量化。火灾演绎分析结果获得了6个火灾场景,分析各火灾场景对核电厂始发事件和系统设备的影响。风险定量化结果表明:电气设备间火灾导致的堆芯损坏频率为1.42×10~(-9)/(堆·年)。     

核电厂内部火灾概率安全评价现状

核电厂运行经验表明火灾对其安全具有严重威胁,各国安全监管当局也加强了对核电厂火灾安全的监管,要求核电厂实施火灾危害性分析,并对火灾风险进行评估。详细介绍了核电厂内部火灾概率安全评价(PSA)的发展历史与开展情况,并对主要方法和标准做了简要介绍。     

关于核电厂火灾概率风险的研究

火灾是核电厂安全面临的重要威胁之一。应用概率风险评价(PRA)方法对其进行分析,能找出电厂薄弱环节,优化电厂的设计。通过研究国际广泛使用的火灾PRA方法,以典型的二代压水堆核电厂为对象,开展了火灾概率风险分析,计算得到了火灾引起的堆芯损坏频率(CDF)为4.03×10-6(堆·年)-1。在此基础上,开展了敏感性分析,讨论了人因事件和定量筛选值对结果的影响。     

核电厂起火频率分析

通过研究美国核管理委员会(NRC)和电力研究院(EPRI)的《核电厂的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(PSA)过程中,各类点火源起火频率的分析方法和步骤。以大亚湾核电厂的主变压器为例,介绍了起火频率的具体分析过程。经定量计算分析,大亚湾核电厂主变压器的起火频率是4.32×10-3/(堆.年),是反应堆堆芯损坏频率(CDF)的203倍。一旦起火,发生破坏性火灾的概率高达83%。     

基于FDS的核电厂主控室火灾PSA研究

主控室火灾是核安全领域的重要课题,一旦发生事故,会危及操纵员和设备安全,进而影响到核电厂运行和安全停堆的控制,导致堆芯损坏。本文在介绍主控室火灾特点的基础上,在国内率先将火灾动力学模型FDS(Fire Dynamics Simulator)应用到主控室火灾概率安全评价(PSA)中,通过实例分析主控室火灾情境中关键参数的变化规律,讨论主控室火源的热释放速率及操纵员撤离要求,得到主控室主专用安全盘和次专用安全盘火灾导致的堆芯损坏频率分别为1.0×10-7/(堆·年)和2.5×10-8/(堆·年),体现FDS在核电厂火灾应用领域的优势。     

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。     

火灾情景下的人员可靠性分析

国内外各核电厂火灾概率安全评价（PSA）表明,人员操作对火灾情景下的电厂风险有重要影响,因此,有必要采用系统的人员可靠性分析（HRA）方法来评价火灾情景下的人员失误概率。本文阐述了HCR/ORE和CBDTM模型的基本理论和在火灾情景下的特殊考虑。将HCR/ORE和CBDTM方法与THERP方法相结合应用于火灾情景下的人员可靠性分析,并进行了实例分析。为建立更符合工程实际的火灾PSA模型奠定了基础。     

基于主控室电气柜火灾的CFAST程序参数不确定性分析

以核电厂主控室电气柜火灾为研究对象,利用蒙特卡洛抽样法对热释放速率和产烟率这2个参数进行抽样,并输入CFAST程序进行计算。通过统计烟气层温度和光学密度2个输出量的分布,获得主控室人员撤离时间和概率信息,为火灾概率安全分析当中事件序列定量分析提供基础数据。     

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.     

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.     

Considerations of uncertainties in evaluating dynamic reliability by GO-FLOW methodology – example study of reliability monitor for PWR safety system in the risk-monitor system

The uncertainty analyses have been considered as a relevant topic since WASH-1400 and analysis was performed for identifying the risk measure, e.g. plant- and core-damage frequency or the frequency of a large early release of radioactivity in the probabilistic safety assessment (PSA) or probabilistic risk assessment. There are two main sources of uncertainty such as aleatory uncertainty and epistemic uncertainty (parameter uncertainty, model uncertainty and completeness uncertainty) for risk analysis in PSA or risk-monitor system. A sensitivity analysis is related field to uncertainty, which can provide information of the most effective on those inputs of PSA, which are mostly contributed to the uncertainty.

In this paper, uncertainty analysis (epistemic) has been conducted in the evaluation of dynamic reliability of safety-related subsystem for risk analysis. GO-FLOW methodology has been employed for the procedure of uncertainty analysis alternatively to Fault Tree Analysis and Even Tree because it is success-oriented system-analysis technique and comparatively easy to conduct the reliability analysis of the complex system. The method used sample data from Monte Carlo simulation to quantify uncertainty in terms of appropriate estimates for analysis results. Pressurized water reactor containment spray system has been taken as an example of safety-related subsystem. The results of this paper show that the uncertainty analysis is an important part for the practical evaluation of the system dynamic reliability and makes the reliability prediction more accurate compared with the result without the uncertainty analysis. The GO-FLOW methodology can be employed easily for uncertainty analysis with its advance functions.     

Effect of uncertainties in best-estimate thermal hydraulic analysis on core damage frequency for PSA

Generally, thermal hydraulic (TH) analyses have been performed as part of a probabilistic safety assessment (PSA) to construct event trees and to evaluate success criteria. Even though an accident scenario in an event tree for PSA is exceedingly dependent on many uncertainty parameters, TH analysis in PSA, up to now, has been performed without considering the uncertainties for the important parameters. In the present study, TH analysis was carried out using the MARS code to simulate the large break loss of coolant accident (LBLOCA) which is one of the event sequences of level 1 PSA in an optimized power reactor 1000 MWe (OPR1000). First, the phenomena identification and ranking table (PIRT) for LBLOCA were established, and the candidate parameters were set-up. Once the input file for the MARS code was made with consideration of the uncertainties of the candidate parameters, and a parameter assessment was carried out with the MARS code to rank the candidate parameters according to the effect on peak cladding temperature (PCT). For the five highest-ranking parameters resulting from parameter assessment, the probability density function (PDF) of PCT was derived by the response surface method (RSM), and comparative Monte Carlo calculations were also performed to assess the accuracy of the RSM. As a result, it was shown that by considering the uncertainties of the TH analysis, the accident sequence, which had filed in the PSA result in the established PSA results, had a possibility of succeeding, and thus, be able to modify the core damage frequency (CDF).     

核电厂火灾PSA方法浅析

作为PSA工作中不可缺少的一部分,核电厂火灾PSA正在发挥着越来越重要的作用。本文对核电厂火灾PSA的发展、应用和研究的基本情况进行了论述。     

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。     

基于FDS程序的钠火事故情景模拟技术研究

钠火是钠冷快堆的典型事故,钠火事故情景计算机模拟仿真是对钠火事故风险评价的有力工具。本文以常规火灾三维计算流体力学软件FDS为平台,增加钠火燃烧模型,包括燃烧热模型、燃烧速率模型、喷射液钠粒径分布模型等,完成了钠火情景模型的开发,并通过与SPHINCS钠火试验和计算结果的温度分布与氧气含量对比,验证了模拟技术和模型开发方案的可行性。本文的研究成果能为后续钠火仿真模拟程序的开发提供研究基础和经验参考。     

Sodium Fire Accident Scenario Simulation Technology Study Based on FDS Code

Sodium fire is a typical accident of sodium-cooled fast reactor. Simulation of sodium fire accident scenario by software is a powerful tool for risk assessment of sodium fire accident. In this paper, the conventional fire three-dimensional computational fluid dynamics software FDS was used as a platform to add a sodium fire combustion model, including combustion heat model, combustion rate model, spray liquid sodium particle size distribution model, etc., and complete the development of sodium fire scenario modeling analysis program. And through the comparison with SPHINCS sodium fire test and calculation results, the feasibility of the method and development plan was verified. The research results of this paper can provide the research basis and experience reference for the development of the subsequent sodium fire simulation program.