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人的认知失误事件定量分析法的进展及应用 总被引:3,自引:0,他引:3
认知可靠性与人误分析法(即认知失误分析法,CREAM)是具有代表性的第2代可靠性分析(HRA)方法,它可从回顾式和预测式进行班组人误事件概率的定量分析.本工作除描述了通用的CREAM方法外,还建立了用环境影响指数β与共同绩效条件(CPC)因子关系的人误事件概率简化的定量化公式,可用于计算核电厂人误事故中班组的人误事件概率.并假想以秦山一期蒸汽发生器传热管破裂(SGTR)事故为例,说明人的认知失误事件概率的计算过程及结果,为核电厂概率安全评价(PSA)的班组人因分析提供了另一种有效的途径,使核电厂的风险的概率估计值更为客观、更有参考价值. 相似文献
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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 相似文献
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WANO人因事件统计及分析 总被引:11,自引:0,他引:11
对世界核电厂营运者联合会1993~2002年940份运行事件分析报告进行分析.发现有551件与人因相关.人因失误仍然是核电站事故最主要的诱因之一运用统计技术和数据挖掘技术对这551件人因事件进行分析.获得如下结论:①维修、调试,试验活动中所产生的人误导致系统潜在失效而最终诱发系统事故已成为人因事故最重要的原因;②反应堆启动阶段和停堆阶段人误概率较高,且易诱发严重事故;③人因事件发生概率与反应堆类型没有必然的联系;④对事故征兆或事故判断失误和操作失误是人误的最主要表现形式.也是导致人因事故的最主要原因,造成严重电站瞬态和安全系统的故障或者不恰当投运的事故的可能性较大;⑤理论知识欠缺、基本操作能力较差、组织管理缺陷、规程不足以及粗心大意、缺乏相互检查是导致人因失误最主要的根原因. 相似文献
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概率安全分析(PSA)是核电站安全分析的一种有效方法,在核电站的安全评价、运行、维修及为安全管理部门的决策提供技术支持等方面具有重要的作用。PSA方法已广泛应用于在役核电厂的运行管理、系统及部件的维修以及核电厂的风险监控之中,取得了很好的经济效益与社会效益。长期以来,我国有关PSA的工作主要是在核电厂的应用方面,在PSA方法学研究及PSA技术研发方面做得工作很少。没有开发自己的PSA分析软件,因而,在开展核电厂PSA工作时,采用的都是国外的PSA软件。 相似文献
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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。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(11):1184-1193
This paper deals with a research on human factors performed by the Japanese BWR group, focusing mainly on the work done in Phase I (1984–86) of the research project. As a first simulator study of operator performance during accidents in the nuclear field in Japan, it was necessary to develop analysis methods to identify and quantify operator errors. The performance of operators under plant abnormal conditions was analyzed using full-scale BWR plant simulators. By utilizing on-line plant data collection systems and audio/video devices, data were gathered from actual plant operators in the retraining courses on two full-scale simulators at the BWR operator training center in Japan. More than a hundred cases of operator performance data were analyzed to identify human errors and to classify the types of errors. For application to probabilistic safety assessment (PAS), error probabilities were estimated and compared with existing data. Quantification of cognitive errors was also performed based on the Time-Reliability Correlation. An assessment of the effects of human factors in PSA suggested the importance of errors in the operators' cognitive processes, which accounted for the major effects of errors in tasks of high safety significance. The results of the analysis esrve as a basis for building a data bank to be utilized for multiple purposes of PSA, man-machine interface improvement and operator training. 相似文献
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ZHOU Tao SUN Canhui LI Zhenyang WANG Zenghui Institute of Nuclear Thermal-Hydraulic Safety St ardization North China Electricity Power University Beijing China Graduate University of Chinese Academy of Sciences Beijing China 《核技术(英文版)》2011,(5):316-320
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. 相似文献
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Operator error in diagnosis and execution of task have significant impact on Nuclear Power Plant (NPP) safety. These human errors are classified as mistakes (rule base and knowledge based errors), slip (skill based) and lapses (skill based). Depending on the time of occurrence, human errors have been categorized as i) Category ‘A’ (Pre-Initiators): actions during routine maintenance and testing wherein errors can cause equipment malfunction ii) Category ‘B’ (Initiators): actions contributing to initiating events or plant transients iii) Category ‘C’ (Post-Initiators): actions involved in operator response to an accident. There have been accidents in NPPs because of human error in an operator's diagnosis and execution of an event. These underline the need to appropriately estimate HEP in risk analysis. There are several methods that are being practiced in Probabilistic Safety Assessment (PSA) studies for quantification of human error probability. However, there is no consensus on a single method that should be used. In this paper a method for estimating HEP is proposed which is based on simulator data for a particular accident scenario. For accident scenarios, the data from real NPP control room is very sparsely available. In the absence of real data, simulator based data can be used. Simulator data is expected to provide a glimpse of probable human behavior in real accident situation even though simulator data is not a substitute for real data. The proposed methodology considers the variation in crew performance time in simulator exercise and in available time from deterministic analysis, and couples them through their respective probability distributions to obtain HEP. The emphasis is on suitability of the methodology rather than particulars of the cited example. 相似文献
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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. 相似文献
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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。 相似文献
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Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis (BEPU) is increasingly being used for deterministic calculation in NPPs. The PSA methodology integrates information about the postulated accident, plant design, operating practices, component reliability and human behavior. The deterministic and probabilistic methodologies are combined by analyzing the accident sequences within design basis in the event trees of a postulated initiating event (PIE) by BEPU. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. 相似文献
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核电厂传统人员可靠性分析方法中引入班组因素的研究 总被引:1,自引:0,他引:1
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。 相似文献