人的认知失误事件定量分析法的进展及应用

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.     

概率风险评价在反应堆保护逻辑改进中的应用

用概率风险评价方法对大亚湾核电站反应堆保护逻辑设计进行了分析，证实原安全壳压力高通道及换料水储存箱水位低通道的逻辑设计存在失电误动的严重问题。该保护误动将直接导致轴封水丧失事故发生。由此引起的轴封水丧失事故事件概率为６．５×１０－６。     

一种基于HRA的数字化人-机界面评价方法研究

为识别数字化人-机界面中可能诱发人因失误或弱化操纵员绩效的设计缺陷,建立了一种基于HRA的人-机界面评价方法：HCR+CREAM+HEC。首先,采用HCR方法从事件整体中识别出失误概率高的风险场景;然后,针对高风险场景采用CREAM方法确定各种失误模式及其失误概率,并对失误概率进行排序;最后,依据数字化人-机界面特征建立人因工程检查表,对失误概率高的人-机界面进行审查,以识别人-机界面设计中存在的缺陷,并提出改进建议。结果表明,该方法能快速有效地识别出数字化人-机界面设计中存在的容易诱发人因失误的缺陷,通过设计优化提高核电站数字控制系统运行的安全性。     

反应堆系统人因事件分析与预防方法及应用

以人因工程学、控制论和安全科学为工具,采用综合集成及实证性方法,研究了人因(事件)分析学科面临的问题和发展趋势、人因事件分类体系、人因事件成因理论、人因事件定量评价方法、人因事件数据采集、处理及数据库、组织管理因素对人因事件的作用和影响、人因事件根原因分析技术、人因事件模式与影响及严重度分析、人因事故防御方法与策略,构建了反应堆系统中人因事件分析理论和方法,且在大亚湾核电站、岭澳核电站获得了成功应用。     

核电厂临时设备投运人员可靠性评估模型

核电厂临时设备作为严重事故缓解的重要设施,其接入工序大多较为复杂。为了分析核电厂人员在临时设备投运时的可靠性,通过研究福岛核事故后改进项所增设临时设备接入行为的特征,基于人因失误模式和影响分析,定义人因失误发生概率、人因失误影响程度、人因失误可恢复概率为风险因子,结合专家评价与模糊语言理论提出一种临时设备投运人员可靠性评估模型。以全厂断电事故下移动电源的接入任务为例,应用所建模型获得了该任务中的人误模式重要度排序及合理的风险见解,验证了模型的可行性。      

考虑组织和管理因素影响的人误概率计算方法

采用三级影响图模型，对人的失误概率计算进行了描述，模型中能考虑组织和管理因素对人误概率的影响。概率影响图最初只是作为构造和表征决策树或事件树模型的一种工具，本文按三级影响水平对产人识的原因建立了分析模型，介绍了定量求解影响图模型的方法。它可用于核电厂概率安全评价中的人误概率计算，也可用于复杂树的定量化，它是一种工程决策分析中的重要方法。     

广东核电站的人可靠性分析

人的可靠性分析是广东核电站PSA研究中的一个重要部分,其目的是研究电站中人的行为对反应堆堆芯熔化频率的影响,文章对广东核电站事故序列中可能出现的人误进行了定性和定量化分析,给出了PSA研究中需要的人误概率值。详细的人因分析将在以后的工作中进行。     

SGTR事故人员可靠性DFM模型定量化方法研究

以SGTR事故人员可靠性DFM模型求解为基础,对模型定量化中的技术难点进行分析;结合THERP、HCR等第一代人员可靠性方法中人误数据库对DFM模型进行定量化分析和讨论。结果表明:质蕴含PI#5和PI#6人误概率占SGTR事故中人误的主要部分,前几个时间段的执行失误和诊断/决策失误的FV及RAW重要度相对较大,将时间划分为2步长、3步长和1步长的总体人误概率无显著差异,这都与如何获得的人误数据及处理质蕴含内部相关性等密切相关。     

核电厂激发事故初因的人因事件分析

人因可靠性分析(HRA)已成为概率安全分析(PSA)不可或缺的内容.激发事故初因的人因事件(B类人因事件)分析作为HRA的重要组成部分,在国内外尚无正式的分析报告.本文描述了B类人因事件的定义和分类,建立了B类人因事件分析基本程序和方法,该方法已在国内某核电厂最近的HRA分析中得到应用.文中还对1993年～2002年WANO 940件运行事件和国内某核电厂运行事件进行了B类人因事件统计分析和发生原因分析,并据此提出了预防和减少B类人因事件的措施.     

CREAM追溯法及其在根原因分析中的应用

认知可靠性和失误分析方法(CREAM)是第二代人因可靠性分析方法中的代表方法之一,它具有追溯和预测的双向分析功能。介绍了CREAM追溯分析方法的基本思想,给出了追溯分析的实现框架和具体步骤。应用CREAM追溯法对三哩岛事故进程中的重要人因失误事件进行了根原因追溯分析,结果表明根原因是和电站的情景环境有关的,是迫使人因失误发生的因素,也说明了CREAM追溯分析方法的实用性和有效性。     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.     

Human error probability estimation by coupling simulator data and deterministic analysis

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.     

基于氢气风险分析的ADS手动卸压策略研究

先进非能动压水堆设计采用自动卸压系统(ADS)对一回路进行卸压,严重事故下主控室可手动开启ADS,缓解高压熔堆风险。然而ADS的设计特点可能导致氢气在局部隔间积聚,带来局部氢气风险。本文基于氢气负面效应考虑,对利用ADS进行一回路卸压的策略进行研究,为严重事故管理提供技术支持。选取全厂断电始发的典型高压熔堆严重事故序列,利用一体化事故分析程序,评估手动开启第1~4级ADS、手动开启第1~3级ADS、手动开启第4级ADS 3种方案的卸压效果,并分析一回路卸压对安全壳局部隔间的氢气负面影响。研究结果表明,3种卸压方案均能有效降低一回路压力。但在氢气点火器不可用时,开启第1~3级ADS以及开启第1~4级ADS卸压会引起内置换料水箱隔间氢气浓度迅速增加,可能导致局部氢气燃爆。因此,基于氢气风险考虑,建议在实施严重事故管理导则一回路卸压策略时优先考虑采用第4级ADS进行一回路卸压。     

应用ECF图对控制棒驱动机构电源全部丧失导致反应堆自动停堆事件分析

国内某核电厂发生了控制棒驱动机构电源全部丧失导致反应堆自动停堆运行事件,为系统分析导致事件发生的原因,本文运用事件原因因素(ECF)图分析法对其进行了事件调查和原因分析。通过梳理事件序列,确定了事件发展过程中的失效点,通过对各失效点深入分析,确定了事件发生的促成原因和根本原因。发生该事件的根本原因是:隔离经理不了解触碰3RAM601JA可能导致停堆的风险,执行了程序规定以外的动作,在操作过程中未对操作进行自检或采取其他防人因失误方法,工作技能不足;电厂相关人员核安全文化存在缺陷,将尽快恢复电厂运行置于优先位置,主动违反了SOP程序规定。     

AP1000核电厂大量放射性释放源项分析

在AP1000核电厂的某些严重事故情景中,安全壳可能发生失效或旁通,导致大量放射性物质释放到环境中,造成严重的放射性污染。针对大量放射性释放频率贡献最大的3种释放类别（安全壳旁通、安全壳早期失效和安全壳隔离失效）,分别选取典型的严重事故序列（蒸汽发生器传热管破裂、自动卸压系统阀门误开启和压力容器破裂）,使用MAAP程序计算分析了释放到环境中的裂变产物源项。该分析结果为量化AP1000核电厂的放射性释放后果和厂外剂量分析提供了必要的输入。     

Quantification of unavailability caused by random failures and maintenance human errors in nuclear power plants

Risk analysis of any equipment or system estimates the unavailability of redundant components due to hardware failure, periodic test and repair work, and human errors in maintenance tasks. A model has been developed in this study to estimate the unavailability of a periodically repairable component of a system that considers hardware failure and maintenance errors due to human failure, and which is applicable to estimate system unavailability in nuclear power plants as well as other industries. A sensitivity analysis has been performed based on the newly developed model considering multiple levels of human errors in maintenance tasks to observe the effects of maintenance errors on overall unavailability of a component. The effect of human error on the optimal frequency of periodic maintenance of a device has been analyzed based on the newly developed unavailability quantification model.     

基于认知模型与故障树的核电厂严重事故下人因失误分析

为分析核电厂应急人员在处理严重事故时可能发生的人因失误,通过建立不同应急人员的认知模型及识别相应的行为影响因子,在认知功能的基础上识别出13种人因失误模式:信息来源不足、信息可靠性不佳、过早结束对参数的获取、重要数据处理不正确、缓解措施负面影响评估失误、选择不适用当前情景的策略、延迟决策、遗漏重要信息/警报、延迟发觉、软操作失误、信息反馈失效、设备安装/连接/操作失误、延迟实施,并基于故障树分析得出人因失误模式的主要根原因:交流失效、时间压力、事故发展的不确定性、信息接收延误、监视失误、人-机界面不佳和环境因素。分析结果可用于预测严重事故缓解进程中可能出现的人因失误,为核电厂实施严重事故管理和技术改进,以及保障严重事故工况下核电厂安全提供参考。     

Comprehensive and consistent interpretation of local fault experiments and application to hypothetical local overpower accident in Monju

Experimental studies on local fault (LF) accidents in fast breeder reactors have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Comprehensive and consistent interpretations of in-pile and out-of-pile experiments related to LF were arrived at in this study based on state-of-the-art review and data analysis techniques. Safety margins for a hypothetical local overpower accident, which was evaluated as a LF accident in the licensing document of the construction permit for a prototype fast breeder reactor called Monju, were also studied. Based on comprehensive interpretations of the latest experimental database, including those performed after the permission of Monju construction, it was clarified that the evaluation of the hypothetical local overpower accident in the Monju licensing was sufficiently conservative. Furthermore, it incorporated adequate safety margins in terms of failure thresholds of the fuel pin, molten fuel ejection, fuel sweep-out behavior after molten fuel ejection, and pin-to-pin failure propagation. Moreover, these comprehensive interpretations are valid and applicable to the safety evaluation of LF accidents of other fast breeder reactors with various fuel and core designs.     

Reactor protection system software test-case selection based on input-profile considering concurrent events and uncertainties

Recently, the input-profile-based testing for safety critical software has been proposed for determining the number of test cases and quantifying the failure probability of the software. Input-profile of a reactor protection system (RPS) software is the input which causes activation of the system for emergency shutdown of a reactor. This paper presents a method to determine the input-profile of a RPS software which considers concurrent events/transients. A deviation of a process parameter value begins through an event and increases owing to the concurrent multi-events depending on the correlation of process parameters and severity of incidents. A case of reactor trip caused by feedwater loss and main steam line break is simulated and analyzed to determine the RPS software input-profile and estimate the number of test cases. The different sizes of the main steam line breaks (e.g., small, medium, large break) with total loss of feedwater supply are considered in constructing the input-profile. The uncertainties of the simulation related to the input-profile-based software testing are also included. Our study is expected to provide an option to determine test cases and quantification of RPS software failure probability.     

Containment venting as a mitigation technique for BWR MARK I plant ATWS

Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it.Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure. For both cases the operators are assumed to take action to intentionally restrict injected flow such that fuel in the upper part of the core would be steam cooled. Resulting fuel temperatures are estimated with an off-line calculation and found to be acceptable.