反应堆操纵员培训及考核的有关问题

反应堆操纵员是核电厂安全稳定运行的基本保证,对操纵员的筛选、培训和考核尤为重要。但随着核电的发展,以及社会民众对电厂安全性要求的提高,目前我国的反应堆操纵员的培训与考核在某些方面已逐渐不能适应我国核电发展的需求。笔者分析了现阶段我国操纵员考核制度、培训手段、选拔方式等方面的不足,结合我国核电发展对操纵人员的要求,提出操纵人员考核及管理体系可优化性、加强复杂工况培训以及重视操纵员心理素质等方面的一些观点。     

核电厂运行操纵员人员绩效改进

核电厂运行操纵员的人员绩效是个永恒的话题,也是我们在核电运行领域的管理人员不断挑战自我的主题,操纵员作为安全发电的可变主体,他们不同于普通的设备或构筑物,可以通过大量的技术数据作为衡量、判断其可靠性的依据,如何使操纵员人员行为的不确定性降到最小,就成为关键。文章通过实践应用的方法、分析和效果,讨论了有别于WANO(世界核电营运者协会)防人因工具的一些管理手段,弥补了防人因工具使用的单调性,有利于防人因工具的有效应用。     

韩阿签署核电运行支持服务协议

正【本刊2016年8月综合报道】2016年7月25日,韩国水电核电公司(KHNP)与阿联酋核能公司(Enec)签署核电运行支持服务协议。根据这份协议,韩国专业人员将在巴拉卡核电厂4台机组全部投运的前10年内为该电厂提供运行支持服务。巴拉卡3号机组近期完成反应堆压力容器安装。核电运行支持服务协议根据协议,水电核电将派遣富有经验的合格核电厂工作人员(包括主控室操纵员和各岗位操纵员)为阿核能公司负责核电运营的子公司Nawah能源公司(Nawah Energy)提     

核动力装置运行故障诊断系统设计研究

为了有效提高核动力装置运行时的安全性与可靠性,减少人因失误,采用基于专家系统的故障诊断技术并结合人因工程概念,对运行故障诊断系统的设计进行了初步研究。首先,对系统应具有的功能进行了分析,提出了关键安全参数及其分级的思想,从方便操纵员由宏观到细节掌控核动力装置运行安全状态的目标出发,对安全状态监督功能模块进行了概要设计;然后,对运行故障诊断专家系统的结构、数据库和知识库的构成进行了设计,设计了分级推理和前后向推理相结合的推理机制,以便最大程度地提高推理过程的快速性与准确性,并符合操纵员运行操纵思维过程和经验知识的表达方式。     

核电站主控室数字化操纵员评价系统

为准确和高效地进行核电站主控室人机系统匹配关系的评价,分析了操纵员评价人体模型的表达形式和人体参数,确定了数字化操纵员应具备的评价功能。基于二次开发技术开发了数字化操纵员评价人体模型,该模型可根据具体评价的需要,选择操纵员的生成方式、性别、操作姿态、单/双眼视域、左/右手可达域,用于评价显示信息是否在操纵员视域范围内以及评价操纵设备是否可触及;同时还具备作业空间和字符可视性评价等关键功能。实例应用表明,该方法能准确、高效和快捷地完成核电站主控室人机匹配关系的评价工作,使二者的协调关系达到最优。     

提高预备操纵员现场操作的参与率

中核辽宁核电有限公司首批预备操纵员正在代培核电站进行操纵员培训。公司"锐意进取"QC小组对现场操作培训中存在的问题进行了分析,选择"提高预备操纵员现场操作的参与率"作为活动课题,并结合学员两个月内操作跟踪信息挖掘出导致现场操作参与率低的两大症结:"无法第一时间获取操作信息"和"与当前操作冲突"。小组使用关联图查找末端因素,并采用头脑风暴、资料统计、现场调查等方法确定了"内部学习工作计划制定欠妥"和"操作信息获取渠道单一"两条要因,经讨论制定出两条对策内的九项措施,通过对策实施圆满实现课题目标。     

在核电厂操纵人员资格核准工作中对其所要求文化程度的一点认识

本文阐述了核安全法规中对操纵人员文化程度的要求,并针对当前在操纵员执照核准工作中出现的新问题,结合对《高等教育法》的理解,提出了一点个人认识,并建议尽快改进和完善核电厂操纵人员执照核准工作,保证核电厂运行安全.     

核电厂先进测量控制系统发展：ADIOS和ASICS

文章中讨论了两个系统：自动启动智能控制系统(ASICS)，它自动地控制压水堆核电厂从冷停堆到反应堆功率的5％；警报与综合诊断操纵员支持系统(ADIOS)，它把报警过程值和诊断信息集成到一个进行警报处理的专家系统中。根据核电厂启动操作的常规规程，核电厂是手动控制从冷态到5％功率的.在核电厂和其它电厂中，检查异常主要利用的资源是警报信息。传统的硬件报警系统以单传感器单指示器为特点。这将使控制室中的操纵员在面对纷至沓来的警报的瞬间产生思维混乱、设计自动启动智能控制系统(ASICS)和警报与综合诊断操纵员支持系统(ADIOS)以减轻操纵员的负担。计算机软、硬件和信息处理技术的发展，给核电厂和有相关过程的电厂的控制系统与报警系统的改善提供了一个好的机会。对于像ASICS和ADIOS这样以计算机技术或以软件技术为基础的系统来说，测试与评价它们的功能和性能是非常必要的ASICS和ADIOS的功能和性能的评价是在实时功能测试装置上进行的。结果表明：开发的系统对于运行和操纵员的支持是有效的。     

基于人员绩效的核电厂数字化人机界面评价研究

与传统的操作界面相比,核电厂数字化人机界面有了很大变化,在诸多方面都变得更为复杂,这无疑影响了操纵员的情景意识和人员绩效。为识别当前数字化界面存在的问题,为核电厂人机界面的设计和优化提供理论支持,本文对数字化核电厂相关人员进行问卷调查,问卷内容涉及数字化人机界面的6个方面,分别为信息显示系统、数字化SOP系统、软控制系统、报警系统、操纵员支持系统和情景交互。通过分析,发现当前数字化界面在操作画面的信息分配、风险提示、操作难度、辅助功能等方面存在不足,整体情景交互效果不佳,需进行优化完善。     

基于认知的数字化核电站人机接口设计

计算机技术的广泛应用和自动化程度的提高改变了操纵员监视和控制电厂系统及设备的行为方式。在数字化仪控系统中,人机接口(HSI)的设计思路和方法也应随之转变。人机接口是操纵员与仪控系统进行交互的主要媒介。本文提出一种基于认知过程的分析和设计人机接口系统的方法。为了实现"以人为中心"的人机接口设计,根据人的行为过程和Rasmussen三级行为模型开发了适用于核电站操纵员的认知行为模型。按NUREG 0711中对于运行概念的要求,定义了操纵员与人机接口系统的信息交互方式,通过人机接口资源与操纵员认知模型的映射关系,设计符合操纵员认知决策需求的人机接口系统。最后,给出了各项人机接口资源的设计关键点,帮助识别设计过程中的需要特别关注的关键问题。     

Development of advanced I&C in nuclear power plants: ADIOS and ASICS

In this paper Automatic Startup Intelligent Control System (ASICS) that automatically controls the PWR plant from cold shutdown to 5% of reactor power and Alarm and Diagnosis-Integrated Operator Support System (ADIOS) that is integrated with alarms, process values, and diagnostic information to an expert system focused on alarm processing are described. Nuclear Power Plant is manually controlled from cold shutdown to 5% according to the general operation procedures for startup operation of nuclear power plant. Alarm information is the primary sources to detect abnormalities in nuclear power plants or other process plants. The conventional hardwired alarm systems, characterized by one sensor-one indicator may lead the control room operators to be confused with avalanching alarms during plant transients. ASICS and ADIOS are designed to reduce the operator burden. The advances in computer software and hardware technology and also in information processing provide a good opportunity to improve the control systems and the annunciator systems of nuclear power plants or other similar process plants. It is very important to test and evaluate the performance and the function of the computer- or software-based systems like ASICS and ADIOS. The performance and the function of ASICS and ADIOS are evaluated with the real-time functional test facility and the results have shown that the developed systems are efficient and useful for operation and operator support.     

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。      

核电厂智能操纵员支持系统的设计与开发

为了减轻核电厂操纵员在处理异常或事故工况时的工作压力和心理负担，以避免误判或误操作，本研究针对核电厂反应堆一回路系统及其关键辅助系统设计并开发了一套集数据采集与存储、在线监测、故障检测与诊断、关键参数趋势预测等功能于一体的核电厂智能操纵员支持系统（NPPIOSS），仿真验证结果表明NPPIOSS能够准确检测并识别核电厂的典型故障，从而帮助操纵员准确判断核电厂状态，以减少人因失误。因此，NPPIOSS能够在核电厂故障后辅助操纵员进行后续的决策处理，达到提高核电厂运行安全性的目的。     

核电厂操纵员心理选拔系统研究

核电厂操纵员心理选拔系统研究是职业适宜性研究的一个重要方面．,通过研究找出核电厂操纵员应具有的心理胜任特质,并据此选拔适合的人员．对提高核电厂的安全性和可靠性具有重要意义。本文在分析核电厂操纵员工作和任务特点基础上．找出操纵员的工作行为特点。参考其他危险职业领域人员心理选拔的研究成果,将这些工作行为特点、重要的行为影响因素与特定的心理评价维度相对应,从认知能力、人格特质、心理健康等3个方面给出了操纵员心理选拔系统所需要的21个维度．并提出各个维度的测量方法。     

Analysis of gross error rates in operation of commercial nuclear power stations

Experience in operation of US commercial nuclear power plants is reviewed over a 25-moth period. The reports accumulated in that period on events of human error and component failure are examined to evaluate gross operator error rates. The impact of such errors on plant operation and safety is examined through the use of proper taxonomies of error, tasks and failures. Four categories of human errors are considered; namely, operator, maintenance, installation and administrative. The computed error rates are used to examine appropriate operator models for evaluation of operator reliability. Human error rates are found to be significant to a varying degree in both BWR and PWR. This emphasizes the import of considering human factors in safety and reliability analysis of nuclear systems. The results also indicate that human errors, and especially operator errors, do indeed follow the exponential reliability model.     

Integrated software safety analysis method for digital I&C systems

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: (1) software common-cause failure, (2) the interaction failure between operator and digital instrumentation and control system interface, and (3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.     

Development of the effectiveness measure for an advanced alarmsystem using signal detection theory

Since many alarms which are activated during major process deviations or accidents in nuclear power plants can result in negative effects for operators, various types of advanced alarm systems that can select important alarms for the identification of process deviation have been developed to reduce the operator's workload. However, the irrelevant selection of important alarms could distract the operator from correct identification of process deviation. Therefore, to evaluate the effectiveness of the advanced alarm system, a tradeoff between the alarm reduction rate (how many alarms are reduced?) and informativeness (how many important alarms that are conducive to identifying process deviation are provided?) of an advanced alarm system should be considered. In this paper, a new measure is proposed to evaluate the effectiveness of an advanced alarm system with regard to the identification of process deviation. Here, the effectiveness measure is the combination of informativeness measure and reduction rate, and the informativeness measure means the information processing capability performed by the advanced alarm system including wrong rejection and wrong acceptance, and it can be calculated using the signal detection theory (SDT). The effectiveness of the prototype alarm system was evaluated using the loss of coolant accident (LOCA) scenario, and the validity of the effectiveness measure was investigated from two types of the operator response, such as the identification accuracy and the operator's preference for the identification of LOCA. From the investigation, it can be concluded that the effectiveness measure is appropriately used to evaluate an advanced alarm system because the operator can perform the identification of process deviation more correctly and easily when the effectiveness score is sufficiently high     

华龙一号LPD在线监测系统误差分析

核电厂采用线功率密度（LPD）在线监测系统对电站运行的实测参数在线计算并显示堆芯线功率密度，能够准确、及时地描述堆芯状态，提高核电厂运行的安全性和经济性。LPD在线监测系统报警限值的设定，需要考虑在线监测系统的总体误差，并留有一定裕量。通过研究华龙一号LPD在线监测系统的总体不确定度的分析方法，将系统的各部分误差通过统计方法综合起来，得到系统的总体误差限值。结果表明，华龙一号采用的LPD在线监测系统误差满足工程要求。      

核反应堆故障早期检测和在线状态监测方法

即将颁布的核行业标准《核电厂反应堆堆内构件的振动监测》(简称标准)对早期监测反应堆压力容器堆内构件蜕化的方法、故障检测仪表和监测程序提出了要求,适用于以中子波动信号和反应堆压力容器振动信号为基础的堆内构件和一回路部件的动态特性的监测。本文强调了标准所包含的核电站新型仪表控制系统的监测方法。新系统不同于传统的监测系统,它的主要目的是早期故障检测．以便向电厂操纵员和检查维修人员提供有用的状态信息。