Application of fuzzy decision making method to the evaluation of spent fuel storage options

In this paper, a fuzzy decision making method is proposed which is based on the fuzzy set theory and the hierarchical structure analysis. The method employs two key concepts: linguistic variables and fuzzy numbers. The linguistic variables and fuzzy numbers are used to represent the decision-maker's subjective assessments for the decision criteria and the decision alternatives versus the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments and the total integral value method is used to rank the decision alternatives. As a case study, the method is applied to evaluating the spent fuel storage options in Korea.     

灰色理论在操纵员管理中的应用

科学合理的操纵员管理是核安全的基础,也是核电厂人力资源管理的发展要求。尤其在三哩岛事故发生之后,操纵员管理更加受到重视。科学合理的操纵员管理的前提和基础是操纵员基本行为参数的预测和决策。灰色理论恰好解决了现行操纵员管理预测和决策中所遇到的困境。整个管理过程分为两步：首先,通过以往的操纵员行为参数的记录,建立灰色预测模型,对其今后行为进行预测;然后,使用灰色决策对其进行决策。计算结果对操纵员管理有一定的指导意义,也为操纵员提高自身素质指明了方向。灰色理论方法为今后操纵员管理提供了新的思路和方法。     

Improved TOPSIS decision model for NPP emergencies

In this paper,an improved decision model is developed for its use as a tool to respond to emergencies at nuclear power plants.Given the complexity of multi-attribute emergency decision-making on nuclear accident,the improved TOPSIS method is used to build a decision-making model that integrates subjective weight and objective weight of each evaluation index.A comparison between the results of this new model and two traditional methods of fuzzy hierarchy analysis method and weighted analysis method demonstrates that the improved TOPSIS model has a better evaluation effect.     

An extended value-impact approach for nuclear regulatory decision making

This paper presents an extended value-impact methodology which aids decision makers in ranking various alternative actions for reducing the risk associated with nuclear power reactors. It extends the state-of-the-art value-impact methodology by using the Analytic Hierarchy Process (AHP), a formalized decision making tool for ranking alternatives based on judgment. The AHP reduces some of the limitations present in current value-impact work, such as the inability to include subjective factors in a structured approach, as well as controversial questions such as the importance of onsite versus offsite accident costs averted, uncertainty, and impact of public opinion.In the paper, the method is applied to include a value-impact study of the implementation of either a vented-containment system or an alternative decay heat removal system as a means for reducing risk at the Grand Gulf nuclear power plant. The results of this analysis show that the method provides considerable insight to the solution of topics of interest in the decision making area of nuclear power risk management.     

SAMEX: A severe accident management support expert

A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R&D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.     

核电站事故应急模糊层次决策模型及应用

针对核事故应急决策的复杂性,应用层次分析和模糊决策理论,对核事故应急决策的多个目标进行综合分析,建立了核电站事故应急方案优化模型,将该模型应用于实际事故应急决策中,提高了决策的合理性和可操作性,     

核电厂严重事故缓解进程中应急人员行为分析

为了分析核电厂人员处理严重事故的行为特征，本文通过研究严重事故管理导则的特殊性，结合现场调研和操纵员、应急技术支持人员访谈，建立了严重事故缓解进程中的人员决策模型，识别了决策人员、执行人员的关键影响因子，为严重事故下的人因可靠性分析方法研究奠定基础。      

Modeling and measuring the effects of imprecision in accident management

This paper presents two approaches for evaluating the uncertainties inherent in accident management strategies. Current PRA methodology uses expert opinion in the assessment of rare event probabilities. The problem is that these probabilities may be difficult to estimate even though reasonable engineering judgement is applied. This occurs because expert opinion under incomplete knowledge and limited data is inherently imprecise. In this case, the concept of uncertainty about a probability value is both intuitively appealing and potentially useful. This analysis considers accident management as a decision problem (i.e. “applying a strategy” vs. “do nothing”) and uses an influence diagram. Then, the analysis applies two approaches to evaluating imprecise node probabilities in the influence diagram: “a fuzzy probability” and “an interval-valued subjective probability”. For the propagation of subjective probabilities, the analysis uses a Monte-Carlo simulation approach. In case of fuzzy probabilities, fuzzy logic is applied to propagate them. We believe that these approaches can allow us to better understand uncertainties associated with severe accident management strategies, because they provide additional information regarding the implications of using imprecise input data.     

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。     

二级PSA在严重事故管理中的应用研究

Level 2 Probabilistic Safety Analysis (PSA) can be used to quantitatively assess the risk of severe accident and is a good tool to evaluate the severe accident management. By studying the general method and procedure for the application of level 2 PSA in severe accident management, taking an improved generation-Ⅱnuclear power plant as an example, the “primary loop depressurization operation ” and the “ primary loop emergency water injection” in severe accident management guideline are quantitatively evaluated. Analysis shows that performing the “primary loop depressurization operation” immediately after entering the severe accident management guideline can greatly reduce the risk of large radioactive release, and performing “primary loop emergency water injection operation” contributes greatly to reducing the risk of large radioactive release in the slower accident sequence. The study shows that there still has further improvement room in severe accidents management for nuclear power plants in China.     

二级PSA在严重事故管理中的应用研究

二级概率安全分析(PSA)可用来定量评估严重事故风险,是评价严重事故管理的良好工具。通过研究二级PSA应用于严重事故管理的一般方法与流程,以某二代改进型核电厂二级PSA模型为例,对严重事故管理导则中"一回路卸压"和"一回路应急注水"两个关键操作进行了定量评价。评价表明进入严重事故管理导则后立即执行"一回路卸压操作"可大幅度降低大量放射性释放风险,执行"一回路应急注水操作"对于降低进程较慢的事故序列大量放射性释放风险贡献较大。研究表明国内核电厂针对严重事故的管理还有进一步提升空间。。     

Prediction of major transient scenarios for severe accidents of nuclear power plants

It is very difficult for nuclear power plant operators to predict and identify the major severe accident scenarios following an initiating event by staring at temporal trends of important parameters. In this regard, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to classify accidents into groups of initiating events such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Also, a fuzzy neural network (FNN) is designed to identify their major severe accident scenarios after the initiating events. The inputs to PNN and FNN are initial time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. In cases that an initiating event develops into a severe accident, this may happen when plant operators do not follow the appropriate accident management guidance or plant safety systems do not work, the proposed algorithm showed accurate classification of initiating events. Also, it well predicted timings for important occurrences during severe accident progression scenarios, which is very helpful to perform severe accident management.     

基于RS-FNN的核电厂设备智能故障诊断方法的研究

将粗糙集(RS)理论与模糊神经网络(FNN)相结合,能充分发挥各自的优点.本文利用RS方法对知识的约简技术,从大量的原始数据中提取精简的规则,基于这些规则建立的FNN网络具有更好的拓扑结构,学习速度大大提高、判断准确、容错能力强,具有更高的实用价值.为了验证该方法的有效性,以核电厂设备蒸汽发生器U形管破裂等故障为例,进行了仿真实验研究.诊断结果表明,将基于RS理论的FNN智能故障诊断方法引入核电厂设备故障诊断中是可行的,并且具有简单方便、计算量小、诊断结果可靠等特点.     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.     

Filtered vented containments

Adding a filtered vented containment system (FVCS) to an existing nuclear power plant has been suggested as one approach to mitigating the effects of a severe accident. The integration of a new system into a plant appears simple on the surface but may be expensive to install and complex in developing operating procedures and restrictions. A number of designs have been proposed, and many are currently being installed at European nuclear facilities. Risk assessments and cost benefits for an FVCS installation in a typical US plant are discussed. An approach to developing an FVCS and strategies for system operation (venting) are also included, along with a conceptual design of a sand/gravel filter system. Estimated installed costs are included.     

VVER机组LBLOCA始发严重事故工况下堆芯出口温度BEPU分析

最佳估算加不确定性（BEPU）方法目前广泛应用于核电厂设计基准事故（DBA）的分析。考虑到严重事故现象复杂及不确定性较大,BEPU方法在严重事故领域应用较少。堆芯出口温度（CET）是核电厂安全运行的重要监测参数,本文以VVER1000压水堆核电厂为研究对象,采用BEPU方法对大破口失水事故（LBLOCA）始发严重事故工况下包壳破裂对应的CET进行不确定性分析,并对输入参数进行敏感性分析。计算结果表明：气隙释放对应CET的单侧统计容忍限值（95/95）为430.85 ℃;CET对输入参数中的衰变热系数和包壳厚度较为敏感。     

Emergency management for a nuclear power plant using fuzzy cognitive maps

This paper explores the application of fuzzy cognitive maps (FCM) to emergency operating procedures (EOPS), to represent the decision-making process during abnormal situations in a nuclear power plant (NPP). The decision-making process in a NPP is a complex process, due to the many elements involved in its operation, and the permanent attention demanded by its maintenance. At the present time, the decision making process in a NPP is analyzed and developed by reactor operators, based on a set of instructions as well as flow charts to mitigate the consequences of a broad range of transients, accidents and multiple equipment failures, whose main characteristic is to be linear representations of events within a scenario. One of the main objectives of this paper is to present a method based in FCM that could be applied in the development of EOPS, and show some simulations, specifically the loss of coolant accident (LOCA) scenario in a boiling water reactor (BWR) with the Mark II containment design was studied. The FCM-based method represents with high fidelity the expert reasoning (the human expert is very important) and the interpretation of the results aids instantly to the reactor operators in the surveillance of the reactor proper functionality due that they have the responsibility of the decision taking in emergency situations. The simulations results show that the FCM predict properly the phenomenon in the reactor vessel and primary containment.     

Classification method of severe accident condition for the development of severe accident instrumentation and monitoring system in nuclear power plant

Instrumentation and monitoring systems in a nuclear power plant are very important to monitor plant conditions for safe operations and a plant shutdown. The severe accident at TOKYO ELECTRIC POWER COMPANY's Fukushima Daiichi Nuclear Power Station (hereinafter called as TF1) in March 2011 caused several severe situations such as core damage, hydrogen explosion, etc. Lessons learned from the severe accident at TF1 show that an appropriate operable instrumentation and monitoring system for a severe accident should be developed so that the system will deliver an appropriate performance for mitigation of severe accident condition in a nuclear power plant.

This paper proposes the classification method of severe accident condition for the development of an appropriate operable instrumentation and monitoring system for a severe accident based on the problem analysis of monitoring variables during the severe accident at TF1. The classification is formed on the basis of the integrity of boundary for plant safety and the successful (or unsuccessful) condition of the cooling water injection, and is used for an establishment of defining severe accident environmental conditions for the instrumentation and monitoring system. Examples of the establishment method are also shown in this paper.     

核电厂严重事故情况下氢气控制分析

文章首先阐述了核电厂严重事故情况下安全壳内的氢气风险,研究现状,以及缓解、控制氢气风险的具体措施.在此基础上,介绍了田湾核电站严重事故情况下氢气控制的系统和方法,调试结果及历次大修对氢气控制系统的检查结果,表明该方法具备严重事故预防和缓解能力,安全风险处于受控状态,安全是有保障的,符合国家核安全局针对福岛核事故后对核电厂改进行动的通用技术要求.     

基于模糊神经网络的核动力装置设备故障诊断系统研究

将模糊逻辑与神经网络相结合,对模糊逻辑和神经网络的区别和联系进行了阐述,并探讨了它们之间的结合．模糊神经网络结构及其实现算法,并将这一理论应用于核动力装置故障诊断,建立了基于模糊神经网络的船用核动力装置故障诊断系统。为了验证该系统的有效性,以蒸汽发生器U形管破裂事故为例,进行了仿真实验研究。诊断结果表明该理论方法对此事故完全可以正确识别。