Quantum evolutionary algorithm applied to transient identification of a nuclear power plant

When transients occur during the operation of Nuclear Power Plants (NPPs), their identification is critically important for both operational and safety reasons. Thus, plant operators have to identify an event based upon the evaluation of several distinct process variables, which might difficult operators’ actions and decisions. Transient identification systems have been proposed in order to support the analysis with the aim of achieving successful or effective courses of action, as well as to reduce the time interval for a decision and corrective actions. This article presents a system for accident and transient identification in a pressurized water reactor NPP whose optimization step of the classification algorithm is based upon the paradigm of the Quantum Computing. In this case, the optimization metaheuristic Quantum Inspired Evolutionary Algorithm (QEA) was implemented and tested. The system is able to identify anomalous events related to transients of the time series of process variables related to postulated accidents. The results of the classification of transients/accidents are compared with other results in the literature.     

Transient identification in nuclear power plants: A review

A transient is defined as an event when a plant proceeds from a normal state to an abnormal state. In nuclear power plants (NPPs), recognizing the types of transients during early stages, for taking appropriate actions, is critical. Furthermore, classification of a novel transient as “don't know”, if it is not included within NPPs collected knowledge, is necessary. To fulfill these requirements, transient identification techniques as a method to recognize and to classify abnormal conditions are extensively used. The studies revealed that model-based methods are not suitable candidates for transient identification in NPPs. Hitherto, data-driven methods, especially artificial neural networks (ANN), and other soft computing techniques such as fuzzy logic, genetic algorithm (GA), particle swarm optimization (PSO), quantum evolutionary algorithm (QEA), expert systems are mostly investigated. Furthermore, other methods such as hidden Markov model (HMM), and support vector machines (SVM) are considered for transient identification in NPPs. By these modern techniques, NPPs safety, due to accidents recognition by symptoms rather than events, is improved. Transient identification is expected to become increasingly important as the next generation reactors being designed to operate for extended fuel cycles with less operators' oversight. In this paper, recent studies related to the advanced techniques for transient identification in NPPs are presented and their differences are illustrated.     

核电厂停堆状态下的安全问题

通过对核电厂运行经验反馈和停堆状态下某些事件或事故序列的分析以及国外在这方面研究的进展［１］，探讨进一步提高核电厂安全性的一些措施以及核安全监督管理方面的新的要求。     

核电厂停堆状态下的安全问题

核电厂运行经验反馈和概率安全分析表明，核电厂在停堆状态下具有相当大的堆芯熔化的潜在风险。本文叙述了核电厂运行经验反馈，概率安全分析和事故分析的结果，以及相关的措施和研究课题，特别涉及到停堆状态下的非可控硼稀释事故的维修冷停堆下推动余热排出系统。     

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.     

健康效应模型在核电站事故后果分析中的应用

健康效应模型是核电站事故后果分析中使用的重要模型之一。健康效应模型的有效性，直接影响到后果评价的可信性，从而影响到为缓解事故后果、保护公众健康而采取的应急措施的有效性。本研究在ＮＵＲＥＧ／ＣＲ４２１４模型的基础上，建立了适合于我国核电站事故后果分析的健康效应模型参数，并将该模型应用于广东大亚湾核电站的事故后果分析中，研究中选择香港居民为对象，计算和分析了大亚湾核电站的严重事故对香港居民带来的健康风险的增加。计算结果表明，该电站对香港地区带来的健康风险很低。     

Identification of nuclear power plant transients using the Particle Swarm Optimization algorithm

In order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition, transient identification systems have been devised to help operators identify possible plant transients and take fast and right corrective actions in due time. In the design of classification systems for identification of nuclear power plants transients, several artificial intelligence techniques, involving expert systems, neuro-fuzzy and genetic algorithms have been used. In this work we explore the ability of the Particle Swarm Optimization algorithm (PSO) as a tool for optimizing a distance-based discrimination transient classification method, giving also an innovative solution for searching the best set of prototypes for identification of transients. The Particle Swarm Optimization algorithm was successfully applied to the optimization of a nuclear power plant transient identification problem. Comparing the PSO to similar methods found in literature it has shown better results.     

Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code

Incorporation of full three-dimensional models of the reactor core into system thermal–hydraulic transient codes allows better estimation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations to verify and validate the capability of the so-called coupled codes technique. For these purposes appropriate Light Water Reactor (LWR) transient benchmarks based upon programmed transients performed in Nuclear Power Plants (NPP) were recently developed on a higher ‘best-estimate’ level. The reference problem considered in the current framework is a Main Coolant Pump (MCP) switching-on transient in a VVER1000 NPP. This event is characterized by a positive reactivity addition as consequence of the increase of the core flow. In the current study the coupled RELAP5/PARCS code is used to reproduce the considered test. Results of calculation were assessed against experimental data and also through the code-to-code comparison.     

Approach to accident management in RBMK-1500

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK.Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed.This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account.     

A procedure to optimize the timing of operator actions of accident management procedures

The analysis of beyond design basis accidents (BDBA) is an essential component of the safety concept of nuclear power plants (NPP). Goal of the analysis is to achieve a set of actions aimed to prevent the escalation into a severe accident, to mitigate consequences of a severe accident, and to achieve a long term controllable state of the NPP. This paper presents an analytical procedure to optimize the timing of operator interventions. The procedure is demonstrated based on four sets of parameters, first, parameters which define the operator actions are chosen. Second, parameters which define the system availability are chosen. Third, parameters which define in a continuous way the status of the plant are chosen. Finally, one looks for a functional dependency of the accident management (AM)-parameters and the parameters describing the plant status. Once a function could be found, this function is “optimized” in the sense that the AM-parameters are varied to find a optimal overall condition for the plant. In the first part, the paper presents the analytical procedure in a general way, in the second part, an initiating event is chosen. The procedure is applied to a station black out (SBO) transient, and as operator action secondary side bleed and feed, followed by primary side bleed and feed, is foreseen. As result, the optimal timing to initiate both actions is achieved.     

Applications of fault detection and diagnosis methods in nuclear power plants: A review

Nuclear power industries have increasing interest in using fault detection and diagnosis (FDD) methods to improve safety, reliability, and availability of nuclear power plants (NPP). A brief overview of FDD methods is presented in this paper. FDD methods are classified into model-based methods, data-driven methods, and signal-based methods. While practical applications of model-based methods are very limited, various data-driven methods and signal-based methods have been applied for monitoring key subsystems in NPPs. In this paper, six areas of such applications are considered. They are: instrument calibration monitoring, instrumentation channel dynamic performance monitoring, equipment monitoring, reactor core monitoring, loose part monitoring, and transient identification. The principles of using FDD methods in these applications are explained and recent studies of advanced FDD methods are examined. Popularity of FDD applications in NPPs will continuously increase as FDD theories advance and the safety and reliability requirement for NPP tightens     

基于模糊熵的核电站瞬态识别方法

为保障核电站安全经济运行,需及时准确地识别核电站出现的异常。本文通过处理关键变量的时间序列数据,对瞬态过程进行识别：利用模糊熵度量时间序列复杂度的能力,判断系统是否处于正常状态;进而利用互模糊熵度量两时间序列相似度的能力,对出现的瞬态进行类型识别。利用模块式高温气冷堆核电站仿真机的数据验证了本方法的可行性和有效性,结果表明本文方法可有效进行瞬态识别,且不需复杂的训练过程。     

福岛核事故后国内外核电厂应急准备的改进研究

福岛核事故3年后,国内外大量经验反馈指出,为了确保核电厂的安全运行,并在事故后尽量减少放射性物质释放,降低事故引发的人员伤亡和财产损失,对目前的应急准备进行改进是十分必要的。本文首先详细研究了国内外针对福岛后的应急准备改进要求,其次对我国3个典型核电厂对《福岛核事故后核电厂改进行动通用技术要求（试行）》的落实情况进行调研,最后初步归纳了《应急准备改进的技术要求》关键点。本调研报告将为技术要求的最终制定提供重要依据,并可为国家核安全局及各核电业主决策提供参考。     

RALOC4 code validation against measured data at Ignalina NPP during single Main Safety Valve opening

Although RALOC4 code is validated against many experiments with regard to Western Nuclear Power Plants (NPPs) the code validation problem for the Accident Localization System (ALS) of Ignalina NPP modeling is of special importance because the condensing pools at NPP with RBMK-1500 differ from the pressure suppression systems installed in NPPs with German BWR. The response of Ignalina NPP ALS to the unintentional opening of single Main Safety Valve, which occurred in 1998, is analyzed by employing code RALOC4. The results of post-event calculations compared with the measured data available after the event. The performed analysis showed that RALOC4 code could be applied for the simulation of Ignalina NPP ALS. Nevertheless, the spray modeling in RALOC4 should be improved allowing the simulation of sprays in NON_EQUILIBRIUM zone model and to consider the diameter of water droplet diameter and height of droplet fall.     

Development of a standard communication protocol for an emergency situation management in nuclear power plants

Correct communication between main control room (MCR) operators is an important factor in the management of emergency situations in nuclear power plants (NPPs). For this reason, a standard communication protocol for the management of emergency situations in NPPs has been developed, with the basic direction of enhancing the safety of NPPs and the standardization of communication protocols. To validate the newly developed standard communication protocol, validation experiments with 10 licensed NPP MCR operator teams was performed. From the validation experiments, it was found that the use of the standard communication protocol required more time, but it can contribute to the enhancement of the safety of NPPs by an operators’ better grasp of the safety-related parameters and a more efficient and clearer communication between NPP operators, while imposing little additional workloads on the NPP MCR operators. The standard communication protocol is expected to be used to train existing NPP MCR operators without much aversion, as well as new operators.     

Application and In-depth Assessment of RETRAN-3D for Best Estimate Analysis of Nuclear Power Plant Transients

The code RETRAN-3D has been extensively applied in the project STARS at PSI to perform BE analysis for a range of operational and other (non-LOCA) transients for the Swiss NPPs, which include both PWRs and BWRs. Extensive assessment employing experimental and plant data has provided confidence in the applicability and accuracy of RETRAN-3D for the analysis of transients in both types of LWRs. In this context, this paper presents an in-depth study of the performance of the code for two types of applications. First, a detailed analysis of BWR/6 recirculation pump trip tests shows that the code is able to accurately predict the coupled neutronic and thermal-hydraulic behaviour during an operational transient in which plant-specific system features (e.g. SRI insertion, response of the recirculation line valves to changes in flow, etc.) play an important role. Second, RETRAN-3D is applied to BE analysis of a PWR beyond-design basis scenario, namely the failure of the RHR system during reactor shutdown. In-depth assessment of the results obtained demonstrates the applicability of RETRAN-3D and the need for employing a BE approach to predict a more accurate, shorter “time window” for operator remedial action than that estimated using the assumption of uniform primary temperature distribution.     

核电厂操纵员职业适宜性研究

文章涉及核电厂操纵员的工作任务分析和核电厂事故中相关人误分析的结果。从人机工程学原理出发，提出了包括认知、人格特质、心理健康三个方面的核电厂操纵员职业适宜性心理测评系统，并讨论了该系统的应用方式以及重要性。     

Developing and modeling of the “Laguna Verde” BWR CRDA benchmark

Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant — unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The “Laguna Verde” (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTRÉE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions.     

海上浮动核电站应急撤离分析

核事故应急撤离是核应急响应的重要组成部分, 目的在于快速有效地将可能受到事故影响的人员转移至安全地区。本文根据海上浮动核电站的运行场址与运行特点, 对海上浮动核电站应急响应特征进行分析, 给出了浮动核电站应急等级划分和应急计划区范围。结合陆地核电站场区撤离与海洋平台撤离疏散方法, 制定了海上浮动核电站应急撤离情景与撤离分析假设。对浮动核电站人员撤离的分析结果表明, 浮动核电站人员撤离满足客船撤离要求, 及海上浮动核电站应急撤离的时间要求。关键词: 海上浮动核电站; 核应急; 应急计划区;应急响应; 应急撤离