Development and analysis of LOCA sequences for severe accident risk database

Although a Level 2 PSA has been performed for the Korean Standard Power Plants (KSNPs), and that it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access of the results. At present, KAERI is intensively calculating the severe accident sequences for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviors. The developed database (DB) system includes a graphical display for a plant and equipment status, previous research results by a knowledge-based technique, and the expected plant behavior. The plant model used in this paper is oriented to the cases of LOCAs related to severe accident phenomena and thus can simulate the plant behaviors of a severe accident. Therefore, the developed system may play a central role as an information's source during the decision-making for a severe accident management, and be used as a training simulator for a severe accident management.     

Plant Operator Performance Evaluation System

A plant operator performance evaluation system to analyze plant operation records during accident training and to identify and classify operator errors has been developed for the purpose of supporting realization of a training and education system for plant operators. A knowledge engineering technique was applied to evaluation of operator behavior by both event-based and symptom-based procedures, in various situations including event transition due to multiple failures or operational errors The system classifies the identified errors as to their single and double types based on Swain's error classification and the error levels reflecting Rasmussen's cognitive level, and it also evaluates the effect of errors on plant state and then classifies error influence, using “knowledge for phenomena and operations”, as represented by frames. It has additional functions for analysis of error statistics and knowledge acquisition support of “knowledge for operations”.

The system was applied to a training analysis for a scram event in a BWR plant, and its error analysis function was confirmed to be effective by operational experts.     

Development of a risk-informed accident diagnosis and prognosis system to support severe accident management

RISARD, risk-informed severe accident risk diagnosis system, is a computerized tool developed to improve a severe accident management (SAM) for a nuclear power plant and to effectively support the MCR and the TSC in executing the relevant SAM activities. In order to provide a diagnostic capability to a state of the plant and a prognostic capability for an anticipated accident progression, the system examines (a) a symptom-based diagnosis of a plant damage state (PDS) sequence in a risk-informing way and (b) a PDS sequence-based prognosis of key plant parameter behavior, through a prepared database (DB) containing plant-specific severe accident risk (SAR)-related information. For a given accident, the replicated use of these two processes makes it possible to obtain information about the functional states of the plant and containment safety systems expected at the time of a severe accident as well as future trend of the key plant parameters that are essentially required for taking the relevant SAM action, eventually leading to an answer about the best strategy for SAM. The foregoing concept for an accident diagnosis and prognosis can give the SAM practitioners more time to take action for mitigating the consequences of the potential accident scenarios since they are made in a simple, fast, and efficient way through a prepared SAR database and it is useful especially when the plant information available for SAM is incomplete and limited. The main purpose of this paper is to (a) introduce the concept of the RISARD system proposed to support SAM and its implementation through a prepared OPR1000 plant- and MAAP code-specific SAR database and (b) assess prediction capabilities of major events expected during the evolution of a severe accident through the system.     

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.     

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences.The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA.A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach.This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR,which is the representative case for high pressure sequences.MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident.In addition,a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.     

Two-phase flow degradation on Fukushima-Daiichi Unit 2 RCIC turbine performance

After the Fukushima accident, several investigation reports, including experiments and simulations have been done for each of the affected units to completely understand the accident progression and use their results to improve the knowledge of severe accident management and the severe codes performance. In Unit 2, the major uncertainties are related with the reactor core isolation cooling (RCIC) system performance during the accident progression especially focused in the RCIC turbine, which is assumed to work in two-phase flow. The main objective of this study is to analyze the RCIC turbine performance under two-phase flow scenarios under the assumption that the power produced by the turbine is lower than expected due to the liquid phase in the flow. A degradation coefficient quantifying the turbine power reduction is developed as a function of the flow quality by using the sonic speed reduction at critical flow conditions principle obtained by applying the non-homogeneous equilibrium model (NHEM). The degradation coefficient was applied to RELAP/ScdapSIM severe accident code showing a drastic reduction of the turbine-generated power during two-phase flow and obtaining a RCIC system behavior closer to the Tokyo electric power company (TEPCO) investigation report conclusions.     

核电站培训模拟器的综合利用

核电站培训模拟器是一个高投资设备,同时它的利用潜力又远超过一个核电站操作员培训的需要。因而,扩大人员培训范围和开辟培训之外的其它用途以提高核电站模拟器的经济效益,是一个很值得考虑的重要问题。本文将介绍几个利用核电站模拟器作为实验工具开展核电安全科研的例子,其中包括:①人因工程的研究,②安全仪表的检验,③应急规程的验证。     

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

A five-step methodology was developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information that personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment.     

Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its’ own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.     

Simulator performance upgrade and evaluation of loss of RHR during midloop operation for Maanshan PWR plant

Because a greater risk than expected is introduced during midloop operation for typical PWRs, the performance of training simulator of Taiwan’s Maanshan 3-loop PWR plant was verified and upgraded for midloop operation (MLO) simulation. Besides, plant specific midloop abnormal operation procedure (AOP) also was quantitatively evaluated. Instead of modifying existing RCS module, a thermal-hydraulic code, namely ROSE (Reactor Outage Simulation and Evaluation) has been developed and transplanted into the training simulator of Maanshan PWR plant. A two-region approach with a modified two-fluid model was adopted as the theoretical basis of the ROSE code. The success of the simulator performance upgrade for the MLO was demonstrated by comparisons to W-GOTHIC MLO calculation as well as the original simulator performance. Moreover, regarding the evaluation of the associated AOP for MLO after loss of RHR, the most risky plant configuration as well as associated crucial timing before core uncovery was also identified by the upgraded training simulator.     

核电站操纵人员模拟机培训与取照考试

"大亚湾核电操纵人员模拟机培训与执照考核体系"针对模拟机教员的配备、培训、岗位资格考核、教学效果反馈等环节制定了管理方法,对教员提出了专业知识、工作经验、外语能力等方面的具体要求.设计了教员培养流程,培训项目包括:培训者培训、教学法培训、时间管理培训、机组大修期间运行活动培训、模拟机培训课程的"影子"培训和在岗培训等.实现了模拟机教员与现场持照人员的岗位轮换;新教员须经管理部门进行资格确认和审查,符合条件者实行聘任上岗制度,任期2年.任期届满后,需重新对其进行资格认定.在引进法国电力公司操纵人员培训办法的基础上,开发了新课程,对模拟机初训、复训以及利用模拟机进行技术支持与经验反馈方面进行了改进.     

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.     

Accident management strategy in Sweden — implementation and verification

A comprehensive program for severe accident mitigation was completed in Sweden by the end of 1988. As described in this paper, this program included plant modifications such as the introduction of filtered containment venting, and an accident management system comprising emergency operating strategies and procedures, training and emergency drills. The accident management system at Vattenfall has been further developed since 1988 and some results and experience from this development are reported in this paper. The main aspects covered concern the emergency organization and the supporting tools developed for use by the emergency response teams, the radiological implications such as accessibility to various locations and the long-term aspects of accident management.     

压水堆核电站严重事故紧凑型仿真机开发

为了缓解压水堆核电站可能发生的严重事故的后果,也为了满足安全分析工程师和概率风险评价人员的需求,并在与国际原子能机构合作框架协议内,研制开发了紧凑型的严重事故仿真分析机MELSIM－PC。该仿真系统主要由仿真核心程序、同步通讯程序、人机界面程序等几个部分组成,可以工作在一台普通的微型计算机上,成功地实现MELCOR程序变量的运行数据库管理、电站动态图形显示、仿真计算控制、再启动和仿真重演等重要功能。     

数字化核电厂主控室操纵员班组行为研究

核电厂主控室数字化后将引起班组交流合作等一些列变化,但以往针对班组的研究较少且主要以经验研究为主,较少有实证研究,以致在进行人因可靠性分析（HRA）时难以恰当引用或修正行为形成因子。本文使用行为分析软件以模拟机培训中一个班组为研究对象,首次从行为分析的角度研究了数字化后主控室班组行为特征和组织结构的有效性。对班组成员在一次事故培训中的交流行为进行了统计分析,研究结果表明,数字化后操纵员班组为一个分工明确、层次清晰的组织结构。      

Behavioral simulation of a nuclear power plant operator crew for human–machine system design

This article proposes an architecture of behavioral simulation of an operator crew in a nuclear power plant including group processes and interactions between the operators and their working environment. An operator model was constructed based on the conceptual human information processor and then substantiated as a knowledge-based system with multiple sets of knowledge base and blackboard, each of which represents an individual operator. From a trade-off between reality and practicality, we adopted an architecture of simulation that consists of the operator, plant and environment models in order to consider operator–environment interactions. The simulation system developed on this framework and called OCCS was tested using a scenario of BWR plant operation. The case study showed that operator–environment interactions have significant effects on operator crew performance and that they should be considered properly for simulating behavior of human–machine systems. The proposed architecture contributed to more realistic simulation in comparison with an experimental result, and a good prospect has been obtained that computer simulation of an operator crew is feasible and useful for human–machine system design.     

核电厂数字化主控室操纵员界面管理任务特征的研究

首先采用系统性人因失误减少和预测方法（SHERPA）分析操纵员界面管理任务中的关键行为；再用行为分析软件（INTERACT9）分析国内某数字化核电厂全范围模拟机上操纵员的操作录像视频，之后对INTERACT9采集的关键行为数据进行统计分析，得到4个操纵员界面管理任务的一般特征:①一、二回路操作员操作菜单栏、选择监视目标和打开参数界面的频率最高；②操纵员在选择进入不同系统界面的方式上趋向于选择从菜单栏进入；③一回路操纵员在正常工况和事故工况下的界面管理任务没有明显差异；二回路操纵员在正常工况下的界面管理任务明显少于事故工况；④正常工况下，一回路操纵员的界面管理任务显著多于二回路操纵员；事故工况下，一回路操纵员的界面管理任务与二回路操纵员的的界面管理任务相当。      

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。     

核电工作人员模拟机培训和考核方法探讨

核电厂操作员模拟机培训和考核，是核电厂运行安全中极为重要的问题，本文介绍了清华大学核电厂模拟培训中心自１９８８年建成以来举办的各种类型的核电工作人员培训班，并详细分析了各类培训班，老操作员在职再培训班和各种类型核电管理人员培训班等的不同特点，详细介绍了清华大家核电厂模拟培训中心对不同培训班的课程设置、考核方法和评分标准等问题。     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...