Analysis of pressurizer level control system using MAAP4 code

The code initialization effort has been troubling code users for decades for system transient and severe accident analyses using codes such as RETRAN, MAAP4, MAAP5 and MELCOR. The purpose of this work is to demonstrate an approach that could be considered a generic method to address the code initialization problem. This was demonstrated by developing a pressurizer level control model and temperature dependent level control logic in MAAP4 without re-compiling with the source code. The method would enhance the simulation capability and accuracy of a severe accident analysis by transient and severe accident analyses codes. The demonstration case used MAAP4 to show that the adopted proportional-integral controller with the temperature dependent level control logic would reduce its code steady state errors to zero. The subsequent transient response would become more realistic. The proposed method provides a convenient and exemplified approach for code initialization which is applicable to the next generation of codes that couple with the balance of plant models. These codes include the MAAP5 code and others future codes that could simulate the whole plant by a single and elaborate plant model with exhausting component and phenomenological models.     

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.     

Insights and comparisons of the level-2 results of recent probabilistic safety analyses

Observations and insights based on the review of a number of recent level-2 probabilistic safety analyses (PSAs) and individual plant examinations (IPEs) are provided. Observations and comparisons are made regarding plant and containment design characteristics, methods for the analysis of containment response to severe accident loads, modeling of the uncertain phenomenological processes impacting containment response, accident progression and containment analysis, source term calculation, and uncertainty analysis. Insights are obtained which attempt to relate the various plant and containment design characteristics to expected containment performance, though these relationships can often be obscured by the large inherent uncertainties associated with quantification of most level-2 PSA issues.     

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。     

An integrated approach to source term uncertainty and sensitivity analyses for nuclear reactor severe accidents

Large-scale computer programs simulate severe accident phenomena and often have a moderate-to-large number of models and input variables. Analytical solutions to uncertainty distributions of interested source terms are impractical, and influential inputs on outputs are hard to discover. Runs of such integral codes for complex severe accidents are generally time-consuming and hence computationally expensive. This article presents an integrated approach to uncertainty and sensitivity analyses for nuclear reactor severe accident source terms, with an example which simulates an accident sequence similar to that occurred at Unit 2 of the Fukushima Daiichi Nuclear Power Plant using an integral code, MELCOR. Monte-Carlo-based uncertainty analysis has been elaborated to investigate the released fractions of representative radionuclides, Cs and CsI. In order to estimate the sensitivity of inputs, which have a substantial influence on the core melt progression and the transportation process of radionuclides, a variance decomposition method is applied. Stochastic process, specifically a Dirichlet process, is applied to construct a surrogate model in sensitivity analysis as a substitute of the code. The surrogate model is cross-validated by comparing with corresponding results of MELCOR. The analysis with the simpler model avoids laborious computational cost/load, so that the importance measures for input factors are obtained successfully.     

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。     

Level-1, -2 and -3 PSA for AHWR

In nuclear reactor probabilistic safety analyses (PSAs), risk is usually defined by the frequency and magnitude of radioactive releases to the environment (Generic CANDU, 2002). An integrated Level-1, -2 and -3 PSA have been carried out for thorium based natural circulation driven advanced heavy water reactor (AHWR). A Level-1 PSA models accident sequences up to the point at which the reactor core either reaches a stable condition or becomes severely damaged, releasing large amounts of radionuclides into the containment. The probabilistic aspects of the analysis focus on the performance and reliability of nuclear plant systems and station staff in response to plant upsets. A Level-2 PSA examines severe reactor accidents through a combination of probabilistic and deterministic approaches, in order to determine the release of radionuclides from containment, including the physical processes that are involved in the loss of structural integrity of the reactor core (Generic CANDU, 2002). A Level-3 PSA goes through the short and long term (radiological) effects on the public (Fullwood, 2000). In this study the risk associated with internal events is only addressed. In the first phase, Level-1 PSA has been carried out to identify postulated initiating events (PIEs) which may lead to severe core damage (SCD) for the reactor. In the second phase, a Level-2 PSA examines two enveloping severe accidents through a combination of probabilistic and deterministic approaches and determines the release of radionuclides from containment. In the third phase, a Level-3 PSA is carried out for the transport of radionuclides through the environment and for the evaluation of public health risk for the two scenarios considered. The salient findings are presented in the paper.     

动态可靠性评价方法在AP1000核电厂严重事故中的应用研究

动态可靠性评价方法能模拟系统状态发生连续或多重变化的情况,是核电厂概率安全研究的一个新发展点。本文利用动态可靠性评价方法,使用严重事故程序MAAP对AP1000核电厂全厂断电事故进行分析,并将动态可靠性评价结果应用于二级概率安全评价（PSA）分析,最终评价对放射性裂变产物的影响。研究结果表明,系统动态特性对核电厂PSA的分析结果有一定影响,且动态可靠性评价过程可挖掘更多信息,有利于更好地指导核电厂设计及提高核电厂的安全性。     

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.     

Generic CANDU 6 plant severe accident analysis employing SCAPSIM/RELAP5 code

The purpose of the present study is to assess the capability of SCDAPSIM/RELAP5 to perform the deterministic analysis for postulated severe accidents for CANDU plant and to gain information for potential improvements in code modelling. SCDAPSIM/RELAP5 is a widespread and detailed computer code for severe accident analysis that can be adapted to benchmark the CANDU dedicated tools, MAAP4–CANDU and ISAAC. Simulations of station blackout (SBO) and large loss-of-coolant accident (LOCA) scenarios, which, through further system failures, may eventually lead to severe core damage (SCD) accident in a CANDU 6, are presented. The paper provides details concerning the methodology and nodalization used, and interprets the results obtained. Comparisons of the SCDAPSIM/RELAP5 simulations with the MAAP4–CANDU code reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAPSIM/RELAP5 and MAAP4–CANDU code predictions.     

严重事故条件下堆芯升温模拟

使用FLUENT计算流体程序数值模拟了AP1000在严重事故条件下的堆芯升温过程,目的是对堆芯裸露后并在其显著熔化前对堆芯升温的均匀程度进行比一体化事故程序MAAP更为详尽的研究,进行围筒和吊篮温度分析,同时评估MAAP程序堆芯升温计算结果。分析结果表明：在堆芯显著熔化时刻,堆芯围筒和吊篮已熔化,因此熔融堆芯将从侧面迁移进入下封头,同时对比证明MAAP程序关于堆芯升温的计算结果也是可接受的。     

大破口触发的严重事故分析及缓解措施研究

采用严重事故一体化分析程序MELCOR,对国产先进压水堆核电厂进行系统建模,选取大破口触发的严重事故进行校核计算研究,获得了严重事故工况下核电厂关键参数的瞬态特性和非能动系统响应特性,并与安全分析报告中MAAP的计算结果进行了对比分析。结果表明:虽然校核计算结果与安全分析报告中的结果存在一定差异,但总体上事故序列和主要参数的变化趋势吻合良好,并且都能够在严重事故情况下保持压力容器和安全壳的完整性,放射性裂变产物释放量极低,缓解措施的设计能够有效缓解事故进程,满足核电厂的安全要求。     

二级PSA中人员可靠性分析方法研究

在概率安全分析(PSA)中,人员可靠性分析(HRA)是必不可少的组成部分。国内在一级PSA中的HRA做了大量的研究工作,已有良好的基础和工程实践,但由于核电厂严重事故下人员响应的复杂性,有关二级PSA的HRA还处于摸索阶段。通过研究二级PSA中人员响应特点,调研国内外在二级PSA中采用的HRA方法,最后以我国某三代压水堆核电厂严重事故下一回路快速卸压为例,采用THERP、HCR+THERP以及SPAR-H三种方法,分别进行了HRA,并给出相应的结论和建议。     

大亚湾核电厂全厂"断电"事故裂变产物行为计算

使用 MELCOR 程序模拟大亚湾核电厂假想全厂断电事故早期进程,计算出安全壳内源项的最大存量,同KORIGEN 程序结合推导出安全壳内主要裂变产物的活性,为核电厂PSA 分析提供保守性数据.     

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...     

Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO

Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.     

Development of linked analysis system for PWR severe accidents utilizing RELAP5 and MAAP4

The purpose of this study is to develop a severe accident (SA) analysis method that is more reliable thorough transferring the physical status of the plant predicted by RELAP5 computer code to MAAP4 computer code. The methodology of the linkage analysis is developed and the criterion of linkage time is suggested to utilize the RELAP5 thermal–hydraulic calculation to the maximum degree possible and thereby guarantee the continuity of calculation for hydrogen generation. The MAAP4 calculations after data transfer show the physically proper results based on RELAP5 data. Comparison with other code results for TMI-2 accident reveals that the result from the RELAP5–MAAP4 linked analysis lay in the span given by a number of results of TMI calculation from other SA code systems. The results of this study are expected to improve the SA analysis methodology by analyzing an SA scenario with more reliable thermal–hydraulic initial conditions.     

核电厂二级PSA释放类划分及代表性事故序列选取研究

国内AP1000、EPR、华龙一号等核电工程项目已将二级概率安全分析（PSA）源项用于应急输入,但二级PSA释放类的划分以及各释放类代表性事故序列的选取尚无明确可操作的方法,需要进一步开展研究。对比研究国内先进核电厂二级PSA释放类划分和代表性事故序列选取情况,以国内某三代先进压水堆核电厂为例,在同一释放类中根据频率和后果选取4个不同的严重事故序列开展源项计算。结果表明,同一释放类4个不同事故序列的源项结果差别较大,建议释放类划分以应用为导向,根据分析目的进行迭代,对同一释放类应选取多个事故序列进行对比分析,以论证释放类划分的合理性和事故序列的代表性。      

基于MAAP5程序的秦山核电站严重事故分析

应用MAAP5程序建立了秦山核电站一、二回路,安全系统以及安全壳的模型,并以冷段双端断裂叠加高高、高、低压安注失效,安全壳喷淋系统失效为例,对该严重事故序列进行了模拟计算,给出了瞬态过程一些重要参数随时间的变化规律。结果表明：在72 h内无能动干预手段的条件下,安全壳的完整性可得到保证,相关数据可为秦山核电站严重事故预防和事故缓解措施的制定提供重要参考。     

The development and demonstration of integrated models for the evaluation of severe accident management strategies—SAMEM

This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the generic modelling framework was refined to enable a number of applications. Some recommendations have also been made regarding the applicability of these approaches to existing operating reactors and future reactors. The need for further research and development in the area of human reliability quantification was identified.