Impact of thicker cladding on the nuclear parameters of the NPP Krško fuel

To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krško that uses 16 × 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.     

Animation model of Krško nuclear power plant for RELAP5 calculations

Today most software applications, also in the nuclear field, come with a graphical user interface. The first graphical user interface for the RELAP5 thermal-hydraulic computer code was called the Nuclear Plant Analyzer (NPA). Later, Symbolic Nuclear Analysis Package (SNAP) was developed. The purpose of the present study was to develop SNAP animation model of Krško nuclear power plant (NPP) for RELAP5 calculations with the aim to help analyze the results. In addition, the reference calculations for Krško full scope simulator validation were performed with the latest RELAP5/MOD3.3 Patch 03 code and compared to previous RELAP5 versions to provide verified source data, needed to demonstrate animation model. In total six scenarios were analyzed: two scenarios of the small-break loss-of-coolant accident, two scenarios of the loss of main feedwater, a scenario of the anticipated transient without scram, and a scenario of the steam generator tube rupture. The use of SNAP for animation of Krško nuclear power plant analyses showed several benefits, especially better understanding of the calculated physical phenomena and processes. It can be concluded that an animation tool was created, which enables to analyze very complex accident scenarios. The graphical surface helps keeping the overview and focusing on the main influences. Also, the use of such support tools to system codes may significantly contribute to better quality of safety analysis.     

The NUREG-1150 probabilistic risk assessment for the Sequoyah Nuclear Plant

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Sequoyah Nuclear Plant performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results of this PRA indicate that the offsite risk from internal initiating events at Sequoyah are quite low with respect to the safety goals. The containment appears likely to withstand the loads that might be placed upon it if the reactor vessel fails. A good portion of the risk, in this analysis, comes from initiating events which bypass the containment. These events are estimated to have a relatively low frequency of occurrence, but their consequences are quite large. Other events that contribute to offsite risk involve early containment failures that occur during degradation of the core or near the time of vessel breach. Considerable uncertainty is associated with the risk estimates produced in this analysis. Offsite risk from external initiating events was not included in this analysis.     

核电厂风险指引型管道在役检查方法研究及应用

本文对核电厂风险指引型管道在役检查(RI-ISI)方法开展研究。RI-ISI方法将风险见解融入到在役检查,采用风险矩阵从两个维度(管道失效可能性和管道失效后果)考虑安全重要度,提高了在役检查的有效性和针对性。将该方法应用于田湾核电站1、2号机组一回路压力边界内的管道在役检查,评价结果表明,在役检查工作量和成本得到明显降低,减少了在役检查人员的放射性照射。本文在最后给出了开展RI-ISI的一些建议。     

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components.This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.     

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.     

部分参数统计方法在30万kW核电厂应用的初步研究

偏离泡核沸腾(DNB)设计基准是反应堆热工水力设计中的重要基准之一,为评价该设计基准是否满足热工水力设计要求,需确定堆芯偏离泡核沸腾比(DNBR)设计限值。本文研究了使用统计学方法确定不确定性的部分参数统计方法原理,并应用该方法和堆芯子通道分析程序对30万kW核电厂DNBR设计限值进行计算。计算结果表明,反应堆冷却剂流量全部丧失事故最小DNBR分析采用部分参数统计较STDP额外获得了约5%的裕量。本文结果为DNBR设计基准的验证提供了关键判据。     

Radiological monitoring results of the ambient environment around Qinshan Nuclear Power Plant

1 Introduction Qinshan Nuclear Power Plant (QNPP), a300 MW pressurized-water reactor, was built in 1983,and put into operation in December 1991. In order toestimate the impact QNPP exerted on the ambient en-vironment and the radiation dose the public received,the lab monitoring system and instantaneous landgamma radiation dose-rate monitoring system wereestablished in 1985, and worked from 1988. This pa-per provides the part results of the lab monitoring sys-tem.2 Monitoring pr…     

基于GO法的核电厂电气主接线系统可靠性分析

将GO法用于核电厂电气主接线的可靠性分析,建立了核电厂典型3/2断路器主接线结构考虑有无检修状态(工作/失效/检修)的系统GO图并进行定性分析和定量计算.最后通过与故障树法所得结果的对比,验证了GO法在电气主接线可靠性分析领域的正确性和优势.     

利用核电厂模拟器进行操纵员可靠性研究

为提高核电厂运动可靠性,需要对核电厂操纵员进行可靠性研究。本文结合我国核电厂操纵员可靠性研究的状况,并参考国际上流行的核电厂操纵员可靠性研究方法,利用两参数威布尔分布的理论在核电厂模拟器上对我国核电厂操纵员进行认知可靠性研究,将该方法得到的结果与其他理论模型的结果进行了比较和讨论,得到了一致的认知。本文的研究方法可为真实核电厂运行提供参考。     

A multibarrier safety concept for the reactor pressure vessel of the Stade Nuclear Power Plant

Two specific problems within the safety case of Stade RPV have been analysed: brittle fracture initiation and arrest under strip type emergency core cooling conditions and safety margins against ductile failure from deep cracks as postulated by ASME- and German KTA-rules. For EOL material conditions exclusion of initiation is shown for cracks of more than twice the size which is safely detectable by NDE; for arbitrarily postulated large cracks it is demonstrated that they are arrested well within the allowed depth of of the wall thickness; therefore no critical crack size exists for Stade RPV under strip cooling. Growth in depth of an assumed circumferential flaw in the girth weld embrittled at EOL could occur only at upper shelf temperatures and by loads higher than about twice the service pressure; leak before break was demonstrated in a constraint-modified J − R-curve crack-growth analysis. But neither a transient nor the plant itself would be able to provide the necessary high loads. The LEFM and EPFM proofs are summarized in a multibarrier safety scheme.     

Summary of NET plant probabilistic safety approach and results by means of ENEA fusion plant safety assessment (EFPSA)

This paper summarizes the probabilistic safety assessment for the main accident scenarios associated with failures originating in the In-Vessel Plant Area of the Next European Torus (NET). The assessment refers to the Basic Performance Phase of operation under normal running and conditioning. For the corresponding accident sequences, the values of the annual expected frequency and the seriousness of consequences expressed as early dose to the Most Exposed Individual (MEI) of the public are listed.     

大型集成概率安全分析软件系统的研究与发展

FDS团队在广泛调研和深入分析国际现有概率安全分析软件及其关键技术的基础上,研发了具有自主知识产权的大型概率安全分析软件系统RiskA.该软件提供了系统建模、故障树分析、事件树分析、不确定性分析、可靠性数据管理与分析、敏感性分析和重要度分析等概率安伞分析所需的基本功能.介绍RiskA的设计思想、总体结构、主要功能、技术特点和相关测试与应用等.     

AFAL分析方法在核电厂仪表标定周期延长论证中的应用

介绍核电厂的仪表漂移特性(AFAL)分析法以及在核电厂仪表标定周期延长论证中的应用,以及AFAL的主要技术手段和实施过程。通过对AFAL的研究,论证仪表标定周期延长的可行性。在确保核安全的前提下,实现核电厂机组换料大修周期的延长。     

The erection and commissioning of the Mülheim-Kärlich Nuclear Power Plant

The contract for the Mülheim-Kärlich Nuclear Power Plant, equipped with a 1300 MWe pressurized water reactor, was awarded in 1973. The erection phase of the plant had been characterized in the mid-seventies by an aggravation of circumstances in connection with the nuclear energy controversy in the Federal Republic of Germany, the tightening of stipulations regarding safety philosophy, regulations and documentation, and by the consequences of the TMI accident in 1979. These led to considerable additional difficulties and delays. The commissioning phase on the other hand proceeded smoothly and speedily without major disturbances. The Mülheim-Kärlich Nuclear Power Plant has some major technical features distinguishing it from other pressurized water reactor plants built in the Federal Republic of Germany. Its nuclear steam system is based on a license from the Babcock & Wilcox Company, USA, but it was adapted to German rules and regulations. The Mülheim-Kärlich power plant is the first of this type and size built and put into operation. Its main technical features are described and, after a brief survey of the erection phase, the results of the start-up operations are discussed.     

地下核电研究现状

日本福岛核事故以后,能否探索一种安全性更好、公众接受度更高的核电站,成为未来安全发展核电的重要方向。地下核电站的研究经过几十年的蛰伏,重新进入研究人员的视野。本文总结了上世纪50年代以来国外地下核电领域的研究成果,介绍了近年来我国地下核电方面的研究历程和研究现状。地下核电利用洞室围岩增加了一道实体屏障,有利于实现从设计上实际消除大量放射性物质释放的可能性。     

高温气冷堆核电站的经济分析

本文采用动态分析法对中外合资高温气冷堆(HTGR)电站作了经济分析,在此基础上整理出一个核电站经济分析通用的模型,可用于气冷堆,但所编程序具有通用性。     

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.     

核电站凝汽器仿真模型研究

凝汽器是核电站汽轮发电机组的主要设备之一,采用仿真手段研究核电站凝汽器的运行特性对于优化设计和运行、提高运行的安全性和经济性具有实际意义。本文以核电站中普遍采用的表面式单流程对分制凝汽器为研究对象,运用分布式热动力学的建模思想,考虑不凝结气体（空气）对换热特性的影响,建立了凝汽器的动态数学模型。通过仿真计算,得到不同控制参数阶跃情况下凝汽器的动态响应,仿真结果与实验结果符合良好,表明所建立的仿真模型正确、合理。