基于随机模拟一致危险性谱的核电厂抗震性能分析

为改善概率地震危险性分析对震源传播特性考虑的不足,提出采用随机模拟与概率地震危险性分析结合的方法,充分考虑反应谱生成中震源机制、传播路径和场地效应等影响,生成更为精确的一致危险性谱。结合核电厂具体场地条件对场地近两千年的历史地震进行模拟,并生成同一超越概率下的一致危险性谱（UHS）。为了比较已有的厂址谱（SL-2）和安评报告中的UHS及美国RG1.60谱所生成的地震动对结构抗震性能的影响,以某核电结构为例,建立三维有限元模型,进行动力时程分析。结果表明：不同反应谱对结构的动力响应差别较大,UHS与SL-2对结构的响应较为接近,且略大于SL-2,但小于美国RG1.60谱。基于随机模拟方法生成的一致危险性谱可为核电厂抗震设计提供参考。     

Methodology and results of the seismic probabilistic safety assessment of Krško Nuclear Power Plant

A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Kr

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ko plant. The methodology adopted is the seismic PSA (probabilistic safety assessment). The Kr

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ko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of site hazard, calculation of plant structures response including soil–structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Relay chatter analysis and soil stability studies were also performed. The seismic PSA described here is limited to the analysis of CDF (level 1 PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Kr

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ko NPP but are not further described in this paper. The results of the seismic PSA study indicates that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable with most US and Western Europe NPPs located in high seismic areas.     

Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value.     

Seismic analysis of nuclear power plants by using three-dimensional finite element models: a review

ABSTRACT

Seismic design of nuclear power plants (NPPs) is important for ensuring their integrity during earthquakes. Seismic analysis has been conducted using lumped mass beam models (LMBMs) for the design of plants in Japan, whereas three-dimensional (3D) finite element models (FEMs) have been used for novel plants outside Japan. The purposes of this study are to organize issues related to the development and application of 3D FEMs for seismic analysis of Japanese NPPs and to indicate future study directions. To organize these issues, the authors systematically investigated: (1) international guides and standards related to seismic analysis and (2) 3D FEMs of novel NPPs outside Japan. By considering other studies on the issues, the authors suggest directions for future studies. Resolving the issues will contribute to application of 3D FEMs for seismic analysis in the design of Japanese NPPs.     

福岛核事故后对我国核电厂应急控制中心抗震设防的思考

结合福岛核事故后对我国核电厂进行的核安全检查,分析了我国核安全法规关于核电厂应急控制中心的要求以及福岛核事故的经验教训,提出目前我国核电厂应急控制中心采用民用抗震设防标准进行抗震设防,无法保证在由地震引发的应急事故工况下应急控制中心的功能,应该适当提高其抗震设防级别。     

Seismic probabilistic risk analysis based on stochastic simulation of accelerograms for nuclear power plants in the UK

This article presents an approach to probabilistically assess the seismic risk of nuclear power plants (NPPs) in the UK. The approach proposed is based on direct stochastic simulation of the seismic input to conduct nonlinear dynamic analysis of a structural model of the NPP analysed. Therefore, it does not require the use of ground motion prediction equations and scaling/matching procedures to define suitable accelerograms as is done in conventional approaches. Additionally, as the structural response is directly calculated, it does not require the use of Monte Carlo-type algorithms to simulate the damage state of the NPP analysed. However, it demands longer use of computer resources as a relatively large number of nonlinear dynamic analyses are needed to perform. The approach is illustrated using an example of a 1000 MW Pressurised Water Reactor building located in a representative UK nuclear site. A comparison of risk assessment is made between the conventional and proposed approaches. Results obtained are reasonable and well constrained by conventional procedures; hence, it can confidently be used by the UK New Build Programme in the next two decades to generate 16 GWe of new nuclear capacity.     

Realistic seismic qualification using the updated finite element model for in-core components of reactors

It is now mandatory to seismically qualify the safety-related structures and components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However, the estimated dynamic behaviour by FE model of a structure is known to show significant deviations from the dynamic behaviour of the ‘as-installed’ structure in many cases. Considering such limitations, few researchers have advocated re-qualification of such structures after their installation at site to enhance the confidence in qualification vis-à-vis plant safety. For such an exercise, validation of FE model with experimental modal data is important. A validated FE model can be obtained by the model updating methods in conjugation with the in situ experimental modal data. Such a model can then be used for qualification. However, for the reactor in-core components such a modal testing and FE model updating may not be straightforward. Hence, the complication involved in the reliable seismic qualification of in-core components and the advantage of using the FE model updating has been brought out in the paper through an example of a typical in-core component—a perforated horizontal tube recently installed in a nuclear reactor in India.     

Probability-based load combinations for design of Category I structures — Overview of research program and recent results

This paper discusses the probability-based load combinations for the program dealing with the design of Category I structures, currently being worked on at Brookhaven National Laboratory (BNL) for the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission (NRC). The objective of this program is to develop a probabilistic approach for the safety evaluations of reactor containments and other seismic Category I structures subjected to multiple static and dynamic loadings. Furthermore, on the basis of the developed probabilistic approach, a load combination methodology for the design of seismic Category I structures will also be established.The major tasks of this program are: (1) establish probabilistic representations for various loads and structural resistance, (2) select appropriate structural analysis methods and identify limit states of structures, (3) develop a reliability analysis method applicable to nuclear structures, (4) apply the developed methodology to existing Category I structures in order to evaluate the reliability levels implied in the current design criteria, and (5) recommend load combination design criteria for Category I structures. When the program is completed, it will be possible to (1) provide a method that can evaluate the safety margins of existing containment and other Category I structures and (2) recommend probability-based load combinations and load factors for the design of Category I structures.At the present time, a reliability analysis method for seismic Category I concrete structures has been completed. By utilizing this method, it is possible to evaluate the safety of structures under various static and dynamic loads. In this paper, results of a reliability analysis of a realistic reinforced concrete containment structure under dead load, accidental pressure, and earthquake ground acceleration are presented to demonstrate the feasibility of the methodology.     

Modeling of slabs in seismic analysis of nuclear power plant buildings

Safety related nuclear power plant buildings are commonly represented as lumped mass weightless elastic beam stick models to determine their dynamic behavior under seismic ground motions. Implicit in this analysis procedure is the assumption that the floor slabs are rigid. This paper critically evaluates the slab flexibilities in typical power plant buildings and presents a practical approach to include these in the seismic analysis. Vertical as well as horizontal earthquake components are considered. Results presented include amplified floor response spectra for equipment qualification and design forces in floor slabs and the supporting walls. A satisfactory analysis procedure would consist of traditional stick model analysis to obtain overall seismic responses, force distribution by static analysis using suitable methods such as the finite element method and subsystem analysis to evaluate local amplifications, if necessary.     

核电厂大型组合结构的有限元抗震分析方法研究

在现代核电站抗震设计中,有限元法是各类相关设备抗震分析与评价的重要数值仿真工具。对于形状复杂、部件众多的大型组合结构,采用整体三维建模的有限元模型通常需要很大的存储和计算规模,超出现有的计算条件。因此需要首先研究组合结构各个部件的动力学特性,从而建立合理的三维简化力学模型,并以该模型为基础进行有限元数值仿真。本文以某地车-吊车组合结构为例,给出此类大型组合结构的抗震分析方法,并将等效静力法与反应谱法相结合,对该结构进行分析,最后根据相关法规对各子结构进行评价,以确保总体组合结构在极限安全地震条件下能够保持结构完整性。     

Seismic analysis of RCS with finite element model for advanced PWR

Reactor Coolant Pumps (RCPs) are very important to the safe operation of Nuclear Power Plants (NPPs), especially during the earthquake, which needs detailed seismic analysis of individual RCPs and the boundary conditions, for example, at the nozzles. In this paper, three-dimensional finite element model of Reactor Coolant System (RCS) is constructed from a systematic perspective to perform dynamic evaluation, in which the boundary conditions could be given. The seismic spectrum analysis with three orthotropic directions is performed to obtain the stress and displacement response, which shows that the maximum Tresca stress locates in the connection part of SG with RCP and the maximum displacement occurs at the surge line. Sensitivity analysis of spectrum input angle and stiffness of supports is performed, which may be useful to further design and analysis. Furthermore, direct integration method is used to perform time-history analysis, and the boundary conditions of RCP, the loads, acceleration and displacement at nozzles are obtained, which could support the detailed analysis of RCP components. Besides, the lumped mass model of RCS is also constructed to compare with three-dimensional finite element model, which means that for the complicated geometry the 3-D model is better than the lumped mass model.     

不同法规关于核电厂设计地震动合成的技术要求比较

详细介绍了我国核电厂地震安全评价及抗震分析与设计中用到的多部国内外的法规、标准和导则的技术规定,对其技术背景和要求进行了深入比较和分析,结合工程实践给出了相关的评述和应用建议.     

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     

基于神经网络与表决融合的核动力装置故障诊断方法

针对单神经网络（ANN）故障诊断方法的不足,将多神经网络诊断与表决融合方法结合起来,研究了基于多神经网络与表决融合的核动力装置故障诊断方法。在该方法中,多个不同类型的神经网络训练后用于核动力装置的故障诊断。选择对核动力装置安全有重要影响的运行参数作为各神经网络的输入变量,神经网络的输出是核动力装置的故障模式。用表决融合方法对不同神经网络的诊断结果进行融合,从而得到核动力装置故障诊断的最后结果。利用核动力装置典型的运行模式来验证所提出的诊断方法的效果。结果表明,与单神经网络相比,该方法可提高核动力装置故障诊断结果的精度和可靠性。     

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.     

Performance demonstration for an automated ultrasonic testing system for piping welds

This paper addresses the implementation of an automated ultrasonic testing (AUT) system qualification by performance demonstration (PD) as imposed by the ASME Boiler and Pressure Vessel Code Section XI. To improve the reliability of the ultrasonic testing results for nuclear power plant (NPP) components, almost all engineering codes related to NPP inspection require the ultrasonic inspection systems to be qualified by passing a PD examination. In this study, an AUT system developed to inspect pipe welding parts in NPPs is introduced. To acquire a Korean Performance Demonstration (KPD) qualification, the developed system had a KPD. System obtained the qualification for flaw detection, length, and depth sizing from KPD.     

An overview of polymer ageing studies in the nuclear power industry

Polymeric components are widely used in nuclear power plants (NPPs) in equipment which is important to the safety of the plant. The degradation of such components is therefore of considerable interest to the industry and its regulatory bodies, generating a large number of studies worldwide. Some of these components need to remain functional over the full operational life of the plant, which may span up to 60 years. Predictive modelling of their behaviour is therefore of key importance. This paper outlines the main areas of research, particularly relating to the use of elastomeric seals and polymeric cable insulation in NPP.     

The IAEA accident management programme

Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff.