Seismic safety of nuclear power plants in eastern Europe

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on "Benchmark study for the seismic analysis and testing of WWER type nuclear power plants". These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.     

Seismic re-evaluation of EDF Bugey 900 PWR nuclear power plant in the frame of the 3rd periodic safety review

As compared to the previous Periodic Safety Review (PSR), main features of the seismic facet of the Bugey NPP (France) 3rd PSR are relating to (i) a 45% increase of the Review Level Earthquake (RLE), (ii) a dramatic increase of computer code capabilities and (iii) a certain evolution of safety standards. This paper presents the method implemented to cope with this situation. Seismic input motion is determined according to an updated Basic Safety Rule published in 2001. Concerning buildings, 3D models are established, accounting for torsion effects, which were not represented by the initial design spring-mass models; analyses are carried out as per the IAEA methodology, accounting for ductile capacity through a F_μ factor; finally the major impact of the seismic re-evaluation is a reinforcement of electrical building, so that it is less torsion sensitive. Regarding equipment, the concept of Safe Shutdown Equipment List (SSEL) is enlarged to all the Safety Related Equipment List (SREL). An equipment hierarchy classification is introduced, based on two complementary approaches: a system approach that identifies components associated to success paths for a safe shutdown, and a ruggedness approach that considers experience feedback about seismic behavior of components. On these bases, all the safety related equipments are classified in two categories: High Hierarchy level (HH) and Medium Hierarchy level (MH). MH equipment is mainly verified by walkdown, whereas HH equipment is by calculation, taking benefit from design margins. The optimization procedures enable to focus on the more relevant components. Bugey NPP seismic re-evaluation stands out by its realistic calculation chain and the completeness of its assessment, and eventually, the 3rd PSR is an opportunity to improve the plant safety and increase design margins.     

Testing to determine relay seismic ruggedness

The seismic qualification of equipment in operating nuclear plants has been identified as a potential safety concern in U.S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46, “Seismic Qualification of Equipment in Operating Nuclear Power Plants”. In response to this concern, the Seismic Qualification Utility Group (SQUG), with support from the Electric Power Research Institute (EPRI), has undertaken a program to demonstrate the seismic adequacy of essential equipment by the use of actual experience with such equipment in plants which have undergone significant earthquakes and by the use of available test data for similar equipment. An important part of this program is the development of the methodology and test data for verifying the functionality of electrical relays used in essential circuits needed for plant shutdown during a seismic event. This paper describes the EPRI supported relay testing program to supplement existing relay test data. Many old relays which are used in safe shutdown systems of SQUG plants and for which seismic test data do not exist have been shake-table tested. The testing performed on these relays and the test results for two groups of relays are summarized in this paper.     

Earthquake response of inelastic structures

The most commonly used method in the seismic analysis of structures is the response spectrum method. For seismic re-evaluation of existing facilities elastic response spectrum method cannot be used directly as large deformation above yield may be observed under Safe Shutdown Earthquake (SSE). The plastic deformation, i.e. hysteretic characteristics of various elements of the structure cause dissipation of energy. Hence the values of damping given by the code, which does not account hysteretic energy dissipation cannot be directly used. In this paper, appropriate damping values are evaluated for 5-storey, 10-storey and 15-storey shear beam structures, which deform beyond their yield limit. Linear elastic analysis is performed for the same structures using these damping values and the storey forces are compared with those obtained using inelastic time history analysis. A damping model, which relates ductility of the structure and damping, is developed. Using this damping model, a practical structure is analysed and results are compared with inelastic time history analysis and the comparison is found to be good.     

10MW 高温气冷实验堆的地震监测系统

为了保障地震发生时10MW高温气冷实验堆（HTR－10）的正常运行和安全停堆,设置了地震监测系统。该系统按无人值守方式设计,主要由高灵敏度的加速度传感器、高可靠性的工业控制计算机和成熟的地震分析软件组成,信号采集处理采用了最新技术。系统监测三个测点的九路加速度时程信号,判断为地震信号后计算峰值加速度,超过报警阈值时向主控室提供声光报警信号;系统同时计算响应谱和设计响应谱,并与理论设计响应谱进行比较,最后打印出地震报告。抗震试验和性能试验结果表明,该系统符合抗震设计要求,其功能覆盖并超过了组合使用的多种地震监测仪表。     

Seismic re-evaluation of nuclear facilities worldwide: overview and status

Existing nuclear facilities throughout the world are being subjected to severe scrutiny of their safety in the event of an earthquake. In the United States, there have been several licensing and safety review issues for which industry and regulatory agencies have cooperated to develop rational and economically feasible criteria for resolving the issues. Currently, all operating nuclear power plants in the United States are conducting an Individual Plant Examination of External Events, including earthquakes beyond the design basis. About two-thirds of the operating plants are conducting parallel programs for verifying the seismic adequacy of equipment for the design basis earthquake. The U.S. Department of Energy is also beginning to perform detailed evaluations of their facilities, many of which had little or no seismic design. Western European countries also have been re-evaluating their older nuclear power plants for seismic events often adapting the criteria developed in the United States. With the change in the political systems in Eastern Europe, there is a strong emphasis from their Western European neighbors to evaluate and upgrade the safety of their operating nuclear power plants. Finally, nuclear facilities in Asia are also being evaluated for seismic vulnerabilities. This paper focuses on the methodologies that have been developed for re-evaluation of existing nuclear power plants and presents examples of the application of these methodologies to nuclear facilities worldwide.     

基于FPGA和MCU的低成本地震信号数字滤波器设计

针对地震数据采集系统提出一种低成本、高灵活性的数字滤波设计方案。不仅对于地震数据采集系统有很大的实用意义,对于其它传感器网络也是可供参考的解决方案。设计的目的是在保持优秀的低通性能的同时实现高度抽取。该滤波器系统由四级CIC滤波和两级FIR滤波构成,分别在FP-GA和MCU内部实现。滤波器有着良好的低通性能,在430 Hz～500 Hz范围内幅度响应下降了137 dB。采用实际信号测试,信噪比好于110 dB,能够很好地满足地震数据采集的需求。     

IMS地震台站数据模拟发送系统

连续数据子系统可以保证国际监测系统中台站的监测数据安全可靠地发送到国际数据中心。设计了一种地震台站数据模拟发送系统,在无法接收国际监测系统地震台站数据的情况下,可以用来调试连续数据子系统。这套模拟系统由5台工作站和4台服务器组成,可以连续不断地读取和发送CD1.0格式的地震数据,按照数据接收方的要求任意更改记录的起始时间,能够进行数据发送、转发和归档等实验。     

高温气冷堆蓄电池组地震易损性研究

为验证核电厂发生地震外部事件时的电力安全,需要对蓄电池组进行抗震鉴定试验。本文以高温气冷堆(HTR)核电厂安全级蓄电池组为研究对象、以安全级蓄电池组抗震鉴定试验数据和工程经验为基础,通过识别、量化蓄电池组的地震易损性变量,并应用基于试验的易损性分析法推导出地震易损性曲线和高置信度低失效概率(HCLPF)抗震能力。研究结果表明,安全级蓄电池组的抗震能力远高于核电厂设计基准地震动需求。     

反应堆堆顶和控制棒驱动机构整体抗震分析

通过建立反应堆堆顶和控制棒驱动机构(CRDM)的整体三维有限元简化抗震分析模型,对反应堆堆顶和CRDM进行了三维非线性抗震分析。该分析模型可真实反映CRDM顶部抗震支承板相互之间、抗震支承板与抗震支承环之间的碰撞作用,以及抗震拉杆的拉压非线性特性,进而获得堆顶和CRDM结构各位置的精确分析结果。通过分析,得到反应堆堆顶和CRDM各部位的载荷,可为部件的应力分析提供必要的地震载荷,为CRDM抗震鉴定试验提供加速度时程等输入数据。     

Parametric study of relay seismic capacity

An evaluation of the existing relay test data base at Brookhaven National Laboratory (BNL) has indicated that the seismic capacity of a relay may depend on various parameters related to the design or the input motion. In order to investigate the effect of these parameters on the seismic fragility level, BNL has conducted a relay test program. Establishing the correlation between the single frequency fragility test input and the corresponding multifrequency response spectrum (TRS) is also an objective of this test program. The testing has been performed at Wyle Laboratories. This paper discusses the methodology used for testing and presents a brief summary of important test results.     

钢筋混凝土框架结构弹塑性动力响应分析

基于一榀空间混凝土框架结构的地震响应实验结果,采用以力法为基础的纤维梁单元模拟材料非线性及粘结-滑移特征,进行弹塑性动力响应分析,并将分析结果与线弹性反应分析及实测结果进行对比.结果表明,弹塑性分析足提高钢筋混凝土结构地震响应分析精度的有效手段,对于在役核电站的地震安全评价,引入有效的非线性分析方法是有价值的.     

压水堆燃料组件抗震试验研究

燃料组件属I类抗震物项,其抗震问题直接关系核电厂运行安全,通常需通过抗震试验验证反应堆燃料组件抗震分析方法的合理性。本文模拟反应堆实际堆芯燃料组件安装方式,设计压水堆燃料组件抗震试验件与试验装置,针对不同组件数量布置方案,在高性能地震模拟振动台上开展试验研究。结果表明,水介质中燃料组件的第一阶频率为2.96 Hz,最大冲击力出现在燃料组件偏中间位置处,试验获取了地震作用下燃料组件的格架冲击力、格架相对位移、模拟堆芯板与围板的加速度等响应。试验结果可用于设计基准事故工况中燃料组件抗震分析模型的建立与分析软件的验证。     

基于分布式系统的地震数据处理平台

基于分布式应用控制系统的地震数据处理平台是针对地震数据处理的特点设计的.该平台为地震数据处理提供了一个分布处理的框架,支持地震数据的自动处理和交互分析,把分布于各个工作站上的UNIX软件进程连接为一个统一的、一致的处理分析平台.该平台的建立,大大加快了地震数据处理的速度.     

Some research needs for improved seismic qualifications tests of electrical and mechanical equipment

There is currently a wide variety of methods of seismic simulation for qualifications tests for components to be used in nuclear power plants. This paper poses some questions with regard to the relationships among these methods of seismic simulation, the influence of laboratory seismic simulators required to produce the test seismic environment, and the way one demonstrates component functionality during a qualifications test. In particular, evidence obtained through experience in the laboratory indicates the need for research to improve currently accepted seismic qualifications tests and to show the relationship between previous and present test methods.     

Strength and stiffness of reinforced concrete containments subjected to seismic loading: Research results and needs

The primary purpose of this paper is to present results of an experimental investigation on the strength and stiffness of reinforced concrete subjected to combined biaxial tension and simulated seismic forces. The test specimens represent a section of a wall of a containment structure carrying combined pressurization and seismic loading. Shear stiffness and strength, and their degradation with shear cycling, are given, along with simple expressions for predicting strength and extensional stiffness. The secondary purpose of the paper is to discuss research needs for improved prediction of the response of containment structures to seismic effects.     

Seismic individual plant examination of external events of US nuclear power plants: insights and implications

As part of the implementation of the severe accident policy, nuclear power plants in the US are conducting the individual plant examination of external events (IPEEE). Seismic events are treated in these IPEEEs by either a seismic probabilistic risk assessment (PRA) or a seismic margin assessment. The major elements of a seismic PRA are the seismic hazard analysis, seismic fragility evaluation of structures and equipment and systems analysis using event tree and fault tree analysis techniques to develop accident sequences and calculate their frequencies of occurrence. The seismic margin assessment is a deterministic evaluation of the seismic margin of the plant beyond the design basis earthquake. A review level earthquake is selected and some of the components that are on the success paths are screened out as exceeding the review level earthquake; the remaining ones are evaluated for their seismic capacity using information from the original plant design criteria, test data and plant walkdown. The IPEEEs of over 100 operating nuclear power plants are nearing completion. This paper summarizes the lessons learned in conducting the IPEEEs and their applicability to nuclear power plants outside of the United States.     

基于经验-解析数据的某核电安全壳地震易损性分析

本文基于混合数据的地震易损性分析方法,对我国已运行核电厂地震易损性分析进行研究。首先基于地震危险性分析和分解结果,生成了我国华南地区某核电厂厂址条件谱;然后采用贪心优化算法,选取符合厂址危险性的地震动记录;基于增量动力分析方法,生成我国某核电厂安全壳地震易损性安全系数FS和FSA的解析数据;地震易损性其他参数采用经验数据,基于经验-解析数据,生成了我国某核电厂安全壳地震易损性曲线。建议将基于经验-解析数据的地震易损性分析方法应用于我国核电厂安全壳初步地震易损性分析中。     

Vertical seismic response of overhead crane

Vertical seismic response behavior is an important issue for the seismic design of equipments. The equipment, which is comparatively soft and unrestrained vertically, may resonate and its response is significantly magnified under vertical seismic excitation. Overhead crane is an example of equipment that is unrestrained vertically. The dynamic behavior of an 150-ton-capacity overhead crane under vertical seismic excitation was investigated by scale model excitation test and nonlinear time history analysis. The excitation tests were performed with several input levels and the vertical response with each input level was obtained. The simulation analysis approximately corresponded to the results of the excitation test.     

An evaluation of structural failure modes for prestressed concrete containments

Prestressed Concrete Containment Vessels (PCCVs) refer to a popular type of containment used in the United States for Pressurized Water Reactors (PWRs).This paper presents analytically predicted ultimate pressures and seismic levels for PCCVs, considering various modes of failures. Results for six containments are presented, and correlated with the available test data.The analytical methods use either classical techniques or finite element analyses. The ultimate capacity calculations are based upon conservative deterministic estimates of strength of the structure, under both internal pressure and earthquake loads.The results indicate the following: internal pressure capacities of PCCVs built in the US are almost uniformly equal to 2.5 times the design pressure; seismic capacities are at least two times the design level, but they vary widely among the PCCVs depending on the foundation characteristics; seismic capacity of a PCCV decreases with internal pressure; and a PCCV is expected to contribute very little to the overall seismic risk of a nuclear power plant.