基于统计的核电厂地震裕量分析建议反应谱

基于美国新一代地震动衰减关系(NGA)数据库中350条基岩强震动加速度记录数据,以及我国汶川M_W7.9地震和芦山M_W6.6地震中获取的14条基岩强震动加速度记录,获得可用于核电厂地震裕量分析的基岩水平向加速度反应谱谱型。获得的反应谱充分考虑了地震规模(震级)对地震动反应谱频率成分的显著影响,在核电厂地震裕量分析中可以考虑厂址所处地震构造环境对输入地震动反应谱谱型的影响。相对于RG1.60谱,给出的反应谱能够更加可靠地反映近场中强地震产生的地震动高频成分。     

基于近场加速度记录的弥散地震影响评价方法

基于地震动近场饱和的现象,根据已有的地震加速度记录样本,以及随距离衰减特征和震源尺度等因素综合确定的震中附近加速度记录样本,进行直接统计获得可以作为核工程安全性分析所使用的弥散地震加速度反应谱及其离散特征。为获得足够的样本数量,根据已有的地震动场地条件修正关系将非基岩场地上的加速度记录转换成为等效基岩场地上的记录,与基岩记录同时用于统计分析。     

非基岩核电厂结构地震响应振动台试验研究

为研究在非基岩场地条件下核电厂结构的适用性和地震响应特征,以CAP1400型核电厂结构为例,开展非基岩场地核电厂结构振动台试验。结果表明:模型场地对各方向上的地震动均放大,场地反应谱低频部分受结构影响较大;在低于基准地震动作用下场地出现裂缝,在设计基准地震动作用下结构与土体分离。试验结束后,场地表面裂缝连通,结构无裂缝,地基失稳破坏。核电厂结构地震响应受场地条件的影响明显,在进行核电厂结构地震响应分析时应考虑场地条件和进行土-结构相互作用(SSI)分析。     

核电厂厂址设计地震动参数影响关键因素分析

基于中国核电厂选址的46个工程场地地震安全性评价资料,分析不同地震危险性分析方法计算结果对厂址设计地震动参数确定的控制作用,并对地震危险性分析概率方法计算结果及确定性方法中的构造地震影响、弥散地震影响计算结果进行统计分析。研究表明:在地震活动性较弱地区,厂址设计地震动参数主要由确定性方法计算结果控制,峰值加速度和高频加速度反应谱值由弥散地震计算结果控制,在这类地区基于厂址设计地震动的核电工程建设将具有更高的抗震安全裕度;在地震活动性相对较强地区,厂址设计地震动参数更可能由概率方法计算结果控制,部分厂址的概率方法计算结果(特别是低频加速度反应谱值)远大于确定性方法计算结果;中国核电厂厂址设计地震动参数确定总体上具有较高保守性。     

AP1000设计地震反应谱在具体厂址评价中的应用

分析AP1000设计地震反应谱（CSDRS）与各相关导则中定义的反应谱的对应关系,指出在特定厂址评价中,应基于同一标高比较厂址特定设计反应谱（SRS）和AP1000 CSDRS。基于5种设计场地模型将AP1000 CSDRS反演至设计基岩处和核岛结构基础底部,计算得到设计基岩处和结构基础底部的AP1000设计谱。计算结果表明,AP1000 CSDRS不能包络已有核电厂核岛结构抗震设计采用的0.2g标定的RG1.60的设计反应谱;若在非硬质基岩场地建造AP1000核岛结构,应进行AP1000 CSDRS的保守性分析。     

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。     

考虑一致危险性谱的核电厂房结构振动台试验研究

为得到适合特定核电厂所需要的反应谱，考虑具体的场地条件及地震动参数，采用随机模拟方法与概率危险性分析相结合的方式，建立了生成超越概率为10-4的一致危险性谱（UHS）的方法。为进一步研究核电结构的抗震性能及UHS在实际核电结构中的适用性，设计和制作了1∶20的核电厂房结构模型进行振动台试验，采用2条天然波及UHS、厂址谱（SL-2）、RG1.60谱所生成的人工波对结构的响应进行比对分析。结果表明，不同地震波对核电结构的响应有所差异，UHS生成的人工波对上部结构加速度放大效应以及位移影响较大，对应的楼层反应谱幅值相对其他反应谱较高，进行结构及设备抗震设计时应予以考虑。      

考虑竖向地震动影响的某核电安全壳地震易损性研究

核电厂等重要基础设施的抗震设计和评估需要考虑竖向地震动影响,目前竖向地震动对核电安全壳地震易损性影响研究还较少。本文进行了考虑竖向地震动影响的核电安全壳地震易损性研究,分析了以水平向场地相关谱为目标谱选取的地震动记录的不足,提出了同时匹配水平和竖向场地相关谱的地震动选取方法,选取了指定场址的水平和竖向地震动记录。采用增量动力分析方法,基于选取的水平和竖向地震动,分别进行核电安全壳水平向地震动作用下与水平和竖向地震动联合作用下的易损性分析。采用基于混合易损性数据的易损性分析方法,得到了具有置信度的易损性曲线和高置信度低失效概率。分析结果表明：竖向地震动对安全壳抗震能力和地震易损性有较大影响。     

不同输入界面对AP1000核岛结构设计地基地表地震动的影响

在核电厂地震安全性评价中,中国规范是依据剪切波速定义的基岩面,与美国规范不同。本文基于AP1000核岛结构设计地基的场地参数模型,分别选取各个规范中定义的剪切波速700、1100、2438 m/s基岩层作为地震反应分析的输入界面,采用中美两国通用的土层地震反应分析程序计算,定量分析选取不同地震输入界面时同一地震波、同一特定场地模型的地表加速度峰值和反应谱的差异值,结果表明地震输入界面的不同,AP1000核岛结构设计地基的同一场地土层模型地表地震加速度反应谱频谱特性产生较大的变化,地表加速度峰值差异高达2.25倍,故本文建议在AP1000核电厂地震安全性评价中应基于剪切波速为2438 m/s的基岩层作为土层地震反应分析程序的地震输入界面。本文的研究结果可供后续研究和核电工程建造应用参考。     

匹配设计反应谱的目标功率谱密度的确定方法

美国核管会标准审查大纲（SRP）3.7.1节要求，核电厂结构、系统和部件（SSCs）抗震设计时程需同时满足包络设计反应谱和匹配设计反应谱的目标功率谱密度（PSD）的要求。本文结合2014版SRP 3.7.1，对匹配设计反应谱的目标PSD的确定方法进行介绍，并根据其算法编写相应的计算程序，通过算例分析对程序结果进行验证。结果表明:计算所得RG1.60谱、美国中东部和西部基岩厂址谱的目标PSD与SRP 3.7.1结果具有较好的一致性，且基于本文所得目标PSD和三角级数叠加法所构造的加速度时程反应谱与设计反应谱匹配良好。本文所给出的目标PSD的确定方法可为核电厂抗震设计时程的PSD检验提供重要依据，且采用本文方法生成目标PSD，设计时程的PSD检验仅需包络该目标PSD的70%。      

Lower bound earthquake magnitude for probabilistic seismic hazard evaluation

This paper presents the results of a study that develops an engineering and seismological basis for selecting a lower-bound magnitude (LBM) for use in seismic hazard assessment. As part of a seismic hazard analysis the range of earthquake magnitudes that are included in the assessment of the probability of exceedance of ground motion must be defined. The upper-bound magnitude is established by earth science experts based on their interpretation of the maximum size of earthquakes that can be generated by a seismic source. The lower-bound or smallest earthquake that is considered in the analysis must also be specified.The LBM limits the earthquakes that are considered in assessing the probability that specified ground motion levels are exceeded. In the past there has not been a direct consideration of the appropriate LBM value that should be used in a seismic hazard assessment. This study specifically looks at the selection of a LBM for use in seismic hazard analyses that are input to the evaluation/design of nuclear power plants (NPPs). Topics addressed in the evaluation of a LBM are earthquake experience data at heavy industrial facilities, engineering characteristics of ground motions associated with small-magnitude earthquakes, probabilistic seismic risk assessments (seismic PRAs), and seismic margin evaluations. The results of this study and the recommendations concerning a LBM for use in seismic hazard assessments are discussed.     

An extensive study of a simple method for estimating the response spectrum based on a simulated spectrum

A simple method to simulate the response spectrum is proposed and developed. The response spectrum of the single-degree-of-freedom system shows generally multi-peaks at ground predominant periods. The simulated response spectrum of a building and a building-appendage structure system is made a standard. It is statistically computed using the theory of random vibration assuming earthquake motion with a single ground predominant period. Based on the fact that the simulated response spectrum for two ground predominant periods fits well with the spectrum for the earthquake motion, it is proposed that the simulated response spectrum can also be given by adequate summation of the standard spectrum, which is carried out by the method of square root of sum of squares. This makes it possible to simulate the spectrum which has generally plural ground predominant periods. This was virtually impossible using the statistical computation. It is shown that the method is generally applicable to multi-degree-of-freedom building systems and appendages. This means that if the average response spectrum for a single-degree-of-freedom at a site is given, the so-called floor response spectrum can be estimated from the standard spectrum.     

Numerical computation of fragility curves for NPP equipment

The seismic probabilistic risk assessment (PRA) methodology is a popular approach for evaluating the risk of failure of engineering structures due to earthquake. In this framework, fragility curves express the conditional probability of failure of a structure or component for a given seismic input motion parameter A, such as peak ground acceleration (PGA) or spectral acceleration. The failure probability due to a seismic event is obtained by convolution of fragility curves with seismic hazard curves. In general, a log-normal model is used in order to estimate fragilities. In nuclear engineering practice, these fragilities are determined using safety factors with respect to design earthquake. This approach allows to determine fragility curves based on design study but largely draws on expert judgement and simplifying assumptions. When a more realistic assessment of seismic fragility is needed, simulation-based statistical estimation of fragility curves is more appropriate. In this paper, we will discuss statistical estimation of parameters of fragility curves and present results obtained for a reactor coolant system of nuclear power plant. We have performed non-linear dynamic response analyses using artificially generated strong motion time histories. Uncertainties due to seismic loads as well as model uncertainties are taken into account and propagated using Monte Carlo simulation.     

Prediction equations for the standardized version of cumulative absolute velocity as adapted for use in the shutdown of U.S. nuclear power plants

The U.S. Nuclear Regulatory Commission (USNRC) uses the standardized version of the cumulative absolute velocity (CAV) together with the response spectra of the recorded ground motion at a site to determine whether a nuclear power plant must be shut down for inspection after an earthquake. In order to better understand the impact of these criteria on potential plant shutdowns, we used several subsets of the PEER-NGA strong motion database to develop empirical prediction equations between a CAV intensity measure that incorporates these criteria and the geometric mean horizontal component of CAV. This particular approach was used because, after applying the USNRC shutdown criteria, we found that there were an insufficient number of records remaining to reliably develop a ground motion prediction equation (GMPE) for this new CAV intensity measure directly from the physical parameters of an earthquake. We use these prediction equations to demonstrate how a lower-bound value of this intensity measure can be used to prevent non-damaging ground motions from both nearby small-magnitude earthquakes and distant large-magnitude earthquakes from contributing to the ground motion hazard computed from a probabilistic seismic hazard analysis (PSHA). We suggest that the use of a CAV criterion, in addition to a criterion based on either peak ground motion parameters or response spectral ordinates, could also be used to prevent the unnecessary shutdown of nuclear power plants outside of the U.S. Furthermore, with some adjustment, a similar approach could be used to rapidly assess the potential damage to conventional structures after an earthquake to aid in emergency response and loss assessment.     

Probabilistic seismic hazard analysis for a site

Traditionally, the seismic design basis ground motion has been specified by normalised response spectral shapes and peak ground acceleration (PGA). The mean recurrence interval (MRI) used to be computed for PGA only. The present work develops uniform hazard response spectra, i.e., spectra having the same MRI at all frequencies for Tarapur Atomic Power Station Site. Sensitivity of the results to the changes in various parameters has also been presented. These results determine the seismic hazard at the given site and the associated uncertainties.     

Preliminary seismic analysis of an innovative near term reactor: Methodology and application

Nuclear power plant (NPP) design is strictly dependent on seismic hazard and safety aspects concerned with the external events of the site. Earthquake resistant structures design requires realistic and accurate physical and theoretical models to describe the response of the nuclear power plants (NPPs) that depend on both the ground motion characteristics and the dynamic properties of the structures themselves. In order to improve the design of new NPPs and, at the same time, to retrofit existing ones the dynamic behaviour of structures subjected to critical seismic excitations that may occur during their expected service life must be evaluated.The aim of this work is to select new effective methods to assess NPPs vulnerability by properly capturing the effects of a safe shutdown earthquake (SSE) event on nuclear structures, like the near term deployment IRIS reactor, and to evaluate the seismic resistance capability of as-built structures systems and components. To attain the purpose a validated deterministic methodology based on an accurate finite element modelling coupled to substructure and time history approaches was employed for studying the overall dynamic behaviour of the NPP relevant components. Moreover the set up three-dimensional model was also validated to evaluate the performance and reliability of the adopted FEM code (mesh refinements and type element influence). This detailed numerical assessment, involving the most widely used finite element numerical codes (MSC.Marc^® and Ansys^®), allowed to solve, perform and simulate as accurately as possible the dynamic behaviour of structures which may withstand a lot of more or less complicate structural problems.To evaluate the accuracy and the reliability as well as to determine the related error of the set-up procedure, the obtained seismic analyses results in term of accelerations, propagated from the ground to the auxiliary building systems and components, and displacements were compared highlighting a very good agreement.     

Comments on earthquake response spectra

Statistical summaries have been carried out on a large number of earthquake records to examine the influences of geological conditions, duration of strong motion, and peak ground acceleration on ground motion and response spectra. The results indicate that the peak ground velocity-acceleration ratio is substantially lower for records on rock deposits than those on alluvium, and it is lower for records with a peak ground acceleration greater than 0.20 g than those with acceleration less than 0.20 g. For each influence the spectral bounds defined as the product of mean ground motion and mean plus one standard deviation amplifications are computed for five damping coefficients and compared to those for alluvium deposits without consideration of duration and acceleration level.     

Effect of near-fault earthquakes on North American nuclear design spectra

Ground motion records from recent earthquakes show that near-fault ground motions are different from far-field ground motions in that they often contain strong coherent dynamic long period pulses and permanent ground displacements. The objective of the current study is to assess the safety implications of near-fault earthquakes (NFE) on nuclear power plant structures designed according to North American codes. Fifty-four fault-normal near-fault records in the forward directivity were selected for this investigation. Spectral comparison of the selected earthquake records with the US and Canadian design spectra was conducted. From the investigation it was concluded that the current nuclear design spectra may need to be adjusted to better reflect the effect of near-fault earthquake ground motions.     

Artificial ground motion compatible with specified peak ground displacement and target multi-damping response spectra

With respect to the design ground motion of nuclear power plant (NPP), the Regular Guide 1.60 of the US not only defined the standard multi-damping response spectra, i.e. the RG1.60 spectra, but also definitely prescribed the peak ground displacement (PGD) value corresponding to the standard spectra. However, in the engineering practice of generating multi-damping-spectra-compatible artificial ground motion for the seismic design of NPP, the PGD value had been neglected. Addressing this issue, this paper proposed a synthesizing method which generates the artificial ground motion compatible with not only the target multi-damping response spectra but also the specified PGD value. Firstly, by the transfer formula between the power spectrum and the response spectrum, an initial uniformly modulated acceleration time history is synthesized by multiplying the stationary Gaussian process with the prescribed intensity envelope to simulate the amplitude-non-stationarity of earthquake ground motion. And then by superimposing a series of narrow-band time histories in the time domain, the initial time history is modified in the iterative manner to match the target PGD as well as the target multi-damping spectra with the pre-specified matching precisions. Numerical examples are provided to demonstrate the matching precisions of the proposed method to the target values.     

Seismic considerations in siting and design of power plants

A methodology which provides guidelines for the preliminary evaluation of the safety of nuclear power stations subjected to strong vibratory ground motions from earthquakes is outlined. The methodology includes a procedure for estimating a spectral envelope of ground motion at the reactor site. On the basis of this ground motion the seismic response of structural systems and equipment of the power plant can be estimated. A comparison of the expected seismic response of these systems with their strength and functional capabilities yields an evaluation of the safety of the power plant systems studied.