地震波空间相干效应对核电厂土与结构相互作用分析的影响

为研究地震波空间相干效应对核岛厂房土与结构相互作用分析的影响，采用ACS-SASSI软件对典型的压水堆厂房的土与结构相互作用分析时，分析不同土层、不同标高处，空间相干性对楼面反应谱的影响。结果表明，在高频段考虑空间相干性，对于中硬质岩石场地和坚硬土至软质岩石场地将降低楼层响应10%~70%；在高频段考虑空间相干性，对于中硬土场地将降低楼层响应10%~40%。因此，不考虑空间相干性，其楼层响应分析结果高频段偏保守，低频段偏不安全。      

基于统计数据的非基岩核电厂抗震设计谱研究

为获取更为准确的抗震设计谱，统计分析了美国NGA-West2数据库中2661条强震动加速度记录数据，研究了相关地震参数对地震动反应谱的中长周期的影响，并建立了基岩（I类）与非基岩（II类、III类）场地上水平向归一化加速度反应谱。结果表明，地震动反应谱受场地条件和矩震级的影响十分显著，受震源距离参数的影响较弱；与RG1.60谱、GB50267-97谱相比，本文获得的地震动反应谱能够更加可靠地估计场地覆盖层特性以及矩震级对地震动反应谱的中长周期的影响。最后，本文确定了考虑场地条件与地震构造环境影响的抗震设计谱，可作为非基岩核电厂选址及抗震设计的地震动输入。      

核安全审评中的竖向地震反应谱

介绍了民用核设施抗震设计的基本要求.从地震反应谱的基本概念入手,分析了核安全审评中应该关注的厂址地震反应谱和设计地震反应谱.从区分绝对反应谱和相对反应谱入手,剖析了竖向地震反应谱和水平地震反应谱之间的区别及联系,并就我国目前实际情况,对厂址竖向地震反应谱的确定提出了建议.     

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。     

核电厂电气厂房地震响应敏感性分析

利用人工地震波生成算法,探讨考虑土壤-结构相互作用的核电厂电气厂房地震响应动力分析模型和计算方法。通过比较楼层反应谱,研究岩土材料参数和载荷的不确定性对结构响应的影响。结果表明:岩土材料参数对核电厂电气厂房地震响应的影响更大,单一岩土材料参数下计算得到的拓宽后的楼层反应谱不能完全包络参数变化带来的地震响应差别。即使最终的反应谱大于或等于各种不同岩土参数下的楼层反应谱,仍有必要对不同岩土参数下的楼层反应谱做包络。     

反应堆中子注量率谱的谱参数法解

按照费米年龄理论,提出包含6个待定参数的数学表达式来描述反应堆堆内空间某点完整的中子注量率谱,并由此使用非线性规划可变容差法建立了新的解谱方法和相应解谱程序SNAIL。对于热中子反应堆内部的中子注量率谱,根据测量的反应率数据计算得到的整个能区中子注量率谱结果很满意,与NEUSPAC解谱程序解得的中子注量率谱符合得很好。对于中能中子场和快中子反应堆,在未使用可裂变材料探测器情况下,计算结果与文献公布的结果基本符合。     

三向地震载荷下HTR-10厂房土壤-结构相互作用分析

土壤-结构相互作用（SSI）会影响核电厂厂房的地震响应。本文充分考虑SSI效应的影响，对10 MW高温气冷堆（HTR-10）厂房在三向地震载荷下的响应进行了分析。建立了土壤-结构耦合有限元模型，通过构造人工边界实现对地震波在无限域内传播过程的模拟，并对模型的准确性进行了验证。利用该模型计算了HTR-10厂房的地震响应，并对不同楼层的反应谱计算结果进行了分析。对于水平向反应谱，各楼层的反应谱谱型类似，SSI影响规律基本一致。在竖直方向上，结构的响应特点与楼板自身的竖向频率特性有明显关系，不同楼板的响应差别较大。一般情况下，SSI效应对竖向响应有抑制作用，且随着楼层增加更为明显。当楼板与土壤的固有频率接近时，竖向响应与其他楼层相比会有显著放大。     

考虑SSI效应的核电站泵房结构楼层反应谱分析

采用Super SAP和CLASSI程序对某在建核电站泵房结构进行了极限安全地震震动和运行安全地震震动情况下的土壤-结构相互作用(SSI)的地震分析,揭示了结构在时域内的特性;通过傅立叶变换(FFT)分析了结构在频域内的特性,求得建筑结构中各楼层反应谱,结合核电厂设计规范分别给出了在较硬地基岩土条件下,结构考虑和不考虑SSI时各楼层反应谱,并对其进行了比较分析.结果表明,SSI效应对结构楼层反应谱的谱形、谱值以及零周期平台高度有一定的影响.     

大亚湾核电站反应堆厂房楼层反应谱分析评估

大亚湾核电站核岛厂房的抗震分析遵循技术输出国-法国M310型机组的土建技术规范RCC-G,采用简化的阻抗函数法计算地基岩土的作用.根据大亚湾厂址的地基岩土特点,拟采用更为精确的三维连续半空间边界子结构法来考虑地基岩土的作用,并与原设计进行对比.另外,在原设计中采用多组时程作为地震输入,取各组计算结果的平均值作为设计值的基础(称为"平均"法).在研究中基于相同的时程,拟分别采用"平均"法和更为常用的"包络"法,处理多组时程的响应.基于上述两方面,通过反应堆厂房的地震响应计算,得到核电站系统设备重要的设计基础数据-楼层反应谱(FRS),并将计算的楼层反应谱同设计谱进行比较,从而对设计方法及其结果进行评估,为电站的抗震设计裕量评估和安全管理提供可资参考的结论.     

试验堆主厂房楼板谱计算与比较

土壤－结构动力相互作用分析是核工程抗震设计与安全分析中的重要环节。核在地震动中的反应主要取决于地震特性、地基土特征与结构特性三个方面。对于非岩石地基上,由于构筑物基础面的运动受到土壤－结构动力相互作用的影响,与地表自由场运动会有显著的差别。在抗震设计规范中要求考虑这一相互作用的影响。本文采用FJUSH、SASSI1000程序进行了试验堆土壤－结构的动力分析,给出堆厂址地表自由场运动与各楼层的反应谱,以供结构与设备抗震设计之用,并进行了比较分析。     

Seismic response analysis of deeply embedded nuclear reactor buildings considering frequency-dependent soil impedance in time domain

In order to estimate the seismic behavior of deeply embedded nuclear power buildings, it is important to accurately transform the soil impedance in the frequency domain to the impulse response in the time domain. Although the transform is important for some nuclear buildings because they are deeply embedded in the soil, there are few practical and accurate methods at present. The author has proposed practical transform methods. In this paper, seismic response analyses considering frequency-dependent soil impedance in the time domain are shown. First, the formulation of the proposed transform methods is described. Then, the response analysis of a nuclear reactor building deeply embedded in inhomogeneous soil was performed considering the full matrix soil impedance as the example of practical problems. Through these analyses, the validity and efficiency of the methods were confirmed.     

Development of the site-specific uniform hazard spectra for Korean nuclear power plant sites

A seismic risk analysis has been performed to evaluate the seismic safety of a nuclear power plant for strong earthquakes beyond a design earthquake level. A site-specific median spectrum has generally been used for a seismic fragility analysis of structures and equipment in Korean nuclear power plants as a part of a probabilistic seismic risk assessment. The site-specific response spectrum, however, does not represent the same probability of an exceedance over entire frequency range of interest. The site-specific uniform hazard spectrum (UHS) is more appropriate for use in a seismic probabilistic risk assessment (SPRA) than the site-specific spectrum, since there are only a few strong motion data and seismological information for the nuclear plant sites in Korea. In this study, the uniform hazard spectra were developed using the available seismic hazard data for four Korean NPP sites.     

Improvement of the seismic safety of existing nuclear power plants by an increase of the component seismic capacity: A case study

This case study produces the scenario earthquakes for an example nuclear power plant (NPP) site and suggests the effective seismic capacity of safety-related equipment and components which significantly contribute to a core damage to improve the seismic safety of an existing NPP by using a probabilistic safety assessment. The response spectra for the scenario earthquakes show greater spectral accelerations than those for the design response spectrum in the frequency range higher than about 12 Hz. In order to improve the seismic safety of an example NPP, the effects of the seismic capacity of safety-related equipment and components on the core damage frequency (CDF) are investigated, and their effective seismic capacities are determined. The results of the case study show that an increase of the seismic capacity of the equipment reduces the CDF considerably. The effective seismic capacities for the diesel generator, offsite power, condensate storage tank and battery rack are determined as 0.84, 0.35, 0.63 and 0.63 g, respectively.     

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。     

The seismic reliability of VVER-1000 NPP prestressed containment building

The failure probability assessment of the containment building is an essential feature of the Level 2 PSA studies of nuclear power plants. The primary purpose of this paper is to demonstrate the methodology of evaluating containment seismic induced probability of failure without containment pressurization. The Loviisa, Finland site is one of the most seismically stable in the world and the numerically evaluated seismic induced failure probabilities are not representative for other sites. In addition, the containment concept described in this paper is not the typical Russian design which uses helical tendons in the cylindrical part of the structure and has a ring girder at the spring line of the structure. So the conclusions reached are applicable only to the containment configuration described in the paper. The geometry of the containment was determined by its preliminary design. The seismic hazard of the plant site was assessed during Level 1 PSA of the Loviisa plant. The initial information for seismic fragility analysis of the containment is the seismic response of the structure. The structural model for response analysis was the stick model. The stress analysis of the containment was carried out using the shell element model. The fragility evaluation of the containment was performed with the PROSAN-program. The structure was modeled as a parallel system consisting of the most heavily stressed elements. The resulting fragility curve gives the conditional probability of failure as a function of peak ground acceleration. The seismic hazard and the fragility were convolved to obtain the annual nonexceedance probability distribution for the collapse frequency of the structure.     

核电厂地震PSA中的风险定量化研究

论述了核电厂地震概率安全评价（PSA）定量化方法和工具的现状,指出了定量化工具面临的挑战和存在的问题。根据定量化的概率论本质,提出了计算方法。以我国某核电厂厂址多方案概率地震危险性分析（PSHA）结果和核电厂地震响应分析给出的最小割集为例,展示了计算方法的应用过程,分析了地震动参数和置信度参数对定量化计算结果的影响。结果表明,针对置信度参数进行拉丁超立方采样,采样次数较小时即可给出地震导致的核电厂堆芯损坏频率（SCDF）的稳定估计值;通常情况下,设备失效对SCDF的贡献最大,厂房失效的影响相对较小;地震动年发生率对SCDF的贡献需要根据工程场地的位置进行具体分析。      

利用HHT产生与多阻尼楼板谱匹配的地震波

美国原子能管理委员会（USNRC）规范规定了用于核电厂抗震分析和设计的地震波要求。在抗震分析和设计中，采用的地震波可与多阻尼目标反应谱匹配，也可与单阻尼目标反应谱匹配。然而，在对核电设备和部件进行动力时程分析时，则需要与多阻尼目标楼板谱匹配的地震波。基于此问题，提出利用希尔伯特-黄变换（HHT）方法，通过修改种子地震波的频率和振幅信息，使之与多阻尼目标楼板谱匹配，且完全符合USNRC规范的匹配标准，从而为核电设备和部件的地震安全评估提供合适的地震激励。      

Discussion of the paper: “Dynamic decoupling of secondary systems, by A.K. Gupta and J.M. Tembulkar, in: Nucl. Engrg. Des. 18 (1984) 359”

Technical aspects of seismic isolation systems show merit for their use in nuclear power plants. Less quantifiable non-technical aspects must be evaluated in the decision to employ a seismic isolation system.First, non-technical aspects are discussed. An historical and applications perspective is given, and it is suggested that the number of applications of seismic isolation systems is correlated with the amount of research activity in this area. For nuclear plants, it is suggested that application of seismic isolation systems is in part related to standardized plant designs in high seismic regions. Also, for nuclear plants, it is suggested that direct capital cost, enhanced seismic safety, regulatory licensing and unknown locations of nearby active faults are all factors which can weigh in favor and/or not in favor for seismic isolation application.Second, technical aspects are discussed. The technical results show that seismic isolation reduces building response, and reduces floor response spectra/equipment response. These results combine in application to reduce seismic risk and thus enhance safety for nuclear plants.     

Seismic response analysis of soil-structure interactive system using a coupled three-dimensional FE-IE method

This paper proposes a slightly new three-dimensional radial-shaped dynamic infinite elements fully coupled to finite elements for an analysis of soil-structure interaction system in a horizontally layered medium. We then deal with a seismic analysis technique for a three-dimensional soil-structure interactive system, based on the coupled finite-infinite method in frequency domain. The dynamic infinite elements are simulated for the unbounded domain with wave functions propagating multi-generated wave components. The accuracy of the dynamic infinite element and effectiveness of the seismic analysis technique may be demonstrated through a typical compliance analysis of square surface footing, an L-shaped mat concrete footing on layered soil medium and two kinds of practical seismic analysis tests. The practical analyses are (1) a site response analysis of the well-known Hualien site excited by all travelling wave components (primary, shear, Rayleigh waves) and (2) a generation of a floor response spectrum of a nuclear power plant. The obtained dynamic results show good agreement compared with the measured response data and numerical values of other soil-structure interaction analysis package.