A study on the connection effects of subsurface slabs to exterior walls in the KNGR seismic analysis

The soil-structure interaction (SSI) analyses are being performed for the Korean next generation reactor (KNGR) design, because the KNGR is developed as a standard nuclear power plant concept enveloping various soil conditions. The SASSI program, which adopts the flexible volume method is used for the SSI analyses. Since the impedances are computed based on the assumption of a rigid foundation in the KNGR design, the impedances are computed at a single point around the center of the basemat. The base model used in the SSI analyses assumed that the subsurface slabs are free and not connected to the exterior walls. This results in a more flexible structure than reality. This study is performed to address the concern of the potential dynamic effect of this modeling assumption on the overall structural response. The super-structure model is modified with the subsurface portions of the sticks being laterally rigid, so that the entire subsurface cavity together with the super-structure model acts laterally rigid. For the comparison, the in-structure response spectra using the rigid subsurface model are compared with the original spectra without considering the connection effects.     

考虑地基岩土参数不确定性的核电厂结构随机地震反应分析

在考虑土-结构相互作用(SSI)效应的情况下,引入随机地震反应分析方法,探讨地基岩土参数的不确定性对核电厂地震响应的影响.基于ANSYS程序,采用常数阻抗法,通过设置边界弹簧和阻尼来考虑地基土的作用,并通过设置弹簧和阻尼参数的不确定性,来模拟岩土动态参数的不确定性.针对某1000MW级压水堆核电站反应堆厂房结构,进行随机地震反应的数值仿真分析,并将随机反应结果与确定论分析结果进行了对比.结果表明,随机分析方法是确定论分析方法的有益补充,二者结合能更合理地反映参数的不确定性对结构地震响应的影响.     

The frequency-dependent elements in the code SASSI: A bridge between civil engineers and the soil–structure interaction specialists

After four decades of the intensive studies of the soil-structure interaction (SSI) effects in the field of the NPP seismic analysis there is a certain gap between the SSI specialists and civil engineers. The results obtained using the advanced SSI codes like SASSI are often rather far from the results obtained using general codes (though match the experimental and field data). The reasons for the discrepancies are not clear because none of the parties can recall the results of the “other party” and investigate the influence of various factors causing the difference step by step. As a result, civil engineers neither feel the SSI effects, nor control them. The author believes that the SSI specialists should do the first step forward (a) recalling “viscous” damping in the structures versus the “material” one and (b) convoluting all the SSI wave effects into the format of “soil springs and dashpots”, more or less clear for civil engineers. The tool for both tasks could be a special finite element with frequency-dependent stiffness developed by the author for the code SASSI. This element can represent both soil and structure in the SSI model and help to split various factors influencing seismic response. In the paper the theory and some practical issues concerning the new element are presented.     

Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.     

设计基准内压下混凝土安全壳的有效预应力作用研究

核电厂安全壳建造过程中大量采用预应力技术,预应力在设计基准内压下的分布状况、损失规律直接影响到安全壳结构的耐久性。介绍了某核电厂安全壳结构和预应力系统的布置情况和预应力损失的分析过程,以闸门洞口附近水平预应力钢柬为例进行了预应力损失计算,同时计算了5年打压试验时安全壳结构的有效预应力。基于以上分析,利用ANSYS程序建立预应力混凝土安全壳有限元模型进行结构计算,对设计基准内压下的有效预应力作用进行了总结。结果表明,预应力系统承担了打压试验下大部分设计内压,安全壳整体结构是安全的,这些结论与安全壳的预应力系统设计理念一致,可供工程设计人员参考。     

基于地基土等效线性的三维厂址反应分析

随着核电厂选址条件的日趋复杂化,土-结构相互作用（SSI）成为核电厂抗震分析需要考虑的重要因素之一。目前经典的自由场厂址反应分析采用的是一维层状地基土的分析,比如SHAKE91、EERA和SASSI等,很难考虑土层的非均质层状因素。因此随着核电安全的监管要求越来越高,抗震的精细化分析成为趋势。本文采用有限元程序ABAQUS编写的UMAT材料子程序,实现了基于地基土材料的等效线性,开展均质层状土的三维自由场厂址反应分析。其计算分析结果与SHAKE91计算结果进行对比表明,在均质层状土条件下吻合较好。因此,本研究为求解复杂非均质地基条件的SSI分析提供了良好的工程适用性。      

Deterministic and stochastic earthquake response analysis of the containment shell of a nuclear power plant

Response of the containment shell of a nuclear plant to earthquake ground motion is considered. A finite element model of the structure is developed and SAP IV structural analysis program is employed for the determination of the frequencies and the corresponding mode shapes of the structure. The response of the containment shell to several past earthquakes are analyzed and the results are discussed. Stochastic models of earthquake ground acceleration are then considered and the general expressions for the power spectra, cross correlations and the mean-square responses are derived. The root mean-square of the relative displacement responses of various nodal points of the containment shell structure subjected to stationary as well as nonstationary random support motion are evaluated. The stochastically estimated maximum displacement responses are compared with those obtained from a deterministic analysis and reasonable agreements are observed.     

Simplified method for evaluating the seismic buckling capacity of unstiffened steel containment structures

A simplified method is presented for evaluating the seismic buckling capacity of unstiffened, free-standing steel containment structures. The method is consistent with current US Nuclear Regulatory Commission seismic design standards and with containment buckling interaction equations given in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code which includes the influence of geometrical imperfections of the shell on buckling. Stresses to be considered in the interaction equations are determined from beam theory using standard response spectrum analysis. An empirical correction factor is developed to account for hoop stresses that are not explicitly represented in the beam theory. As the results of these analyses are very sensitive to the damping that is assumed, the extensive three-dimensional finite element analyses that were performed to develop the hoop stress reduction factor were also used to study the sensitivity of containment buckling to the assumed damping. Experiments on model containment structures were then performed to further investigate the damping properties exhibited by these structures. The study in concluded by showing that the simplified method reasonably predicts seismic buckling capacities when compared with independently determined predictions from detailed finite element analyses.     

基于ANSYS的核电厂安全壳结构非线性有限元分析

对核电厂预应力混凝土安全壳结构进行了内压作用下的非线性有限元分析.详细介绍了ANSYS中的混凝土单元SOLID65及混凝土材料的本构关系,并对非线性求解过程中影响收敛的因素进行了分析；同时,以福清核电厂5、6号机组内层安全壳为工程实例进行有限元计算.结果表明,15 m至30 m标高范围内的径向位移大于其他高度的径向位移,标高25 m左右径向位移最大；内压加至0.42MPa,模型结构仍处于受压状态,满足使用要求.分析表明,福清核电厂5、6号机组安全壳结构在设计内压作用下是安全的,可为安全壳整体性试验提供参考.     

Soil-structure interaction in explosive testing of model containments

The SIMQUAKE series of field tests on model containment structures provides a data base which may be used to validate analytic models of soil-structure interaction (SSI). In addition, the test produced significant evidence of nonlinear rocking response of 1/4-embedded model structures due to debonding and rebonding at the soil-structure interface. This paper describes the SIMQUAKE II test and an analytic method for soil-structure interaction based on explicit finite difference techniques, which consider both nonlinear behavior at the soil-structure interface and nonlinear constitutive behavior of the site. The analytic method is applied to two-dimensional analyses of both 1/8- and 1/12-size structures and results are compared with test measurements. Structural rocking response is shown to be sensitive to the inelastic compaction characteristic of the soil adjacent to the structure. A three-dimensional pretest analysis of the SIMQUAKE II test is also briefly reviewed.     

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。      

混凝土安全壳整体性能试验峰值压力持续时间探讨

核电站建造阶段必须进行安全壳整体性能试验（CTT）,验证在设计基准事故时安全壳结构的完整性。本文针对某核电厂3号机组预应力混凝土安全壳CTT进行非线性有限元分析。结果表明：筒体闸门洞口标高附近径向变形最大,预应力钢束承担了峰值压力0.483 MPa作用下大部分设计内压,安全壳整体结构处于受压状态,与实际试验状态基本吻合。同时,对国内外法规标准关于安全壳峰值压力持续时间的规定进行总结,提出相关结论及建议,可为安全壳CTT方案设计提供参考。     

Lotung large-scale seismic experiment and soil-structure interaction method validation

This paper describes the Electric Power Research Institute's (EPRI) current on-site large-scale soil-structure interaction (SSI) research. The objectives of the research are: (1) to obtain an earthquake database which can be used to substantiate SSI models and analysis methods; (2) to develop realistic SSI analysis guidelines and procedures based on experimental-analytical correlation; and (3) to quantify nuclear power plant reactor containment and internal component's seismic margin based on earthquake experience data.To meet the objectives, two model structures were sited in a high seismic region, Lotung, Taiwan, under the joint sponsorship of EPRI and the Taiwan Power Company (Taipower). The model structures which simulate scaled-down nuclear containments ( ) were constructed and instrumented within an existing strong motion array (SMART-1) deployed by U.C. Berkeley under a U.S. National Science Foundation grant. The instrumentation layout associated with the model structures included accelerometers on the models, on the internal components, on the ground surface, and in the ground. These instrumentations are to provide information required for validation and qualification of SSI models. Between September 1985 and November 1986, 18 earthquakes ranging from Richter magnitude 4.5 to 7.0 were recorded.The analysis phase of the research was conducted with the cooperation of the U.S. Nuclear Regulatory Commission (NRC) and Taipower. A round-robin approach was utilized with emphasis on blind predictions and independent assessment of existing methodologies. A total of 13 research teams from the United States, the Republic of China, Japan, and Switzerland participated in the effort. A workshop was held in December 1987 where research results and findings were presented. Further effort is ongoing to synthesize the results and findings for providing technical bases of developing improved SSI analyses guidelines and procedures.     

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.     

Analysis of a stiffener intersection eccentricity in a steel containment

Containment structures have several regions in which the continuity of the cylindrical pressure boundary is interrupted, e.g., shell penetrations, discontinuous stiffeners, and changes in the shell thickness. Significant strain concentrations can occur in these areas of discontinuity. The Sandia National Laboratories 1:8-scale steel containment equipment hatch was analyzed as an example of an eccentricity at a stiffener intersection.A portion of the as-built 1:8-scale model was modeled with the ANSYS general purpose finite element program using triangular, thin shell finite elements. The overall size of the model was determined from Saint-Venant type considerations of the stress field around the hatch. Shell elements were used to model the ring and formed stiffeners. Geometric and material nonlinear behavior were included. The model was loaded using discrete load steps up to a pressure of 165 psig. At this pressure, the maximum strain was 19.7 percent in the formed stiffener near its intersection with the ring stiffener. The finite element solution demonstrated the very localized nature of the strain field near the ring/formed stiffener intersection.In an attempt to reduce analysis costs, a small portion of the 1:8-scale model immediately surrounding the ring/formed stiffener intersection was selected for further analysis. Two smaller models, a ring/formed stiffener intersection and a ring/circular stiffener intersection, were studied. The models were significantly smaller than the regions used previously. A comparison of the two intersection models showed that the circular stiffener is a more efficient configuration.     

某核电厂LOCA下预应力混凝土安全壳响应规律初探

核电厂LOCA发生后,预应力混凝土安全壳结构内温度场分布具有明显的非线性特征,但现行的混凝土安全壳设计规范未对LOCA下温度和应力的组合作用提出具体的计算方法。基于用ANSYS程序建立的包含预应力钢束的混凝土安全壳结构的有限元模型,本文计算了LOCA下不同时刻安全壳壳壁内的温度场分布,并与理论值进行了比较,验证了计算模型的正确性。初步分析了高温、高压作用下安全壳结构变形的规律,总结了混凝土温度效应和预应力系统的作用,可为安全壳结构设计提供参考。     

核反应堆厂房结构楼层反应谱的敏感性分析

以某千兆瓦级压水堆核电站反应堆厂房结构为对象,研究了考虑土一结构动力相互作用的硬土场地条件下地基土动态剪切模量的变化对楼层反应谱计算的影响,定量分析了厂房结构楼层加速度反应谱对地基土动态参数变化的敏感性,从而为评估类似硬土场地条件下核反应堆厂房结构抗震安全性提供了一种可供参考的计算方法。     

大亚湾核电厂反应堆厂房地震响应分析评估

大亚湾核电厂核反应堆厂房的抗震分析基本沿用法国M310型机组的标准分析方法（RCC—G）,对于土-结构相互作用（SSI）效应的考虑,采用简化的阻抗函数法。本文拟采用新的相对精确的基于Green函数的三维连续半空间边界子结构法考虑地基岩土的作用,进行SSI耦合系统的地震响应分析计算,并将计算的楼层反应谱（FRS）同设计值进行比较,对设计方法及其结果的趋向性（偏于安全／或不安全）进行评估。结果表明,与基于三维连续半空间边界子结构法的计算结果相比较,电厂设计偏于安全。     

Sizing rules for the pressure design of pump casings

The paper presents the theoretical and experimental studies undertaken by AFCEN (French Society for Design and Construction Rules for nuclear island components) and SCP (French Union of Pump Manufacturers) in order to develop sizing rules for the pressure design of radially split pump casings.Shape factors used in simplified design formulas have been developed for different configurations using results of finite element analyses and experimental extensometric analyses.