核电厂主设备开孔结构有限元分析方法对比研究

核电厂主设备会使用许多结构复杂、形式不一的开孔结构。为了探索其力学分析及评价方法,对开孔结构进行有限元分析方法与等效解析方法对比研究。研究中考虑了影响开孔结构计算分析的主要因素(包括各种开孔形式、排列方式等),系统地梳理出开孔结构的有限元分析方法。     

AP1000核电厂抗震Ⅰ类大型结构模块吊装分析研究

模块化设计和建造方法是AP1000第三代核电厂特点之一,但同时也面临着大型模块的吊装挑战。本文基于ANSYS大型有限元计算软件,对AP1000核电厂中大型结构模块CA01进行了吊装工况下的有限元分析,计算了结构的应力和变形,并对构件进行了规范验算。计算结果表明,在合理设置临时支撑的情况下,CA01模块能够安全吊装就位,为大型结构模块的顺利吊装提供强有力的理论支撑。     

考虑液晃效应的大型储罐有限元抗震分析方法

在美国土木工程协会规范ASCE-4-98的基础上,扩展建立了大型储罐三维有限元模型,采用质量-弹簧模型模拟液晃效应,应用ANSYS软件进行抗震分析。有限元计算的液晃频率与ASCE-4-98规范公式计算结果一致,验证了有限元模型的合理性和可用性。该方法可直观地从三维模型上获取大型储罐在地震下的应力分布,是一种简单适用的工程处理方法。     

快堆堆芯抗震实验研究与有限元分析进展

堆芯的安全评价是快中子增殖反应堆抗震设计的一个重要问题。发生地震时,应该确保堆芯组件的结构完整性和核电厂能按要求紧急停堆。数百根堆芯组件之间存在着间隙,组件与堆芯支承处也存在间隙,整个堆芯被液钠包围,堆芯的抗震计算比较困难。本文重点介绍近年来法国、日本、意大利以及中国等国家针对快堆做过的一系列实验和理论研究进展情况。     

核电厂抗震系统的耦合性分析

核电厂设施是由构筑物、管系、设备和部件(SSCs)等组成的十分复杂的系统,抗震I类设施的抗震设计分析是在安全停堆地震(SSE)设计基准事故下确保核电厂安全的重要措施之一。为了将核电厂中复杂的构筑物、系统、部件的抗震分析开展得全面、可靠又深入,最有效和通用的方法是在抗震分析中将整个构筑物、系统、部件合理地分解成若干抗震主系统和子系统。本文从将主、子系统简化为二自由度的基本振动原理出发,论证主、子系统解耦的条件,证实了美国核管会安全分析报告标准审查大纲3.7.2中提出解耦条件的依据。同时又论证了耦合系统中将子系统独立解耦并进行抗震分析时所必须满足的必要条件。本文的结论可为核电厂抗震设计分析工程师以及安全评审人员提供一个重要的设计分析及评审依据。     

基于人工透射边界的核电厂结构抗震分析

目前核电厂结构-地基地震动力相互作用模型仍局限为规范建议的简单集总参数模型,或以粘弹性边界为代表的基本有限元地基模型,缺乏精度相对较高的分析方法。本文将高精度的二次透射人工边界地基模型引入核电结构抗震的分析领域,针对厂房结构的复杂约束,采用隐显式方法求解,显式积分时域结果后处理等关键问题,基于ANSYS软件提出了模型解决方案,并以实例形式验证了本文方法与模型的适用性。结果表明,文中模型适用于核电厂建筑物动力响应的有限元分析。     

核电厂应急负压系统风阀的抗震分析

风阀是用以调节通风系统中气体流量和压力的常用设备,在核电厂中得到广泛使用。为确保核电机组结构安全,采用有限元方法(FEM)对核电厂应急负压系统中的风阀进行分析,在计算风阀结构固有频率的基础上,得到了风阀结构在地震作用下的应力及变形。分析结果表明,风阀的应力和变形均满足相关抗震规范的要求,风阀结构在地震作用下是安全可靠的。     

核电厂轴流风机抗震分析与评定

本文利用ANSYS程序,采用单点谱分析方法,对在地震、自重、内压等多种载荷组合作用下的核电厂轴流风机进行抗震性能计算分析,幵按照ASME-AG-1《核气体处理规范》和RCC-M《压水堆核岛机械设备设计和建造规则》对计算结果进行应力评定。结果表明,轴流风机的设计满足抗震要求,应力和变形均在允许限值内。     

核电厂用LSJH-31-1碘吸附器排架抗震分析与评定

利用ANSYS程序,采用谱分析方法,对在地震、自重、内压等多种载荷组合作用下的LSJH-31-1碘吸附器排架进行了抗震性能计算分析,并根据EJ/T 1116-2000"核设施通风空调和气体处理系统机械设备设计规范"进行了应力评定.结果表明,排架的设计满足抗震要求,应力和形变均在允许限值内.     

AP1000核电厂OLP型预埋件有限元分析方法研究

作为世界上最先进的第三代核电技术,AP1000首堆正在中国建造。AP1000核电站厂房的一大特点是模块化程度高,以钢板混凝土模块墙结构取代传统的钢筋混凝土结构,模块墙上需要布置大量的OLP(Overlay Plate)型预埋件以连接其它结构构件,比如支撑工艺管道、设备支架、操作平台、预制构件等,因此预埋件的设计是AP1000结构设计中十分重要的环节。本文介绍了预埋件的有限元分析方法,将开发的GTStrudl命令流模板和基于Microsoft Excel环境下的VBA宏处理程序应用于预埋件的设计,显著提高了工作效率,对工程设计具有一定的帮助和借鉴意义。     

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.     

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.     

Evaluation of torsional effects in the seismic analysis of symmetric and unsymmetric nuclear power plant structures

This paper summarizes evaluation of torsional effects of symmetric and unsymmetric structures in the seismic analysis of nuclear power plants. The idealization of structures to predict the response due to earthquakes is described. The determination of stiffness or flexibility matrices of structures, using an approximate method and finite element technique, is outlined. The evaluation of foundation-structure interaction for structures founded on soft soils, hard rocks, and piles is presented. The analytical methods of analysis such as time-history and spectrum approach are discussed.     

Effects of foundation rotation on seismic inertia forces of nuclear power plant structures

Effects of foundation rotation on seismic inertia forces are studied in two different plane strain finite element investigations. A three-mass model is used to approximate the dynamic characteristics of a containment vessel of a nuclear power plant and its internal structure. In the first study a large finite element mesh is specified in order to eliminate unwanted reflections from the boundary. Motion input to the structure is limited to 2 sec duration. A quiet boundary technique is employed in the second investigation. As a result, earthquake motion inputs of any duration could be specified. Most results are based on portions of two recorded earthquake motions of 4 sec in length. Effects of foundation rotation and lateral soil-structure interaction are evaluted with program output. Results are presented in graphs and tables.     

Criteria for seismic analysis of nuclear plant structures and substructures

This paper encompasses criteria used for seismic analysis of nuclear power plant structures such as supporting structures founded on ground, as well as substructures. Nuclear power plant equipment and systems can be treated as substructures. Modeling of structures and substructures is described. Since instructure response spectra play an important role in the design and analysis of nuclear power plant equipment, systems and components, methods for development of instructure response spectra as well as variations of input parameters considered in determining these spectra are described.When the principal contribution to the equipment response is due to flexibility of the supporting substructures, an analytical approach to the problem for obtaining reduced stiffness and associated mass matrices of supporting substructures with finite element representation for use in the dynamic analysis of equipment and supporting structures is presented. When supporting structures and equipment, that have inherently different damping properties, are included as intergral parts of the dynamic models, the approximate evaluation of the modal damping based on the weighted damping according to the modal energy stored in each component is outlined. Use of time history and response spectrum analyses is presented. The effects of relative displacements due to different motion of the support points of substructures in each significant mode of the supporting structures as well as procedures of combining modal responses are detailed.     

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.     

Methodologies for rapid evaluation of seismic demand levels in nuclear power plant structures

A methodology for rapid assessment of both acceleration spectral peak and “zero period acceleration” (ZPA) values for virtually any major structure in a nuclear power plant is presented. The methodology is based on spectral peak and ZPA amplification factors, developed from regression analyses of an analytical database. The developed amplification factors are applied to the plant's design ground spectrum to obtain amplified response parameters. A practical application of the methodology is presented.This paper also presents a methodology for calculating acceleration response spectrum curves at any number of desired damping ratios directly from a single known damping ratio spectrum. The methodology presented is particularly useful and directly applicable to older vintage nuclear power plant facilities (i.e. such as those affected by USI A-46). The methodology is based on principles of random vibration theory. The methodology has been implemented in a computer program (SPECGEN). SPECGEN results are compared with results obtained from time history analyses.     

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.     

Study of ultimate seismic response and fragility evaluation of nuclear power building using nonlinear three-dimensional finite element model

The probabilistic safety assessment (PSA) is important for nuclear power buildings in Japan because the risk of the occurrence of seismic ground motions beyond the design assumption cannot be denied. In this paper, the building fragility of the seismic PSA was evaluated using a high accuracy analysis model (three-dimensional nonlinear FEM building model considering soil-structure interaction and basemat uplift behavior). First, the response analyses were conducted increasing the input acceleration up to 3500 Gal, until the damage of the building reached the ultimate condition. The damage of the building was estimated from the shear strain, the axial stress, and the consumed strain energy of the shear walls. Then, the influence on the response given by the vertical ground motion and the basemat uplift was evaluated. In addition, considering the shear destruction of the web wall and compressive crash of the flange wall as the fracture modes, the building fragility was evaluated. As a result, it was shown that the investigated method is efficient for more accurate seismic PSA estimation.     

有限元在杆式支撑结构力学分析中的应用

介绍有限元分析软件ANSYS在一种杆式支撑结构设计中的应用,通过模拟各种载荷工况对结构的影响分析,对HT-7U超导托卡马克冷质部件支撑结构初步设计方案做进一步的优化和完善。