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。     

中国实验快堆堆内构件应力分析与评定

本文是中国实验快堆堆内构件主要部件的应力分析与评定汇总报告.主要构件包括堆内支承结构、堆芯支承结构、堆内热屏蔽等7类设备.堆内各部件采用有限元方法按其特点进行整体分析或部件分析.文章首先建立结构的计算模型,然后,对有限元计算模型进行在自重、流体流动压差、冷却剂流动引起的结构振动和温差载荷条件下的静态分析计算和结构的模态分析以及地震载荷下的动态分析.最后,按规范要求对堆内各结构在承受的各种载荷条件下进行载荷组合与评定.     

中国实验快堆新燃料组件装载机的抗震分析

中国实验快堆新燃料组件装载机是一机构部件,结构复杂,装载机的许多运动件和传动件难以用有限元进行精细模拟。本文引入了一种简化有限元计算的思路,即针对装载机主要部件的主体框架建立简化的有限元模型,通过反应谱分析,计算出各个主要部件在地震条件下的加速度响应。采用手算方法对主要的连接螺栓进行力学计算,并根据相关法规对螺栓进行评价,以确保总体结构的完整性。     

反应堆结构三维非线性抗震分析研究

对反应堆结构抗震分析软件进行研究,编制必要的接口程序和载荷组合及敏感性分析程序,将抗震分析中的商用软件、国外引进软件和编制的载荷组合及敏感性分析程序等组合成一个软件系统。对反应堆结构三维非线性抗震分析的建模技术和分析技术进行研究,对采用三维非线性时程法进行反应堆结构抗震分析的必要性和可行性进行研究。建立反应堆结构三维非线性模型,利用ANSYS软件采用时程法中的直接积分法完成了反应堆结构抗震分析计算,给出了应力分析评定所需的载荷和控制棒驱动机构抗震鉴定试验所需的加速度时程。     

基于参数化-超单元法的HFETR乏燃料贮存格架抗震分析

为解决HFETR乏燃料贮存格架结构抗震设计中存在的结构迭代设计速度较慢、抗震计算时间较长、多格架分析困难等问题,提出将参数化方法和超单元法组合应用于HFETR格架抗震分析。基于超单元法建立了格架有限元模型;利用APDL和VB对有限元模型的建立和计算进行了参数化处理;对HFETR乏燃料贮存格架进行了多格架抗震计算,并对格架在满载、半载、空载3种装载情况下的应力、滑移和倾倒进行了分析。结果表明：所设计的HFETR格架抗震性能满足要求;参数化-超单元法可用于乏燃料贮存格架的迭代设计和多格架抗震分析。     

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。     

先进压水堆一回路地震反应谱分析及参数敏感性研究

压水堆一回路系统包含压力容器、蒸汽发生器、主泵、稳压器、主管道和波动管等重要部件,各部件在地震激励下的动态响应与整个系统的结构形式密切相关。本文从系统的角度,以非能动先进压水堆一回路为研究对象,运用ANSYS建立了其三维有限元模型,在模态分析的基础上,进行了三正交方向输入下的反应谱分析,得到了系统在地震载荷下的响应。并对反应谱输入角度和支撑刚度进行了敏感性研究,给出了这些特性参数对结构设计和分析的指导性意见。此外,通过直接积分法得到系统的地震时程响应,并与谱分析的模拟结果进行了对比,同时也为主泵等单个部件的详细地震分析提供位移、加速度输入。最后通过三维实体模型与集中质量模型抗震计算结果的比较,说明建立三维实体模型的必要性。本文为核电站一回路重要设备的结构分析提供了技术支持。     

基于DSEM的核电厂结构-土-结构相互作用模型

核电厂结构的抗震能力是保障安全的重要内容,而考虑结构-土-结构相互作用(SSSI)的影响效应是必要的且意义重大。本文以阻尼溶剂抽取法(DSEM)为理论基础,考虑相邻工程结构与无限土体的动力特性,利用位移协调与力平衡机制,建立了相邻结构-土体相互作用计算模型,给出了具体数值实现公式,并通过UPFs二次开发功能在通用有限元程序ANSYS中实现该模型的嵌入。进而,以国内某核电工程为例,建立一系列SSSI系统的三维模型,并就不同的地基条件、埋置效应对核电厂反应堆SSSI规律的影响进行探讨,结果可为类似核电厂址地基的抗震适应性分析及优化设计提供借鉴与参考。     

余热排出热交换器抗震分析中实体和壳体单元的组合应用

余热排出系统是压水堆中重要的安全系统之一,需要对该系统中的主要设备进行详细的抗震分析。采用实体单元和壳体单元组合对余热排出系统的余热排出热交换器建立模型,对设备应用多点响应谱法进行详细的应力分析。使用RCC-M规范评价分析结果,表明应用此方法对余热排出热交换器设备建模计算方便合理,可减少计算时间。     

供热堆控制棒驱动线试验模型动力分析

对200MW低温核供热堆的控制棒及其相关部件（包括燃料组件等）试验模型进行抗震分析,包括有限元模型的简化、建立模型、模态分析、响应谱分析及时程响应分析等。通过分析得到结构的基频以及不同因素对基频的影响;根据频谱响应分析和时程分析的结果,得到结构受到地震载荷冲击时的最大位移响应和加速度响应。     

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。     

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.     

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.     

核电厂管线上设备地震加速度响应放大系数评估方法与应用

为了方便快捷地确定核电厂管线上设备的地震载荷,基于弹簧-质量模型、管线与安装于管线上的刚性设备的关系,提出了评估管线对设备的地震加速度响应放大系数的方法,以及该评估方法在确定设备抗震鉴定地震输入方面的应用方法,并通过有限元模型对该评估方法和应用予以了验证。结果表明,该评估方法在确定管线上设备抗震鉴定地震输入方面提供了易于操作又不至于太过保守而降低经济性的可行的应用办法。      

核电站环形吊车抗震计算分析

应用有限元分析软件ANSYS建立了核电站环形吊车结构的三维计算模型,在模态分析的基础上,以环形吊车所在的安全壳标高40.0 m处的地震反应谱作为输入,对环形吊车结构进行了地震响应分析计算.计算结果表明,地震动作用下环形吊车的垂直位移和应力响应比较小,但水平位移和应力响应比较大,原因是环形吊车水平方向1阶弯曲振动固有频率位于水平地震反应谱最大值频率区间附近;环形吊车结构在地震动作用下能满足抗震设计强度要求,应力集中处的最大应力小于材料屈服极限.     

Elastic–plastic analysis of nuclear structures with millions of DOFs using the hierarchical domain decomposition method

The hierarchical domain decomposition method (HDDM) proposed by Comp. Sys. Eng. 4 (1993) 495 is applied to the large scale elastic–plastic finite element (FE) analysis of nuclear structures. The HDDM is a method to implement the finite element method (FEM) on various kinds of parallel environments. The substructure-based iterative methods can effectively be used with the HDDM to solve the large scale linear algebraic equations derived from the implicit FEM. In this paper, some key techniques to parallelize the static elastic–plastic FE analysis by the HDDM are described. As illustrative examples, a support structure of the high temperature engineering test reactor (HTTR), a pressure vessel, and an internal pump of a pressure vessel are analyzed. The structure of HTTR and the pressure vessel are modeled by hexahedral solid elements whose total degrees of freedom (DOFs) are about 1.3 millions (M) and 3 M, respectively. The internal pump is modeled by quadratic tetrahedral elements whose total DOFs are about 2 M. The elastic–plastic analysis of a simple cube with 10 M DOFs is also carried out. Both the conjugate gradient method for solving the linear equations and the Newton–Raphson method for solving nonlinear problems successfully converge.     

Seismic analysis of RCS with finite element model for advanced PWR

Reactor Coolant Pumps (RCPs) are very important to the safe operation of Nuclear Power Plants (NPPs), especially during the earthquake, which needs detailed seismic analysis of individual RCPs and the boundary conditions, for example, at the nozzles. In this paper, three-dimensional finite element model of Reactor Coolant System (RCS) is constructed from a systematic perspective to perform dynamic evaluation, in which the boundary conditions could be given. The seismic spectrum analysis with three orthotropic directions is performed to obtain the stress and displacement response, which shows that the maximum Tresca stress locates in the connection part of SG with RCP and the maximum displacement occurs at the surge line. Sensitivity analysis of spectrum input angle and stiffness of supports is performed, which may be useful to further design and analysis. Furthermore, direct integration method is used to perform time-history analysis, and the boundary conditions of RCP, the loads, acceleration and displacement at nozzles are obtained, which could support the detailed analysis of RCP components. Besides, the lumped mass model of RCS is also constructed to compare with three-dimensional finite element model, which means that for the complicated geometry the 3-D model is better than the lumped mass model.     

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.