反应堆结构的三维非线性地震分析

建立反应堆结构三维非线性有限元模型,采用总体瑞利阻尼、局部材料阻尼和弹簧单元阻尼相结合的方法解决部件间的阻尼差异问题,导出燃料组件的等效间隙与等效刚度计算公式并真实地模拟了带间隙的同心圆部件间的接触、动态变化的预紧力和摩擦效应。随后,以同时满足响应谱和功率谱包络性要求的非平稳地震位移时程作输入,开展反应堆结构的非线性地震分析,得到结构在安全停堆地震(SSE)下的地震响应,为反应堆结构的三维非线性抗震研究提供借鉴。     

核电反应堆在地震和失水事故下的结构动力响应分析

以核电厂反应堆在地震和失水事故下的结构动力响应分析为工程背景,对反应堆结构的动力响应分析开展了研究。分析研究形成了-套可用于新核电站反应堆结构设计的完整设计分析系统,介绍了该方法所使用的分析模型,明确了动力学分析模型中考虑的重要因素。克服了结构复杂性、载荷多样性,结构非线性、材料非线性、接触非线性和流固耦合等多重非线性因素,更加精确的模拟了反应堆结构,提高了计算精度。该方法对我国自主开展反应堆事故工况下的设计和安全分析具有良好的推广和应用前景。     

反应堆系统地震载荷下动力响应计算的不确定性分析

反应堆结构在地震载荷下的动力分析是反应堆系统安全设计的重要环节。反应堆系统的关键参数由于计算误差、制造工艺和安装过程中的随机和其他不可控误差影响,通常存在一定的不确定性。本文开展了反应堆系统中结构参数不确定性对地震载荷下系统动力响应的影响研究。应用最大熵原理,建立描述反应堆系统部件间接触刚度和间隙的概率密度分布函数。应用马尔科夫链蒙特卡洛采样（MCMC）技术,对系统关键参数进行采样,建立不同参数下地震响应的输入输出数据池,基于数据池考察反应堆系统不同位置和部件的动力响应分布。研究表明,上、下堆芯板附近的动力响应的不确定性存在不同分布。本文所述研究对评价反应堆系统地震分析模型的鲁棒性和动力响应结果的可靠性提供了定量分析手段。      

振动条件下平行孔板间流体的作用力实验

乏燃料贮存格架自由放置在乏燃料水池内,格架 格架、格架 池壁之间有一定间隙。在地震载荷下,这些间隙中流体的流固耦合作用耗散了结构的能量,保证了格架的结构完整性。根据AP1000和CAP1400系列反应堆型的格架储存腔设计,格架的侧壁有平板或孔板两种方案。对格架进行结构动力学分析时,为了简化流体间隙附加质量的计算,工程上一般将孔板直接简化为平板。这样的方法并不能精确反映出实际的流固耦合效应。为获得格架与格架、格架与池壁间流体的流固耦合特性,搭建实验台架,通过振动实验,测量孔板的间隙流体在不同激振频率、不同间隙条件下的流体作用力。实验最终给出孔板的间隙流体的流体力,并给出了不同间隙条件下附加质量,本文可为AP1000和CAP1400系列乏燃料储存格架的流固耦合参数选取提供依据。     

双弹性管流固耦合振动的数值模拟

为研究反应堆结构中诸如燃料棒、蒸汽发生器和其他换热器等管束类结构的流固耦合振动问题,利用有限体积法离散大涡模拟的流体控制方程及有限元方法离散结构动力学方程,结合动网格技术,建立了三维流体诱发弹性管束振动的数值模型,实现了计算结构动力学与计算流体力学之间的双向耦合。得到横流作用下单管的振动响应,并与已有的实验数据比较,证明了本文模型的合理性;对横流作用下的两串列管、两并列管的流固耦合振动进行了数值模拟,着重研究了节径比为1.2、1.6、2、3、4的两弹性管在不同流速作用下的动力学响应及流场特性;得到串列管、并列管的临界间距与临界流速。     

地震载荷下反应堆系统的不确定性量化

反应堆结构力学分析中,由于设计变更、制造安装、计算偏差等因素的影响,会导致力学分析关键输入参数存在一定的不确定性,这种不确定性将直接影响到动力响应、载荷分配与最终的力学评价结果。为量化参数不确定性对载荷计算的影响,本文采用不确定性量化的方法,以反应堆系统为研究对象,开展了地震载荷下系统关键结构参数对系统动力响应与载荷分配的不确定性量化研究。首先依据关键参数的基本特性,利用最大熵原理,建立了描述反应堆系统部件间接触刚度和间隙的概率密度函数。随后,应用马尔科夫链蒙特卡罗采样技术对系统关键参数进行采样,并通过有限元瞬态计算获得了输入输出数据池。最后,以样本数据为基础,考察了不确定性参数对部件动力响应统计分布的影响,开展了名义模型的可靠性与不确定性量化分析。研究发现,结构参数不确定性对系统响应的影响在不同部位、不同频域内呈现不同的分布。在考察名义模型的可靠性时应根据响应具体形式有针对性地进行量化。本文所提出的不确定性量化方法对核动力装置其他系统和设备的动力分析具有推广价值。     

地震载荷下反应堆系统的不确定性量化

反应堆结构力学分析中,由于设计变更、制造安装、计算偏差等因素的影响,会导致力学分析关键输入参数存在一定的不确定性,这种不确定性将直接影响到动力响应、载荷分配与最终的力学评价结果。为量化参数不确定性对载荷计算的影响,本文采用不确定性量化的方法,以反应堆系统为研究对象,开展了地震载荷下系统关键结构参数对系统动力响应与载荷分配的不确定性量化研究。首先依据关键参数的基本特性,利用最大熵原理,建立了描述反应堆系统部件间接触刚度和间隙的概率密度函数。随后,应用马尔科夫链蒙特卡罗采样技术对系统关键参数进行采样,并通过有限元瞬态计算获得了输入输出数据池。最后,以样本数据为基础,考察了不确定性参数对部件动力响应统计分布的影响,开展了名义模型的可靠性与不确定性量化分析。研究发现,结构参数不确定性对系统响应的影响在不同部位、不同频域内呈现不同的分布。在考察名义模型的可靠性时应根据响应具体形式有针对性地进行量化。本文所提出的不确定性量化方法对核动力装置其他系统和设备的动力分析具有推广价值。     

基于两种模型下的储水罐应力计算

分别采用弹簧质量模型和流固耦合模型2种方式对大型储水罐进行相互验证计算。弹簧质量模型基于HOUSNER理论,计算出脉冲液体质量M0、脉冲液体重心高度H0、对流液体质量M1、对流液体重心高度H1以及弹簧模拟对流液体晃动频率的等效刚度k1,并以此作为有限元模型的输入参数;流固耦合模型采用ANSYS的流体单元fluid80模拟液体,壳单元shell181模拟储水罐,考虑液体及结构阻尼,计算结构在地震时程下的响应。     

基于三维核热耦合方法的吸气式核反应堆瞬态特性研究

吸气式反应堆具有大比冲、高功率密度的优良特点,在未来空天推进领域具有广阔的应用前景.该反应堆在正常运行条件下需要完成各类功率快速响应的瞬态工况,并在极短时间内完成启停堆过程,采用传统的数值方法研究其瞬态三维核热耦合特性存在较大的困难.本文基于开源OpenFOAM平台,开发了三维瞬态核热耦合求解程序,提出了新型流固热耦合...     

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。     

Homogenisation method for the dynamic analysis of a complete nuclear steam generator with fluid-structure interaction

The present paper deals with the dynamic analysis of a steam generator tube bundle with fluid-structure interaction modelling. As the coupled fluid-structure problem involves a huge number of degrees of freedom to account for the tube displacements and the fluid pressure evolutions, classical coupled method cannot be applied for industrial studies. In the present case, the three-dimensional fluid-structure problem is solved with an homogenisation method, which has been previously exposed and successfully validated for FSI modelling in a nuclear reactor [Sigrist, J.F., Broc, D., 2007a. Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid-structure interaction coupling. Nuclear Engineering and Design 237, 431-440]. Formulation of the homogenisation method for general two- and three-dimensional cases is exposed in the paper. Application to a simplified, however representative, model of an actual industrial nuclear component (steam generator) is proposed. The problem modelling, which includes tube bundle, primary and secondary fluids and pressure vessel, is performed with an engineering finite element code in which the homogenisation technique has been implemented. From the practical point of view, the analysis highlights the major fluid-structure interaction effects on the dynamic behaviour of the steam generator; from the theoretical point of view, the study demonstrates the efficiency of the homogenisation method for periodic fluid-structure problems modelling in industrial configurations.     

Seismic model of the gas cooled fast breeder reactor core support structure

The modelling technique for the seismic analysis of the core support structure of a gas-cooled fast breeder reactor is developed. The core support structure consists of the support cylinder and a perforated grid plate to which 265 fuel and blanket elements are clamped as cantilevered beams. The analysis of the core support structure consists of three steps: (1) analysis of the grid plate, (2) analysis of the core elements, and (3) modal synthesis.The first step in developing a solution to the problem is to assume that the core elements (fuel and blanket) are attached to the grid plate as rigid rods. In this case the influence of the rigid rods can be represented by their masses and rotary inertias. The solution of this problem was developed by applying the dynamic theory of grid plates. This was accomplished by generalizing the Reissner-Mindlin thick-plate theory with orthotropic constants and then modifying the formulations of the rotary inertia expressions to include the rotary inertia effects of the core elements. The numerical results showed that the grid plate's fundamental frequency is in the range of the fundamental frequencies of the core elements so that a dynamic coupling effect exists. Because of this dynamic coupling effect the elastic properties of the grid plate must be included in the seismic analysis of the GCFR'The second step was to develop a mathematical model of the grid-plate core-element system using the method of Rayleigh-Ritz. In this model the elastic coupling effect of the core elements was included.For the final application of the theory, the exact solution of the elastic plate with rigid rods was simulated on the computer by applying the elastic rotary inertias of the core elements to the model of the grid plate. With this technique it is possible to model the grid plate with a reasonable number of fuel and blanket elements and to replace the missing core elements with their equivalent effective rotary inertias. The method includes the capability of modeling the different mass, damping and elastic properties of the fuel and blanket elements.Comparing the results of the present analysis with the preliminary simple spring-mass core model, the amplitudes of vibration obtained, in some cases in the present analysis, are a factor of ten smaller than was previously computed. Applying this more elaborate analysis will lead to a simpler and less expensive design.     

Impacts and fluid-structure interaction in pressurized water reactor safety analysis

A numerical method is described for the analysis of coupled three-dimensional fluid-structure motion with impacts between structural parts at rigid or flexible supports with small clearances. The method is used for the analysis of the blowdown loadings and the response of internal structures in the vessel of a pressurized water reactor (PWR) in the hypothetical event of a sudden break of a coolant inlet pipe. The method is a generalization of the existing code FLUX which simulates the three-dimensional fluid-structure motion by means of an implicit time integration scheme. The additional supports with clearances are taken into account by applying support forces to the freely moving fluid-structure system. The forces are determined such that the kinematic constraints are enforced at each time step. Numerically, these forces are determined efficiently using a precomputed influence matrix which defines the dynamic displacement per time step at each support due to a unit force at each other support. According to the actually “active” supports the relevant influence matrix in constructed. Energy is conserved for rigid supports and for supports which are so flexible that the impact time is large in comparison to the time steps. Treatment of plastic supports is possible.An application of the new method is demonstrated by analysis of the core barrel motion in a PWR with and without impacts at the lower core barrel edge and at the upper flange. The results show the large effects of such impacts in changing the global motions. Large local impact forces and accelerations appear. The interaction with the fluid reduces these loads. By proper design of the supports the resultant stresses can be minimized. Thus the method can be used to demonstrate and enlarge nuclear reactor safety.     

Experimental study on dynamic response of reactor vessel to an HCDA

This paper presents a discussion on the model experiments results for reactor structure dynamic response on FBR hypothetical core disruptive accident (HCDA) and the results of analysis using the dynamic response analysis code under experimental conditions. The purpose of this study is to clarify experimentally the dynamic response by use of scale models, as well as to attempt to confirm and improve the dynamic response analysis code on the basis of experimental data. The experimental results have clarified the inner barrel effects on reactor vessel deformation and its behavior due to a impact load. On the other hand, dynamic analysis was made of the 1/7.5 scale complex model by a dynamic response analysis code “PISCES-2DL”, using the explosive combustion characteristics as inputs. Obtained values were compared with experimental values. Results showed that this method was fairly capable of evaluating radial deformation behavior in lower cylindrical parts of the vessel.     

A survey of computational methods for fluid-structure analysis of reactor safety

The development of computational techniques and computer codes for the analysis of coupled fluid-structure behavior in the field of breeder reactor safety analysis is described. Methods for linear fluid-structure analysis, which are limited to acoustic or small-displacement, incompressible fluids are summarized first. Nonlinear methods have been found necessary for many of the breeder reactor safety problems because large displacements and nonlinear material behavior are encountered. Two problem classes are considered in detail: hypothetical core disruptive accidents, or containment analysis; subassembly analysis. The evolution of codes for these problems is described along with current capabilities and difficulties, and requirements for future developments.     

Transient failure analysis of liquid-filled shells Part I: Theory

In this paper, the governing equations which consider dynamic fluid-structure interaction, modal coupling in both axial and circumferential directions, and dynamic buckling are derived. The various pressure components acting on the shell wall due to a seismic event are also analyzed. The matrix equation of motion for liquid-filled shells is obtained through a Galerkin/Finite Element discretization procedure. The modal coupling among the various combinations of axial and circumferential modes are identified with a particular reference to the fluid-structure system under seismic excitation. Finally, the equations for the dynamic stability analysis of liquid-filled shells are presented.     

Analysis of nonlinear fluid-structure interaction in LMFBR containment

A generalized Eulerian method has been incorporated into ICECO for analyzing the nonlinear fluid-structure interaction in the primary containment of an LMFBR, consisting of complicated structural components such as the radial shield, core barrel, core-support structure, and the primary vessel. The method employs a Poisson equation to determine the hydrodynamic pressure in the fluid region, while using a relaxation equation to compute the pressure adjacent to the structure. A generalized coupling scheme is developed for treating the sliding condition at the fluid-structure interface, modeling the perforated structure, and analyzing the fluid motion at the geometrical discontinuities. Detailed formulations are given. Sample problems concerning wave propagation in a typical reactor containment are presented. It is shown from the results that this implicit, iterative method is unconditionally stable, and is especially suitable for excursions involving large material distortions.     

Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation.This paper presents a discussion of the history and some of the results of this effort, with respect to advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The purpose is to describe the nonlinear dynamic analysis techniques employed to analyze the HTGR core. Correlation of the codes is beyond the scope of the paper and will be discussed in subsequent publications.Each fuel column in the HTGR core is composed of stacked elements doweled together to ensure alignment of the coolant channels. Gaps exist between columns, allowing the elements to impact during a seismic disturbance. Analysis of this type of structure by standard structural dynamics techniques is not possible since both nonlinearities and discontinuities exist. One- and two-dimensional models of the three-dimensional core have been developed with explicit time integration methods. Various methods to treat the impact between elements are discussed.Three computer codes were developed. CRUNCH-1D models a one-dimensional horizontal strip through the core; CRUNCH-2D, a two-dimensional horizontal planar section; and MCOCO, a two-dimensional vertical planar section. The dynamic characteristics of these three representations of the full core structure are compared and the methods evaluated in the text. Plans for additional development and work to improve the techniques are also discussed.