基于Johnson-Cook本构模型的压紧板弹簧刚度特性研究

燃料组件压紧板弹簧的刚度设计对其安全服役起着至关重要的作用。通过引入INCONEL 718合金的Johnson-Cook非线性本构关系,拟合了不同中子辐照剂量下INCONEL 718合金的Johnson-Cook非线性本构模型;建立了压紧板弹簧系统的有限元模型,开展了不同因素对压紧板弹簧刚度特性的影响研究。结果表明:温度、加载次数、幅值的增大会导致压紧板弹簧在循环加载下出现不同程度的软化,而加载速率的增大会使压紧板弹簧的刚度出现硬化。本文研究可为燃料组件中的压紧板弹簧刚度设计提供参考。      

燃料组件格架弹簧刚度模型研究

格架弹簧是压水堆燃料组件的关键零部件,其为燃料棒提供夹持功能。刚度是格架弹簧的关键特性,其关系到燃料棒的堆内运行性能。本文基于格架弹簧的结构以及受力特点,给出了格架弹簧的受力分析模型,并推导了理论刚度计算公式,即得到了格架弹簧的理论刚度模型。另外,在商业有限元软件ABAQUS中建立了多种尺寸格架弹簧实体结构的有限元刚度模型,并计算获得了格架弹簧的变形量以及刚度曲线。通过有限元结果与理论刚度模型结果的对比,证明了理论刚度模型的合理性,并分析讨论了理论刚度模型的优缺点。本文首次提出的格架弹簧理论刚度模型可用于代替格架方案设计期间的有限元迭代过程,并能快速获得优化方案的主要设计尺寸。但本文分析方法不能替代实验,在方案固化后仍需开展实验确定格架弹簧的刚度。本文得到的格架弹簧理论模型可为燃料组件格架弹簧参数的快速优化设计提供新的思路。     

核电厂高静低动三维隔震系统的力学性能试验与减震效果研究

为实现隔震结构在静载阶段隔震层位移较小的同时满足动载阶段良好的减震效果，设计了一种由水平隔震单元和高静低动隔震系统（斜置橡胶支座和负刚度装置构成）组成的高静低动三维隔震系统。针对核电厂结构建立了该系统的竖向动力模型，分析了参数对系统传递率的影响，结果表明随着刚度比、阻尼比和力激励幅值比的增大，弹簧压缩比减小，力传递率幅值越小，在共振区体现出更好的隔震效果。通过对高静低动隔震系统进行静力加载试验，结果表明高静低动隔震系统在动载阶段滞回曲线饱满，具有较低刚度特征。通过理论模型与试验结果的对比，表明所提出的高静低动隔震系统理论模型能较好反映该装置系统力学特性。     

235U快中子核反应截面协方差评价

本文基于最小二乘不确定度传递方法,建立²³⁵U中子裂变核反应截面模型依赖型与非模型依赖型协方差评价体系。通过针对实验测量较丰富的中子反应总截面、辐射俘获、(n, 2n)等核反应实验数据不确定度源项分析,为协方差评价提供实验基础,并给出对应核反应截面的非模型依赖型协方差评价数据。通过开展快中子能区²³⁵U核反应理论模型参数灵敏度计算与分析,导出实验测量缺乏的核反应截面模型依赖型协方差评价数据。经上述系统评价,所得协方差数据与核反应截面中心值研究过程自洽、物理合理,并按国际标准ENDF-6格式输出,便于核工程用户使用。     

235U快中子核反应截面协方差评价

本文基于最小二乘不确定度传递方法,建立²³⁵U中子裂变核反应截面模型依赖型与非模型依赖型协方差评价体系。通过针对实验测量较丰富的中子反应总截面、辐射俘获、(n,2n)等核反应实验数据不确定度源项分析,为协方差评价提供实验基础,并给出对应核反应截面的非模型依赖型协方差评价数据。通过开展快中子能区²³⁵U核反应理论模型参数灵敏度计算与分析,导出实验测量缺乏的核反应截面模型依赖型协方差评价数据。经上述系统评价,所得协方差数据与核反应截面中心值研究过程自洽、物理合理,并按国际标准ENDF-6格式输出,便于核工程用户使用。     

蒸汽发生器下部水平支承改进型缓冲结构刚度性能试验

介绍核电厂蒸汽发生器(SG)下部水平支承改进型缓冲结构,以及截面尺寸为10 mm×10 mm缓冲结构的刚度试验选材、试验设备、载荷施加方式等,并对缓冲结构刚度试验的结果进行处理,得到此改进型缓冲结构在不同区间内的加载载荷与位移之间的关系式、对应的弹塑性区的刚度值以及加载载荷与残余变形之间的变化曲线。     

蒸汽发生器下部水平支承改进型缓冲结构刚度性能试验

介绍核电厂蒸汽发生器(SG)下部水平支承改进型缓冲结构,以及截面尺寸为10 mm×10 mm缓冲结构的刚度试验选材、试验设备、载荷施加方式等,并对缓冲结构刚度试验的结果进行处理,得到此改进型缓冲结构在不同区间内的加载载荷与位移之间的关系式、对应的弹塑性区的刚度值以及加载载荷与残余变形之间的变化曲线。     

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。     

采用组合质量-弹簧模型的快堆卧式贮钠罐晃动分析

在核电工程中广泛使用各类形式的储液容器,储液容器的抗震分析必须考虑液体晃动的影响。针对矩形储液容器,不同于传统的单向质量-弹簧模型将液体晃动对容器侧壁、底部的作用都等效为对侧壁的作用,本文提出一种组合质量-弹簧模型及计算公式,模拟了液体晃动分别对容器的侧壁、底部的作用。组合质量-弹簧模型在三维有限元模型上的加载位置更加合理,容器底部的应力结果更加真实。利用组合质量-弹簧模型对中国实验快堆的卧式贮钠罐进行横向晃动有限元计算,算例表明了计算结果的可靠性。组合质量-弹簧模型为储液容器的有限元抗震分析提供了一种有效的方法。     

核电厂负刚度阻尼隔震结构的地震响应研究

隔震技术能有效减小核电厂上部结构的加速度响应,但强地震作用下隔震层位移过大会导致管道断裂。本文基于曲面运动原理及预压弹簧伸缩特性提出了一种负刚度阻尼系统,通过球铰在拱球面曲线运动实现负刚度特性,并在弹簧压缩方向提供黏滞阻尼性能。提出了负刚度系统的理论恢复力模型并进行了力学特性分析,设计了负刚度装置并完成了静力试验,结果显示理论恢复力模型与试验结果的一致性较理想。将核电厂负刚度阻尼隔震结构与核电厂隔震结构进行了地震响应对比分析,比较了不同地震波输入下的地震响应。结果表明负刚度阻尼系统可有效同时减小核电厂上部加速度响应和隔震层位移响应。     

非岩性地基条件下核岛厂房-桩-土非线性动力相互作用特性研究

采用核电厂核岛厂房-地基相互作用高性能大规模计算方法,建立了同时考虑钢制安全壳、附属厂房、桩基础以及覆盖层地基的第三代核电厂整体精细化模型,结合了考虑覆盖层非线性效应的粘弹性边界的地震波动输入方法,开展了地基模型尺寸、桩基单元类型、桩-土接触效应等对核岛厂房地震动力响应影响的研究,建议了地基尺寸的合适范围。结果表明:桩基处理方案与覆盖层地基结果相比,厂房水平方向楼层反应谱峰值频率向高频方向发生较大偏移;采用实体单元模拟的桩基能更准确地反应楼层反应谱(FRS)的规律;桩-土接触面对厂房楼层谱影响不大,却对桩基自身的应力分布有着显著改变。     

Large strain, viscoelastic and elasto-viscoplastic numerical analysis by means of the finite element method

Finite element procedures and illustrative numerical examples for linear and nonlinear viscoelastic and elasto-viscoplastic bodies are discussed. According to the well-known theory in rheology, the constitutive equation of viscoelastic bodies can be described by the first-order simultaneous differential equation system. Using this result, the finite element equation leads to an equation similar to the Hookean elastic body except for the inclusion of the time increment. The procedure is applicable to certain forms of nonlinear viscoelastic and elasto-viscoplastic bodies. To solve the nonlinear simultaneous equation system, it is convenient to employ the incremental displacement method for the nonlinear problems such as large deformation loading, unloading and cyclic loading behavior.     

Mechanical testing of a FW panel attachment system for ITER

An objective of experiments and finite element simulations was to check the stiffness, the strength and the fatigue resistance of the attachment of the First Wall panels onto a shield block of blanket modules according to the ITER 2001 design. The panel has a poloidal key at the rear side (in so-called option A with the rear access bolting) and it is attached by means of special studs located on a key-way in the shield block. Special device for a test of stud tensile pre-load relaxation during a thermal cycling was developed. True-to-scale panels, the shield block mock-up and simplified studs were fabricated and the assembly was loaded alternatively by radial moment, poloidal force or poloidal moment simulating the loading during off-normal plasma operations. Thermal cycling led to an acceptable stud pre-load relaxation. Mechanical cycling caused neither the pre-load relaxation nor the loss of the contact in the key-way nor a damage of the attachment system. The combination of poloidal moment and radial force during vertical displacement events (VDEs) seems to be a most dangerous case because it could lead to the loss of the key–key-way contact.     

控制棒驱动机构用电磁轴承本体电磁结构设计研究

电磁结构设计是控制棒驱动机构用电磁轴承本体设计的核心环节。为了获得合理的电磁结构,针对不同磁通分布形式进行了电磁结构选型,建立了用于电磁分析的解析模型和有限元模型;然后采用有限元仿真分析的方法确定了电磁轴承的工作点,并对电流刚度和位移刚度等电磁轴承关键性能参数进行了深入分析研究;最后采用有限元仿真分析结果与解析计算结果对比的方式进行了互算验证。验证结果表明,控制棒驱动机构用电磁轴承本体电磁结构设计合理,各项性能指标均可满足设计要求。      

Predictions of displacements and stresses in cracked reinforced concrete containment vessels subjected to seismic loads

A three dimensional, nonlinear finite element analysis capability for analysis of cracked reinforced concrete nuclear containment vessels is described. A bilinear shear stress-shear strain model, based on experiments, is used to study the distribution of radial and tangential shear produced by horizontal seismic loading. Results show substantial redistribution of tangential shear, and significant increases in radial shear, peak bending stress, and vessel displacement, as compared to conventional linear analysis results.     

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.     

基于小位移旋量的浮动式核电站反应堆装卸料定位误差分析与优化

燃料组件装卸是核电站反应堆换料检修的一项重要操作，浮动式核电站由于运行环境特殊，其装卸料定位精度要求更高。本文基于小位移旋量（SDT）的公差建模方法对浮动式核电站反应堆装卸料的导向定位误差进行了分析，采用刚体动力学坐标系变换的方法得到了装卸料定位误差的表达式；采用MATLAB程序进行了模拟计算。对海洋条件下反应堆装卸料的极限倾斜量进行了分析，给出了最大倾角与海浪参数的关系，对燃料组件导向间隙量等关键参数进行了优化，并将其结果与试验数据进行了比较，相符性较好。      

超临界水堆燃料棒流致振动简化模型

在超临界水堆中,当超临界水流过带有绕丝的燃料棒时可能诱发其发生振动,使得燃料包壳发生疲劳现象。带有的接触的非线性有限元模型使得计算量大大增加,而且其计算精度仍有待实验验证。本文针对超临界水堆流致振动实验,将绕丝的影响简化为弹簧,建立燃料棒流致振动的简化模型,并通过有限元模型对燃料棒的固有特性进行分析,验证了模型的正确性。最后,以功率谱对模型加载,求得了超临界水堆燃料棒的位移响应和1δ解。     

Evaluation of aseismic integrity in the HTTR core-bottom structure V. On the static and dynamic behavior of graphitic HTTR key-keyway structures

The graphite components in high temperature gas-cooled reactors are connected to each other through a key-keyway structure that has gaps between the key and the keyway to accomodate thermal expansion. Because a dynamic load concentrates on the key-keyway structure during earthquakes, it is considered to be a crucial element for assessing the integrity of the graphite components. A combination of experiments and analyses was employed to investigate the dynamic behavior of the key-keyway structure, i.e. the equivalent stiffness associated with vibrational characteristics of the graphite components and the stress distribution under dynamic loading. The experiments were performed using a graphite scale model and a dynamic photo-elastic method. The analysis was carried out using the finite element method (FEM) code Abaqus, taking account of the contact between the key and the keyway. The following conclusions were derived. (1) The equivalent stiffness of the key-keyway structure shows nonlinearity, owing to the contact deformation. (2) The equivalent stiffness evaluated by the FEM analysis, taking account of the non-inear contact deformation, is applicable for predicting the vibrational characteristics of ky-keyway structure. (3) The stress concentration under dynamic loading is lower than or nearly equal to that under static loading. The maximum stress concentration of the seismic load can be sufficiently evaluated under static loading conditions.     

Damage-plastic loading surface model for concrete

The paper presents a plasticity-type model which is characterized by normality, continuity, convexity and associatedness. It utilizes two loading surfaces, of which one describes the behavior associated with damage due to inelastic volume change (expansion from micro-cracking, or contraction from pore collapse or closure) and the other plasticity at no inelastic volume change. Since only one of these two surfaces is active at the same time, the formulation is equivalent to a single loading surface. The deviatoric section of the damage loading surface is a rounded triangle, and the Rendulić section meridians have the shape of a slanted ellipse. The loading surfaces are smooth and have no corners. Although the model involves approximately the same number of material constants as the previous models of similar capability, identification of these constants from test data is much simpler. The properties of the model are also more advantageous for numerical finite element applications. The model is shown to be capable of describing the basic uniaxial and multiaxial test data for concrete except degradation of material stiffness.