蒸汽发生器试验本体快速冷却过程中主蒸汽法兰泄漏有限元分析

采用线弹性瞬态热固耦合有限元方法对高温气冷堆蒸汽发生器试验本体主蒸汽法兰在快速降温试验过程中出现的泄漏现象进行了分析。建立了主蒸汽联箱及法兰螺栓连接模型，模拟了主蒸汽法兰的预紧、加压、升温和瞬态降温过程，分析得出了导致法兰密封结构泄漏的主要因素是快速降温过程中法兰的局部变形及螺栓残余预紧力降低，导致密封面张开量大于金属O型环的可靠密封回弹量。在此基础上模拟了不同降温速率下法兰密封面的张开位移，结果表明，限制蒸汽降温速率可改善压力容器法兰的密封性能。     

高温高压容器封头安装中螺柱应力试验

针对高温高压容器封头法兰连接中,加载后螺柱预紧力分布不均匀问题,提出在螺柱表面粘贴电阻应变片测量螺柱应变变化的测试过程与基本操作方法.采用正弦模型和指数衰减模型预测螺柱应力的变化,以探索适合的预紧方式和制定合理预紧程序.对试验数据的分析结果表明,螺柱的应力沿法兰周向成正弦规律变化;单根螺柱应力在预紧过程中满足指数衰减规...     

基于泄漏率的核设备密封技术研究

本文以稳压器双锥密封结构为对象,采用多线性随动强化模型模拟了垫片的压缩回弹力学行为,探究了降温速率和螺栓预紧力变化对密封面接触压力的影响,利用基于泄漏率的密封模型对密封性能进行了科学表征,建立了以最大允许泄漏率为准则的螺栓法兰垫片连接设计方法,解决了依据ASME和GB-150规范密封设计系统在使用过程中多次发生泄漏的问题,为工业领域螺栓法兰密封系统设计提供了新的思路。     

蒸汽发生器一次侧人孔螺栓连接结构三维有限元分析

采用三维有限元模型,对核电厂蒸汽发生器一次侧人孔螺栓进行运行工况下的瞬态分析。综合考虑螺栓温度滞后、法兰转角及法兰与垫片嵌合面的弹塑性接触等因素,解决螺栓孔及人孔盖与法兰之间的空气传热问题,实现螺栓预紧模拟及螺栓载荷的动态提取,较真实地模拟出系统的变形协调性及传热特性,并根据规范对螺栓密封状态进行验证,对法兰连接结构及螺栓本身进行疲劳评价。     

基于初应变理论的螺栓预紧力加载方法在应力分析中的应用

本文针对在ANSYS中建立起来的下泄热交换器下法兰设计载荷下的计算模型,提出一种用初应变法解决下泄热交换器下法兰螺栓预紧力加载的方法.在螺栓三维梁单元参数中定义梁单元的初应变,等同获得预紧力引起的初始应力场,成功解决计算模型的螺栓预紧力加载问题.并对下泄热交换器下法兰在设计载荷条件下给出了应力计算分析结果.     

10MW高温气冷堆压力容器主螺栓液压拉伸机

为了配合１０ＭＷ高温气冷堆３个压力容器的安装与拆卸，清华大学核能技术设计研究院研制了３种主螺栓拉伸机以代替昂贵的进口设备。本文对此拉伸机的研制情况、结构特点、设计参数、试验情况等方面作了描述，并给出螺栓预紧力的计算方法。这批拉伸机已成功地应用于１０ＭＷ高温堆３个压力容器的出厂水压试验。     

水平孔道“O”形环密封结构有限元接触分析

利用ABAQUS软件对研究堆水平实验孔道中异种材料法兰联接的密封结构进行了弹塑性接触计算.利用有限元分步加载技术,模拟了主螺栓预紧和加压过程,研究了主法兰的应力分布和结构的密封性能. .计算结果表明,在预紧状态和设计压力状态下,水平孔道中的异种材料法兰联接.双道"O"形环密封结构完全可以满足强度和密封要求.     

反应堆压力容器主螺栓预紧工艺优化仿真分析

应用商业软件ANSYS模拟了反应堆压力容器主螺栓十字拉伸3级预紧过程,对其预紧工艺进行了仿真优化分析。分析结果表明:3级加载方式中第1级采用对称加载、第2级采用间隔加载、第3级采用顺序加载得到的主螺栓不均匀度和离散度最小;第3级预紧采用变载荷能明显减小主螺栓的不均匀度和离散度,提高反应堆压力容器的密封性能。      

螺栓加载拉伸技术的应用

主要介绍了直径较大的螺栓实现预紧的一种新方法－螺栓加载拉伸技术，讨论了螺栓加载拉伸技术原理，拉伸量计算方法，螺栓拉伸实现的步骤及校验的方法，这种技术对设备损伤小，密封组受力均匀，从而使容器密封的可靠性大大提高。     

核级承压设备密封结构的有限元分析

在AP1000反应堆系统中,很多设备具有承压的功能,其密封性能直接关系到系统能否正常运行,因而密封失效是较之弹塑性失效、疲劳失效等更为基本的失效形式。在ASME规范中采用的密封结构设计方法是华脱尔斯法,此方法采用了一些保守的经验和假设,无法对密封结构处的变形和应力进行细致的计算。本文采用ANSYS有限元分析软件对核承压设备典型的密封结构进行了建模计算,提出了在有限元模型中螺栓预紧力和垫片的等效处理方法,能够对密封结构处垫片的回弹量、法兰的变形及应力分布进行预测。模型分析了采用华脱尔斯法进行密封设计时的设计余量,得到了垫片回弹量与设备内压之间的关系,对于核级承压设备密封结构的设计具有一定的借鉴意义。     

基于单变量求解方法的华龙一号内置换料水箱过滤器的设计和研究

为了解决华龙一号（HPR1000）事故后安全壳内置换料水箱（IRWST）过滤器设计中的压降求解问题,本文提出了一种单变量求解IRWST过滤器压降的方法,通过在过滤模块和汇流槽之间增加阻力部件,将IRWST过滤器压降求解中的多组变量转化为阻力部件的流通面积这一单组变量,实现了IRWST过滤器的压降求解。结果表明:采用单变量求解方法,可使每个过滤模块的碎渣量和流量相同,通过对IRWST过滤器的压降值计算,可确定IRWST过滤器的初步过滤面积;通过碎渣压降试验对IRWST过滤器的初步过滤面积进行了验证,其结果满足安全系统的设计要求。      

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor

A highly reliable control rod drive mechanism driven by an electric motor installed inside the reactor vessel (INV- CRDM) for a very small reactor has been designed. The INV-CRDM contributes to the compactness and simplicity of the reactor system, and can eliminate the possibility of a rod ejection accident. In the design, a new type of latch mechanism using an electromagnetic force to directly connect both of the shafts, one of which was the motor driven shaft and the other the control rod driving shaft, was applied so as to make the INV-CRDM very compact. The cable supplying current remained stationary, even when both of the shafts was moving. The required functions of the latch mechanism are to maintain an adequate latching force for the control rod shaft to move within a stroke of 370 mm, and to release the shafts in a shorter time than 0.2 s after a scram signal is received. A functional test with a model that approximately simulated the design was conducted to test the latching force and de-latching at room temperature. The test showed that the latching force increased with the current of the magnet coil, as did the de-latch time. The post-test analysis with a finite element analysis code revealed that the clearance between the two shafts greatly affected the latching force. With the same analysis method, the design analysis of the latch mechanism at a high temperature condition of 300°C was conducted, and it was confirmed that the latch mechanism contained enough latching force.     

压水堆核容器密封分析(英文)

本文作为核容器密封性能综合研究中心课题之一,给出容器密封分析基本方程及程序系统。经多种试验校核证实程序可信。根据多个容器分析计算,提出了就密封性能而言的压力容器类型概念,这对容器设计选定合宜预紧系数、保证密封并改善主螺栓疲劳性能有重要意义。     

超临界压水堆改进型控制棒组件落棒过程计算分析研究

为验证超临界压水堆改进型控制棒组件能否实现预期水力缓冲功能,采用计算流体力学分析软件Fluent、基于6自由度（6DOF）模型的铺层法动网格技术,对其落棒过程进行研究,分析了控制棒组件落棒时间和落棒末速度。结果表明：相比改进前的设计,改进型控制棒组件落棒时间虽有所增大,但仍然能满足安全要求;落棒末速度大幅下降,落棒冲击力降低,从而能够保证控制棒组件及燃料组件的结构完整性。改进型控制棒组件的设计能够实现预期的水力缓冲功能,可用于超临界压水堆堆芯设计。     

Identification of damping mechanism of TRR-II reactor control rod during free fall insertion

In light water reactors, control rods are in general inserted into reactors by gravity. In order to achieve a rapid shutdown, it is required to insert control rods as fast as possible. On the other hand, a control rod with a fast falling velocity would impose a substantial impact to reactor structure as well as to the rod itself. Therefore, a damping force must come into effect, especially during the final stage of the free fall of the control rod. The purpose of this study is to develop a mathematical model and a numerical simulation to describe and identify the damping mechanism; and apply this model to the design of the control rod used in TRR-II reactor of the Institute of Nuclear Energy Research (INER) of Taiwan.The damping effect of a falling control rod comes from two factors: the viscous shear stress occurred in a narrow gap between the rod and an outer tube which confines the lateral movement of the rod, and the pressure force exerted on the rod by the compressed water under the rod. The viscous shear stress can be analyzed by assuming a couette flow between the rod and the outer tube similar to the viscous force occurred in rheology. In doing this, the flow rate in each flow path is closely related to the pressure gradient in the flow path and can be evaluated using an electrical circuit analogy. The results of the code prediction were compared to the experimental results as carried out by the INER. Finally, a parametric study was applied to estimate the effects of the various factors including gap thickness, size of the flow holes, and other geometric considerations on the rod falling velocity. The results of this study can serve some technical support during the stage of rod design and manufacture.     

秦山核电厂反应堆压力容器水压试验应变测量与分析

对秦山核电厂反应堆压力容器出厂水压试验测点布置作了说明,与一回路系统水压试验位移测量的主要结果作了分析对比,给出了实测载荷与主螺栓测试数据,讨论了表征密封性能的法兰转动,认为出厂水压试验此项结果有错.     

Three-dimensional transient sealing analysis of the bolted flange connections of reactor pressure vessel

The failure of sealing system of the bolt flange connections is the primary failure mode of the nuclear reactor pressure vessel (RPV). For the safety and integrity of RPV, it is important to predict the sealing behaviour of the bolt flange connections under various loading conditions. Based on the finite element (FE) method for coupled thermal elastoplastic contact problems, a three-dimensional (3D) transient sealing analysis program of nuclear reactor pressure vessels is developed with the consideration of the non-linearity from both surface and material, transient heat transfer and multiple coupled effects. A contact correction approach is proposed to simulate the loading of the bolt connection under the condition of pre-stressing. An automatic pre-processing program is developed for FE modelling of RPVs. Using these programs, a 1:4 scaled model of a 300 MW RPV is analyzed under the loading conditions including pre-stressing, pressurization, heating and cooling. The computational results obtained are in a good agreement with the data of experimental tests. These programs are also successfully used in analyzing the full-scale model of the RPV in a nuclear power plant.