10MW高温气冷堆压力容器主法兰结构的有限元接触分析

10MW高温气冷堆（HTR－10）压力容器主法兰是HTR－10的关键结构部件,对HTR－10的正常运行起着重要作用。HTR－10压力容器主法兰采用主螺栓进行联接,采用一道金属“O”形环和一道“Ω”环进行密封。该研究课题对此主法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预紧以及加压过程中主法兰的应力和位移情况。计算采用二维轴对称结构模型和MSC MARC2000有限元程序。结果表明,无论是预紧状态还是设计压力状态。HTR－10主法兰能满足强度要求,“O”形环和“Ω”环也都有满足密封要求。     

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。     

反应堆压力容器主螺栓粘扣问题的分析与处理

基于反应堆堆压力容器主螺栓的结构及其工况,提出一种均载的、安全系数高的削峰均载螺纹结构。建立大螺栓螺纹连接的数学模型进行限元分析,并与普通螺纹结构比较,验证了该结构的可行性。解决现在的压力容器主螺栓连接结构在使用一段时间后的粘扣现象,使得压力容器主螺栓连接具有更高的安全性、可靠性以及更长的寿命。     

两种不同密封面结构反应堆压力容器的密封性能对比研究

反应堆压力容器(RPV)密封面结构是影响RPV密封性能的重要因素。建立2种不同密封面结构的RPV三维有限元模型,研究其对RPV密封性能的影响,并得到上、下法兰轴向分离量以及主螺栓载荷等分析评价RPV密封性能的关键指标,同时,对比分析2种密封面结构形式的安全裕量,为优化RPV密封面结构设计提供理论依据。     

反应堆压力容器螺栓孔与螺栓连接件专用检修设备

简要叙述了中国核动力研究院进行的核岛专用检修设备的研制工作,描述了反应堆压力容器（ＲＰＶ）螺栓孔与螺栓连接件的检修工艺流程,围绕该工艺流程,从功能,结构,特点等方面介绍了为大亚湾核电站研制的ＲＰＶ法兰模拟体,螺栓孔螺纹闭路电视（ＣＣＴＶ）自动检查仪,螺栓孔螺纹膨胀梳刀,螺栓孔螺栓面抛光机,螺栓孔螺纹面上油机,螺栓孔螺纹铣削机,螺栓孔螺纹观察镜及螺栓螺母清洗机,并对专用检修设备的实际使用情况作了介绍     

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.     

CPR1000反应堆压力容器密封性能模拟技术研究

压水型核反应堆压力容器的密封性能是保证核电厂安全运行的关键因素之一。为了探索反应堆压力容器密封性能的数值模拟技术,本文建立了CPR1000反应堆压力容器(RPV)密封结构的热弹塑性三维有限元分析模型,考虑了运行期间的载荷及载荷组合,得到了反应堆压力容器在升温、运行和降温瞬态过程中上下法兰的轴向分离量、径向滑移量以及螺栓载荷等。分析结果表明热弹塑性三维有限元密封分析模型能够较好地模拟密封结构的性能。     

核反应堆压力容器主密封瞬态性能研究

为研究核反应堆压力容器主密封瞬态力学特性和密封性能，本文建立了主密封结构三维数值模型，分析了主密封组件在典型瞬态条件下的温度和应力分布特性，从法兰和主螺栓变形协调机理角度，研究了主螺栓应力在瞬态条件下的变化规律及内在原因，总结了密封面处法兰轴向分离量变化机制，并对瞬态循环条件下密封面累积塑性变形和法兰分离量演化规律进行了预测研究。研究结果表明，温度滞后效应导致主螺栓在瞬态条件下应力交变幅值大；瞬态温度和压力对密封面处分离量影响很大，急速升压会使得分离量快速增大；在启停堆瞬态循环作用下，密封面处分离量曲线呈现周期性特征，经历若干次循环后分离量曲线达到稳定，密封面局部弹塑性变形达到安定，整体塑性变形分布趋于均匀。     

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。     

AP1000核电厂主泵法兰螺栓在役检查系统开发

AP1000核电厂反应堆主泵法兰螺栓是在役检查重要监督项目之一,目前国内尚无针对该部件的在役检查系统及应用案例。本文结合AP1000主泵法兰螺栓结构特点、现场高剂量环境及复杂检查条件分析,设计开发了一套从螺栓中心孔内壁实施超声检测、适用于在役检查要求的主泵法兰螺栓在役超声检查系统。主泵模拟体上的调试试验结果表明,该系统可实现周向运行、垂直方向避障、专用超声探头与螺栓孔精确对中调节等功能,进而实现对主泵法兰螺栓的超声扫查。工程应用结果证明本系统满足AP1000核电厂主泵法兰螺栓在役检查现场要求,具有较高的可靠性和良好的适用性。      

CPR1000核电厂核反应堆压力容器应力分析

本文所计算的核反应堆压力容器是保证核安全的一道重要屏障,因此,要参照相应的规范和标准对其进行强度方面的分析和校核.通过有限元软件ANSYS建立压力容器的三维模型,计算压力容器在设计工况以及试验工况下,在压力、温度、堆内构件重力和接管载荷等各种载荷作用下的应力强度,并严格参照规范标准RCC-M B篇规定的各种工况下的应力准则,对压力容器进行强度评定.评定的结果表明,压力容器在计算的几类工况下,均符合规范标准RCC-M的强度要求.本工作的计算和分析也为我国核工业未来的设备设计制造走上国产化、标准化奠定了一定的基础.     

Berechnung des räumlichen spannungszustandes in druckbehältern mit ausschnitten mit hilfe der dynamischen relaxation

The general axisymmetric stress and strain fields of prestressed concrete reactor pressure vessels usually are perturbed by a series of large openings in the vessel walls. While up to now the usual practice was to supplement the axisymmetric structural analysis of a fictive truly axisymmetric vessel structure by supplementing considerations on the areas of the surroundings of the holes, now there is a trend for performing an integral three-dimensional vessel analysis. The mathematical difficulties do not allow analytical solutions, and numerical solutions of such problems have been performed up to now solely by use of the finite element method.

In this paper it is shown that it is also possible to determine the three-dimensional stress and strain fields of prestressed concrete reactor pressure vessels by using the so-called dynamic relaxation method. The mathematical basis for such calculations is outlined, and the characteristics peculiar to this numerical procedure are discussed. Besides of some two-dimensional examples demonstrating the degree of exactness achievable for fulfilling the boundary conditions, the results of a three-dimensional analysis of a hollow cylinder with a quadratic opening are given. These examples are confined to the elastic region. Expansions of the mathematical basis for considering plastic and viscoelastic deformations are possible.     

Development of Debris Coolability Analysis Module in Severe Accident Analysis Code SAMPSON for IMPACT Project

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability.     

Considerations for in-service inspection on pressure vessel of 200MW nuclear heating reactor

The 200MW nuclear heating reactor adopts an integrated arrangement for the primary circuit. It is designed to be operated at lower temperature and lower pressure as compared to large reactors. A steel containment serves a barricade for the reactor pressure vessel. The pressure vessel has some safety characteristics, such as low stress level, low induced integral neutron flux, and high toughness etc. Among them, the most important is its LBB behavior. Based on the safety analysis for the pressure, the requirements and procedures of in-service inspection are layed-out accordingly.     

A large electro-magnetic pump for high temperature LMFBR applications

A 14 500 gal/min (3300 m³/hr) electro-magnetic (EM) axial annular linear induction pump with center return (ALIP-CR), was designed to satisfy the requirements of Section III and Code Case 1592 of the ASME Boiler and Pressure Vessel Code. One of the important features of this EM pump is that there are no moving parts. Every significant loading condition was considered in the stress analysis of this pump for a liquid metal (Na) fast breeder reactor (LMFBR) application. An axisymmetric overall stress model (finite element) was first set up to analyze the stress state within the pump duct which has a pressure boundary with the conventional pressure vessel features. Axisymmetric loading such as pressure and thermal loads as well as non-axisymmetric loading such as seismic and vibration loads were taken into account. Local analyses were then performed for specific critical areas to supplement the overall model and to substantiate the adequacy of the design. Analytical results show that the EM pump is an acceptable and an attractive design for high temperature LMFBR applications.     

与流体晃动相耦合的快堆主容器地震响应的数值模拟

快堆的主容器内存在着自由表面流体,当发生长周期地震时,该流体的晃动会冲击到容器壁,对反应堆造成威胁.本文采用ADINA软件建立快堆主容器的三维有限元模型,模拟了正弦三波激励下液面晃动对容器壁的冲击现象,得到的冲击压力为容器结构完整性分析提供了载荷,验证了运用ADINA软件对自由表面流动进行分析的可行性以及在处理流固耦合问题上的优越性.     

The numerical simulation of the WWER440/V213 reactor pressure vessel internals response to maximum hypothetical Large-break Loss of Coolant Accident

The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident.This paper presents preliminary results of the numerical simulation of the WWER440/V213 reactor vessel internals (RVI) dynamic response to maximum hypothetical Large-break Loss of Coolant Accident (LOCA). The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such permanent (plastic) deformations occur in the RVI which would prevent timely and proper activation of the emergency control assemblies.In the case of the LOCA accident it is assumed rapid “guillotine” break of one of the main coolant pipes and rapid depressurization of the primary circuit. The pressure wave spreads at the speed of sound, enters the reactor pressure vessel and causes deformation and stress in reactor vessel internals.The finite element model was created by MSC.Patran (Patran, 2010) and dynamic response was solved using MSC.Dytran (Dytran, 2008) finite element code. The model consists of reactor vessel internals (Lagrangian solid elements) and water coolant (Euler elements) inside the reactor. Arbitrary Lagrangian Eulerian (Belytschko et al., 2003) coupling was used for simulation of the fluid-structure interaction. The calculation assumes no phase change in the water. No comparison with the experiment was performed up to now, because the required experimental data are not accessible for this type of the reactor.The most important acceptance criteria for the reactor internals demands that the movement of the emergency control assemblies under all operating conditions including accident is ensured (BNS, 2008). The numerical simulation of the WWER440/V213 reactor internals response to a LOCA accident showed that the acceptance criteria for RVI is fulfilled and required NPP safety standards are satisfied.     

基于OpenFOAM的熔融池自然对流传热与凝固数值研究

熔融物堆内滞留是第3代核电技术重要的严重事故缓解措施之一,堆芯熔融池在压力容器下封头壁面的热流密度分布直接影响该策略的有效性。本文基于开源的数值计算流体力学软件平台OpenFOAM,应用相变模型和浮升力模型二次开发了用于模拟堆芯熔融物由内热源或温差驱动的自然对流传热与相变求解器。应用该求解器模拟了瑞典皇家理工学院开展的二维氧化池与金属层耦合传热试验,获得了氧化池和金属层硬壳的相场,以及熔融池内的温度分布及沿容器壁面的热流密度分布。计算结果表明,该模型可用于熔融物凝固与自然对流的模拟,为深入分析核电厂采用熔融物堆内滞留措施后熔融池的行为奠定了基础。     

Implications of the differences in neutron spectra predicted for reactor pressure vessel walls by transport and the P1MG codes

Neutron-energy spectra were calculated for the interface between the vessel wall and cladding of the Army SM-1A Reactor pressure vessel using the transport theory code Program S and the diffusion code P1MG. Different sets of basic nuclear data and microscopic cross sections were used for the two calculations. Spectra were normalized to the same amount of activation in an iron, neutron flux detector. The transport code predicted a higher flux of neutrons in the energy groups between 6 and 10 MeV resulting in a lower overall intensity for the transport theory spectrum versus the P1MG spectrum. This was found to be consistent with the predictions of two transport codes versus the P1MG code for the PM-2A reactor vessel wall and for a simulated reactor vessel wall experiment. Such divergence of results for a given reactor using two different code analysis techniques raises important questions as to their usually unqualified acceptance and use for projecting the lifetime fluence for a reactor pressure vessel. Strong support is thus generated for establishment of one “standard” set of basic nuclear data from which all reactor physics analysts can draw to generate specific cross sections for reactor physics calculations, and for the writing of a new reactor physics spectrum code specifically for deep penetration analysis of reactor pressure vessel walls.