后处理厂乏燃料储运吊篮设计与分析

在后处理厂乏燃料贮存水池,采用吊篮储存和转运乏燃料组件可以提升转运能力,同时具有较好的安全性和经济性。本文针对国内商业核电站服役使用的典型乏燃料组件结构参数,设计了可一次性装载8盒乏燃料组件的储运吊篮,利用SuperMC软件对满载乏燃料组件的储运吊篮进行了临界安全分析,再利用ANSYS有限元软件对其进行抗震分析。临界安全分析结果表明:即使满载乏燃料组件的吊篮在无限大乏燃料贮存水池内紧密排列,贮存水池keff最高仅0.906;抗震分析结果表明:在0.3 g设计地震震动加速度下,吊篮最大结构应力为7.92 MPa,最大形变量为0.32 mm,不影响其功能。因此储运吊篮设计合理,可满足后处理厂乏燃料水池储运吊篮的基本安全要求。     

吊篮下挂分体式堆内构件设计研究

通过对模块式多用途小型压水堆（ACP100）结构特点进行分析,提出了一种吊篮下挂的分体式堆内构件,其吊篮组件悬挂于反应堆压力容器中部并由压紧筒组件压紧,该结构具有易于制造安装、冷却剂分流密封效果好、流量分配合理等优点。力学分析、控制棒驱动线综合实验、堆内构件流致振动实验结果表明,该结构具有很好安全性和可靠性,满足ACP100的功能需求。      

反应堆堆内构件流致振动特性研究

以某新型工程试验堆为研究对象,采用试验分析与仿真计算相结合的手段探索该试验堆堆内构件流致振动特性。在流致振动试验中,根据相似准则建立了1/2的缩比试验模型,并在整体水力模拟台架上开展了100%额定流量工况下的堆内构件流致振动试验,测量了吊篮组件和二次支承组件在流体作用下的动力响应;根据吊篮组件和二次支承组件所受激励的不同,结合试验结果分别采用不同的计算方法得到了流体力作用下结构的动力响应,分别获得了100%额定流量工况下的最大应力值。     

堆芯跌落事故下水力缓冲性能分析研究

对采用"水力缓冲+机械缓冲"技术的反应堆堆内构件二次支承结构缓冲性能进行分析,研究假想堆芯跌落事故(吊篮断裂)下反应堆堆内构件二次支承结构对吊篮组件的水力缓冲作用机理。基于Fluent动网格技术对吊篮组件跌落过程进行数值模拟,分析不同竖直间隙、冷却剂温度及初始流速下吊篮组件跌落过程的运动规律;基于LS-DYNA非线性动力分析程序分析跌落末端的冲击过程,研究二次支承结构水力缓冲作用效果。分析显示,堆芯跌落事故下,水力缓冲可以吸收缓解大部分跌落冲击能量,与传统缓冲结构相比,缓冲效果更佳,确保了反应堆压力容器(RPV)的结构完整性和堆芯稳定性。     

反应堆吊篮在泵致脉动压力载荷下的响应研究

反应堆结构的流致振动问题一直受到核工程界的广泛关注。主泵的泵致脉动压力是一个重要激励源,其将导致反应堆吊篮等部件周期性振动,长期运行会导致结构的疲劳损坏。为研究新设计的“华龙一号”反应堆吊篮在泵致脉动压力作用下的振动响应,本文首先分析反应堆吊篮所受的泵致脉动压力,而后建立吊篮有限元模型,对其在泵致脉动压力载荷下的动力学响应进行研究,并综合考虑湍流激励,评价吊篮在堆内构件流体作用下的整体影响。应力分析表明,吊篮各位置流致振动的最大应力强度小于疲劳应力限值,结构是安全的。但对于新设计的反应堆,或反应堆冷却剂系统更换新的主泵,则反应堆吊篮及堆内构件的泵致振动需受到重视。     

反应堆吊篮的非轴对称温度场和热应力分析

反应堆吊篮在运行时受到γ射线辐射后会引起内部发热,由于燃料组件在堆内排列是非轴对称的,因此在吊篮内壁所产生的内热源也是非轴对称的。本文推导了由非轴对称内热源引起的温度场求解方程,并用热弹性位移势与应力函数相融合方法求解非轴对称的热应力解析式。以30万千瓦秦山核电站所设计的反应堆吊篮为例,计算了稳态温度场和热应力。计算结果表明温度应力环向分布不均匀发比较严重,如把问题简化为轴对称处理,则计算所得的热应力作为设计依据是不恰当的。     

反应堆吊篮在正常和劣化支撑条件下的振动模态特性

反应堆正常运行时,吊篮组件受到冷却剂的作用而诱发振动,往往会造成吊篮及其支撑结构发生疲劳破坏或松动脱落,危及反应堆的安全。通过长期对运行中的反应堆吊篮振动特性的监测发现:随着反应堆的运行,吊篮的固有频率,特别是梁式频率会发生较大的变化,而常规吊篮模态的计算方法是无法模拟和预测该梁式频率的变化规律。针对该问题提出了在劣化支撑条件下吊篮结构模态分析的计算方法和力学模型,该方法准确模拟了不同的劣化状态下吊篮的约束边界。以国内某堆型的吊篮结构为研究对象,分别计算了在空气和静水中支撑条件劣化5%、10%、15%、20%、25%、30%、35%、40%时的振动模态,并与国外学者相关研究的试验结果进行对比,结果表明:本文提出的反应堆吊篮在劣化支撑条件下的振动模态计算方法与相关研究的试验结果趋势一致,吻合较好。因此,该计算模型是合理、可行的,能够满足工程计算分析的需求。     

压紧弹簧两点接触下的吊篮法兰LOCA工况应力分析

对欧洲三代压水堆(EPR)吊篮法兰在两相失水事故(LOCA)瞬态下压紧弹簧"两点接触"状态下的载荷计算方法和EPR堆内构件吊篮法兰在LOCA工况下的受载特点进行分析。采用SYSTUS程序建模计算两相LOCA时吊篮法兰与压紧弹簧"两点接触"时的应力水平,通过与LOCA动态响应等效换算获得LOCA工况下的应力响应,并根据压水堆核岛机械设备设计和建造规范(RCC-M)的具体要求进行吊篮法兰应力评定。     

聚变次临界堆双冷嬗变包层第一壁结构和氦气流道优化分析

在聚变次临界堆双冷嬗变包层第一壁结构初步设计基础上 ,对第一壁结构尺寸和氦气流道形状进行优化分析 ,利用有限元分析软件对第一壁结构进行应力数值模拟 ,在满足结构应力及部件可靠性的前提下 ,给出最佳优化方案。     

严重事故下堆芯围板及吊篮熔融行为研究

为支撑堆芯熔融物压力容器内滞留有效性评估,采用计算流体动力学方法,建立了某先进压水堆堆芯辐射传热数值模型,对严重事故下围板及吊篮的熔融行为及其影响因素进行了研究。研究结果表明,在靠近堆芯燃料组件轴向功率分布因子峰值的节点,围板及吊篮的熔融行为较为显著;在同一节点处,围板的熔融并不是均衡发展的,最先熔穿的区域多发生在外围多个燃料组件交汇处,而吊篮的熔融则呈现出由内向外均衡扩展的变化趋势;压力容器外壁面的换热条件对堆芯围板及吊篮的熔融行为的影响并不显著,而燃料组件发射率的设置对堆芯围板及吊篮的熔融行为具有显著影响。可以为堆芯熔融物压力容器内滞留有效性评估提供技术支持。     

FCM燃料堆内行为模拟及结构设计研究

The thermal mechanical performance of the fully ceramics microencapsulated fuel (FCM) with different non-fuel part size was simulated using two-dimensional characteristic unit. When the fissile loading meet the requirements of the reactor core, the stress condition of SiC matrix and SiC layers were investigated for FCM pellets with different structures. Non-fuel parts and SiC layers suffered relative lower stress by optimizing FCM pellet structure and adjusting distance between different TRISO particles. The stress distribution of matrix, non-fuel part and SiC layer was discussed for the FCM pellets with non-fuel part size from 100 μm to 500 μm. The results indicate that, the maximum hoop stress of the matrix and SiC layer increased with the increasing of non-fuel part size, while the non-fuel parts exhibited crosscurrent. Non-fuel parts and SiC layer possessed lower stress when the non-fuel part was 400 μm. The stress of non-fuel part was about 400 MPa, and the maximum hoop stress of the SiC layers were about 200 MPa. The failure probability was 2.5×10-4. The structure integrity was maintained for the pellets with 400 μm non-fuel part, at the same time the failure probability SiC layer was low. Structural optimization is the basis for the application of FCM pellet.     

FCM燃料堆内行为模拟及结构设计研究

本文采用二维特征模型模拟不同无燃料区厚度全陶瓷微封装弥散(FCM)燃料的热力学行为,在保证堆芯装载要求的条件下,研究不同结构FCM燃料SiC基体和包覆燃料颗粒SiC层的应力状态。通过优化无燃料区厚度,调整TRISO颗粒间的间距,保证无燃料区和SiC层同时具有较低的应力水平。分析了无燃料区厚度为100～500μm时基体SiC、无燃料区以及SiC层的应力分布,结果表明,基体SiC和SiC层最大应力随无燃料区厚度增大而增大,而无燃料区的最大应力则随其厚度增大而降低。当无燃料区厚度为400μm时,无燃料区和SiC层均处于较低的应力状态,无燃料区SiC基体应力约为400 MPa,而SiC层的最大环向应力约为200MPa,其失效概率约为2.5×10-4。因此,当无燃料区厚度为400μm时,FCM燃料既能维持芯块结构完整,又能保证SiC层具有较低的失效概率。结构优化为FCM燃料的应用提供了基础。     

A structural analysis of the circular cylinder with multi holes under thermal loading

A structural analysis of the circular cylinder with multi holes is performed using the finite element analysis program . This structure is an analytical model of the capsule used for material irradiation tests. The temperature distributions of the cylinder due to gamma heating are obtained and various parameters, such as specimen size, quantity of specimens and gap sizes between the holder and the specimen are considered in the analysis to obtain the thermal and mechanical characteristics. To assess the structural integrity of the capsule, stress analysis under thermal loading is also performed. The analysis results show that, in all specimens, the peak temperature occurred, and is significantly dependent on gap sizes between the holder and the external tube or the specimen. The stress of the cylinder, under thermal loading, is lower than the allowable stress of the material used.     

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.     

Creep rupture at superimposed non-stationary stress and temperature loading

For superimposed stress and temperature loading the conditions are analysed for which the combined loading can be considered as the result of an independent superposition of the individual loading types. Superimposed time linear ramp loading as well as superimposed cyclic (triangle) loading is treated as a special case. In general, as far as one loading type strongly dominates as being the most life time consumpting type, the same will essentially determine the life time for the superimposed loading. In that case the loading types can be considered as mutually independent. However, the differences between the results of rigorous calculations based on the life fraction rule and approximate estimate (independent superposition) as shown for the case of Zircaloy-4 may be within the limits of reproducibility of usual stress rupture tests. The calculations are compared to experiments performed on Zircaloy-4. At loading conditions for which the grain structure remains constant very good agreement is achieved between calculated and measured life times.     

Thermal and mechanical characteristics of an instrumented capsule for a material irradiation test

In this study, the thermal and mechanical characteristics are analyzed for the structural integrity evaluation of the instrumented capsule used for the irradiation test of reactor vessel materials in the research reactor, hi-flux advanced neutron application reactor (HANARO). The temperature of test specimens inserted in the capsule mainbody by γ-flux is calculated using a heat transfer code, HEATING 7.2f. The maximum temperature is 556.75 K at the center of the capsule mainbody, thus the temperature satisfies the user's requirement. To estimate the mechanical characteristics of the capsule due to the pressure and thermal loading, stress analysis is carried out with a finite element analysis program, ANSYS. The strength of the capsule's external tube is also evaluated by considering the buckling stress of the capsule mainbody under coolant pressure loading. The results of the analysis show that the temperature distributions are significantly affected by the gap size between the holder and the specimen. The calculated stresses of the capsule structure are well within the allowable stress values of the ASME code. It is expected that the results presented in this paper will be useful in the design and safety evaluation of instrumented capsules for material irradiation tests.     

Modified crack closure integral based computation of stress intensity factors for 2-D thermoelastic problems through boundary element method

A modified crack closure integral (MCCI) based computation of stress intensity factors (SIFs) for thermal loading through boundary element method (BEM) is presented. Simple relations are given for the determination of stress intensity factors (SIFs) using the BEM results for linear, quadratic and quarter point elements employed around the crack tip. Examples of crack under mode I, mode II and mixed mode thermal and/or mechanical loading are examined. The computed SIFs are compared wherever possible with solutions available in the literature. The agreement is good. The effect of crack tip element size on the accuracy of results is reported.     

Fracture assessment of shallow-flaw cruciform beams tested under uniaxial and biaxial loading conditions

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow surface flaws. Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant depth), shallow, surface flaws. The cruciform beam specimens were developed at Oak Ridge National Laboratory (ORNL) to introduce a far-field, out-of-plane biaxial stress component in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock or pressure–temperature loading of an RPV. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for an RPV material. The cruciform fracture toughness data were used to evaluate fracture methodologies for predicting the observed effects of biaxial loading on shallow-flaw fracture toughness. Initial emphasis was placed on assessment of stress-based methodologies, namely, the J–Q formulation, the Dodds–Anderson toughness scaling model, and the Weibull approach. Applications of these methodologies based on the hydrostatic stress fracture criterion indicated an effect of loading-biaxiality on fracture toughness; the conventional maximum principal stress criterion indicated no effect. A three-parameter Weibull model based on the hydrostatic stress criterion is shown to correlate with the experimentally observed biaxial effect on cleavage fracture toughness by providing a scaling mechanism between uniaxial and biaxial loading states.     

核电站用贮气罐抗震分析

采用有限元法,对台山核电站SBO柴油发电机组用贮气罐进行了结构抗震计算分析。介绍了贮气罐结构的力学建模、载荷条件和地震响应谱分析方法,计算得到了结构的固有频率、振型和地震载荷下的响应,并根据RCCM规范对结构在重力、内压、地震等多种载荷组合下的应力进行评定。结果表明,贮气罐的结构强度满足规范要求。计算结果为贮气罐的抗震设计提供了依据。     

Probabilistic assessment of thermal fatigue in nuclear components

A probabilistic framework is set up to assess the fatigue life of components of nuclear power plants. It intends to incorporate all kinds of uncertainties, such as those appearing in the specimen fatigue strength (number-of-cycles-to-failure of specimens), design margin factors (taking into account the size, surface finish and environmental effects), mechanical model (precisely, the uncertainty on the model input parameters) and the thermal loading. This paper presents the global methodology and details the statistical treatment of the fatigue specimen test data. A first analytical example shows that the reliability of a structure submitted to a periodic stress cycle S changes significantly with respect to the value of S, although the codified (deterministic) design criterion is equally fulfilled. A more comprehensive example involving a mechanical model of a pipe submitted to a deterministic inner temperature loading is finally analysed. The use of the first-order reliability method (FORM) allows to compute the probability of failure as a function of the foreseen lifetime and to rank the input random variables according to their importance in response sensitivity.