板状燃料元件堆芯流量分配及不对称冷却计算研究

通过选择合适的数学物理模型,应用Compaq Visual Fortran 6.O软件编制程序,对全堆芯流量分配和板状燃料元件的不对称冷却进行了耦合迭代求解.构造了3种流量分配和两种板状燃料元件不对称冷却的迭代方法,以一个组件为计算对象,对这几种方法作了计算对比.计算表明:堆芯结构和功率分布对流量分配都有影响,但是主要起作用的还是堆芯结构;不对称冷却将显著的影响到板状燃料元件的温度场和功率分配.     

多层环形流道阻塞事故三维数值分析初步探索

由环形燃料板组成的多层环形燃料组件容易发生流道阻塞事故,对流道阻塞情况下的流动与换热特性进行研究非常有必要。本文对多层环形流道阻塞事故在进行三维数值分析研究中的基础问题进行探讨,验证尺度化壁面函数在流量持续降低工况下的适用性。对流道入口发生90%堵塞的工况,分别运用速度入口与压力入口的边界条件进行计算,证明压力入口是流道阻塞事故计算流体动力学(CFD)研究保守、可信的边界条件。对堵塞工况进行瞬态与稳态计算,分析流量与功率的再分配。     

板状燃料组件入口堵流事故下流场和温度场的瞬态数值计算

板状燃料组件具有结构紧凑、换热效率高、深燃耗等特点,故被广泛应用在一体化反应堆和实验用研究堆中。在堆芯窄矩形流道中,冷却剂一般采用自上向下的强迫循环方式。在某些事故工况下,譬如由于燃料元件的辐照肿胀、堆内材料碎片或异物随冷却剂循环流入堆芯,可能引发堵流事故。该事故将造成燃料板失冷,板温升高,可能导致局部冷却剂蒸干,威胁燃料包壳的完整性,甚至造成放射性外泄,引发严重事故后果。本文采用CFD软件ANSYS FLUENT 12.1对板状燃料组件在入口95%部分堵塞和全部堵塞的工况进行了瞬态数值模拟。计算中考虑了冷却剂和燃料板的流固耦合传热问题,并对所得三维流场、温度场及影响因素进行了分析。     

板状燃料元件流道堵塞事故预防与探测技术研究

板状燃料元件用于研究堆中表现出良好的辐照性能。通过对国内外一些使用板状燃料元件研究堆堵流事故实例的调研,发现板状燃料元件板间的栅距通常很小,堆芯冷却剂流道狭窄,堵流事故的发生大都由异物进入流道或燃料肿胀引起。选取中国先进研究堆(China Advanced Research Reactor,CARR)作为特征研究对象,采用RELAP5/MOD3.2热工计算程序,对CARR堆芯、堆本体、单盒组件、堆外冷却回路等进行了热工水力模拟计算,结果表明:当反应堆功率提升时,堵塞的流道内燃料组件温度上升,冷却剂开始发生沸腾,功率会发生明显波动。通过中子注量率与功率的监控以及燃料温度的分析,有助于及早探知和预防堵流事故的进一步发展扩大。     

基于环形燃料的低温供热堆热工水力特性数值模拟研究

为建立低温供热堆热工水力系统的计算流体力学(CFD)仿真模型,针对供热堆堆芯燃料组件结构复杂的特点,采用多孔介质模型对堆芯环形燃料组件进行简化建模,多孔介质的孔隙率、渗透率以及惯性阻力系数通过对1组环形燃料组件精细化CFD模拟结果,采用多孔模型进行拟合得到。典型运行工况的计算结果表明：针对复杂几何采用多孔介质模型简化能大幅提高计算的经济性,多孔介质模型能正确反映参数整体分布趋势,堆芯入口最大流量分配不均匀系数为1.07。本文研究结果对基于环形燃料组件的低温供热堆中热工水力安全设计具有参考价值。     

小型压水堆压力容器内部三维流场计算

反应堆安全分析过程中，获得反应堆压力容器内部准确的流场至关重要。以小型压水堆为研究对象，运用计算流体力学（CFD）方法对反应堆压力容器内部流场进行计算分析，获得燃料组件流量分配和下封头混合特性。结果表明:两泵高速对称入口条件下，燃料组件流量分配系数最大值为1.032，最小值为0.934，且流量整体分布呈现“中间大、边缘小”的特点；一泵高速非对称入口条件下，下封头流动漩涡增强，燃料组件流量分配的不均性增大；下封头混合特性计算得到堆芯入口冷却剂流量混合因子最小值为0.022，下封头冷却剂混合能力不足。      

Calculation of Transverse Acceptance of CYCIAE-100

CARR燃料组件是板状元件，共有21块燃料板和20个矩形冷却流道。流道宽度不均匀，从外到内分别有2．59、2．45、2．32和2．20mm4种宽度，长度均为66．6mm。燃料组件的主要特点是热流密度高、传热性能好。良好的组件流量分配特性是充分发挥这些优点的重要保障。对于这种窄缝流道的水力特性和流量分配实验，国内外均进行不多，加之结构复杂，流速很高，因此，很有必要对它们进行实验验证。     

板状燃料组件窄缝流道流速分布实验研究

实验研究了平行窄缝流道板状燃料组件的流速分布。实验以无离子水作流动介质，在常压及不同Re条件下进行。实验结果表明：在不同Re条件下，横向单通道流速分布呈中间流速高、边缘流速陡减的梯形分布，符合常规流道流速分布特点；纵向平行多通道间的流型呈中间流速低、边缘流速高的凹型分布，这种凹型分布是由入口结构件造成的，对反应堆堆芯流量的均匀分配不利。     

并联通道瞬态流量分配方法研究

应用多通道模型对闭式燃料栅格反应堆进行热工水力分析时,首先需要解决流量分配问题。本文提出了3种流量分配方法,编写了瞬态流量分配程序,求解了1个并联通道流量分配问题,并对这3种方法做了计算对比。结果表明,方法1只适合流量缓慢变化工况;而方法2和方法3适用于流量剧烈变化工况;但在计算同一工况时方法3比方法2更稳定;因此,方法3可作为板状燃料堆芯冷却剂通道流量分配的计算方法。     

环形燃料混合堆芯横向流动特性数值模拟研究

环形燃料混合堆芯横向流动特性对原堆芯的热工安全具有重要影响。本文基于计算流体力学(CFD)方法建立了3×3环形燃料混合堆芯，通过计算混合堆芯的速度场、局部阻力特性与各组件的出入口流量守恒性，对环形燃料与原堆芯燃料之间的横向流动进行了评价。结果表明，当环形燃料与原堆芯燃料轴向各处阻力一致时，原堆芯燃料出入口冷却剂流量相对偏差小于0.8%,环形燃料出入口冷却剂流量相对偏差小于1.8%,混合堆芯各格架段无显著横向流动。     

CFD methodology and validation for single-phase flow in PWR fuel assemblies

This paper presents the CFD modeling methodology and validation for steady-state, normal operation in a PWR fuel assembly. This work is part of a program that is developing a CFD methodology for modeling and predicting single-phase and two-phase flow conditions downstream of structural grids that have mixing devices. The purpose of the mixing devices (mixing vanes in this case) is to increase turbulence and improve heat transfer characteristics of the fuel assembly. The detailed CFD modeling methodology for single-phase flow conditions in PWR fuel assemblies was developed using the STAR-CD CFD code. This methodology includes the details of the computational mesh, the turbulence model used, and the boundary conditions applied to the model. The methodology was developed by benchmarking CFD results versus small-scale experiments. The experiments use PIV to measure the lateral flow field downstream of the grid, and thermal testing to determine the heat transfer characteristics of the rods downstream of the grid. The CFD results and experimental data presented in the paper provide validation of the single-phase flow modeling methodology. Two-phase flow CFD models are being developed to investigate two-phase conditions in PWR fuel assemblies, and these can be presented at a future CFD Workshop.     

窄流道中柔性单板流固耦合数值模拟

板状燃料组件在先进核反应堆中得到了广泛应用。流体以一定流速轴向掠过平行板组件可能导致板的流致振动（FIV），而板的振动又会影响流场的重新分布，两者之间构成强烈的流固耦合（FSI）关系。针对板状燃料组件的FIV现象开发了计算程序。程序基于物理组成贴体坐标系（PCBFC），结合任意拉格朗日欧拉坐标法（ALE）实现网格的移动。本工作详细模拟了在窄通道中移动边界条件下流场的分布;数值求解板在流体压力下的梁式振动方程，从而实现窄流道中柔性单板流固耦合的数值模拟。     

燃料组件少群参数计算程序AFGPB及其基准验证

介绍了自行开发的反应堆燃料组件少群参数计算程序AFGPB的理论方法和模型,并利用IAEA板元件组件基准问题对其进行了验证分析,同时给出了TPFAP和CPM程序对美国CE公司燃料组件功率分布的校算结果.结果表明,AFGPB的计算值与国际上其他机构的结果符合良好.     

Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.     

Extension of Askew's Coarse Mesh Method to Few-Group Problems for Calculating Two-Dimensional Power Distribution in Fast Breeder Reactors

A few-group coarse mesh method has been developed for the calculation of power distribution in 2-dimensional geometry of a fast breeder reactor by extending Askew's one-group coarse mesh method. This method employs modified macroscopic cross sections including group-dependent corrections for coarse meshes of one point per hexagonal assembly and can be easily incorporated into conventional diffusion codes.

Results obtained in few-group 2-dimensional test cases on a prototype fast breeder reactor indicate that this method is as accurate as fine mesh calculations with six mesh points per assembly and the computing time is about ¼ of that of fine mesh calculations.     

CEFR堆芯组件安装方式对组件间流量分配的影响

为研究堆芯组件安装方式对组件间流量分配的影响,通过CFD数值模拟软件FLUENT,对中国实验快堆(CEFR)1型小栅板联箱上的7盒组件在22种安装情况下进行三维数值模拟。由计算结果可知,组件在22种布置情况下的流量与设计值的相对偏差为0.41%~2.03%。根据计算结果可更准确地预测堆内最热管的流量分配,可为今后CEFR的组件安装方式提供参考。     

CFD在核能系统分析中应用的最新进展

介绍了国内外计算流体力学(CFD)方法在核能系统分析中应用的最新进展.CFD已经可以应用到一些三维单相瞬态流动情况中,其中包括堆芯及组件内的流场模拟,以及堆芯外空腔和其他领域的模拟分析.CFD的应用还需要进行进一步验证和基准化,并需要针对CFD的应用方法进行研究,包括建立最佳实践导则(BPG).CFD程序还被用于与热工水力系统程序的多尺度耦合以及与中子物理程序的耦合.CFD程序在两相流领域的应用还处于初级阶段,在湍流和多相流的模拟中需要更进一步发展.     

数值堆热工流体程序CVR-PACA验证及典型应用

谱元方法是一种高精度的数值计算方法,采用该方法开发了数值堆高精度热工水力并行CFD计算程序CVR-PACA。应用CVR-PACA对单棒光棒通道湍流流场、3×3光棒棒束湍流流场、Matis-H压水堆棒束通道基准题、19棒带绕丝组件通道湍流流场进行了仿真计算。通过与实验测量值对比,研究定量验证了大涡模拟(LES)模型及非稳态雷诺时均(URANS)模型对各类棒束通道流场预测的准确性。算例在建模过程中采用网格分裂技术实现了复杂几何的纯六面体网格划分,用于支撑谱元方法计算。研究较为全面地积累了高精度谱元方法模拟流场流动及换热的建模经验,获取了各类棒束通道内丰富的流动和换热细节,获得的建模经验能更加精准有力地指导相关设计的优化改进。     

板型燃料元件反应堆瞬态热工水力分析程序的开发与验证

针对核动力系统瞬态分析的需求，建立板型燃料反应堆的热工水力数学物理模型，开发了具有自主知识产权的核动力系统瞬态热工水力分析程序SYSTRAN，并采用中国先进研究堆（CARR堆）的设计工况和国际原子能机构（IAEA）基准题的堵流瞬态数据对程序进行了验证。计算结果表明，堆芯流量分配、出口温度等关键参数与验证数据吻合良好，初步证明了本程序适用于板型燃料反应堆系统瞬态热工水力分析。      

Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.