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。     

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。     

组合阀步降流道三维流场数值分析

运用计算流体力学程序CFX对组合阀步降流道内三维流场进行了分析和计算,流道水力学特性计算结果与组合阀步降流道流动阻力实验结果符合良好。在此基础上,对流道内流速和压力分布进行了研究,结果表明:阀芯流道侧壁入口上端流体流速达到峰值,回零水腔内流体的流速分布均匀,流道内压力损失最大的部位处于入口流道竖管段与横管段之间,可通过增大管径或加缓变过渡段的方法来减少该部位的流动阻力损失。     

花瓣形燃料元件棒束通道内过冷流动沸腾特性数值研究

花瓣形燃料元件具有换热性能强和无需定位格架等优点，能进一步提高反应堆的功率密度和经济性。为此，本文利用欧拉两流体模型，同时结合RPI壁面沸腾模型，对2×2花瓣形燃料元件棒束通道内过冷流动沸腾特性开展数值研究。通过圆管过冷沸腾实验数据验证了模型的准确性。开展了流速和热流密度参数对花瓣形燃料元件棒束通道内流动、换热及空泡份额分布影响的数值研究。结果表明，通道内冷却剂的流动速度分布不均匀；横向流动沿主流方向存在波动；空泡份额在燃料元件的内凹弧与外凸弧处表现出较大差异；同时，由于流场和换热形式的不同，导致燃料元件的周向壁面温度呈现不均匀分布，横向流动的存在影响着壁面热流分配情况。     

先进中子学栅格计算程序KYLIN-Ⅱ共振计算基准验证

针对先进栅格计算程序KYLIN-Ⅱ的共振计算模块，开展了多个基准题数值验证，包括单栅元、IAEA板型燃料组件、钍基组件、多层套管型燃料组件、带可燃毒物的燃料栅格、带中心大水腔的超临界水堆燃料栅格、AFA3G含钆燃料组件基准题。验证结果表明，本文的共振计算模块适用于棒栅元方型组件、板型燃料组件、六角形组件等几何结构较复杂的问题，同时可以正确计算含有铀、钍、毒物等复杂材料的共振问题，满足未来工程使用的需求。      

池式反应堆堆内流场数值模拟

以板型燃料组件池式反应堆为研究对象,采用计算流体动力学程序CFX5对堆内流场进行了数值模拟,结果表明:流过堆芯燃料组件的流速较大,在燃料组件出口位置流速达到最大值;无论是否带有围桶,堆内压降均主要集中在堆芯燃料组件上,入口流量增大,堆芯燃料组件上的压降随之增加;堆芯上部腔室和下部腔室的压力变化很小;在相同的入口流量下,带与不带围桶的堆芯进出口差压非常接近。     

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。     

中国先进研究堆矩形通道流场数值计算分析

通过SIMPLE数值方法,编制程序,对中国先进研究堆(CARR)全流道进行流场数值模拟.采用对CARR的单个冷却剂通道进行单相水的数值传热计算,并递增地改变流道入口流速,计算获得与入口流速对应的流道速度场与温度场分布,展现其变化规律,分析入口流速对流道热工水力参数分布的影响.采用所编制的程序,对板式燃料组件构成的窄矩形通道进行数值模拟,由此来确定热工水力设计需要的一些反应堆安全参数.这些安全参数为反应堆事故监测系统提供必要的热工过程状态信息,也为CARR提供必要的数据参考.     

环形燃料元件内冷却剂流动换热特性的数值研究

相较于传统棒束燃料元件，内外双冷却通道的环形燃料元件具有堆芯功率密度高同时燃料温度低的优点，研究其热工水力特性具有重要意义。本文采用计算流体动力学(CFD)方法对内外冷却的环形燃料元件内外冷却流道的流动沸腾进行数值模拟，根据模拟结果对内外冷却流道的温度场、二次流速度及换热系数等参数进行分析。结果表明：最大二次流速度出现在燃料棒近壁面处；环形燃料元件外流道温度场分布呈现间隙处温度高，各子通道温度低的分布趋势；固体燃料棒表面温度在轴向同一位置处，沿周向以90°为周期变化；换热系数呈现规律性波动，单棒的不同周向角度换热系数存在较大差异，沿周向以90°为周期变化，周向角度为45°、135°、225°和315°位置处均出现温度极大值。本文结果可为环形燃料元件工程应用提供理论参考。     

5×5花瓣形燃料棒束组件内单相流动与换热特性数值模拟研究

相较于传统圆柱形燃料棒,花瓣形燃料棒具有安全裕量高等优点,研究其在压水堆运行工况下的热工水力特性具有重要意义。本文通过STAR-CCM+对5×5花瓣形燃料棒束组件进行数值模拟研究,计算并分析了组件内二次流速度、温度、换热系数等关键热工参数,获得了入口流速、螺旋节距对组件内部流动与换热特性的影响规律。计算结果表明：花瓣形燃料棒的螺旋结构可增强冷却剂的横向流动,同一高度上燃料棒表面温度分布具有周期性,增大入口流速可增强燃料棒的表面换热,消除温度分布的不均匀性。此外,螺旋节距大于750 mm,燃料棒换热性能与无扭转的燃料棒相差不大,甚至更低。     

A study of non-Boussinesq integral boundary layer analysis for an MTR research reactor

In the present work, a non-Boussinesq (variable physical properties) integral boundary layer analysis is accomplished. The model analyzes laminar free convection between nuclear fuel plates having large fuel plate length to gap between plate ratio. The coolant channels are undergoing to a uniform, symmetric, heat flux and varying fluid properties. In the present study the flow is assumed to be fully developed. This is a good assumption for channels with large fuel plate length to gap between plate ratios. To describe the velocity and temperature distributions of the coolant the non-Boussinesq approximation is introduced into the integral boundary layer equations of flow between parallel plates. The fuel plate temperature is related to the adjacent coolant fluid temperature by a principle in conduction heat transfer. Fluids considered here are air and water. The obtained results show that the present heat transfer problem encountered in nuclear research reactor such Tehran nuclear research reactor (TRR) is characterized by high temperature ratios and thereby rendering the commonly applied Boussinesq approximation invalid. As a result, the use of the Boussinesq approximation (constant fluid properties) for high temperature ratios is not suggested.     

Three-dimensional analysis of flow blockage accident in Tehran MTR research reactor core using CFD

In the present study, a 3D simulation of flow blockage accident which may occur in the coolant channels of a fuel assembly of Tehran research reactor (TRR) is investigated using CFD code. Consideration is given to the scenario in which partial blockage of hot channel occurs due to buckling of its fuel plates towards the inside. Governing conservation laws are solved using Control volume approach and pressure field is coupled to the velocity field through the SIMPLE algorithm. Flow convergence is considered when the residual for all flow variables are less than 10⁻⁵. The simulation is performed under four different obstruction levels of the nominal flow area, i.e., 0%, 20%, 50% and 70%. By solving momentum and energy equation in three channels with their fuel plates, it is found that heat transfer is substantially affected by channels flow field. In the blockage accident, decrease in flow rate of the obstructed channel decreases cooling capacity of the obstructed channel as a result of hydraulic resistance augmentation. The obtained results show that above the 50% blockage, critical phenomena will appear which may compromise the clad integrity. Moreover, in the 70% blockage scenario, the clad temperature in the obstructed channel reaches the value associated with nucleate boiling temperature at the operative pressure.     

Numerical Analysis of the Lifting Beam's Influence on Flow Fields in Planar Fuel Assemblies

This study used the Computational Fluid Dynamics (CFD) program CFX5 to investigate the flow field in planar fuel assemblies with two types of lifting beams, and to determine the effect of the two types of lifting beams on the flow field in planar fuel assemblies. This information is helpful in optimizing the design of a planar fuel assembly. The results indicated that the first type of lifting beam had a similar effect on the planar fuel assembly as the other type, the flow distribution in narrow flow channels between fuel plates with the first lifting beam was similar to the other, and the pressure difference of the inlet and outlet of the planar fuel assembly was also similar for the two types of lifting beams. The results also indicated that the two types of lifting beams' effect on the flow field in narrow flow channels between fuel plates is small, so either of the lifting beams could be used in the planar fuel assembly's design.     

Integral boundary layer heat transfer modeling for Tehran Research Reactor

In the present work the validity of applying the Boussinesq approximation in the analysis of natural convection heat transfer along nuclear fuel plates with large coolant channel aspect ratios is evaluated. The Boussinesq approximation is introduced into the integral boundary layer equations governing the system to describe the velocity and temperature distributions of the coolant in the cooling channels. The fuel plate temperature is related to the adjacent coolant fluid temperature by a fundamental law in conduction heat transfer. Air and water are considered as fluids. The coolant flow is assumed to be fully developed which is a convenient assumption for coolant channels having large aspect ratios. Obtained results indicate that the Boussinesq approximation is merely applicable over a limited range of coolant channel outlet fluid temperatures. The use of this approximation produces conservative estimation of the critical plate power for air flow and non-conservative estimation of the critical plate power for water flow.     

Hydraulic study of turbulent flow in MTR-type nuclear fuel assembly

For a more accurate safety analysis for the McMaster nuclear reactor (MNR), local velocity measurements in a mock-up of the 18-plate fuel assembly were conducted in the low Reynolds number region with emphasis on edge effects. With modification of the curved plates in the assembly to flat ones, velocity in the narrow channels was measured by laser Doppler velocimetry (LDV). The velocity profile in individual channels and the channel-to-channel flow distribution were obtained. The effect of the sudden transition from the larger rectangular channel to the circular pipe on the velocity profile and the effect of the handle presence in the top end fitting on the flow distribution were investigated. The pressure drop along the assembly was measured and compared to a simple 1D correlation prediction.     

钠冷快堆乏燃料组件自然循环冷却瞬态过程的数值模拟研究

应用Fluent程序,对处于氩气中的钠冷快堆乏燃料组件自然循环冷却瞬态过程进行了三维数值模拟。计算获得了乏燃料组件内部冷却剂通道和外部区域的热工水力学现象及变化规律。结果表明:利用标记区域分割方法,将燃料棒间隙网格划分为绕丝网格和绕丝周边流体域网格,能在棒束区生成高质量结构化网格;在氩气自然循环冷却瞬态过程中,棒束区内子通道氩气流量增加速度落后于边子通道,内子通道升温更快;乏燃料组件棒束区温度在轴向呈现中心高、边缘低的分布特征;为避免包壳温度超过设计值,乏燃料组件处于氩气中的时间不宜超过6min。     

Flow-induced plastic collapse of stacked fuel plates

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental nuclear reactors such as the Oak Ridge National Laboratory High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by the nuclear fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. In this paper a methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.     

Laminar flow induced deflections of stacked plates

Some research and power reactors such as the Engineering Test Reactor (ETR), the Materials Test Reactor (MTR) and the Shippingport Reactor have core designs which consist of parallel, flat or curved plate fuel assemblies. The fuel is contained in the thin plates which are separated by narrow channels through which coolant flows to remove heat generated within the plates. Since the plates are flexible, the coolant flowing through the channels causes the plates to deflect. At high coolant velocities large deflections have been observed causing the plates to deform plastically leading to structural failure or plate collapse. This work examines a single plate bounded by two channels and determines the static plate deflection as a function of plate, channel and flow parameters. The deflection is due to differences in pressure and flow velocity in the channels bounding the plate and also due to different channel dimensions caused by tolerance effects. The classical thin plate equations are used with a nonlinear hydrodynamic loading function expressing the external fluid forces on the plate surfaces.     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods.In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels.Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.