基于图像算法的棒束通道流动交混研究

反应堆燃料组件内冷却剂流动特性是反应堆热工安全分析重要内容,本文针对带定位格架的棒束通道内可视化实验图像数据,基于图像算法对棒束通道内流动交混现象进行了定性与定量的分析,为堆芯热工水力及安全分析提供参考。从数据验证、规律总结、机理分析3个层面上提出了图像处理方法,并采用这些方法对棒束通道内实验数据进行了分析,分析结果表明:定位格架下游流体的流动交混在5 cm开始逐渐缓和;Canny算子适用于定位格架湍流图形的边缘识别。     

湍流模型对5×5格架棒束通道流动传热数值模拟影响分析

同一软件工具采用不同湍流模型进行燃料组件格架棒束通道CFD分析时会得到不同的数值结果,本文采用ANSYS CFX软件,建立了包含典型5×5格架的棒束通道CFD模型,研究了涡粘和雷诺应力两大类6种典型湍流模型对燃料组件压降与换热特性数值结果的影响,计算了压降和Nu分布结果与相似的实验结果进行对比,通过分析3个典型搅混效果评价因子,探讨了搅混翼流动与换热的内在影响关系,同时对比了不同湍流模型对结果的影响。通过与相似实验数据对比分析,认为雷诺应力模型较适宜计算本文所研究的定位格架及棒束通道内流动传热特性。     

棒束通道内定位格架搅混特性PIV可视化研究

定位格架作为燃料组件中重要的组成部件之一,不仅在结构上固定燃料棒,而且在燃料组件内热工水力性能同样显著,特别是对工质的搅混性能直接关系到反应堆的经济性和安全性,因此有必要对燃料组件内定位格架搅混特性进行研究。本文通过粒子图像测速（PIV）技术开展了棒束通道内定位格架上下游流场的可视化研究,对比了有无格架棒束通道内流场的分布特征,定量分析了定位格架对棒束通道流场搅混的贡献。对不同流速下定位格架下游横纵速度的沿程变化特性进行研究,发现了不同流速作用下定位格架对横向、轴向速度的促进和抑制规律。另外,通过速度均方根对下游的湍流特性进行了评估。实验结果可为数值计算提供全场的数据验证,并可为定位格架设计和优化提供基础。     

液态铅铋合金在绕丝燃料棒组件子通道间湍流交混数值模拟

液态铅铋合金（LBE）是第四代液态金属核反应堆候选冷却剂,由于LBE热物性具有一定的特殊性,亟待对LBE在燃料组件子通道中的流动与传热过程开展研究。本文对LBE在带绕丝燃料棒组件中湍流流动进行数值模拟与分析,将燃料棒壁面温度的数值模拟结果与响应的实验数据相比较,2者具有较高的吻合度,说明数学模型及数值结果具有较高的可靠性与准确性;使用湍流交混系数β表征LBE在不同子通道间、不同燃料棒间隙宽度与燃料棒直径比（S/D）结构下的湍流交混情况,结果表明,不同子通道间β波动程度具有差异性,β的大小与S/D呈负相关。基于不同S/D与雷诺数的计算结果,拟合出不同子通道间β关联式,为绕丝燃料棒三角形排列方式的燃料组件子通道分析程序开发提供交混模型。      

定位格架模型对子通道分析程序的影响研究

子通道分析程序只能进行简单的定位格架模型分析,对格架的考虑仅以形阻系数表示,仅能反映轴向的平均阻力效应,难以反映格架上不同角度和排布的交混翼对流场及温度场的影响,不能准确反映格架中的交混翼对局部参数的影响。本文通过选用适当的阻力经验关系式,引入交混翼几何尺寸及交混翼角度,将交混翼对流体产生的力定量地表示出来,添加至动量方程中,建立了新的交混翼分布式阻力模型,并将其耦合到ATHAS子通道分析程序中。程序经过对5×5棒束组件在不同结构下的计算,对比有交混翼和无交混翼子通道内的流场,研究不同形状、角度、排列方式的交混翼产生的效应。     

带MVG和MSMG的5×5全长棒束单相流场和温度行为数值分析

定位格架作为燃料组件的关键部件之一,直接影响到燃料组件的热工性能。本文对带结构格架（MVG）和跨间搅混格架（MSMG）的5×5全长加热棒束单相流场和温度场采用计算流体力学（CFD）程序进行数值分析研究,获得该特征棒束组件出口二次流场以及温度场分布特性。研究表明,定位格架下游流场受定位格架和距离的影响,定位格架上游流场对下游二次流几乎无影响,定位格架导致流体强烈的横向二次流,增强了流体和加热棒之间的换热能力,使得棒束子通道截面流体温度更加均匀。与5×5全长棒束出口子通道温度的实验数据对比分析表明,获得的计算模型可以较好地分析该型棒束组件结构温度场行为。      

脉动流下定位格架下游时均流场分布特性研究

研究流量波动下棒束通道内定位格架下游瞬时流场演变特性对于揭示海洋条件下燃料组件内流动换热机理具有重要意义。本文应用粒子图像测速（PIV）技术获得了脉动流下棒束通道内定位格架下游时空演变流场结构,分析了脉动参数（脉动周期和脉动振幅）对定位格架下游速度分布和湍流特性的影响。结果表明,脉动流下定位格架下游时均速度与定常流动下时均速度差异较小,且基本不随脉动振幅和脉动周期变化而变化;脉动流下的定位格架下游横向速度和轴向速度均方根与定常流动下的速度均方根存在明显差异,且随脉动参数变化呈现出不同的变化趋势。本文研究结果有助于揭示燃料组件在非稳态条件下瞬态特性,并为燃料组件的设计和优化奠定基础。      

棒束定位格架空泡份额分布特性实验研究

带定位格架棒束通道内的空泡份额分布特性是反应堆热工水力特性研究的重要内容。对AFA 2G3× 3定位格架组成的棒束通道在空气 水两相流动工况下用RBI光学探针测得了通道内的横向空泡份额分布 ,分析了其横向分布的一般规律。结果表明 ,定位格架结构 ,特别是交混叶片对定位格架附近区域两相流动和空泡份额分布特性有重要影响 ,从而为进一步研究棒束定位格架加热工况下两相流动特性 ,发展新型高热工水力性能燃料组件打下基础     

棒束内湍流流动特性的PIV与CFD研究

在压水堆燃料组件的定位格架下游,局部扰动沿流动方向逐渐衰减,流场最终趋于稳定。光滑棒束区冷却剂的湍流流动和交混特性是影响反应堆经济性和安全性的重要因素,有必要进行深入研究。本文采用粒子图像测速（PIV）与数值模拟(CFD)相结合的方法,对3×3小规模棒束内水的流动特性进行研究,得到了一阶平均流速以及二阶湍流统计信息。结果表明,中心子通道的速度明显高于棒间隙区,但轴向均方根速度呈现出相反的变化趋势。在相邻子通道横向速度梯度的作用下,棒束内出现了大尺度的流量脉动现象,且脉动波长随雷诺数的增加而增大。此外,实验得到的湍流交混系数较压水堆采用的Castellana公式预测值偏高10%左右,这一偏差随雷诺数的增加有减小的趋势。     

单相4×4棒束流动试验的CFD方法验证

为研究计算流体力学（CFD）方法预测棒束通道内流场分布的准确性,基于网格敏感性分析所确定的网格方案,采用标准k-ε模型（SKE）、可实现k-ε模型（RKE）、标准k-ω模型（SKW）和剪切应力传输模型（SST模型）对单相棒束流动进行模拟,并将横向速度与轴向速度与试验结果进行量化比较。结果表明:4种湍流模型均能较好地预测棒束通道内的流场分布,其中SKE与RKE的在横向速度预测上相对偏差较小,为19.6%;对于近格架区域的横向流场分析,SKE模拟较优,反之RKE模拟较优;对于轴向速度的预测,SKE的相对偏差最小为4.9%;4种湍流模型均低估均方根（RMS）速度,但能够预测棒束通道内RMS速度的分布规律,近格架区域采用RKE,反之SST较优。本文的计算结果可为单相棒束流动CFD分析的最佳实践导则建立提供参考。      

定位格架下游湍流特性测量研究

在子通道雷诺数为6 600、13 200、26 400和39 600下,使用粒子成像测速仪对5×5棒束分流型交混翼定位格架下游横向和纵向流动进行测量。平均速度和湍流脉动速度均方根的实验结果最大不确定度低于1%的主流平均速度。格架下游二次流结构经历了交混翼脱落涡结构耗散、剪切产生双涡结构、双涡结构向单涡结构的转变及单涡结构沿程衰减过程,横向平均速度和湍流脉动速度均方根沿程变化均受涡结构演进影响。格架近场湍流统计量迅速衰减;格架远场湍流统计量缓慢衰减,流动趋于光棒束充分发展流动。横向流动受雷诺数效应和格架交混效应共同影响。     

PIV and CFD Study of Turbulent Flow Behavior in Rod Bundle

At the downstream of the spacer grid in a PWR fuel assembly, local disturbance damps out along the flow direction and the flow returns to stable eventually. The turbulent flow and mixing behavior of the coolant are key factors affecting the economy and safety of a nuclear reactor, and need in-depth investigations. In the present paper, the turbulent flow of water in a 3×3 rod bundle was studied using PIV (particle image velocimetry) and CFD. First-order mean velocity and second-order turbulent statistics were obtained. It is found that the velocity in the central subchannel is higher than that in the gap region, but the streamwise root-mean-square velocity behaves inversely. Large-scale flow pulsation induced by the strong streamwise velocity gradient between adjacent subchannels, is observed in the rod bundle, and the wave length increases with Reynolds numbers. In addition, the measured turbulent mixing coefficient is 10% higher than that predicted by the Castellana correlation for PWRs, but this deviation reduces with the increase of Reynolds numbers.     

Analytical prediction of turbulent friction factor for a rod bundle

An analytical calculation has been performed to predict the turbulent friction factor in a rod bundle. For each subchannel constituting a rod bundle, the geometry parameters are analytically derived by integrating the law of the wall over each subchannel with the consideration of a local shear stress distribution. The correlation equations for a local shear stress distribution are supplied from a numerical simulation for each subchannel. The explicit effect of a subchannel shape on the geometry parameter and the friction factor is reported. The friction factor of a corner subchannel converges to a constant value, while the friction factor of a central subchannel steadily increases with a rod distance ratio. The analysis for a rod bundle shows that the friction factor of a rod bundle is largely affected by the characteristics of each subchannel constituting a rod bundle. The present analytic calculations well predict the experimental results from the literature with rod bundles in circular, hexagonal, and square channels.     

棒束通道单相和两相交混模型评估

本研究利用子通道程序,基于已有的实验数据,对棒束通道的单相和两相交混模型进行了评估。单相交混主要考虑横流和湍流交混,横流由守恒方程决定并在流量分布中占主导作用,湍流交混取决于交混系数,对湍流交混研究发现Sadatomi模型预测结果与实验结果吻合较好。两相交混由横流、湍流交混和空泡漂移共同作用,通过已有模型预测结果与实验数据对比分析,推荐两相交混中空泡漂移采用Hotta模型、湍流交混系数采用Sadatomi模型和两相乘子采用Beus模型,这是一个预测结果较为保守的组合模型,有利于反应堆安全的保守性评估。      

Numerically simulating the thermal-hydraulic characteristics within the fuel rod bundle using CFD methodology

With the dramatic progress in the computer processing power, computational fluid dynamics (CFD) methodology can be applied in investigating the detailed knowledge of thermal-hydraulic characteristics in the rod bundle, especially with the spacer grid. These localized information, including flow, turbulence, and heat transfer characteristics, etc., can assist in the design and the improvement of rod bundles for nuclear power plants. In this paper, a three-dimensional (3D) CFD model with the Reynolds stresses turbulence model is proposed to simulate these characteristics within the rod bundle and subsequently to investigate the effects of different types of grid on the turbulent mixing and heat transfer enhancement. Two types of grid designs are used herein, including the standard grid and split-vane pair one, respectively. Based on the CFD simulations, the secondary flow can be reasonably captured in the rod bundle with the grid. The split-vane pair grid would enhance both the flow mixing and the heat transfer capability more than the standard grid does, as clearly shown in the simulation results. In addition, compared with the results of experiment and correlation, the present predicted result for the Nusselt (Nu) number distribution downstream the grid shows reasonable agreement for the standard grid design. However, there is discrepancy in the decay trend of Nu number between the prediction and measurement for the split-vane pair gird. This would be improved by adopting the finer mesh (y⁺ < 1) simulation and Low-Reynolds form turbulence model, which is our future research work.     

棒束通道内温度场分布特性研究

可视化实验技术越来越多地被应用于核反应堆系统参数的测量,本文基于激光诱导荧光（LIF）技术的特点,介绍该技术的难点和解决方案,并对棒束通道定位格架下游稳态流和脉动流下温度分布进行了研究。结果显示,通过对系统光学特性和染色剂特性研究,可提高LIF技术的应用范围和测量精度。同时采用后处理技术,可获得更准确的温度场分布。通过对棒束通道定位格架下游全场温度进行测量,获得了稳态流和脉动流两种工况下温度的分布。定位格架能显著增强下游的流动搅混,提高换热能力。流速的波动也会对温度分布产生显著影响。研究表明,LIF技术可实现对棒束通道内流体温度分布的全场测量,根据温度分布特性研究可实现对定位格架性能的评价。     

Temperature Field Distribution Characteristics in Rod Bundle Channel

Visualized experimental techniques are increasingly used in the measurement of nuclear reactor system parameters. Based on the characteristics of laser induced fluorescence (LIF) technique, the difficulties and solutions of the LIF technique were introduced in this paper. And the temperature distributions downstream of spacer grid in rod bundle channel under the steady flow and pulsating flow were analyzed. The results show that the application range and measurement accuracy of LIF technique can be improved by studying the optical properties and dye characteristics. At the same time, post-processing technology can be used to obtain more accurate temperature field distribution. The full-field temperature distributions downstream of spacer grid in rod bundle channel under steady flow and fluctuating flow conditions were obtained. Spacer grid can significantly enhance flow mixing and improve heat exchange capacity. Temperature distribution is also affected by fluctuations in velocity. In summary, the LIF technique can achieve the full-field measurement of the temperature distribution in the rod bundle channel. According to the temperature distribution characteristics, the performance of the spacer grid can be evaluated.     

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k–ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.     

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k-ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.