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     

带定位格架棒束通道内两相流型研究

在空气-水两相流动工况下,将RBI光学探针测得的时序波形和目测相结合,对AFA-2G 33定位格架组成的棒束通道内存在的两相流型进行了识别。通道水力当量直径为8.98mm,元件的棒径为9.5mm,栅距为12.6mm,棒壁距为2.65mm。液相和气相表观速度范围分别为0.40-2.69m/s,0.02-2.99m/s。试验获得了流型图。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型变化有重要影响,在棒束通道内的同一截面上存在不同种类流型。     

棒束定位格架两相CFD模拟方法研究

考虑气泡合并分裂,采用MUSIG模型,对3×3格架内空气-水两相分布进行计算流体力学(CFD)数值模拟研究发现,计算对入口两相分布预计不敏感,但对气泡直径大小敏感;在定位格架下游不远处,空泡份额分布由较小直径气泡起主导作用,格架下游较远处,空泡份额分布由较大直径气泡起主导作用。考虑空气-水两相流量、几何条件和压力对气泡直径的影响,本文提出针对棒束定位格架的数值模拟气泡最大直径设置关系式,并对模型选取和模拟方法给出建议。计算表明空泡份额分布曲线形状与峰值均和实验符合较好,该模拟方法能合理预测复杂通道两相数值分布。     

7棒束紧密栅元流体流动传热数值研究

紧密栅元内的流体流动传热研究对高转化比反应堆燃料组件的优化有十分重要的意义。本文采用CFD方法对7棒束紧密栅元棒束通道内流体流动传热现象进行了数值模拟,并与7棒束紧密栅元内氟利昂流体传热的实验结果进行对比分析,详细分析了定位格架对棒束内流体传热流动的影响。结果表明：数值计算所得的非加热棒的壁面温度和实验吻合良好,定位格架的存在对其下游流体流动、棒束最高温度分布及交混系数有明显的影响,棒束某些位置因流动滞止导致温度大幅上升,在设计中应加以注意。     

燃料组件精细化定位格架模型开发及评价

为提高燃料组件子通道内两相局部参数预测的准确性,本文基于分布式阻力方法建立精细化定位格架模型,选用合适的摩擦阻力表达式,对格架上的交混翼进行精细化建模,采用Carlucci湍流交混模型计算湍流交混速率,引入阻塞因子计算由定位格架引起的湍流交混效应,并将建立的精细化定位格架模型植入子通道分析程序（ATHAS）,对压水堆子通道和棒束实验（PSBT）基准题进行计算分析。结果表明,本文开发的精细化定位格架模型能够提高燃料组件子通道内空泡份额和温度分布的预测准确性,为棒束通道流场、焓场计算和临界热流密度（CHF）预测奠定了基础。      

带格架四棒束超临界水流动传热数值分析

棒束内超临界水流动传热是超临界水堆堆芯热工水力研究的重要内容,但对其认识还十分有限。本文针对四棒束内超临界水的流动传热现象开展数值模拟,特别分析了定位格架对棒束通道内流动和传热的影响。结果表明,采用SSG湍流模型计算所得到的棒束壁面温度和实验结果吻合良好,定位格架的存在影响下游流体的速度分布,显著提高格架下游的传热特性,交混系数有大幅上升,使得加热棒周向壁面温度分布更加平均,最高温度出现位置发生改变。     

全长尺寸5×5格架棒束通道两相流动特性研究

采用两相计算流体动力学(CFD)分析的方法,对全长尺寸格架棒束通道内过冷沸腾两相流动进行了数值模拟。将模拟得到的棒束通道中心4个子通道的平均空泡份额与实验值进行对比发现,在高空泡份额区域与实验值符合较好;在低空泡份额区域,计算值略高于实验值。两相CFD方法模拟得到了棒束通道内空泡份额的详细分布,观察到格架上游空泡份额集中在加热棒的周围,但在格架下游,子通道中心的空泡份额增加,加热棒周围的空泡份额减小,间接地证明了格架对临界热流密度(CHF)的提升作用。     

压水堆棒束定位格架两相搅混特性数值研究

本文提出一种CFD方法用于评价压水堆燃料棒束定位格架两相搅混特性。针对两种典型的定位格架,采用CFX12.0进行了空气-水两相流动的数值模拟,并与采用氟里昂工质开展的临界热流密度（CHF）实验进行对比。结果表明,CFD方法可初步应用于评价格架下游汽泡的分布特性。     

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。     

URANS与LES对带分裂式交混叶片定位格架5×5棒束通道流场数值模拟研究

利用非定常雷诺平均纳维斯托克斯模拟（URANS）和大涡模拟（LES）对带分裂式交混叶片定位格架5×5棒束通道流动特性进行了研究。数值计算中建模考虑了格架条带、交混叶片等几何结构对流场的影响,并将模拟结果与MATiS-H基准实验进行了对比。结果表明,URANS与LES均能较好地模拟格架下游3个流速分量时均值;对于格架下游流速分量脉动值,URANS中非定常SST k?–?ω模型几乎不能够模拟出流速脉动值,非定常RSM模型对于流速脉动值模拟比实验值偏低。与URANS相比,LES能相对较为准确地模拟流速脉动值,然而LES对格架附近流速脉动值模拟结果与MATiS-H基准实验相比仍然偏低。      

CFD prediction of flow and phase distribution in fuel assemblies with spacers

This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.     

Experimental and numerical simulation of air-water two-phase flow in the rod bundle with grid spacer

1 Introduction Grid spacer is the key part of reactor fuel assem-bly. The presence of spacers in fuel assemblies affectsvarious thermal-hydraulic characteristics of the reactorcore. The grid spacer with fine performance can im-prove thermal-hydraulic performance of the core fuelassembly and enhance the critical heat flux withouttoo much augment of the pressure loss. As a result,the implementation of grid spacer with high thermalperformance provides more thermal margin, then in-creases s…     

Development of drift-flux model based on 8 × 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 × 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 × 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data.     

棒束定位格架内单相流体三维流场研究

利用计算流体动力学(CFD)方法对带定位格架棒束内单相流体三维流场进行研究,以了解定位格架对流体流动的影响。用激光多普勒测速系统测量的实验数据对预测结果进行了校验,两者吻合较好。     

Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles

The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90° the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.     

Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60° with respect to the horizontal. The measured phase distributions indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape; (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region.     

CFD application to prediction of void distribution in two-phase bubbly flows in rod bundles

This paper is concerned with the prediction of the void fraction distribution in two-phase bubbly flows in fuel rod bundles. Special attention has been devoted to the phenomena which govern the void fraction distribution in the lateral direction of a channel. A two-fluid model of two-phase flow has been formulated and implemented into a commercial computational fluid dynamics (CFD) code. The model has been used for the prediction of the void distribution in three different channels: a circular channel (inside diameter (ID), 34.5 mm) with a single heated rod of 13.9 mm outside diameter (OD), and circular channels (ID, 71 mm) with six heated rods (13.8 and 13.9 mm OD each). The predicted axial and lateral avoid fraction distributions in subcooled and bulk boiling regions have been area averaged in three lateral zones and compared with experimental data: in all cases, satisfactory agreement between the predictions and measurements has been obtained.     

计算流体动力学在燃料组件热工水力性能分析和格架研制中的应用

国外使用商用计算流体动力学(CFD)软件分析燃料组件中流体的三维流场和温度场,并将验证的方法用于燃料组件格架设计,获得了成功。中国核动力研究设计院空泡物理和自然循环重点实验室用CFX程序对带格架棒束内流场进行了计算,解决了小尺寸复杂结构几何体的模拟,边界条件的选取和CFX计算能力的评价,然后完成了单相,空气一水两相流场和流动特性的计算分析及试验对比验证。已完成的研究表明,尽管CFX程序目前在计算两相流动和传热方面还存在不足,但通过比较单相流场的湍流,旋涡和棒束附近流体温度分布基本可以评价格架对流体的交混性能;格架上的弹簧和刚突对于流动有相当的作用,对其进行模拟是必要的。研究还建议在使用CFD方法进行燃料组件格架热工水力分析前要先进行基准练习以保证分析结果的正确性。