棒束通道定位格架沸腾临界特性数值研究

为研究核反应堆中定位格架及搅混翼对沸腾临界现象产生的影响,本文采用计算流体动力学（CFD）分析方法,探讨了棒束通道中定位格架的数目、位置和搅混翼的角度对于沸腾临界现象的影响。结果表明:定位格架会对主流流动产生阻力,同时定位格架数目越多,沸腾临界发生的温度也越高,但将定位格架布置在沸腾临界发生位置时,则可有效改善壁面传热环境并降低沸腾临界发生时的峰值温度。搅混翼的存在则会有效降低加热面附近空泡份额,改善传热环境,但搅混翼角度过大时会导致沸腾临界提前发生。      

Modifications to COBRA-TF to model dispersed flow film boiling with two flow,four field Eulerian–Eulerian approach – Part 1

In the case of a postulated loss of coolant accident (LOCA) in a nuclear reactor, an accurate prediction of clad temperature is needed to determine the safety margins. During the reflood phase of the LOCA, when the local void fraction is greater than 80% with the wall temperature above minimum film boiling temperature (T_min), the heat transfer process is dispersed flow film boiling (DFFB). This study has been performed to model DFFB in the reflood phase of a LOCA in a pressurized water reactor (PWR) rod bundle. The COBRA-TF computer code is utilized, since it has a detailed reflood package which takes into account the effect of spacer grids on the local heat transfer. The COBRA-TF code has also been improved to include a four field Eulerian–Eulerian modeling for the two-phase dispersed flow film boiling heat transfer regime. The modifications include adding a small droplet field to COBRA-TF as the fourth field. In addition, the spacer grid models of COBRA-TF have been revised and modified. In the first part of the paper, the results of the code predictions are presented by comparing the experimental data from rod bundle heat transfer (RBHT) experiments with the results of code simulations performed with original and modified code. Measurements and calculations for the heater rod, vapor temperatures and quench front progression have been compared and the results are described in detail. The results of the analysis performed with the modified code indicate the improvement in code predictions for the rod surface temperature, vapor temperature and quench front behavior. The results also indicate the need for improvement in the entrainment and interfacial drag models for the drop fields. The effects of spacer grids on the heat transfer, the models improved and developed for spacer grids and the results of the code calculations with these models are described in the part 2 of the paper.     

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。     

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of droplets after collision with a wet grid spacer depending on flow conditions.     

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.     

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。     

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。     

Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.     

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.     

Modele de comportement mecanique d'element combustible dans un assemblage PWR

Irradiated fuel in pressurized water reactors (PWRs) frequently displays rod bowing, due to two kinds of assymmetry. The first originates in the fabrication of the sheath, causing eccentricity, ovalization and thickness non-uniformity. The other comes from in-pile fuel element conditions such as off-line grids compressing the rods, circumferential thermal gradients on the sheath, and pellet-clad interactions.The MAC code was developed for parametric studies of some of these effects. It shows that:In the case of fuel rods undergoing compressive forces by the spacer grids the usual friction forces are unable to bow the rods significantly, except when the rods are blocked by the spacer grid springs.In some assembly configurations, the temperature difference between adjacent rods is able to bow them, requiring an increase in number of spacer grids.Localized pellet-clad interactions may cause significant bowing, particularly when they occur near the grids.     

Flow Characteristics in Spacer Grids Measured by Rod-embedded Fiber Laser Doppler Velocimetry

Precise measurement of velocity in fuel bundles is required to improve the thermal-hydraulic properties of Pressurerized Water Reactor (PWR) spacer grids. To better understand the cross-flow characteristics in rod bundles for developing spacer grids, we used the rod-embedded fiber laser Doppler velocimetry (rod LDV) to measure the flow velocities inside the spacer grid flow channels. As the result of measurement, we found that the flow distribution inside the spacer grid depends on the local flow resistance of the grid straps and is clearly affected by the presence of a mixing vane. We also clarified the relationship between cross-flow velocity in the fuel bundle downstream of the spacer grid and the axial flow inside the spacer grid.     

激光诱导荧光技术对棒束通道内定位格架搅浑特性研究

基于激光诱导荧光（LIF）技术开展了5×5棒束通道内定位格架搅浑特性的可视化研究。常温常压下,通过示踪染色剂（RhB）浓度分布表征流体微团的搅浑行为,清晰展现染色剂溶液在定位格架作用下的搅浑扩散过程,获取格架下游流场的搅浑信息。采用自验证方法分析验证LIF技术测量的准确性,重构棒束通道内径向与轴向染色剂浓度分布,对比带定位格架与不带定位格架的实验结果,得到定位格架对其下游流场的影响范围及不同棒束子通道所受搅浑程度的差异,并以变异系数量化格架对流场搅浑性能的强弱。实验结果表明：定位格架能快速搅浑流动工质,其搅浑翼片分布形式的差异是造成不同子通道交叉搅浑强弱及各向异性的主要原因。本实验工况（Re＝10 478）下,格架对其下游流场的作用范围约为8倍当量直径(D_h),流动工质在格架下游5D_h附近所受搅浑最为剧烈。     

关于核电站燃料棒束临界热流密度研究中一些问题的讨论

本文分析了定位格架对临界热流密度(CHF)影响的机理,讨论了如何判断定位格架热工性能的好坏;对我院已做过的几种带不同定位格架的核电站燃料棒束的 CHF 实验结果作了对比分析,并与国外最新的 CHF 经验公式作了对比。     

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。     

反应堆棒束通道搅混翼数值研究

定位格架上的搅混翼是核反应堆燃料组件中的关键部件,其性能对棒束通道热工水力特性有重要的影响。以带单层定位格架的5×5棒束为研究对象,对搅混翼排布方式及末端形状对格架下游的流场和温度场的影响进行数值模拟研究。计算结果表明,改变搅混翼的排布方式,压降几乎不受影响,但格架下游流场和传热情况却因排布方式的改变而发生显著变化;将搅混翼末端形状改为弧形,压降较典型撕裂型搅混翼没有明显差异,但换热情况得到明显改善。      

The influence of walls and upper tie plate slots on the flooding mechanism in fuel elements with and without heat transfer between steam and water

The counter-current flow of steam and water was experimentally investigated for the upper part of a PWR fuel element. The actual geometrical shape of the nuclear equipment was simulated by various types of plates, in which bore holes and slots were arranged in different positions. The experiments were performed with and without an installed, unheated rod bundle below the plates. The water was injected at saturated and subcooled temperatures in order to observe the effects of heat transfer on counter-current flow.

With increasing steam velocity the flooding occurs initially in the tie-plate area. If the rod bundle is installed in the flow duct, a part of the downwards flowing water is transported upwards from the region of the upper grid spacer to the plate. Heat transfer between the phases can cause in the counter-current flow region an instable transition from downward to near complete upward directed liquid flow. In comparison to experiments with saturated water injection, flooding occurs at larger steam velocities. Different flooding correlations, which are known from the literature, were compared with the experimental data to appraise their applicability to counter-current flow in the core of PWRs.     

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。     

Single-phase CFD applicability for estimating fluid hot-spot locations in a 5 × 5 fuel rod bundle

High-thermal performance PWR spacer grids require both of low pressure loss and high critical heat flux (CHF) properties. Therefore, a numerical study using computational fluid dynamics (CFD) was carried out to estimate pressure loss in strap and mixing vane structures. Moreover, a CFD simulation under single-phase flow condition was conducted for one specific condition in a water departure from nucleate boiling (DNB) test to examine the applicability of the CFD model for predicting the CHF rod position. Energy flux around the rod surface in a water DNB test is the sum of the intrinsic energy flux from a rod and the extrinsic energy flux from other rods, and increments of the enthalpy and decrements of flow velocity near the rod surface are assumed to affect CHF performance. CFD makes it possible to model the complicated flow field consisting of a spacer grid and a rod bundle and evaluate the local velocity and enthalpy distribution around the rod surface, which are assumed to determine the initial conditions for the two-phase structure. The results of this study indicate that single-phase CFD can play a significant role in designing PWR spacer grids for improved CHF performance.     

The Effect of Mixing-vane Arrangements in a Subchannel Turbulent Flow

Large eddy simulation (LES) of developed turbulent flows in a rod bundle was carried out for four spacer designs. The mixing-vanes attached at the spacer were inclined at 30° or 20° they were arranged to promote the swirling or convective flow. These arrangements are possible elements to compose an actual rod bundle. Our LES technique with a consistent higher-order immersed boundary method and a one-equation dynamic sub-grid scale model contributed to an efficient treatment of the complex wall configurations of rods and spacers. The computational results reasonably reproduced experimental results for the drag coefficient and the decay rate of swirling flow. The profiles of the axial velocities and the turbulence intensities indicated reasonable trend for the turbulent flow in the rod bundle. The effect of mixing-vane arrangement on the lateral flows was successfully clarified: the cross flow took the longer way on the rod surface than the swirling flow and then was more significantly influenced by momentum diffusion at the no-slip wall. Therefore, the largely inclined mixing-vanes promoted the cross flow only in the neighborhood of the spacer; the swirling flow inside a subchannel could reach farther downstream than the cross flow.     

Evaluation of fluid flow in the lower core of a PWR with Code_Saturne

In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. The first spacer grid region is of particular interest, as fuel rod fretting has sometimes been observed at that level. Entry conditions depend on the geometry of the lower core plate and of the assembly nozzles. Distribution of flow in the downcomer and lower plenum is also a factor. A series of calculations are thus run with the incompressible Navier–Stokes solver, Code_Saturne, using a classical RANS turbulence model. The first calculations involve a global geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate and the fuel rod assemblies above it cannot be well represented within a practical mesh size, so that a head loss model is used. Different types of assemblies can be represented through different head loss coefficients. We make full use of Code_Saturne’s non-conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Steady-state or near steady-state results are obtained, which may be used as realistic entry conditions for full core calculations at assembly width resolution, and beyond those, mechanical strain calculations. We are especially interested in more detailed flow conditions and in the lower core area, so as in the future to quantify vibrational input. This requires a much higher resolution, which is limited to a scale of a few assemblies for practical reasons. At this scale, most of the features of the fuel rods, nozzles, and guide tubes are represented, though the geometry of the spacer grids is still much simplified, and details such as debris-trapping grids are ignored. Different meshes are used for different fuel types. For the moment, a constant velocity upstream of the lower core plate is used as an inlet condition. We have also built a small lower fuel rod assembly mock-up (1/5 scale 7 × 7 tube, 3 × 3 assemblies) with which we plan to obtain detailed flow information, and better qualify the use of our CFD codes with regards to this type of application.