Void distribution and bubble motion in bubbly flows in a 4×4 rod bundle. Part I: Experiments

Lack of local void fraction data in a rod bundle makes it difficult to validate a numerical method for predicting gas–liquid two-phase flow in the bundle. Distributions of local void fraction and bubble velocity in each subchannel in a 4×4 rod bundle were, therefore, measured using a double-sensor conductivity probe. Liquid velocity in the subchannel was also measured using laser Doppler velocimetry (LDV) to obtain relative velocity between bubbles and the liquid phase. The size and pitch of rods were 10 and 12.5 mm, respectively. Air and water at atmospheric pressure and room temperature were used for the gas and liquid phases, respectively. The volume fluxes of gas and liquid phases ranged from 0.06 to 0.15 m/s and from 0.9 to 1.5 m/s, respectively. Experimental results showed that the distributions of void fraction in inner and side subchannels depend not only on lift force acting on bubbles but also on geometrical constraints on bubble dynamics, i.e. the effects of rod walls on bubble shape and rise velocity. The relative velocity between bubbles and the liquid phase in the subchannel forms a non-uniform distribution over the cross-section, and the relative velocity becomes smaller as bubbles approach the wall due to the wall effects.     

Loss-of-flow experiment in a 37-pin bundle LMFBR fuel assembly simulator

An experimental study was conducted on transient sodium boiling in an LMFBR fuel subassembly mockup under loss-of-flow conditions. In the test section, an electrically heated 37-pin bundle was centered in a hexagonal tube. The measured maximum IB wall superheat was 36°C, and the effects of heat flux, temperature rise rate, and system pressure were unclear. Boiling was initiated at the end of the heated section, the bubble expanded mainly to the upstream central subchannels and to the downstream unheated section according to the expansion of the saturated temperature region. When the voided zone covered the whole flow cross-section, the void pattern changed to the one-dimensional slug ejection-type and the inlet flow decreased rapidly. Dryout occurred after the inception of flow reversal in the wide region of the bundle.     

On the transport mechanisms in simulated heterogeneous rod bundle subchannels

Experimental results are presented on fully developed turbulent flow through simulated heterogeneous rod bundle subchannels. The emphasis of this study is on the universality of the cross-gap turbulence convection transport with respect to symmetric versus asymmetric subchannels. The flow passage was formed by a rod asymmetrically mounted in a trapezoidal duct. The Reynolds number based on the equivalent hydraulic diameter and bulk average axial velocity is 26 300. The measurements include mean axial velocities, r.m.s. values of the fluctuating velocity components and the energy density spectra. The results demonstrate the existence of an unusual region near the asymmetric rod-to-wall gap characterized by high levels of axial turbulence intensity with a remarkably different type of distribution compared with a normal boundary layer. It is also shown that the strength of the cross-gap transport is subchannel geometry dependent. The distributions of wall shear stress and turbulence kinetic energy indicate that mean convection by secondary flow is also an important transport mechanism that should be taken into account in the analysis of momentum/heat transfer in rod bundle subchannels.     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

基于丝网传感器的5×5棒束通道空泡分布测量研究

为研究压水反应堆燃料组件棒束通道内的两相分布规律,设计并制造了适用于棒束通道的丝网传感器模块,开展了5×5棒束通道内空气-水泡状流的空泡分布测量实验,分析了棒束通道内空泡份额的分布规律及气泡尺寸对空泡分布的影响。实验结果表明,发生横升力方向反转的小气泡在壁面附近聚集、大尺寸气泡则聚集在子通道中心;常温常压下发生横升力方向反转的临界气泡直径在4~6 mm之间,证明了横升力模型在棒束通道中的适用性。      

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.     

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。     

Results of the temperature noise measurements in subchannels of sodium cooled electrically heated 28 rod bundle behind a blockage

For LMFBR safety studies a 28 rod bundle has been built at Petten (cooperation of GfK and ECN), representing a 60-degrees section of an SNR-300 fuel element having a 70% flat type central blockage. The aims of the temperature noise measurements were to determine the subchannel coolant velocities behind the blockage to study the mixing of coolant in subchannels of different temperature from behind the blockage to the outlet and to study the temperature noise due to boiling in a subchannel. The temperature noise measurements were carried out in parallel to the other measurements (temperature distribution, etc.), using signals of fourteen subchannel thermocouples placed in five measuring planes behind the blockage. The single phase measurements were made with several heat fluxes (5 W/cm² to 120 W/cm²), inlet flows (0.25 to 3 m/s) and inlet temperatures (250°C to 600°C). Two phase flow is initiated and sustained either by a slow and continuous pressure reduction or by stepwise reduction of the main flow. The temperature noise signals were amplified and recorded in analog form. Later the signals were digitized and analysed by digital computers. Part of the signals was also processed by a hardware correlator. The experimental results of the temperature noise measurements will be shown for the different conditions of the loop. Measurements clearly show the following effects:

• - the recirculation flow pattern due to the vortex in the wake behind the blockage;

• - the dependence of r.m.s. value of the noise on the heat flux and the coolant flow;

• - the increase in noise and change in power spectra when going from single phase to the boiling condition.

     

定位绕丝结构对棒束通道热工水力特性影响数值分析

定位绕丝设计广泛应用于金属快堆堆芯设计及气冷快堆堆芯设计中,本文基于三维精细化绕丝定位棒束通道网格模型模拟分析了定位绕丝螺距、定位绕丝数量及定位绕丝形状对超临界二氧化碳在棒束通道中流动换热的影响。模拟结果表明定位绕丝螺距比定位绕丝数量及定位绕丝形状对温场流场的影响更大,定位绕丝螺距小于200 mm时,进出口压降大幅增加,表面换热系数增加,温度不均匀度大幅降低;随着定位绕丝数量增加,进出口压降线性增加,表面换热系数变化不大;圆形定位绕丝可以以较小截面积达到与方形定位绕丝相似的效果,梯形定位绕丝对流场影响不如矩形定位绕丝。     

CFD analysis of flow field in a triangular rod bundle

The flow field was investigated in subchannels of VVER-440 pressurized water cooled reactors’ fuel assemblies (triangular lattice, P/D = 1.35). Impacts of the mesh resolution and turbulence model were studied in order to obtain guidelines for CFD calculations of VVER-440 rod bundles. Results were compared to measurement data published by Trupp and Azad in 1975. The study pointed out that RANS method with BSL Reynolds stress model using a sufficient fine grid can provide an accurate prediction for the turbulence quantities in this lattice. Applying the experiences of the sensitivity study thermal hydraulic processes were investigated in VVER-440 rod bundle sections. Based on the examinations the spacer grids have important effects on the cross flows, axial velocity and outlet temperature distribution of subchannels therefore they have to be modeled satisfactorily in CFD calculations.     

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…     

Void distribution and bubble motion in bubbly flows in a 4×4 rod bundle. Part II: numerical simulation

Numerical simulations of bubbly flows in a four by four rod bundle are carried out using a multi-fluid model to examine effects of the numerical treatment of phase distribution and drag model. The transport equations of bubble number density and void fraction are used as the continuity equation of the gas phase. Two drag models are tested: one of them accounts for the bubble deformation (aspect ratio), whereas the other does not. The rod diameter, the rod pitch and the hydraulic diameter of the rod bundle are 10, 12.5 and 9.1 mm, respectively. The gas and liquid volume fluxes are J_G = 0.06 m/s and J_L = 0.9 and 1.5 m/s, respectively. The bubble diameter ranges from 1 to 5 mm. Comparisons between the numerical and measured data show that (1) the restriction on bubble lateral motion due to the presence of rods can be taken into account by using the transport equation of bubble number density, whereas that of the void fraction cannot deal with the restriction and causes large errors in the distribution of void fraction and (2) the reduction in the bubble-relative velocity near the wall is predictable by using the drag model accounting for the bubble deformation effect.     

CFD analyses in tight-lattice subchannels and seven-rod bundle geometries of a Super Fast Reactor

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.     

Turbulent momentum transport in rod bundles

An experimental investigation was performed with air to obtain detailed information on the velocity and turbulence distribution for parallel turbulent flows through subchannels of rod bundles. Experimental results were obtained for wall and corner subchannels of rod bundles of four parallel rods. The pitch-to-diameter ratios were varied between 1.07 and 1.4. The Reynolds numbers ranged from 6 × 10⁴ to 2 × 10⁵, depending on the rod bundle arrangement.On the basis of the data measured, the eddy viscosities in the directions normal and parallel to the wall were calculated. The experimental results of the velocity and wall shear stress distributions are compared with the predictions by the VELASCO code. There are considerable differences between computed and experimental results especially for low pitch-to-diameter ratios. The reasons for the discrepancies are discussed together with the results of attempts to adjust the VELASCO code against the experimental data.     

棒束通道内气液两相流流型的实验研究

常温、常压条件下,在7×7矩形截面棒束通道内进行了垂直向上气液两相流动实验,气液两相折算速度的变化范围分别为0.04~14 m/s和0.238~1.860 m/s。实验中用高速摄像仪对流型进行记录,观察到了泡状流、泡状-搅混流、搅混流和搅混-环状流4种流型,发现搅混流是主要流型,并对Hewitt流型图的界限进行了修正。分析实验数据发现,摩擦压降在泡状流和搅混流区域的变化是相反的。根据实验数据,参考前人的研究得到棒束通道中泡状流向搅混流转变的边界。     

Transient Boiling of Sodium in Seven-Pin Bundle under Transient Overpower Conditions

Transient sodium boiling experiments were conducted in an electrically heated 7-pin bundle under transient overpower conditions. In each run the heater power was gradually raised at almost constant rate under forced convection.

The observed coolant voiding was initially limited to the central subchannel on account of an appreciable time lag in temperature rise occurring between the central and peripheral subchannels. This would appear to call for calculations with two-dimensional voiding model.

The bulk pressure rises registered upon initial vaporization were markedly lower than the vapor pressure corresponding to the incipient-boiling (IB) wall superheat. The pressure pulse generated upon vapor bubble collapse correlated reasonably well with the re-entrant liquid velocity, but the measured value was very much smaller than predicted theoretically from sodium hammer analysis.     

Local Sodium Boiling behind Local Flow Blockage in Simulated LMFBR Fuel Subassembly

This paper deals with local sodium boiling in the downstream of a six-subchannel blockage in an electrically heated LMFBR fuel subassembly mock-up.

The first series of experiments were conducted to measure temperature distributions in the downstream of the blockage under non-boiling conditions. The measured temperature rise due to the blockage agreed fairly well with the calculation by the LOCK code.

The second series of experiments were performed to investigate local boiling phenomena. In the local boiling region, no flow instability was observed since the sub-channels near the wrapper wall were still filled with sub-cooled liquid. In the nearly bulk boiling region, however, considerable upstream voiding occurred and then the inlet flow decreased, leading to final dryout.

The boiling caused a considerable increase in acoustic noise intensity. The root-mean-square (RMS) noise level of approximately 20 mbar obtained in the present local boiling experiments with sodium was much higher than that (approximately 0.5 mbar) in the ordinary nucleate boiling experiments with water. The peak observed in the hertz ranges was due to the repetition of bubble formation and collapse. In the kilohertz ranges, however, resonance peaks were superposed on a smooth curve with a broad peak at approximately 7 kHz.

The frequency (2.9 and 20.2 sec^?1) of bubble formation decreased with the increase of the bubble size at its point of maximum development. The product of the bubble frequency and the equivalent diameter was found to be constant.     

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。      

Flow structure and bubble characteristics of steam–water two-phase flow in a large-diameter pipe

The flow structure and bubble characteristics of steam–water two-phase upward flow were observed in a vertical pipe 155 mm in inner diameter. Experiments were conducted under volumetric flux conditions of J_G<0.25 m s⁻¹ and J_L<0.6 m s⁻¹, and three different inlet boundary conditions to investigate the developing state of the flow. The radial distributions of flow structure, such as void fraction, bubble chord length and gas velocity, were obtained by horizontally traversing optical dual void probes through the pipe. The spectra of bubble chord length and gas velocity were also obtained to study the characteristics of bubbles in detail. Overall, an empirical database of the multi-dimensional flow structure of two-phase flow in a large-diameter pipe was obtained. The void profiles converged to a so-called core-shaped distribution and the flow reached a quasi-developed state within a relatively short height-to-diameter aspect ratio of about H/D=4 compared to a small-diameter pipe flow. The PDF histogram profiles of bubble chord length and gas velocity could be approximated fairly well by a model function using a gamma distribution and log–normal distribution, respectively. Finally, the correlation of Sauter mean bubble diameter was derived as a function of local void fraction, pressure, surface tension and density. With this correlation, cross sectional averaged bubble diameter was predicted with high accuracy compared to the existing constitutive equation mainly being used in best-estimate codes.     

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.