摇摆对气-液两相流流型及空泡份额的影响

对摇摆状态下竖直上升管内气-液两相流的流型及空泡份额变化进行了实验和理论研究.研究发现,摇摆使两相流的流型发生改变,使泡状流提前转变为弹状流,使搅混流的区域加宽.实验还发现,在弹状流型区摇摆状态下两相流的空泡份额小于非摇摆状态下的空泡份额.通过对两相流滑速比的分析并应用分相流的动量方程,合理解释了产生这种结果的原因.     

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。      

气液两相流空泡份额计算

一、引言 气液两相流的空泡份额是两相流的基本参数之一,它对于反应堆和蒸汽发生装置的设计有重要作用。空泡份额α=Ag/A、体积含汽量β=Qg/Q、重量含汽量χ=Wg/W和滑速比S=u_g/u_1之间的关系式为:     

海洋条件下反应堆热工水力参数的子通道计算

针对海洋条件下反应堆的子通道热工水力分析,建立了海洋运动附加力模型和瞬态入口边界,将起伏、摇摆及复合运动的附加力关系式用于子通道模型的轴向和横向动量方程,并应用到COBRAⅢC程序将其改造为适应海洋条件的反应堆子通道分析程序。作为验证,计算了加热实验通道和"奥陆"堆在起伏运动情况下热通道的临界热流密度比(CHFR)、出口空泡份额和冷却剂流量,并与文献结果对比。还详细计算了"奥陆"堆在起伏、不同摇摆中心和复合运动情况下,热通道的CHFR和不同位置子通道出口的热工水力参数。研究表明:海洋条件下反应堆的子通道热工水力参数随运动呈周期性变化;起伏运动对子通道的压降影响较大,摇摆运动对子通道冷却剂的流量和温度影响较大。     

COSINE子通道均相流分析程序的研发

COSINE一体化软件包的子通道安全分析程序cosSubc基于子通道控制体三维网格模型,采用轴向及横向的热工水力控制方程,包括两流体和均相流两种求解算法。本文介绍了子通道均相流程序的物理模型和数值算法,并用cosSubc均相流程序和参考程序COBRA-TF分别对典型1 000MW核电厂稳态算例进行计算分析,结果表明:cosSubc均相流程序与COBRA-TF吻合较好,具备堆芯子通道的热工水力计算能力。     

基于双区域模型的钠冷快堆组件子通道分析程序的开发与验证

子通道分析程序是钠冷快堆堆芯热工水力设计和安全分析的重要工具。本文为计算和分析钠冷快堆组件在径向均匀与倾斜功率分布工况下的热工水力特性,利用双区域绕丝交混模型开发了一款适用于钠冷快堆组件分析的子通道程序SPLICA,并与FFM2A 19棒束实验数据与WARD 61棒束实验数据进行了对比验证。由于本文开发的子通道分析程序SPLICA使用了详细的绕丝交混模型,与经过二次开发后的COBRA程序的计算结果相比,对于FFM2A实验SPLICA程序计算得到的结果与实验结果符合得更好。这两个实验数据的验证结果证明了本文开发的子通道分析程序的准确性以及对高流量工况和低流量工况均具有良好的适用性。本程序能为钠冷快堆组件热工水力分析提供有效的设计和研究手段。     

垂直上升横掠水平管束的两相流空泡份额模型研究

与管内两相流空泡份额模型相比,垂直上升横掠水平管束的两相流空泡份额研究成果相对有限。利用垂直上升的气-水两相流横掠水平管束的实验数据,对现有的空泡份额计算模型进行对比分析,并对2种现有模型的拟合公式进行修正。采用其他实验结果对本文重新修正的空泡份额模型进行验证,结果表明:与原始模型相比,修正的空泡份额计算模型给出的空泡份额预测值更好。     

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。     

钠冷快堆自然循环组件子通道程序初步研发

本文为计算和分析钠冷快堆自然循环组件的热工水力特性,开发了钠冷快堆堆芯自然循环冷却组件子通道分析程序。基于61棒单组件模型,通过将本程序的结果与COBRA程序进行比较,验证了钠冷快堆堆芯自然循环冷却组件子通道分析程序对自然循环冷却组件的适用性。基于多盒组件模型,初步验证了本程序具备自然循环冷却组件的流量分配和盒间换热计算的功能。本程序能为池式快堆自然循环冷却组件提供有效的设计和分析工具。     

摇摆状态下矩形通道内弹状流压力模型

本文通过可视化方法对竖直与倾斜条件下矩形通道内弹状流单元的参数进行研究,尝试给出摇摆状态下矩形通道内弹状流压力模型。通过图像处理给出气弹段空泡份额以及两相速度的计算关系式,并验证漂移流模型在液弹段的适用性,给出弹状流单元的长度份额以及空泡份额的计算关系式。根据实验结果给出摇摆条件下矩形通道内弹状流压力组分的模型,并重点分析摩擦压降模型的适用程度。结果表明,弹状流压力模型可很好地预测摇摆条件下矩形通道内的压力。     

Numerical simulation of secondary flow in bubbly turbulent flow in sub-channel

Secondary flow in bubbly turbulent flow in sub-channel was simulated by using an algebraic turbulence stress model. The mass, momentum, turbulence energy and bubble diffusion equations were used as fundamental equation. The basis for these equations was the two-fluid model: the equation of liquid phase was picked up from the equation system theoretically derived for the gas-liquid two-fluid turbulent flow. The fundamental equation was transformed onto a generalized coordinate system fitted to the computational domain in sub-channel. It was discretized for the SIMPLE algorithm using the finite-volume method. The shape of sub-channel causes a distortion of the computational mesh, and orthogonal nature of the mesh is sometimes broken. An iterative method to satisfy a requirement for the contra-variant velocity was introduced to represent accurate symmetric boundary condition. Two-phase flow at a steady state was simulated for different magnitudes of secondary flow and void fraction. The secondary flow enhanced the momentum transport in sub-channel and accelerated the liquid phase in the rod gap. This effect was slightly mitigated when the void fraction increased. The acceleration can contribute to effective cooling in the rod gap. The numerical result implied a phenomenon of industrial interest. This suggested that experimental approach is necessary to validate the numerical model and to identify the phenomenon.     

COBRA-Ⅳ对8×8棒束计算的不确定性分析

综合现有不确定性分析方法的原理及特点，采用Wilks公式确定需要计算的最小次数，研究针对子通道程序不确定性的分析方法，并编写程序。运用该程序对8×8棒束的出口空泡份额实验的子通道计算进行了不确定性分析与研究，得到了每个子通道出口空泡份额的计算预测值，以及满足容忍限的不确定性上、下限。计算结果表明：边角子通道的计算不确定性较小，约为±5.5%；而水棒周围不规则形状的子通道的不确定性较大，约为±9%。堆芯热工水力问题中最关注的高空泡子通道的出口空泡份额，其不确定性为-5.5%～6%。     

A study of the performance of void fraction correlations used in the context of drift-flux two-phase flow models

Drift-flux models have traditionally been and are currently used in thermal-hydraulic analysis codes in the nuclear and other industries to analyze the behavior of systems during a wide variety of transient conditions. Their simplicity and closeness to experimental data, compared to two-fluid models, and their robustness, make them a cost-effective and efficient choice, although these models are generally limited to co-current flow. The drift-flux models are based on correlations to compute the void fraction distribution and slip in two-phase flow needed to obtain the relative velocity between the phases. Thus, the accuracy of the correlations has a decisive role in determining the correct transport of vapor along the system and, subsequently, in the prediction of the correct response of nuclear or industrial systems. This paper presents the results of an evaluation of the accuracy of a range of widely used void fraction correlations based on the Findlay–Zuber drift-flux model. The 13 correlations presented in this paper, a sub-set of all considered, can loosely be termed as ‘wide range void correlations’, since, as shown in this paper, they are those able to perform reasonably well for the wide range of experimental conditions used in the assessment. The size of the experimental database allowed a detailed statistically based comparison of the performance of all the correlations assessed. The void fraction data is taken from rod bundle, level swell and boil-off experiments performed within the last 10–15 years at 9 experimental facilities in France, Japan, Switzerland, the UK and the USA. The pressure and mass fluxes of the analyzed experiments range from 0.1 to 15 MPa and from 1 to 2000 kg m⁻² s⁻¹, respectively. Finally, the assessment of a widely used correlation against experimental transient void fraction data has been performed. The selected correlation is that of Chexal–Lellouche, currently used in the system codes RETRAN-3D and RELAP-5. The results show the performance of the correlation when used in the context of a system code and two different drift-flux model approaches, namely, an algebraic slip calculation and the calculation of the slip velocity based on the solution of a differential slip equation. The accuracy of the predictions shows that it is possible to use a drift-flux approach even for the analysis of rapid transients.     

Void fraction estimation within rod bundles based on three-fluid model and comparison with x-ray CT void data

An interfacial shear stress equation in the dispersed-annular two-phase flow regime has been developed, which is based on a three-fluid model consisting of a liquid film on a rod, vapor and entrained liquid associated with a vapor flow. It is an extension of J.G.M. Andersen's procedure that provides a two-fluid interfacial shear stress equation using the drift flux parameters C₀ and V_gj. This interfacial shear stress equation can take into account a phase and velocity distribution through an equivalence between the drift flux parameters and the interfacial shear stress.

Using the three-fluid subchannel analysis code TEMPO with the three-fluid interfacial shear stress model, the capability of a three-fluid calculation using the drift flux parameters C₀ and V_gj that reproduce a measured void fraction is demonstrated. A comparison was made with advanced X-ray computed tomography (CT) void fraction data within a 4×4 rod bundle in diabatic 1 MPa pressure conditions. The three-fluid velocity field was estimated to be in good agreement with the experimental result of a void fraction.     

Modeling of distribution parameter,void fraction covariance and relative velocity covariance for upward steam-water boiling flow in vertical rod bundle

In order to improve the prediction accuracy of one-dimensional interfacial force formulated by ‘Andersen’ approach, the distribution parameter in a drift–flux correlation, void fraction covariance, and relative velocity covariance has been modeled for dispersed boiling two-phase flow in a vertical rod bundle. The distribution parameter has been derived by a bubble-layer thickness model. The correlations of void fraction covariance and relative velocity covariance have been developed based on prototypic 8 × 8 rod bundle data. The correlation of void fraction covariance agrees with the bundle data with the mean absolute error, standard deviation, mean relative deviation, and mean absolute relative deviation being 0.00120, 0.0415, ?0.173%, and 1.80%, respectively. The correlation of relative velocity covariance agrees with the bundle data with the mean absolute error, standard deviation, mean relative deviation, and mean absolute relative deviation being ?0.00241, 0.0452, ?0.0316%, and 2.52%, respectively. In view of the great importance of void fraction covariance and relative velocity covariance on the one-dimensional interfacial drag force formulation, it is highly recommended to include the void fraction covariance and relative velocity covariance in the one-dimensional formulation of interfacial drag force used in nuclear thermal-hydraulic system analysis codes.     

Experimental study on void fraction in a simulated BWR fuel assembly (evaluation of cross-sectional averaged void fraction)

Void fraction measurement of a vertical (4 x 4) rod bundle has been conducted in a steam-water two phase flow, using an advanced X-ray CT scanner. A large amount of rod bundle data was obtained. It was found from the results that the cross-sectional averaged void fraction data for a rod bundle can be correlated by the Drift-Flux model and that the Zuber-Findlay correlation underestimates the data in a void fraction area of 80% or more. This is because the data range over which this correlation was developed, does not cover this experimental range. Therefore, a modified correlation was developed based on the authors' data.     

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.     

含弯曲棒的全长燃料棒束沸腾传热CFD计算分析

采用计算流体动力学（CFD）分析方法模拟了含一根弯曲燃料棒（简称“弯曲棒”）的5×5全长燃料棒束内的沸腾传热现象。基于欧拉两流体模型和改进的壁面沸腾模型进行计算,并基于压水堆子通道和棒束实验( PSBT )基准题中的试验数据对计算方法进行了验证,计算所得截面平均空泡份额与试验数据吻合良好,说明了现有计算方法的可靠性。基于计算结果考察了弯曲棒对棒束通道内流场、温度场、空泡份额等关键参数的影响。研究结果表明,弯曲棒的存在对截面横向流动、流体温度、空泡份额等均未产生显著影响,但弯曲棒表面温度增加,气泡也易发生聚集,增加了发生临界热流密度（CHF）的风险。      

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.