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。     

空泡份额模型在矩形通道沸腾压降计算中的适用性评价

分析了各空泡份额模型随质量流量、系统压力及质量含气率的变化关系,并对各空泡份额计算模型在窄矩形通道内的应用进行了评价。研究结果表明:尽管空泡份额模型选取对重位压降及加速压降计算影响极大,但由于窄矩形通道内的饱和沸腾流动以环状流为主,重位压降及加速压降在两相总压降中的份额极小,因此在两相摩擦压降计算过程中,空泡份额模型的影响非常轻微。采用Zivi模型计算得到的沸腾摩擦压降与其他关系式计算值相对偏差在±5%范围内,因此建议采用Zivi模型计算窄矩形通道内空泡份额。     

窄矩形通道内搅混流和环状流相界面参数的计算方法

空泡份额和界面浓度是两相流动中重要的相界面参数,准确获取窄矩形通道内搅混流和环状流工况下空泡份额和界面浓度是构建和完善两流体模型的关键。本文针对横截面为65 mm×2 mm的矩形通道开展了气液两相流动特性可视化实验研究,气相折算速度j_g＝1～9 m/s,液相折算速度j_f＝0.1～1.5 m/s,流型包含搅混流和环状流。提出了基于高速摄像法获取搅混流和环状流下空泡份额和界面浓度的分析计算方法,利用该方法所得空泡份额与窄矩形通道内经验关系式计算值的相对偏差约在10%以内。此计算方法可为研究复杂流型下窄矩形通道内的相界面参数提供理论依据。     

垂直环形管中气—液两相流空泡份额的计算与实验研究

本文提出了环形通道中的空泡份额计算模型,并在实验数据基础上从模型得出了在较广参数范围内适用的空泡份额关系式,计算结果与实验数据符合较好;介绍了一种γ射线衰减法及“快速关闭阀门法”对同一环形通道中空泡份额的测量;分析了不同质量流速,不同容积含气率、不同流向时的流动情况及空泡份额值的变化规律。     

关于一维两流体模型中界面阻力计算的研究

一维两流体模型中,界面阻力是决定相间耦合程度的关键参数,其计算方法目前有漂移流模型法和阻力系数法。本研究利用子通道程序,基于圆管空气-水两相实验数据,对这2种计算方法进行了评估,结果表明一维两流体模型中漂移流模型法的预测能力要优于阻力系数法。同时评估了两相流动中分布效应对界面阻力计算的影响,结果表明在低空泡份额区分布效应影响较小,而高空泡份额区其影响明显。      

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。     

强迫对流沸腾模型中的分布参数C0的新计算公式

强制对流沸腾在能源和加工工业中占有重要地位，在核电站中尤其重要。对于不同类型的压水堆，在启动过程、正常或者事故工况下冷却剂在管道内可能发生沸腾。粗略估计，此时的空泡份额甚至能达到0．9。本研究的目的是研究在上述热工水力条件下的两相流动模型。根据法国原子能委员会（CEA）在格勒诺布尔的氟里昂12（R12）回路上得到的实验数据发现，在大空泡份额情况下，流动特性近似于起泡的乳状液。此时，无论空泡份额多大，液相都保持连续的状态。在此结论基础上，我们为漂移流模型中的分布参数C0建立了一个新的求解模型，用低过冷、低空泡份额情况下R12回路的数据对该模型进行了校验，结果表明与高空泡份额下R12以及高压矩形通道内流体的沸腾实验数据非常吻合。     

矩形通道中两相流变密度模型空泡份额计算

基于Bankoff的圆管内和无限长平板间两相流变密度模型空泡份额计算式的推导，结合流体在管道中的流场分布特征，建立了矩形通道中两相流流场分布规律方程，导出了变密度模型在矩形通道中空泡份额的计算式，并对3种通道计算的结果进行了对比分析。计算结果与原有Bankoff模型符合得很好。     

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%。     

矩形通道内泡状流-弹状流转变判定准则研究

以空气和去离子水为工质,对竖直矩形通道内两相流流型特性进行实验研究;矩形通道的横截面为1.41mm×40 mm和10 mm×40 mm,实验压力为常压,气、液相表观速度分别为0.015⁰.59 m/s和0.0250.59 m/s和0.025³.74 m/s。利用获得的实验数据及文献数据,对4种典型泡状流-弹状流转变判定准则进行评价,结果表明4种准则都存在一定局限性。从实验数据及文献数据可以看出,泡状流-弹状流转变临界空泡份额为通道窄边与宽边比(宽高比s/w)的函数。为此,以当量直径10 mm为界,分别提出临界空泡份额计算关系式,从而得到修正转变判定准则。与本文及文献中实验数据的比较,修正准则较4种典型准则精度和适用性有一定提高。     

An analytical model for the analysis of BWR/4 long-term cooling with either intact or broken jet pump seals

An analytical model for the long-term emergency core cooling (ECC) of a boiling water nuclear reactor (BWR) has been developed. This one dimensional drift-flux model, is an extension of a previous study by Lahey and Kamath [1]. It considers both subcooled and bulk boiling in the core, allows the drift-flux parameters, C₀ and V_gj, to be functions of void fraction (α), and can accommodate both broken and intact jet pump seals. The results of this analytical model compare well with data from simulated full scale BWR fuel rod bundles, and experiments in the PCE facility at RPI.It has been found that the unlikely failure of jet pump seals can have a detrimental effect on the long term cooling capabilities of a BWR/4.     

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.     

Drift-Flux Parameters for Upward Gas Flow in Stagnant Liquid

The drift-flux model is widely used for gas-liquid two phase flow analysis, because it is applicable to various flow patterns and a wide range of void fractions. The drift- flux parameters for upward gas flow in stagnant liquid, however, have not been well examined. In this study, the distribution parameter C_o and the drift velocity V_gj for stagnant liquid were derived from the void fraction correlation and boundary conditions of drift-flux parameters, and then compared with C_o and V_gj for high liquid velocities. Also using the two region model where a circular flow area was divided into an inner region of cocurrent up-flow and an outer annulus region of liquid down flow, C_o and V_gj for stagnant liquid and for high liquid velocity were compared. The results showed that C_o values for stagnant liquid were larger than values for high liquid velocity, while V_gj values were almost the same for both cases.     

The analysis of CCFL using drift-flux models

Counter-current flow limited (CCFL) situations can be analyzed using one-dimensional drift-flux techniques. Indeed such techniques can be used to prove that CCFL conditions occur at the point of kinematic ‘choking.’ This paper derives drift-flux parameters (C₀ and V_gj) which automatically predict CCFL conditions. These parameters are applied to the special case of simulated inlet flow blockage, and are shown to accurately predict internal flooding.     

Effects of Mass Transfer Resistances in Fluid and Permeable Walls on Hydrogen Permeation in Tube-Type Metal Window

The gas lift pump concept based on the bubbling of an inert gas into the primary reactor coolant to enhance natural circulation is currently considered in a number of PbBi-cooled reactor concepts. Thus, the analysis of available void fraction data and the development of two-phase heavy liquid metal/gas flow calculational models have become an important issue in the study of advanced nuclear reactor systems. In the absence of the detailed two-phase flow information needed to develop a flow regime map and the associated interfacial relations, drift-flux models have often been used in the thermal-hydraulic analysis of nuclear and other systems. Accordingly, we consider, in the current paper, the analysis of five sets of experimental data with different geometries, working fluids, flow rates and void fraction ranges, with a view to obtaining a best fit to the data in the form of a drift-flux model. The results of the analysis show that, for systems with flowing fluid, it is possible to represent the heavy liquid metal void fraction data in the form of a drift-flux correlation with a residual error of as low as 0.016, thus offering an improvement over existing void correlations.     

High-order fully implicit SIMPLE-based model for fully implicit simulation of upward two-phase flow

The drift-flux model has a practical importance in two-phase flow analysis.In this study,a finite volume solution is developed for a transient four-equation drift-flux model through the staggered mesh,leading to the development of a fully implicit discretization method.The main advantage of the fully implicit method is its unconditional stability.Newton's scheme is a popular method of choice for the solution of a nonlinear system of equations arising from fully implicit discretization of field equations.However,the lack of convergence robustness and the construction of Jacobian matrix have created several difficulties for the researchers.In this paper,a fully implicit model is developed based on the SIMPLE algorithm for two-phase flow simulations.The drawbacks of Newton's method are avoided in the developed model.Different limiter functions are considered,and the stabilized method is developed under steady and transient conditions.The results obtained by the numerical modeling are in good agreement with the experimental data.As expected,the results prove that the developed model is not restricted by any stability limit.     

Verification of the two-phase stratified-flow model in by separate effect tests

The code which is being developed by the Gesellschaft für Anlagen- und Rcaktorsicherheit (GRS) mbH is intended to cover, by means of a single code, the entire spectrum of loss-of-coolant and transient accidents in pressurized and boiling water reactors. The actual version Mod 1.1-Cycle A has a five-equation two-phase model based on the conservation laws for liquid mass, liquid energy, vapor energy and overall momentum. The relative velocity between liquid and vapor is determined by a full-range drift-flux model for two-phase flow in horizontal and vertical pipes. The verification of this drift-flux model is carried out by both large-scale experiments and single-effect tests. The single-effect test ECTHOR investigates stratified flow during the clearance of a water-filled loop seal by a forced air flow through the loop. ECTHOR is a French test for the consideration of two-phase flow regimes in pipes for the development of the codes. The experiments are dedicated to investigating typical two-phase flow during small break loss of coolant accidents (LOCA) in pressurized water reactors (PWR).As a measure, the remaining water level in the loop is determined as a function of the air flow rate. For the verification, a comparison between and computations, on the one hand, and experiments on the other hand is carried out. The results compare very well to each other. Test runs on different numerical grids show convergence to an asymptotic limit with increasing grid refinement.     

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.     

A problem in the COBRA-EN code related to the void fraction calculation

When a certain void fraction value is reached in the two-phase flow regime, a problem occurs in the COBRA-EN code. This problem was observed in the drift-flux model option and interrupts code execution. Two solutions are proposed to solve the problem.