棒束通道内定位格架的两相流动局部阻力实验研究

在常温、常压条件下,对竖直3×3棒束通道内定位格架的单相及两相局部阻力特性进行了实验研究。单相流动实验时,水雷诺数的变化范围为290~18 007;两相实验时,气相、液相表观速度变化范围分别为0.013~3.763m/s和0.076~1.792m/s。利用单相实验数据得到的定位格架局部阻力系数计算关系式,用两相实验数据对均相流模型中8种不同的两相等效黏度计算方法进行了评价。Rel9 000时,Dukler模型的预测效果最好;Rel≥9 000时,McAdams计算方法预测效果最好;基于所有数据,Dukler模型的计算值与实验值吻合最好,平均相对误差为29.03%。考虑了质量含气率、两相雷诺数及气液相密度的影响,对Rel9 000时的实验数据进行了拟合,得到的经验关系式的计算值与实验值符合较好。     

Numerical simulation of multi-dimensional two-phase flow based on flux vector splitting

This paper describes a new approach to the numerical simulation of transient, multidimensional two-phase flow. The development is based on a fully hyperbolic two fluid model of two-phase flow using separated conservation equations for the two phases. Features of the new model include the existence of real eigenvalues, and a complete set of independent eigenvectors which can be expressed algebraically in terms of the major dependent flow parameters. This facilitates the application of numerical techniques specifically developed for high speed single-phase gas flows which combine signal propagation along characteristic lines with the conservation property with respect to mass, momentum and energy. Advantages of the new model for the numerical simulation of 1- and 2-dimensional two-phase flow are discussed.     

An improved mixing-length model for annular two-phase flow with liquid entrainment

A differential model for adiabatic, fully developed, annular two-phase flow with liquid entrainment has been developed. The model is based on a modified form of the single-phase mixing length function to account for turbulence intensity attenuation due to the presence of entrained liquid droplets in the gas core. Simultaneous predictions of pressure drop and average film thickness for given flow rates show good agreement with a wide range of experimental results for various fluids at several pressures in a range of tube diameters for both horizontal and vertical (upwards and downwards) flows.     

Steady-state performances and scaling analyses for an open flashing-driven natural circulation system

Simulating investigations are carried out to study the steady-state performances, the pressure resistance distributions and the scaling methods of the single-phase and flashing-induced two-phase flow in the open natural circulation system, which is designed for the passive containment cooling system. The results show that the steady-state mass flow rate changes with the heat transfer regularly both in the single-phase and flashing-induced two-phase flow under a certain inlet subcooling. From the sensitivity analysis, it can be found that the riser height only has impact on the single-phase flow but has little influence on the flashing-driven two-phase flow. Both increasing the diameters of the riser and downcomer can enhance the flow and heat transfer in sing-phase and two-phase flow when keeping the structure of the heat exchanger unchanged, but the influence degree for each flow type is different. The flow resistance distributions of the loops under different flow modes have been studied to provide the foundation for improving the heat transfer capacity by choosing the structural parameters reasonably. The pressure resistance distribution of the steady single-phase flow only relates to the geometrical, but the pressure resistance distribution of the two-phase flow relates both to the pipe diameters and to the external conditions. The acceleration pressure resistance in the riser section is the main resistance under the higher-quality two-phase conditions. Therefore, the influence of the riser diameter on the flashing-induced two-phase flow is far greater than that of downcomer diameter and the most effective method to improve the two-phase flow and heat transfer is to increase the diameter of the riser. Finally, the scaling analysis is performed for the penetration and economy considerations after selecting the optimal dimensions. The scaling of the cold and hot sections is considered separately to insure the driven force of the system unchanged, and different scaling criterions are given for the single-phase and flashing-induced two-phase flow according to the analyses of the pressure resistance distributions. The results show that the scaling criterion of the two-phase flow can deal with the scaling problem accurately both in the single-phase and two-phase flow. However, the scaling criterion of the single-phase flow only can solve the single-phase scaling problem, but it will overestimate the operating results in the scaling model.     

Assessment of the CATHARE 1D pump model

The development of an advanced model to determine the dynamic pump performance under two-phase flow conditions is presented. This model is included in CATHARE 2, version V1.3. It is based on the two-fluid six-equation CATHARE model which describes the mechanical and thermal non-equilibria.In a previous review (P. Van den Hove and G. Geffraye, The CATHARE code— one-dimension pump model, Fifth Int. Topical Meet. on Nuclear Reactor Thermal Hydraulics (NURETH-5), Salt Lake City, USA, September, 1992), various calculations were presented concerning Eva single-phase and two-phase steam–water test results in the first three quadrants. Here, the range of assessment of the first quadrant is enlarged with Eva air–water tests and Bethsy pump steam–water tests. Both pumps are mixed flow pumps, the Bethsy one being radial at the impeller outlet.Some improvements suggested in the above cited paper are tested against all single-phase liquid, single-phase vapor, two-phase steam–water, and two-phase air–water data in the first quadrant. They concern a new deviation model and head losses model, and the model of mechanical interaction between phases.     

Scaling of two-phase flow transients using reduced pressure system and simulant fluid

Scaling criteria for a natural circulation loop under single-phase and two-phase flow conditions are derived. Based on these criteria, practical applications for designing a scaled-down model are considered. Particular emphasis is placed on scaling a test model at reduced pressure levels compared to a prototype and on fluid-to-fluid scaling. The large number of similarity groups which are to be matched between model and prototype makes the design of a scale model a challenging task. The present study demonstrates a new approach to this classical problem using two-phase flow scaling parameters. It indicates that a real time scaling is not a practical solution and a scaled-down model should have an accelerated (shortened) time scale. An important result is the proposed new scaling methodology for simulating pressure transients. It is obtained by considering the changes of the fluid property groups which appear within the two-phase similarity parameters and the single-phase to two-phase flow transition parameters.Sample calculations are performed for modeling two-phase flow transients of a high-pressure water system by a low-pressure water system or a Freon system. It is shown that modeling is possible for both cases for simulating pressure transients. However, simulation of phase change transitions is not possible by a reduced pressure water system without distortion in either power or time.     

两相条件下主泵特性试验研究

针对自主研发百万千瓦级核电主泵项目,本研究在专设台架上对经过模化缩比后的模型试验泵开展特性研究,首先开展泵单相液体条件下正常工况、水轮机工况、耗能工况、卡轴工况、飞逸工况的流动特性试验,之后逐一针对两相空泡份额为0.1、0.2、0.3、0.4、0.5、0.6、0.7、0.8、0.9、1.0条件下的泵在上述具体运行工况开展研究,最终利用比例定律,对多个转速下的实验数据进行整理,进而掌握了其扬程在不同空泡、不同流量下的变化规律,在单相液体基础上的泵扬程比例定律曲线随着空泡份额的增大会存在不断“漂移”和再回归的过程,当空泡份额达到0.7左右时,“漂移”程度最大,当流体变为单相气体时,实现再回归,上述实验结果为后续核主泵设计的深入研究及一回路安全分析提供了数据支撑。      

棒束通道单相和两相交混模型评估

本研究利用子通道程序,基于已有的实验数据,对棒束通道的单相和两相交混模型进行了评估。单相交混主要考虑横流和湍流交混,横流由守恒方程决定并在流量分布中占主导作用,湍流交混取决于交混系数,对湍流交混研究发现Sadatomi模型预测结果与实验结果吻合较好。两相交混由横流、湍流交混和空泡漂移共同作用,通过已有模型预测结果与实验数据对比分析,推荐两相交混中空泡漂移采用Hotta模型、湍流交混系数采用Sadatomi模型和两相乘子采用Beus模型,这是一个预测结果较为保守的组合模型,有利于反应堆安全的保守性评估。      

Nuclear magnetic resonance: A new tool for the validation of multiphase multidimensional CFD codes

Turbulent transport models and data in bubbly flows are briefly reviewed since they play an important role in the modeling of boiling flows in forced convection. Shortcomings of earlier measurements of the eddy diffusivity by NMR (Lemonnier and Leblond, 2007b) are analyzed and a new procedure is presented which is now consistent with the procedure of Gatenby and Gore (1994) developed for single-phase turbulent flow characterization. The newly estimated eddy diffusivity agrees now with that previously obtained by Serizawa et al. (1975b) with a thermal method and that of the model of Sato and Sadatomi (1981). This procedure also provides the liquid velocity fluctuation RMS and the Lagrangian correlation time of velocity fluctuations. In addition, the same NMR technique provides also the area-averaged liquid velocity and void fraction. Bubbly flow data up to transition to slug flow are provided which also agree with existing drift-flux models (Ishii and Hibiki, 2006). It is finally discussed how the NMR method can be extended to local measurements and may provide a fully non-intrusive diagnostic in two-phase flows and which is not limited to bubbly flow.     

Validation of the sub-channel code F-COBRA-TF: Part I. Recalculation of single-phase and two-phase pressure loss measurements

The present paper deals with recalculations of single-phase and two-phase pressure loss measurements with the advanced two-phase, three-field sub-channel code F-COBRA-TF. Thereby, experimental data of both the OECD/NRC BFBT benchmark and in-house tests in AREVA NP's KATHY loop are used. The main goal of this paper is not to focus on a special new model or correlation but to give an overview how a complete pressure loss calculation for practical purposes can be carried out being based on a simplified and straightforward method to estimate sub-channel spacer pressure loss coefficients on the one hand and an advanced sub-channel code on the other hand.The pressure loss coefficients are calculated analytically and calibrated at available measurements of total single-phase bundle pressure loss. Thus, they are not adapted to any two-phase measurement and also do not depend on the sub-channel code they are used in.The results of the recalculations of the measurements especially demonstrate the capability of a three-field code to predict both single-phase and two-phase pressure losses with high accuracy, whereas the code is not based on conventional pressure loss correlations using two-phase multipliers but rather on interfacial friction correlations for each flow regime. Thereby, the F-COBRA-TF standard models - which are usually applied for all sorts of calculations (pressure loss, void distribution, lateral mixing, critical heat flux, etc.) - were used. It was not necessary to do special code tuning with respect to certain experiments.     

The discrete equation method (DEM) for fully compressible, two-phase flows in ducts of spatially varying cross-section

For the simulation of light water nuclear reactor coolant flows, general two-phase models (valid for all volume fractions) have been generally used which, while allowing for velocity disequilibrium, normally force pressure equilibrium between the phases (see, for example, the numerous models of this type described in (Städtke, 2006)). These equations are not hyperbolic, their physical wave dynamics are incorrect, and their solution algorithms rely on dubious truncation error induced artificial viscosity to render them numerically well-posed over a portion of the computational spectrum. The inherent problems of the traditional approach to multiphase modeling, which begins with an averaged system of (ill-posed) partial differential equations (PDEs) which are then discretized to form a numerical scheme, are avoided by employing a new homogenization method known as the discrete equation method (DEM) (Abgrall and Saurel, 2003). This method results in well-posed hyperbolic systems, this property being important for transient flows. This also allows a clear treatment of non-conservative terms (terms involving interfacial variables and volume fraction gradients) permitting the solution of interface problems without conservation errors, this feature being important for the direct numerical simulation of two-phase flows.Unlike conventional methods, the averaged system of PDEs for the mixture are not used, and the DEM method directly obtains a well-posed discrete equation system from the single-phase conservation laws, producing a numerical scheme which accurately computes fluxes for arbitrary number of phases. The method effectively uses a sequence of single-phase Riemann problem solutions. Phase interactions are accounted for by Riemann solvers at each interface. Non-conservative derivative product terms are correctly approximated. Some of the closure relations missing from the traditional approach are automatically obtained. Lastly, the continuous equation system resulting from the discrete equations can be identified by taking the continuous limit with weak-wave assumptions. In this work, this approach is tested by constructing a DEM model for the flow of two compressible phases in 1-D ducts of spatially varying cross-section with explicit time integration. An analytical equation of state is included for both water vapor and liquid phases, and a realistic interphase mass transfer model is developed based on interphase heat transfer. A robust compliment of boundary conditions is developed and discussed. Though originally conceived as a first step toward implict time integration of the DEM method (to relieve time step size restrictions due to stiffness and to achieve tighter coupling of equations) in multidimensions, this model offers some unique capabilities for incorporation into next generation light water reactor safety analysis codes. We demonstrate, on a converging-diverging two-phase nozzle, that this well-posed, 2-pressure, 2-velocity DEM model can be integrated to a realistic and meaningful steady state with both phases treated as compressible.     

Effect of depressurization on reactor vessel inventory in the absence of ECCS injection

A series of tests were performed to evaluate inventory depletion as a reactor vessel undergoes depressurization in the absence of any emergency core coolant system injection (ECCS). These tests were carried out in a scaled representation of a reactor vessel which was initially filled with saturated water up to the elevation of the hot legs. Depressurization valves installed on take-off lines from the hot legs were opened and level swell ensued in the reactor vessel initiating a two-phase blowdown. This was followed by subsequent single-phase discharge transient which in some cases led to core uncovery. A combined model encompassing the two-phase and single-phase discharge portions of the transient is proposed. The inventory-versus-pressure traces obtained from the model compare well with the experimental results. These traces are discussed as bounding trajectories for a large class of small break loss of coolant accident (LOCA) transients which otherwise must be considered individually.     

CFD methodology and validation for single-phase flow in PWR fuel assemblies

This paper presents the CFD modeling methodology and validation for steady-state, normal operation in a PWR fuel assembly. This work is part of a program that is developing a CFD methodology for modeling and predicting single-phase and two-phase flow conditions downstream of structural grids that have mixing devices. The purpose of the mixing devices (mixing vanes in this case) is to increase turbulence and improve heat transfer characteristics of the fuel assembly. The detailed CFD modeling methodology for single-phase flow conditions in PWR fuel assemblies was developed using the STAR-CD CFD code. This methodology includes the details of the computational mesh, the turbulence model used, and the boundary conditions applied to the model. The methodology was developed by benchmarking CFD results versus small-scale experiments. The experiments use PIV to measure the lateral flow field downstream of the grid, and thermal testing to determine the heat transfer characteristics of the rods downstream of the grid. The CFD results and experimental data presented in the paper provide validation of the single-phase flow modeling methodology. Two-phase flow CFD models are being developed to investigate two-phase conditions in PWR fuel assemblies, and these can be presented at a future CFD Workshop.     

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.     

一维自然循环比例分析的结果与讨论

根据一维自然循环比例分析理论模型推导的试验装置与实际电站热工水力特性的相似准则,对整体性能试验装置主要参数的确定方法进行了深入讨论。结果表明:采用小尺度、等压力、同工质的实验装置模拟实际系统自然循环现象更为准确实际,单相和两相自然循环比例准则可同时满足,不存在复杂比例变化带来的失真,不利因素是试验成本偏高。同工质非等物性(不等压)模拟能够降低试验成本,但比例参数不能满足从单相自然循环到两相自然循环的平滑过渡。如保持功率连续,其速度比和特征时间比会有所差异。     

Development of a pump performance model for an integral effect test facility

Modeling methods for the reactor coolant pumps of the existing integral facilities are reviewed from the viewpoint of scaling, single- and two-phase characteristics. A series of separate effect tests was performed for the pumps of the ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) in order to obtain a complete set of single-phase homologous curves in all the quadrant operating regions. Besides the friction loss tests for the sheared shaft and the locked rotor conditions were conducted to extend the homologous curves to the limiting cases. A method which can take into account the two-phase degradation effects of the ATLAS pumps was suggested as a first approximation based on similarity principles. The present data and model can completely replace the pump input model of the MARS 3.1 code both for single- and two-phase conditions.     

Experimental study on the natural circulation of a passive residual heat removal system for an integral reactor following a safety related event

Natural circulation characteristics of an integral type reactor during the operation of a passive residual heat removal system (PRHRS) following a safety related event has been experimentally investigated by using the VISTA facility. A PRHRS actuation trip signal is generated by a high power trip signal following a steam flow increasing event. The experimental results show that the single-phase coolant flows steadily in the primary loop by a natural convection process and that it effectively removes the decay heat from the core through a steam generator during the PRHRS operation. The heat transfers through the PRHRS heat exchanger and the emergency cooldown tank (ECT) are sufficient enough to enable a two-phase natural circulation of the coolant in the PRHRS loop.     

Pressure drops for steam and water flow in heated tubes

Measurements of pressure drop for a steam and water flow inside several heated tubes were obtained at two different laboratories (Chalk River Laboratories and École Polytechnique). The test sections used to carry out the experiments were constructed with Inconel-600 tubes, and were cooled internally with an upward flow of water. A reduction in frictional pressure gradient with an increase in heat flux was observed for single-phase flows at the same cross-sectional-average conditions. After boiling was initiated, the two-phase friction multiplier increased with increasing heat flux in the bubbly-flow region, but decreased in the annular-flow region. As the heated surface approached dryout, the two-phase friction multiplier exhibited a maximum value and decreased with increasing thermodynamic quality, until dryout occurred. The decreasing trend was attributed to a reduction in the liquid-film thickness and the termination of liquid entrainment. Following the occurrence of critical heat flux, a significant drop in frictional pressure gradient was noted, caused by the change in the near-wall sublayer to a low-viscosity vapor blanket.     

A review of heat exchanger tube bundle vibrations in two-phase cross-flow

Flow-induced vibration is an important concern to the designers of heat exchangers subjected to high flows of gases or liquids. Two-phase cross-flow occurs in industrial heat exchangers, such as nuclear steam generators, condensers, and boilers, etc. Under certain flow regimes and fluid velocities, the fluid forces result in tube vibration and damage due to fretting and fatigue. Prediction of these forces requires an understanding of the flow regimes found in heat exchanger tube bundles. Excessive vibrations under normal operating conditions can lead to tube failure.

Relatively little information exists on two-phase vibration. This is not surprising as single-phase flow induced vibration; a simpler topic is not yet fully understood. Vibration in two-phase is much more complex because it depends upon two-phase flow regime, i.e. characteristics of two-phase mixture and involves an important consideration, which is the void fraction. The effect of characteristics of two-phase mixture on flow-induced vibration is still largely unknown. Two-phase flow experiments are much more expensive and difficult to carry out as they usually require pressurized loops with the ability to produce two-phase mixtures. Although convenient from an experimental point of view, air–water mixture if used as a simulation fluid, is quite different from high-pressure steam–water. A reasonable compromise between experimental convenience and simulation of steam–water two-phase flow is desired.

This paper reviews known models and experimental research on two-phase cross-flow induced vibration in tube bundles. Despite the considerable differences in the models, there is some agreement in the general conclusions. The effect of tube bundle geometry, random turbulence excitations, hydrodynamic mass and damping ratio on tube response has also been reviewed. Fluid–structure interaction, void fraction modeling/measurements and finally Tubular Exchanger Manufacturers Association (TEMA) considerations have also been highlighted.     

Two-phase flow characterization by nuclear magnetic resonance

The results presented in this paper demonstrate the ability of the PFGSE-NMR to obtain a complete characterization of two-phase flows. Different methods are proposed to characterize air-water flows in different regimes: stationary two-phase flows and flows in transient condition. Finally a modified PFGSE is proposed to analyze the fluctuations in stationary air-water bubbly flows.