Steam-water Two-phase Flow Research Articles

Pressure Drop and Void Fraction in Steam-Water Two-Phase Flow at High Pressure

For a steam generator (SG) in a commercialized sodium-cooled fast breeder reactor (FBR), flow instability in the water side is one of the most important items needing research. As the first step of this research, thermal-hydraulic experiments using water as the test fluid were performed under high pressure conditions at the Japan Atomic Energy Agency (JAEA) by using a circular tube. Void fraction, pressure drop, and heat transfer coefficient data were obtained under 15, 17, and 18 MPa. This paper discusses the steam-water pressure drop and void fraction. Using the obtained data, we evaluated existing correlations for void fraction and two-phase flow multipliers under high pressure. As a result, the drift flux model implemented in the TRAC-BF1 code was confirmed to suitably predict the void fraction well under the present high pressure conditions. For the two-phase flow multiplier, the Chisholm correlation and the homogeneous model were confirmed to be the best under the present high-pressure conditions.

Interfacial area transport of vertical upward steam–water two-phase flow in an annular channel at elevated pressures

The interfacial area transport of vertical, upward, steam-water two-phase flows in a vertical annular channel has been investigated. The inner and outer diameters of the annular channel were 19.1 and 38.1mm, respectively. The test section had a 2845mm heated section followed by a 1632mm unheated section. Fifty seven experimental conditions were selected, which cover bubbly, cap-slug, and churn-turbulent flows. Each one of flow conditions was obtained by achieving different inlet sub-cooling temperatures, liquid velocities, wall heat flux or system pressures. The local flow parameters, such as void fraction, interfacial area concentration, and bubble interface velocity, were measured at different radial positions for the five axial locations. The radial and axial evolutions of local flow structure were interpreted based on presence of wall superheat, wall nucleation, bulk condensation and evaporation, bubble sizes, coalescence and break-up mechanisms. The measured data can be used for both the assessment of the bubble coalescence/breakup models and the development of closure models for computational fluid dynamics.

High pressure steam–water two-phase flow measurements by flow division and separation method

This paper proposes a method for high pressure steam–water two-phase flow measurements by flow division and separation. The flowrate of a steam–water mixture is metered by diverting a small fraction (about 2%) of total mass flow to a division loop, where the diverted mixture is separated into steam and water flows by means of a small separator and they are metered with single-phase flow meters respectively. The metered values are then converted to total flowrates according to the extraction ratios. The first equation of the extraction ratios was derived from the relation of resistance between the main loop and the division loop, and the second equation of the extraction ratios was developed from the distribution characteristics of the division loop. The extraction ratios were then determined by solving these two equations. Experiments were carried out in an once-through boiler for steam flooding. The inside diameter of pipe was 50mm, and the experimental flowrate ranged from 2000kg/h to 8000kg/h, the steam mass quality varied from 60% to 81%, and pressure ranged from 7.6MPa to 16MPa. Experiments show that the flow division and separation method is a reliable technique. The flowrate measurement error is less than ±2.5%, and the steam quality measurement error is less than ±3.5%.

The Numerical Simulation on Steam-Water Two-Phase Flow Field of Car Exhaust Pipe

The flow field of car exhaust pipe using ethanol gasoline has been researched, the continuity equation of a mixed two-phase model have been established by means of the mathematical model of two-phase flow field, and the simulation for the velocity distribution of droplets radial slipping in exhaust gas have been achieved. By using the divided grid and setting these four boundary conditions as conditions of the entrance, flow conditions at the end of the export, overflow conditions of the export, solid-wall boundary conditions, the iterative steps of calculation process and the relationship among the residuals of various parameters have been obtained. State of exhaust flow have been calculated using algebraic slip model (ASM) ,obtaining the distribution rule of the internal flow field in exhaust pipe for the disconnection of steam-water ,and the characteristics relationship of separation for two-phase flow field of steam-water. The characteristic relationship among flow, split ratio and separation efficiency have been numerically simulated. The existence of the best range of split ratio for a particular exhaust pipe has been confirmed.

The Numerical Simulation on Steam-Water Two-Phase Flow Field of Car Exhaust Pipe

The flow field of car exhaust pipe using ethanol gasoline has been researched, the continuity equation of a mixed two-phase model have been established by means of the mathematical model of two-phase flow field, and the simulation for the velocity distribution of droplets radial slipping in exhaust gas have been achieved. By using the divided grid and setting these four boundary conditions as conditions of the entrance, flow conditions at the end of the export, overflow conditions of the export, solid-wall boundary conditions, the iterative steps of calculation process and the relationship among the residuals of various parameters have been obtained. State of exhaust flow have been calculated using algebraic slip model (ASM) ,obtaining the distribution rule of the internal flow field in exhaust pipe for the disconnection of steam-water ,and the characteristics relationship of separation for two-phase flow field of steam-water. The characteristic relationship among flow, split ratio and separation efficiency have been numerically simulated. The existence of the best range of split ratio for a particular exhaust pipe has been confirmed.

Investigation of Temperature Fluctuations Caused by Steam-Water Two-Phase Flow in Pressurizer Spray Piping

Investigation of Temperature Fluctuations Caused by Steam-Water Two-Phase Flow in Pressurizer Spray Piping

TWO-PHASE CFD: THE VARIOUS APPROACHES AND THEIR APPLICABILITY TO EACH FLOW REGIME

Two-phase Computational Fluid Dynamics (CFD) or CMFD (Computational Multi-Fluid Dynamics) is being applied to some nuclear reactor thermalhydraulic investigations. The NURESIM and NURISP projects of the European 5th and 6th Framework programs made some pioneering efforts to start some applications to a few selected reactor issues such as the Critical Heat Flux or the Pressurized Thermal shock. The extreme variety of flow configurations in steam-water two-phase flow makes the modeling issue very complex. Myriads of model options are available in the vast domain of CMFD, with various treatments of turbulence and of interfaces; the methods for choosing model options depending on the application have to be clarified. The purpose of this paper is to list and classify model options, to discuss some conditions and limits of applicability of the various options, and to identify the modeling needs for closure of the system of equations. The proposed classification of the modeling approaches is based on the space and time filtering or averaging of basic equations and on the number of phases and fields that are distinguished. The various flow processes that must be modeled by closure relations are identified for each type of approach. The applicability of each of these methods to each flow regime is defined and the present degree of maturity of the models is evaluated. Considering only Eulerian approaches used in an open medium, it is shown that five main approaches are considered: the RANS approach, the pseudo-DNS approach, and three types of space-filtered methods. Some of these methods can be applied only to a reduced number of flow regimes. Attention is drawn on the need to specify the model options, to guaranty the consistency between options, and to ensure completeness of closure terms. © 2011 by Begell House, Inc.

ICONE19-43043 Investigation of Wall Temperature Fluctuations by Visualization Tests for Steam-Water Two-Phase Flow in the Pressurizer Spray Piping

ICONE19-43043 Investigation of Wall Temperature Fluctuations by Visualization Tests for Steam-Water Two-Phase Flow in the Pressurizer Spray Piping

Condensation Behavior in a Microchannel Heat Exchanger

A small and high performance heat exchanger for small size energy equipments such as fuel cells and CO2 heat pumps is required in these days. In author's previous studies, the heat exchanger consisted of microchannels stacked in layers has been developed. It has resistance to pressure of larger than 15 MPa since it is manufactured by diffusion bond technique. Thus this device can be applied for high flow rate and pressure fluctuation conditions as boiling and condensation. The objectives of the present study are to clarify the heat transfer performance of the prototype heat exchanger and to investigate the thermal hydraulic behavior in the microchannel for design optimization of the device. As the results, it is clarified that the present device attained high heat transfer as 7 kW at the steam condensation, despite its weight of only 230 g. Furthermore, steam condensation behavior in a glass capillary tube, as a simulated microchannel, in a cooling water pool was observed with various inlet pressure and temperature of surrounding water. Relation between steam-water two-phase flow structure and the overall heat transfer coefficient is discussed.

F109 EXPERIMENTAL INVESTIGATION ON THE FRICTIONAL PRESSURE DROP OF HIGH PRESSURE STEAM-WATER TWO-PHASE FLOWS IN AN OPTIMIZED INTERNALLY-RIBBED TUBE(Multiphase Flow-1)

Experimental investigation on the frictional pressure drop of high pressure steam-water two-phase flows in an optimized six-head internally-ribbed tube has been carried out in Xi'an Jiaotong University, China. Two adiabatic test sections were utilized, with one installed horizontally and another installed vertically. The frictional pressure drops of an adiabatic horizontal smooth tube were also measured to provide a data base for comparison. According to experimental data, the two-phase friction multiplier was calculated and its variations with the pressure, the mass flux and the steam quality were obtained. By comparison, it was found that the two-phase friction multiplier of the horizontal internally-ribbed tube had an obvious peak value, which was much larger than that of the horizontal smooth tube, in the region with high steam qualities. It was also found that the two-phase friction multiplier of the horizontal internally-ribbed tube was always greater than that of the vertically-upward internally-ribbed tube.

An investigation on heat transfer characteristics of different pressure steam-water in vertical upward tube

Within the range of pressure from 9 to 30 MPa, mass velocity from 600 to 1200 kg/(m 2 s), and heat flux at inner wall from 200 to 600 kW/m 2, experiments have been performed to investigate the heat transfer characteristics of steam-water two-phase flow in vertical upward tube. The outer diameter of the tube is 32 mm, and the wall thickness is 3 mm. Based on results, it was found that Dryout is the main mechanism of the heat transfer deterioration in the sub-critical pressure region. Near the critical pressure, when the heat transfer deterioration occurs, the steam quality of water is lower than that in the sub-critical pressure region, so that DNB is the main mechanism in this pressure region. At supercritical pressure, the heat transfer performance in circular channel is improved and enhanced. Heat transfer deterioration phenomenon is observed when the fluid bulk temperature approaches to the pseudo-critical value. Nusselt correlation of the forced-convection heat transfer in supercritical pressure region has been provided, which can be used to predict heat transfer coefficient of the vertical upward flow in tube.

Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology.In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

Geometry influence on safety valves sizing in two-phase flow

In the case of two-phase vapour–liquid flow, especially for low vapour quality (<10%), pressure safety valves (PSV) design becomes very difficult due to complex thermal-hydraulic phenomena occurring between the two phases. Currently, there are some calculation methods, based on different simplifying hypotheses, trying to predict the two-phase flow rate through a PSV knowing inlet fluid conditions (pressure, quality or temperature) and the outlet pressure. However, none of them is acknowledged as being reliable for any situation and, therefore, there is still a lacking of standards for PSV design under two-phase conditions. The PSV size is one of the most important parameters used for choosing between the two main prediction models, homogeneous equilibrium model (HEM) and homogeneous non-equilibrium model (HNE). This paper shows the results of an experimental research carried out with steam-water two-phase flow through two PSVs having the same orifice diameter (10 mm), but different discharge coefficients and inlet geometry. The experimental results are compared with the predictions obtained using a calculation method based on a homogeneous model with non-equilibrium hypotheses and another method proposed in API Recommended Practice 520, developed with equilibrium hypotheses. The results show that the PSV geometry and the discharge conditions are important factors for choosing the more suitable model for the sizing of a little PSV.

Experiment and analysis on performance of steam-driven jet injector for district-heating system

The objective of this work is to analyze the performance of steam-driven jet injector for its design and application in the district-heating system. The paper experimentally studied the performance of a steam-driven jet injector with central steam nozzle configuration. Three injectors with throat diameters of 16 and 20 mm were tested. The main inlet parameters tested—steam pressure, water pressure and water temperature—were in the range of 0.20–0.60 MPa, 0.14–0.49 MPa and 18–68 °C, respectively. Experimental results show that the pressures were almost constant for the steam–water two-phase flow in the variable section channel of the mixing chamber tested, whose values were mainly affected by inlet steam pressure, inlet water pressure and temperature, and a shockwave appeared at the outlet of the mixing chamber. The lifting-pressure coefficient and jet coefficient measured were 1.27–2.1 and 7.63–43.65, respectively. A simplified mathematical model has been developed to predict the performance of the injector based on conservation of mass, momentum and energy as well as the present experimental data. The predictions agreed with the experimental data within ±15%.

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.

Convective boiling heat transfer and two-phase flow characteristics inside a small horizontal helically coiled tubing once-through steam generator

The pressure drop and boiling heat transfer characteristics of steam-water two-phase flow were studied in a small horizontal helically coiled tubing once-through steam generator. The generator was constructed of a 9-mm ID 1Cr18Ni9Ti stainless steel tube with 292-mm coil diameter and 30-mm pitch. Experiments were performed in a range of steam qualities up to 0.95, system pressure 0.5–3.5 MPa, mass flux 236–943 kg/m 2s and heat flux 0–900 kW/m 2. A new two-phase frictional pressure drop correlation was obtained from the experimental data using Chisholm’s B-coefficient method. The boiling heat transfer was found to be dependent on both of mass flux and heat flux. This implies that both the nucleation mechanism and the convection mechanism have the same importance to forced convective boiling heat transfer in a small horizontal helically coiled tube over the full range of steam qualities (pre-critical heat flux qualities of 0.1–0.9), which is different from the situations in larger helically coiled tube where the convection mechanism dominates at qualities typically >0.1. Traditional single parameter Lockhart–Martinelli type correlations failed to satisfactorily correlate present experimental data, and in this paper a new flow boiling heat transfer correlation was proposed to better correlate the experimental data.

C216 並列ヘリカルコイル管内気液二相流の流れの可視化

In the steam generator of the fast breeder prototype reactor Monju, heat exchange tubes of helical coil type are utilized. Steam-water two-phase flows in the helical coil tube have unique characteristic due to effects of centrifugal acceleration. In this paper, we visualized gas-liquid two-phase flows in each the helical coil tubes with two curvature radii to investigate the effect of centrifugal acceleration, and also the slug flow in parallel helical coil tubes with same pressure condition at outlet part are visualized.

ICONE11-36033 EXPERIMENTAL STUDY OF THERMAL-HYDRAULIC BEHAVIOR OF COOLANT OUT OF NEUTRON REFLECTOR FOR APWR DURING REFLOODING OF LARGE BREAK LOCA

An experimental study was carried out to investigate the steam-water two-phase flow of coolant in the Neutron Reflector (N/R) during the reflooding of large break Loss OF Coolant Accident (LOCA). The experimental apparatus simulated the geometry of one hole of a typical N/R cooling-hole (including the corresponding thermal capacity), the lower and the upper plenum. The steam and water were used as the practical working fluids. The water flow rate at the inlet was kept constant at a fixed level through the injection pump. The boundary conditions including the LOCA phenomena, which were estimated by an analytical code, were supplied through the measurement of the system pressure and the entrance temperature of the water. The system pressure, temperature, and flow rate of steam and water were measured simultaneously to investigate the thermal-hydraulic behaviors at the end of the cooling-hole. Further, the behavior of liquid lumps and droplets flow during the simulated reflooding was observed at the outlet of the test section. The closed inspection revealed that the flow condition out of the test section were changing during the simulated reflooding and was classified into three stages. At the initial stage, the liquid lumps and the droplets flowed continuously like a fountain. At the medium stage, misty flow appeared, i. e. dispersed small droplets and/or liquid lumps flowed in steam flow. At the final stage, the liquid lumps flowed periodically with a little amount of steam flow. In this study, the effects of the reflooding velocity, the initial metal temperature, the reflooding water temperature and the upper plenum pressure were clarified. For the lower reflooding velocity, the ratio of steam flow rate increased. For the higher initial metal temperature, the integration of steam and water flow rate increased. The reflooding water temperature and the pressure of the upper plenum at the experimental conditions had little effects on the thermal hydraulic behavior of the coolant at the outlet of the test section.

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.

STEAM-WATER STRATIFIED FLOW IN T-JUNCTIONS – EXPERIMENTS AND MODELLING

Experimental data on dividing steam-water two-phase stratified flow in T-junctions with horizontal inlet and both horizontally and vertically downward branches are presented in this paper. The measurements included inlet liquid level and run and branch fully developed void fraction. These measurements were used to describe the effects of the inlet flow conditions and junction geometry (relative branch diameter and orientation) on phase redistribution. The data covered the range of inlet steam and water superficial velocities from 1.5 to 5.0 m/s and 0.05 to 0.09 m/s respectively. Under these conditions, which have not been previously examined, the extent of phase redistribution was greatly affected by the inlet superficial gas velocity (uGS), and the junction geometry. The effect of inlet superficial water velocity (uLS) on the redistribution characteristics was less significant for the range of data tested. A model for the phase redistribution of stratified flow in T-junctions is developed. It consists of...