Film boiling from spheres in single- and two-phase flow

International Nuclear Information System (INIS)

Liu, C.; Theofanous, T.G.; Yuen, W.W.

1992-01-01

Experimental data on film boiling heat transfer from single, inductively heated, spheres in single- and two-phase flow (saturated water and steam, respectively) are presented. In the single-phase-flow experiments water velocities ranged from 0.1 to 2.0 m/s; in the two-phase-flow experiments superficial water and steam velocities covered 0.1 to 0.6 m/s and 4 to 10 m/s, respectively. All experiments were run at atmospheric pressure and with sphere temperatures from 900C down to quenching. Limited interpretations of the single-phase- flow data are possible, but the two-phase-flow data are new and unique

Frictional pressure drop of steam-water two-phase flow in helical coils with small helix diameter of HTR-10

International Nuclear Information System (INIS)

Bi Qincheng; Chen Tingkuan; Luo Yushan; Zheng Jianxue

1996-01-01

Experiments of steam-water two-phase flow frictional pressure drop through five vertically and horizontally positioned helical coils were carried out in the high pressure steam water test loop of Xi'an Jiaotong University. Two kinds of tube with inner diameters of 10 mm and 12 mm were used to form the coils. The helix diameter was 115 mm with coil pitch 22.5 mm. The experimental conditions were: pressure p = 4-14 MPa, mass velocity G = 400-2000 kg/(m 2 ·s), and inner wall heat flux q = 0-750 kW/m 2 . Theoretical analysis with a semi-empirical correlation was made to predict the two-phase flow fictional pressure drop through these kinds of helical coils

Effects of phase transformation of steam-water relative permeabilities

Energy Technology Data Exchange (ETDEWEB)

Verma, A.K.

1986-03-01

A combined theoretical and experimental study of steam-water relative permeabilities (RPs) was carried out. First, an experimental study of two-phase concurrent flow of steam and water was conducted and a set of RP curves was obtained. These curves were compared with semi-empirical and experimental results obtained by other investigators for two-phase, two-component flow (oil/gas; gas/water; gas/oil). It was found that while the wetting phase RPs were in good agreement, RPs for the steam phase were considerably higher than the non-wetting phase RPs in two-component systems. This enhancement of steam RP is attributed to phase transformation effects at the pore level in flow channels. The effects of phase transformation were studied theoretically. This study indicates that there are two separate mechanisms by which phase transformation affects RP curves: (1) Phase transformation is converging-diverging flow channels can cause an enhancement of steam phase RP. In a channel dominated by steam a fraction of the flowing steam condenses upstream from the constriction, depositing its latent heat of condensation. This heat is conducted through the solid grains around the pore throat, and evaporation takes place downstream from it. Therefore, for a given bulk flow quality; a smaller fraction of steam actually flows through the throat segments. This pore-level effect manifests itself as relative permeability enhancement on a macroscopic level; and (2) phase transformation along the interface of a stagnant phase and the phase flowing around it controls the irreducible phase saturation. Therefore, the irreducible phase saturation in steam-water flow will depend, among other factors, on the boundary conditions of the flow.

Experimental research on density wave oscillation of steam-water two-phase flow in parallel inclined internally ribbed pipes

International Nuclear Information System (INIS)

Gao Feng; Chen Tingkuan; Luo Yushan; Yin Fei; Liu Weimin

2005-01-01

At p=3-10 MPa, G=300-600 kg/(m 2 ·s), Δt sub =30-90 degree C, and q=0-190 kW/m 2 , the experiments on steam-water two-phase flow instabilities have been performed. The test sections are parallel inclined internally ribbed pipes with an outer diameter of φ38.1 mm, a wall thinkness of 7.5 mm, a obliquity of 19.5 and a length more than 15 m length. Based on the experimental results, the effects of pressure, mass velocity, inlet subcooling and asymmetrical heat flux on steam-water two-phase flow density wave oscillation were analyzed. The experimental results showed that the flow system were more stable as pressure increased. As an increase in mass velocity, critical heat flux increased but critical steam quality decreased. Inlet subcooling had a monotone effect on density wave oscillation, when inlet subcooling decreased, critical heat flux decreased. Under a certain working condition, critical heat flux on asymmetrically heating parallel pipes is higher than that on symmetrically heating parallel pipes, that means the system with symmetrically heating parallel pips was more stable. (authors)

R 12 two-phase flow in throttle capillaries in critical flow conditions

International Nuclear Information System (INIS)

Petry, G.

1983-01-01

In this dissertation, the state of knowledge on two phase flow, its use and measurement processes are given from an extensive search of the literature. In the experimental part of the work, a continuously working experimental circuit was built up, by which single component two phase flow can be examined in critical flow conditions. Using the maintenance equations, a system of equations was produced, by which the content of steam flow, the content of steam volume and the slip between the phases at the end corssection of the capillary can be determined. The transfer of the experimental results into the Baker diagram shows that the experimental values lie in the region of mist, bubble and foam flow. (orig.) [de

Numerical analysis of gas-liquid two-phase flow in secondary side of steam generator

Energy Technology Data Exchange (ETDEWEB)

Murase, Michio; Nakamura, Akira; Yagi, Yoshinori [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2002-09-01

The steam generator (SG) in a pressurized water reactor (PWR) is an important two-phase flow component as the boundary between the primary loop and the secondary loop. In this study, we performed gas-liquid two-phase flow analyses for SG reliability tests conduced by Nuclear Power Engineering Corporation (NUPEC) using the two-fluid model of a thermal-hydraulic computer code PHOENICS. In order to calculate the location of the boiling initiation accurately, detailed inputs were required for the friction coefficients affecting the velocity distribution and the heat transfer distribution. However, the velocity and heat transfer distributions did not greatly affect the void fractions in the upper region of the heat transfer tubes. The calculated void fractions agreed with the measured values within 4% as the local average and within 2% as an average in a cross-section, except the region of low void fractions. (author)

Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

Energy Technology Data Exchange (ETDEWEB)

Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Shaver, Dillon [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Vegendla, Prasad [Argonne National Lab. (ANL), Argonne, IL (United States); Tentner, Adrian [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-30

The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.

Two-Phase Instability Characteristics of Printed Circuit Steam Generator for the Low Pressure Condition

International Nuclear Information System (INIS)

Kang, Han-Ok; Han, Hun Sik; Kim, Young-In; Kim, Keung Koo

2015-01-01

Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall manufacturing cost for the components. A PCHE(Printed Circuit Heat Exchanger) is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. PCSG(Printed Circuit Steam Generator) is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. For the introduction of new steam generator, design requirement for the two-phase flow instability should be considered. This paper describes two-phase flow instability characteristics of PCSG for the low pressure condition. PCSG is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. Interconnecting flow path was developed to mitigate the two-phase flow instability in the cold side. The flow characteristics of two-phase flow instability at the PCSG is examined experimentally in this study

Models for assessing the relative phase velocity in a two-phase flow. Status report

International Nuclear Information System (INIS)

Schaffrath, A.; Ringel, H.

2000-06-01

The knowledge of slip or drift flux in two phase flow is necessary for several technical processes (e.g. two phase pressure losses, heat and mass transfer in steam generators and condensers, dwell period in chemical reactors, moderation effectiveness of two phase coolant in BWR). In the following the most important models for two phase flow with different phase velocities (e.g. slip or drift models, analogy between pressure loss and steam quality, ε - ε models and models for the calculation of void distribution in reposing fluids) are classified, described and worked up for a further comparison with own experimental data. (orig.)

Entrainment and deposition studies in two-phase cross flow: comparison between air-water and steam-water in a square horizontal duct. Technical report (final)

International Nuclear Information System (INIS)

Berryman, R.J.; Ralph, J.C.; Wade, C.D.

1981-03-01

Air-water simulation studies of two phase steam water flow relevant to the upper plenum of a PWR during reflood situations have recently been undertaken at Harwell for the US Nuclear Regulatory Commission. In order to give confidence that the simulation fluids were capable of modelling the important features of the actual system, a relatively basic comparison experiment has been carried out. Water entrainment and deposition tests have been carried out on a pair of 2.5 cm diameter vertical rods mounted in a cross flow of steam or air in a 10.2 cm x 10.2 cm tunnel. The air and steam systems exhibited similar characteristics to one another. A 'critical' film flowrate was identified for the rods which, once reached, either by injection through the sinters or by entrainment from the main two phase stream, was not exceeded with further water addition. The 'critical' film flowrate decreased with increase of cross flow velocity and was lower for air than steam at the same velocity. The results from the air and steam tests were found to be reasonably well correlated on the basis of the cross flow momentum flux of the air or steam

Experiments of steady state head and torque of centrifugal pumps in two-phase flow

International Nuclear Information System (INIS)

Minato, Akihiko; Tominaga, Kenji.

1988-01-01

Circulation pump behavior has large effect on coolant discharge flow rate in case of reactor pipe break. Experiment of two-phase pump performance was conducted as a joint study of Japanese BWR user utilities and makers. Two-phase head and torque of three centrifugal pumps in high temperature and high pressure (around 6 MPa) steam/water were measured. Head was decreased from single-phase characteristics when gas was mixed in liquid flow in condition with normal flow and normal rotation directions. When flow rate was large enough, two-phase head was about the same as single-phase one in reversal flow conditions. Two-phase head was smoothly increased as flowing steam volumetic concentration increased when flow rate was small and flow direction was reversal. Changes of torque with gas concentration were correspondent to those of head. This suggested that changes of interaction between flow and impellers due to phase slip effected on torque which caused head differences between single- and two-phase flows. Dependence of dimensionless head and torque of three test pumps on steam concentration were almost the same as each other. (author)

Two-phase flow in a diverging nozzle

International Nuclear Information System (INIS)

Wadle, M.

1986-05-01

Stationary two-phase flow experiments were performed with steam-water and air-water mixtures in a well-instrumented horizontal diverging nozzle. The test section consisted of a constant diameter tube, the friction-section, followed by an expansion, the diffusor, which has a tanh-contour and finally another constant diameter tube. The diameter ratio sigma=D1/D2 is 16/80. For the steam-water experiments the flow parameters were: 0 2 and for air-water mixtures (0 2 ). The initial conditions were varied to achieve subcritical and critical mass flow rates. A new model for the pressure recovery in an abrupt expansion is presented. It is based on the superficial velocity concept and agrees well with the steam-water and the water-air experimental data as well as with the experiments of other authors. The experiments were also calculated with the two-phase code DUESE. The Drift-Flux models in this code as well as the constitutive correlations and their empirical constants could be tested. It is shown, that a 1D Drift-Flux code can handle the highly transient flow in the diffusor if the proper drift model is used. In a 1D simulation it is only necessary that the computational flow area is expanded to its full width within an axial length which is equivalent to the real contour. (orig./GL) [de

Two-phase flow characteristics analysis code: MINCS

International Nuclear Information System (INIS)

Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

1992-03-01

Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

Studies on turbulence structure and liquid film behavior in annular two-phase flow flowing in a throat section

International Nuclear Information System (INIS)

Yoshida, Kenji; Miyabe, Masaya; Matsumoto, Tadayoshi; Kataoka, Isao; Ohmori, Shuichi; Mori, Michitsugu

2004-01-01

Experimental studies on turbulence structure and liquid film behavior in annular two-phase flow were carried out concerned with the steam injector systems for a next-generation nuclear reactor. In the steam injector, steam/water annular two-phase flow is formed at the mixing nozzle. To make an appropriate design for high-performance steam injector system, it is very important to accumulate the fundamental data of thermo-hydro dynamic characteristics of annular flow in the steam injector. Especially, the turbulence modification in multi-phase flow due to the phase interaction is one of the most important phenomena and has attracted research attention. In this study, the liquid film behavior and the resultant turbulence modification due to the phase interaction were investigated. The behavior of the interfacial waves on liquid film flow such as the ripple or disturbance waves were observed to make clear the interfacial velocity and the special structure of the interfacial waves by using the high-speed video camera and the digital camera. The measurements for gas-phase velocity profiles and turbulent intensity in annular flow passing through the throat section were precisely performed to investigate quantitatively the turbulent modification in annular flow by using the constant temperature hot-wire anemometer. The measurements for liquid film thickness by the electrode needle method were also carried out. (author)

State of the art: two-phase flow calibration techniques

International Nuclear Information System (INIS)

Stanley, M.L.

1977-01-01

The nuclear community faces a particularly difficult problem relating to the calibration of instrumentation in a two-phase flow steam/water environment. The rationale of the approach to water reactor safety questions in the United States demands that accurate measurements of mass flows in a decompressing two-phase flow be made. An accurate measurement dictates an accurate calibration. This paper addresses three questions relating to the state of the art in two-phase calibration: (1) What do we mean by calibration. (2) What is done now. (3) What should be done

Pressure drop, steam content and turbulent cross exchange in water/steam flows

International Nuclear Information System (INIS)

Teichel, H.

1978-01-01

For describing the behaviour of two-phase flows of water and steam with the help of calculating patterns, a number of empirical correlations are required. - In this article, correlations for the friction pressure drop in water/steam flows are compared, as well as for the steam mass and the volumetric steam content with each other and with the test results on simple geometries. As the mutual effect between cooling chanels plays an important part at the longitudinal flow through bar bundles, the appertaining equations are evaluated, in addition. (orig.) 891 HP [de

Two-phase flow boiling pressure drop in small channels

International Nuclear Information System (INIS)

Sardeshpande, Madhavi V.; Shastri, Parikshit; Ranade, Vivek V.

2016-01-01

Highlights: • Study of typical 19 mm steam generator tube has been undertaken in detail. • Study of two phase flow boiling pressure drop, flow instability and identification of flow regimes using pressure fluctuations is the main focus of present work. • Effect of heat and mass flux on pressure drop and void fraction was studied. • Flow regimes identified from pressure fluctuations data using FFT plots. • Homogeneous model predicted pressure drop well in agreement. - Abstract: Two-phase flow boiling in small channels finds a variety of applications in power and process industries. Heat transfer, boiling flow regimes, flow instabilities, pressure drop and dry out are some of the key issues related to two-phase flow boiling in channels. In this work, the focus is on pressure drop in two-phase flow boiling in tubes of 19 mm diameter. These tubes are typically used in steam generators. Relatively limited experimental database is available on 19 mm ID tube. Therefore, in the present work, the experimental set-up is designed for studying flow boiling in 19 mm ID tube in such a way that any of the different flow regimes occurring in a steam generator tube (from pre-heating of sub-cooled water to dry-out) can be investigated by varying inlet conditions. The reported results cover a reasonable range of heat and mass flux conditions such as 9–27 kW/m 2 and 2.9–5.9 kg/m 2 s respectively. In this paper, various existing correlations are assessed against experimental data for the pressure drop in a single, vertical channel during flow boiling of water at near-atmospheric pressure. A special feature of these experiments is that time-dependent pressures are measured at four locations along the channel. The steady-state pressure drop is estimated and the identification of boiling flow regimes is done with transient characteristics using time series analysis. Experimental data and corresponding results are compared with the reported correlations. The results will be

Flow-induced vibration of steam generator helical tubes subjected to external liquid cross flow and internal two-phase flow

International Nuclear Information System (INIS)

Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Hho Jung Kim

2005-01-01

Full text of publication follows: This paper addresses the potential flow-induced vibration problems in a helically-coiled tube steam generator of integral-type nuclear reactor, of which the tubes are subjected to liquid cross flow externally and multi-phase flow externally. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted using a general purpose computational fluid dynamics code employing the finite volume element modeling. To get the natural frequency and corresponding mode shape of the helical type tubes with various conditions, a finite element analysis code is used. Based on the results of both helical coiled tube steam generator thermal-hydraulic and coiled tube modal analyses, turbulence-induced vibration and fluid-elastic instability analyses are performed. And then the potential for damages on the tubes due to either turbulence-induced vibration or fluid-elastic instability is assessed. In the assessment, special emphases are put on the detailed investigation for the effects of support conditions, coil diameter, and helix pitch on the modal, vibration amplitude and instability characteristics of tubes, from which a technical information and basis needed for designers and regulatory reviewers can be derived. (authors)

Numerical simulation of two phase flows in heat exchangers

International Nuclear Information System (INIS)

Grandotto Biettoli, M.

2006-04-01

The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)

Modelling aspects of two phase flow

International Nuclear Information System (INIS)

Mayinger, F.

1977-01-01

In two phase flow scaling is much more limited to very narrowly defined physical phenomena than in single phase fluids. For complex and combined phenomena it can be achieved not by using dimensionless numbers alone but in addition a detailed mathematical description of the physical problem - usually in the form of a computer program - must be available. An important role plays the scaling of the thermodynamic data of the modelling fluid. From a literature survey and from own scaling experiments the conclusion can be drawn that Freon is a quite suitable modelling fluid for scaling steam-water mixtures. However, whithout a theoretical description of the phenomena nondimensional numbers for scaling two phase flow must be handled very carefully. (orig.) [de

CFD Simulations of Pb-Bi Two-Phase Flow

International Nuclear Information System (INIS)

Dostal, Vaclav; Zelezny, Vaclav; Zacha, Pavel

2008-01-01

In a Pb-Bi cooled direct contact steam generation fast reactor water is injected directly above the core, the produced steam is separated at the top and is send to the turbine. Neither the direct contact phenomenon nor the two-phase flow simulations in CFD have been thoroughly described yet. A first attempt in simulating such two-phase flow in 2D using the CFD code Fluent is presented in this paper. The volume of fluid explicit model was used. Other important simulation parameters were: pressure velocity relation PISO, discretization scheme body force weighted for pressure, second order upwind for momentum and CISCAM for void fraction. Boundary conditions were mass flow inlet (Pb-Bi 0 kg/s and steam 0.07 kg/s) and pressure outlet. The effect of mesh size (0.5 mm and 0.2 mm cells) was investigated as well as the effect of the turbulent model. It was found that using a fine mesh is very important in order to achieve larger bubbles and the turbulent model (k-ε realizable) is necessary to properly model the slug flow. The fine mesh and unsteady conditions resulted in computationally intense problem. This may pose difficulties in 3D simulations of the real experiments. (authors)

Two-phase flow instability and propagation of disturbances

International Nuclear Information System (INIS)

Yadigaroglu, G.

1984-01-01

Various mechanisms of static and dynamic macroinstabilities, appearing in two-phase flows, have been considered. Types of instabilities, conditioned by the form of hydraulic characteristics of the channel and density waves are analyzed in detail. Problems of instabilities in nuclear reactor circuits, in particular problems of instabilities, conditioned by water and steam mixing and vapour condensation, and problems of steam generator operation instability are discussed

OTSGI--a program analysing two-phase flow instabilities in helical tubes of once-through steam generator

International Nuclear Information System (INIS)

Shi Shaoping; Zhou Fangde; Wang Maohua

1998-01-01

The author has studied the two-phases flow instabilities of the helical tubes of once-through steam generator. Using liner-frequency-domain analytical method, the authors have derived out a mathematical model and designed the program. In this model, the authors also have considered the thermal dynamic characteristics of the tube's wall. The program is used to calculate the threshold of the stability and the influences of some factors, such as entrance throttling coefficient, system pressure, entrance supercooling degree, et al. The outcomes are compared with other studies

Visualization of large waves in churn and annular two-phase flow

International Nuclear Information System (INIS)

Dasgupta, Arnab; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.; Kshirasagar, S.; Reddy, B.R.; Walker, S.P.

2015-01-01

The study of churn and annular two-phase flow regimes is important for boiling systems like nuclear reactors, U-tube steam generators etc. In this paper, visualization studies on air-water churn and annular two-phase flow regimes are reported. Though there are differences between air-water and boiling steam water systems, the major flow-pattern characteristics are similar (if not same).The specific object of study is the large waves which exist in both churn and annular regimes. These waves are responsible for majority of the momentum and mass dispersion across the phases. The differentiating characteristics of these waves in the chum and annular flow regimes are reported. The visualization also leads to a more quantitative representation of the transition from churn to annular flow. A new interpretation of the criterion for onset of entrainment is also evolved from the studies. (author)

Structure of the gas-liquid annular two-phase flow in a nozzle section

International Nuclear Information System (INIS)

Yoshida, Kenji; Kataoka, Isao; Ohmori, Syuichi; Mori, Michitsugu

2006-01-01

Experimental studies on the flow behavior of gas-liquid annular two-phase flow passing through a nozzle section were carried out. This study is concerned with the central steam jet injector for a next generation nuclear reactor. In the central steam jet injector, steam/water annular two-phase flow is formed at the mixing nozzle. To make an appropriate design and to establish the high-performance steam injector system, it is very important to accumulate the fundamental data of the thermo-hydro dynamic characteristics of annular flow passing through a nozzle section. On the other hand, the transient behavior of multiphase flow, in which the interactions between two-phases occur, is one of the most interesting scientific issues and has attracted research attention. In this study, the transient gas-phase turbulence modification in annular flow due to the gas-liquid phase interaction is experimentally investigated. The annular flow passing through a throat section is under the transient state due to the changing cross sectional area of the channel and resultantly the superficial velocities of both phases are changed compared with a fully developed flow in a straight pipe. The measurements for the gas-phase turbulence were precisely performed by using a constant temperature hot-wire anemometer, and made clear the turbulence structure such as velocity profiles, fluctuation velocity profiles. The behavior of the interfacial waves in the liquid film flow such as the ripple or disturbance waves was also observed. The measurements for the liquid film thickness by the electrode needle method were also performed to measure the base film thickness, mean film thickness, maximum film thickness and wave height of the ripple or the disturbance waves. (author)

Theoretical investigation of flow regime for boiling water two-phase flow in horizontal rectangular narrow channels

International Nuclear Information System (INIS)

Zhang Chunwei; Qiu Suizheng; Yan Mingyu; Wang Bulei; Nie Changhua

2005-01-01

The flow regime transition criteria for the boiling water two-phase flow in horizontal rectangular narrow channels (1 x 20 mm, 2 x 20 mm) were theoretically explored. The discernible flow patterns were bubble, intermittent slug, churn, annular and steam-water separation flow. By using two-fluid model, equations of conservation of momentum were established for the two-phase flow. New flow-regime criteria were obtained and agreed well with the experiment data. (authors)

Macroscopic balance equations for two-phase flow models

International Nuclear Information System (INIS)

Hughes, E.D.

1979-01-01

The macroscopic, or overall, balance equations of mass, momentum, and energy are derived for a two-fluid model of two-phase flows in complex geometries. These equations provide a base for investigating methods of incorporating improved analysis methods into computer programs, such as RETRAN, which are used for transient and steady-state thermal-hydraulic analyses of nuclear steam supply systems. The equations are derived in a very general manner so that three-dimensional, compressible flows can be analysed. The equations obtained supplement the various partial differential equation two-fluid models of two-phase flow which have recently appeared in the literature. The primary objective of the investigation is the macroscopic balance equations. (Auth.)

Nonlinear dynamics of two-phase flow

International Nuclear Information System (INIS)

Rizwan-uddin

1986-01-01

Unstable flow conditions can occur in a wide variety of laboratory and industry equipment that involve two-phase flow. Instabilities in industrial equipment, which include boiling water reactor (BWR) cores, steam generators, heated channels, cryogenic fluid heaters, heat exchangers, etc., are related to their nonlinear dynamics. These instabilities can be of static (Ledinegg instability) or dynamic (density wave oscillations) type. Determination of regions in parameters space where these instabilities can occur and knowledge of system dynamics in or near these regions is essential for the safe operation of such equipment. Many two-phase flow engineering components can be modeled as heated channels. The set of partial differential equations that describes the dynamics of single- and two-phase flow, for the special case of uniform heat flux along the length of the channel, can be reduced to a set of two coupled ordinary differential equations [in inlet velocity v/sub i/(t) and two-phase residence time tau(t)] involving history integrals: a nonlinear ordinary functional differential equation and an integral equation. Hence, to solve these equations, the dependent variables must be specified for -(nu + tau) ≤ t ≤ 0, where nu is the single-phase residence time. This system of nonlinear equations has been solved analytically using asymptotic expansion series for finite but small perturbations and numerically using finite difference techniques

Research on one-dimensional two-phase flow

International Nuclear Information System (INIS)

Adachi, Hiromichi

1988-10-01

In Part I the fundamental form of the hydrodynamic basic equations for a one-dimensional two-phase flow (two-fluid model) is described. Discussions are concentrated on the treatment of phase change inertial force terms in the equations of motion and the author's equations of motion which have a remarkable uniqueness on the following three points. (1) To express force balance of unit mass two-phase fluid instead of that of unit volume two-phase fluid. (2) To pick up the unit existing mass and the unit flowing mass as the unit mass of two-phase fluid. (3) To apply the kinetic energy principle instead of the momentum low in the evaluation of steady inertial force term. In these three, the item (1) is for excluding a part of momentum change or kinetic energy change due to mass change of the examined part of fluid, which is independent of force. The item (2) is not to introduce a phenomenological physical model into the evaluation of phase change inertial force term. And the item (3) is for correctly applying the momentum law taking into account the difference of representative velocities between the main flow fluid (vapor phase or liquid phase) and the phase change part of fluid. In Part II, characteristics of various kinds of high speed two-phase flow are clarified theoretically by the basic equations derived. It is demonstrated that the steam-water two-phase critical flow with violent flashing and the airwater two-phase critical flow without phase change can be described with fundamentally the same basic equations. Furthermore, by comparing the experimental data from the two-phase critical discharge test and the theoretical prediction, the two-phase discharge coefficient, C D , for large sharp-edged orifice is determined as the value which is not affected by the experimental facility characteristics, etc. (author)

Modeling of two-phase slug flow

International Nuclear Information System (INIS)

Fabre, J.; Line, A.

1992-01-01

When gas and liquid flow in a pipe, over a range of flow rates, a flow pattern results in which sequences of long bubbles, almost filling the pipe cross section, are successively followed by liquid slugs that may contain small bubbles. This flow pattern, usually called slug flow, is encountered in numerous practical situations, such as in the production of hydrocarbons in wells and their transportation in pipelines; the production of steam and water in geothermal power plants; the boiling and condensation in liquid-vapor systems of thermal power plants; emergency core cooling of nuclear reactors; heat and mass transfer between gas and liquid in chemical reactors. This paper provides a review of two phase slug flow modeling

Seabrook simulator model upgrade: Implementation and validation of two-phase, nonequilibrium RCS and steam generator models

International Nuclear Information System (INIS)

Kao, S.

1990-01-01

A number of deficiencies in the original RCS and steam generator models on the Seabrook simulator were found to give unrealistic results under some off-normal and accident conditions. These deficiencies are attributed to the simplistic assumptions used in the original models, such as the homogeneous, equilibrium equations used in the pressurizer and steam generator models, and the single-phase flow model used in the RCS thermal-hydraulic model. To improve the fidelity of the simulator, efforts have been made to upgrade the RCS and steam generator models to include two-phase, nonequilibrium features. In the new RCS model, the following major assumptions are used to derive the finite difference form of the conservation equations: a donor-cell differencing scheme is adopted to allow flow reversal; a single pressure is used to evaluate properties; a single mass flow rate is assumed in each loop; enthalpy is assumed to vary linearly within each control volume; a homogeneous flow is assumed under two-phase conditions. The pressurizer is divided into a vapor region and a liquid region, each of which is represented by a set of mass and energy conservation equations. Interfacial mass and energy exchange mechanisms (condensation and flashing), thermal interactions between the vessel and fluids, and thermal nonequilibrium between the phases are included in the pressurizer model. The steam generator is divided into the vapor dome, riser, and downcomer regions. The assumptions applied are similar to those of the RCS and pressurizer models. A momentum model is incorporated to calculate the recirculation flow and simulate the downcomer level shrink/swell phenomenon. The new RCS and steam generator models are validated by comparing the simulator calculations against sister plant data and FSAR vendor analysis. The results show the new models give realistic and reliable calculations under off-normal and accident conditions

Numerical Simulation of Non-Equilibrium Two-Phase Wet Steam Flow through an Asymmetric Nozzle

Directory of Open Access Journals (Sweden)

Miah Md Ashraful Alam

2017-11-01

Full Text Available The present study reported of the numerical investigation of a high-speed wet steam flow through an asymmetric nozzle. The spontaneous non-equilibrium homogeneous condensation of wet steam was numerically modeled based on the classical nucleation theory and droplet growth rate equation combined with the field conservations within the computational fluid dynamics (CFD code of ANSYS Fluent 13.0. The equations describing droplet formations and interphase change were solved sequentially after solving the main flow conservation equations. The calculations were carried out assuming the flow two-dimensional, compressible, turbulent, and viscous. The SST k-ω model was used for modeling the turbulence within an unstructured mesh solver. The validation of numerical model was accomplished, and the results showed a good agreement between the numerical simulation and experimental data. The effect of spontaneous non-equilibrium condensation on the jet and shock structures was revealed, and the condensation shown a great influence on the jet structure.

Identification of two-phase flow regimes by time-series modeling

International Nuclear Information System (INIS)

King, C.H.; Ouyang, M.S.; Pei, B.S.

1987-01-01

The identification of two-phase flow patterns in pipes or ducts is important to the design and operation of thermal-hydraulic systems, especially in the nuclear reactor cores of boiling water reactors or in the steam generators of pressurized water reactors. Basically, two-phase flow shows some fluctuating characteristics even at steady-state conditions. These fluctuating characteristics can be analyzed by statistical methods for obtaining flow signatures. There have been a number of experimental studies conducted that are concerned with the statistical properties of void fraction or pressure pulsation in two-phase flow. In this study, the authors propose a new technique of identifying the patterns of air-water two-phase flow in a vertical pipe. This technique is based on analyzing the statistic characteristics of the pressure signals of the test loop by time-series modeling

Numerical methods on flow instabilities in steam generator

International Nuclear Information System (INIS)

Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki

2008-06-01

The phenomenon of two-phase flow instability is important for the design and operation of many industrial systems and equipment, such as steam generators. The designer's job is to predict the threshold of flow instability in order to design around it or compensate for it. So it is essential to understand the physical phenomena governing such instability and to develop computational tools to model the dynamics of boiling systems. In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of Sodium-cooled Fast Breeder Reactor. As one part of the research work, the evaluations of two-phase flow instability in the steam generator are being carried out experimentally and numerically. In this report, the numerical methods were studied for two-phase flow instability analysis in steam generator. For numerical simulation purpose, the special algorithm to calculate inlet flow rate iteratively with inlet pressure and outlet pressure as boundary conditions for the density-wave instability analysis was established. There was no need to solve property derivatives and large matrices, so the spurious numerical instabilities caused by discontinuous property derivatives at boiling boundaries were avoided. Large time-step was possible. The flow instability in single heat transfer tube was successfully simulated with homogeneous equilibrium model by using the present algorithm. Then the drift-flux model including the effects of subcooled boiling and two phase slip was adopted to improve the accuracy. The computer code was developed after selecting the correlations of drift velocity and distribution parameter. The capability of drift flux model together with the present algorithm for simulating density-wave instability in single tube was confirmed. (author)

Two-phase flow steam generator simulations on parallel computers using domain decomposition method

International Nuclear Information System (INIS)

Belliard, M.

2003-01-01

Within the framework of the Domain Decomposition Method (DDM), we present industrial steady state two-phase flow simulations of PWR Steam Generators (SG) using iteration-by-sub-domain methods: standard and Adaptive Dirichlet/Neumann methods (ADN). The averaged mixture balance equations are solved by a Fractional-Step algorithm, jointly with the Crank-Nicholson scheme and the Finite Element Method. The algorithm works with overlapping or non-overlapping sub-domains and with conforming or nonconforming meshing. Computations are run on PC networks or on massively parallel mainframe computers. A CEA code-linker and the PVM package are used (master-slave context). SG mock-up simulations, involving up to 32 sub-domains, highlight the efficiency (speed-up, scalability) and the robustness of the chosen approach. With the DDM, the computational problem size is easily increased to about 1,000,000 cells and the CPU time is significantly reduced. The difficulties related to industrial use are also discussed. (author)

Two Phase Flow Split Model for Parallel Channels | Iloeje | Nigerian ...

African Journals Online (AJOL)

The model and code are capable of handling single and two phase flows, steady states and transients, up to ten parallel flow paths, simple and complicated geometries, including the boilers of fossil steam generators and nuclear power plants. A test calculation has been made with a simplified three-channel system ...

Experimental Study about Two-phase Damping Ratio on a Tube Bundle Subjected to Homogeneous Two-phase Flow

Energy Technology Data Exchange (ETDEWEB)

Sim, Woo Gun; Dagdan, Banzragch [Hannam Univ., Daejeon (Korea, Republic of)

2017-03-15

Two-phase cross flow exists in many shell-and-tube heat exchangers such as condensers, evaporators, and nuclear steam generators. The drag force acting on a tube bundle subjected to air/water flow is evaluated experimentally. The cylinders subjected to two-phase flow are arranged in a normal square array. The ratio of pitch to diameter is 1.35, and the diameter of the cylinder is 18 mm. The drag force along the flow direction on the tube bundles is measured to calculate the drag coefficient and the two-phase damping ratio. The two-phase damping ratios, given by the analytical model for a homogeneous two-phase flow, are compared with experimental results. The correlation factor between the frictional pressure drop and the hydraulic drag coefficient is determined from the experimental results. The factor is used to calculate the drag force analytically. It is found that with an increase in the mass flux, the drag force, and the drag coefficients are close to the results given by the homogeneous model. The result shows that the damping ratio can be calculated using the homogeneous model for bubbly flow of sufficiently large mass flux.

Operation of pumps in two-phase steam-water flow

International Nuclear Information System (INIS)

Grison, P.; Lauro, J.F.

1978-01-01

Determining the two-phase flow (critical or not) through a pump is an esential element for a complete description of loss of coolant accident in a PWR reactor. This article descibes the theoretical and experimental research being done on this subject in France. The model of the pump is first described and its behaviour is examined in different possible cases, particularly that of critical flow. The analysis of the behaviour of the pump is then used to define the experimental conditions for the tests. Two test loops, EVA and EPOPEE, were built. The experimental results are then compared with the theoretical forecasts [fr

Measurements of local two-phase flow parameters in a boiling flow channel

International Nuclear Information System (INIS)

Yun, Byong Jo; Park, Goon-CherI; Chung, Moon Ki; Song, Chul Hwa

1998-01-01

Local two-phase flow parameters were measured lo investigate the internal flow structures of steam-water boiling flow in an annulus channel. Two kinds of measuring methods for local two-phase flow parameters were investigated. These are a two-conductivity probe for local vapor parameters and a Pitot cube for local liquid parameters. Using these probes, the local distribution of phasic velocities, interfacial area concentration (IAC) and void fraction is measured. In this study, the maximum local void fraction in subcooled boiling condition is observed around the heating rod and the local void fraction is smoothly decreased from the surface of a heating rod to the channel center without any wall void peaking, which was observed in air-water experiments. The distributions of local IAC and bubble frequency coincide with those of local void fraction for a given area-averaged void fraction. (author)

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

2016-12-15

Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

Control-volume-based model of the steam-water injector flow

Science.gov (United States)

Kwidziński, Roman

2010-03-01

The paper presents equations of a mathematical model to calculate flow parameters in characteristic cross-sections in the steam-water injector. In the model, component parts of the injector (steam nozzle, water nozzle, mixing chamber, condensation wave region, diffuser) are treated as a series of connected control volumes. At first, equations for the steam nozzle and water nozzle are written and solved for known flow parameters at the injector inlet. Next, the flow properties in two-phase flow comprising mixing chamber and condensation wave region are determined from mass, momentum and energy balance equations. Then, water compression in diffuser is taken into account to evaluate the flow parameters at the injector outlet. Irreversible losses due to friction, condensation and shock wave formation are taken into account for the flow in the steam nozzle. In two-phase flow domain, thermal and mechanical nonequilibrium between vapour and liquid is modelled. For diffuser, frictional pressure loss is considered. Comparison of the model predictions with experimental data shows good agreement, with an error not exceeding 15% for discharge (outlet) pressure and 1 K for outlet temperature.

Instrumentation for two-phase flow measurements in code verification experiments

International Nuclear Information System (INIS)

Fincke, J.R.; Anderson, J.L.; Arave, A.E.; Deason, V.A.; Lassahn, G.D.; Goodrich, L.D.; Colson, J.B.; Fickas, E.T.

1981-01-01

The development of instrumentation and techniques for the measurement of mass flow rate in two-phase flows conducted at the Idaho National Engineering Laboratory during the past year is briefly described. Instruments discussed are the modular drag-disc turbine transducer, the gamma densitometers, the ultrasonic densitometer, Pitot tubes, and full-flow drag screens. Steady state air-water and transient steam-water data are presented

Shock wave of vapor-liquid two-phase flow

Institute of Scientific and Technical Information of China (English)

Liangju ZHAO; Fei WANG; Hong GAO; Jingwen TANG; Yuexiang YUAN

2008-01-01

The shock wave of vapor-liquid two-phase flow in a pressure-gain steam injector is studied by build-ing a mathematic model and making calculations. The results show that after the shock, the vapor is nearly com-pletely condensed. The upstream Mach number and the volume ratio of vapor have a great effect on the shock. The pressure and Mach number of two-phase shock con-form to the shock of ideal gas. The analysis of available energy shows that the shock is an irreversible process with entropy increase.

Simulation of a two phase boiling flow in Poseidon geometry with Astrid steam-water software

International Nuclear Information System (INIS)

Larrauri, D.

1997-01-01

After different validation test runs in tube an annular geometries, the simulation of a subcooled boiling flow in a rod bundle geometry has been achieved with ASTRID Steam-Water software. The experiment we have simulated is the Poseidon experiment. It is a three heating tube geometry. The thermohydraulic conditions of the simulated flow are closed to the DNB conditions. The simulation results are analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water behaviour in such a geometry brings satisfaction. The wall and the liquid temperatures are well predicted in the different parts of the flow. The void fraction reaches 40 % in the vicinity of the heating rods. Besides, the evolution of the different calculated variables shows that a three-dimensional simulation gives capital information for the analyse of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry lead us to think about simulating and analyzing rod bundle flows with ASTRID Steam-Water code. (author)

Experimental study on the convective heat transfer enhancement in single-phase steam flow by a support grid

International Nuclear Information System (INIS)

Kim, Byoung Jae; Kim, Kihwan; Kim, Dong-Eok; Youn, Young-Jung; Park, Jong-Kuk; Moon, Sang-Ki; Song, Chul-Hwa

2014-01-01

Highlights: • The convective heat transfer enhancement by support grids is investigated. • Experiments were performed in a square array 2 × 2 rod bundle. • The enhancement was affected not only by the blockage ratio also by the Reynolds number. • For low Reynolds numbers, the enhancement depends on the Reynolds number (Re). • For high Reynolds numbers, the enhancement is nearly independent of Re. - Abstract: Single-phase flow occurs in the fuel rod bundle of a pressurized water reactor, during the normal operation period or at the early stage of the reflood phase in a loss-of-coolant accident scenario. In the former period, the flow is single-phase water flow, but in the latter case, the flow is single-phase steam flow. Support grids are required to maintain a proper geometry configuration of fuel rods within nuclear fuel assemblies. This study was conducted to elucidate the effects of support grids on the convective heat transfer in single-phase steam flow. Experiments were made in a square array 2 × 2 rod bundle. The four electrically-heating rods were maintained by support grids with mixing vanes creating a swirl flow. Two types of support grids were considered in this study. The two types are geometrically similar except the blockage ratio by different mixing vane angles. For all test runs, 2 kW power was supplied to each rod. The working fluid was superheated steam with Re = 2,301–39,594. The axial profile of the rod surface temperatures was measured, and the convective heat transfer enhancement by the presence of the support grids was examined. The peak heat transfer enhancement was a function of not only the blockage ratio but also the Reynolds number. Given the same blockage ratio, the heat transfer enhancement was sensitive to the Reynolds number in laminar flow, whereas it was nearly independent of the Reynolds number in turbulent flow

Large scale steam flow test: Pressure drop data and calculated pressure loss coefficients

International Nuclear Information System (INIS)

Meadows, J.B.; Spears, J.R.; Feder, A.R.; Moore, B.P.; Young, C.E.

1993-12-01

This report presents the result of large scale steam flow testing, 3 million to 7 million lbs/hr., conducted at approximate steam qualities of 25, 45, 70 and 100 percent (dry, saturated). It is concluded from the test data that reasonable estimates of piping component pressure loss coefficients for single phase flow in complex piping geometries can be calculated using available engineering literature. This includes the effects of nearby upstream and downstream components, compressibility, and internal obstructions, such as splitters, and ladder rungs on individual piping components. Despite expected uncertainties in the data resulting from the complexity of the piping geometry and two-phase flow, the test data support the conclusion that the predicted dry steam K-factors are accurate and provide useful insight into the effect of entrained liquid on the flow resistance. The K-factors calculated from the wet steam test data were compared to two-phase K-factors based on the Martinelli-Nelson pressure drop correlations. This comparison supports the concept of a two-phase multiplier for estimating the resistance of piping with liquid entrained into the flow. The test data in general appears to be reasonably consistent with the shape of a curve based on the Martinelli-Nelson correlation over the tested range of steam quality

Instrumentation for localized measurements in two-phase flow conditions

International Nuclear Information System (INIS)

Neff, G.G.; Averill, R.H.; Shurts, S.W.

1979-01-01

Three types of instrumentation that have been developed by EG and G Idaho, Inc., and its predecessor, Aerojet Nuclear company, at the Idaho National Engineering Laboratory to investigate two-phase flow phenomenon in a nuclear reactor at the Loss-of-Fluid Test (LOFT) facility are discussed: (a) a combination drag disc-turbine transducer (DTT), (b) a multibeam nuclear hardened gamma densitometer system, and (c) a conductivity sensitive liquid level transducer (LLT). The DTT obtains data on the complex problem of two-phase flow conditions in the LOFT primary coolant system during a loss-os-coolant experiment (LOCE). The discussion of the DTT describes how a turbine, measuring coolant velocity, and a drag disc, measuring coolant momentum flux, can provide valuable mass flow data. The nuclear hardened gamma densitometer is used to obtain density and flow regime information for two-phase flow in the LOFT primary coolant system during a LOCE. The LLT is used to measure water and steam conditions within the LOFT reactor core during a LOCE. The LLT design and the type of data obtained are described

A theoretical model for measuring mass flowrate and quality of two phase flow by the noise of throttling set

International Nuclear Information System (INIS)

Tong Yunxian; Wang Wenran

1992-03-01

The mass flowrate and steam quality measuring of two phase flowrate is an essential issue in the tests of loss-of-coolant accident (LOCA). The spatial stochastic distribution of phase concentration would cause a differential pressure noise when two phase flow is crossing a throttling set. Under the assumption of that the variance of disperse phase concentration is proportional to its mean phase concentration and by using the separated flow model of two phase flow, it has demonstrated that the variance of noise of differential pressure square root is approximately proportional to the flowrate of disperse phase. Thus, a theoretical model for measuring mass flowrate and quality of two phase flow by noise measurement is developed. It indicates that there is a possibility to measure two phase flowrate and steam quality by using the simple theoretical model and a single throttling set

Operation of pumps in two-phase steam-water flow. [PWR

Energy Technology Data Exchange (ETDEWEB)

Grison, P; Lauro, J F [Electricite de France, 78 - Chatou

1978-01-01

Determining the two-phase flow (critical or not) through a pump is an esential element for a complete description of loss of coolant accident in a PWR reactor. This article descibes the theoretical and experimental research being done on this subject in France. The model of the pump is first described and its behaviour is examined in different possible cases, particularly that of critical flow. The analysis of the behaviour of the pump is then used to define the experimental conditions for the tests. Two test loops, EVA and EPOPEE, were built. The experimental results are then compared with the theoretical forecasts.

Void fraction and interfacial velocity in gas-liquid upward two-phase flow across tube bundles

International Nuclear Information System (INIS)

Ueno, T.; Tomomatsu, K.; Takamatsu, H.; Nishikawa, H.

1997-01-01

Tube failures due to flow-induced vibration are a major problem in heat exchangers and many studies on the problem of such vibration have been carried out so far. Most studies however, have not focused on two-phase flow behavior in tube bundles, but have concentrated mainly on tube vibration behavior like fluid damping, fluid elastic instability and so on. Such studies are not satisfactory for understanding the design of heat exchangers. Tube vibration behavior is very complicated, especially in the case of gas-liquid two-phase flow, so it is necessary to investigate two-phase flow behavior as well as vibration behavior before designing heat exchangers. This paper outlines the main parameters that characterize two-phase behavior, such as void fraction and interfacial velocity. The two-phase flow analyzed here is gas-liquid upward flow across a horizontal tube bundle. The fluids tested were HCFC-123 and steam-water. HCFC-123 stands for Hydrochlorofluorocarbon. Its chemical formula is CHCl 2 CF 3 , which has liquid and gas densities of 1335 and 23.9 kg/m 3 at a pressure of 0.40 MPa and 1252 and 45.7 kg/m 3 at a pressure of 0.76 MPa. The same model tube bundle was used in the two tests covered in this paper, to examine the similarity law of two-phase flow behavior in tube bundles using HCFC-123 and steam-water two-phase flow. We also show numerical simulation results for the two fluid models in this paper. We do not deal with vibration behavior and the relationship between vibration behavior and two-phase flow behavior. (author)

Development and calibration of instruments for measurements in transient two-phase flow

International Nuclear Information System (INIS)

Banerjee, S.; Heidrick, T.R.

1981-01-01

For validation and development of theoretical models for transient two-phase flow, it is necessary to measure local and cross-sectionally averaged thermalhydraulic parameters. Of these parameters, void fraction and mass velocity are the most difficult to measure. In this paper, we present our recent work on various techniques for determining these quantities. The possibility of determining flow regime by using fast neutron transmission is discussed. The development of a miniaturized electrical resistivity probe for measuring local void fraction is described, together with calibrations obtained by integrating the void fraction profile and comparing the cross-sectionally averaged void fraction with direct measurements using two quick closing valves. Results on the calibration of combinations of full-flow turbine meters, Pitot tube rakes and gamma densitometers for measuring cross-sectionally averaged mass velocity in steady steam-water flow are presented. The results are interpreted with a simple model using single-phase calibration factors for the Pitot tube rakes and turbine meters. Calibration experiments were also done in transient steam-water flows and interpretation of the results with the steady state models is also discussed

Transient two-phase flow

International Nuclear Information System (INIS)

Hsu, Y.Y.

1974-01-01

The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

Entrainment in vertical annular two-phase flow

International Nuclear Information System (INIS)

Sawant, Pravin; Ishii, Mamoru; Mori, Michitsugu

2009-01-01

Prediction of amount of entrained droplets or entrainment fraction in annular two-phase flow is essential for the estimation of dryout condition and analysis of post dryout heat transfer in light water nuclear reactors and steam boilers. In this study, air-water and organic fluid (Freon-113) annular flow entrainment experiments have been carried out in 9.4 and 10.2 mm diameter test sections, respectively. Both the experiments covered three distinct pressure conditions and wide range of liquid and gas flow conditions. The organic fluid experiments simulated high pressure steam-water annular flow conditions. In each of the experiments, measurements of entrainment fraction, droplet entrainment rate and droplet deposition rate have been performed by using a liquid film extraction method. A simple, explicit and non-dimensional correlation developed by Sawant et al. (2008a) for the prediction of entrainment fraction is further improved in this study in order to account for the existence of critical gas and liquid flow rates below which no entrainment is possible. Additionally, a new correlation is proposed for the estimation of minimum liquid film flow rate at the maximum entrainment fraction condition. The improved correlation successfully predicted the newly collected air-water and Freon-113 entrainment fraction data. Furthermore, the correlations satisfactorily compared with the air-water, helium-water and air-genklene experimental data measured by Willetts (1987). (author)

Joint test rig for testing and calibrating of different methods of two-phase mass flow measurement

International Nuclear Information System (INIS)

Reimann, J.; Demski, A.; Hahn, H.; Harten, U.; John, H.; Megerle, A.; Mueller, S.; Pawlak, L.; Wanner, E.

1977-01-01

The steam-water loop was completed by building in two throttling valves upstream of the mixing chamber. By producing steam by throttling the total mass flow may be increased up to 35% compared to the former method of operating the loop. Furthermore, throttling stabilizes the single phase mass flow measurement. The data aquisition system and computation of the reference values has been finished. The computer program contains the equations of state of steam/water and the calibration curves for all signal transducers. The 5 beam γ-densitometer has been finished mechanically and supplied with the electronics. First calibration tests are fully satisfactory. The instrumentation of the air-water loop completed. At low quality the mass fluxes are increased by a factor of 5 compared with the steam-water-loop. The regime of dispersed bubble flow is fully reached in the test section. To detect flow regimes air-water as well as in steam-water flow, a local impedance probe was used. In addition, the phase distribution across the channel could be detected by traversing the probe. The boundaries of the air-water flow regimes detected by the probe are in good correspondance with other investigations. For the first time, such experiments have been carried out in horizontal steam-water flow. The results indicate that the region of slug flow becomes smaller with increasing pressure. (orig./RW) [de

The development of two-phase flow instrumentation at PNC O-arai Engineering Center

International Nuclear Information System (INIS)

Obata, T.; Kobori, T.; Hayamizu, Y.

1975-10-01

This paper reviews development works on the two-phase flow instrumentation carried out at PNC Oarai Engineering Center for FUGEN safety test. The paper describes heater surface temperature measurement, four types of void meters and two steam quality meters. (auth.)

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

International Nuclear Information System (INIS)

Tentner, A.

2009-01-01

A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

Reverse primary-side flow in steam generators during natural circulation cooling

International Nuclear Information System (INIS)

Stumpf, H.; Motley, F.; Schultz, R.; Chapman, J.; Kukita, Y.

1987-01-01

A TRAC model of the Large Scale Test Facility with a 3-tube steam-generator model was used to analyze natural-circulation test ST-NC-02. For the steady state at 100% primary mass inventory, TRAC was in excellent agreement with the natural-circulation flow rate, the temperature distribution in the steam-generator tubes, and the temperature drop from the hot leg to the steam-generator inlet plenum. TRAC also predicted reverse flow in the long tubes. At reduced primary mass inventories, TRAC predicted the three natural-circulation flow regimes: single phase, two phase, and reflux condensation. TRAC did not predict the cyclic fill-and-dump phenomenon seen briefly in the test. TRAC overpredicted the two-phase natural-circulation flow rate. Since the core is well cooled at this time, the result is conservative. An important result of the analysis is that TRAC was able to predict the core dryout and heatup at approximately the same primary mass inventory as in the test. 4 refs., 8 figs., 2 tabs

Void fraction and flow regime determination by optical probe for boiling two-phase flow in a tube subchannel

International Nuclear Information System (INIS)

Cheng Huiping; Wu Hongtao; Ba Changxi; Yan Xiaoming; Huang Suyi

1995-12-01

In view of the need to determine void fraction and flow regime of vapor-liquid two-phase flow in the steam generator test model, domestic made optical probe was applied on a small-scale freon two-phase flow test rig. Optical probe signals were collected at a sampling rate up to 500 Hz and converted into digital form. Both the time signal, and the amplitude probability density function and FFT spectrum function calculated thereof were analysed in the time and frequency domains respectively. The threshold characterizing vapor or liquid contact with the probe tip was determined from the air-water two-phase flow pressure drop test results. Then, the boiling freon two-phase flow void fraction was determined by single threshold method, and compared with numerical heat transfer computation. Typical patterns which were revealed by the above-mentioned time signal and the functions were found corresponding to distinct flow regimes, as corroborated by visual observation. The experiment shows that the optical probe was a promising technique for two-phase flow void fraction measurement and flow regime identification (3 refs., 15 figs., 1 tab.)

Review of two-phase instabilities

Energy Technology Data Exchange (ETDEWEB)

Kang, Han Ok; Seo, Han Ok; Kang, Hyung Suk; Cho, Bong Hyun; Lee, Doo Jeong

1997-06-01

KAERI is carrying out a development of the design for a new type of integral reactors. The once-through helical steam generator is important design features. The study on designs and operating conditions which prevent flow instability should precede the introduction of one-through steam generator. Experiments are currently scheduled to understand two-phase instability, evaluate the effect of each design parameter on the critical point, and determine proper inlet throttling for the prevention of instability. This report covers general two-phase instability with review of existing studies on this topics. The general classification of two phase flow instability and the characteristics of each type of instability are first described. Special attention is paid to BWR core flow instability and once-through steam generator instability. The reactivity feedback and the effect of system parameters are treated mainly for BWR. With relation to once-through steam generators, the characteristics of convective heating and dryout point oscillation are first investigated and then the existing experimental studies are summarized. Finally chapter summarized the proposed correlations for instability boundary conditions. (author). 231 refs., 5 tabs., 47 figs

Application results of a prototype ultrasonic liquid film sensor to a 7 MPa steam-water two-phase flow experiment

International Nuclear Information System (INIS)

Aoyama, Goro; Fujimoto, Kiyoshi; Katono, Kenichi; Nagayoshi, Takuji; Baba, Atsushi; Yasuda, Kenichi

2016-01-01

A prototype ultrasonic liquid film sensor was applied to a high-temperature steam-water two-phase flow experiment. The liquid film sensor was vertically installed in a loop which was connected to HUSTLE, a multi-purpose steam source test facility. The hydraulic diameter of the measurement section was 9.4 mm. The output waveforms of the sensor were acquired with a digital oscilloscope. The fluid temperature and system pressure were kept at 288°C and 7.2 MPa, respectively, during the experiment. The pulse-echo method was used to calculate the liquid film thickness. The cross-correlation calculation was utilized to determine the time difference between the pulse reflected at the sensor surface and the pulse reflected at the liquid film surface. The time-averaged liquid film thicknesses were less than 0.055 mm in the annular flow condition. The increase of the time-averaged thickness was small with the change of the gas momentum flux. The film thicknesses measured with the sensor were compared with the past experimental results; the former were smaller than one-fourth of the thickness estimated as the mean film thickness. The comparison results suggested that the continuous liquid sublayer thickness was measured with the liquid film sensor. (author)

Gulping phenomena in transient countercurrent two-phase flow

International Nuclear Information System (INIS)

Tehrani, Ali A.K.

2001-04-01

Apart from previous work on countercurrent gas-liquid flow, transient tank drainage through horizontal off-take pipes is described, including experimental procedure, flow pattern on observations and countercurrent flow limitation results. A separate chapter is devoted to countercurrent two-phase flow in a pressurised water reactor hot-leg scaled model. Results concerning low head flooding, high head and loss of bowl flooding, transient draining of the steam generator and pressure variation and bubble detachment are presented. The following subjects are covered as well: draining of sealed tanks of vertical pipes, unsteady draining of closed vessel via vertical tube, unsteady filling of a closed vessel via vertical tube from a constant head reservoir. Practical significance of the results obtained is discussed

Two-phase flow models in unbounded two-phase critical flows

International Nuclear Information System (INIS)

Celata, G.P.; Cumo, M.; Farello, G.E.

1985-01-01

With reference to a Loss-of-Coolant Accident in Light Water Reactors, an analysis of the unbounded two-phase critical flow (i.e. the issuing two-phase jet) has been accomplished. Considering jets external shape, obtained by means of photographic pictures; pressure profiles inside the jet, obtained by means of a movable ''Pitot;'' and jet phases distribution information, obtained by means of X-rays pictures; a characterization of the flow pattern in the unbounded region of a two-phase critical flow is given. Jets X-ray pictures show the existence of a central high density ''core'' gradually evaporating all around, which gives place to a characteristic ''dartflow'' the length of which depends on stagnation thermodynamic conditions

Two Phase Flow Stability in the HTR-10 Steam Generator

Institute of Scientific and Technical Information of China (English)

居怀明; 左开芬; 刘志勇; 徐元辉

2001-01-01

A 10 MW High Temperature Gas Cooled Reactor (HTR-10) designed bythe Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.

An analytical model for prediction of two-phase (noncondensable) flow pump performance

International Nuclear Information System (INIS)

Furuya, O.

1985-01-01

During operational transients or a hypothetical LOCA (loss of coolant accident) condition, the recirculating coolant of PWR (pressurized water reactor) may flash into steam due to a loss of line pressure. Under such two-phase flow conditions, it is well known that the recirculation pump becomes unable to generate the same head as that of the single-phase flow case. Similar situations also exist in oil well submersible pumps where a fair amount of gas is contained in oil. Based on the one dimensional control volume method, an analytical method has been developed to determine the performance of pumps operating under two-phase flow conditions. The analytical method has incorporated pump geometry, void fraction, flow slippage and flow regime into the basic formula, but neglected the compressibility and condensation effects. During the course of model development, it has been found that the head degradation is mainly caused by higher acceleration on liquid phase and deceleration on gas phase than in the case of single-phase flows. The numerical results for head degradations and torques obtained with the model favorably compared with the air/water two-phase flow test data of Babcock and Wilcox (1/3 scale) and Creare (1/20 scale) pumps

A contribution to the study of two-phase steam-water critical flow

International Nuclear Information System (INIS)

Reocreux, M.

1975-06-01

Conservation equations were derived to describe two phase flow systems and conditions were established in order to satisfy critical flow. The theoretical analysis performed to establish the above condition has demonstrated the important part played by transfer terms. Experimental studies on glass and metal channels showed the importance of the way evaporation was initiated. (R.L.)

Raman scattering temperature measurements for water vapor in nonequilibrium dispersed two-phase flow

International Nuclear Information System (INIS)

Anastasia, C.M.; Neti, S.; Smith, W.R.; Chen, J.C.

1982-09-01

The objective of this investigation was to determine the feasibility of using Raman scattering as a nonintrusive technique to measure vapor temperatures in dispersed two-phase flow. The Raman system developed for this investigation is described, including alignment of optics and optimization of the photodetector for photon pulse counting. Experimentally obtained Raman spectra are presented for the following single- and two-phase samples: liquid water, atmospheric nitrogen, superheated steam, nitrogen and water droplets in a high void fraction air/water mist, and superheated water vapor in nonequilibrium dispersed flow

Numerical methods for limit problems in two-phase flow models

International Nuclear Information System (INIS)

Cordier, F.

2011-01-01

Numerical difficulties are encountered during the simulation of two-phase flows. Two issues are studied in this thesis: the simulation of phase transitions on one hand, and the simulation of both compressible and incompressible flows in the other hand. Un asymptotic study has shown that the loss of hyperbolicity of the bi fluid model was responsible for the difficulties encountered by the Roe scheme during the simulation of phase transitions. Robust and accurate polynomial schemes have thus been developed. To tackle the occasional lack of positivity of the solution, a numerical treatment based on adaptive diffusion was proposed and allowed to simulate with accuracy the test-cases of a boiling channel with creation of vapor and a tee-junction with separation of the phases. In a second part, an all-speed scheme for compressible and incompressible flows have been proposed. This pressure-based semi-implicit asymptotic preserving scheme is conservative, solves an elliptic equation on the pressure, and has been designed for general equations of state. The scheme was first developed for the full Euler equations and then extended to the Navier-Stokes equations. The good behaviour of the scheme in both compressible and incompressible regimes have been investigated. An extension of the scheme to the two-phase mixture model was implemented and demonstrated the ability of the scheme to simulate two-phase flows with phase change and a water-steam equation of state. (author) [fr

Evaluation method for two-phase flow and heat transfer in a feed-water heater

International Nuclear Information System (INIS)

Takamori, Kazuhide; Minato, Akihiko

1993-01-01

A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)

Experimental investigation of a two-phase nozzle flow

International Nuclear Information System (INIS)

Kedziur, F.; John, H.; Loeffel, R.; Reimann, J.

1980-07-01

Stationary two-phase flow experiments with a convergent nozzle are performed. The experimental results are appropriate to validate advanced computer codes, which are applied to the blowdown-phase of a loss-of-coolant accident (LOCA). The steam-water experiments present a broad variety of initial conditions: the pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of critical as well as subcritical experiments with different flow pattern is investigated. Additional air-water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiment. The layout of the nozzle and the applied measurement technique allow for a separate testing of blowdown-relevant, physical models and the determination of empirical model parameters, respectively. The measured quantities are essentially the mass flow rate, quality, axial pressure and temperature profiles as well as axial and radial density/void profiles obtained by a γ-ray absorption device. Moreover, impedance probes and a pitot probe are used. Observed phenomena like a flow contraction, radial pressure and void profiles as well as the appearance of two chocking locations are described, because their examination is rather instructive about the refinement of a program. The experimental facilities as well as the data of 36 characteristic experiments are documented. (orig.) [de

Flow structure of steam-water mixed spray

International Nuclear Information System (INIS)

Sanada, Toshiyuki; Mitsuhashi, Yuki; Mizutani, Hiroya; Saito, Takayuki

2010-01-01

In this study, the flow structure of a steam-water mixed spray is studied both numerically and experimentally. The velocity and pressure profiles of single-phase flow are calculated using numerical methods. On the basis of the calculated flow fields, the droplet behavior is predicted by a one-way interaction model. This numerical analysis reveals that the droplets are accelerated even after they are sprayed from the nozzle. Experimentally, the mixed spray is observed using an ultra-high-speed video camera, and the velocity field is measured by using the oarticle image velocimetry (PIV) technique. Along with this PIV velocity field measurement, the velocities and diameters of droplets are measured by phase Doppler anemometry. Furthermore, the mixing process of steam and water and the atomization process of a liquid film are observed using a transparent nozzle. High-speed photography observations reveal that the flow inside the nozzle is annular flow and that most of the liquid film is atomized at the nozzle throat and nozzle outlet. Finally, the optimum mixing method for steam and water is determined.

Flow structure of steam-water mixed spray

Energy Technology Data Exchange (ETDEWEB)

Sanada, Toshiyuki, E-mail: ttsanad@ipc.shizuoka.ac.j [Department of Mechanical Engineering, Shizuoka University, 3-5-1 Johoku, Naka-ku, Hamamatsu 432-8561, Shizuoka (Japan); Mitsuhashi, Yuki; Mizutani, Hiroya; Saito, Takayuki [Department of Mechanical Engineering, Shizuoka University, 3-5-1 Johoku, Naka-ku, Hamamatsu 432-8561, Shizuoka (Japan)

2010-12-15

In this study, the flow structure of a steam-water mixed spray is studied both numerically and experimentally. The velocity and pressure profiles of single-phase flow are calculated using numerical methods. On the basis of the calculated flow fields, the droplet behavior is predicted by a one-way interaction model. This numerical analysis reveals that the droplets are accelerated even after they are sprayed from the nozzle. Experimentally, the mixed spray is observed using an ultra-high-speed video camera, and the velocity field is measured by using the oarticle image velocimetry (PIV) technique. Along with this PIV velocity field measurement, the velocities and diameters of droplets are measured by phase Doppler anemometry. Furthermore, the mixing process of steam and water and the atomization process of a liquid film are observed using a transparent nozzle. High-speed photography observations reveal that the flow inside the nozzle is annular flow and that most of the liquid film is atomized at the nozzle throat and nozzle outlet. Finally, the optimum mixing method for steam and water is determined.

Development of a generalized correlation for phase-velocity measurements obtained from impedance-probe pairs in two-phase flow systems

International Nuclear Information System (INIS)

Hsu, C.T.; Keshock, E.G.; McGill, R.N.

1983-01-01

A flag type electrical impedance probe has been developed at the Oak Ridge National Lab (ORNL) to measure liquid- and vapor-phase velocities in steam-water mixtures flowing through rod bundles. Measurements are made by utilizing the probes in pairs, installed in line, parallel to the flow direction, and extending out into the flow channel. The present study addresses performance difficulties by examining from a fundamental point of view the two-phase flow system which the impedance probes typically operate in. Specifically, the governing equations (continuity, momentum, energy) were formulated for both air-water and steam-water systems, and then subjected to a scaling analysis. The scaling analysis yielded the appropriate dimensionless parameters of significance in both kinds of systems. Additionally, with the aid of experimental data obtained at ORNL, those parameters of significant magnitude were established. As a result, a generalized correlation was developed for liquid and vapor phase velocities that makes it possible to employ the impedance probe velocity measurement technique in a wide variety of test configurations and fluid combinations

Two-phase flow measurements with advanced instrumented spool pieces and local conductivity probes

International Nuclear Information System (INIS)

Turnage, K.G.; Davis, C.E.

1979-01-01

A series of two-phase, air-water and steam-water tests performed with instrumented spool pieces and with conductivity probes obtained from Atomic Energy of Canada, Ltd. is described. The behavior of the three-beam densitometer, turbine meter, and drag flowmeter is discussed in terms of two-phase models. Application of some two-phase mass flow models to the recorded spool piece data is made and preliminary results are shown. Velocity and void fraction information derived from the conductivity probes is presented and compared to velocities and void fractions obtained using the spool piece instrumentation

Numerical simulation of two phase flows in heat exchangers; Simulation numerique des ecoulements diphasiques dans les echangeurs

Energy Technology Data Exchange (ETDEWEB)

Grandotto Biettoli, M

2006-04-15

The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)

Experimental studies on flow-induced vibration to support steam generator design

International Nuclear Information System (INIS)

Pettigrew, M.J.; Gorman, D.J.

1977-06-01

Vibration experiments were done on small tube bundles of triangular and square lattice configurations in both liquid and two-phase (air-water) cross-flow. The effects of flow velocity, simulated steam quality, lattice orientation, tube location and tube frequency were explored. Tube response to random flow turbulence excitation and fluidelastic instability were observed in both liquid and two-phase cross-flow. Fluidelastic instability criteria and random forcing function characterizations are derived from this work. This information may be used in the vibration analysis of shell-and-tube heat exchanger components. (author)

Heat transfer in the over-crisis region in a steam-generating channel with steam-droplet flow

International Nuclear Information System (INIS)

Nigmatulin, B.I.; Kukharenko, V.N.

1991-01-01

Statement and results of numerical solution of the problem on heat transfer in the over-crisis region under stationary and nonstationary conditions are studied. Two-temperature model, accounting for influence of phase slipping on heat exchange between steam and droplets and direct heat interaction with heated surface is used for describing steam-droplet flow. Comparison is made of calculational results and experimental data published in the literature on heated surface temperature within broad range of parameters: pressure - p=0.3-18.5 MPa; specific mass flow rate of the mixture - G=30-5000 kg/m 2 xs; specific heat flows - q=0.02-3 MW/m 2 and tube diameter - D=4-20 mm

Analysis of flow induced valve operation and pressure wave propagation for single and two-phase flow conditions

International Nuclear Information System (INIS)

Nagel, H.

1986-01-01

The flow induced valve operation is calculated for single and two-phase flow conditions by the fluid dynamic computer code DYVRO and results are compared to experimental data. The analysis show that the operational behaviour of the valves is not only dependent on the condition of the induced flow, but also the pipe flow can cause a feedback as a result of the induced pressure waves. For the calculation of pressure wave propagation in pipes of which the operation of flow induced valves has a considerable influence it is therefore necessary to have a coupled analysis of the pressure wave propagation and the operational behaviour of the valves. The analyses of the fast transient transfer from steam to two-phase flow show a good agreement with experimental data. Hence even these very high loads on pipes resulting from such fluid dynamic transients can be calculated realistically. (orig.)

Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

Energy Technology Data Exchange (ETDEWEB)

Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case.

Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

International Nuclear Information System (INIS)

Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y.

2015-01-01

The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case

Study of nonequilibrium dispersed two phase flow

International Nuclear Information System (INIS)

Reyes, J.N. Jr.

1986-01-01

Understanding the behavior of liquid droplets in a superheated steam environment is essential to the accurate prediction of nuclear fuel rod surface temperatures during the blowdown and reflood phase of a loss-of-coolant-accident (LOCA). In response to this need, this treatise presents several original and significant contributions to the field of thermofluid physics. The research contained herein presents a statistical derivation of the two-phase mass, momentum, and energy-conservation equations using a droplet continuity equation analogous to that used in the Kinetic Theory of Gases. Unlike the Eulerian volume and time-averaged conservation equations generally used to describe dispersed two-phase flow behavior, this statistical averaging approach results in an additional mass momentum or energy term in each of the respective conservation equations. Further, this study demonstrates that current definitions of the volumetric vapor generation rate used in the mass conservation equation are inappropriate results under certain circumstances. The mass conservation equation derived herein is used to obtain a new definition for the volumetric vapor-generation rate. Last, a simple two phase phenomenological model, based on the statistically averaged conservation equations, is presented and solved analytically. It is shown that the actual quality and vapor temperature, under these circumstances, depend on a single dimensionless group

Coupling two-phase fluid flow with two-phase darcy flow in anisotropic porous media

KAUST Repository

Chen, J.

2014-06-03

This paper reports a numerical study of coupling two-phase fluid flow in a free fluid region with two-phase Darcy flow in a homogeneous and anisotropic porous medium region. The model consists of coupled Cahn-Hilliard and Navier-Stokes equations in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition on the interface between the free flow and the porous media regions. Obtained results have shown the anisotropic properties effect on the velocity and pressure of the two-phase flow. 2014 Jie Chen et al.

Coupling Two-Phase Fluid Flow with Two-Phase Darcy Flow in Anisotropic Porous Media

Directory of Open Access Journals (Sweden)

Jie Chen

2014-06-01

Full Text Available This paper reports a numerical study of coupling two-phase fluid flow in a free fluid region with two-phase Darcy flow in a homogeneous and anisotropic porous medium region. The model consists of coupled Cahn-Hilliard and Navier-Stokes equations in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition on the interface between the free flow and the porous media regions. Obtained results have shown the anisotropic properties effect on the velocity and pressure of the two-phase flow.

Model to calculate mass flow rate and other quantities of two-phase flow in a pipe with a densitometer, a drag disk, and a turbine meter

International Nuclear Information System (INIS)

Aya, I.

1975-11-01

The proposed model was developed at ORNL to calculate mass flow rate and other quantities of two-phase flow in a pipe when the flow is dispersed with slip between the phases. The calculational model is based on assumptions concerning the characteristics of a turbine meter and a drag disk. The model should be validated with experimental data before being used in blowdown analysis. In order to compare dispersed flow and homogeneous flow, the ratio of readings from each flow regime for each device discussed is calculated for a given mass flow rate and steam quality. The sensitivity analysis shows that the calculated flow rate of a steam-water mixture (based on the measurements of a drag disk and a gamma densitometer in which the flow is assumed to be homogeneous even if there is some slip between phases) is very close to the real flow rate in the case of dispersed flow at a low quality. As the steam quality increases at a constant slip ratio, all models are prone to overestimate. At 20 percent quality the overestimates reach 8 percent in the proposed model, 15 percent in Rouhani's model, 38 percent in homogeneous model, and 75 percent in Popper's model

A simple and rational numerical method of two-phase flow with volume-junction model. 2. The numerical method for general condition of two-phase flow in non-equilibrium states

International Nuclear Information System (INIS)

Okazaki, Motoaki

1997-11-01

In the previous report, the usefulness of a new numerical method to achieve a rigorous numerical calculation using a simple explicit method with the volume-junction model was presented with the verification calculation for the depressurization of a saturated two-phase mixture. In this report, on the basis of solution method above, a numerical method for general condition of two-phase flow in non-equilibrium states is presented. In general condition of two-phase flow, the combinations of saturated and non-saturated conditions of each phase are considered in the each flow of volume and junction. Numerical evaluation programs are separately prepared for each combination of flow condition. Several numerical calculations of various kinds of non-equilibrium two-phase flow are made to examine the validity of the numerical method. Calculated results showed that the thermodynamic states obtained in different solution schemes were consistent with each other. In the first scheme, the states are determined by using the steam table as a function of pressure and specific enthalpy which are obtained as the solutions of simultaneous equations. In the second scheme, density and specific enthalpy of each phase are directly calculated by using conservation equations of mass and enthalpy of each phase, respectively. Further, no accumulation of error in mass and energy was found. As for the specific enthalpy, two cases of using energy equations for the volume are examined. The first case uses total energy conservation equation and the second case uses the type of the first law of thermodynamics. The results of both cases agreed well. (author)

Two-phase pressurized thermal shock investigations using a 3D two-fluid modeling of stratified flow with condensation

International Nuclear Information System (INIS)

Yao, W.; Coste, P.; Bestion, D.; Boucker, M.

2003-01-01

In this paper, a local 3D two-fluid model for a turbulent stratified flow with/without condensation, which can be used to predict two-phase pressurized thermal shock, is presented. A modified turbulent K- model is proposed with turbulence production induced by interfacial friction. A model of interfacial friction based on a interfacial sublayer concept and three interfacial heat transfer models, namely, a model based on the small eddies controlled surface renewal concept (HDM, Hughes and Duffey, 1991), a model based on the asymptotic behavior of the Eddy Viscosity (EVM), and a model based on the Interfacial Sublayer concept (ISM) are implemented into a preliminary version of the NEPTUNE code based on the 3D module of the CATHARE code. As a first step to apply the above models to predict the two-phase thermal shock, the models are evaluated by comparison of calculated profiles with several experiments: a turbulent air-water stratified flow without interfacial heat transfer; a turbulent steam-water stratified flow with condensation; turbulence induced by the impact of a water jet in a water pool. The prediction results agree well with the experimental data. In addition, the comparison of three interfacial heat transfer models shows that EVM and ISM gave better prediction results while HDM highly overestimated the interfacial heat transfers compared to the experimental data of a steam water stratified flow

Definition of a facility for experimental studies of two-phase flows and heat transfer in porous materials

International Nuclear Information System (INIS)

Reda, D.C.; Eaton, R.R.

1981-01-01

A facility-development effort is currently underway at Sandia National Laboratories in order to create an experimental capability for the study of two-phase, steam/water flows through a variety of porous media. The facility definition phase of this project is described. Equations are derived for the steady, adiabatic, macroscopically-linear two-phase flow of a single-component fluid through a porous medium, including energy transfer both by convection and conduction. These equations are then solved to give relative permeabilities for the steam and water phases as functions of known and/or measurable quantities. A viable experimental approach was thereby formulated, leading to the definition of facility components and instrumentation requirements, including the application of gamma-beam densitometry for the measurement of liquid-saturation distributions in porous media. Finally, a state-of-the-art computer code was utilized to numerically simulate the proposed experiments, providing an estimate of the facility operating envelope

Two-phase flow in the upper plenum of a boiling water nuclear reactor

International Nuclear Information System (INIS)

Tinoco, Hernan

2003-01-01

The end part of the Emergency Core Spray System (ECSS) of the Boiling Water Reactors (BWRs) at Forsmark Nuclear Power Plant (NPP) is situated in the Upper Plenum. It consists of a pipe network equipped with water injection nozzles. In case of Lost-of-Coolant Accidents (LOCAs), the ECSS should maintain the core covered by water and, at the same time, rapidly cool and decompress the reactor by means of cold water injection. In similar reactors, some welds belonging to the ECSS support have, after a period of time, shown crack indications. Inspection, repair or replacement of these welds is time consuming and expensive. For this reason, it has now been decided to permanently remove the end part of the ECSS and to replace it by water injection in the Downcomer. However, this removal should not be accompanied by undesirable effects like an increase in the moisture of the steam used for operating the turbines. To investigate the effect of this removal on the steam moisture, a CFD analysis of the two-phase flow in the Upper Plenum of Unit 3, with and without ECSS, has been carried out by means of a two-phase Euler model in FLUENT 6.0. The inlet conditions are given by an analysis of the core kinetics and thermal hydraulics by mean of the POLCA-code. The outlet conditions, i. e. the steam separator pressure drops, are given by empirical correlations from the experiments carried out at the SNORRE facility. The predicted the mass flow-rates to each separator, together with empirical correlations for the moisture content of the steam leaving the separators and the steam dryer, indicate a slight decrease in the steam moisture when the ECSS is removed. Also, a minor decrease in pressure losses over the Upper Plenum is achieved with this removal. On the other hand, rounding the sharp edges of the inlet openings to the steam separators at the shroud cover may give a large reduction in pressure losses

Investigation of temperature fluctuation phenomena in a stratified steam-water two-phase flow in a simulating pressurizer spray pipe of a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Miyoshi, Koji, E-mail: miyoshi.koj@inss.co.jp; Takenaka, Nobuyuki; Ishida, Taisuke; Sugimoto, Katsumi

2017-05-15

Highlights: • Thermal hydraulics phenomena were discussed in a spray pipe of pressurizer. • Temperature fluctuation was investigated in a stratified steam-water two-phase. • Remarkable liquid temperature fluctuations were observed in the liquid layer. • The observed temperature fluctuations were caused by the internal gravity wave. • The temperature fluctuations decreased with increasing dissolved oxygen. - Abstract: Temperature fluctuation phenomena in a stratified steam-water two-phase flow in a horizontal rectangular duct, which simulate a pressurizer spray pipe of a pressurized water reactor, were studied experimentally. Vertical distributions of the temperature and the liquid velocity were measured with water of various dissolved oxygen concentrations. Large liquid temperature fluctuations were observed when the water was deaerated well and dissolved oxygen concentration was around 10 ppb. The large temperature fluctuations were not observed when the oxygen concentration was higher. It was shown that the observed temperature fluctuations were caused by the internal gravity wave since the Richardson numbers were larger than 0.25 and the temperature fluctuation frequencies were around the Brunt-Väisälä frequencies in the present experimental conditions. The temperature fluctuations decreased by the non-condensable gas since the non-condensable gas suppressed the condensation and the temperature difference in the liquid layer was small.

Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

Improvement of estimation method of two-phase flow in a large diameter pipe. 2. Development of mechanistic interfacial drag force model

International Nuclear Information System (INIS)

Okawa, Tomio; Yoneda, Kimitoshi

1998-01-01

It is experimentally clarified that behavior of gas-liquid two-phase flow in large diameter pipe is different from one occurred in small diameter pipe. However, no special model for large diameter pipe is used in existing nuclear reactor safety analysis codes. In the present study, detailed investigation about the two-phase flow model used in the safety analysis was carried out to specify the physical phenomena which should be modeled more precisely. Based on the investigation, steam-water two-phase flow experiments using large diameter pipe was conducted to obtain new models. As a result, new evaluation methods for bubble size, heterogeneous distribution of void fraction, and wake formed behind bubble were developed. These new models were applied to the prediction of steam-water two-phase flow experiments using large diameter pipes to clarify their validity. It was consequently demonstrated that the accuracy of the numerical solution is remarkably improved not only for the experiment used for model development but also for the experiment where the pipe diameter, pressure, velocities, void fraction are different. (author)

Identification of two-phase flow patterns in a nuclear reactor by the high-frequency contribution fraction

International Nuclear Information System (INIS)

Wang, Y.W.; Pei, B.S.; King, C.H.; Lee, S.C.

1989-01-01

Recently, King et al. and Wang et al. analyzed the fluctuating characteristics of differential pressure and void fraction by the optimum modeling method and by spectral analysis, respectively. These two investigations presented some new concepts and deterministic criteria, which are based on purely empirical formulas, to identify two-phase flow patterns. These deterministic criteria on two-phase flow patterns' identification seem to show reasonable performance. In King's and Wang's studies, there are at least three problems that need further investigations for the applications to the nuclear reactor engineering field. These three problems are the following: 1. Is the response to a certain two-phase flow pattern, i.e., the fluctuating characteristics, of neutrons the same as that of differential pressure or void fraction? 2. Could those criteria developed from air/water flow be allowed to identify steam/water two-phase flow patterns? 3. Could those criteria be applied to identify two-phase flow patterns in rod bundles? In this paper, parts of the investigated results answer the first problem, and detailed comparisons with the previous work of the authors are given on a variety of items

STUDY OF IDENTIFICATION OF TWO-PHASE FLOW PARAMETERS BY PRESSURE FLUCTUATION ANALYSIS

Directory of Open Access Journals (Sweden)

Ondrej Burian

2016-12-01

Full Text Available This paper deals with identification of parameters of simple pool boiling in a vertical rectangular channel by analysis of pressure fluctuation. In this work is introduced a small experimental facility about 9 kW power, which was used for simulation of pool boiling phenomena and creation of steam-water volume. Several pressure fluctuations measurements and differential pressure fluctuations measurements at warious were carried out. Main changed parameters were power of heaters and hydraulics resistance of channel internals. Measured pressure data was statistically analysed and compared with goal to find dependencies between parameters of two-phase flow and statistical properties of pressure fluctuation. At the end of this paper are summarized final results and applicability of this method for parameters determination of two phase flow for pool boiling conditions at ambient pressure.

System identification on two-phase flow stability

International Nuclear Information System (INIS)

Wu Shaorong; Zhang Youjie; Wang Dazhong; Bo Jinghai; Wang Fei

1996-01-01

The theoretical principle, experimental method and results of interrelation analysis identification for the instability of two-phase flow are described. A completely new concept of test technology and method on two-phase flow stability was developed by using he theory of information science on system stability and system identification for two-phase flow stability in thermo-physics field. Application of this method would make it possible to identify instability boundary of two-phase flow under stable operation conditions of two-phase flow system. The experiment was carried out on the thermohydraulic test system HRTL-5. Using reverse repeated pseudo-random sequences of heating power as input signal sources and flow rate as response function in the test, the two-phase flow stability and stability margin of the natural circulation system are investigated. The effectiveness and feasibility of identifying two-phase flow stability by using this system identification method were experimentally demonstrated. Basic data required for mathematics modeling of two-phase flow and analysis of two-phase flow stability were obtained, which are useful for analyzing, monitoring of the system operation condition, and forecasting of two-phase flow stability in engineering system

Development of a Program for Predicting Flow Instability in a Once-through Sodium-Heated Steam Generator (III)

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Yoon, Jung; Kim, Jong Bum; Jeong, Jiyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Two-phase flow systems can be subjected to several types of instability problems. Density-wave oscillation is the most common and important type of instability in boiling channels. Such instability gives difficulties in predictions of system performance and system control, and component failure due to thermal fatigue. A computer program developed for predicting two-phase flow instability in a steam generator heated by liquid sodium was presented in the previous works. Limit cycle was predicted even in a fixed node system. The amplitude of inlet flow rate is larger than that of outlet flow rate. The amplitude of phase change location oscillation within boiling-to-vapor boundary node is larger than that of liquid-to-boiling boundary node. Sodium and steam temperature are invariant at tube exit.

Investigation of two-phase flow structure in model of draught pipe of water boiling reactor VK-300

International Nuclear Information System (INIS)

Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P.

2001-01-01

VK-300 reactor represents a vessel-type boiling reactor with integral arrangement of assemblies and in-vessel steam separation at one-circuit scheme. The circuit consists of core, draught pipes, and separation facilities. The vessel of VK-300 reactor is chosen on the base of the dimensions of that of VVER-1000 reactor. The following thermal-hydraulic parameters of nuclear power plant (NPP) were investigated experimentally: dependence of void fraction upon the steam quality in mixing chamber (on the draught section input); pressure losses at different, specific zones of up-flow and down-flow sections of the circuit with free circulation; degree of steam separation in the separating chamber (at the first step of phase separation) and its dependence upon steam quality; structure of steam-water flow in draught pipes (distribution of phases over the draught pipe cross- section); presence of steam hovering and height of this hovering in inter-pipe space of draught section. (author)

Turbine flow meter response in two-phase flows

International Nuclear Information System (INIS)

Shim, W.J.; Dougherty, T.J.; Cheh, H.Y.

1996-01-01

The purpose of this paper is to suggest a simple method of calibrating turbine flow meters to measure the flow rates of each phase in a two-phase flow. The response of two 50.8 mm (2 inch) turbine flow meters to air-water, two-phase mixtures flowing vertically in a 57 mm I.D. (2.25 inch) polycarbonate tube has been investigated for both upflow and downflow. The flow meters were connected in series with an intervening valve to provide an adjustable pressure difference between them. Void fractions were measured by two gamma densitometers, one upstream of the flow meters and the other downstream. The output signal of the turbine flow meters was found to depend only on the actual volumetric flow rate of the gas, F G , and liquid, F L , at the location of the flow meter

Fluid-Elastic Instability of U-Tube Bundle in Air-Water Two-Phase Flow

International Nuclear Information System (INIS)

Chu, In Cheol; Lee, Chang Hee; Yun, Young Jung; Chung, Heung June

2007-03-01

Using steam generator U-tube flow-induced vibration test facility, the flow-induced vibration characteristics of U-tube in row 34-44 and line 71-77 were investigated. Air and water at room temperature and near atmospheric pressure were used as working fluids. In the present experiments, followings were evaluated under two-phase cross-flow condition: the fundamental vibration responses and the critical gap velocity for a fluid-elastic instability of U-tubes, the damping ratio and hydrodynamic mass of U-tubes. In addition, the fluid-elastic instability factor, K, was preliminary assessed using Connors' relation. In the case of the U-tubes which are not supported by partial egg-crate in OPR100 steam generator, it has been found that the vibration displacement of those U-tubes are highly possible to exceed the design limit even by a turbulent excitation mechanism. The damping ratio of U-tubes measured in the present experiments was significantly higher than the OPR1000 steam generator design value. The fluid-elastic instability factor of U-tube bundle obtained in the present experiments were preliminary evaluated to be mostly in the range of 6.5-10.5

Developmental assessment of RELAP5/MOD3 code against ROSA-IV/TPTF horizontal two-phase flow experiments

International Nuclear Information System (INIS)

Kukita, Yutaka; Asaka, Hideaki; Anoda, Yoshinari; Ishiguro, Misako; Tasaka, Kanji; Mimura, Yuichi; Nemoto, Toshiyuki.

1990-03-01

A developmental version of the RELAP5/Mod3 code (as of June 1989) was assessed for accuracy using experimental data taken for high-pressure (7MPa) steam-water two-phase flow in a large-diameter (0.18 m) horizontal-pipe test section of the ROSA-IV Two-Phase Flow Test Facility (TPTF). The agreement between the measured and calculated test section void fractions was much better than that for the previous generation of RELAP5 (MOD2). The improvement was achieved primarily due to the code changes with respect to the flow stratification criterion and interfacial-drag calculation scheme. (author)

Study of two-phase underexpanded jets by gas jet

International Nuclear Information System (INIS)

Uchida, Mitsunori; Someya, Satoshi; Okamoto, Koji

2008-01-01

When a heat exchange in a Fast Breeder Reactor cracks, a sodium-water reaction occurs. When a tube cracks, highly pressurized water or steam escapes into the surrounding liquid sodium and a sodium-water reaction occurs forming the disodium oxide. The disodium oxide caught in the steam jet strikes other tubes in the reactor. The struck disodium oxide can then cause these tubes to crack. The release of steam into the liquid sodium media is a two-phase flow involving underexpansion. In this paper qualitative measurement of the underexpanded gas jet which injected into water was carried our for the purpose of analyzing the behavior of the two-phase flow. (author)

One-dimensional transient unequal velocity two-phase flow by the method of characteristics

International Nuclear Information System (INIS)

Rasouli, F.

1981-01-01

An understanding of two-phase flow is important when one is analyzing the accidental loss of coolant or when analyzing industrial processes. If a pipe in the steam generator of a nuclear reactor breaks, the flow will remain critical (or choked) for almost the entire blowdown. For this reason the knowledge of the two-phase maximum (critical) flow rate is important. A six-equation model--consisting of two continuity equations, two energy equations, a mixture momentum equation, and a constitutive relative velocity equation--is solved numerically by the method of characteristics for one-dimensional, transient, two-phase flow systems. The analysis is also extended to the special case of transient critical flow. The six-equation model is used to study the flow of a nonequilibrium sodium-argon system in a horizontal tube in which the nonequilibrium sodium-argon system in a horizontal tube in which the critical flow condition is at the entrance. A four-equation model is used to study the pressure-pulse propagation rate in an isothermal air-water system, and the results that are found are compared with the experimental data. Proper initial and boundary conditions are obtained for the blowdown problem. The energy and mass exchange relations are evaluated by comparing the model predictions with results of void-fraction and heat-transfer experiments. A simplified two-equation model is obtained for the special case of two incompressible phases. This model is used in the preliminary analysis of batch sedimentation. It is also used to predict the shock formation in the gas-solid fluidized bed

Modeling of two-phase flow with thermal and mechanical non-equilibrium

International Nuclear Information System (INIS)

Houdayer, G.; Pinet, B.; Le Coq, G.; Reocreux, M.; Rousseau, J.C.

1977-01-01

To improve two-phase flow modeling by taking into account thermal and mechanical non-equilibrium a joint effort on analytical experiment and physical modeling has been undertaken. A model describing thermal non-equilibrium effects is first presented. A correlation of mass transfer has been developed using steam water critical flow tests. This model has been used to predict in a satisfactory manner blowdown tests. It has been incorporated in CLYSTERE system code. To take into account mechanical non-equilibrium, a six equations model is written. To get information on the momentum transfers special nitrogen-water tests have been undertaken. The first results of these studies are presented

Two-phase Flow Ejector as Water Refrigerant by Using Waste Heat

International Nuclear Information System (INIS)

Yamanaka, H; Nakagawa, M

2013-01-01

Energy saving and the use of clean energy sources have recently become significant issues. It is expected that clean energy sources such as solar panels and fuel cells will be installed in many private dwellings. However, when electrical power is generated, exhaust heat is simultaneously produced. Especially for the summer season, the development of refrigeration systems that can use this waste heat is highly desirable. One approach is an ejector that can reduce the mechanical compression work required in a normal refrigeration cycle. We focus on the use of water as a refrigerant, since this can be safely implemented in private dwellings. Although the energy conversion efficiency is low, it is promising because it can use heat that would otherwise be discarded. However, a steam ejector refrigeration cycle requires a large amount of energy to change saturated water into vapour. Thus, we propose a more efficient two-phase flow ejector cycle. Experiments were carried out in which the quality of the two-phase flow from a tank was varied, and the efficiency of the ejector and nozzle was determined. The results show that a vacuum state can be achieved and suction exerted with a two-phase flow state at the ejector nozzle inlet.

Investigation of Steam Flow Behavior During Horizontal Injection into Vertical Annulus

International Nuclear Information System (INIS)

Yoon, Sang H.; Kim, Won J.; Ku, Ja H.; Suh, Kune Y.; Song, Chul H.

2004-01-01

Qualification of uncertainty margins for accidents, which are classified as the design basis accidents, requires thermal hydraulic codes and related code models with an enhanced level of sophistication. In a cold leg break accident, the flow in downcomer is multidimensional and the velocity distribution of the steam flow in downcomer serves as a good example. For observation of the flow behavior near the break, experiments are performed to measure the velocity of the steam flow in a vessel scaled down from the APR1400 (Advanced Power Reactor 1400 MWe). In this case, the steam has a quality approaching unity and thus is dealt with as a single-phase gas. The velocity of the steam flow is measured by micro-Pitot tubes arranged horizontally and vertically around the outer shell of the 1/20 scaled-down test vessel OMEGA (Optimized Multidimensional Experiment Geometric Apparatus). A commercial computational fluid dynamics code yields analytic results of multidimensional flow motion in the complex annular passage with flow obstacles. CFX is run with well-defined boundary conditions to obtain velocity profiles of the steam flow in the annular downcomer. Results of CFX shed light on the experimental setup as to fixing the location and angle of the micro-Pitot tubes, and correcting the sensitivity of the micro- Pitot tubes, for instance. This study aims to improve the multidimensional capability of the MARS code, which is based on RELAP5 and COBRA-IV, in predicting the multiphase flow behavior in the reactor downcomer. MARS is currently based on one- and two-dimensional flow analyses, which tends to distort total flow due to misrepresentation of the local phenomena. It is thus necessary to scrutinize the steam flow path and mechanistically model the momentum variation. These experimental and analytical results can locally be applied to developing the models of specific forms and essential phenomena treated in MARS. (authors)

Stratified steady and unsteady two-phase flows between two parallel plates

International Nuclear Information System (INIS)

Sim, Woo Gun

2006-01-01

To understand fluid dynamic forces acting on a structure subjected to two-phase flow, it is essential to get detailed information about the characteristics of two-phase flow. Stratified steady and unsteady two-phase flows between two parallel plates have been studied to investigate the general characteristics of the flow related to flow-induced vibration. Based on the spectral collocation method, a numerical approach has been developed for the unsteady two-phase flow. The method is validated by comparing numerical result to analytical one given for a simple harmonic two-phase flow. The flow parameters for the steady two-phase flow, such as void fraction and two-phase frictional multiplier, are evaluated. The dynamic characteristics of the unsteady two-phase flow, including the void fraction effect on the complex unsteady pressure, are illustrated

Experimental CFD grade data for stratified two-phase flows

Energy Technology Data Exchange (ETDEWEB)

Vallee, Christophe, E-mail: c.vallee@fzd.d [Forschungszentrum Dresden-Rossendorf e.V., Institute of Safety Research, D-01314 Dresden (Germany); Lucas, Dirk; Beyer, Matthias; Pietruske, Heiko; Schuetz, Peter; Carl, Helmar [Forschungszentrum Dresden-Rossendorf e.V., Institute of Safety Research, D-01314 Dresden (Germany)

2010-09-15

Stratified two-phase flows were investigated at two test facilities with horizontal test-sections. For both, rectangular channel cross-sections were chosen to provide optimal observation possibilities for the application of optical measurement techniques. In order to show the local flow structure, high-speed video observation was applied, which delivers the high-resolution in space and time needed for CFD code validation. The first investigations were performed in the Horizontal Air/Water Channel (HAWAC), which is made of acrylic glass and allows the investigation of air/water co-current flows at atmospheric pressure and room temperature. At the channel inlet, a special device was designed for well-defined and adjustable inlet boundary conditions. For the quantitative analysis of the optical measurements performed at the HAWAC, an algorithm was developed to recognise the stratified interface in the camera frames. This allows to make statistical treatments for comparison with CFD calculation results. As an example, the unstable wave growth leading to slug flow is shown from the test-section inlet. Moreover, the hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was investigated in this closed channel. The structure of the hydraulic jump over time is revealed by the calculation of the probability density of the water level. A series of experiments show that the hydraulic jump profile and its position from the inlet vary substantially with the inlet boundary conditions due to the momentum exchange between the phases. The second channel is built in the pressure chamber of the TOPFLOW test facility, which is used to perform air/water and steam/water experiments at pressures of up to 5.0 MPa and temperatures of up to 264 {sup o}C, but under pressure equilibrium with the vessel inside. In the present experiment, the test-section represents a flat model of the hot leg of the German Konvoi pressurised water reactor scaled at

Experimental CFD grade data for stratified two-phase flows

International Nuclear Information System (INIS)

Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Pietruske, Heiko; Schuetz, Peter; Carl, Helmar

2010-01-01

Stratified two-phase flows were investigated at two test facilities with horizontal test-sections. For both, rectangular channel cross-sections were chosen to provide optimal observation possibilities for the application of optical measurement techniques. In order to show the local flow structure, high-speed video observation was applied, which delivers the high-resolution in space and time needed for CFD code validation. The first investigations were performed in the Horizontal Air/Water Channel (HAWAC), which is made of acrylic glass and allows the investigation of air/water co-current flows at atmospheric pressure and room temperature. At the channel inlet, a special device was designed for well-defined and adjustable inlet boundary conditions. For the quantitative analysis of the optical measurements performed at the HAWAC, an algorithm was developed to recognise the stratified interface in the camera frames. This allows to make statistical treatments for comparison with CFD calculation results. As an example, the unstable wave growth leading to slug flow is shown from the test-section inlet. Moreover, the hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was investigated in this closed channel. The structure of the hydraulic jump over time is revealed by the calculation of the probability density of the water level. A series of experiments show that the hydraulic jump profile and its position from the inlet vary substantially with the inlet boundary conditions due to the momentum exchange between the phases. The second channel is built in the pressure chamber of the TOPFLOW test facility, which is used to perform air/water and steam/water experiments at pressures of up to 5.0 MPa and temperatures of up to 264 o C, but under pressure equilibrium with the vessel inside. In the present experiment, the test-section represents a flat model of the hot leg of the German Konvoi pressurised water reactor scaled at 1

Experimental and Analytical Study of Lead-Bismuth-Water Direct Contact Boiling Two-Phase Flow

Science.gov (United States)

Novitrian; Dostal, Vaclav; Takahashi, Minoru

The characteristics of lead-bismuth(Pb-Bi)-water boiling two-phase flow were investigated experimentally and analytically using a Pb-Bi-water direct contact boiling two-phase flow loop. Pb-Bi flow rates and void fraction were measured in a vertical circular tube at conditions of system pressure 7MPa, liquid metal temperature 460°C and injected water temperature 220°C. The drift-flux model with the assumption that bubble sizes were dependent on the fluid surface tension and the density ratio of Pb-Bi to steam-water mixture was chosen and modified by the best fit to the measured void fraction. Pb-Bi flow rates were analytically estimated using balance condition between buoyancy force and pressure losses, where the buoyancy force was calculated from void fraction estimated using the modified drift-flux model. The deviation of the analytical results of the flow rates from the experimental ones was less than 10%.

Influence of the degree of simplification of the two-phase hydrodynamic model on the simulated behaviour dynamics of a steam generator

International Nuclear Information System (INIS)

Dupont, J.F.

1979-03-01

The principal simplifications of a mathematical model for the simulation of behaviour dynamics of a two-phase flow with heat exchange are examined, as it appears in a steam generator. The theoretical considerations and numerical solutions permit the evaluation of the validity limits and the influence of these simplifications on the results. (G.T.H.)

Interaction between local parameters of two-phase flow and random forces on a cylinder

International Nuclear Information System (INIS)

Sylviane Pascal-Ribot; Yves Blanchet; Franck Baj; Phillippe Piteau

2005-01-01

Full text of publication follows: In the frame of assessments of steam generator tube bundle vibrations, a study was conducted in order to investigate the effects of an air/water flow on turbulent buffeting forces induced on a cylinder. The main purpose is to relate the physical parameters characterizing an air/water two-phase crossflow with the structural loading of a fixed cylindrical tube. In this first approach, the experiments are carried out in a rectangular acrylic test section supplied with a vertical upward bubbly flow. This flow is transversally impeded by a fixed rigid 12,15 mm diameter cylinder. Different turbulence grids are used in order to modify two-phase characteristics such as bubble diameter, void fraction profile, fluctuation parameters. Preliminarily, a dimensional analysis of fluid-structure interaction under two-phase turbulent solicitations has enabled to identify a list of physically relevant variables which must be measured to evaluate the random forces. The meaning of these relevant parameters as well as the effect of flow patterns are discussed. Direct measurements of two-phase flow parameters are performed simultaneously with measurements of forces exerted on the cylinder. The main descriptive parameters of a two-phase flow are measured using a bi-optical probe, in particular void fraction profiles, interfacial velocities, bubble diameters, void fraction fluctuations. In the same time, the magnitude of random forces caused by two-phase flow is measured with a force transducer. A thorough analysis of the experimental data is then undertaken in order to correlate physical two-phase mechanisms with the random forces exerted on the cylinder. The hypotheses made while applying the dimensional analysis are verified and their pertinence is discussed. Finally, physical parameters involved in random buffeting forces applied on a transverse tube are proposed to scale the spectral magnitude of these forces and comparisons with other authors

Two-phase flow dynamics in a model steam generator under vertical acceleration oscillation field

International Nuclear Information System (INIS)

Ishida, T.; Teshima, N.; Sakurai, S.

1992-01-01

The influence of periodically varying acceleration on hydrodynamic response has been studied experimentally using an experimental rig which models a marine reactor subject to vertical motion. The effect on the primary loop is small, but the effect on the secondary loop is large. The variables of the secondary loop, such as circulation flow rate and water level, oscillate with acceleration. The variation of gains in frequency response is analysed. The variations of flow in the secondary loop and in the downcome water level, increase in proportion to the acceleration. The effect of the flow resistance in the secondary loop on the two-phase flow dynamics is clarified. (7 figures) (Author)

Implicit approximate Riemann solver for two fluid two phase flow models

International Nuclear Information System (INIS)

Raymond, P.; Toumi, I.; Kumbaro, A.

1993-01-01

This paper is devoted to the description of new numerical methods developed for the numerical treatment of two phase flow models with two velocity fields which are now widely used in nuclear engineering for design or safety calculations. These methods are finite volumes numerical methods and are based on the use of Approximate Riemann Solver's concepts in order to define convective flux versus mean cell quantities. The first part of the communication will describe the numerical method for a three dimensional drift flux model and the extensions which were performed to make the numerical scheme implicit and to have fast running calculations of steady states. Such a scheme is now implemented in the FLICA-4 computer code devoted to 3-D steady state and transient core computations. We will present results obtained for a steady state flow with rod bow effect evaluation and for a Steam Line Break calculation were the 3-D core thermal computation was coupled with a 3-D kinetic calculation and a thermal-hydraulic transient calculation for the four loops of a Pressurized Water Reactor. The second part of the paper will detail the development of an equivalent numerical method based on an approximate Riemann Solver for a two fluid model with two momentum balance equations for the liquid and the gas phases. The main difficulty for these models is due to the existence of differential modelling terms such as added mass effects or interfacial pressure terms which make hyperbolic the model. These terms does not permit to write the balance equations system in a conservative form, and the classical theory for discontinuity propagation for non-linear systems cannot be applied. Meanwhile, the use of non-conservative products theory allows the study of discontinuity propagation for a non conservative model and this will permit the construction of a numerical scheme for two fluid two phase flow model. These different points will be detailed in that section which will be illustrated by

Numerical simulation of gas-liquid two-phase flow behavior with condensation heat transfer

International Nuclear Information System (INIS)

Takamori, Kazuhide; Murase, Michio; Baba, Yoshikazu; Aihara, Tsuyoshi.

1995-01-01

In this study, condensation heat transfer experiments were performed in order to verify a condensation heat transfer model coupled with a three-dimensional two-phase flow analysis. In the heat transfer model, the liquid film flow rate on the heat transfer tubes was calculated by a mass balance equation and the liquid film thickness was calculated from the liquid film flow rate using Nusselt's laminar flow model and Fujii's equation for steam velocity effect. In the experiments, 112 horizontal staggered tubes with an outer diameter of 16 mm and length of 0.55 m were used. Steam and spray water were supplied to the test section, and inlet quality was controlled by the spray water flow rate. The temperature was 100degC and the pressure was 0.1 MPa. The overall heat transfer coefficients were measured for inlet quality of 13-100%. From parameter calculations for the falling liquid film ratio from the upper tubes to the lower tubes, the calculated overall heat transfer coefficients agreed with the data to within ±5% at the falling liquid film ratio of 0.7. (author)

Preliminary Two-Phase Terry Turbine Nozzle Models for RCIC Off-Design Operation Conditions

Energy Technology Data Exchange (ETDEWEB)

Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, James [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-06-12

This report presents the effort to extend the single-phase analytical Terry turbine model to cover two-phase off-design conditions. The work includes: (1) adding well-established two-phase choking models – the Isentropic Homogenous Equilibrium Model (IHEM) and Moody’s model, and (2) theoretical development and implementation of a two-phase nozzle expansion model. The two choking models provide bounding cases for the two-phase choking mass flow rate. The new two-phase Terry turbine model uses the choking models to calculate the mass flow rate, the critical pressure at the nozzle throat, and steam quality. In the divergent stage, we only consider the vapor phase with a similar model for the single-phase case by assuming that the liquid phase would slip along the wall with a much slower speed and will not contribute the impulse on the rotor. We also modify the stagnation conditions according to two-phase choking conditions at the throat and the cross-section areas for steam flow at the nozzle throat and at the nozzle exit. The new two-phase Terry turbine model was benchmarked with the same steam nozzle test as for the single-phase model. Better agreement with the experimental data is observed than from the single-phase model. We also repeated the Terry turbine nozzle benchmark work against the Sandia CFD simulation results with the two-phase model for the pure steam inlet nozzle case. The RCIC start-up tests were simulated and compared with the single-phase model. Similar results are obtained. Finally, we designed a new RCIC system test case to simulate the self-regulated Terry turbine behavior observed in Fukushima accidents. In this test, a period inlet condition for the steam quality varying from 1 to 0 is applied. For the high quality inlet period, the RCIC system behaves just like the normal operation condition with a high pump injection flow rate and a nominal steam release rate through the turbine, with the net addition of water to the primary system; for

Flow characteristics in nuclear steam turbine blade passage

International Nuclear Information System (INIS)

Ahn, H.J.; Yoon, W.H.; Kwon, S.B.

1995-01-01

The rapid expansion of condensable gas such as moist air or steam gives rise to nonequilibrium condensation. As a result of irreversibility of condensation process in the nuclear steam turbine blade passage, the entropy of the flow increases, and the efficiency of the turbine decreases. In the present study, in order to investigate the flow characteristics of moist air in two-dimensional turbine blade passage which is made from the configuration of the last stage tip section of the actual nuclear steam turbine moving blade, the static pressures along both pressure and suction sides of blade are measured by static pressure taps and the distribution of Mach number on both sides of the blade are obtained by using the measured static pressure. Also, the flow field is visualized by a Schlieren system. From the experimental results, the effects of the stagnation temperature and specific humidity on the flow properties in the two dimensional steam turbine blade passage are clearly identified

Improvement of Estimation method for two-phase flow in a large-diameter pipe. Pt. 4. Effect of the inlet boundary condition of the upward flow section on flow characteristics

International Nuclear Information System (INIS)

Yoneda, Kimitoshi; Okawa, Tomio; Zhou, Shirong

1999-01-01

In nuclear power plants, many large-diameter pipes are subject to gas-liquid two-phase flow. For rational design and performance estimation, the flow in the pipes should be predicted accurately. With the correlation used at present, however, the flow analysis can not reach desirable precision. This is partly due to the lack of understanding of the two-phase flow characteristics in large-diameter pipes. Therefore, steam-water two-phase flow in a vertical pipe (155 mm i.d.) was investigated empirically. Lateral distribution data of phase volume fraction, gas velocity and bubble diameter were obtained. The effects of the inlet boundary condition were also observed. The drift velocity in the developing region was considerably affected by the inlet boundary condition. By deriving the correlation of mean bubble diameter as a function of void fraction and pressure, the empirical data was predicted with high accuracy compared with the existing correlation used in best-estimate codes of nuclear reactor safety analysis. (author)

Comparison of heat transfer in straight and corrugated minichannels with two-phase flow

Directory of Open Access Journals (Sweden)

Peukert P.

2014-03-01

Full Text Available Measurements of heat transfer rates performed with an experimental condensation heat exchanger are reported for a corrugated minichannel tube and for a straight minichannel tube. The two cases were compared at same flow regimes. The corrugation appears advantageous for relatively low steam pressures and flow rates where much higher heat transfer rates were observed close to the steam entrance, thus allowing shortening the heat exchanger with the associated advantages of costs lowering and smaller built-up space. At high steam pressures and high flow rates both tubes performed similarly.

Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher

International Nuclear Information System (INIS)

Tahara, M.; Suzuki, Y.; Abe, N.; Kurita, T.; Hamazaki, R.; Kojima, Y.

2008-01-01

Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in

Joint test rig for tests and calibration of different methods of two-phase mass flow measurement

International Nuclear Information System (INIS)

John, H.; Erbacher, F.; Wanner, E.

1975-01-01

On behalf of the Federal Ministry of Research and Technology, the Institute of Reactor Components (IRB) has begun building a test rig which will be used for testing and calibrating the methods of measuring non-steady state two-phase mass flows developed by various research agencies. The test rig is designed for the generation of steam-water mixtures of any mixing ratio and a maximum pressure of 160 data. Depending on the mixing ratio, the mass flow will reach a maximum level of 10 to 20 t/h. The conceptual design phase of the test rig has largely been finished, the component ordering phase has begun. (orig.) [de

Thermo-fluid dynamics of two-phase flow

CERN Document Server

Ishii, Mamoru; Ishii, Mamoru; Ishii, M

2006-01-01

Provides a very systematic treatment of two phase flow problems from a theoretical perspectiveProvides an easy to follow treatment of modeling and code devlopemnt of two phase flow related phenomenaCovers new results of two phase flow research such as coverage of fuel cells technology.

Numerical simulation of countercurrent flow based on two-fluid model

Energy Technology Data Exchange (ETDEWEB)

Chen, H.D. [Institute of Nuclear Engineering and Technology, Sun Yat-sen University, Zhuhai 519082 (China); School of Electric Power, South China University of Technology, Guangzhou 510640 (China); Zhang, X.Y., E-mail: zxiaoying@mail.sysu.edu.cn [Institute of Nuclear Engineering and Technology, Sun Yat-sen University, Zhuhai 519082 (China)

2017-03-15

Highlights: • Using one-dimensional two-fluid model to help understanding counter-current flow two-phase flows. • Using surface tension model to make the one-dimensional two-fluid flow model well-posed. • Solving the governing equations with a modified SIMPLE algorithm. • Validating code with experimental data and applying it to vertical air/steam countercurrent flow condition - Abstract: In order to improve the understanding of counter-current two-phase flows, a transient analysis code is developed based on one-dimensional two-fluid model. A six equation model has been established and a two phase pressure model with surface tension term, wall drag force and interface shear terms have been used. Taking account of transport phenomenon, heat and mass transfer models of interface were incorporated. The staggered grids have been used in discretization of equations. For validation of the model and code, a countercurrent air-water problem in one experimental horizontal stratified flow has been considered firstly. Comparison of the computed results and the experimental one shows satisfactory agreement. As the full problem for investigation, one vertical pipe with countercurrent flow of steam-water and air-water at same boundary condition has been taken for study. The transient distribution of liquid fraction, liquid velocity and gas velocity for selected positions of steam-water and air-water problem were presented and discussed. The results show that these two simulations have similar transient behavior except that the distribution of gas velocity for steam-water problem have larger oscillation than the one for air-water. The effect of mesh size on wavy characteristics of interface surface was also investigated. The mesh size has significant influence on the simulated results. With the increased refinement, the oscillation gets stronger.

Two-phase flow in refrigeration systems

CERN Document Server

Gu, Junjie; Gan, Zhongxue

2013-01-01

Two-Phase Flow in Refrigeration Systems presents recent developments from the authors' extensive research programs on two-phase flow in refrigeration systems. This book covers advanced mass and heat transfer and vapor compression refrigeration systems and shows how the performance of an automotive air-conditioning system is affected through results obtained experimentally and theoretically, specifically with consideration of two-phase flow and oil concentration. The book is ideal for university postgraduate students as a textbook, researchers and professors as an academic reference book, and b

Numerical analysis of water hammer induced by injection of subcooled water into steam flow in a horizontal pipe

International Nuclear Information System (INIS)

Minato, Akihiko; Nagoyoshi, Takuji; Nakamura, Akira; Fujii, Yuzo; Aya, Izuo; Yamane, Kenji

2004-01-01

Subcooled water injection into steam flow in piping systems may generate a water column containing a large steam slug. The steam slug collapses due to rapid condensation and interfaces on both sides collides with each other. Water hammer takes place and sharp pressure pulse propagates through the pipe. The purpose of this study is to show capability of the present numerical simulation method for predictions of pressure transient and loads on a piping system following steam slug collapse. A three-dimensional computer code for transient gas-liquid two-phase flow was applied to simulate an experiment of steam-condensation-induced water hammer with a horizontal polycarbonate pipe. The code was based on the extended two-fluid model, which treated interface motion using the VOF (Volume of Fluid) technique. The Godunov scheme of highly compressible single-phase flow was modified for application to the Riemann problem solution of gas-liquid mixture. Analysis of local steam slug collapse resulted in comparable peak pressure and pulse width of pressure transients with the observation. The calculation of pressure pulse propagation and impact load on piping system showed the quasi-steady pressure load was imposed especially on elbow at 1/10 of water hammer peak pressure. (author)

Refrigeration. Two-Phase Flow. Flow Regimes and Pressure Drop

DEFF Research Database (Denmark)

Knudsen, Hans-Jørgen Høgaard

2002-01-01

The note gives the basic definitions used in two-phase flow. Flow regimes and flow regimes map are introduced. The different contributions to the pressure drop are stated together with an imperical correlation from the litterature.......The note gives the basic definitions used in two-phase flow. Flow regimes and flow regimes map are introduced. The different contributions to the pressure drop are stated together with an imperical correlation from the litterature....

Burnout specific features in steam-water mixture annular flow in a tube

International Nuclear Information System (INIS)

Doroshchuk, V.E.

1981-01-01

Some unexplained burnout specific features in a steam-generating tube are analysed on the basis of experimental data. The following problems are considered: 1) the effect of the tube length and the state of the working medium (single-phase, two-phase) on burnout at the tube inlet; 2) the character of the specific thermal flow dependence at the moment of burnout appearance on the mass steam content q=f(x). It is found that the effect of the tube length on the burnout exists only in a relatively narrow range of the operating parameters. The run of the q=f(x) dependence is also explained [ru

Analysis of Two Phase Natural Circulation Flow in the Cooling Channel of the PECS

Energy Technology Data Exchange (ETDEWEB)

Park, R. J; Ha, K. S; Rhee, B. W; Kim, H. Y [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Decay heat and sensible heat of the relocated and spread corium are removed by the natural circulation flow at the bottom and side wall of the core catcher and the top water cooling of the corium. The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. As shown in Fig. 1, the engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting in the PECS. To maintain the integrity of the ex-vessel core catcher, however, it is necessary that the coolant be sufficiently circulated along the inclined cooling channel to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. For this reason, a verification experiment on the cooling capability of the EU-APR1400 core catcher has been performed in the CE (Cooling Experiment)-PECS facility at KAERI. Preliminary simulations of two-phase natural circulation in the CE-PECS were performed to predict two-phase flow characteristics and to determine the natural circulation mass flow rate in the flow channel. In this study, simulations of two-phase natural circulation in a real core catcher of the PECS have been performed to determine the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code.

Industrial aspects of gas-liquid two-phase flow

International Nuclear Information System (INIS)

Hewitt, G.F.

1977-01-01

The lecture begins by reviewing the various types of plant in which two phase flow occurs. Specifically, boiling plant, condensing plant and pipelines are reviewed, and the various two phase flow problems occurring in them are described. Of course, many other kinds of chemical engineering plant involve two phase flow, but are somewhat outside the scope of this lecture. This would include distillation columns, vapor-liquid separators, absorption towers etc. Other areas of industrial two phase flow which have been omitted for space reasons from this lecture are those concerned with gas/solids, liquid/solid and liquid/liquid flows. There then follows a description of some of the two phase flow processes which are relevant in industrial equipment and where special problems occur. The topics chosen are as follows: (1) pressure drop; (2) horizontal tubes - separation effects non-uniformites in heat transfer coefficient, effect of bends on dryout; (3) multicomponent mixtures - effects in pool boiling, mass transfer effects in condensation and Marangoni effects; (4) flow distribution - manifold problems in single phase flow, separation effects at a single T-junction in two phase flow and distribution in manifolds in two phase flow; (5) instability - oscillatory instability, special forms of instability in cryogenic systems; (6) nucleate boiling - effect of variability of surface, unresolved problems in forced convective nucleate boiling; and (7) shell side flows - flow patterns, cross flow boiling, condensation in cross flow

Unsteady coupling effects of wet steam in steam turbines flows

International Nuclear Information System (INIS)

Blondel, Frederic

2014-01-01

In addition to conventional turbomachinery problems, both the behavior and performances of steam turbines are highly dependent on the vapour thermodynamic state and the presence of a liquid phase. EDF, the main French electricity producer, is interested in further developing its' modelling capabilities and expertise in this area to allow for operational studies and long-term planning. This PhD thesis explores the modelling of wetness formation and growth in a steam turbine and an analysis of the coupling between the liquid phase and the main flow unsteadiness. To this end, the work in this thesis took the following approach. Wetness was accounted for using a homogeneous model coupled with transport equations to take into account the effects of non-equilibrium phenomena, such as the growth of the liquid phase and nucleation. The real gas attributes of the problem demanded adapted numerical methods. Before their implementation in the 3D elsA solver, the accuracy of the chosen models was tested using a developed one-dimensional nozzle code. In this manner, various condensation models were considered, including both poly-dispersed and monodispersed behaviours of the steam. Finally, unsteady coupling effects were observed from several perspectives (1D, 1D - 3D, 3D), demonstrating the ability of the method of moments to sustain unsteady phenomena which were not apparent in a simple monodispersed model. (author)

The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

International Nuclear Information System (INIS)

Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

2005-01-01

The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

Mechanism of falling water limitation in two-phase counter flow through single hole vertical channel

International Nuclear Information System (INIS)

Sudo, Yukio; Ohnuki, Akira

1983-01-01

In the safety evaluation at the time of loss coolant accident, which is a credible accident in LWRs, recently main effort has been concentrated to the optimum evaluation calculation, and the grasp of vapor-liquid two-phase flow phenomena has become important. As one of the important phenomena, there is the limitation of falling water in two-phase counter flow through a vertical channel. This phenomenon is divided into the limitation of falling water stored in an upper plenum to a core through an upper core-supporting plate and a tie plate at the time of reflooding, and the limitation of falling emergency core-cooling water in downcomer channels at the time of reflooding in PWRs, under the presence of rising steam flow. In both cases, the evaluation of the quantity of falling water is important, because it contributes directly to core cooling. In this research, in order to clarify the mechanism of limitation of falling water in two-phase vertical counter flow, first, two-phase flow of air-water system through a single-hole vertical channel was taken up, and the effect of main parameters was experimentally studied. At the same time, the theoretical investigation was performed, and the comparison with the experimental results obtained so far was carried out. The different mechanisms for short and long channels gave the good results. (Kako, I.)

Thermo-Fluid Dynamics of Two-Phase Flow

CERN Document Server

Ishii, Mamrou

2011-01-01

"Thermo-fluid Dynamics of Two-Phase Flow, Second Edition" is focused on the fundamental physics of two-phase flow. The authors present the detailed theoretical foundation of multi-phase flow thermo-fluid dynamics as they apply to: Nuclear reactor transient and accident analysis; Energy systems; Power generation systems; Chemical reactors and process systems; Space propulsion; Transport processes. This edition features updates on two-phase flow formulation and constitutive equations and CFD simulation codes such as FLUENT and CFX, new coverage of the lift force model, which is of part

Visualization of cross-sectional flow structure during condensation of steam in a slightly inclined horizontal tube

Energy Technology Data Exchange (ETDEWEB)

Puseya, Andree; Kim, H. [Kyung Hee University, Yongin (Korea, Republic of); Kwon, T. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

These flow characteristics called flow patterns still depend on a proper visualization technique in order to identify such local distribution. These proper distributions will have a dependence on the inclination of the tube as well, as it was demonstrated by Lips and Mayer. This work is focused on presenting an experimental investigation to visualize the cross sectional two-phase flow structure for condensation of steam in a horizontal tube and identify the liquid-gas interface using the axial-viewing technique. This innovative technique developed by Hewitt and more recently used in visualization works by Badie, permits the achievement to identify those systems in the area of interest by looking directly into the two-phase flow system during condensation of steam inside a pipe with technology such a high speed camera. An experimental work to visualize and locate the liquid-gas interface for steam condensation in horizontal tubes with slightly inclination was developed on this research The experimental results shows that the axial viewing technique works well with condensation phenomena and can be used for further developments in the field such as determination of liquid film geometry and calculation of void fraction.

A review of damping of two-phase flows

International Nuclear Information System (INIS)

Hara, Fumio

1993-01-01

Damping of two-phase flows has been recognized as one of the most unknown parameters in analyzing vibrational characteristics of structures subjected to two-phase flows since it seems to be influenced by many physical parameters involved in the physics of dynamic energy dissipation of a vibrating structure, for example, liquid viscosity, surface tension, flow velocity, mass ratio, frequency, void fraction, flow regime and so forth. This paper deals with a review of scientific works done to date on the damping of two phase flows and discussions about what has been clarified and what has not been known to us, or what kinds of research are needed about two-phase flow damping. The emphasis is put on the definition of two-phase fluid damping, damping measurement techniques, damping characteristics in relation to two phase flow configurations, and damping generation mechanisms

Two-phase flow in fractured rock

International Nuclear Information System (INIS)

Davies, P.; Long, J.; Zuidema, P.

1993-11-01

This report gives the results of a three-day workshop on two-phase flow in fractured rock. The workshop focused on two-phase flow processes that are important in geologic disposal of nuclear waste as experienced in a variety of repository settings. The goals and objectives of the workshop were threefold: exchange information; describe the current state of understanding; and identify research needs. The participants were divided into four subgroups. Each group was asked to address a series of two-phase flow processes. The following groups were defined to address these processes: basic flow processes; fracture/matrix interactions; complex flow processes; and coupled processes. For each process, the groups were asked to address these four issues: (1) describe the two-phase flow processes that are important with respect to repository performance; (2) describe how this process relates to the specific driving programmatic issues given above for nuclear waste storage; (3) evaluate the state of understanding for these processes; and (4) suggest additional research to address poorly understood processes relevant to repository performance. The reports from each of the four working groups are given here

Two-phase flow modeling in the rod bundle subchannel analysis

International Nuclear Information System (INIS)

Hisashi, Ninokata

2006-01-01

the NASCA code capabilities for BT is described. There a combination of experimental and computational fluid dynamics approaches is undertaken to construct a two-phase fluid dynamics database. The experimental approach consists of 1) high-resolution air-water tests performed under the room-temperature and atmospheric pressure conditions for the inter-subchannel exchanges, three-dimensional behaviors of liquid films, and spacer effects; and 2) integral steam-water tests performed at high-temperature and at higher pressure. In the integral tests, state-of the- arts of multi-phase flow measurement technologies are applied in order to obtain local and instantaneous data that reveal underlying detailed physical processes including high resolution void distributions inside a 4 x 4 bundle, liquid film thickness and two-phase flow regime. The analytical approach consists of computational multi-phase fluid dynamics (CMFD) applicable to two-phase flows. A physical interpretation of the equilibrium two-phase flow redistribution inside a rod bundle is discussed that is considered to closely be related to the void drift phenomena. Identification of interactions among dominant factors is a main objective of the integral test and acquired data will be utilized in verifying the improved subchannel code. Construction of a complete set of two-phase fluid dynamics database will be made by supplementing missing data regions with the aid of numerical analyses. Dependency on important state variables is extracted from the database and prototype constitutive equations are going to be proposed in the final stage of the project. (author)

Two-phase flow modeling in the rod bundle subchannel analysis

International Nuclear Information System (INIS)

Hisashi, Ninokata

2004-01-01

methodology adopted to improve the NASCA code capabilities for BT is described. There a combination of experimental and computational fluid dynamics approaches is undertaken to construct a two-phase fluid dynamics database. The experimental approach consists of 1) high-resolution air-water tests performed under the room-temperature and atmospheric pressure conditions for the inter-subchannel exchanges, three-dimensional behaviors of liquid films, and spacer effects; and 2) integral steam-water tests performed at high-temperature and at higher pressure. In the integral tests, state-of-the- arts of multi-phase flow measurement technologies are applied in order to obtain local and instantaneous data that reveal underlying detailed physical processes including high resolution void distributions inside a 4 x 4 bundle, liquid film thickness and two-phase flow regime. The analytical approach consists of computational multi-phase fluid dynamics (CMFD) applicable to two-phase flows. A physical interpretation of the equilibrium two-phase flow redistribution inside a rod bundle is discussed that is considered to closely be related to the void drift phenomena. Identification of interactions among dominant factors is a main objective of the integral test and acquired data will be utilized in verifying the improved subchannel code. Construction of a complete set of two-phase fluid dynamics database will be made by supplementing missing data regions with the aid of numerical analyses. Dependency on important state variables is extracted from the database and prototype constitutive equations are going to be proposed in the final stage of the project. (author)

Modeling of annular two-phase flow using a unified CFD approach

Energy Technology Data Exchange (ETDEWEB)

Li, Haipeng, E-mail: haipengl@kth.se; Anglart, Henryk, E-mail: henryk@kth.se

2016-07-15

Highlights: • Annular two-phase flow has been modeled using a unified CFD approach. • Liquid film was modeled based on a two-dimensional thin film assumption. • Both Eulerian and Lagrangian methods were employed for the gas core flow modeling. - Abstract: A mechanistic model of annular flow with evaporating liquid film has been developed using computational fluid dynamics (CFD). The model is employing a separate solver with two-dimensional conservation equations to predict propagation of a thin boiling liquid film on solid walls. The liquid film model is coupled to a solver of three-dimensional conservation equations describing the gas core, which is assumed to contain a saturated mixture of vapor and liquid droplets. Both the Eulerian–Eulerian and the Eulerian–Lagrangian approach are used to describe the droplet and vapor motion in the gas core. All the major interaction phenomena between the liquid film and the gas core flow have been accounted for, including the liquid film evaporation as well as the droplet deposition and entrainment. The resultant unified framework for annular flow has been applied to the steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show good agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.

Modeling of annular two-phase flow using a unified CFD approach

International Nuclear Information System (INIS)

Li, Haipeng; Anglart, Henryk

2016-01-01

Highlights: • Annular two-phase flow has been modeled using a unified CFD approach. • Liquid film was modeled based on a two-dimensional thin film assumption. • Both Eulerian and Lagrangian methods were employed for the gas core flow modeling. - Abstract: A mechanistic model of annular flow with evaporating liquid film has been developed using computational fluid dynamics (CFD). The model is employing a separate solver with two-dimensional conservation equations to predict propagation of a thin boiling liquid film on solid walls. The liquid film model is coupled to a solver of three-dimensional conservation equations describing the gas core, which is assumed to contain a saturated mixture of vapor and liquid droplets. Both the Eulerian–Eulerian and the Eulerian–Lagrangian approach are used to describe the droplet and vapor motion in the gas core. All the major interaction phenomena between the liquid film and the gas core flow have been accounted for, including the liquid film evaporation as well as the droplet deposition and entrainment. The resultant unified framework for annular flow has been applied to the steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show good agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.

Advanced numerical methods for three dimensional two-phase flow calculations in PWR

International Nuclear Information System (INIS)

Toumi, I.; Gallo, D.; Royer, E.

1997-01-01

This paper is devoted to new numerical methods developed for three dimensional two-phase flow calculations. These methods are finite volume numerical methods. They are based on an extension of Roe's approximate Riemann solver to define convective fluxes versus mean cell quantities. To go forward in time, a linearized conservative implicit integrating step is used, together with a Newton iterative method. We also present here some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. This kind of numerical method, which is widely used for fluid dynamic calculations, is proved to be very efficient for the numerical solution to two-phase flow problems. This numerical method has been implemented for the three dimensional thermal-hydraulic code FLICA-4 which is mainly dedicated to core thermal-hydraulic transient and steady-state analysis. Hereafter, we will also find some results obtained for the EPR reactor running in a steady-state at 60% of nominal power with 3 pumps out of 4, and a thermal-hydraulic core analysis for a 1300 MW PWR at low flow steam-line-break conditions. (author)

Microgravity Two-Phase Flow Transition

Science.gov (United States)

Parang, M.; Chao, D.

1999-01-01

Two-phase flows under microgravity condition find a large number of important applications in fluid handling and storage, and spacecraft thermal management. Specifically, under microgravity condition heat transfer between heat exchanger surfaces and fluids depend critically on the distribution and interaction between different fluid phases which are often qualitatively different from the gravity-based systems. Heat transfer and flow analysis in two-phase flows under these conditions require a clear understanding of the flow pattern transition and development of appropriate dimensionless scales for its modeling and prediction. The physics of this flow is however very complex and remains poorly understood. This has led to various inadequacies in flow and heat transfer modeling and has made prediction of flow transition difficult in engineering design of efficient thermal and flow systems. In the present study the available published data for flow transition under microgravity condition are considered for mapping. The transition from slug to annular flow and from bubbly to slug flow are mapped using dimensionless variable combination developed in a previous study by the authors. The result indicate that the new maps describe the flow transitions reasonably well over the range of the data available. The transition maps are examined and the results are discussed in relation to the presumed balance of forces and flow dynamics. It is suggested that further evaluation of the proposed flow and transition mapping will require a wider range of microgravity data expected to be made available in future studies.

One-dimensional two-phase thermal hydraulics (ENSTA course)

International Nuclear Information System (INIS)

Olive, J.

1995-11-01

This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends

Mixed convection in a two-phase flow cooling loop

International Nuclear Information System (INIS)

Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

2002-03-01

This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N PCh - N Sub ) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the modelling of the interfacial area

Mixed convection in a two-phase flow cooling loop

Energy Technology Data Exchange (ETDEWEB)

Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

2002-03-01

This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N{sub PCh} - N{sub Sub}) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the

An experimental study of two-phase natural circulation in an adiabatic flow loop

International Nuclear Information System (INIS)

Tan, M.J.; Lambert, G.A.; Ishii, Mamoru.

1988-01-01

An experimental investigation was conducted to study the two-phase flow aspect of the phenomena of interruption and resumption of natural circulation, two-phase flow patterns and pattern transitions in the hot legs of B and W light water reactor systems. The test facility was a scaled adiabatic loop designed in accordance with the scaling criteria developed by Kocamustafaogullari and Ishii. The diameter and the height of the hot leg were 10 cm and 5.5 m, respectively; the working fluid pair was nitrogen-water. The effects of the thermal center in the steam generators, friction loss in the cold leg, and configuration of the inlet to the hot leg on the flow conditions in the hot leg were investigated by varying the water level in a gas separator, controlling the size of opening of a friction loss control valve, and using two inlet geometries. Methods for estimating the distribution parameter and the average drift velocity are proposed so that they may be used in the application of one-dimensional drift-flux model to the analysis of the interruption and resumption of natural circulation in a similar geometry. 7 refs., 17 figs., 4 tabs

Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

International Nuclear Information System (INIS)

Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

2005-01-01

prediction of boiling flow regimes in a BWR fuel bundle. This strategy includes the use of local flow regime maps and flow regime specific phenomenological models in conjunction with an interface transport and topology transport approach. Specific features and capabilities which have been implemented in the CFD-BWR code are described, and the results of preliminary calculations for a typical BWR channel are presented. A set of experiments focused on two-phase flow and phase change phenomena has been identified for the validation of the CFD-BWR code. The validation experiments are described and the results of the initial analysis of an experiment focused on coolant flow in a pipe with heated wall are presented. Two cases were considered: constant and variable heat flux along the pipe. The steam generation onset location, temperature profile and void profile are determined and compared with experimental data. The paper concludes with a discussion of results and plans for future work. (authors)

High-velocity two-phase flow two-dimensional modeling

International Nuclear Information System (INIS)

Mathes, R.; Alemany, A.; Thilbault, J.P.

1995-01-01

The two-phase flow in the nozzle of a LMMHD (liquid metal magnetohydrodynamic) converter has been studied numerically and experimentally. A two-dimensional model for two-phase flow has been developed including the viscous terms (dragging and turbulence) and the interfacial mass, momentum and energy transfer between the phases. The numerical results were obtained by a finite volume method based on the SIMPLE algorithm. They have been verified by an experimental facility using air-water as a simulation pair and a phase Doppler particle analyzer for velocity and droplet size measurement. The numerical simulation of a lithium-cesium high-temperature pair showed that a nearly homogeneous and isothermal expansion of the two phases is possible with small pressure losses and high kinetic efficiencies. In the throat region a careful profiling is necessary to reduce the inertial effects on the liquid velocity field

Zero-G two phase flow regime modeling in adiabatic flow

International Nuclear Information System (INIS)

Reinarts, T.R.; Best, F.R.; Wheeler, M.; Miller, K.M.

1993-01-01

Two-phase flow, thermal management systems are currently being considered as an alternative to conventional, single phase systems for future space missions because of their potential to reduce overall system mass, size, and pumping power requirements. Knowledge of flow regime transitions, heat transfer characteristics, and pressure drop correlations is necessary to design and develop two-phase systems. This work is concerned with microgravity, two-phase flow regime analysis. The data come from a recent sets of experiments. The experiments were funded by NASA Johnson Space Center (JSC) and conducted by NASA JSC with Texas A ampersand M University. The experiment was on loan to NASA JSC from Foster-Miller, Inc., who constructed it with funding from the Air Force Phillips Laboratory. The experiment used R12 as the working fluid. A Foster-Miller two phase pump was used to circulate the two phase mixture and allow separate measurements of the vapor and liquid flow streams. The experimental package was flown 19 times for 577 parabolas aboard the NASA KC-135 aircraft which simulates zero-G conditions by its parabolic flight trajectory. Test conditions included bubbly, slug and annular flow regimes in 0-G. The superficial velocities of liquid and vapor have been obtained from the measured flow rates and are presented along with the observed flow regimes and several flow regime transition predictions. None of the predictions completely describe the transitions as indicated by the data

List of benchmarks for simulation tools of steam-water two-phase flows

International Nuclear Information System (INIS)

Mimouni, S.; Serre, G.

2001-01-01

A physical-numerical benchmarks matrix was drawn up in the context of the ECUME co-development action. Its purpose is to test the different potentialities required for the numerical methods to be used in the codes of the future which will benefit from advanced physics simulations. This benchmarks matrix is to be used for each numerical method in order to answer the following questions: What is the two-phase flow field that the combination of physics model + numerical scheme can process? What is the accuracy of the scheme for each type of physics situation? What is the numerical efficiency (computing time) of the numerical scheme for each type of physics situation? (author)

List of benchmarks for simulation tools of steam-water two-phase flows

Energy Technology Data Exchange (ETDEWEB)

Mimouni, S. [Electricite de France (EDF), Div. R and D, 78 - Chatou (France); Serre, G. [CEA Grenoble, Dept. de Thermohydraulique et de Physique, DTP, 38 (France)

2001-07-01

A physical-numerical benchmarks matrix was drawn up in the context of the ECUME co-development action. Its purpose is to test the different potentialities required for the numerical methods to be used in the codes of the future which will benefit from advanced physics simulations. This benchmarks matrix is to be used for each numerical method in order to answer the following questions: What is the two-phase flow field that the combination of physics model + numerical scheme can process? What is the accuracy of the scheme for each type of physics situation? What is the numerical efficiency (computing time) of the numerical scheme for each type of physics situation? (author)

Flow patterns in vertical two-phase flow

International Nuclear Information System (INIS)

McQuillan, K.W.; Whalley, P.B.

1985-01-01

This paper is concerned with the flow patterns which occur in upwards gas-liquid two-phase flow in vertical tubes. The basic flow patterns are described and the use of flow patter maps is discussed. The transition between plug flow and churn flow is modelled under the assumption that flooding of the falling liquid film limits the stability of plug flow. The resulting equation is combined with other flow pattern transition equations to produce theoretical flow pattern maps, which are then tested against experimental flow pattern data. Encouraging agreement is obtained

1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Kim, Jae Cheol

2007-02-01

During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

The role of two-phase coolant in moderating fretting in nuclear steam generators

International Nuclear Information System (INIS)

Dyke, J.M.

2004-01-01

This paper expands the principal of coolant-cushioning in Nuclear Steam Generators whereby the two-phase coolant, especially the bubble film on the tube surface, moderates the vibration of coolant tubes against their supports. The current paper addresses tube bundle and anti-vibration bars (AVB) geometry issues; examines the tube bundle-coolant-AVB interfaces and examines implications for recirculation flow, AVB design and boiler size. In a T(sat) fluid, the tube surface is uniformly coating with growing bubbles whose momentum is perpendicular to the surface at first, then they are swept away by the bulk flow. The combination of this momentum, the phase change and the water film remaining on the surface, counteract the vibration energy of the tube-AVB system, reducing the likelihood of metal-to-metal contact and consequent fretting. To maximize the benefit of the cushioning effect, the following design inputs are needed: 1) the AVB-tube interface should have sufficient clearance for the T(sat) solution to operate, 2) The AVB should be wide enough to generate the necessary cushioning force, and 3) the AVB should be thin enough to be flexible and absorb some of the transferred vibration energy. Furthermore, fretting and crude deposition at the AVB-tube interface can be reduced or eliminated by reducing the number of AVBs, increasing clearances and making the AVBs limber

Two-phase-flow models and their limitations

International Nuclear Information System (INIS)

Ishii, M.; Kocamustafaogullari, G.

1982-01-01

An accurate prediction of transient two-phase flow is essential to safety analyses of nuclear reactors under accident conditions. The fluid flow and heat transfer encountered are often extremely complex due to the reactor geometry and occurrence of transient two-phase flow. Recently considerable progresses in understanding and predicting these phenomena have been made by a combination of rigorous model development, advanced computational techniques, and a number of small and large scale supporting experiments. In view of their essential importance, the foundation of various two-phase-flow models and their limitations are discussed in this paper

Droplet phase characteristics in liquid-dominated steam--water nozzle flow

International Nuclear Information System (INIS)

Alger, T.W.

1978-01-01

An experimental study was undertaken to determine the droplet size distribution, the droplet spatial distribution and the mean droplet velocity in low-quality, steam-water flow from a rectangular cross-section, converging-diverging nozzle. A unique forward light scattering technique was developed for droplet size distribution measurements. Droplet spatial variations were investigated using light transmission measurements, and droplet velocities were measured with a laser-Doppler velocimeter (LDV) system incorporating a confocal Fabry-Perot interferometer. Nozzle throat radius of curvature and height were varied to investigte their effects on droplet size. Droplet size distribution measurements yielded a nominal Sauter mean droplet diameter of 1.7 μm and a nominal mass-mean droplet diameter of 2.4 μm. Neither the throat radius of curvature nor the throat height were found to have a significant effect upon the nozzle exit droplet size. The light transmission and LDV measurement results confirmed both the droplet size measurements and demonstrated high spatial uniformity of the droplet phase within the nozzle jet flow. One-dimensional numerical calculations indicated that both the dynamic breakup (thermal equilibrium based on a critical Weber number of 6.0) and the boiling breakup (thermal nonequilibrium based on average droplet temperature) models predicted droplet diameters on the order of 7.5 μm, which are approximately equal to the maximum stable droplet diameters within the nozzle jet flow

Numerical fluid dynamics calculations of nonequilibrium steam-water flows with entrained droplets

International Nuclear Information System (INIS)

Williams, K.A.

1984-01-01

The present work has developed a computational fluid dynamics formulation that efficiently solves the conservation laws for a vapor field, a continuous liquid field, and two dispersed droplet fields. The thermal-hydraulic effects resulting from the exchange of mass, momentum and energy between the vapor and the dispersed droplet phases has been accurately modeled. This work is an advancement of the state-of-the-art for engineering analyses of nonequilibrium steam-water-droplet flows in heated channels. It is particularly applicable for boiling steam-water flows in which it is important to represent the effects of significant thermal nonequilibrium between the vapor and the liquid phases. This work was shown to be in good agreement with unique experimental measurements of significant thermal nonequilibrium between the vapor and dispersed droplets. The tests analyzed covered a range of mass fluxes and wall heating rates, and were all at low pressures where nonequilibrium effects are most pronounced

Fluid-elastic vibration in two-phase cross flow

International Nuclear Information System (INIS)

Sasakawa, T.; Serizawa, A.; Kawara, Z.

2003-01-01

The present work aims at clarifying the mechanisms of fluid elastic vibration of tube bundles in two-phase cross flow. The experiment is conducted using air-water two-phase flow under atmospheric pressure. The test section is a 1.03m long transparent acrylic square duct with 128 x 128 mm 2 cross section, which consists of 3 rod-rows with 5 rods in each row. The rods are 125mm long aluminum rods with 22 mm in diameter (p/D=1.45). The natural frequency of rod vibration is about 30Hz. The result indicated a diversion of observed trend in vibration behavior depending on two-phase flow patterns either bubbly flow or churn flow. Specifically, in churn flow, the fluid elastic vibration has been observed to occur when the frequency in void fraction fluctuation approached to the natural frequency of the rods, but this was not the case in fluid elastic vibration in bubbly flow. This fact suggests the existence of mechanisms closely coupled with two-phase flow structures depending on the flow patterns, that is, static two-phase character-controlled mechanism in bubbly flow and dynamic character- controlled in churn flow

Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

International Nuclear Information System (INIS)

Tingkuan, C.; Xuanzheng, C.

1987-01-01

It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

Equivalence of two models in single-phase multicomponent flow simulations

KAUST Repository

Wu, Yuanqing

2016-02-28

In this work, two models to simulate the single-phase multicomponent flow in reservoirs are introduced: single-phase multicomponent flow model and two-phase compositional flow model. Because the single-phase multicomponent flow is a special case of the two-phase compositional flow, the two-phase compositional flow model can also simulate the case. We compare and analyze the two models when simulating the single-phase multicomponent flow, and then demonstrate the equivalence of the two models mathematically. An experiment is also carried out to verify the equivalence of the two models.

Equivalence of two models in single-phase multicomponent flow simulations

KAUST Repository

Wu, Yuanqing; Sun, Shuyu

2016-01-01

In this work, two models to simulate the single-phase multicomponent flow in reservoirs are introduced: single-phase multicomponent flow model and two-phase compositional flow model. Because the single-phase multicomponent flow is a special case of the two-phase compositional flow, the two-phase compositional flow model can also simulate the case. We compare and analyze the two models when simulating the single-phase multicomponent flow, and then demonstrate the equivalence of the two models mathematically. An experiment is also carried out to verify the equivalence of the two models.

Investigation of grid-enhanced two-phase convective heat transfer in the dispersed flow film boiling regime

International Nuclear Information System (INIS)

Miller, D.J.; Cheung, F.B.; Bajorek, S.M.

2013-01-01

Highlights: • Experiments were done in the RBHT facility to study the droplet flow in rod bundle. • The presence of a droplet field was found to greatly enhance heat transfer. • A second-stage augmentation was observed downstream of a spacer grid. • This augmentation is due to the breakup of liquid ligaments downstream of the grid. - Abstract: A two-phase dispersed droplet flow investigation of the grid-enhanced heat transfer augmentation has been done using steam cooling with droplet injection experimental data obtained from the Penn State/NRC Rod Bundle Heat Transfer (RBHT) facility. The RBHT facility is a vertical, full length, 7 × 7-rod bundle heat transfer facility having 45 electrically heated fuel rod simulators of 9.5 mm (0.374-in.) diameter on a 12.6 mm (0.496-in.) pitch which simulates a portion of a PWR fuel assembly. The facility operates at low pressure, up to 4 bars (60 psia) and has over 500 channels of instrumentation including heater rod thermocouples, spacer grid thermocouples, closely-spaced differential pressure cells along the test section, several fluid temperature measurements within the rod bundle flow area, inlet and exit flows, absolute pressure, and the bundle power. A series of carefully controlled and well instrumented steam cooling with droplet injection experiments were performed over a range of Reynolds numbers and droplet injection flow rates. The experimental results were analyzed to obtain the axial variation of the local heat transfer coefficients along the rod bundle. At the spacer grid location, the flow was found to be substantially disrupted, with the hydrodynamic and thermal boundary layers undergoing redevelopment. Owing to this flow restructuring, the heat transfer downstream of a grid spacer was found to be augmented above the fully developed flow heat transfer as a result of flow disruption induced by the grid. Furthermore, the presence of a droplet field further enhanced the heat transfer as compared to single

Transition from annular flow to plug/slug flow in condensation of steam in microchannels

Energy Technology Data Exchange (ETDEWEB)

Quan, Xiaojun; Cheng, Ping; Wu, Huiying [School of Mechanical and Power Engineering, Shanghai Jiaotong University, 800 Dong Chuan Road, Shanghai 200240 (China)

2008-02-15

A visualization study has been conducted to investigate the transition from annular flow to plug/slug flow in the condensation of steam in two different sets of parallel microchannels, having hydraulic diameters of 90 {mu}m and 136 {mu}m, respectively. The steam in the parallel microchannels was cooled on the bottom by forced convection of water and by natural convection of air from the top. It is found that the location, where the transition from annular flow to plug/slug flow takes place, depends on mass flux and cooling rate of steam. The effects of mass flux and cooling rate on the occurrence frequency of the injection flow in a single microchannel, having a hydraulic diameter of 120 {mu}m and 128 {mu}m, respectively, are investigated. It is found that two different shapes of injection flow occur in the smooth annular flow in microchannels: injection flow with unsteady vapor ligament occurring at low mass flux (or high cooling rate) and injection flow with steady vapor ligament occurring at high mass flux (or low cooling rate). It is also found that increase of steam mass flux, decrease of cooling rate, or decrease of the microchannel diameter tends to enhance instability of the condensate film on the wall, resulting in occurrence of the injection flow further toward the outlet with an increase in occurrence frequency. (author)

The effect of vortex mixing grids on the behaviour of steam phase in fuel assemblies

International Nuclear Information System (INIS)

Sergeev, V.V.; Fedotovsky, V.S.; Shcherbakov, S. I.

2013-01-01

The paper examines the behavior of the steam phase in two-phase flow in the space between fuel elements of a transparent model of WWER FA with “Vortex”-type mixing grids. The model was a real subchannel scaled-up in the ratio of 5 to 1 and allowed for gas feeding through the walls of displacers to simulate boiling on fuel surface. Hydraulic resistances of intensifier simulators were determined during experiments and bubble paths were photographed at different velocities of the water flow. Experimental results made it possible to verify calculation models developed at SSC RF - IPPE for fuel assemblies with “Vortex”-type mixing grids. These models let calculate the effect of steam removal from fuel surface. (authors)

POST: a postprocessor computer code for producing three-dimensional movies of two-phase flow in a reactor vessel

International Nuclear Information System (INIS)

Taggart, K.A.; Liles, D.R.

1977-08-01

The development of the TRAC computer code for analysis of LOCAs in light-water reactors involves the use of a three-dimensional (r-theta-z), two-fluid hydrodynamics model to describe the two-phase flow of steam and water through the reactor vessel. One of the major problems involved in interpreting results from this code is the presentation of three-dimensional flow patterns. The purpose of the report is to present a partial solution to this data display problem. A first version of a code which produces three-dimensional movies of flow in the reactor vessel has been written and debugged. This code (POST) is used as a postprocessor in conjunction with a stand alone three-dimensional two-phase hydrodynamics code (CYLTF) which is a test bed for the three-dimensional algorithms to be used in TRAC

Analysis of phase dynamics in two-phase flow using latticegas automata

International Nuclear Information System (INIS)

Ohashi, H.; Hashimoto, Y.; Tsumaya, A.; Chen, Y.; Akiyama, M.

1998-01-01

In this paper, we describe lattice gas automaton models appropriate for two-phase flow simulation and their applications to study various phase dynamics of two-fluid mixtures. Several algorithms are added to the original immiscible Lattice Gas model to adjust surface tension and to introduce density difference between two fluids. Surface tension is controlled by the collision rules an difference in density is due to nonlocal forces between automaton particles. We simulate the relative motion of the dispersed phase in another continuous fluid. Deformation and disintegration of rising drops are reproduced. The interaction between multiple drops is also observed in calculations. Furutre, we obtain the transition of the two-phase flow pattern from bubbly, slug to annular flow. Density difference of two phase is one of the key ingredients to generate the annular flow pattern

CFD Modeling of Wall Steam Condensation: Two-Phase Flow Approach versus Homogeneous Flow Approach

International Nuclear Information System (INIS)

Mimouni, S.; Mechitoua, N.; Foissac, A.; Hassanaly, M.; Ouraou, M.

2011-01-01

The present work is focused on the condensation heat transfer that plays a dominant role in many accident scenarios postulated to occur in the containment of nuclear reactors. The study compares a general multiphase approach implemented in NEPTUNE C FD with a homogeneous model, of widespread use for engineering studies, implemented in Code S aturne. The model implemented in NEPTUNE C FD assumes that liquid droplets form along the wall within nucleation sites. Vapor condensation on droplets makes them grow. Once the droplet diameter reaches a critical value, gravitational forces compensate surface tension force and then droplets slide over the wall and form a liquid film. This approach allows taking into account simultaneously the mechanical drift between the droplet and the gas, the heat and mass transfer on droplets in the core of the flow and the condensation/evaporation phenomena on the walls. As concern the homogeneous approach, the motion of the liquid film due to the gravitational forces is neglected, as well as the volume occupied by the liquid. Both condensation models and compressible procedures are validated and compared to experimental data provided by the TOSQAN ISP47 experiment (IRSN Saclay). Computational results compare favorably with experimental data, particularly for the Helium and steam volume fractions.

Two-phase flow

International Nuclear Information System (INIS)

Olive, J.

1990-01-01

The design, operation and safety of nuclear components requires increasingly accurate knowledge of two-phase flows. This knowledge is also necessary for some studies related to electricity applications. The author presents some concrete examples showing the range of problems and the complexity of the phenomena involved in these types of flows. Then, the basic principles of their numerical modelling are explained, as well as the new tendency to use increasingly local and refined models. The newest computer codes developed at EDF are briefly presented. Experimental studies dealing with twophase flow are also referred to, and their connections to numerical modelling are explained. Emphasis is placed on the major efforts devoted to the development of new test rigs and instrumentation [fr

Experimental study of centrifugal pump performance under steam-water two-phase flow conditions at elevated pressures

International Nuclear Information System (INIS)

Chan, A.M.C.; Barreca, S.L.; Hartlen, R.T.

1991-01-01

The performance of a centrifugal pump under two-phase flow conditions was studied in a closed loop. System voids of increasing magnitude were attained by draining water from the loop in steps. The operating temperature/pressure were varied from 110 degrees C/0.15 MPa to 260 degrees C/4.7 MPa. Only tests in the first quadrant were conducted. In this paper the head-flow characteristics and pump head degradation data are presented and discussed

Fluid dynamics of cryogenic two-phase flows

International Nuclear Information System (INIS)

Verfondern, K.; Jahn, W.

2004-01-01

The objective of this study was to examine the flow behavior of a methane hydrate/methane-liquid hydrogen dispersed two-phase fluid through a given design of a moderator chamber for the ESS target system. The calculations under simplified conditions, e.g., taking no account of heat input from outside, have shown that the computer code used, CFX, was able to simulate the behavior of the two-phase flow through the moderator chamber, producing reasonable results up to a certain level of the solid phase fraction, that allowed a continuous flow process through the chamber. Inlet flows with larger solid phase fractions than 40 vol% were found to be a ''problem'' for the computer code. From the computer runs based on fractions between 20 and 40 vol%, it was observed that with increasing solid phase fraction at the inlet, the resulting flow pattern revealed a strong tendency for blockage within the chamber, supported by the ''heavy weight'' of the pellets compared to the carrying liquid. Locations which are prone to the development of such uneven flow behavior are the areas around the turning points in the semispheres and near the exit of the moderator. The considered moderator chamber with horizontal inlet and outlet flow for a solid-liquid two-phase fluid does not seem to be an appropriate design. (orig.)

A fast response miniature probe for wet steam flow field measurements

International Nuclear Information System (INIS)

Bosdas, Ilias; Mansour, Michel; Abhari, Reza S; Kalfas, Anestis I

2016-01-01

Modern steam turbines require operational flexibility due to renewable energies’ increasing share of the electrical grid. Additionally, the continuous increase in energy demand necessitates efficient design of the steam turbines as well as power output augmentation. The long turbine rotor blades at the machines’ last stages are prone to mechanical vibrations and as a consequence time-resolved experimental data under wet steam conditions are essential for the development of large-scale low-pressure steam turbines. This paper presents a novel fast response miniature heated probe for unsteady wet steam flow field measurements. The probe has a tip diameter of 2.5 mm, and a miniature heater cartridge ensures uncontaminated pressure taps from condensed water. The probe is capable of providing the unsteady flow angles, total and static pressure as well as the flow Mach number. The operating principle and calibration procedure are described in the current work and a detailed uncertainty analysis demonstrates the capability of the new probe to perform accurate flow field measurements under wet steam conditions. In order to exclude any data possibly corrupted by droplets’ impact or evaporation from the heating process, a filtering algorithm was developed and implemented in the post-processing phase of the measured data. In the last part of this paper the probe is used in an experimental steam turbine test facility and measurements are conducted at the inlet and exit of the last stage with an average wetness mass fraction of 8.0%. (paper)

Switching moving boundary models for two-phase flow evaporators and condensers

Science.gov (United States)

Bonilla, Javier; Dormido, Sebastián; Cellier, François E.

2015-03-01

The moving boundary method is an appealing approach for the design, testing and validation of advanced control schemes for evaporators and condensers. When it comes to advanced control strategies, not only accurate but fast dynamic models are required. Moving boundary models are fast low-order dynamic models, and they can describe the dynamic behavior with high accuracy. This paper presents a mathematical formulation based on physical principles for two-phase flow moving boundary evaporator and condenser models which support dynamic switching between all possible flow configurations. The models were implemented in a library using the equation-based object-oriented Modelica language. Several integrity tests in steady-state and transient predictions together with stability tests verified the models. Experimental data from a direct steam generation parabolic-trough solar thermal power plant is used to validate and compare the developed moving boundary models against finite volume models.

French steam generator design developments

International Nuclear Information System (INIS)

Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

1986-01-01

From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

Apparatus for monitoring two-phase flow

Science.gov (United States)

Sheppard, John D.; Tong, Long S.

1977-03-01

A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

Two-phase flow characterisation by nuclear magnetic resonance

International Nuclear Information System (INIS)

Leblond, J.; Javelot, S.; Lebrun, D.; Lebon, L.

1998-01-01

The results presented in this paper demonstrate the performance of the PFGSE-NMR to obtain a complete characterisation of two-phase flows. Different methods are proposed to characterise air-water flows in different regimes: stationary two-phase flows and flows in transient condition. Finally a modified PFGSE is proposed to analyse the turbulence of air-water bubbly flow. (author)

Effect of steam quality on two—phase flow in a netural circulation loop

Institute of Scientific and Technical Information of China (English)

贾海军; 吴少融; 等

1996-01-01

Test pressures are 1.0-4.0MPa,heating powers 27-190kW,inlet subcoolings 5-80℃,water used as coolant,and steam quality at the outlet of test section is less than 0.05,These test conditions cover the parameters for a typical 200MW heating reactor.The experimental results show that the stema quality is the dominant factor in a natural circulation system with low pressure and low steam quality about the effect of system pressure,heating power and inlet subcooling on the flow rate,relative oscilatroy amplitude and oscilatory region of flow rate.

Numerical method for two-phase flow discontinuity propagation calculation

International Nuclear Information System (INIS)

Toumi, I.; Raymond, P.

1989-01-01

In this paper, we present a class of numerical shock-capturing schemes for hyperbolic systems of conservation laws modelling two-phase flow. First, we solve the Riemann problem for a two-phase flow with unequal velocities. Then, we construct two approximate Riemann solvers: an one intermediate-state Riemann solver and a generalized Roe's approximate Riemann solver. We give some numerical results for one-dimensional shock-tube problems and for a standard two-phase flow heat addition problem involving two-phase flow instabilities

Numerical method for solution of transient, homogeneous, equilibrium, two-phase flows in one space dimension

International Nuclear Information System (INIS)

Shin, Y.W.; Wiedermann, A.H.

1979-10-01

A solution method is presented for transient, homogeneous, equilibrium, two-phase flows of a single-component fluid in one space dimension. The method combines a direct finite-difference procedure and the method of characteristics. The finite-difference procedure solves the interior points of the computing domain; the boundary information is provided by a separate procedure based on the characteristics theory. The solution procedure for boundary points requires information in addition to the physical boundary conditions. This additional information is obtained by a new procedure involving integration of characteristics in the hodograph plane. Sample problems involving various combinations of basic boundary types are calculated for two-phase water/steam mixtures and single-phase nitrogen gas, and compared with independent method-of-characteristics solutions using very fine characteristic mesh. In all cases, excellent agreement is demonstrated

Analysis of water hammer in two-component two-phase flows

International Nuclear Information System (INIS)

Warde, H.; Marzouk, E.; Ibrahim, S.

1989-01-01

The water hammer phenomena caused by a sudden valve closure in air-water two-phase flows must be clarified for the safety analysis of LOCA in reactors and further for the safety of boilers, chemical plants, pipe transport of fluids such as petroleum and natural gas. In the present work water hammer phenomena caused by sudden valve closure in two-component two-phase flows are investigated theoretically and experimentally. The phenomena are more complicated than in single phase-flows due to the fact of the presence of compressible component. Basic partial differential equations based on a one-dimensional homogeneous flow model are solved by the method of characteristic. The analysis is extended to include friction in a two-phase mixture depending on the local flow pattern. The profiles of the pressure transients, the propagation velocity of pressure waves and the effect of valve closure on the transient pressure are found. Different two-phase flow pattern and frictional pressure drop correlations were used including Baker, Chesholm and Beggs and Bril correlations. The effect of the flow pattern on the characteristic of wave propagation is discussed primarily to indicate the effect of void fraction on the velocity of wave propagation and on the attenuation of pressure waves. Transient pressure in the mixture were recorded at different air void fractions, rates of uniform valve closure and liquid flow velocities with the aid of pressure transducers, transient wave form recorders interfaced with an on-line pc computer. The results are compared with computation, and good agreement was obtained within experimental accuracy

Measurement of Two Phase Flow

Directory of Open Access Journals (Sweden)

J. Novotný

2005-01-01

Full Text Available This paper presents the results of experiments with moist wet steam. The aim of the experiment was to measure the velocity of the growth of a condensing nucleus in wet steam dependent on the velocity of condensation. For the experiments in wet steam an experimental setup was designed and constructed, which generated superheated steam at lowered pressure and a temperature of 50 °C. Low pressure and temperature of the hot vapour was chosen in order to minimize the risk of accidental disruption of the wall. The size of the condensing nucleus was measured by the method of Interferometric Particle Imaging (IPI. The IPI method is a technique for determining the particle size of transparent and spherical particles based on calculating the fringes captured on a CCD array. The number of fringes depends on the particle size and on the optical configuration. The experimental setup used is identical with the setup for measuring flow by the stereo PIV method. The only difference is the use of a special camera mount comprising a transparent mirror and enabling both cameras to be focused to one point. We present the results of the development of the growth of a condensing nucleus and histograms of the sizes of all measured particles depending on position and condensation velocity. 

Assessment of theoretical flow pattern maps for vertical upward two-phase flow

International Nuclear Information System (INIS)

Khare, Rajesh; Vijayan, P.K.; Saha, D.; Venkat Raj, V.

1997-04-01

Taitel-Dukler (1980), Mishima-Ishii (1984) and Solbrig (1986) flow pattern maps have been assessed against an experimental data bank compiled from different sources. The data bank consisted of a total of 1411 data points with 368 bubbly, 474 slug/churn and 545 annular flow points, the rest being transition points. The data bank consisted of mainly steam water data; some amount of air-water data are included as there were no steam-water data at low pressure ( gs - U ls plane. (author)

Apparatus for monitoring two-phase flow

International Nuclear Information System (INIS)

Sheppard, J.D.; Tong, L.S.

1977-01-01

A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods. 3 claims, 9 figures

Two-phase flow through small branches in a horizontal pipe with stratified flow

International Nuclear Information System (INIS)

Smoglie, C.

1984-12-01

This report presents the description and results of experiments designed to determine the mass flow rate and quality through a small break at the bottom, the top or the side of a main pipe with stratified gas-liquid flow. If the interface level is far below (above) the branch, only single-phase gas (liquid) flow enters the branch. For smaller distances the interface is locally deformed because of the pressure decrease due to the fluid acceleration near the branch inlet (Bernoulli effect) and liquid (gas) can be entrained. This report contains photographs illustrating the flow phenomena as well as a general correlation to determine the beginning of entrainment. Results are presented on the branch mass flow rate and quality as a function of a normalized distance between the interface and the branch inlet. A model was developed which enables to predict the branch quality and mass flux. Results from air-water flow through horizontal branches, were extrapolated for steam water flow at high pressure with critical branch mass flux. (orig./HP) [de

Calculations of the nozzle coefficient of discharge of wet steam turbine stages

International Nuclear Information System (INIS)

Jinling, Z.; Yinian, C.

1989-01-01

A method is presented for calculating the coefficient of discharge of wet steam turbine nozzles. The theoretical formulation of the problem is rigorously in accordance with the theory of two-phase wet steam expansion flow through steam turbine nozzles. The computational values are plotted as sets of curves in accordance with orthogonality test principles. They agree satisfactorily both with historical empirical data and the most recent experimental data obtained in the wet steam two-phase flow laboratory of Xian Jiaotong University. (author)

Two-phase flow patterns and their relationship to two-phase heat transfer

International Nuclear Information System (INIS)

Hewitt, G.F.

1977-01-01

The objective of this lecture was to discuss the general nature of two phase flows, to define the various regimes of flow and to discuss the influence of these regimes on the heat transfer processes taking place. The methods of regime delineation are briefly described and regime descriptions introduced for both vertical and horizontal flows in tubes. ''Flow regime maps'' have been widely used as an aid to determination of the regime which occurs in a given situation. Some of the more widely used maps are described and the limitations of this approach discussed. There have been many attempts to obtain a better phenomenological description of two phase flow patterns. In this lecture, these attempts will be reviewed in the context of the bubble/plug, plug/churn and churn/annular flow transitions in vertical flow. The latter two transitions are related to the flooding/flow reversal phenomena. For horizontal flows, recent work on the onset of slugging will be reviewed. In flows with evaporation or condensation, the situation is influenced by departures from thermodynamic equilibrium and the types of departure observed are discuss briefly. Flow patterns and their relationships with heat transfer regimes are then reviewed for the case of condensation in horizontal tubes and evaporation in vertical tubes

CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

Energy Technology Data Exchange (ETDEWEB)

Liu, T. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2009-07-01

CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

A CFD Analysis of Steam Flow in the Two-Stage Experimental Impulse Turbine with the Drum Rotor Arrangement

Directory of Open Access Journals (Sweden)

Yun Kukchol

2016-01-01

Full Text Available The aim of the paper is to present the CFD analysis of the steam flow in the two-stage turbine with a drum rotor and balancing slots. The balancing slot is a part of every rotor blade and it can be used in the same way as balancing holes on the classical rotor disc. The main attention is focused on the explanation of the experimental knowledge about the impact of the slot covering and uncovering on the efficiency of the individual stages and the entire turbine. The pressure and temperature fields and the mass steam flows through the shaft seals, slots and blade cascades are calculated. The impact of the balancing slots covering or uncovering on the reaction and velocity conditions in the stages is evaluated according to the pressure and temperature fields. We have also concentrated on the analysis of the seal steam flow through the balancing slots. The optimized design of the balancing slots has been suggested.

Horizontal liquid film-mist two phase flow, (1)

International Nuclear Information System (INIS)

Akagawa, Koji; Sakaguchi, Tadashi; Fujii, Terushige; Nakatani, Yoji; Nakaseko, Kosaburo.

1979-01-01

The characteristics of liquid film in annular spray flow, the generation of droplets from liquid film and the transport of droplets to a wall are the important matters in the planning and design of nuclear reactor cooling system and the channels of steam generators. The study on the liquid film spray flow is scarce, and its characteristics are not yet elucidated. The purpose of this series of studies is to clarify the characteristics of liquid film, the generation, diffusion and distribution of droplets and pressure loss in the liquid film spray flow composed of the liquid film on the lower wall and spraying gas flow in a rectangular, horizontal channel. In this paper, the concentration distribution and the diffusion coefficient of droplets on a cross section in the region of flow completion are reported. The experimental apparatuses and the experimental method, the flow rate of droplets and the velocity distribution of gas phase, the concentration distribution and the diffusion coefficient of droplets, and the diameter of generated droplets are explained. The equation for the concentration distribution of droplets using dimensionless characteristic value was derived. The mean diffusion coefficient of droplets was constant on a cross section, and the effects of gravity and turbulent diffusion can be evaluated. (Kako, I.)

Ferrofluid-in-oil two-phase flow patterns in a flow-focusing microchannel

Science.gov (United States)

Sheu, T. S.; Chen, Y. T.; Lih, F. L.; Miao, J. M.

This study investigates the two-phase flow formation process of water-based Fe3O4 ferrofluid (dispersed phase) in a silicon oil (continuous phase) flow in the microfluidic flow-focusing microchannel under various operational conditions. With transparent PDMS chip and optical microscope, four main two-phase flow patterns as droplet flow, slug flow, ring flow and churn flow are observed. The droplet shape, size, and formation mechanism were also investigated under different Ca numbers and intended to find out the empirical relations. The paper marks an original flow pattern map of the ferrofluid-in-oil flows in the microfluidic flow-focusing microchannels. The flow pattern transiting from droplet flow to slug flow appears for an operational conditions of QR < 1 and Lf / W < 1. The power law index that related Lf / W to QR was 0.36 in present device.

DISTRIBUTION OF TWO-PHASE FLOW IN A DISTRIBUTOR

Directory of Open Access Journals (Sweden)

AZRIDJAL AZIZ

2012-02-01

Full Text Available The flow configuration and distribution behavior of two-phase flow in a distributor made of acrylic resin have been investigated experimentally. In this study, air and water were used as two-phase flow working fluids. The distributor consists of one inlet and two outlets, which are set as upper and lower, respectively. The flow visualization at the distributor was made by using a high–speed camera. The flow rates of air and water flowing out from the upper and lower outlet branches were measured. Effects of inclination angle of the distributor were investigated. By changing the inclination angle from vertical to horizontal, uneven distributions were also observed. The distribution of two-phase flow through distributor tends even flow distribution on the vertical position and tends uneven distribution on inclined and horizontal positions. It is shown that even distribution could be achieved at high superficial velocities of both air and water.

A device for pre-separating water-drops in a two-phase flow

International Nuclear Information System (INIS)

Andro, Jean; Peyrelongue, J.-P.

1974-01-01

The invention relates to the mechanical pre-separation of water-drops in suspension in a flow of saturated steam. To this end, the method comprises the steps of carrying out rough separations by directing the flow towards curved surfaces adapted to deflect that flow and to project the drops onto said surfaces, sucking the film formed by the water-drops displaced by centrifugal force on the outer periphery of said surfaces, directing the steam separated from the water-drops onto five separators so as to extract dry steam and discharging the water provided by the sucking of said surfaces and the five separators. The invention applies to the drying of steam issuing from the high-pressure bodies of nuclear steam-turbines [fr

K-FIX: a computer program for transient, two-dimensional, two-fluid flow

International Nuclear Information System (INIS)

Rivard, W.C.; Torrey, M.D.

1976-11-01

The transient dynamics of two-dimensional, two-phase flow with interfacial exchange are calculated at all flow speeds using the K-FIX program. Each phase is described in terms of its own density, velocity, and temperature. The six field equations for the two phases couple through mass, momentum, and energy exchange. The equations are solved using an Eulerian finite difference technique that implicitly couples the rates of phase transitions, momentum, and energy exchange to determination of the pressure, density, and velocity fields. The implicit solution is accomplished iteratively without linearizing the equations, thus eliminating the need for numerous derivative terms. K-FIX is written in a highly modular form to be easily adaptable to a variety of problems. It is applied to growth of an isolated steam bubble in a superheated water pool

Review on two-phase flow instabilities in narrow spaces

International Nuclear Information System (INIS)

Tadrist, L.

2007-01-01

Instabilities in two-phase flow have been studied since the 1950s. These phenomena may appear in power generation and heat transfer systems where two-phase flow is involved. Because of thermal management in small size systems, micro-fluidics plays an important role. Typical processes must be considered when the channel hydraulic diameter becomes very small. In this paper, a brief review of two-phase flow instabilities encountered in channels having hydraulic diameters greater than 10 mm are presented. The main instability types are discussed according to the existing experimental results and models. The second part of the paper examines two-phase flow instabilities in narrow spaces. Pool and flow boiling cases are considered. Experiments as well as theoretical models existing in the literature are examined. It was found that several experimental works evidenced these instabilities meanwhile only limited theoretical developments exist in the literature. In the last part of the paper an interpretation of the two-phase flow instabilities linked to narrow spaces are presented. This approach is based on characteristic time scales of the two-phase flow and bubble growth in the capillaries

Parallel two-phase-flow-induced vibrations in fuel pin model

International Nuclear Information System (INIS)

Hara, Fumio; Yamashita, Tadashi

1978-01-01

This paper reports the experimental results of vibrations of a fuel pin model -herein meaning the essential form of a fuel pin from the standpoint of vibration- in a parallel air-and-water two-phase flow. The essential part of the experimental apparatus consisted of a flat elastic strip made of stainless steel, both ends of which were firmly supported in a circular channel conveying the two-phase fluid. Vibrational strain of the fuel pin model, pressure fluctuation of the two-phase flow and two-phase-flow void signals were measured. Statistical measures such as power spectral density, variance and correlation function were calculated. The authors obtained (1) the relation between variance of vibrational strain and two-phase-flow velocity, (2) the relation between variance of vibrational strain and two-phase-flow pressure fluctuation, (3) frequency characteristics of variance of vibrational strain against the dominant frequency of the two-phase-flow pressure fluctuation, and (4) frequency characteristics of variance of vibrational strain against the dominant frequency of two-phase-flow void signals. The authors conclude that there exist two kinds of excitation mechanisms in vibrations of a fuel pin model inserted in a parallel air-and-water two-phase flow; namely, (1) parametric excitation, which occurs when the fundamental natural frequency of the fuel pin model is related to the dominant travelling frequency of water slugs in the two-phase flow by the ratio 1/2, 1/1, 3/2 and so on; and (2) vibrational resonance, which occurs when the fundamental frequency coincides with the dominant frequency of the two-phase-flow pressure fluctuation. (auth.)

Two-phase flow model with nonequilibrium and critical flow

International Nuclear Information System (INIS)

Sureau, H.; Houdayer, G.

1976-01-01

The model proposed includes the three conservation equations (mass, momentum, energy) applied to the two phase flows and a fourth partial derivative equation which takes into account the nonequilibriums and describes the mass transfer process. With this model, the two phase critical flow tests performed on the Moby-Dick loop (CENG) with several geometries, are interpreted by a unique law. Extrapolations to industrial dimension problems show that geometry and size effects are different from those obtained with earlier models (Zaloudek, Moody, Fauske) [fr

Multiparticle imaging velocimetry measurements in two-phase flow

International Nuclear Information System (INIS)

Hassan, Y.A.

1998-01-01

The experimental flow visualization tool, Particle Image Velocimetry (PIV), is being extended to determine the velocity fields in two and three-dimensional, two-phase fluid flows. In the past few years, the technique has attracted quite a lot of interest. PIV enables fluid velocities across a region of a flow to be measured at a single instant in time in global domain. This instantaneous velocity profile of a given flow field is determined by digitally recording particle (microspheres or bubbles) images within the flow over multiple successive video frames and then conducting flow pattern identification and analysis of the data. This paper presents instantaneous velocity measurements in various two and three- dimensional, two-phase flow situations. (author)

On intermittent flow characteristics of gas–liquid two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Thaker, Jignesh; Banerjee, Jyotirmay, E-mail: jbaner@gmail.com

2016-12-15

Highlights: • Unified correlations for intermittent flow characteristics are developed. • Influence of inflow conditions on intermittent flow characteristics is analysed. • Developed correlations can be used for effective design of piping components. - Abstract: Flow visualisation experiments are reported for intermittent regime of gas–liquid two-phase flow. Intermittent flow characteristics, which include plug/slug frequency, liquid plug/slug velocity, liquid plug/slug length, and plug/slug bubble length are determined by image processing of flow patterns captured at a rate of 1600 frames per second (FPS). Flow characteristics are established as a function of inlet superficial velocity of both the phases (in terms of Re{sub SL} and Re{sub SG}). The experimental results are first validated with the existing correlations for slug flow available in literature. It is observed that the correlations proposed in literature for slug flow do not accurately predict the flow characteristics in the plug flow regime. The differences are clearly highlighted in this paper. Based on the measured database for both plug and slug flow regime, modified correlations for the intermittent flow regime are proposed. The correlations reported in the present paper, which also include plug flow characteristics will aid immensely to the effective design and optimization of operating conditions for safer operation of two-phase flow piping systems.

Homogeneous non-equilibrium two-phase critical flow model

International Nuclear Information System (INIS)

Schroeder, J.J.; Vuxuan, N.

1987-01-01

An important aspect of nuclear and chemical reactor safety is the ability to predict the maximum or critical mass flow rate from a break or leak in a pipe system. At the beginning of such a blowdown, if the stagnation condition of the fluid is subcooled or slightly saturated thermodynamic non-equilibrium exists in the downstream, e.g. the fluid becomes superheated to a degree determined by the liquid pressure. A simplified non-equilibrium model, explained in this report, is valid for rapidly decreasing pressure along the flow path. It presumes that fluid has to be superheated by an amount governed by physical principles before it starts to flash into steam. The flow is assumed to be homogeneous, i.e. the steam and liquid velocities are equal. An adiabatic flow calculation mode (Fanno lines) is employed to evaluate the critical flow rate for long pipes. The model is found to satisfactorily describe critical flow tests. Good agreement is obtained with the large scale Marviken tests as well as with small scale experiments. (orig.)

Inlet effects on vertical-downward air–water two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Qiao, Shouxu; Mena, Daniel; Kim, Seungjin, E-mail: skim@psu.edu

2017-02-15

Highlights: • Inlet effects on two-phase flow parameters in vertical-downward flow are studied. • Flow regimes in the vertical-downward two-phase flow are defined. • Vertical-downward flow regime maps for three inlet configurations are developed. • Frictional pressure loss analysis for three different inlets is performed. • Database of local two-phase flow parameters for each inlet configuration. - Abstract: This paper focuses on investigating the geometric effects of inlets on global and local two-phase flow parameters in vertical-downward air–water two-phase flow. Flow visualization, frictional pressure loss analysis, and local experiments are performed in a test facility constructed from 50.8 mm inner diameter acrylic pipes. Three types of inlets of interest are studied: (1) two-phase flow injector without a flow straightener (Type A), (2) two-phase flow injector with a flow straightener (Type B), and (3) injection through a horizontal-to-vertical-downward 90° vertical elbow (Type C). A detailed flow visualization study is performed to characterize flow regimes including bubbly, slug, churn-turbulent, and annular flow. Flow regime maps for each inlet are developed and compared to identify the effects of each inlet. Frictional pressure loss analysis shows that the Lockhart–Martinelli method is capable of correlating the frictional loss data acquired for Type B and Type C inlets with a coefficient value of C = 25, but additional data may be needed to model the Type A inlet. Local two-phase flow parameters measured by a four-sensor conductivity probe in four bubbly and near bubbly flow conditions are analyzed. It is observed that vertical-downward two-phase flow has a characteristic center-peaked void profile as opposed to a wall-peaked profile as seen in vertical-upward flow. Furthermore, it is shown that the Type A inlet results in the most pronounced center-peaked void fraction profile, due to the coring phenomenon. Type B and Type C inlets

Modeling two-phase flow in PEM fuel cell channels

Energy Technology Data Exchange (ETDEWEB)

Wang, Yun; Basu, Suman; Wang, Chao-Yang [Electrochemical Engine Center (ECEC), and Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States)

2008-05-01

This paper is concerned with the simultaneous flow of liquid water and gaseous reactants in mini-channels of a proton exchange membrane (PEM) fuel cell. Envisaging the mini-channels as structured and ordered porous media, we develop a continuum model of two-phase channel flow based on two-phase Darcy's law and the M{sup 2} formalism, which allow estimate of the parameters key to fuel cell operation such as overall pressure drop and liquid saturation profiles along the axial flow direction. Analytical solutions of liquid water saturation and species concentrations along the channel are derived to explore the dependences of these physical variables vital to cell performance on operating parameters such as flow stoichiometric ratio and relative humility. The two-phase channel model is further implemented for three-dimensional numerical simulations of two-phase, multi-component transport in a single fuel-cell channel. Three issues critical to optimizing channel design and mitigating channel flooding in PEM fuel cells are fully discussed: liquid water buildup towards the fuel cell outlet, saturation spike in the vicinity of flow cross-sectional heterogeneity, and two-phase pressure drop. Both the two-phase model and analytical solutions presented in this paper may be applicable to more general two-phase flow phenomena through mini- and micro-channels. (author)

Mathematical modeling of disperse two-phase flows

CERN Document Server

Morel, Christophe

2015-01-01

This book develops the theoretical foundations of disperse two-phase flows, which are characterized by the existence of bubbles, droplets or solid particles finely dispersed in a carrier fluid, which can be a liquid or a gas. Chapters clarify many difficult subjects, including modeling of the interfacial area concentration. Basic knowledge of the subjects treated in this book is essential to practitioners of Computational Fluid Dynamics for two-phase flows in a variety of industrial and environmental settings. The author provides a complete derivation of the basic equations, followed by more advanced subjects like turbulence equations for the two phases (continuous and disperse) and multi-size particulate flow modeling. As well as theoretical material, readers will discover chapters concerned with closure relations and numerical issues. Many physical models are presented, covering key subjects including heat and mass transfers between phases, interfacial forces and fluid particles coalescence and breakup, a...

Measurement of transient two-phase flow velocity using statistical signal analysis of impedance probe signals

International Nuclear Information System (INIS)

Leavell, W.H.; Mullens, J.A.

1981-01-01

A computational algorithm has been developed to measure transient, phase-interface velocity in two-phase, steam-water systems. The algorithm will be used to measure the transient velocity of steam-water mixture during simulated PWR reflood experiments. By utilizing signals produced by two, spatially separated impedance probes immersed in a two-phase mixture, the algorithm computes the average transit time of mixture fluctuations moving between the two probes. This transit time is computed by first, measuring the phase shift between the two probe signals after transformation to the frequency domain and then computing the phase shift slope by a weighted least-squares fitting technique. Our algorithm, which has been tested with both simulated and real data, is able to accurately track velocity transients as fast as 4 m/s/s

The Condensation effect on the two-phase flow stability

International Nuclear Information System (INIS)

Abdou Mohamed, Hesham Nagah

2005-01-01

considering riser condensation and of correcting the localized friction due to the presence of the two-phase mixture in the two-phase region.These effects are more important for high heating power and high inlet subcooling. CAREM 25 nuclear power reactor is investigated to get the stability boundary map. The flow instability regions are appeared at low and high core power. In the low heat flux range, the trends of the thermal equilibrium - equal velocity (homogeneous) model and the thermal non equilibrium - non equal velocity model are the same because the steam quality is small.In the high heat flux range, for the subcooled boiling number and the phase change number, the marginal stability boundaries are crossed in a point, determining tow different regions, of high and low inlet subcooling.For the first region, the steam quality calculation of the first model is greater and has the effect of stabilizing the system more than the second one.For the second region, the two-phase region length calculation of the first model is smaller and has the effect of stabilizing the system less than the second one. In general, the model predicts a more stable system with an increase in inlet restriction or riser condensation or system pressure or a decrease in exit restriction [es

Magnetic liquid metal two-phase flow research. Phase 1. Final report

International Nuclear Information System (INIS)

Graves, R.D.

1983-04-01

The Phase I research demonstrates the feasibility of the magnetic liquid metal (MLM) two-phase flow concept. A dispersion analysis is presented based on a complete set of two-phase-flow equations augmented to include stresses due to magnetic polarization of the fluid. The analysis shows that the stability of the MLM two-phase flow is determined by the magnetic Mach number, the slip ratio, geometry of the flow relative to the applied magnetic field, and by the voidage dependence of the interfacial forces. Results of a set of experiments concerned with magnetic effects on the dynamics of single bubble motion in an aqueous-based, viscous, conducting magnetic fluid are presented. Predictions in the theoretical literature are qualitatively verified using a bench-top experimental apparatus. In particular, applied magnetic fields are seen to lead to reduced bubble size at fixed generating orifice pressure

Two-phase flow through small branches in a horizontal pipe with stratified flow

International Nuclear Information System (INIS)

Smoglie, C.

1985-02-01

In the field of reactor safety the occurrence of a small break in a horizontal primary coolant pipe is of great importance. This report presents the description and results of experiments designed to determine the mass flow rate and quality through a small break at the bottom, the top or the side of a main pipe with stratified gas-liquid flow. If the interface level is far below (above) the branch, only single-phase gas (liquid) flow enters the branch. For smaller distances the interface is locally deformed because of the pressure decrease due to the fluid acceleration near the branch inlet (Bernoulli effect) and liquid (gas) can be entrained. This report contains photographs illustrating the flow phenomena as well as a general correlation to determine the beginning of entrainment. Results are presented on the branch mass flow rate and quality as a function of a normalized distance between the interface and the branch inlet. A model was developed which enables to predict the branch quality and mass flux. Results from air-water flow through horizontal branches, were extrapolated for steam water flow at high pressure with critical branch mass flux. (orig./HS) [de

Dynamic Modeling Strategy for Flow Regime Transition in Gas-Liquid Two-Phase Flows

Directory of Open Access Journals (Sweden)

Xia Wang

2012-12-01

Full Text Available In modeling gas-liquid two-phase flows, the concept of flow regimes has been widely used to characterize the global interfacial structure of the flows. Nearly all constitutive relations that provide closures to the interfacial transfers in two-phase flow models, such as the two-fluid model, are flow regime dependent. Current nuclear reactor safety analysis codes, such as RELAP5, classify flow regimes using flow regime maps or transition criteria that were developed for steady-state, fully-developed flows. As two-phase flows are dynamic in nature, it is important to model the flow regime transitions dynamically to more accurately predict the two-phase flows. The present work aims to develop a dynamic modeling strategy to determine flow regimes in gas-liquid two-phase flows through introduction of interfacial area transport equations (IATEs within the framework of a two-fluid model. The IATE is a transport equation that models the interfacial area concentration by considering the creation of the interfacial area, fluid particle (bubble or liquid droplet disintegration, boiling and evaporation, and the destruction of the interfacial area, fluid particle coalescence and condensation. For flow regimes beyond bubbly flows, a two-group IATE has been proposed, in which bubbles are divided into two groups based on their size and shapes, namely group-1 and group-2 bubbles. A preliminary approach to dynamically identify the flow regimes is discussed, in which discriminators are based on the predicted information, such as the void fraction and interfacial area concentration. The flow regime predicted with this method shows good agreement with the experimental observations.

Fluid-elastic instability in tube arrays subjected to air-water and steam-water cross-flow

Science.gov (United States)

Mitra, D.; Dhir, V. K.; Catton, I.

2009-10-01

Flow induced vibrations in heat exchanger tubes have led to numerous accidents and economic losses in the past. Efforts have been made to systematically study the cause of these vibrations and develop remedial design criteria for their avoidance. In this research, experiments were systematically carried out with air-water and steam-water cross-flow over horizontal tubes. A normal square tube array of pitch-to-diameter ratio of 1.4 was used in the experiments. The tubes were suspended from piano wires and strain gauges were used to measure the vibrations. Tubes made of aluminum; stainless steel and brass were systematically tested by maintaining approximately the same stiffness in the tube-wire systems. Instability was clearly seen in single phase and two-phase flow and the critical flow velocity was found to be proportional to tube mass. The present study shows that fully flexible arrays become unstable at a lower flow velocity when compared to a single flexible tube surrounded by rigid tubes. It is also found that tubes are more stable in steam-water flow as compared to air-water flow. Nucleate boiling on the tube surface is also found to have a stabilizing effect on fluid-elastic instability.

Experimental on two sensors combination used in horizontal pipe gas-water two-phase flow

International Nuclear Information System (INIS)

Wu, Hao; Dong, Feng

2014-01-01

Gas-water two phase flow phenomenon widely exists in production and living and the measurement of it is meaningful. A new type of long-waist cone flow sensor has been designed to measure two-phase mass flow rate. Six rings structure of conductance probe is used to measure volume fraction and axial velocity. The calibration of them have been made. Two sensors have been combined in horizontal pipeline experiment to measure two-phase flow mass flow rate. Several model of gas-water two-phase flow has been discussed. The calculation errors of total mass flow rate measurement is less than 5% based on the revised homogeneous flow model

Prediction of two-phase choked-flow through safety valves

International Nuclear Information System (INIS)

Arnulfo, G; Bertani, C; De Salve, M

2014-01-01

Different models of two-phase choked flow through safety valves are applied in order to evaluate their capabilities of prediction in different thermal-hydraulic conditions. Experimental data available in the literature for two-phase fluid and subcooled liquid upstream the safety valve have been compared with the models predictions. Both flashing flows and non-flashing flows of liquid and incondensable gases have been considered. The present paper shows that for flashing flows good predictions are obtained by using the two-phase valve discharge coefficient defined by Lenzing and multiplying it by the critical flow rate in an ideal nozzle evaluated by either Omega Method or the Homogeneous Non-equilibrium Direct Integration. In case of non-flashing flows of water and air, Leung/Darby formulation of the two-phase valve discharge coefficient together with the Omega Method is more suitable to the prediction of flow rate.

Mechanistic multidimensional analysis of horizontal two-phase flows

International Nuclear Information System (INIS)

Tselishcheva, Elena A.; Antal, Steven P.; Podowski, Michael Z.

2010-01-01

The purpose of this paper is to discuss the results of analysis of two-phase flow in horizontal tubes. Two flow situations have been considered: gas/liquid flow in a long straight pipe, and similar flow conditions in a pipe with 90 deg. elbow. The theoretical approach utilizes a multifield modeling concept. A complete three-dimensional two-phase flow model has been implemented in a state-of-the-art computational multiphase fluid dynamics (CMFD) computer code, NPHASE. The overall model has been tested parametrically. Also, the results of NPHASE simulations have been compared against experimental data for a pipe with 90 deg. elbow.

Measurement of gas-liquid two-phase flow around horizontal tube bundle using SF6-water. Simulating high-pressure high-temperature gas-liquid two-phase flow of PWR/SG secondary coolant side at normal pressure

International Nuclear Information System (INIS)

Ishikawa, Atsushi; Imai, Ryoj; Tanaka, Takahiro

2014-01-01

In order to improve prediction accuracy of analysis code used for design and development of industrial products, technology had been developed to create and evaluate constitutive equation incorporated in analysis code. The experimental facility for PWR/SG U tubes part was manufactured to measure local void fraction and gas-liquid interfacial velocity with forming gas-liquid upward two-phase flow simulating high-pressure high-temperature secondary coolant (water-steam) rising vertically around horizontal tube bundle. The experimental facility could reproduce flow field having gas-liquid density ratio equivalent to real system with no heating using SF6 (Sulfur Hexafluoride) gas at normal temperature and pressure less than 1 MPa, because gas-liquid density ratio, surface tension and gas-liquid viscosity ratio were important parameters to determine state of gas-liquid two-phase flow and gas-liquid density ratio was most influential. Void fraction was measured by two different methods of bi-optical probe and conductivity type probe. Test results of gas-liquid interfacial velocity vs. apparent velocity were in good agreement with existing empirical equation within 10% error, which could confirm integrity of experimental facility and appropriateness of measuring method so as to set up original constitutive equation in the future. (T. Tanaka)

Visual Analysis of Inclusion Dynamics in Two-Phase Flow.

Science.gov (United States)

Karch, Grzegorz Karol; Beck, Fabian; Ertl, Moritz; Meister, Christian; Schulte, Kathrin; Weigand, Bernhard; Ertl, Thomas; Sadlo, Filip

2018-05-01

In single-phase flow visualization, research focuses on the analysis of vector field properties. In two-phase flow, in contrast, analysis of the phase components is typically of major interest. So far, visualization research of two-phase flow concentrated on proper interface reconstruction and the analysis thereof. In this paper, we present a novel visualization technique that enables the investigation of complex two-phase flow phenomena with respect to the physics of breakup and coalescence of inclusions. On the one hand, we adapt dimensionless quantities for a localized analysis of phase instability and breakup, and provide detailed inspection of breakup dynamics with emphasis on oscillation and its interplay with rotational motion. On the other hand, we present a parametric tightly linked space-time visualization approach for an effective interactive representation of the overall dynamics. We demonstrate the utility of our approach using several two-phase CFD datasets.

Interfacial structures in downward two-phase bubbly flow

International Nuclear Information System (INIS)

Paranjape, S.S.; Kim, S.; Ishii, M.; Kelly, J.

2003-01-01

Downward two-phase flow was studied considering its significance in view of Light Water Reactor Accidents (LWR) such as Loss of Heat Sink (LOHS) by feed water loss or secondary pipe break. The flow studied, was an adiabatic, air-water, co-current, vertically downward two-phase flow. The experimental test sections had internal hydraulic diameters of 25.4 mm and 50.8 mm. Flow regime map was obtained using the characteristic signals obtained from an impedance void meter, employing neural network based identification methodology to minimize the subjective judgment in determining the flow regimes. A four sensor conductivity probe was used to measure the local two phase flow parameters, which characterize the interfacial structures. The local time averaged two-phase flow parameters measured were: void fraction (α), interfacial area concentration (a i ), bubble velocity (v g ), and Sauter mean diameter (D Sm ). The flow conditions were from the bubbly flow regime. The local profiles of these parameters as well as their axial development revealed the nature of the interfacial structures and the bubble interaction mechanisms occurring in the flow. Furthermore, this study provided a good database for the development of the interfacial area transport equation, which dynamically models the changes in the interfacial area along the flow field. An interfacial area transport equation was developed for downward flow based on that developed for the upward flow, with certain modifications in the bubble interaction terms. The area averaged values of the interfacial area concentration were compared with those predicted by the interfacial area transport model. (author)

Mechanics of occurrence of critical flow in compressible two-phase flow

International Nuclear Information System (INIS)

Katto, Yoshiro; Sudo, Yukio

1976-01-01

Fundamental framework of mechanics for the occurrence of critical flow is investigated, following the principle that the critical flow appears as a limit in a continuous change of state of flow along a nozzle (or a pipe) and should be derived only from simultaneous mechanical equations concerned with the flow. Mathematical procedures with which the critical flow: (i) the single phase flow of an arbitrary fluid, unrestricted by the equation of state of ideal gas, where the number of simultaneous equations is equal to the number of independent variables, and (ii) the one-component, separated two-phase flow under saturated condition, where the number of equations exceeds that of variables. In each case, interesting mechanism of leading to the occurrence of a limiting state of flow at a definite cross-section in a nozzle (incl. a pipe) is clarified, and a definite state of flow at the critical cross-section is also determined. Then, the analysis is extended to the critical flow which should appear in the completely isolated and the homogeneously dispersed, two-component, two-phase flow (composed of a compressible and an incompressible substance). It is found that the analyses of these special flow patterns provide several supplementary information to the mechanics of critical flow. (auth.)

Measurement of mass flux in high temperature high pressure steam-water two-phase flow using a combination of Pitot tubes and a gamma densitometer

International Nuclear Information System (INIS)

Chan, A.M.C.; Bzovey, D.

1990-01-01

The design and calibration of a two-phase mass-flux measurement device making use of a Pitot-tube rake and a gamma densitometer are described. Five Pitot tubes and three chordal void-fraction measurements are used. Similar devices have been reported previously. The present device is designed for easy operation and simple data interpretation for both axisymmetric and non-axisymmetric flows under high pressure transient two-phase flow conditions. The device was calibrated using a vertical two-phase flow loop as well as a model-scale pump loop in horizontal orientation. Good agreement between the measured two-phase mass fluxes and the single-phase values was obtained in both cases. (orig.)

Stability of interfacial waves in two-phase flows

Energy Technology Data Exchange (ETDEWEB)

Liu, W S [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

The influence of the interfacial pressure and the flow distribution in the one-dimensional two-fluid model on the stability problems of interfacial waves is discussed. With a proper formulation of the interfacial pressure, the following two-phase phenomena can be predicted from the stability and stationary criteria of the interfacial waves: onset of slug flow, stationary hydraulic jump in a stratified flow, flooding in a vertical pipe, and the critical void fraction of a bubbly flow. It can be concluded that the interfacial pressure plays an important role in the interfacial wave propagation of the two-fluid model. The flow distribution parameter may enhance the flow stability range, but only plays a minor role in the two-phase characteristics. (author). 20 refs., 3 tabs., 4 figs.

Damping and fluidelastic instability in two-phase cross-flow heat exchanger tube arrays

Science.gov (United States)

Moran, Joaquin E.

An experimental study was conducted to investigate damping and fluidelastic instability in tube arrays subjected to two-phase cross-flow. The purpose of this research was to improve our understanding of these phenomena and how they are affected by void fraction and flow regime. The model tube bundle had 10 cantilevered tubes in a parallel-triangular configuration, with a pitch ratio of 1.49. The two-phase flow loop used in this research utilized Refrigerant 11 as the working fluid, which better models steam-water than air-water mixtures in terms of vapour-liquid mass ratio as well as permitting phase changes due to pressure fluctuations. The void fraction was measured using a gamma densitometer, introducing an improvement over the Homogeneous Equilibrium Model (HEM) in terms of void fraction, density and velocity predictions. Three different damping measurement methodologies were implemented and compared in order to obtain a more reliable damping estimate. The methods were the traditionally used half-power bandwidth, the logarithmic decrement and an exponential fitting to the tube decay response. The decay trace was obtained by "plucking" the monitored tube from outside the test section using a novel technique, in which a pair of electromagnets changed their polarity at the natural frequency of the tube to produce resonance. The experiments showed that the half-power bandwidth produces higher damping values than the other two methods. The primary difference between the methods is caused by tube frequency shifting, triggered by fluctuations in the added mass and coupling between the tubes, which depend on void fraction and flow regime. The exponential fitting proved to be the more consistent and reliable approach to estimating damping. In order to examine the relationship between the damping ratio and mass flux, the former was plotted as a function of void fraction and pitch mass flux in an iso-contour plot. The results showed that damping is not independent of mass

An assessment of void fraction correlations for vertical upward steam-water flow

International Nuclear Information System (INIS)

Vijayan, P.K.; Maruthi Ramesh, N.; Pilkhwal, D.S.; Saha, D.

1997-01-01

An assessment of sixteen void fraction correlations have been carried out using experimental void fraction data compiled from open literature for vertical upward steam-water flow. Nearly 80% of all the data pertained to natural circulation flow. This assessment showed that best prediction is obtained by Chexal et al. (1996) correlation followed by Hughmark (1965) and the Mochizuki and Ishii (1992) correlations. The Mochizuki-Ishii correlation is found to satisfy all the three limiting conditions whereas Chexal et al. (1996) correlation satisfies all the limiting conditions at moderately high mass fluxes (greater than 140 kg/m 2 s) while Hughmark correlation satisfies only one of the three limiting conditions. The available void fraction data in the open literature for steam-water two-phase flow lies predominantly in the low quality region. This is the reason why correlations like Hughmark which do not satisfy the upper limiting condition (i.e. at x=1, α=1) perform rather well in assessments. Additional work is required for the generation of high quality (greater than 40%) void fraction data. (author)

Regimes of Two-Phase Flow in Short Rectangular Channel

Science.gov (United States)

Chinnov, Evgeny A.; Guzanov, Vladimir V.; Cheverda, Vyacheslav; Markovich, Dmitry M.; Kabov, Oleg A.

2009-08-01

Experimental study of two-phase flow in the short rectangular horizontal channel with height 440 μm has been performed. Characteristics of liquid motion inside the channel have been registered and measured by the Laser Induced Fluorescence technique. New information has allowed determining more precisely the characteristics of churn regime and boundaries between different regimes of two-phase flow. It was shown that formation of some two-phase flow regimes and transitions between them are determined by instability of the flow in the lateral parts of the channel.

Steam generator thermal-hydraulics

International Nuclear Information System (INIS)

Inch, W.W.; Scott, D.A.; Carver, M.B.

1980-01-01

This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)

Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

International Nuclear Information System (INIS)

Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

1982-01-01

Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

Unsteady State Two Phase Flow Pressure Drop Calculations

OpenAIRE

Ayatollahi, Shahaboddin

1992-01-01

A method is presented to calculate unsteady state two phase flow in a gas-liquid line based on a quasi-steady state approach. A computer program for numerical solution of this method was prepared. Results of calculations using the computer program are presented for several unsteady state two phase flow systems

Two-Phase Annular Flow in Helical Coil Flow Channels in a Reduced Gravity Environment

Science.gov (United States)

Keshock, Edward G.; Lin, Chin S.

1996-01-01

A brief review of both single- and two-phase flow studies in curved and coiled flow geometries is first presented. Some of the complexities of two-phase liquid-vapor flow in curved and coiled geometries are discussed, and serve as an introduction to the advantages of observing such flows under a low-gravity environment. The studies proposed -- annular two-phase air-water flow in helical coil flow channels are described. Objectives of the studies are summarized.

Qualitative behaviour of incompressible two-phase flows with phase ...

Indian Academy of Sciences (India)

Jan Prüss

2017-11-07

Nov 7, 2017 ... Qualitative behaviour of incompressible two-phase flows with phase ... Germany. 2Graduate School of Human and Environmental Studies, Kyoto University, ... Note that j is a dummy variable as it can be eliminated from the ...

Study on reverse flow characteristics under natural circulation in inverted U-tube steam generator

International Nuclear Information System (INIS)

Duan Jun; Zhou Tao; Zhang Lei; Hong Dexun; Liu Ping

2013-01-01

Natural circulation is important for application in the nuclear power industry. Aiming at the steam generator of AP1000 pressurized water reactor loop, the mathematical model was established to analysis the reverse flow of single-phase water in the inverted U-tubes of a steam generator in a natural circulation system. The length distribution and the mass flow rates in both tubes with normal and reverse flow were determined respectively. The research results show that the reverse flow may result in sharp decrease of gravity pressure head, circulation mass flow rate and heat release rate of natural circulation. It has adverse influence on natural circulation. (authors)

Two-phase flow operational maps for multi-microchannel evaporators

International Nuclear Information System (INIS)

Szczukiewicz, Sylwia; Borhani, Navid; Thome, John Richard

2013-01-01

Highlights: • New operational maps for several different micro-evaporators are presented. • Inlet micro-orifices prevented flow instability, back flow, and flow maldistribution. • Eight different operating regimes were distinguished. • The flashing two-phase flow without back flow operating regime is preferred. -- Abstract: The current paper presents new operational maps for several different multi-microchannel evaporators, with and without any inlet restrictions (micro-orifices), for the two-phase flow of refrigerants R245fa, R236fa, and R1234ze(E). The test fluids flowed in 67 parallel channels, each having a cross-sectional area of 100 × 100 μm 2 . In order to emulate the power dissipated by active components in a 3D CMOS CPU chip, two aluminium microheaters were sputtered onto the back-side of the test section providing a 0.5 cm 2 each. Without any inlet restrictions in the micro-evaporator, significant parallel channel flow instabilities, vapor back flow, and flow maldistribution led to high-amplitude and high-frequency temperature and pressure oscillations. Such undesired phenomena were then prevented by placing restrictions at the inlet of each channel. High-speed flow visualization distinguished eight different operating regimes of the two-phase flow depending on the tested operating conditions. Therefore, the preferred operating regimes can be easily traced. In particular, flashing two-phase flow without back flow appeared to be the best operating regime without any flow and temperature instabilities

Present status of numerical analysis on transient two-phase flow

International Nuclear Information System (INIS)

Akimoto, Masayuki; Hirano, Masashi; Nariai, Hideki.

1987-01-01

The Special Committee for Numerical Analysis of Thermal Flow has recently been established under the Japan Atomic Energy Association. Here, some methods currently used for numerical analysis of transient two-phase flow are described citing some information given in the first report of the above-mentioned committee. Many analytical models for transient two-phase flow have been proposed, each of which is designed to describe a flow by using differential equations associated with conservation of mass, momentum and energy in a continuous two-phase flow system together with constructive equations that represent transportation of mass, momentum and energy though a gas-liquid interface or between a liquid flow and the channel wall. The author has developed an analysis code, called MINCS, that serves for systematic examination of conservation equation and constructive equations for two-phase flow models. A one-dimensional, non-equilibrium two-liquid flow model that is used as the basic model for the code is described. Actual procedures for numerical analysis is shown and some problems concerning transient two-phase analysis are described. (Nogami, K.)

Numerical calculation of two phase flow in a shock tube

International Nuclear Information System (INIS)

Rivard, W.C.; Travis, J.R.; Torrey, M.D.

1976-01-01

Numerical calculations of the dynamics of initially saturated water-steam mixtures in a shock tube demonstrate the accuracy and efficiency of a new solution technique for the transient, two-dimensional, two-fluid equations. The dependence of the calculated results on time step and cell size are investigated. The effects of boiling and condensation on the flow physics suggest the merits of basic fluid dynamic measurements for the determination and evaluation of mass exchange models

Characterization of horizontal air–water two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Kong, Ran; Kim, Seungjin, E-mail: skim@psu.edu

2017-02-15

Highlights: • A visualization study is performed to develop flow regime map in horizontal flow. • Database in horizontal bubbly flow is extended using a local conductivity probe. • Frictional pressure drop analysis is performed in horizontal bubbly flow. • Drift flux analysis is performed in horizontal bubbly flow. - Abstract: This paper presents experimental studies performed to characterize horizontal air–water two-phase flow in a round pipe with an inner diameter of 3.81 cm. A detailed flow visualization study is performed using a high-speed video camera in a wide range of two-phase flow conditions to verify previous flow regime maps. Two-phase flows are classified into bubbly, plug, slug, stratified, stratified-wavy, and annular flow regimes. While the transition boundaries identified in the present study compare well with the existing ones (Mandhane et al., 1974) in general, some discrepancies are observed for bubbly-to-plug/slug, and plug-to-slug transition boundaries. Based on the new transition boundaries, three additional test conditions are determined in horizontal bubbly flow to extend the database by Talley et al. (2015a). Various local two-phase flow parameters including void fraction, interfacial area concentration, bubble velocity, and bubble Sauter mean diameter are obtained. The effects of increasing gas flow rate on void fraction, bubble Sauter mean diameter, and bubble velocity are discussed. Bubbles begin to coalesce near the gas–liquid layer instead of in the highly packed region when gas flow rate increases. Using all the current experimental data, two-phase frictional pressure loss analysis is performed using the Lockhart–Martinelli method. It is found that the coefficient C = 24 yields the best agreement with the data with the minimum average difference. Moreover, drift flux analysis is performed to predict void-weighted area-averaged bubble velocity and area-averaged void fraction. Based on the current database, functional

Bubbly flows around a two-dimensional circular cylinder

Science.gov (United States)

Lee, Jubeom; Park, Hyungmin

2016-11-01

Two-phase cross flows around a bluff body occur in many thermal-fluid systems like steam generators, heat exchangers and nuclear reactors. However, our current knowledge on the interactions among bubbles, bubble-induced flows and the bluff body are limited. In the present study, the gas-liquid bubbly flows around a solid circular cylinder are experimentally investigated while varying the mean void fraction from 5 to 27%. The surrounding liquid (water) is initially static and the liquid flow is only induced by the air bubbles. For the measurements, we use the high-speed two-phase particle image velocimetry techniques. First, depending on the mean void fraction, two regimes are classified with different preferential concentration of bubbles in the cylinder wake, which are explained in terms of hydrodynamic force balances acting on rising bubbles. Second, the differences between the two-phase and single-phase flows (while matching their Reynolds numbers) around a circular cylinder will be discussed in relation to effects of bubble dynamics and the bubble-induced turbulence on the cylinder wake. Supported by a Grant (MPSS-CG-2016-02) through the Disaster and Safety Management Institute funded by Ministry of Public Safety and Security of Korean government.

Two-phase flow patterns in horizontal rectangular minichannel

Directory of Open Access Journals (Sweden)

Ron’shin Fedor

2016-01-01

Full Text Available The two-phase flow in a short horizontal channel of rectangular cross-section of 1 × 19 mm2 has been studied experimentally. Five conventional two-phase flow patterns have been detected (bubble, churn, stratified, annular and jet and transitions between them have been determined. It is shown that a change in the width of the horizontal channels has a substantial effect on the boundaries between the flow regimes.

Turbulence in two-phase flows

International Nuclear Information System (INIS)

Sullivan, J.P.; Houze, R.N.; Buenger, D.E.; Theofanous, T.G.

1981-01-01

Hot film Anemometry and Laser Doppler Velocimetry have been employed in this work to study the turbulence characteristics of Bubbly and Stratified two-phase flows, respectively. Extensive consistency checks were made to establish the reliability and hence the utility of these experimental techniques for the measurement of turbulence in two-phase flows. Buoyancy-driven turbulence in vertical bubbly flows has been identified experimentally and correlated in terms of a shear velocity superposition approach. This approach provides a criterion for the demarcation of the buoyancy-driven turbulence region from the wall shear-generated turbulence region. Our data confirm the roughly isotropic behavior expected for buoyancy-driven turbulence. Upgrading of our experimental system will permit investigations of the wall-shear dominated regime (i.e., isotropy, superposition approach, etc.). The stratified flow data demonstrate clearly that the maximum in the mean velocity profile does not coincide with the zero shear plane, indicating the existence of a negative eddy viscosity region. Previous studies do not take into account this difference and thus they yield incorrect friction factor data in addition to certain puzzling behavior in the upper wall region. The conditioned turbulence data in the wavy region indicate interesting trends and that an appropriate normalization of intensities must take into account the shear velocity at the interfacial (wavy) region

DSMC simulation of two-phase plume flow with UV radiation

Energy Technology Data Exchange (ETDEWEB)

Li, Jie; Liu, Ying; Wang, Ning; Jin, Ling [College of Aerospace Science and Engineering, National University of Defense Technology, Changsha, Hunan, 410073 (China)

2014-12-09

Rarefied gas-particle two-phase plume in which the phase of particles is liquid or solid flows from a solid propellant rocket of hypersonic vehicle flying at high altitudes, the aluminum oxide particulates not only impact the rarefied gas flow properties, but also make a great difference to plume radiation signature, so the radiation prediction of the rarefied gas-particle two-phase plume flow is very important for space target detection of hypersonic vehicles. Accordingly, this project aims to study the rarefied gas-particle two-phase flow and ultraviolet radiation (UV) characteristics. Considering a two-way interphase coupling of momentum and energy, the direct simulation Monte Carlo (DSMC) method is developed for particle phase change and the particle flow, including particulate collision, coalescence as well as separation, and a Monte Carlo ray trace model is implemented for the particulate UV radiation. A program for the numerical simulation of the gas-particle two-phase flow and radiation in which the gas flow nonequilibrium is strong is implemented as well. Ultraviolet radiation characteristics of the particle phase is studied based on the calculation of the flow field coupled with the radiation calculation, the radiation model for different size particles is analyzed, focusing on the effects of particle emission, absorption, scattering as well as the searchlight emission of the nozzle. A new approach may be proposed to describe the rarefied gas-particle two-phase plume flow and radiation transfer characteristics in this project.

DSMC simulation of two-phase plume flow with UV radiation

Science.gov (United States)

Li, Jie; Liu, Ying; Wang, Ning; Jin, Ling

2014-12-01

Rarefied gas-particle two-phase plume in which the phase of particles is liquid or solid flows from a solid propellant rocket of hypersonic vehicle flying at high altitudes, the aluminum oxide particulates not only impact the rarefied gas flow properties, but also make a great difference to plume radiation signature, so the radiation prediction of the rarefied gas-particle two-phase plume flow is very important for space target detection of hypersonic vehicles. Accordingly, this project aims to study the rarefied gas-particle two-phase flow and ultraviolet radiation (UV) characteristics. Considering a two-way interphase coupling of momentum and energy, the direct simulation Monte Carlo (DSMC) method is developed for particle phase change and the particle flow, including particulate collision, coalescence as well as separation, and a Monte Carlo ray trace model is implemented for the particulate UV radiation. A program for the numerical simulation of the gas-particle two-phase flow and radiation in which the gas flow nonequilibrium is strong is implemented as well. Ultraviolet radiation characteristics of the particle phase is studied based on the calculation of the flow field coupled with the radiation calculation, the radiation model for different size particles is analyzed, focusing on the effects of particle emission, absorption, scattering as well as the searchlight emission of the nozzle. A new approach may be proposed to describe the rarefied gas-particle two-phase plume flow and radiation transfer characteristics in this project.

Two-Phase Quality/Flow Meter

Science.gov (United States)

Moerk, J. Steven (Inventor); Youngquist, Robert C. (Inventor); Werlink, Rudy J. (Inventor)

1999-01-01

A quality and/or flow meter employs a capacitance probe assembly for measuring the dielectric constant of flow stream, particularly a two-phase flow stream including liquid and gas components.ne dielectric constant of the flow stream varies depending upon the volume ratios of its liquid and gas components, and capacitance measurements can therefore be employed to calculate the quality of the flow, which is defined as the volume ratio of liquid in the flow to the total volume ratio of gas and liquid in the flow. By using two spaced capacitance sensors, and cross-correlating the time varying capacitance values of each, the velocity of the flow stream can also be determined. A microcontroller-based processing circuit is employed to measure the capacitance of the probe sensors.The circuit employs high speed timer and counter circuits to provide a high resolution measurement of the time interval required to charge each capacitor in the probe assembly. In this manner, a high resolution, noise resistant, digital representation of each of capacitance value is obtained without the need for a high resolution A/D converter, or a high frequency oscillator circuit. One embodiment of the probe assembly employs a capacitor with two ground plates which provide symmetry to insure that accurate measurements are made thereby.

Study of two-phase flow redistribution between two passes of a heat exchanger

International Nuclear Information System (INIS)

Mendes de Moura, L.F.

1989-04-01

The object of the present thesis deals with the study of two-phase flow redistribution between two passes of a heat exchanger. Mass flow rate measurements of each component performed at each channel outlet of the second pass allowed us to determine the influence of mass flow, gas quality, flow direction (upward or downward) and common header geometry upon flow redistribution. Local void fraction inside common header was measured with an optical probe. A two-dimensional two-phase flow computational code was developed from a two-fluid model. Modelling of interfacial momentum transfer was used in order to take into account twp-phase flow patterns in common headers. Numerical simulation results show qualitative agreement with experimental results. Present theoretical model limitations are analysed and future improvements are proposed [fr

Two-phase flow induced parametric vibrations in structural systems

International Nuclear Information System (INIS)

Hara, Fumio

1980-01-01

This paper is divided into two parts concerning piping systems and a nuclear fuel pin system. The significant experimental results concerning the random vibration induced in an L-shaped pipe by air-water two-phase flow and the theoretical analysis of the vibration are described in the first part. It was clarified for the first time that the parametric excitation due to the periodic changes of system mass, centrifugal force and Coriolis force was the mechanism of exciting the vibration. Moreover, the experimental and theoretical analyses of the mechanism of exciting vibration by air-water two-phase flow in a straight, horizontal pipe were carried out, and the first natural frequency of the piping system was strongly related to the dominant frequency of void signals. The experimental results on the vibration of a nuclear fuel pin model in parallel air-water two-phase flow are reported in the latter part. The relations between vibrational strain variance and two-phase flow velocity or pressure fluctuation, and the frequency characteristics of vibrational strain variance were obtained. The theoretical analysis of the dynamic interaction between air-water two-phase flow and a fuel pin structure, and the vibrational instability of fuel pins in alternate air and water slugs or in large bubble flow are also reported. (Kako, I.)

A new method for the measurement of two-phase mass flow rate using average bi-directional flow tube

International Nuclear Information System (INIS)

Yoon, B. J.; Uh, D. J.; Kang, K. H.; Song, C. H.; Paek, W. P.

2004-01-01

Average bi-directional flow tube was suggested to apply in the air/steam-water flow condition. Its working principle is similar with Pitot tube, however, it makes it possible to eliminate the cooling system which is normally needed to prevent from flashing in the pressure impulse line of pitot tube when it is used in the depressurization condition. The suggested flow tube was tested in the air-water vertical test section which has 80mm inner diameter and 10m length. The flow tube was installed at 120 of L/D from inlet of test section. In the test, the pressure drop across the average bi-directional flow tube, system pressure and average void fraction were measured on the measuring plane. In the test, fluid temperature and injected mass flow rates of air and water phases were also measured by a RTD and two coriolis flow meters, respectively. To calculate the phasic mass flow rates : from the measured differential pressure and void fraction, Chexal drift-flux correlation was used. In the test a new correlation of momentum exchange factor was suggested. The test result shows that the suggested instrumentation using the measured void fraction and Chexal drift-flux correlation can predict the mass flow rates within 10% error of measured data

Non-local two phase flow momentum transport in S BWR

International Nuclear Information System (INIS)

Espinosa P, G.; Salinas M, L.; Vazquez R, A.

2015-09-01

The non-local momentum transport equations derived in this work contain new terms related with non-local transport effects due to accumulation, convection, diffusion and transport properties for two-phase flow. For instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where the local volume averaging equations fail. The S BWR was considered to study the non-local effects on the two-phase flow thermal-hydraulic core performance in steady-state, and the results were compared with the classical local averaging volume conservation equations. (Author)

Non-local two phase flow momentum transport in S BWR

Energy Technology Data Exchange (ETDEWEB)

Espinosa P, G.; Salinas M, L.; Vazquez R, A., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Apdo. Postal 55-535, 09340 Ciudad de Mexico (Mexico)

2015-09-15

The non-local momentum transport equations derived in this work contain new terms related with non-local transport effects due to accumulation, convection, diffusion and transport properties for two-phase flow. For instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where the local volume averaging equations fail. The S BWR was considered to study the non-local effects on the two-phase flow thermal-hydraulic core performance in steady-state, and the results were compared with the classical local averaging volume conservation equations. (Author)

Three-dimensional numerical modeling of turbulent single-phase and two-phase flow in curved pipes

International Nuclear Information System (INIS)

Xin, R.C.; Dong, Z.F.; Ebadian, M.A.

1996-01-01

In this study, three-dimensional single-phase and two-phase flows in curved pipes have been investigated numerically. Two different pipe configurations were computed. When the results of the single-phase flow simulation were compared with the experimental data, a fairly good agreement was achieved. A flow-developing process has been suggested in single-phase flow, in which the turbulence is stronger near the outer tube wall than near the inner tube wall. For two-phase flow, the Eulerian multiphase model was used to simulate the phase distribution of a three-dimensional gas-liquid bubble flow in curved pipe. The RNG/κ-ε turbulence model was used to determine the turbulence field. An inlet gas void fraction of 5 percent was simulated. The gas phase effects on the liquid phase flow velocity have been examined by comparing the results of single-phase flow and two-phase flow. The findings show that for the downward flow in the U bend, the gas concentrates at the inner portion of the cross section at φ = π/18 - π/6 in most cases. The results of the phase distribution simulation are compared to experimental observations qualitatively and topologically

Void fraction fluctuations in two-phase gas-liquid flow

International Nuclear Information System (INIS)

Ulbrich, R.

1987-01-01

Designs of the apparatus in which two-phase gas-liquid flow occurs are usually based on the mean value of parameters such as pressure drop and void fraction. The flow of two-phase mixtures generally presents a very complicated flow structure, both in terms of the unsteady formation on the interfacial area and in terms of the fluctuations of the velocity, pressure and other variables within the flow. When the gas void fraction is near 0 or 1 / bubble or dispersed flow regimes / then oscillations of void fraction are very small. The intermittent flow such as plug and slug/ froth is characterized by alternately flow portions of liquid and gas. It influences the change of void fractions in time. The results of experimental research of gas void fraction fluctuations in two-phase adiabatic gas-liquid flow in a vertical pipe are presented

Technical description of the test section SUCOT to investigate a water/steam two-phase flow

International Nuclear Information System (INIS)

Daubner, M.; Janssens-Maenhout, G.; Knebel, J.U.

2002-02-01

Within the POOLTHY Project of the European Union (Euratom Fourth Framework Programme contract FJ4J-CT95-0003) an active/passive concept (SUCO-Programme) was investigated which controls the heat removal after a potential core melt-down accident in an evolutionary light water reactor by spreading and stabilising the core melt in the reactor sump and flooding the melt with sump water from above. The experiments were performed in the test facility SUCOT which was designed and erected at the Institute for Nuclear and Energy Technologies (IKET). The report gives an overview on the SUCO-Programme and the scaling analysis, which was applied to design the test facility SUCOT. A detailed technical description of the test facility SUCOT is given, in which the natural circulation driven two-phase flow within the reactor sump and relevant phenomena such as flow boiling, disperse bubbly flow with and without mass transfer, and geysering are investigated. The major components of the test facility, the three-loop system and the instrumentation are described. Finally, a perspective for future application of the gained knowledge is given. (orig.) [de

On the origin of burnout in tubes during subheated water and wet steam flow

International Nuclear Information System (INIS)

Doroshchuk, V.E.

1980-01-01

Mecahnisms of arising the burnouts of the first and second kinds during water and steam-water mixture flow in a tube have been studied. It is shown that the burnout of the first kind arises in the cases when the main part is palyed by the thermal processes providing a possibility of the film boiling or destruction of near-wall liquid film. The high value of critical heat flux qsub(cr) is typical for this kind of burnout. In arising the burnout of the second kind the determining part is played by the hydrodynamic processes in the channel but not by the thermal ones. In this case the burnout is related with the formation of disperse structure of the flow in the pipe. The thermal load does not play the determining part in this case. The burnout arises at any q value (within the limits qsub(cr)sup(0)>q>qsub(gr)sup(0)) but always at the certain steam content. On the base of the analysis of conditions of burnout in steam-generating tubes it is concluded that determination of the two-phase flow structure in heating tubes, determination of the regularities of flow rate and film thickness changes in annular flows, investigation of the moisture carrying out by bubbles from a near-wall liquid film are of the greatest importance

Oscillatory two-phase flows

International Nuclear Information System (INIS)

Boure, J.A.

1974-12-01

Two-phase flow instabilities are classified according to three criteria: the static or dynamic nature of the phenomenon, the necessity or not of a triggering phenomenon, and the pure or compound character of the phenomenon. Tables give the elementary instability phenomena, and the practical types of instability. Flow oscillations (or dynamic instabilities) share a number of characteristics which are dealt with, they are caused by the dynamic interactions between the flow parameters (flow rate, density, pressure, enthalpy and their distributions). Oscillation types are discussed: pure oscillations are density wave oscillations, acoustic oscillations may also occur, various compound oscillations involve either the density wave or the acoustic wave mechanism, interacting with some of the boundary conditions in the device. The analysis of slow oscillations has been made either by means of a simplified model (prediction of the thresholds) or of computer codes. Numerous computer codes are available [fr

Flow regime classification in air-magnetic fluid two-phase flow.

Science.gov (United States)

Kuwahara, T; De Vuyst, F; Yamaguchi, H

2008-05-21

A new experimental/numerical technique of classification of flow regimes (flow patterns) in air-magnetic fluid two-phase flow is proposed in the present paper. The proposed technique utilizes the electromagnetic induction to obtain time-series signals of the electromotive force, allowing us to make a non-contact measurement. Firstly, an experiment is carried out to obtain the time-series signals in a vertical upward air-magnetic fluid two-phase flow. The signals obtained are first treated using two kinds of wavelet transforms. The data sets treated are then used as input vectors for an artificial neural network (ANN) with supervised training. In the present study, flow regimes are classified into bubbly, slug, churn and annular flows, which are generally the main flow regimes. To validate the flow regimes, a visualization experiment is also performed with a glycerin solution that has roughly the same physical properties, i.e., kinetic viscosity and surface tension, as a magnetic fluid used in the present study. The flow regimes from the visualization are used as targets in an ANN and also used in the estimation of the accuracy of the present method. As a result, ANNs using radial basis functions are shown to be the most appropriate for the present classification of flow regimes, leading to small classification errors.

Complex network analysis in inclined oil–water two-phase flow

International Nuclear Information System (INIS)

Zhong-Ke, Gao; Ning-De, Jin

2009-01-01

Complex networks have established themselves in recent years as being particularly suitable and flexible for representing and modelling many complex natural and artificial systems. Oil–water two-phase flow is one of the most complex systems. In this paper, we use complex networks to study the inclined oil–water two-phase flow. Two different complex network construction methods are proposed to build two types of networks, i.e. the flow pattern complex network (FPCN) and fluid dynamic complex network (FDCN). Through detecting the community structure of FPCN by the community-detection algorithm based on K-means clustering, useful and interesting results are found which can be used for identifying three inclined oil–water flow patterns. To investigate the dynamic characteristics of the inclined oil–water two-phase flow, we construct 48 FDCNs under different flow conditions, and find that the power-law exponent and the network information entropy, which are sensitive to the flow pattern transition, can both characterize the nonlinear dynamics of the inclined oil–water two-phase flow. In this paper, from a new perspective, we not only introduce a complex network theory into the study of the oil–water two-phase flow but also indicate that the complex network may be a powerful tool for exploring nonlinear time series in practice. (general)

Measurement of phase interaction in dispersed gas-particle two-phase flow by phase-doppler anemometry

OpenAIRE

Mergheni Ali Mohamed; Ben Ticha Hmaied; Sautet Jen-Charles; Godard Gille; Ben Nasrallah Sassi

2008-01-01

For simultaneous measurement of size and velocity distributions of continuous and dispersed phases in a two-phase flow a technique phase-Doppler anemometry was used. Spherical glass particles with a particle diameter range from 102 up to 212 µm were used. In this two-phase flow an experimental results are presented which indicate a significant influence of the solid particles on the flow characteristics. The height of influence of these effects depends on the local position in the jet. Near t...

Flow-pattern identification and nonlinear dynamics of gas-liquid two-phase flow in complex networks.

Science.gov (United States)

Gao, Zhongke; Jin, Ningde

2009-06-01

The identification of flow pattern is a basic and important issue in multiphase systems. Because of the complexity of phase interaction in gas-liquid two-phase flow, it is difficult to discern its flow pattern objectively. In this paper, we make a systematic study on the vertical upward gas-liquid two-phase flow using complex network. Three unique network construction methods are proposed to build three types of networks, i.e., flow pattern complex network (FPCN), fluid dynamic complex network (FDCN), and fluid structure complex network (FSCN). Through detecting the community structure of FPCN by the community-detection algorithm based on K -mean clustering, useful and interesting results are found which can be used for identifying five vertical upward gas-liquid two-phase flow patterns. To investigate the dynamic characteristics of gas-liquid two-phase flow, we construct 50 FDCNs under different flow conditions, and find that the power-law exponent and the network information entropy, which are sensitive to the flow pattern transition, can both characterize the nonlinear dynamics of gas-liquid two-phase flow. Furthermore, we construct FSCN and demonstrate how network statistic can be used to reveal the fluid structure of gas-liquid two-phase flow. In this paper, from a different perspective, we not only introduce complex network theory to the study of gas-liquid two-phase flow but also indicate that complex network may be a powerful tool for exploring nonlinear time series in practice.

Strongly coupled dispersed two-phase flows; Ecoulements diphasiques disperses fortement couples

Energy Technology Data Exchange (ETDEWEB)

Zun, I.; Lance, M.; Ekiel-Jezewska, M.L.; Petrosyan, A.; Lecoq, N.; Anthore, R.; Bostel, F.; Feuillebois, F.; Nott, P.; Zenit, R.; Hunt, M.L.; Brennen, C.E.; Campbell, C.S.; Tong, P.; Lei, X.; Ackerson, B.J.; Asmolov, E.S.; Abade, G.; da Cunha, F.R.; Lhuillier, D.; Cartellier, A.; Ruzicka, M.C.; Drahos, J.; Thomas, N.H.; Talini, L.; Leblond, J.; Leshansky, A.M.; Lavrenteva, O.M.; Nir, A.; Teshukov, V.; Risso, F.; Ellinsen, K.; Crispel, S.; Dahlkild, A.; Vynnycky, M.; Davila, J.; Matas, J.P.; Guazelli, L.; Morris, J.; Ooms, G.; Poelma, C.; van Wijngaarden, L.; de Vries, A.; Elghobashi, S.; Huilier, D.; Peirano, E.; Minier, J.P.; Gavrilyuk, S.; Saurel, R.; Kashinsky, O.; Randin, V.; Colin, C.; Larue de Tournemine, A.; Roig, V.; Suzanne, C.; Bounhoure, C.; Brunet, Y.; Tanaka, A.T.; Noma, K.; Tsuji, Y.; Pascal-Ribot, S.; Le Gall, F.; Aliseda, A.; Hainaux, F.; Lasheras, J.; Didwania, A.; Costa, A.; Vallerin, W.; Mudde, R.F.; Van Den Akker, H.E.A.; Jaumouillie, P.; Larrarte, F.; Burgisser, A.; Bergantz, G.; Necker, F.; Hartel, C.; Kleiser, L.; Meiburg, E.; Michallet, H.; Mory, M.; Hutter, M.; Markov, A.A.; Dumoulin, F.X.; Suard, S.; Borghi, R.; Hong, M.; Hopfinger, E.; Laforgia, A.; Lawrence, C.J.; Hewitt, G.F.; Osiptsov, A.N.; Tsirkunov, Yu. M.; Volkov, A.N.

2003-07-01

This document gathers the abstracts of the Euromech 421 colloquium about strongly coupled dispersed two-phase flows. Behaviors specifically due to the two-phase character of the flow have been categorized as: suspensions, particle-induced agitation, microstructure and screening mechanisms; hydrodynamic interactions, dispersion and phase distribution; turbulence modulation by particles, droplets or bubbles in dense systems; collective effects in dispersed two-phase flows, clustering and phase distribution; large-scale instabilities and gravity driven dispersed flows; strongly coupled two-phase flows involving reacting flows or phase change. Topic l: suspensions particle-induced agitation microstructure and screening mechanisms hydrodynamic interactions between two very close spheres; normal stresses in sheared suspensions; a critical look at the rheological experiments of R.A. Bagnold; non-equilibrium particle configuration in sedimentation; unsteady screening of the long-range hydrodynamic interactions of settling particles; computer simulations of hydrodynamic interactions among a large collection of sedimenting poly-disperse particles; velocity fluctuations in a dilute suspension of rigid spheres sedimenting between vertical plates: the role of boundaries; screening and induced-agitation in dilute uniform bubbly flows at small and moderate particle Reynolds numbers: some experimental results. Topic 2: hydrodynamic interactions, dispersion and phase distribution: hydrodynamic interactions in a bubble array; A 'NMR scattering technique' for the determination of the structure in a dispersion of non-brownian settling particles; segregation and clustering during thermo-capillary migration of bubbles; kinetic modelling of bubbly flows; velocity fluctuations in a homogeneous dilute dispersion of high-Reynolds-number rising bubbles; an attempt to simulate screening effects at moderate particle Reynolds numbers using an hybrid formulation; modelling the two-phase

Rolling effects on two-phase flow pattern and void fraction

International Nuclear Information System (INIS)

Yan Changqi; Yu Kaiqiu; Luan Feng; Cao Xiaxin

2008-01-01

The experimental and theoretical study was carried out for the upward gas-liquid two-phase explained reasonably through the analysis of slip ratio of two-phase flow and theoretical analysis using momentum equation of two-phase flow separating model. (authors)

Fluisd elastic instability and fretting-wear characteristics of steam generator helical tubes subjected to single-phase external flow and two-phase internal flow

International Nuclear Information System (INIS)

Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

2004-01-01

This study investigates the fluid elastic instability characteristics of steam generator (SG) helical type tubes and the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted by a general purpose computational fluid dynamics code employing the finite volume element modeling. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for helical type tubes with various conditions. Special emphases are on the effects of coil diameter and the number of turns on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting wear characteristics of the tube

Design specifications to ensure flow-induced vibration and fretting-wear performance in CANDU steam generators and heat exchangers

International Nuclear Information System (INIS)

Janzen, V.P.; Han, Y.; Pettigrew, M.J.

2009-01-01

Preventing flow-induced vibration and fretting-wear problems in steam generators and heat exchangers requires design specifications that bring together specific guidelines, analysis methods, requirements and appropriate performance criteria. This paper outlines the steps required to generate and support such design specifications for CANDU nuclear steam generators and heat exchangers, and relates them to typical steam-generator design features and computer modeling capabilities. It also describes current issues that are driving changes to flow-induced vibration and fretting-wear specifications that can be applied to the design process for component refurbishment, replacement or new designs. These issues include recent experimental or field evidence for new excitation mechanisms, e.g., the possibility of in-plane fluidelastic instability of U-tubes, the demand for longer reactor and component lifetimes, the need for better predictions of dynamic properties and vibration response, e.g., two-phase random-turbulence excitation, and requirements to consider system 'excursions' or abnormal scenarios, e.g., a main steam line break in the case of steam generators. The paper describes steps being taken to resolve these issues. (author)

Investigation of a two-phase nozzle flow and validation of several computer codes by the experimental data

International Nuclear Information System (INIS)

Kedziur, F.

1980-03-01

Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initial conditions: The pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of subcritical as well as critical experiments with different flow pattern is investigated. Additional air/water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiments. The layout of the nozzle and the applied measurement technique allow for a separate testing of physical models and the determination of empirical model parameters, respectively: In the four codes DUESE, DRIX-20, RELAP4/MOD6 and STRUYA the models - if they exist - for slip between the phases, thermodynamic non-equilibrium, pipe friction and critical mass flow rate are validated and criticised in comparison with the experimental data, and the corresponding model parameters are determined. The parameters essentially are a function of the void fraction. (orig.) [de

Two-phase flow instrumentation research at RPI

International Nuclear Information System (INIS)

Lahey, R.T. Jr.; Krycuk, G.

1979-01-01

Novel instrumentation for the measurement of void fraction and phase velocity was developed. An optical digital interferometer and a dual beam x-ray equipment were designed for detection of voids. Pitot tube measurements were made to understand two-phase flow phenomena in liquid phase velocity

Two-phase air-water stratified flow measurement using ultrasonic techniques

International Nuclear Information System (INIS)

Fan, Shiwei; Yan, Tinghu; Yeung, Hoi

2014-01-01

In this paper, a time resolved ultrasound system was developed for investigating two-phase air-water stratified flow. The hardware of the system includes a pulsed wave transducer, a pulser/receiver, and a digital oscilloscope. The time domain cross correlation method is used to calculate the velocity profile along ultrasonic beam. The system is able to provide velocities with spatial resolution of around 1mm and the temporal resolution of 200μs. Experiments were carried out on single phase water flow and two-phase air-water stratified flow. For single phase water flow, the flow rates from ultrasound system were compared with those from electromagnetic flow (EM) meter, which showed good agreement. Then, the experiments were conducted on two-phase air-water stratified flow and the results were given. Compared with liquid height measurement from conductance probe, it indicated that the measured velocities were explainable

Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

International Nuclear Information System (INIS)

Mohany, A.; Feenstra, P.; Janzen, V.P.; Richard, R.

2009-01-01

Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

International Nuclear Information System (INIS)

Choi, Seok Ki; Kim, Seong O

2011-01-01

A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

Phase separation and pressure drop of two-phase flow in vertical manifolds

International Nuclear Information System (INIS)

Zetzmann, K.

1982-01-01

The splitting of a two-phase mass flow in a tube manifold results in a separation between liquid and gas phase. A study is presented of the phase distribution and the related two-phase pressure drop for vertical manifolds in the technically relevant geometry and flow parameter region of an air-water-flow. At the outlet changes in the gas/fluid-radio are observed which are proportional to this ratio at the inlet. The separation characteristic strongly depends on the massflow through the junction. Empirical equations are given to calculate the separation. Measuring the pressure drop at main- and secondary tube of the manifold the additional pressure drop can be obtained. If these results are related with the dynamic pressure at the inlet, two-phase resistance coefficients can be deduced, which may be tested by empirical relations. (orig.) [de

Modeling and numerical analysis of non-equilibrium two-phase flows

International Nuclear Information System (INIS)

Rascle, P.; El Amine, K.

1997-01-01

We are interested in the numerical approximation of two-fluid models of nonequilibrium two-phase flows described by six balance equations. We introduce an original splitting technique of the system of equations. This technique is derived in a way such that single phase Riemann solvers may be used: moreover, it allows a straightforward extension to various and detailed exchange source terms. The properties of the fluids are first approached by state equations of ideal gas type and then extended to real fluids. For the construction of numerical schemes , the hyperbolicity of the full system is not necessary. When based on suitable kinetic unwind schemes, the algorithm can compute flow regimes evolving from mixture to single phase flows and vice versa. The whole scheme preserves the physical features of all the variables which remain in the set of physical states. Several stiff numerical tests, such as phase separation and phase transition are displayed in order to highlight the efficiency of the proposed method. The document is a PhD thesis divided in 6 chapters and two annexes. They are entitled: 1. - Introduction (in French), 2. - Two-phase flow, modelling and hyperbolicity (in French), 3. - A numerical method using upwind schemes for the resolution of two-phase flows without exchange terms (in English), 4. - A numerical scheme for one-phase flow of real fluids (in English), 5. - An upwind numerical for non-equilibrium two-phase flows (in English), 6. - The treatment of boundary conditions (in English), A.1. The Perthame scheme (in English) and A.2. The Roe scheme (in English)

An introduction to two-phase flows

International Nuclear Information System (INIS)

Lemonnier, Herve

2006-01-01

This course aims at proposing the necessary background for a rational approach to two-phase flows which are notably present in numerous industrial devices and equipment designed to perform energy transfer or mass transfer. The first part proposes a phenomenological approach to main two-phase flow structures and presents their governing variables. The second part presents some proven measurement techniques. The third part focuses on modelling. It recalls the equation elaboration techniques which are based on basic principles of mechanics and thermodynamics and on the application of different averaging operators to these principles. Some useful models are then presented such as models of pressure loss in a duct. The last chapter addresses some fundamental elements of heat transfers in ebullition and condensation

Dynamic Phase Boundary Estimation in Two-phase Flows Based on Electrical Impedance Tomography

International Nuclear Information System (INIS)

Lee, Jeong Seong; Muhammada, Nauman Malik; Kim, Kyung Youn; Kim, Sin

2008-01-01

For the dynamic visualization of the phase boundary in two-phase flows, the electrical impedance tomography (EIT) technique is introduced. In EIT, a set of predetermined electrical currents is injected through the electrodes placed on the boundary of the flow passage and the induced electrical potentials are measured on the electrodes. With the relationship between the injected currents and the induced voltages, the electrical conductivity distribution across the flow domain is estimated through the image reconstruction algorithm where the conductivity distribution corresponds to the phase distribution. In the application of EIT to two-phase flows where there are only two conductivity values, the conductivity distribution estimation problem can be transformed into the boundary estimation problem. This paper considers phase boundary estimation with EIT in annular two-phase flows. As the image reconstruction algorithm, the unscented Kalman filter (UKF) is adopted since from the control theory it is reported that the UKF shows better performance than the extended Kalman filter (EKF) that has been commonly used. For the present problem, the formulation of UKF algorithm involved its incorporation in the adopted image reconstruction algorithm. Also, phantom experiments have been conducted to evaluate the improvement reported by UKF

Numerical simulation for gas-liquid two-phase flow in pipe networks

International Nuclear Information System (INIS)

Li Xiaoyan; Kuang Bo; Zhou Guoliang; Xu Jijun

1998-01-01

The complex pipe network characters can not directly presented in single phase flow, gas-liquid two phase flow pressure drop and void rate change model. Apply fluid network theory and computer numerical simulation technology to phase flow pipe networks carried out simulate and compute. Simulate result shows that flow resistance distribution is non-linear in two phase pipe network

Governing equations for a seriated continuum: an unequal velocity model for two-phase flow

International Nuclear Information System (INIS)

Solbrig, C.W.; Hughes, E.D.

1975-05-01

The description of the flow of two-phase fluids is important in many engineering devices. Unexpected transient conditions which occur in these devices cannot, in general, be treated with single-component momentum equations. Instead, the use of momentum equations for each phase is necessary in order to describe the varied transient situations which can occur. These transient conditions can include phases moving in the opposite directions, such as steam moving upward and liquid moving downward, as well as phases moving in the same direction. The derivation of continuity and momentum equations for each phase and an overall energy equation for the mixture are presented. Terms describing interphase forces are described. A seriated (series of) continuum is distinguished from an interpenetrating medium by the representation of interphase friction with velocity differences in the former and velocity gradients in the latter. The seriated continuum also considers imbedded stationary solid surfaces such as occur in nuclear reactor cores. These stationary surfaces are taken into account with source terms. Sufficient constitutive equations are presented to form a complete set of equations. Methods are presented to show that all these coefficients are determinable from microscopic models and well known experimental results. Comparison of the present deviation with previous work is also given. The equations derived here may also be employed in certain multiphase, multicomponent flow applications. (U.S.)

Numerical simulation of two-phase flow with front-capturing

International Nuclear Information System (INIS)

Tzanos, C.P.; Weber, D.P.

2000-01-01

Because of the complexity of two-phase flow phenomena, two-phase flow codes rely heavily on empirical correlations. This approach has a number of serious shortcomings. Advances in parallel computing and continuing improvements in computer speed and memory have stimulated the development of numerical simulation tools that rely less on empirical correlations and more on fundamental physics. The objective of this work is to take advantage of developments in massively parallel computing, single-phase computational fluid dynamics of complex systems, and numerical methods for front capturing in two-phase flows to develop a computer code for direct numerical simulation of two-phase flow. This includes bubble/droplet transport, interface deformation and topology change, bubble-droplet interactions, interface mass, momentum, and energy transfer. In this work, the Navier-Stokes and energy equations are solved by treating both phases as a single fluid with interfaces between the two phases, and a discontinuity in material properties across the moving interfaces. The evolution of the interfaces is simulated by using the front capturing technique of the level-set methods. In these methods, the boundary of a two-fluid interface is modeled as the zero level set of a smooth function φ. The level-set function φ is defined as the signed distance from the interface (φ is negative inside a droplet/bubble and positive outside). Compared to other front-capturing or front-tracking methods, the level-set approach is relatively easy to implement even in three-dimensional flows, and it has been shown to simulate well the coalescence and breakup of droplets/bubbles

Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

Energy Technology Data Exchange (ETDEWEB)

Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

2011-09-15

In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

International Nuclear Information System (INIS)

Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

2011-09-01

In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

Identification of the two-phase flow in the upper part of a boiling water reactor core using reactor noise analysis

International Nuclear Information System (INIS)

Miteff, L.

1983-08-01

The starting point of this work were neutron flux correlation measurements in the core of the Muehleberg boiling water reactor. During these measurements a new effect was observed i.e. that the cross power spectral density (CPSD) phases could be approximated by two straight lines of different slopes. The two linear domains of the CPSD-phases are separated along the frequency axis by a transition interval. It was supposed that such CPSD-phases could account for the propagation in the axial direction of two different thermohydraulic perturbations in the upper part of the Muehleberg BWR-core. It was also assumed that the second linear domain of these CPSD-phases could be related to a characteristic property of an annular flow regime of the steam-water two-phase flows in the bundles around the neutron detectors. The author attempts to give an explanation for the existence of a second transport phenomenon. This was achieved by in-core and out-core correlation measurements as well as by theoretical work. The out-core measurements were performed on an water-air simulation loop by use of laser beams. (Auth.)

Metrology of two-phase flow: different methods

International Nuclear Information System (INIS)

Delhaye, J.M.; Galaup, J.P.; Reocreux, M.; Ricque, R.

Nine papers are presented concerning different methods of measuring two-phase flow. Some of the methods and equipment discussed include: radiation absorption, electromagnetic flowmeter, anemometry, resistance probes, phase indicating microthermocouples, optical probes, sampling methods, and pitot tubes

Measurement of phase interaction in dispersed gas-particle two-phase flow by phase-doppler anemometry

Directory of Open Access Journals (Sweden)

Mergheni Ali Mohamed

2008-01-01

Full Text Available For simultaneous measurement of size and velocity distributions of continuous and dispersed phases in a two-phase flow a technique phase-Doppler anemometry was used. Spherical glass particles with a particle diameter range from 102 up to 212 µm were used. In this two-phase flow an experimental results are presented which indicate a significant influence of the solid particles on the flow characteristics. The height of influence of these effects depends on the local position in the jet. Near the nozzle exit high gas velocity gradients exist and therefore high turbulence production in the shear layer of the jet is observed. Here the turbulence intensity in the two-phase jet is decreased compared to the single-phase jet. In the developed zone the velocity gradient in the shear layer is lower and the turbulence intensity reduction is higher. .

Characteristics of low-mass-velocity vertical gas-liquid two-phase flow

International Nuclear Information System (INIS)

Adachi, Hiromichi; Abe, Yutaka; Kimura, Ko-ji

1995-01-01

In the present paper, characteristics of low mass velocity two-phase flow was analyzed based on a concept that pressure energy of two-phase flow is converted into acceleration work, gravitational work and frictional work, and the pressure energy consumption rate should be minimum at the stable two-phase flow condition. Experimental data for vertical upward air-water two-phase flow at atmospheric pressure was used to verify this concept and the turbulent model used in this method is optimized with the data. (author)

Probabilistic physical characteristics of phase transitions at highway bottlenecks: incommensurability of three-phase and two-phase traffic-flow theories.

Science.gov (United States)

Kerner, Boris S; Klenov, Sergey L; Schreckenberg, Michael

2014-05-01

Physical features of induced phase transitions in a metastable free flow at an on-ramp bottleneck in three-phase and two-phase cellular automaton (CA) traffic-flow models have been revealed. It turns out that at given flow rates at the bottleneck, to induce a moving jam (F → J transition) in the metastable free flow through the application of a time-limited on-ramp inflow impulse, in both two-phase and three-phase CA models the same critical amplitude of the impulse is required. If a smaller impulse than this critical one is applied, neither F → J transition nor other phase transitions can occur in the two-phase CA model. We have found that in contrast with the two-phase CA model, in the three-phase CA model, if the same smaller impulse is applied, then a phase transition from free flow to synchronized flow (F → S transition) can be induced at the bottleneck. This explains why rather than the F → J transition, in the three-phase theory traffic breakdown at a highway bottleneck is governed by an F → S transition, as observed in real measured traffic data. None of two-phase traffic-flow theories incorporates an F → S transition in a metastable free flow at the bottleneck that is the main feature of the three-phase theory. On the one hand, this shows the incommensurability of three-phase and two-phase traffic-flow theories. On the other hand, this clarifies why none of the two-phase traffic-flow theories can explain the set of fundamental empirical features of traffic breakdown at highway bottlenecks.

Geometrical automata for two phase flow simulation

International Nuclear Information System (INIS)

Herrero, V.; Guido-Lavalle, G.; Clausse, A.

1996-01-01

An automaton is an entity defined by a mathematical state which changes following iterative rules representing the interaction with the neighborhood. A model of automata for two-phase flow simulation consisting in a field of disks which are allowed to change their radii and move in a plane is presented. The model is more general than the classical cellular automata in two respects: (1) the grid of cellular automata is dismissed in favor of a trajectory generator; and (2) the rules of interaction involve parameters intended to represent some of the most relevant variables governing the actual physical interactions between phases. Computational experiments show that the algorithm captures the essential physics underlying two-phase flow problems such as bubbly-slug pattern transition and void fraction development along tubes. A comparison with experimental data of void fraction profiles is presented, showing excellent agreement. (orig.)

Interfacial heat transfer in countercurrent flows of steam and water

International Nuclear Information System (INIS)

Megahed, M.M.

1987-04-01

A study was conducted to examine the departure from equilibrium conditions with respect to direct contact condensation. A simple analytical model, which used an equilibrium factor, K, was derived. The model was structured to represent the physical dimensions of a nuclear reactor downcomer annulus, water subcooling, wall temperature, and water flow rate. In a two step process the model was first used to isolate the average interfacial heat transfer coefficient from vertical countercurrent steam/water data of Cook et al., with the aid of a Stanton number correlation. In the second step the model was assessed by regeneration of measured steam flow rates in the experiments by Cook et al., and an additional experiment of Kim. This report documents the analytical model, the derived Stanton number correlation, and the comparison of the calculated and measured steam flow rates by which the accuracy of the model was assessed

Random signal tomographical analysis of two-phase flow

International Nuclear Information System (INIS)

Han, P.; Wesser, U.

1990-01-01

This paper reports on radiation tomography which is a useful tool for studying the internal structures of two-phase flow. However, general tomography analysis gives only time-averaged results, hence much information is lost. As a result, it is sometimes difficult to identify the flow regime; for example, the time-averaged picture does not significantly change as an annual flow develops from a slug flow. A two-phase flow diagnostic technique based on random signal tomographical analysis is developed. It extracts more information by studying the statistical variation of the measured signal with time. Local statistical parameters, including mean value, variance, skewness and flatness etc., are reconstructed from the information obtained by a general tomography technique. More important information are provided by the results. Not only the void fraction can be easily calculated, but also the flow pattern can be identified more objectively and more accurately. The experimental setup is introduced. It consisted of a two-phase flow loop, an X-ray system, a fan-like five-beam detector system and a signal acquisition and processing system. In the experiment, for both horizontal and vertical test sections (aluminum and steel tube with Di/Do = 40/45 mm), different flow situations are realized by independently adjusting air and water mass flow. Through a glass tube connected with the test section, some typical flow patterns are visualized and used for comparing with the reconstruction results

Two-phased flow component loss data

International Nuclear Information System (INIS)

Fairhurst, C.P.

1983-01-01

Pressure loss measurements were made for valves and orifice plates under horizontal and vertical two-phase, air/water flow. The results displayed similar trends and were successfully correlated using a semi-empirical approach. (author)

Film boiling heat transfer from a hot sphere falling in two-phase pool

International Nuclear Information System (INIS)

Bang, K. H.; Kim, K. Y.

1998-01-01

The purpose of the present study is to experimentally investigate film boiling heat trasfer from a hot sphere falling in steam-water two-phase pool, which is the key heat transfer mode in molten fuel and coolant mixing. To measure film boiling heat transfer coefficients on a spere falling in two-phase pool, a heated sphere with a thermocouple embedded at the center is dropped in a vertical tube filled with steam-water mixture. The present experiment is unique in making the heated sphere fall through the two-phase pool while the previous experiments were performed with stationary spheres in flowing fluid. The falling speed of the sphere is measured using a set of magnet pickup coils distributed along the tube. The ranges of the experimental conditions are: spere fall speed 0-0.5 m/s, average void fraction 0-25,% steam superficial velocity 0-0.25 m/s. The results show that the forced convection film boiling heat transfer coefficient decrease slightly as the steam superficial velocity (void fraction) is increased

Flow visualization of two-phase flows using photochromic dye activation method

International Nuclear Information System (INIS)

Kawaji, M.; Ahmad, W.; DeJesus, J.M.; Sutharshan, B.; Lorencez, C.; Ojha, M.

1993-01-01

A non-intrusive flow visualization technique based on light activation of photochromic dye material has been used to obtain velocity profiles in gas-liquid flows including annular, slug and stratified flows. The preliminary results revealed several important two-phase flow mechanisms that have not been clearly seen previously. (orig.)

Hydrodynamics of single- and two-phase flow in inclined rod arrays

International Nuclear Information System (INIS)

Ebeling-Koning, D.B.; Todreas, N.E.

1983-09-01

Required inputs for thermal-hydraulic codes are constitutive relations for fluid-solid flow resistance, in single-phase flow, and interfacial momentum exchange (relative phase motion), in two-phase flow. An inclined rod array air-water experiment was constructed to study the hydrodynamics of multidimensional porous medium flow in rod arrays. Velocities, pressures, and bubble distributions were measured in square rod arrays of P/d = 1.5, at 0, 30, 45, and 90 degree inclinations to the vertical flow direction. Constitutive models for single-phase flow resistance are reviewed, new comprehensive models developed, and an assessment with previously published and new data made. The principle of superimposing one-dimensional correlations proves successful for turbulent single-phase inclined flow. For bubbly two-phase incline flow a new flow separation phenomena was observed and modeled. A two-region liquid velocity model is developed to explain the experimentally observed phenomena. Fundamental data for bubbles rising in rod arrays were also taken

Interfacial Instability in Two-Phase Flow: Manipulating Coalescence and Condensation

Data.gov (United States)

National Aeronautics and Space Administration — Two-phase flow under microgravity conditions presents a number of technical challenges ( and ). Life support and habitation depend on systems that use two-phase flow...

Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system

Energy Technology Data Exchange (ETDEWEB)

Song, Ki Won [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Nguyen, Thanh Hung [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Ha, Kwang Soon; Kim, Hwan Yeol; Song, Jinho [Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Park, Hyun Sun [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Revankar, Shripad T., E-mail: shripad@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Institute of Nuclear Safety, Daejeon 305-338 (Korea, Republic of)

2017-05-15

Highlights: • Two-phase flow regimes and transition behavior were observed in the coolant channel. • Test were conducted for natural circulation with air-water. • Data were obtained on flow regime, void fraction, flow rates and re-wetting time. • The data were related to a cooling capability of core catcher system. - Abstract: Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ‘super-step’ design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system.

Signal analysis of steam line acoustics

International Nuclear Information System (INIS)

Martin, C. Samuel

2003-01-01

The vibration of nuclear steam piping is usually associated with pressure fluctuations emanating from flow disturbances such as steam generator nozzles, bends, or other pipe fittings. Flow separation at pipe tees and within steam chest manifolds or headers generate pressure fluctuations that propagate both upstream to steam generators as well as downstream to the steam turbine. Steady-state acoustic oscillations at various frequencies occur within the piping, possibly exciting structural vibrations. This paper focuses on the assessment of the origin of the disturbances using signal analyses of two dynamic pressure recordings from pressure transducers located along straight runs in the steam piping. The technique involves performing the cross spectrum to two dynamic pressure signals in piping between (1) the steam generator and steam chest header, and (2) between the header and steam turbine outlet. If, at a specified frequency, no causality occurs between the two signals then the cross spectra magnitude will be negligible. Of interest here is the value of the phase between the two signals for frequencies for which the magnitude of the cross spectrum is not negligible. It is shown in the paper that the direction of the dominant waves at all frequencies can be related to the phase angle from the cross spectrum. It has to be realized that pressure waves emanating from one source such as a steam generator will propagate along uniform steam pipes with little transformation or attenuation, but will be reflected at fittings and at inlets and outlets. Hence, the eventual steady-state time record at a given location in the piping is a result of not only the disturbance, but also reflections of earlier pulsations. Cross-spectral analyses has been employed to determine the direction of the dominant acoustic waves in the piping for various frequencies for which there are signals. To prove the technique, synthetic spectra are generated comprised of harmonic waves moving both

Two-phase flow instabilities in a silicon microchannels heat sink

International Nuclear Information System (INIS)

Bogojevic, D.; Sefiane, K.; Walton, A.J.; Lin, H.; Cummins, G.

2009-01-01

Two-phase flow instabilities are highly undesirable in microchannels-based heat sinks as they can lead to temperature oscillations with high amplitudes, premature critical heat flux and mechanical vibrations. This work is an experimental study of boiling instabilities in a microchannel silicon heat sink with 40 parallel rectangular microchannels, having a length of 15 mm and a hydraulic diameter of 194 μm. A series of experiments have been carried out to investigate pressure and temperature oscillations during the flow boiling instabilities under uniform heating, using water as a cooling liquid. Thin nickel film thermometers, integrated on the back side of a heat sink with microchannels, were used in order to obtain a better insight related to temperature fluctuations caused by two-phase flow instabilities. Flow regime maps are presented for two inlet water temperatures, showing stable and unstable flow regimes. It was observed that boiling leads to asymmetrical flow distribution within microchannels that result in high temperature non-uniformity and the simultaneously existence of different flow regimes along the transverse direction. Two types of two-phase flow instabilities with appreciable pressure and temperature fluctuations were observed, that depended on the heat to mass flux ratio and inlet water temperature. These were high amplitude/low frequency and low amplitude/high frequency instabilities. High speed camera imaging, performed simultaneously with pressure and temperature measurements, showed that inlet/outlet pressure and the temperature fluctuations existed due to alternation between liquid/two-phase/vapour flows. It was also determined that the inlet water subcooling condition affects the magnitudes of the temperature oscillations in two-phase flow instabilities and flow distribution within the microchannels.

Experimental study of steam bubble velocities and dimensions in the draught trunk of the AST-500 reactor simulator

International Nuclear Information System (INIS)

Shanin, V.K.; Drobkov, V.P.; Kulakov, I.V.; Khalmeh, M.V.

1988-01-01

Local characteristics for two-phase steam water flow in the vertical channel with 0.45 m diameter and 2 m length, which is the draught trunk of the AST-500 reactor simulator, are investigated. Steam bubble velocities and dimensions were determined by the time-of-flight method using the twinned conductometric transducers. The data obtained testify to the existance of unstable circulation flows in the trunk peripheral region. These flows effect considerably the steam phase motion in the trunk middle part. At the same time the circulation flows to a lesser degree affect steam bubble motion in the trunk low peripheral part and to the lesser degree affect the steam phase in the axial zone near the outlet from the heating section. So the data obtained confirm the conclusion, made earlier, about steam-water flow acceleration in the draught trunk central part

Guidelines for random excitation forces due to cross flow in steam generators

Energy Technology Data Exchange (ETDEWEB)

Taylor, C.E.; Pettigrew, M.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

1998-07-01

Random excitation forces can cause low-amplitude tube motion that will result in long-term fretting-wear or fatigue. To prevent these tube failures in steam generators and other heat exchangers, designers and trouble-shooters must have guidelines that incorporate random or turbulent fluid forces. Experiments designed to measure fluid forces have been carried out at Chalk River Laboratories and at other labs around the world. The data from these experiments have been studied and collated to determine suitable guidelines for random excitation forces. In this paper, a guideline for random excitation forces in single-phase cross flow is presented in the form of normalised spectra that are applicable to a wide range of flow conditions and tube frequencies. In particular, the experimental results used in this study were carried out over the full range of flow conditions found in a nuclear steam generator. The proposed guidelines are applicable to steam generators, condensers, reheaters and other shell-and-tube heat exchangers. They may be used for flow-induced vibration analysis of new or existing components, as input to vibration analysis computer codes and as specifications in procurement documents. (author)

Guidelines for random excitation forces due to cross flow in steam generators

International Nuclear Information System (INIS)

Taylor, C.E.; Pettigrew, M.J.

1998-01-01

Random excitation forces can cause low-amplitude tube motion that will result in long-term fretting-wear or fatigue. To prevent these tube failures in steam generators and other heat exchangers, designers and trouble-shooters must have guidelines that incorporate random or turbulent fluid forces. Experiments designed to measure fluid forces have been carried out at Chalk River Laboratories and at other labs around the world. The data from these experiments have been studied and collated to determine suitable guidelines for random excitation forces. In this paper, a guideline for random excitation forces in single-phase cross flow is presented in the form of normalised spectra that are applicable to a wide range of flow conditions and tube frequencies. In particular, the experimental results used in this study were carried out over the full range of flow conditions found in a nuclear steam generator. The proposed guidelines are applicable to steam generators, condensers, reheaters and other shell-and-tube heat exchangers. They may be used for flow-induced vibration analysis of new or existing components, as input to vibration analysis computer codes and as specifications in procurement documents. (author)

One-dimensional two-phase thermal hydraulics (ENSTA course); Thermo-hydraulique diphasique monodimensionnelle. Cours ENSTA

Energy Technology Data Exchange (ETDEWEB)

Olive, J

1995-11-01

This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends.

Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

International Nuclear Information System (INIS)

Panajotov, D.P.; Gorburov, V.I.

1989-01-01

Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

Test for Jet Flow Induced by Steam Jet Condensation Using the GIRLS Facility

International Nuclear Information System (INIS)

Kim, Yeon Sik; Yoon, Y. J.; Song, C. H.

2007-03-01

To investigate the characteristics of the turbulent jet induced by steam jet condensation in a water tank through a single-hole sparger an experimental investigation was performed using the GIRLS facility. The experiments were conducted with respect to two cases, e.g. horizontal and vertical upward injections. For the measurements, pitot tube and thermocouples were used for turbulent flow velocity and temperatures, respectively. Overall flow shapes of the turbulent jet by the steam jet condensation are similar to those of axially symmetric turbulent jet flows. The angular coefficients of turbulent rays are quantitatively comparable between the traditional turbulent jet flows and the turbulent jet flows induced by the steam jet condensation in this work. Although the turbulent flows were induced by the horizontally injected steam jet condensation, general theory of turbulent jets was found to be applicable to the turbulent flows of this work. But for the vertically upward injection case, experimental data were quite deviated from the theoretical ones, which is considered due to the buoyancy effect

Hydrodynamic Instability and Dynamic Burnout in Natural Circulation Two-Phase Flow. An Experimental and Theoretical Study

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Jahnberg, S; Haga, I; Hansson, P T; Mathisen, R P

1964-09-15

A theoretical model for predicting the threshold of instability for two phase flow in a natural circulation loop is presented. The model calculates the flow transient caused by a step disturbance of the heat input, and is based upon the conservation laws of mass, momentum and energy in one dimensional form. Empirical correlations are used in the model for estimating the void fractions and the two-phase flow pressure drops. The equations are solved numerically in a finite difference approximation coded for a digital computer. An experimental study of the hydrodynamic instability and dynamic burnout in two-phase flow has been performed in a natural circulation loop in the pressure range from 10 to 70 atg. The test sections were round ducts of 20, 30 and 36 mm inner diameter and 4890 mm heated length. The experimental results showed that within the ranges tested, the stability of the flow increases with increasing pressure and increasing throttling before the test section, but decreases with increasing Inlet subcooling and increasing throttling after the test section. Comparing the natural circulation burnout steam qualities with corresponding forced circulation data shoved that the former data were low by a factor up to 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data. The present experimental results as well as data available from other sources have been compared with the stability thresholds obtained with the theoretical model. The comparisons included circular, annular and rod cluster geometries, and the agreement between the experimental and theoretical stability limits was good. Finally the application of the experimental and theoretical results on the assessment of boiling heavy water reactor design is discussed.

Hydrodynamic Instability and Dynamic Burnout in Natural Circulation Two-Phase Flow. An Experimental and Theoretical Study

International Nuclear Information System (INIS)

Becker, Kurt M.; Jahnberg, S.; Haga, I.; Hansson, P.T.; Mathisen, R.P.

1964-09-01

A theoretical model for predicting the threshold of instability for two phase flow in a natural circulation loop is presented. The model calculates the flow transient caused by a step disturbance of the heat input, and is based upon the conservation laws of mass, momentum and energy in one dimensional form. Empirical correlations are used in the model for estimating the void fractions and the two-phase flow pressure drops. The equations are solved numerically in a finite difference approximation coded for a digital computer. An experimental study of the hydrodynamic instability and dynamic burnout in two-phase flow has been performed in a natural circulation loop in the pressure range from 10 to 70 atg. The test sections were round ducts of 20, 30 and 36 mm inner diameter and 4890 mm heated length. The experimental results showed that within the ranges tested, the stability of the flow increases with increasing pressure and increasing throttling before the test section, but decreases with increasing Inlet subcooling and increasing throttling after the test section. Comparing the natural circulation burnout steam qualities with corresponding forced circulation data shoved that the former data were low by a factor up to 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data. The present experimental results as well as data available from other sources have been compared with the stability thresholds obtained with the theoretical model. The comparisons included circular, annular and rod cluster geometries, and the agreement between the experimental and theoretical stability limits was good. Finally the application of the experimental and theoretical results on the assessment of boiling heavy water reactor design is discussed

Modelling of two-phase flow based on separation of the flow according to velocity

International Nuclear Information System (INIS)

Narumo, T.

1997-01-01

The thesis concentrates on the development work of a physical one-dimensional two-fluid model that is based on Separation of the Flow According to Velocity (SFAV). The conventional way to model one-dimensional two-phase flow is to derive conservation equations for mass, momentum and energy over the regions occupied by the phases. In the SFAV approach, the two-phase mixture is divided into two subflows, with as distinct average velocities as possible, and momentum conservation equations are derived over their domains. Mass and energy conservation are treated equally with the conventional model because they are distributed very accurately according to the phases, but momentum fluctuations follow better the flow velocity. Submodels for non-uniform transverse profile of velocity and density, slip between the phases within each subflow and turbulence between the subflows have been derived. The model system is hyperbolic in any sensible flow conditions over the whole range of void fraction. Thus, it can be solved with accurate numerical methods utilizing the characteristics. The characteristics agree well with the used experimental data on two-phase flow wave phenomena Furthermore, the characteristics of the SFAV model are as well in accordance with their physical counterparts as of the best virtual-mass models that are typically optimized for special flow regimes like bubbly flow. The SFAV model has proved to be applicable in describing two-phase flow physically correctly because both the dynamics and steady-state behaviour of the model has been considered and found to agree well with experimental data This makes the SFAV model especially suitable for the calculation of fast transients, taking place in versatile form e.g. in nuclear reactors

Two-phase flow phenomena in broken recirculation line of BWR

International Nuclear Information System (INIS)

Kato, Masami; Arai, Kenji; Narabayashi, Tadashi; Amano, Osamu.

1986-01-01

When a primary recirculation line of BWR is ruptured, a primary recirculation pump may be subjected to very high velocity two-phase flow and its speed may be accelerated by this flow. It is important for safety evaluation to estimate the pump behavior during blowdown. There are two problems involved in analyzing this behavior. One problem concerns the pump characteristics under two-phase flow. The other involves the two-phase conditions at the pump inlet. If the rupture occurs at a suction side of the pump, choking is considered to occur at a broken jet pump nozzle. Then, a void fraction becomes larger downstream from the jet pump nozzle and volumetric flow through the pump will be very high. However, there is little experimental data available on two-phase flow downstream from a choking plane. Blowdown tests were performed using a simulated broken recirculation line and measured data were analyzed by TRAC-PlA. Analytical results agreed with measured data. (author)

Pigging analysis for gas-liquid two phase flow in pipelines

International Nuclear Information System (INIS)

Kohda, K.; Suzukawa, Y.; Furukawa, H.

1988-01-01

A new method to analyze transient phenomena caused by pigging in gas-liquid two-phase flow is developed. During pigging, a pipeline is divided into three sections by two moving boundaries, namely the pig and the leading edge of the liquid slug in front of the pig. The basic equations are mass, momentum and energy conservation equations. The boundary conditions at the moving boundaries are determined from the mass conservation across the boundaries, etc. A finite difference method is used to solve the equations numerically. The method described above is also capable of analyzing transient two-phase flow caused by pressure and flow rate changes. Thus the over-all analysis of transient two-phase flow in pipelines becomes possible. A series of air-water two-phase flow pigging experiments was conducted using 105.3 mm diameter and 1436.5 m long test pipeline. The agreement between the measured and the calculated results is very good

PWR steam generator chemical cleaning. Phase II. Final report

International Nuclear Information System (INIS)

1980-01-01

Two techniques believed capable of chemically dissolving the corrosion products in the annuli between tubes and support plates were developed in laboratory work in Phase I of this project and were pilot tested in Indian Point Unit No. 1 steam generators. In Phase II, one of the techniques was shown to be inadequate on an actual sample taken from an Indian Point Unit No. 2 steam generator. The other technique was modified slightly, and it was demonstrated that the tube/support plate annulus could be chemically cleaned effectively

Molten salt steam generator subsystem research experiment. Volume I. Phase 1 - Final report

Energy Technology Data Exchange (ETDEWEB)

None

1984-10-01

A study was conducted for Phase 1 of a two-phase project whose objectives were to develop a reliable, cost-effective molten salt steam generating subsystem for solar thermal plants, minimize uncertainty in capital, operating, and maintenance costs, and demonstrate the ability of molten salt to generate high-pressure, high-temperature steam. The Phase 1 study involved the conceptual design of molten salt steam generating subsystems for a nominal 100-MWe net stand-alone solar central receiver electric generating plant, and a nominal 100-MWe net hybrid fossil-fueled electric power generating plant that is 50% repowered by a solar central receiver system. As part of Phase 1, a proposal was prepared for Phase 2, which involves the design, construction, testing and evaluation of a Subsystem Research Experiment of sufficient size to ensure successful operation of the full-size subsystem designed in Phase 1. Evaluation of several concepts resulted in the selection of a four-component (preheater, evaporator, superheater, reheater), natural circulation, vertically oriented, shell and tube (straight) heat exchanger arrangement. Thermal hydraulic analysis of the system included full and part load performance, circulation requirements, stability, and critical heat flux analysis. Flow-induced tube vibration, tube buckling, fatigue evaluation of tubesheet junctions, steady-state tubesheet analysis, and a simplified transient analysis were included in the structural analysis of the system. Operating modes and system dynamic response to load changes were identified. Auxiliary equipment, fabrication, erection, and maintenance requirements were also defined. Installed capital costs and a project schedule were prepared for each design.

Studies of thermal-hydrodynamic flow instability, (3)

International Nuclear Information System (INIS)

Suzuoki, Akira

1978-01-01

In the flow system in which large density change occurs midway, sometimes steady flow cannot be maintained according to the conditions, and pulsating flow or the scamper of flow occurs. This phenomenon is called flow instability, and is noticed as one of the causes to obstruct the normal operation in boilers, BWRs and the steam generators for FBRs with parallel evaporating tube system. In the pulsating instability, there are density wave oscillation and pressure wave oscillation. The author has studied the density wave oscillation occurring in the steam generators for FBRs and in this paper, the role played by two-phase flow regarding the occurrence of flow instability, and the effect of the existence of interphase slip on the role played by two-phase flow are reported. The theoretical analysis and the results of the analysis taking a steam generator heated with sodium as the example are described. Regarding flow stability, two-phase flow part generates the variation of weight velocity with different phase in steam single phase part, accepting enthalpy variation in water single phase part. In this action, the effect of interphase slip was observed, and the variation of reverse phase is apt to occur in slip flow as compared with homogeneous flow. Accordingly, flow instability is apt to occur in slip flow. (Kako, I.)

Non-Darcy behavior of two-phase channel flow.

Science.gov (United States)

Xu, Xianmin; Wang, Xiaoping

2014-08-01

We study the macroscopic behavior of two-phase flow in porous media from a phase-field model. A dissipation law is first derived from the phase-field model by homogenization. For simple channel geometry in pore scale, the scaling relation of the averaged dissipation rate with the velocity of the two-phase flow can be explicitly obtained from the model which then gives the force-velocity relation. It is shown that, for the homogeneous channel surface, Dacry's law is still valid with a significantly modified permeability including the contribution from the contact line slip. For the chemically patterned surfaces, the dissipation rate has a non-Darcy linear scaling with the velocity, which is related to a depinning force for the patterned surface. Our result offers a theoretical understanding on the prior observation of non-Darcy behavior for the multiphase flow in either simulations or experiments.

Measurements of liquid-phase turbulence in gas–liquid two-phase flows using particle image velocimetry

International Nuclear Information System (INIS)

Zhou, Xinquan; Doup, Benjamin; Sun, Xiaodong

2013-01-01

Liquid-phase turbulence measurements were performed in an air–water two-phase flow loop with a circular test section of 50 mm inner diameter using a particle image velocimetry (PIV) system. An optical phase separation method-–planar laser-induced fluorescence (PLIF) technique—which uses fluorescent particles and an optical filtration technique, was employed to separate the signals of the fluorescent seeding particles from those due to bubbles and other noises. An image pre-processing scheme was applied to the raw PIV images to remove the noise residuals that are not removed by the PLIF technique. In addition, four-sensor conductivity probes were adopted to measure the radial distribution of the void fraction. Two benchmark tests were performed: the first was a comparison of the PIV measurement results with those of similar flow conditions using thermal anemometry from previous studies; the second quantitatively compared the superficial liquid velocities calculated from the local liquid velocity and void fraction measurements with the global liquid flow rate measurements. The differences of the superficial liquid velocity obtained from the two measurements were bounded within ±7% for single-phase flows and two-phase bubbly flows with the area-average void fraction up to 18%. Furthermore, a preliminary uncertainty analysis was conducted to investigate the accuracy of the two-phase PIV measurements. The systematic uncertainties due to the circular pipe curvature effects, bubble surface reflection effects and other potential uncertainty sources of the PIV measurements were discussed. The purpose of this work is to facilitate the development of a measurement technique (PIV-PLIF) combined with image pre-processing for the liquid-phase turbulence in gas–liquid two-phase flows of relatively high void fractions. The high-resolution data set can be used to more thoroughly understand two-phase flow behavior, develop liquid-phase turbulence models, and assess high

Thermal-hydraulics in recirculating steam generators

International Nuclear Information System (INIS)

Carver, M.B.; Carlucci, L.N.; Inch, W.W.R.

1981-04-01

This manual describes the THIRST code and its use in computing three-dimensional two-phase flow and heat transfer in a steam generator under steady state operation. The manual is intended primarily to facilitate the application of the code to the analysis of steam generators typical of CANDU nuclear stations. Application to other steam generator designs is also discussed. Details of the assumptions used to formulate the model and to implement the numerical solution are also included

Comparison of differential pressure model based on flow regime for gas/liquid two-phase flow

International Nuclear Information System (INIS)

Dong, F; Zhang, F S; Li, W; Tan, C

2009-01-01

Gas/liquid two-phase flow in horizontal pipe is very common in many industry processes, because of the complexity and variability, the real-time parameter measurement of two-phase flow, such as the measurement of flow regime and flow rate, becomes a difficult issue in the field of engineering and science. The flow regime recognition plays a fundamental role in gas/liquid two-phase flow measurement, other parameters of two-phase flow can be measured more easily and correctly based on the correct flow regime recognition result. A multi-sensor system is introduced to make the flow regime recognition and the mass flow rate measurement. The fusion system is consisted of temperature sensor, pressure sensor, cross-section information system and v-cone flow meter. After the flow regime recognition by cross-section information system, comparison of four typical differential pressure (DP) models is discussed based on the DP signal of v-cone flow meter. Eventually, an optimum DP model has been chosen for each flow regime. The experiment result of mass flow rate measurement shows it is efficient to classify the DP models by flow regime.

Development of One Dimensional Hyperbolic Coupled Solver for Two-Phase Flows

International Nuclear Information System (INIS)

Kim, Eoi Jin; Kim, Jong Tae; Jeong, Jae June

2008-08-01

The purpose of this study is a code development for one dimensional two-phase two-fluid flows. In this study, the computations of two-phase flow were performed by using the Roe scheme which is one of the upwind schemes. The upwind scheme is widely used in the computational fluid dynamics because it can capture discontinuities clearly such as a shock. And this scheme is applicable to multi-phase flows by the extension methods which were developed by Toumi, Stadtke, etc. In this study, the extended Roe upwind scheme by Toumi for two-phase flow was implemented in the one-dimensional code. The scheme was applied to a shock tube problem and a water faucet problem. This numerical method seems efficient for non oscillating solutions of two phase flow problems, and also capable for capturing discontinuities

Development of One Dimensional Hyperbolic Coupled Solver for Two-Phase Flows

Energy Technology Data Exchange (ETDEWEB)

Kim, Eoi Jin; Kim, Jong Tae; Jeong, Jae June

2008-08-15

The purpose of this study is a code development for one dimensional two-phase two-fluid flows. In this study, the computations of two-phase flow were performed by using the Roe scheme which is one of the upwind schemes. The upwind scheme is widely used in the computational fluid dynamics because it can capture discontinuities clearly such as a shock. And this scheme is applicable to multi-phase flows by the extension methods which were developed by Toumi, Stadtke, etc. In this study, the extended Roe upwind scheme by Toumi for two-phase flow was implemented in the one-dimensional code. The scheme was applied to a shock tube problem and a water faucet problem. This numerical method seems efficient for non oscillating solutions of two phase flow problems, and also capable for capturing discontinuities.

Phenomenological studies of two-phase flow processes for nuclear waste isolation

International Nuclear Information System (INIS)

Pruess, K.; Finsterle, S.; Persoff, P.; Oldenburg, C.

1994-01-01

The US civilian radioactive waste management program is unique in its focus on a site in the unsaturated zone, at Yucca Mountain, Nevada. Two-phase flow phenomena can also play an important role in repositories beneath the water table where gas is generated by corrosion, hydrolysis, and biological degradation of the waste packages. An integrated program has been initiated to enhance our understanding of two-phase flow behavior in fractured rock masses. The studies include two-phase (gas-liquid) flow experiments in laboratory specimens of natural rock fractures, analysis and modeling of heterogeneity and instability effects in two-phase flow, and design and interpretation of field experiments by means of numerical simulation. We present results that identify important aspects of two-phase flow behavior on different space and time scales which are relevant to nuclear waste disposal in both unsaturated and saturated formations

Application of non-equilibrium thermodynamics to two-phase flows with a change of phase

International Nuclear Information System (INIS)

Delhaye, J.M.

1969-01-01

In this report we use the methods of non-equilibrium thermodynamics in two-phase flows. This paper follows a prior one in which we have studied the conservation laws and derived the general equations of two-phase flow. In the first part the basic ideas of thermodynamics of irreversible systems are given. We follow the classical point of view. The second part is concerned with the derivation of a closed set of equations for the two phase elementary volume model. In this model we assume that the elementary volume contains two phases and that it is possible to define a volumetric local concentration. To obtain the entropy balance we can choose either the reversibility of the barycentric motion or the reversibility of each phase. We adopt the last assumption and our derivation is the same as this of I.Prigogine and P. Mazur about the hydrodynamics of liquid helium. The scope of this work is not to find a general solution to the problems of two phase flows but to obtain a new set of equations which may be used to explain some characteristic phenomena of two-phase flow such as wave propagation or critical states. (author) [fr

A model of single and two-phase flow (critical or not) through cracks

International Nuclear Information System (INIS)

Seynhaeve, J.M.; Giot, M.; Granger, S.; Pages, D.

1995-07-01

The leaks through steam-generator cracks are the subject of research carried out in cooperation between EDF and UCL. A model to predict the mass flow rate with inlet subcooling has been developed, validated and published. The model takes into account the persistence of some metastable liquid in the crack. The present paper improves and extends the model, by making it applicable to all kinds of conditions prevailing in the S.G. tubes: not only subcooled water, but also saturated water, steam-water mixtures, saturated dry steam or superheated steam. Therefore, the flow at the crack inlet is analyzed and appropriate methods to initialize the numerical integration of the flow equations along the crack are proposed. The extensions of the model are still in the process of validation. However, a sensitivity analysis of its results has been made and is presented. (author)

Dynamics of the free surface of stratified two-phase flows in channels with rectangular cross-sections

International Nuclear Information System (INIS)

Vallee, Christophe

2012-01-01

Stratified two-phase flows were investigated at different test facilities with horizontal test sections in order to provide an experimental database for the development and validation of computational fluid dynamics (CFD) codes. These channels were designed with rectangular cross-sections to enable optimal observation conditions for the application of optical measurement techniques. Consequently, the local flow structure was visualised with a high-speed video camera, delivering data with highresolution in space and time as needed for CFD code validation. Generic investigations were performed at atmospheric pressure and room temperature in two air/water channels made of acrylic glass. Divers preliminary experiments were conducted with various measuring systems in a test section mounted between two separators. The second test facility, the Horizontal Air/Water Channel (HAWAC), is dedicated to co-current flow investigations. The hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was studied in this closed channel. Moreover, the instable wave growth leading to slug flow was investigated from the test section inlet. For quantitative analysis of the optical measurements, an algorithm was developed to recognise the stratified interface in the camera frames, allowing statistical treatments for comparison with CFD calculation results. The third test apparatus was installed in the pressure chamber of the TOPFLOW test facility in order to be operated at reactor typical conditions under pressure equilibrium with the vessel atmosphere. The test section representing a flat model of the hot leg of the German Konvoi pressurised water reactor (PWR) scaled at 1:3 is equipped with large glass side walls in the region of the elbow and of the steam generator inlet chamber to allow visual observations. The experiments were conducted with air and water at room temperature and maximum pressures of 3 bar as well as with steam and water at

Dynamics of the free surface of stratified two-phase flows in channels with rectangular cross-sections

Energy Technology Data Exchange (ETDEWEB)

Vallee, Christophe

2012-08-22

Stratified two-phase flows were investigated at different test facilities with horizontal test sections in order to provide an experimental database for the development and validation of computational fluid dynamics (CFD) codes. These channels were designed with rectangular cross-sections to enable optimal observation conditions for the application of optical measurement techniques. Consequently, the local flow structure was visualised with a high-speed video camera, delivering data with highresolution in space and time as needed for CFD code validation. Generic investigations were performed at atmospheric pressure and room temperature in two air/water channels made of acrylic glass. Divers preliminary experiments were conducted with various measuring systems in a test section mounted between two separators. The second test facility, the Horizontal Air/Water Channel (HAWAC), is dedicated to co-current flow investigations. The hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was studied in this closed channel. Moreover, the instable wave growth leading to slug flow was investigated from the test section inlet. For quantitative analysis of the optical measurements, an algorithm was developed to recognise the stratified interface in the camera frames, allowing statistical treatments for comparison with CFD calculation results. The third test apparatus was installed in the pressure chamber of the TOPFLOW test facility in order to be operated at reactor typical conditions under pressure equilibrium with the vessel atmosphere. The test section representing a flat model of the hot leg of the German Konvoi pressurised water reactor (PWR) scaled at 1:3 is equipped with large glass side walls in the region of the elbow and of the steam generator inlet chamber to allow visual observations. The experiments were conducted with air and water at room temperature and maximum pressures of 3 bar as well as with steam and water at

Measurement of two phase flow properties using the nuclear reactor instruments

International Nuclear Information System (INIS)

Albrecht, R.W.; Washington Univ., Seattle; Crowe, R.D.; Dailey, D.J.; Kosaly, G.; Damborg, M.J.

1982-01-01

A procedure is introduced for characterizing one dimensional, two phase flow in terms of three properties; propagation, structure, and dynamics. It is shown that all of these properties can be measured by analyzing the response of the reactor neutron field to a two phase flow perturbation. Therefore, a nuclear reactor can be regarded as a two phase flow instrument. (author)

Void fraction measurements in two-phase flow by transmission and scattering of a neutrons beam

International Nuclear Information System (INIS)

Souza, M.C.L.

1984-01-01

Calibration curves have been obtained which supply average values of void fraction (α) of water-steam two-phase mixtures for bubble, slug, annular and invert annular flow states. The measurements were carried out in simulated models of lucite-air for the steady-state, using the techniques of transmission and diffusion of a thermal neutrons beam. The calibration curves obtained were used for measurements of void fraction in a circuit containing two-phase water-air mixtures, in upward concurrent flow, for slug flow (P sub(max) = 1,06 bar) and annular flow (P sub(max) = 1,33 bar), using the same techniques. In both of the systems, a test section made up of an aluminium (99,9%) tube was used with internal diameter of 25,25 mm and 2,0 mm wall thichness. The beam of neutrons was obtained from a 5 Ci isotopic Am-Be source, thermalised in a cylindrical moderator of paraffin of 500 mm diameter (with H/D=1) which was covered by 2 mm thick cadmium sheets and having in its centre a parallepeliped made from high density polyethilene with the dimensions 240 x 240 x 144 mm. The neutrons escape through a rectangular collimator of 53,0 x 25,25 mm, with a length of 273 mm cut out of a single block of borated paraffin (32% of H 3 BO 3 ). The experimental results are in good agreement with theorical models in published literature. (Author) [pt

Real-time dynamic analysis for complete loop of direct steam generation solar trough collector

International Nuclear Information System (INIS)

Guo, Su; Liu, Deyou; Chu, Yinghao; Chen, Xingying; Shen, Bingbing; Xu, Chang; Zhou, Ling; Wang, Pei

2016-01-01

Highlights: • A nonlinear distribution parameter dynamic model has been developed. • Real-time local heat transfer coefficient and friction coefficient are adopted. • The dynamic behavior of the solar trough collector loop are simulated. • High-frequency chattering of outlet fluid flow are analyzed and modeled. • Irradiance disturbance at subcooled water region generates larger influence. - Abstract: Direct steam generation is a potential approach to further reduce the levelized electricity cost of solar trough. Dynamic modeling of the collector loop is essential for operation and control of direct steam generation solar trough. However, the dynamic behavior of fluid based on direct steam generation is complex because of the two-phase flow in the pipeline. In this work, a nonlinear distribution parameter model has been developed to model the dynamic behaviors of direct steam generation parabolic trough collector loops under either full or partial solar irradiance disturbance. Compared with available dynamic model, the proposed model possesses two advantages: (1) real-time local values of heat transfer coefficient and friction resistance coefficient, and (2) considering of the complete loop of collectors, including subcooled water region, two-phase flow region and superheated steam region. The proposed model has shown superior performance, particularly in case of sensitivity study of fluid parameters when the pipe is partially shaded. The proposed model has been validated using experimental data from Solar Thermal Energy Laboratory of University of New South Wales, with an outlet fluid temperature relative error of only 1.91%. The validation results show that: (1) The proposed model successfully outperforms two reference models in predicting the behavior of direct steam generation solar trough. (2) The model theoretically predicts that, during solar irradiance disturbance, the discontinuities of fluid physical property parameters and the moving back and

Modelling of two-phase flow based on separation of the flow according to velocity

Energy Technology Data Exchange (ETDEWEB)

Narumo, T. [VTT Energy, Espoo (Finland). Nuclear Energy

1997-12-31

The thesis concentrates on the development work of a physical one-dimensional two-fluid model that is based on Separation of the Flow According to Velocity (SFAV). The conventional way to model one-dimensional two-phase flow is to derive conservation equations for mass, momentum and energy over the regions occupied by the phases. In the SFAV approach, the two-phase mixture is divided into two subflows, with as distinct average velocities as possible, and momentum conservation equations are derived over their domains. Mass and energy conservation are treated equally with the conventional model because they are distributed very accurately according to the phases, but momentum fluctuations follow better the flow velocity. Submodels for non-uniform transverse profile of velocity and density, slip between the phases within each subflow and turbulence between the subflows have been derived. The model system is hyperbolic in any sensible flow conditions over the whole range of void fraction. Thus, it can be solved with accurate numerical methods utilizing the characteristics. The characteristics agree well with the used experimental data on two-phase flow wave phenomena Furthermore, the characteristics of the SFAV model are as well in accordance with their physical counterparts as of the best virtual-mass models that are typically optimized for special flow regimes like bubbly flow. The SFAV model has proved to be applicable in describing two-phase flow physically correctly because both the dynamics and steady-state behaviour of the model has been considered and found to agree well with experimental data This makes the SFAV model especially suitable for the calculation of fast transients, taking place in versatile form e.g. in nuclear reactors. 45 refs. The thesis includes also five previous publications by author.

Flooding and flow reversal of two-phase annular flow

International Nuclear Information System (INIS)

Asahi, Y.

1978-01-01

The flooding and flow reversal conditions of two-phase annular flow are mathematically defined in terms of a characteristic function representing a force balance. Sufficiently below the flooding point in counter-current flow, the interface is smooth and the characteristic equation reduces to the Nusselt relationship. Just below flooding point and above the flow reversal point in cocurrent flow, the interface is 'wavy', so that the interfacial shear effect plays an important role. The theoretical analysis is compared with experimental results by others. It is suggested that the various length effects which have been experimentally observed may be accounted for by the spatial variation of the droplet entrainment. (Auth.)

Flow oscillations on the steam control valve in the middle opening condition. Clarification of the phenomena by steam flow experiment and CFD calculation

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio

2006-01-01

A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop improvements by Computational Fluid Dynamics (CFD) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the way to prevent the pressure fluctuations by experiments and CFD calculations. But, as we used air as a working fluid in our previous research instead of steam that is used in the power plant, we couldn't consider effects of condensation and difference of change of the quantity of state between air and steam. In this report, we have conducted steam flow experiments by multi-purpose steam experiment apparatus 'WISSH' and CFD calculations by steam flow code 'MATIS-SC' to clarify those effects. As a result, in the middle opening condition, we have observed rotating pressure fluctuations in the experiment and valve-attached flow and local high-pressure region in the CFD result. These results show the pressure fluctuations in steam experiments and CFD is same kind of the fluctuations found in air experiment and CFD. (author)

Two Phase Flow Simulation Using Cellular Automata

International Nuclear Information System (INIS)

Marcel, C.P.

2002-01-01

The classical mathematical treatment of two-phase flows is based on the average of the conservation equations for each phase.In this work, a complementary approach to the modeling of these systems based on statistical population balances of aut omata sets is presented.Automata are entities defined by mathematical states that change following iterative rules representing interactions with the neighborhood.A model of automata for two-phase flow simulation is presented.This model consists of fie lds of virtual spheres that change their volumes and move around a certain environment.The model is more general than the classical cellular automata in two respects: the grid of cellular automata is dismissed in favor of a trajectory generator, and the rules of interaction involve parameters representing the actual physical interactions between phases.Automata simulation was used to study unsolved two-phase flow problems involving high heat flux rates. One system described in this work consists of a vertical channel with saturated water at normal pressure heated from the lower surface.The heater causes water to boil and starts the bubble production.We used cellular automata to describe two-phase flows and the interaction with the heater.General rule s for such cellular automata representing bubbles moving in stagnant liquid were used, with special attention to correct modeling of different mechanisms of heat transfer.The results of the model were compared to previous experiments and correlations finding good agreement.One of the most important findings is the confirmation of Kutateladze's idea about a close relation between the start of critical heat flux and a change in the flow's topology.This was analyzed using a control volume located in the upper surface of the heater.A strong decrease in the interfacial surface just before the CHF start was encountered.The automata describe quite well some characteristic parameters such as the shape of the local void fraction in the