Experimental study of two-phase pump performance using a full size nuclear reactor pump

The performance of a full-size nuclear reactor primary heat transport pump was investigated experimentally under high pressure, steam–water two-phase flow conditions. A new set of two-phase pump performance test data was obtained with local void fraction and mass flux measurements at the pump suction. The effects of suction temperature and initial flow conditions on the two-phase pump performance characteristics were described.     

Scaled entrainment measurements in ripple-annular flow in a small tube

High pressure steam–water annular flow was simulated with scaled air–water and Freon-113 experiments. A unique Freon-113 loop was built to obtain low surface tension and high vapor density data on entrainment fraction. The results were compared with two correlations available in the open literature. The Ishii and Mishima correlation was capable to collapse the data for air–water, Freon-113 and steam–water experiments satisfactorily. However the correlation needs to be adjusted for high Weber numbers of the gas phase. The data was limited to the ripple-annular regime because the film extraction technique is not reliable when the film becomes frothy.     

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

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

Flow regime transition criteria for upward two-phase flow in vertical narrow rectangular channels

In relation to the cooling system of high performance microelectronics, a high power research reactor with plate type fuels and plasma facing components of a fusion reactor, study of two-phase flow in a narrow rectangular channel has been paid considerable attention, recently. For the two-fluid model, direct geometrical parameters such as the void fraction should be used in flow-regime criteria. From this point of view, flow-regime transition criteria for vertical upward flows in narrow rectangular channels have been developed considering the mechanisms of flow-regime transitions. The basic concept of the present modeling followed the Mishima and Ishii model for vertical upward two-phase flows in round tubes. Newly developed criteria have been compared with the existing experimental data for air–water flows in narrow rectangular channels with the gaps of 0.3–17 mm. The present criteria showed satisfactory agreements with those data. Further comparisons with data for steam–water in a rectangular channel at relatively high system pressures have been made. The results confirmed that the present flow-regime transition criteria could be applied over wide ranges of parameters as well as to boiling flow.     

ALE finite element method for gas–liquid two-phase flow including moving boundary based on an incompressible two-fluid model

This paper proposes an ALE (Arbitrary Lagrangian–Eulerian) finite element method for gas–liquid two-phase flow, based on an incompressible two-fluid model, to analyze the two-phase flow including moving boundaries. The basic equations are derived by describing the two-fluid model in the ALE form. The solution algorithm is parallel to a fractional step method, and the Galerkin method is employed for the formulation. A quadrilateral element with four nodes is used for the discretization of the computational domain. The present method is also applied to calculate the flow around a circular cylinder, which is forced to oscillate in a quiescent air–water two-phase mixture. The drag coefficients of the cylinder exhibit periodical change in accordance with the variation of the flow around the cylinder. The time variations of the flow field and drag coefficients are discussed in relation to the oscillation of the cylinder.     

Evolution of the structure of a gas–liquid two-phase flow in a large vertical pipe

The evolution of the structure of a gas–liquid flow in a large vertical pipe of 195 mm inner diameter was investigated at the TOPFLOW test facility in Rossendorf. Wire-mesh sensors were used to measure sequences of two-dimensional distributions of local instantaneous gas fraction within the complete pipe cross-section. The sensors own a resolution of 3 mm at a frequency of 2500 Hz. Superficial velocities were varied in a range covering flow regimes from bubbly to churn-turbulent flow. The distance between the gas injection and the sensor position was changed using a so-called variable gas injection system. It consists of six gas injection units, each equipped with three rings of injection orifices in the pipe wall (orifice diameter: 1 and 4 mm), which are fed from ring chambers. The gas flow towards these distributor chambers is individually controlled by valves. Measured bubble-size resolved radial gas fraction profiles reveal differences in the lateral migration of bubbles of different size starting from the injection at the wall. The evolution of bubble-size distributions allows to study bubble coalescence and break-up. The influence of the physical properties of the fluid was studied by comparing cold air–water experiments with steam–water tests at 65 bar.     

A study on applicability of similarity rule to performances of centrifugal pumps driven in two-phase flow

It is studied in this report whether the similarity rule which is used for pumps in single-phase flow is applicable to those in two-phase flow using fundamental equations by which the hydraulic torques of pumps in two-phase flow are described pretty well.According to the equations, the similarity rule is applicable to pumps driven in two-phase flow if the force acting on the gas-liquid interface per unit length of the flow channel in the impeller is proportional to the pump size, and according to the study of the force it can be said that the force is proportional to the pump size unless the gas velocity is very small or liquid velocity is extremely large in the impeller.Therefore it is concluded that the similarity rule is applicable to pumps in two-phase flow excepting special operating conditions which are not usually realized.     

Choking flow modeling with mechanical non-equilibrium for two-phase two-component flow

A mechanistic model which considers the mechanical non-equilibrium is described for two-phase choking flow. The choking mass flux is obtained from the momentum equation with the definition of choking. The key parameter for the mechanical non-equilibrium is a slip ratio. The dependent parameters for the slip ratio are identified. In this research, the slip ratio which is defined in the drift flux model is used to identify the impact parameters on the slip ratio. Because the slip ratio in the drift flux model is related to the distribution parameter and drift velocity, the adequate correlations depending on the flow regime are introduced in this study. In this mechanistic modeling approach, the choking mass flow rate is expressed by the function of pressure, quality and slip ratio. The developed model is evaluated by comparing with the air–water experimental data to eliminate the thermal effect. The comparison of predicted choking model for mechanical non-equilibrium with other experimental data in high quality region (up to 80%) is quite reasonable with a small error.     

UPTF experiment: Flow phenomena during full-scale loop seal clearing of a PWR

To investigate the flow phenomena in the primary system of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) occurring with a small or intermediate break, experiments were performed at the full-scale Upper Plenum Test Facility (UPTF). Within the Transient and Accident Management (TRAM) program integral and separate effect tests were carried out to study loop seal clearing and to provide data for the further improvement of computer codes concerning the reactor safety analysis. This paper describes the UPTF tests that focus on the sequence of loop seal clearance in a four-loop operation for two different cold leg break sizes and the residual water levels, the flow patterns in, and the pressure drops across a single loop seal during the clearing. The UPTF results obtained from a single-loop seal operation are compared with experimental data and correlations available in the literature. Two correlations are proposed which allow the quantification of residual water levels in the loop seal under PWR conditions. It is shown that the steam–water test results gained from the full-scale UPTF with realistic PWR loop seal geometry differ from those obtained from the full or small-scale test facilities under air–water conditions. The UPTF experiments indicate the substantial need for steam-water test data from a full-scale facility with realistic PWR geometries in order to validate PWR LOCA thermal-hydraulic system codes to predict loop seal clearing correctly.     

CATHARE phase separation modeling for small breaks in horizontal pipes with stratified flow

Phase separation is analysed for a two-phase mixture flowing through dividing tee junctions with a small ratio of branch to main pipe diameter. The case of a stratified flow in a horizontal main pipe connected to a horizontal, vertical upward or vertical downward branch is considered. A model is developed on the basis of the available data banks. A wide range of parameters is covered (pressure, 0.2–7 MPa; main pipe diameter, D = 80–284 mm; branch diameter, 4–34 mm). The model has been introduced in the CATHARE Pressurized Water Reactor (PWR) safety code, thereby leading to a considerable improvement of the code predictions.     

Improved Italian moisture separators (IIMS)

A highly efficient, low pressure drop and mechanically sturdy steam separator was developed and tested. The separator was tested in an air—water loop and in a steam—water loop under prototypical conditions.The developmental effort is described and the experimental results are given. These separators have the highest known efficiency, and have a very large operational margin.     

Phenomenological investigation of sodium boiling in a SFR core during a postulated ULOF transient with CATHARE 2 system code: a stabilized boiling case

In the framework of the research and development on GEN IV sodium fast reactors (SFRs), the phenomenology of sodium boiling during a postulated unprotected loss of flow (ULOF) transient has been investigated with the CATHARE 2 system code. This study focuses on a stabilized boiling case: in such a regime, no flow redistribution occurs from the subassemblies which have reached the saturation temperature to those that are still single-phase. In this paper, for a subassembly design featuring no restrictive structures above the fuel bundle, a quasi-static approach is first developed to get an upper bound of the reactor core power at boiling onset that would be compatible with the well-known Ledinegg criteria for diphasic flow static equilibrium. Then, dynamics results achieved through simulation with the CATHARE 2 code for a postulated ULOF are presented: boiling is shown to remain stable during the transient for such a core power at boiling onset. Another important outcome of the simulation is the calculation of a dynamic instability, in the form of a two-phase hydrodynamic chugging phenomenon. The predicted phenomenology of this stabilized boiling case should be studied further in order to consider its dependency on the underlying closure laws and to eliminate the possibility of a numerical instability.     

Correlations for predicting single phase and two-phase flow pressure drop in pebble bed flow channels

An experimental study was conducted on the pressure drop of the single phase and the air–water two-phase flow in the bed of rectangular cross sections densely filled with uniform spheres. Three kinds of glass spheres with different equivalent diameters (3 mm, 6 mm, and 8 mm) were used for the establishment of the test sections. The Reynolds number in the experiment ranged from a dozen to thousands for the single-phase flow and from hundreds to tens of thousands for the two-phase flow. In the present flow-regime model, the bed was subdivided into a near-wall region and a central region in order to take the wall effect into account to improve the prediction at low tube-to-particle diameter ratios. Improved correlations are obtained based on the previous study to consider the single-phase flow pressure drops for finite pebble beds with spherical particles and nonspherical particles by fitting the coefficients of that equation to both the database and the present experiment. The correlation is consistent with the observed physical behavior which explains its comparatively good agreement with the experimental data. A new empirical correlation for the prediction of two-phase flow pressure drops was proposed based on the gas phase relative permeability as a function of the gas phase saturation and the void fraction. The correlation fit well for both experimental data of spherical particles and nonspherical particles.     

Development of nonequilibrium pressurizer model with noncondensable gas

To investigate thermal–hydraulic characteristics of a steam–gas pressurizer in the integral type reactor, the steam–gas pressurizer model based on the two-region nonequilibrium concept was developed and introduced into RETRAN-3D/INT code. The model includes an explicit solution method for the one-dimensional governing equations and the equation of the state solution method to determine the thermal–hydraulic state of the steam–gas pressurizer volume. In addition, the wall condensation model based on the diffusion layer modeling was included to consider the effect of the noncondensable gas. The developed model was verified with the results from the pressurizer insurge experiment conducted at Massachusetts Institute of Technology. From the verification results, it was concluded that the developed steam–gas pressurizer model can sufficiently predict the pressurizer transient and it can be used as a component model of the one-dimensional system code based on the homogeneous equilibrium model.     

Natural circulation flow behavior at reduced inventory in a VVER geometry

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

Two-phase flow resistance in flexible metal hoses

This study presents the two-phase flow resistance, hence the friction factor and the pressure drop for air–water mixture flowing in flexible metal hoses. Experiments were performed under the following conditions of two-phase parameters; mass flux from 200 to 1150 kg/m² s, gas quality from 1 to 60% and system pressure from 3 to 10 bar. The inner diameters of the tested hoses were 25, 40, 50 and 65 mm with a ratio of ridge depth to inner diameter (r/d) from 0.02 to 0.1 and a ratio of pitch to inner diameter (p/d) from 0.06 to 0.3. The results demonstrate that the two-phase flow resistance, energy dissipation and friction losses in flexible metal hoses are perceptible greater than that in pipes. Therefore, the two-phase pressure drops of the hoses are two to five times greater than that in smooth pipes. The two-phase friction factor of such hoses increased from 0.035 up to 0.2 in dependence on the influencing flow and geometrical parameters. Based on the energy balance and the presented experimental results, a new model has been developed to calculate the two-phase pressure drops and hence the friction factor of flexible metal hoses. The model includes the relevant primary parameters, fit the data well and is sufficiently accurate for engineering purposes. The results reported enable practical designs with standard products and optimization of the hose geometry for specific conditions.     

Two-phase flow and boiling heat transfer in two vertical narrow annuli

Experimental study associated with two-phase flow and heat transfer during flow boiling in two vertical narrow annuli has been conducted. The parameters examined were: mass flux from 38.8 to 163.1 kg/m² s; heat flux from 4.9 to 50.7 kW/m² for inside tube and from 4.2 to 78.8 kW/m² for outside tube; equilibrium mass quality from 0.02 to 0.88; system pressure from 1.5 to 6.0 MPa. It was found that the boiling heat transfer was strongly influenced by heat flux, while the effect of mass velocity and mass quality were not very significant. This suggested that the boiling heat transfer was mainly via nucleate boiling. The data were used to develop a new correlation for boiling heat transfer in the narrow annuli. In the two-phase flow study, the comparison with the correlation of Chisholm [Chisholm, D., 1967. A theoretical basis for the Lockhart–Martinelli correlation for two-phase flow. Int. J. Heat Mass Transfer 10, 1767–1778] and Mishima and Hibiki [Mishima, K., Hibiki, T., 1996. Some characteristics of air–water two-phase flow in small diameter vertical tubes. Int. J. Multiphase Flow 22, 703–712] indicated that the existing correlations could not predict the two-phase multiplier in the narrow annuli well. Based on the experimental data, a new correlation was developed.     

Investigation of PWR accident situations at BETHSY facility

For several years an extensive programme of separate effect tests related to PWR safety has been conducted in France and in particular at the Nuclear Center of Grenoble. Recently the BETHSY integral test facility - three identical loops, full height and pressure - has been constructed with the main objectives of contributing to the verification of the CATHARE calculation code of accidents and to the validation of the physical bases of Emergency Operating Procedures.So far several tests have been (November 1988) carried out, among which natural circulation under various conditions (single phase, two-phase, symmetric or asymmetric conditions, different core power, variable steam generator liquid level), 2 inches cold leg break, steam bubble formation and collapse in the upper head. Summarized results of some of these tests are introduced and compared to a first CATHARE calculation as far as the 2″ cold leg break is concerned.