Search

The objective of this paper is to investigate the dynamic behavior of air-water fluid transfer across gap regions between fuel rods in LWR bundles, referred to as cross flow. A computational multifluid dynamic simulation of the cross flow is conducted by using an analysis code based on a three-dimensional two-fluid formulation plus an interface-tracking method. This code is applied to two-phase flows inside a single vertical cylindrical tube in order to perform wall friction sensitivity studies with a model that accounts for velocity gradients across boundary layers flowing on walls. Numerical analyses of two-phase cross flow across the gap region of two circular interconnected channels showed that the calculated turbulent mixing rates are in relatively good agreement with the results of experimental tests in the case of increased wall friction at low void fractions. On the other hand, at high void fraction, the agreement is not as good. Spectrum analyses of the calculated mixing velocities of both air and water across the gap region and the calculated differential pressure between channels showed a strong correlation between them. The present numerical study could demonstrate the following observations shown in previous experimental studies. (1) The fluctuation in differential pressure between channels is attributed to the passing of large bubbles through a channel. (2) The fluctuation in differential pressure primarily induces turbulent mixing phenomena.

Numerical simulation of gas-liquid two-phase flow is important for design of nuclear power systems. Conventional codes require experiments from laboratory scale to mock-up to obtain database for constitutive equations and verification. The aim of this study is to provide new simulation technique to substitute for a part of costly experiments. The extended two-fluid model developed by Hitachi, Ltd. was employed to treat inhomogeneous and intermittent two-phase flow behavior, which had been considered through constitutive equations. Validity of the new simulation code for complex two-phase flow was proved by comparison between calculated and experimental void fraction signals from two-phase flows in circular tubes, CCFL of a thick plate, and natural circulation of aerated pool in double cylinders. For measurement of instantaneous, multi-dimensional void fraction distributions, wire-mesh sensor techniques were employed for all test setups. It was found that the two-phase flow has higher orders of dynamics than initially expected, leading to development of robust analysis methods. The comparison was successful for well-structured flow patterns as slug flow, since the large diameter bubbles show low dimensional dynamics. It is also noted that an optimal delay time reconstruction method and a point correlation dimension method together were useful to increase the reliability of the analysis.

The purpose of the present study is to develop numerical tools for two-phase flow ploblems in pumping stations. Reliable simulation methods of two-phase flow are inevitable in nuclear engineering field because computer programs are important design tools for nuclear reactors which are exposed to severe ambient conditions of high temperature, high pressure and gamma ray irradiation. Mock-up tests of nuclear thermal-hydraulics are very difficult. Hitachi Ltd. has created advanced computer programs for multi-dimensional two-phase flow to analize reactor systems and components. The numerical methods enable accurate and stable interface calculation in spite of large density ratio between the phases, because non-staggered finite volume method, which is suitable for treatment of discontinuity, and minimum numerical diffusion are adopted. The multi-dimensional computer programs employ two types of two-phase flow model. The first one is the extended two-fluid model, which has capability of both interface tracking and ordinary two-fluid model calculations. This model is appropriate for the present purpose because it can treat free surface of channels and entrained bubbly mixture due to breaking waves. The second one is the improved VOF (Volume of Fluid) method in which PLIC (Piecewise Linear Interface Calculation) technique was employed. The interface state calculation is simplified by approximating cubic calculation cells to spherical ones. These computer program were applied to hydraulic studies in Hitachi Industries Co., Ltd., a vender of fluid machinery and pumping systems. It is necessary to evacuate air downstream from siphon outlets as bubble entrainment by water flow in order to operate the plant effectively after pumping station start-up. Experiments were carried out by using a reduced scale test-section. The test section was made of transparent acryl and the cross-section was rectangular. Free surface of water layer on the inclined wall and entrained bubbly mixture into the water pool due to water layer collision with free surface were observed by a video camera. The extended two-fluid model results of transient water level and pressures in the simulated siphon outlets were in good agreement with the experimental data. The predicted pressure histories at the crest and the pump site were in good agreement with the experimental data. The sedimentation in a inlet channel of a pump station was carried out. Three dimensional free surface flow was calculated by the improved VOF method and behavior of suspended sediment in the flow field is calculated. The surface wave velocity in the shallow channel was in good agreement with the theoretical one. Sediment particles had size distribution and drag force and gravity and buoyancy was taken into consideration. Critical tractive force of the sediment on floor was estimated from Schields’ correlation. The predictions of sedimentation area and suspended fraction agreed well with experimental data. These results showed that the present computer programs were able to treat key two-phase flow phenomena in pumping stations.

In subchannel analysis, the conservation equations are solved for each channel in a complex fuel bundle, where the effects of fluid exchange between each subchannel are considered. The fluid exchange is commonly referred to as that caused by cross flow. Void drift is considered to be phenomenon resulting from attaining a hydrodynamic equilibrium state. Its mechanism has not been clarified, and the transport due to void drift is therefore estimated through empirical models in conventional subchannel analyses. Therefore, mechanistic model for the void drift phenomenon is required to apply the subchannel analysis to a variety of fuel bundle geometry. In this study, multi-dimensional analysis using two-fluid model was applied to two-phase flow inside a geometry simulating fuel bundle subchannels, for the purpose of clarifying the void drift mechanism. The comparison between the results of the numerical analysis and the experiment confirmed that the reliability of the numerical method used in this study. In this paper, a mechanistic model based on the Stanton number, which expresses the void diffusion coefficient based on the Lahey's proposal, was proposed.

A new gas-liquid two-phase flow simulation method has been developed based on the extended two-fluid model, which has capabilities of both the two-fluid and the interface-tracking models. The VOF (Volume of Fluid) technique has been introduced for suitable interface calculations. Interfaces of free surface and large bubbles are calculated directly by solving transport of a steep void fraction gradient corresponding to interface, while averaged behavior of microscopic dispersed bubbles and droplets are calculated in the two-fluid model scheme. It is expected that the present method can treat effects of significant kinetic interaction between the phases directly without empirical correlations. The calculated propagation of wet front in a dam break problem is close to experimental data. The predicted flow patterns of complex gas-liquid two-phase flow in a flat tube are quite similar to observations with a video camera. The present simulation will be a useful tool for predictions of integral behavior of thermal-hydraulic phenomena in large-scale nuclear power plants.

Detailed gas-liquid two-phase flow prediction is required for development of high performance components of nuclear reactors and oil power generation plants, while performance components of nuclear reactors and oil power generation plants, while accelerated calculation method was developed by combining gas and liquid velocity accelerated calculation method was developed by combining gas and liquid velocity unified volumetric velocity of two-phase mixture was calculated instead of individual phase velocities and the calculation process became simpler. Velocity calculation error for a phase of small fraction could be reduced in convergence criteria of mass conservation because acceleration slip was obtained exactly from the momentum balance equations. Calculation time for two-phase flow in a chimney of a natural circulation reactor was about a half of the previous calculation method.

In order to evaluate the heat transfer performance in a feed water heater, a three-dimensional analysis for the heater of a power plant was conducted with a computational code for two-phase flow using a two-fluid model. In a conventional feed-water heater which was divided into six sections in the axial direction. the steam flowed mainly into both end sections. Therefore the overall heat-transfer coefficient was large there. Based on the calculated results, an improved feed-water heater was proposed in which the support plate had a hole leading the steam into the neighboring sections. In the improved feed-water heater, the steam flowed from the steam inlet section to the neighboring sections through the hole in the support plate, and the steam velocity was generally larger than that in the former heater. Therefore the overall heat-transfer coefficient was also larger.

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.

High-pressure saturated water discharge from orifices (I.D. 17 mm and 30 mm) in a pipe (I.D. 127 mm) is analyzed by using a transient, nonequilibrium, two-dimensional two-fluid model. The numerical results are compared with experimental data These results depict flow patterns which affect the discharge rate, such as a flow contraction in the orifice and a water jet from the orifice with flashing around the peripheral region of the jet. The predicted discharge rate is within the observed data scatter of 15% when the pressure difference across the orifice is over 2.5 MPa, while the prediction is underestimated by 20% when the pressure difference is about 1.0 MPa.

A computer program SENHOR-IV was developed which describes blowdown phenomena associated with a small pipe-break accident in pressure-tube type reactors. Thermal-hydraulic transients of single-phase and two-phase flow in a primary cooling system, which is composed of the pressure tubes, a steam drum, downcomers, a lower header and pipings connecting these components, were calculated from the conservation equations of mass, momentum and energy by assuming pressure propagation and flow rate distribution to be quasi-steady and by applying the method of characteristics to enthalpy transport. The void propagation velocity in two-phase flow was given from Smith's equation for void-quality relationship to the program. Calculation of a flow transient, which has an exact solution, with use of this program showed small deviations from the exact solution. Predicted transients of pressure and water level in the steam drum indicated a good agreement with those observed in a full scale test facility at O-arai Engineering Center.

An analytical model of two-phase head and torque of centrifugal pumps was developed. Pressure differences in gas and liquid phases through the pump were assumed to be equal to each other and a two-fluid model was used for evaluation of the pump performance in the case that two-phase flow entered the pump.Interphase friction had a large effect on calculated results of the two-phase pump performance. Andersen's model of interphase friction in the bubbly flow regime gave the most proper prediction in both normal and reverse direction flows. The analytical results were within data scatter obtained from tests of three model pumps. The agreement between experimental data and the prediction suggests that the present mathematical model is able to describe two-phase flow behavior in the centrifugal pumps.

This paper presents an experimental and analytical study on pressure pulse propagation phenomena in a plastically deforming pipe filled with water. The simple and well-defined experiments have been performed in order to provide the fundamental data to verify a two-dimensional fluid-structure interaction code. The experimental facility consists of a test pipe and a pulse gun filled with water at room temperature. The pressure pulse was generated in the pulse gun by low velocity SK explosive, high velocity pentolite or a low energy detonator. Experiments have been performed on 3 m long, 16 cm OD and 0.287 cm thick stainless steel-316 pipe both with and without elbow section. Transient pressure, transient strain and plastic deformation are measured when the pressure pulse propagates along the test pipe which is jointed rigidly to the pulse gun. On entering the test pipe, the pulse attenuates rapidly to the yield pressure of the pipe (14 MPa) in about 40 cm. Large plastic deformation occurs over wider range in low velocity explosive experiments than high velocity explosive experiments. Peak pressure reduces in the elbow section more in the high velocity explosive experiments than the low velocity explosive experiments. The experimental results have been analyzed by using a two-dimensional code. The comparison between experiment and prediction for the pressure pulse propagation behavior shows fairly good results. However, the strains are 30–100% over-estimated by the code. This discrepancy is attributable to the influence of the strain rate on stress and the limitation of the calculational model for simulating the detailed structure of the test facility.

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.

A numerical calculation technique for the two-dimensional two-fluid model has been developed. The control volume formulation and non-staggered mesh scheme are employed in order to confirm that solutions satisfy the conservation equations of the two-fluid model. Numerical instability due to the non-staggered mesh is overcome by considering additional flows induced from a local pressure gradient on control volume boundaries. The Godunov method and SIMPLE method are used to estimate the additional flow for compressible and incompressible two-phase flows, respectively. These methods are modified to take the density difference between the phases and the added mass effect into account. The present method has been applied to analyses of two-dimensional two-phase flow discharge from a pipe and phase separation due to gravity in a horizontal circular pipe in order to investigate the capability of dealing with fundamental phenomena of two-dimensional two-phase flow. The present technique enables a small scale compu...