• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• 대한기계학회논문집B
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• 제24권6호
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• pp.886-893
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• 2000

The Characteristics of dividing the dispersed bubble, plug, and slug steam-water flow in the horizontal junctions with horizontal branches have been experimentally investigated. The experimental investigation of the separation phenomena in a $45^{\circ}$ horizontal wye with equal pipe inner diameter of 25 mm is presented to provide a data base for the development and verification of the analytical models. The phase separation and pressure distribution in the three legs of each test section are obtained through the set of measurements made in the present work. And the dependence of phase separation on different parameters, such as inlet quality and mass flux, is discussed.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.618-630
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• 1999

An analytical model that includes the steam condensation effect has been derived and a parametric study has been performed. In addition, a series of experiments were performed and a total of 34 experimental data for the onset of CCFL in nearly horizontal countercurrent two-phase How have been obtained for various flow rates of water. Comparisons of the present CCFL data with slug formation models show that the agreement between the present as well as the existing model and the data is about the same. However, the deviation between the Taitel and Dukler's model predictions and the data is the largest when if j$_{f}$<0.04 m/s. A parametric study of the effect of the steam condensation using the present model shows that, when all local conditions are similar, the model predicted local gas velocities that cause the onset of flooding are slightly lower when condensation occurred. Based on the visual observation and the evaluation of the present work, it has been concluded that the criterion derived for the onset of slug flow can be directly used to predict the onset of inner flooding in nearly horizontal two-phase flow within the experimental ranges of the present work.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2004년도 춘계 학술대회논문집
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• pp.49-55
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• 2004

This paper addresses a numerical simulation of the flow and heat transfer in a simplified model of helically coiled tube steam generator using a general purpose computational fluid dynamic analysis computer code. The steam generator model is comprised of a cylindrical shell and helically coiled tubes. A cold feed water entered the tubes is heated up, evaporates. and finally become a superheated steam with a large amount of heat transferred continuously from the hot compressed water at higher pressure flowing counter-currently through the shell side. For the calculation of tube side two-phase flow field formed by boiling, inhomogeneous two-fluid model is used. Both the internal and external turbulent flows are simulated using the standard k-e model. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. The numerical calculations are peformed for helically coiled tubes of steam generator at an integral type pressurized water reactor under normal operation. The effects of tube-side inlet flow velocity are discussed in details. The results of present numerical simulation are considered to be physically plausible based on the data and knowledge from previous experimental and numerical studies where available.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.267-283
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• 2015

The main objective of this study is to investigate experimentally the two-phase flow characteristics in terms of the direct contact condensation of a steam-water stratified flow in a horizontal rectangular channel. Experiments were performed for both air-water and steam-water flows with a cocurrent flow configuration. This work presents the local temperature and velocity distributions in a water layer as well as the interfacial characteristics of both condensing and noncondensing fluid flows. The gas superficial velocity varied from 1.2 m/s to 2.0 m/s for air and from 1.2 m/s to 2.8 m/s for steam under a fixed inlet water superficial velocity of 0.025 m/s. Some advanced measurement methods have been applied to measure the local characteristics of the water layer thickness, temperature, and velocity fields in a horizontal stratified flow. The instantaneous velocity and temperature fields inside the water layer were measured using laser-induced fluorescence and particle image velocimetry, respectively. In addition, the water layer thickness was measured through an ultrasonic method.

• 한국전산유체공학회지
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• 제15권2호
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• pp.86-94
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• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation or condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure distribution that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new reconstruction method to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function, a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the CUPID code.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2010년 춘계학술대회논문집
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• pp.380-388
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• 2010

A thermal-hydraulic code, named CUPID, has been developed for the analysis of transient two-phase flows in nuclear reactor components. A two-fluid three-field model was used for steam-water two-phase flows. To obtain numerical solutions, the finite volume method was applied over unstructured cell-centered meshes. In steam-water two-phase flows, a phase change, i.e., evaporation of condensation, results in a great change in the flow field because of substantial density difference between liquid and vapor phases. Thus, two-phase flows are very sensitive to the local pressure that determines the phase change. This in turn puts emphasis on the accurate evaluation of local pressure gradient. This paper presents a new numerical scheme to evaluate the pressure gradient at cell centers on unstructured meshes. The results of the new scheme for a simple test function a gravity-driven cavity, and a wall boiling two-phase flow are compared with those of the previous schemes in the cupid code.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.889-896
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• 2020

The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents.

• Tribology and Lubricants
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• 제24권6호
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• pp.362-367
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• 2008

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.

• Journal of Mechanical Science and Technology
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• 제20권5호
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• pp.717-727
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• 2006

The present study experimentally investigated the effect of flow direction and other flow parameters on two-phase flow distribution of refrigerants at a T-junction, and also suggested a prediction model for refrigerant in a T-junction by modifying previous model for air-water flow. R-22, R-134a, and R-410A were used as test refrigerants. As geometric parameters, the direction of the inlet or branch tube and the tube diameter ratio of branch to inlet tube were chosen. The measured data were compared with the values predicted by the models developed for air-water or steam-water mixture in the literature. We propose a modified model for application to the reduced T-junction and vertical tube orientation. Among the geometric parameters, the branch tube direction showed the biggest sensitivity to the mass flow rate ratio for the gas phase, while the inlet quality showed the biggest sensitivity to the mass flow rate ratio among the inlet flow parameters.