• Corrosion Science and Technology
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• 제23권3호
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• pp.246-254
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• 2024

Recently, coal-fired power plants are experiencing many problems that they have never experienced before due to an increase in flexible operation. In particular, a two-phase flow accelerated corrosion on water wall tubes in a boiler has not been detected overseas or domestically. There is no response plan to deal with such corrosion problem either. However, oxide film damage and tube material corrosion due to a two-phase flow accelerated corrosion are being discovered on water wall boiler tubes of domestic coal-fired power plants recently. If this situation is severe, it can cause enormous damage such as tube rupture. Therefore, in this paper, in order to prepare a response plan for a two-phase flow accelerated corrosion on water wall tubes in the future, differences between a two-phase flow accelerated corrosion and a single-phase flow accelerated corrosion were investigated and an example of discovery of a two-phase flow accelerated corrosion on water wall tubes was presented.

• Nuclear Engineering and Technology
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• 제2권1호
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• pp.19-25
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• 1970

이 연구는 다봉속내에서의 인접유로간의 냉각재 혼합류를 실험적으로 다룬 것이다. 실험은 19봉속내의 사각형 유로와 삼각형 유로간의 혼합류를 단상 유동과 공기-물 이상 유동에 물질 전달량을 측정하여 얻고 있다. 실험결과는 단상 유동에서 낮은 혼합률을, 공기-물 이상 유동에서 큰 혼합률을 얻고 있으며 공기-물 유동에서의 혼합률은 공기 체적률의 증가에 따라 감소되고 있음을 나타내고 있다. 공기-물 이상 유동에서의 높은 혼합률은 공기류에 의한 충분한 교란효과 때문인 것 같다.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.365-376
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• 2010

This paper summarizes the results of a Writing Group on the Extension of CFD codes to two-phase flow safety problems, which was created by the Group for Analysis and Management of Accidents of the Nuclear Energy Agency' Committee on the Safety of Nuclear Installations (NEA-CSNI). Two-phase CFD used for safety investigations may predict small scale flow processes, which are not seen by system thermalhydraulic codes. However, the two-phase CFD models are not as mature as those in the single phase CFD and potential users need some guidance for proper application. In this paper, a classification of various modelling approaches is proposed. Then, a general multi-step methodology for using two-phase-CFD is explained, including a preliminary identification of flow processes, a model selection, and a verification and validation process. A list of 26 nuclear reactor safety issues that could benefit from investigations at the CFD scale is identified. Then, a few issues are analyzed in more detail, and a preliminary state-of-the-art is proposed and the remaining gaps in the existing approaches are identified. Finally, guidelines for users are proposed.

• Journal of the Korean Society for Industrial and Applied Mathematics
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• 제17권3호
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• pp.209-219
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• 2013

The present paper deals with the extension of AUSMPW+ scheme into two-fluid model for multiphase flow. AUSMPW+ scheme is the improvement of a single-phase AUSM+ scheme by designing pressure-based weighting functions to prevent oscillations near a wall and shock instability after a strong shock. Recently, Kitamura and Liou assessed a family of AUSM-type schemes with two-fluid model governing equations [K. Kitamura and M.-S. Liou, Comparative study of AUSM-Family schemes in compressible multi-phase flow simulations, ICCFD7-3702 (2012)]. It was observed that the direct application of the single-phase AUSMPW+ did not provide satisfactory results for most of numerical test cases, which motivates the current study. It turns out that, by designing pressure-based weighting functions, which play a key role in controlling numerical diffusion for two-fluid model, problems reported in can be overcome. Various numerical experiments validate the proposed modification of AUSMPW+ scheme is accurate and robust to solve multiphase flow within the framework of two-fluid model.

• 한국전산유체공학회지
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• 제20권4호
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• pp.14-20
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• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

• 한국자동차공학회논문집
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• 제4권3호
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• pp.176-186
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• 1996

The phenomena of drag reduction using small quantities of a liner macromolecules has attracted the attention of many experimental investigations. On the other hand drag reduction in two phase flow can be applied to the transport of crude oil, phase change system such as chemical reactor, pool and boiling flow, and to flow with cavitation which occurs pump impellers. But the research on dragreduction in two phase flow is not sufficient. The purpose of the present work is to evaluate the drag reduction by measuring pressure drop, void fraction, mean liquid velocity and turbulent intensity whether polymer additives a horizontal single and two phase system or not. Flow pattern of air-water two phase flow was classified by electrical conductivity probe signal. Velocities and turbulent intensities of signal were measured simultaneously with a Hot-film anemometer.

• 대한기계학회논문집
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• 제19권10호
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• pp.2586-2594
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• 1995

Two phase flow phenomena are observed in many industrial facilities and make much importance of optimum design for nuclear power plant and various heat exchangers. This experimental study has been investigated the classification of the flow pattern, the local void distribution and convective heat transfer in swirl and non-swirl two phase flow under the isothermal and nonisothermal conditions. The convective heat transfer coefficients in the single phase water flow were measured and compared with the calculated results from the Sieder-Tate correlation. These coefficients were used for comparisons with the two-phase heat transfer coefficients in the flow orientations. The experimental results indicate, that the void probe signal and probability density function of void distribution can used into classify the flow patterns, no significant difference in voidage distribution was observed between isothermal and non-isothermal condition in non-swirl flow, the values of two phase heat transfer coefficients increase when superficial air velocities increase, and the enhancement of the values is observed to be most pronounced at the highest superficial water velocity in non-swirl flow. Also two phase heat transfer coefficients in swirl flow are increased when the twist ratios are decreased.

• 한국가시화정보학회지
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• 제18권3호
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• Journal of Advanced Marine Engineering and Technology
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• 제10권4호
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• pp.67-77
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• 1986

The information on the heat transfer characteristics, flow pattern and pressure drop, are very important for the desing of general heat exchanger, refrigerating system, air conditioning system and energy recovery system. In these systems, water or lubricating oil contained in working fluid affects greatly the flow and heat transfer condition and this phenomena must be considered in the practical design. An experiment has been performed for studying the flow and heat transfer characteristics of the forced convective horizontal flow of R-113 under the range of the liquid single phase state to the boiling flow state. Basic experimental results are obtained in the case that water or lubricating oil does not contaminate in the test fluid. Experimental results are as follows; (1) The local heat transfer coefficients in the nucleate boiling region and transition boiling region are almostly ten times as large as that of liquid single phase flow. (2) The measured heat transfer coefficient in the present experimental range is relatively agreed well with the predicted value from the various experimental results for the boiling flow.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2491-2498
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• 2020

The paper concentrates on an experimental study of the pressure drop in double-layered packed beds formed by glass spheres, having the configuration of horizontal and vertical stratification. Both single-phase and two-phase flow tests are performed. The pressure drop during the test is recorded and the measured data are compared with those of homogeneous beds consisting of mono-size particles. The results show that for the horizontally stratified bed with fine particles atop coarse particles, the pressure drop in top layer is found higher than those of homogenous bed consisting of the same smaller size particles, while the measured pressure drop of bottom part is similar with those of similar homogenous bed. But for the homologous bed with upside-down structure, the stratification has little or no effect on the pressure drop of the horizontally stratified bed, and the pressure drop of each layer is almost same as that of homogeneous bed packed with corresponding spheres. Additionally, in vertically stratified bed, the pressure drops on the left and right side is almost equal and between those in homogeneous beds. It is speculated that vertically stratified structure may lead to lateral flow which redistributes the flow rate in different parts of packed bed.