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Lee G.H.,Korea Institute of Nuclear Safety | Bang Y.S.,Korea Institute of Nuclear Safety | Woo S.W.,Korea Institute of Nuclear Safety | Cheong A.J.,Korea Institute of Nuclear Safety | And 2 more authors.
Annals of Nuclear Energy | Year: 2013

A series of 1/5 scale reactor flow distribution tests had been conducted in order to determine the hydraulic characteristics of the APR+ (Advanced Power Reactor Plus) which were used as the input data for the open core thermal margin analysis code. In this study, in order to examine the applicability of computational fluid dynamics with the porous model in the analysis of reactor internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that depending on the shape of flow skirt the flow distribution was locally somewhat different. Standard deviation of mass flow rate (σ) for the original shape of flow skirt was smaller than that for the modified shape of flow skirt. This means that the original shape of flow skirt may give the more uniform distribution of mass flow rate at core inlet plane, which may be more desirable for the core cooling. Porous model for some reactor internal structures could adequately predict the hydraulic characteristics inside reactor in a qualitative manner. However, while the predicted high core inlet flow rate region was located in the core center zone, the measured one was located in the core outer boundary. This difference may result from the fact that some internal structures including the lower support structure assembly were not modeled with the real geometry but treated as the porous domain. If the sufficient computation resource is available, the predicted core inlet mass flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially located in the upstream of core inlet. © 2013 Elsevier Ltd. All rights reserved.


Lee G.H.,Korea Institute of Nuclear Safety | Bang Y.S.,Korea Institute of Nuclear Safety | Woo S.W.,Korea Institute of Nuclear Safety | Kim D.H.,ANFLUX Inc. | Kang M.K.,ANFLUX Inc.
International Congress on Advances in Nuclear Power Plants, ICAPP 2013: Nuclear Power - A Safe and Sustainable Choice for Green Future, Held with the 28th KAIF/KNS Annual Conference | Year: 2013

A series of 1/5 scale reactor flow distribution tests had been conducted in order to determine the hydraulic characteristics of the APR+ (Advanced Power Reactor Plus) which were used as the input data for the open core thermal margin analysis code. In this study, in order to examine the applicability of computational fluid dynamics with the porous model in the analysis of reactor internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that porous domain approach for some reactor internal structures can adequately predict flow characteristics inside reactor in a qualitative manner. However, while the predicted high core inlet flow rate region was located in the core center zone, the measured one was located in the core outer boundary. This difference may result from the fact that some internal structures including the lower support structure assembly were not modeled with the real geometry but treated as the porous domain. If the sufficient computation resource is available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially located in the upstream of core inlet. Finally, depending on the shape of flow skirt the flow distribution was locally somewhat different.


Lee G.H.,Korea Institute of Nuclear Safety | Bang Y.S.,Korea Institute of Nuclear Safety | Woo S.W.,Korea Institute of Nuclear Safety | Kim D.H.,ANFLUX Inc. | Kang M.K.,ANFLUX Inc.
Transactions of the Korean Society of Mechanical Engineers, B | Year: 2013

Even if some CFD software developers and its users think that a state-of-the-art CFD software can be used to reasonably solve at least single-phase nuclear reactor safety problems, there remain limitations and uncertainties in the calculation result. From a regulatory perspective, the Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of commercial CFD software for nuclear reactor safety problems. In this study, to examine the prediction performance of commercial CFD software with the porous model in the analysis of the scale-down APR (Advanced Power Reactor Plus) internal flow, a simulation was conducted with the on-board numerical models in ANSYS CFX R.14 and FLUENT R.14. It was concluded that depending on the CFD software, the internal flow distribution of the scale-down APR was locally somewhat different. Although there was a limitation in estimating the prediction performance of the commercial CFD software owing to the limited amount of measured data, CFX R.14 showed more reasonable prediction results in comparison with FLUENT R.14. Meanwhile, owing to the difference in discretization methodology, FLUENT R.14 required more computational memory than CFX R.14 for the same grid system. Therefore, the CFD software suitable to the available computational resource should be selected for massively parallel computations. © 2013 The Korean Society of Mechanical Engineers.


Lee G.H.,Korea Institute of Nuclear Safety | Bang Y.S.,Korea Institute of Nuclear Safety | Kim D.H.,ANFLUX Inc. | Kang M.G.,ANFLUX Inc.
Transactions of the Korean Society of Mechanical Engineers, B | Year: 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet. © 2013 The Korean Society of Mechanical Engineers.


Jo J.C.,Korea Institute of Nuclear Safety | Do K.S.,Korea Institute of Nuclear Safety | Lee Y.K.,ANFLUX Inc.
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2013

A PWR design incorporates a passive auxiliary feedwater system equipped with one passive condensation heat exchanger (PCHX) which consists of inclined V-shaped tube bundles submerged in a water pool of which the top is open to the atmosphere. During the PCHX operation, saturated steam flows into the PCHX where steam is condensed inside of the tubes by cooling the outer side with the pool water. Then, the condensate flows out passively by gravity. Because the thermal-hydraulic characteristics in the PCHX determine the condensation mass rate and the possibility of thermal stratification-induced fatigue of the pool tank wall, system instability and waterhammer, it is important to understand the phase change flow in the PCHX. In this paper, the complex phase change heat transfer and multi-phase flow in a PCHX tube model were numerically simulated. The single fluid multi-component flow model with the equilibrium phase change model was employed for the condensation phase change flow inside the tube and the two-fluid model with the wall boiling model and the equilibrium phase change model was used for the boiling-induced natural convection outside the tube in the pool. Based on the present numerical simulation, the characteristics of the heat transfer and flow in the PCHX are discussed and illustrated for some typical results. Copyright © 2013 by ASME.

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