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배병언,정재호,유제용 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.3
Hydraulic performance and flow resistance tests were performed to confirm the main parameters of the hydraulic instrumentation that can affect the pump performance of the reactor coolant pump. The flow resistance test offers important experimental data, which are necessary to predict the behavior of the primary coolant when the circulation of the reactor coolant pump is stopped. Moreover, the shape of the hydraulic section of the pump, which was considered in the test, was prepared to compare the mixedflow- and axial-flow-type models, the difference in the number of blades of the impeller and diffuser, the difference in the shape of the impeller blade and its thickness, and the effect of coating at the suction bell. Additionally, five models of the hydraulic part were manufactured for the experiments. In this study, the differences in performance owing to the design factors were confirmed through the experimental results
미포화 비등 유동에 대한 Bubble lift-off model 개발 및 1군 계면면적 수송방정식 해석
배병언(B.U. Bae),윤병조(B.J. Yun),윤한영(H.Y. Yoon),송철화(C.H. Song),박군철(G.C. Park) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5
The interfacial area transport equation for the subcooled boiling flow was developed with a mechanistic model for the wall boiling source term. It included the bubble lift-off diameter model and lift-off frequency reduction factor model. Those models took into account the bubble's sliding on the heated wall after a departure from the nucleate site and the coalescences of sliding bubbles. To implement the model, the two-phase flow CFD code was developed, which was named as EAGLE (Elaborated Analysis of Gas-Liquid Evolution). The developed model and EAGLE code was validated the experimental data of SUBO (Subcooled Boiling) facility. The computational analysis revealed that the interfacial area transport equation with the bubble lift-off diameter model agreed well with the experimental results. It presents that the source term for the wall nucleation enhanced the prediction capability for a multi-dimensional behavior of void fraction or interfacial area concentration.
미포화 비등 이상유동 내 국소 기포 인자에 대한 실험적 연구
배병언(B.U. Bae),윤병조(B.J. Yun),박원만(W.M. Park),송철화(C.H. Song),박군철(G.C. Park) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5
Subcooled boiling test was carried out in the SUBO (Subcooled boiling) test facility to extend a database for a code benchmark. The test section is a vertical annulus with a heater rod at the channel center. For the measurement of local bubble parameters, double optical fiber sensors were applied at the six elevations. Total six test matrixes are chosen for the parametric study of the heat flux, mass flux and inlet subcooling. Void fraction, interfacial area concentration and bubble velocity were measured at 12 radial locations at each elevation. The local bubble parameters show well the characteristics and propagation of a void fraction and an interfacial area concentration along the test section in the subcooled boiling flow. The present data will be suitable for the benchmark, verification and model development for the CFD style codes or existing safety analysis codes.
배병언(B.U. Bae),윤한영(H.Y. Yoon),어동진(D.J. Euh),송철화(C.H. Song),박군철(G.C. Park) 대한기계학회 2007 대한기계학회 춘추학술대회 Vol.2007 No.5
For the analysis of a two-phase flow, the interaction between two phases such as the interfacial momentum or heat transfer is proportional to the interfacial area. So the interfacial area concentration (IAC) is one of the most important parameters governing the behavior of each phase. This study focuses on the development of a computational fluid dynamics (CFD) code for investigating a boiling flow with a one-group IAC transport equation. It was based on the two-fluid model and governing equations were calculated by SMAC algorithm. For checking the robustness of the developed code, the experiment of a subcooled boiling in a vertical annulus channel was analyzed to validate the capability of the IAC transport equation. As the results, the developed code was confirmed to have the capability in predicting multi-dimensional phenomena of vapor generation and propagation in a subcooled boiling.
조윤제,김석,배병언,박유선,강경호,윤병조 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.6
As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data. As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted todevelop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The PassiveAuxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power ReactorPlus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFSremoves decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removalcapability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependenton the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heattransfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate theexperimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default modelfor condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. Toimprove the calculation result, The Thome model and the new version of the Shah model are implemented and compared withthe experimental data.