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MechE Seminar: Dr. Seungjin Kim "Characterization of Geometric Effects in Two-Phase Flow Transport" Department of Mechanical Engineering 2015 Seminar Series Speaker: Seungjin Kim, Ph.D. Characterization of Geometric Effects in Two-Phase Flow Transport November 6, 2015 | 2:00 PM | DeWalt Seminar Room, 2164 Martin Abstract: While the coolant in practical heat transfer systems flows through various junctions and restrictions, most of the existing closure relations for two-phase flow analysis have been developed based on experimental database established in flow channels without such flow restrictions and changes in flow orientation. Ultimately, predictions made by such models may yield erroneous results and limit system capability. In view of lack of knowledge on the geometry-induced effects in twophase flow transport, the present study performs experiments and develops predictive methods to characterize effects of a 90° elbow in vertical-upward-to-horizontal air-water bubbly two-phase flows. Both the global and local effects are investigated, which include the flow regime transition, pressure loss, convection in bubble velocity and distribution of local two-phase flow parameters, such as void fraction and interfacial area concentration. Additionally, a predictive method for the dissipation length of the elbow-induced effects is presented. Bio: Dr. Kim is an Associate Professor of Nuclear Engineering and an Associate Professor of Mechanical Engineering at The Pennsylvania State University. He is also the director of the Advanced Multiphase Flow Laboratory (AMFL). He received his Ph.D. degree from Nuclear Engineering of Purdue University. His research expertise is in the area of thermal-hydraulics and two-phase flow. More specifically, he has performed experiments for two-phase flows in various flow geometries and orientations, developed predictive models, and developed advanced two-phase flow instrumentation. Dr. Kim’s research has led to the development of the interfacial area transport equation that has been implemented in a commercial computational fluid dynamics (CFD) code FLUENT for two-phase flow simulation as well as into the TRACE code employed by the U.S. Nuclear Regulatory Commission for thermal-hydraulic analysis of nuclear reactor systems. For more information: Jungho Kim (kimjh@umd.edu)
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