SEPARATE-EFFECTS TESTS FOR TWO-PHASE FLOW WITH GEOMETRIC RESTRICTIONS
Abstract: While significant efforts have been made to investigate two-phase flows in vertical upward channels, limited work has been performed to study the effects stemming from geometric configurations on two-phase flow transport. This presents significant shortcomings in the application of predictive models to practical systems, including nuclear power reactors. In the presentation, results obtained from various separate-effects test facilities to address the geometric effects on two-phase flow transport will be presented. These include: effects of varying orientations and flow restrictions on two-phase flow transport; effects of inlet geometry in co-current vertical downward two-phase flow; effects of pipe sizes in horizontal two-phase flow; and interfacial area transport modeling. An extensive local two-phase flow database established by the four-sensor conductivity probe in various two-phase flow conditions is presented along with flow visualization studies. Predictive models developed to account for various geometric effects in two-phase flow transport are presented, along with predictions made by the interfacial area transport equation to demonstrate feasibility of interfacial area transport equation in system analysis code.
Bio: Dr. Seungjin Kim is a Professor and McCarthy Head of the School of Nuclear Engineering at Purdue University. He is also the co-director of the Thermal-hydraulics and Reactor Safety Laboratory. Prior to assuming the Head position at Purdue University, he was a professor in the Department of Mechanical and Nuclear Engineering at Penn State University from 2007 through 2017 and in Nuclear Engineering Department at University of Missouri-Rolla from 2003 through 2007. He received his PhD degree in Nuclear Engineering from Purdue University in 1999. Dr. Kim’s research expertise is in the area of reactor thermal-hydraulics and two-phase flow. More specifically, he has done research in reactor thermal-hydraulics, reactor safety analysis, two-phase flow experiments and modeling, interfacial area transport modeling, and 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 code FLUENT for two-phase flow simulation as well as into the system analysis code. His work on advanced two-phase flow instrumentation has made impact on two-phase flow experimental capabilities and is now being employed widely by many research groups worldwide. He has actively taken leadership positions in his field of technical expertise and has served as the Chair of the American Nuclear Society Thermal Hydraulics Division. He is an ANS Fellow and a member of Nuclear Energy Advisory Committee of the Office of Nuclear Energy of U.S. DOE.
