Experimental Investigation on Multifluid Multiphase Flow Phenomena for Computer Code Validation

Thermal hydraulics in nuclear engineering including multifluid, multiphase flow phenomena is indispensable because it is the most crucial mechanism of energy transfer in the current conventional nuclear plants, and it also governs some of the important operational conditions and safety margins of nuclear plants. In this study, two independent subjects of multifluid and multiphase flow closely related to the reactor building transient response of the next generation nuclear plant and conventional boiling water reactors are experimentally investigated. The first part covers the transient response of helium-air binary gas mixture flow to characterize the next generation nuclear plant (NGNP) high temperature gas cooled reactor (HTGR) reactor building (RB) response to hypothetical depressurized loss of forced convection (D-LOFC) accident scenario. 1/28 down-scaled experimental facility is designed and fabricated based on the scaling analysis of simplified NGNP HTGR RB model. The transient responses of pressure, temperature and oxygen concentration for three hypothetical D-LOFC accident scenarios on the facility are measured and analyzed. Characteristics of air ingress into reactor building during the corresponding accident scenarios are analyzed quantitatively and qualitatively. The experimental results suggest an increment in the flow area of the flow path between the reactor cavity and steam generator cavity in the NGNP HTGR conceptual design. The boundary and initial conditions of the experimental work are adopted to establish computer simulation cases for generation of thermal hydraulic information in containment (GOTHIC) code. The reported validation of the GOTHIC simulation results against the experimental results presents consistency in general. The second part deals with the steam-air mixture flow in subcooled water pool to investigate direct contact condensation (DCC) of steam jet with various steam mass flow rate. The DCC process is important in modern boiling water reactors (BWRs) since the RB containment of BWRs employs a pressure suppression chamber as a major safety feature which utilize the DCC process. Steady state gas injection rate, gas temperature, and pool temperature are measured, and their effects on the DCC process are analyzed. A new correlation of heat transfer coefficient is suggested based on the experimental results. The effect of non-condensable gas on heat transfer characteristics is discussed. A novel method of temperature field measurement using backlight aided planar laser induced fluorescence (PLIF) for two-phase flow is developed. Instantaneous velocity and temperature fields during the DCC process near the condensing region are measured by simultaneous particle image velocimetry (PIV) and developed PLIF. Improvement of the DCC closure model in computational fluid dynamics (CFD) code is discussed. The application of the current study is not limited to the nuclear thermal-hydraulics computer code validation, but can be extended to general research of multifluid, multiphase flow phenomena