Symposium on Advanced Sensors and Modeling Techniques for Nuclear Reactor Safety

Advanced Measurement and Modeling of Gas-Dispersed Flows with Phase Change

The two-fluid model has long been the backbone of thermal hydraulics calculations for the nuclear power industry and increasingly relied upon for determination of safety margin, course of accident progression, and design of new reactor concepts and safety systems. The success of the two-fluid model relies on accurate constitutive relations at the interface such as the area concentration, heat transfer and mass transfer. The implementation of the interfacial area transport equation into thermal-hydraulic system analysis and Computational Fluid Dynamics (CFD) codes have been recommended to improve prediction capability and solve current shortcomings including inability to simulate the dynamic changes in interfacial structure across flow regimes and significant compound errors stemming from the two-step flow regime based method. The Multiphase Thermo-fluid Dynamics Laboratory at UIUC has several active projects to acquire validation data for one-dimensional system codes and multi-dimensional CFD codes. The laboratory specializes in measurement of two-phase parameters in gas dispersed flows with phase change, under forced convection and natural circulation, up to Critical Heat Flux (CHF). Through collection of new data and improved constitutive modeling, the two-fluid model with interfacial area transport equation is extended and benchmarked in boiling, condensing, and flashing flows important to nuclear reactor safety.