Accelerating the development and deployment of advanced energy systems is crucial in an increasingly “energy-hunger” world, especially with the advent of AI era. In today’s U.S. energy market, nuclear power accounts for 20% of the total electricity generation capacity and half of its clean energy production. Therefore, further enhancing the safety and economic efficiency of both existing and future nuclear systems is pivotal for achieving our national energy demand. The swift advancement of advanced nuclear technologies has heightened the need for precise, spatiotemporally resolved, multi-scale multi-physics coupled fundamental modeling of fluid flow and heat transfer, along with numerically robust, reliable, and efficient analysis to support the lifecycle of nuclear power plants – both for the legacy and new constructions. The more in-depth understanding of the phenomenological behavior of fluid flow and heat transfer allows for a more cost-effective implementation of new innovations. 

In the present work, leveraging the high-resolution two-phase flow experimental data uniquely obtained from the RBHT reflood test facility in collaboration with U.S. Nuclear Regulatory Commission (NRC), detailed physics-based theoretical modeling, comprehensive numerical analyses and code validations are performed to understand the fuel-to-coolant thermal-hydraulic responses during quenching transients. Valuable insights obtained from these analyses are further enhanced through high-resolution visualization study and physics-based modeling of two-phase flow interface behavior under natural-circulation film boiling conditions, utilizing a small-scale quench test facility as well as extensive system-scale numerical simulations. The use of advanced image processing techniques allows an in-depth investigation to probe the boundary-layer scale as well as to facilitate the development of sophisticated theoretical flow/heat transfer models for code applications. The present work presents a thorough investigation of the transient fluid flow and heat transfer behaviors, as well as reactor thermal-hydraulic responses, thus extending the boundaries of our current knowledge. The experimental and modeling techniques that we are developing and demonstrating provide an exciting avenue for high-spaciotemporal resolution flow and heat transfer characterization. The results obtained are particularly valuable for the future development of thermal-fluid models and for numerical code validation, ultimately supporting the sustainability of nuclear energy.