Reactor Critical Heat Flux Phenomena
This work examines the critical heat flux (CHF) phenomena, both by experiment and by modeling, under high-pressure and low mass flux conditions, in a heated rod bundle for prototypic integral reactor core designs. The experimental results provide important validation data for an updated critical heat flux model for use in thermal-hydraulic computer codes to be used for integral light water reactors, which operate under these conditions.
The conceptual light water reactor (LWR) designs considered as a reference for this study are the NuScale Pressurized Water Reactor Power Plant and the Babcock-Wilcox mPower Reactor Power Plant. There is minimal experimental data at low mass flux conditions and high pressure and not in a rod bundle. Our work provides data and associated modeling as well as a better understanding of the CHF process with prototypic flow and pressure ranges in a rod bundle geometry for integral small modular reactors (SMR) designs. This data currently does not exist.
The critical heat flux phenomenon is one of the key physical phenomena that limit the allowable linear power for a nuclear reactor core design under steady-state operating conditions and can be a limiting condition under transients. The advanced light water reactor designs being considered for modular reactor plants normally involve core flow rates (either under natural circulation or forced circulation) at mass fluxes far below current LWR conditions; i.e., estimated to be 100-400 kg/m2-sec with corresponding velocities. At such flow conditions, the current CHF data is extremely sparse. Based on an open literature review, one finds that no data exists for rod bundles and only a sparse database for heated tubes. For these advanced designs to address the needed safety conditions, some fundamental understanding must be obtained for CHF under these conditions.
