McMaster Transient Critical Heat Flux Facility’s Nodalization and CHF Simulation Using ASYST

Abstract

Critical Heat Flux (CHF) is an important parameter in water-cooled nuclear reactor operations and has been studied both experimentally and empirically for decades. McMaster Transient Critical Heat Flux Facility can perform both steady state and transient CHF experiments in water. Several new CHF data were collected from the facility to be used for code benchmarking in vertically oriented geometry at a scale relevant to boiling water reactors (BWR) and some water-cooled small modular reactors (SMR).

ASYST is a system thermal hydraulic code designed for nuclear safety analysis. ASYST-THA, the thermal hydraulic module, has multidimensional and multi-fluid modeling capability. In the present paper, a nodalization of the experimental facility was developed using ASYST. Steady state CHF simulations were performed. Simulations covered pressure range from 2.0 to 6.0 MPa and mass flux from 1000 to 2500 kg/(m²s). Twenty-three data points were generated. Results show a good agreement with experimental data. A comparison between ASYST prediction and look-up table (LUT) and Biasi’s correlation shows ASYST’s correlations is presented. At the end of the paper, the applicability of ASYST and the CHF data to boiling water SMRs under transient conditions was discussed from the perspective of heated length.