Development and validation of two-phase flow & boiling module based on CorTAF framework

Abstract

Two-phase flow and heat transfer characteristics under boiling condition possesses vital significance for pressurized water reactor (PWR) core design. Traditional system code or subchannel analysis code had been widely applied for safety analysis of reactor core, but the more elaborate distribution of three-dimensional parameters is unable to obtained. In this paper, a two-phase flow & boiling module is developed and implemented based on the previously proposed nuclear reactor core thermal–hydraulic characteristics analysis code CorTAF. The two-phase flow and boiling heat transfer analysis method under drift-flux model is established, combining constitutive model such as bubble formation, grid effect, turbulent mixing, coupled boiling heat transfer and the prediction of critical heat flux under diverse boiling states. The benchmarks including CE5 × 5 and PSBT are selected to perform the comprehensive code validation. Crucial physical parameters are compared with the experiment data. The maximum error of wall temperature is under 4 K in CE5 × 5, maximum error of void fraction and CHF in PSBT is under 0.07 and 15 % respectively, indicating that the two-phase flow & boiling module implemented in CorTAF is capable for accurate prediction of two-phase thermal–hydraulic characteristics in reactor core. Additionally, to visually demonstrate the calculation result by CorTAF, a brief simulation of multiple assemblies under partial blockage is also carried out. This work provides valuable references for safety analysis under reactivity insertion accident and further studies on multi-physics coupling of reactor core.