Prediction of critical heat flux in a rod bundle channel with spacer grids based on the Eulerian two-fluid model

The critical heat flux (CHF) is a vital parameter influencing the safety and efficiency of reactor cores. In this study, the Eulerian two-fluid model coupled with the extended wall boiling model in STAR-CCM+ was employed to simulate the departure from nucleate boiling (DNB) phenomenon in a 5 × 5 pressurized water reactor (PWR) fuel rod bundle channel with spacer grids under non-uniform heating conditions. The transition in boiling curves was used as the criterion of DNB occurrence, while the temperature distribution of rod surfaces was utilized for CHF location predictions. The predicted CHF value and CHF location exhibited good agreement with the experimental data. The deviation between calculated and experimental CHF values was within 15% and the deviation between predicted and experimental CHF locations was within one grid-to-grid span length. The results of this study suggested good prospects for the application of two-phase CFD model in predicting CHF in fuel assemblies with spacer grids.