Abstract
Cooling systems are wildly used in nuclear power plant equipment and safety systems, such as the external reactor vessel cooling (ERVC) system. It is pretty important to understand the heat transfer characteristics and critical heat flux (CHF) to ensure their efficient and safe operation. Challenges till remain in accurately predicting the subcooled flow boiling curve especially in the low pressure and low mass flux conditions due to its complex boiling process. The present study introduces a numerical model to evaluate subcooled flow boiling heat flux. The proposed model can estimate the transition from forced convection, isolated bubble nucleated boiling to fully developed boiling region by taking the growth and interaction of bubbles into consideration. And the flow boiling limit is numerical described with the probability analysis of dry spots. The results from the new model are validated with experimental data of vertical subcooled flow boiling with single-side heating under low pressure and low mass flux conditions. The predicted boiling curve are in well agreement with experimental results corresponding to different thermal-hydraulic parameters, such as pressure (P = 100–360 kPa), mass flux (G = 50–300 kg/m2s), inlet subcooling (ΔTin = 0–15 K) and wall wettability (hydrophilic and hydrophobic), and the prediction error of CHF is within ± 15%.
