Abstract
Critical Heat Flux (CHF) is an important safety operation parameter in nuclear reactors. According to the different flow pattern and heat transfer characteristics, the vertical upward flow boiling crisis can be divided into the Departure form Nucleate Boiling (DNB) type and Dry-Out (DO) type. CHF experiments were performed with Refrigerant (R-134a) flow in an 8-mm vertical heated tube at high-pressure subcooled and low-quality conditions, with a wide range of local conditions including pressure, mass flux, enthalpy, and heating flux. For the DNB occurred in subcooled and low quality flow regime, the Weisman & Pei model was evaluated by using the experimental results from water and R-134a, and the improvement of the constitutive correlations was proposed. The prediction accuracy of the improved model was evaluated with the Weber number range of 2376.8 to 74631, quality range of −0.451 to 0.199. The average error between predicted and experimental CHF values is −4.68% and standard deviation error of 11.74% for 419 data points.
