Investigating Post-CHF Heat Transfer Behavior During Dryout and Rewet Cycles

Abstract

Due to the better thermo-mechanical and oxidation properties of accident tolerant fuel (ATF) materials, a shift from critical heat flux (CHF)-based design safety criteria to a “time and temperature” criterion is being discussed for light water reactors. To make it possible, however, large uncertainties in the post-CHF heat transfer regime need to be addressed. This work investigates the uncertainty of the transition boiling models adapted in three commonly used nuclear fuel performance and thermal-hydraulic codes: BISON, TRACE, and RELAP. This is achieved by comparing the calculation results against the instability tests performed at the Karlstein Thermal Hydraulic Test Loop (KATHY) facility. The data were collected using a 10×10 rod assembly prototypical of a boiling water reactor (BWR) under natural circulation at operational pressure and power. The heat transfer behavior in the transition regime is divided into two phases: the “dry-out” and the “rewet” phases. The results show that the studied models do not fully capture the differences between these phases. The predicted wall superheat does not agree with the experimental values, while heat transfer coefficients show even greater discrepancy. Furthermore, all models are unable to predict the “failure to rewet” behavior at the end of the heating cycle. Therefore, this study concludes that further improvement of the nuclear safety codes is essential to differentiate and capture the transition boiling heat transfer, which significantly impact the transient clad temperature over time after CHF is achieved.