Analysis of COBRA-TF Critical Heat Flux Models for Triangular Fuel Assembly Pitch

The Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code is a best estimate code which is used to perform thermal-hydraulic analyses of a light water reactor vessel. The COBRA-TF heat transfer package has models calculating the critical heat flux (CHF). In order to verify code's predictive capability for CHF for triangular nuclear fuel assembly pitch, experiment results and system parameters of KS-1 test facility that is single loop model of the VVER primary system located in Kurchatov Institute were simulated using COBRA-TF. In order to calculate CHF for triangular nuclear fuel assembly pitch there is a look-up table generated. In addition, there are some correction factors for critical heat flux values listed the table. In this study, once KS-1 experiment results were simulated using COBRA-TF, correction factors were used to correct CHF results from COBRA-TF calculations