CHF Prediction Evaluation for START Sub-channel Code

Critical Heat Flux (CHF) is the value of heat flux at which heat transfer from fuel rod to coolant is suddenly deteriorated due to formation of vapor film. This causes large increase in fuel rod surface temperature and can lead to fuel rod damage. CHF is a limiting thermal phenomenon in nuclear power reactors and is a complex function of geometry and flow conditions. Various correlations such as EPRI, BIASI, CISE-4, W3, to mention a few, have been utilized to determine the CHF. Based on database of many CHF experiments, another approach is to use CHF lookup table (LUT). There have been different versions of CHF LUT i.e. 1986, 1995 and 2006. These LUTs are developed for a wide range of geometrical and flow conditions. In this work, in-house code Steady and Transient Analyzer for Reactor Thermal Hydraulics, START [1] has been equipped with 2006 Groeneveld CHF LUT [2] for LWR conditions. Preliminary validation exercise has been carried out against PWR sub-channel and bundle tests (PSBT) [3] database. For both steady state and transient cases, CHF has been predicted for different test conditions. Comparison with experimental values is carried out and is presented.