Assessment of Biasi and Columbia University CHF correlations with GE 3x3 rod bundle experiment

The critical heat flux (CHF), at which a sudden degradation of heat transfer occurs without corresponding decrease in heat generation, is one of the limiting parameters for safe operation of nuclear reactors. Reactor operation beyond the CHF causes a rapid rise in fuel cladding temperature and thus should be avoided to maintain the fuel element integrity. Reactor power limits are therefore set so that a prescribed safety margin below the CHF is maintained. Two CHF correlations are evaluated for reactor core thermal hydraulic analysis: the Biasi correlation and the Columbia University correlation. The BODYFIT-2PE computer code is used for this assessment. The CHF predicted by the BODYFIT-2PE using the two correlations is compared with GE 3x3 rod bundle CHF experiment