Implementation and validation of CHF-models in the two-phase porous-media code TWOPORFLOW

•Extension of TWOPORFLOW with CHF correlations.•Validation of CHF-models with steady state data.•Validation of CHF-models with transient data.•Improved and validate TWOPORFLOW version.•Potential applications for LWRs. The local deterioration of the heat transfer at certain elevation of the fuel rods of a water-cooled reactor may occur either in case of dry-out of the liquid film (in case of BWR) or due to Departure from Nucleate Boiling (DNB, in case of PWR). Under such local conditions, a sudden cladding temperature escalation may take place challenging the integrity of the cladding material i.e., of a safety barrier. For a reliable evaluation of the safety features of a nuclear reactor under nominal or accidental conditions, the use of validated numerical tools using relevant experimental data is mandatory. In this paper, the implementation of different correlations (Biasi, Bowring, and Groeneveld Look-Up Table) for the prediction of the Critical Heat Flux (CHF) in TWOPORFLOW as well as its validation is presented and discussed. Based on the performed investigations it can be stated that the implemented correlations predict the CHF-value and their location of appearance in acceptable agreement to the measured data.