Assessment of TRACE code against CHF experiments

Full text of publication follows: This paper reports on the validation of the USNRC 'consolidate' code TRACE with data obtained during Critical Heat Flux (CHF) experiments in single channels and round and annular tubes. CHF is one of the key reactor safety parameters, because it determines the conditions for the onset of transition boiling in the core rod bundles, leading to the low heat transfer rates characteristics of the post-CHF heat transfer regime. In the context of the participation of PSI in the the International Programme for uncertainty analysis BEMUSE, we have carried out extensive work for the validation of some important TRACE models. The present work is aimed at assessing the range of validity for the CHF correlations and post-CHF heat transfer models currently included in TRACE. The heat transfer experiments selected for the assessment were performed at the Royal Institute of Technology (RIT) in Stockholm, Sweden and at the Atomic Energy Establishment in Winfrith, UK. The experimental investigations of the CHF and post-CHF heat transfer at RIT for flow of water in vertical tubes and annulus were performed at pressures ranging from 1 to 20 MPa and coolant mass fluxes from 500 to 3000 kg/m2s. The liquid was subcooled by 10 deg. C and 40 deg. C at the inlet of the test section. The experiments were performed on two different types of test sections. Experiments with uniformly heated single 7.0 m long tubes were carried out with three different inner tube diameters of 10, 14.9 and 24.7 mm. A series of experiments with non-uniform axial power distribution were also conducted in order to study the effect of the axial heat flux distribution on the CHF conditions in both 7.0 m long single tubes and 3.65 long annulus. Several different axial power profiles were employed with bottom, middle and top power peaks as well as the double-humped axial power profiles. In total more than 100 experiments with uniform axial heat flux distribution and several hundreds of experiments with varying (non-uniform) axial heat flux were simulated with TRACE. The AEE experiments dealt with forced convection burn-out in high pressure tube configurations both with uniform and non-uniform axial heat flux distributions. The pressure range was 3 to 11 MPa, the mass flux spanned from 1020 to 4070 kg/m2s, and the subcooling varied from 2.8 C to 90 C The analysis of the results of TRACE code assessment against the RIT and AEE CHF and post-CHF experiments allowed to evaluate the parameter validity limits, within which the TRACE predictions for the CHF as well as for the heat transfer in transition and film boiling regions are in good agreement with the experimental data. The analysis of the TRACE calculation results shows that CHF is predicted rather well for the higher pressure range (above 10 MPa) for all investigated flow conditions. For the lower pressure region the TRACE predictions were found to be valid only for a narrow range of the coolant mass flux. (authors)