Thermal-hydraulics analysis of flow blockage events for fuel assembly in a sodium-cooled fast reactor

The subchannel code ATHAS-LMR has been improved for analyzing thermal-hydraulics characteristics of flow blockage events in the fuel assembly of a sodium-cooled fast reactor (SFR). Based on the in-depth study of the phenomenon of flow blockage and combined with the characteristics of fast reactor assembly, enhanced models for the wire-wrap spacer and convective heat transfer have been added to account for changes in transverse flow in the presence of blockages. And appropriate transverse mixing and local resistance model are also implemented in ATHAS-LMR-FB. The improved code has been assessed against experimental data obtained at adiabatic and diabatic conditions in SFR assemblies with simulated flow blockages. Good agreement between predictions and experimental data on velocity and temperature distributions downstream of the blockage has been observed. Comparison of predictions of various analytical tools have also been performed. Predictions of the improved ATHAS-LMR code are shown to be compatible with those of other analytical tools. An in-depth analysis of effects of location and size of blockage has been carried out for the fuel assembly of the China Experimental Fast Reactor. Boiling has been predicted to occur for blockage ratios greater than 55.61% at the central location of the fuel assembly. Peak cladding and coolant temperatures are predicted at the axial location 70% of the rod length (just downstream of the maximum local power position of the non-uniform axial power profile).