Treat sodium loop experiments on performance of unbonded, unirradiated EBR-II Mark I fuel elements

Abstract Description of sodium loop experiments performed in the Transient Reactor Test (TREAT) Facility specifically to provide guidance for analyses on the behavior of an element which either has no sodium thermal bond between fuel and cladding or which, due to some unspecified defect, loses the bond upon reactor startup. Transient heat transfer calculations of the experiments were performed utilizing an idealized model of meltdown in which fuel slumping against the cladding was assumed to occur instantaneously along the length of the pin when the fuel surface at the axial midplane reached 1050°C. The result of TREAT experiments and of the calculations imply that although fuel melting may accompany a loss-of-bond type failure in an EBR-II driver fuel element, the short term consequences are acceptable from the viewpoint of reactor safety. Comparison of results from calculations and experiment indicate that the relatively simple modelling used represents a conservative description of the temperature rise in the cladding following fuel-cladding contact resulting from the fuel melting.