2013:24 Modelling of nuclear fuel cladding under loss-of-coolant accident conditions

We present a unified model for calculation of zirconium alloy fuel cladding rupture during a postulated loss-of-coolant accident in light water reactors. The model treats the Zr alloy solid-to-solid phase transformation kinetics, cladding creep deformation, oxidation and rupture as a function of temperature and time in an integrated fashion during the transient. The fuel cladding material considered here is Zircaloy-4, for which material property data(model parameters) are taken from literature. We have modelled and simulated single-rod transient burst tests in which the rod internal pressure and the heating rate were kept constant during each test. The results are compared with experimental data on cladding rupture strain, temperature and pressure. The effects of heating rate and rod internal pressure on the rupture strain are evaluated by systematic parameter variations of these quantities. The principal uncertainty in our simulations is the treatment of creep deformation in the twophase (α + β) region of the considered alloy, for which no definite constitutive relation is yet available.