Creep properties of Zircaloy-4 for nuclear fuel cladding FEA simulation

Zirconium alloys are commonly used as cladding tube material for nuclear water reactors. To improve the understanding of the creep damage accumulation in thin walled fuel cladding tubes made of Zircaloy-4, data collation (tensile, creep strain and rupture) and material modelling has been performed for use in finite element analysis (FEA). In literature there are two distinct areas of creep modelling: creep strain response to short power transients and long term creep strain evolution for storage purposes. In this paper the short term creep strain response is mainly targeted for FEA simulation of fuel-cladding interaction. In addition, the model performance in predicting long term creep strain is verified from the available public domain data. The creep rupture models are optimized for predicting biaxial deformation (hoop strain) of thin walled tubes. The relevant temperature range is selected for postulated system disturbances, i.e. power transients between 300 and 600°C. For the preliminary FEA simulations the material is assumed to be un-irradiated, cold worked and stress relived. The base material models (constitutive equations) do not at this stage incorporate the effect of anisotropy, however two methods of incorporating irradiation effect are presented. The main models applied for this work are the Wilshire equations (WE) for rupture and the logistic creep strain prediction (LCSP) model for strain.