Ductility and strain rate sensitivity of Zircaloy-4 nuclear fuel claddings

The circumferential mechanical properties of nuclear fuel claddings for Canada deuterium uranium (CANDU) power reactors were examined and the constitutive equation which can predict the temperature dependence of ductility and strain rate sensitivity was developed. The loss of ductility associated with dynamic strain aging was observed between 250°C and 400°C. The elongation minimum results from the concentration of deformation in the necked region in the temperature range of the flow stress plateau. Oxygen atoms actually strengthen the alloy. However, the strengthening of the alloy due to dynamic strain aging decreases the strain rate sensitivity in the temperature range of the flow stress plateau. The decrease in the strain rate sensitivity results in a low ductility. Above the temperature range of the flow stress plateau the elongation of the alloy increases rapidly with temperature. The prediction based on a dynamic strain aging model is in good agreement with the temperature dependence of the circumferential ductility of Zircaloy-4 nuclear fuel claddings.