A MODEL FOR UNIFORM ZIRCALOY CLAD CORROSION IN PRESSURIZED-WATER REACTORS

A phenomenological water-side corrosion model for Zircaloy fuel cladding for pressurized water reactors (PWRs) is considered. The model acounts for the breakaway transition in the Zircaloy oxidation rate that takes place in an isothermal condition and the changes that occur during reactor operation, i.e. the dependence of oxide growth on fast neutron flux and cladding oxide layer thickness. Closed-form analytical solutions of the oxidation kinetics equations are obtained. The corrosion kinetics model is coupled to PWR thermal and hydraulic models which assume a subchannel that is either a closed single channel or a multichannel which accounts for coolant cross-flow and coolant enthalpy mixing. Both single-phase forced convection and subcooled nucleate boiling are accounted for in the thermal-hydraulic models. The model calculates the coolant temperature at the axial midplane of each axial segment of the fuel rod. When an oxide layer is present, the temperature at the metal-oxide interface is determined. This temperature in turn is used to determine the oxide growth via the Arrhenius temperature dependence of the Zircaloy oxidation rate. The predictions of the model have been compared with the measured cladding oxide data obtained in PWRs. The data for a given rod were obtained at various burn-ups (at the end of reactor cycles) and various axial positions of the rod. Our evaluations show that the model predicts the measured data satisfactorily; however, the deviations are discussed. The model has been used to study the effect of core loading patterns on cladding oxide growth. Our analyses show that core nuclear design is an important factor for water-side corrosion of fuel rods.