PWR Zircaloy Fuel Cladding Corrosion Performance, Mechanisms, and Modeling

Abstract

Oxide axial profiles for Zircaloy-4 cladding measured for 17 fuel rods irradiated from 50 to 60 GWD/t in three different pressurized water reactors (PWRs) were studied. Computer simulation of the magnitude and shape of the various profiles involved five contributing interrelated effects: (1) thermal feedback, (2) radial thermal redistribution of corrosion hydrogen, (3) metallurgical variables (tin content and radiation-induced changes in intermetallic particle size distribution), (4) lithium hydroxide exposure history, and (5) radiation effects in the oxide film. Published research results supporting each separate effect are reviewed, and the selection of a predictive simulation for each phenomenon and the effect of varying its relative weight are reported. The effect on cladding corrosion of hydrogen thermal redistribution to the oxide-metal interface, coupled with thermal feedback, was found to be a predominant factor in predicting the magnitude and shape of cladding oxide profiles at high burnups for PWR fuel rods. Tin content, coupled with thermal feedback, was also found to have a very important effect on the predicted magnitude of oxide profiles at all burnups. Radiation-induced dissolution of fine Zr(Cr,Fe)2 intermetallic particles in the Zircaloy-4 cladding primarily affected the predicted shape of the oxide profiles at the colder inlet region of the fuel rods. Lithium hydroxide (LiOH) exposure history and radiation effects in the oxide film appeared to play non-negligible secondary roles in determining the oxide profiles at all burnups.