A model for pellet-cladding interaction (PCI) fracture of light-water reactor (LWR) fuel rods is presented, the basis of which is that Zircaloy cladding fails by iodine stress corrosion cracking (SCC). Laboratory data on iodine SCC of irradiated Zircaloy provide the primary input to the model, but unirradiated Zircaloy SCC data and theoretical analyses are utilized to broaden the regime of validity to encompass the various power reactor observations.

The critical events leading to stress corrosion failure of fuel rods are poorly understood. The aim of the present work is to acquire data on stress corrosion cracking (SCC) initiation and crack growth and to use these data for improving our understanding of fuel rod failures.