The ductility of zirconium-based alloys after steam oxidation at high temperatures, followed by quenching in water, is of significant interest for the nuclear industry. High-temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents. The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation.

Aiming to study possible differences in post-quench ductility caused by differences in chemical composition, extensive testing on Zircaloy-4, ZIRLO^®, and Optimized ZIRLO™ cladding materials as a function of hydrogen content, extent of oxidation, and peak cladding temperature was performed. The results showed similar ductile-to-brittle transition for the alloys studied despite differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile to brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen-enriched layer that forms underneath the oxide, both during the time at high temperatures and during cooling. Significantly different morphologies of the oxygen-enriched layer were observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen-enriched layer are discussed.