Post-quench ductility of zirconium alloy cladding materials - 2016-0088
- Westinghouse Electric Company, Fuel Engineering and Safety Analysis, 1332 Beulah Rd., Pittsburgh, PA 15235 (United States)
- Westinghouse Electric Company, Fuel Engineering and Safety Analysis, 5801 Bluff Rd., Hopkins, SC (United States)
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{sup R}, and Optimized ZIRLO{sup TM} 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. (authors)
- Research Organization:
- ASTM International, 100 Barr Harbor Drive, P.O. Box C700, West Conshohocken, PA, 19428-2959 (United States)
- OSTI ID:
- 22788433
- Country of Publication:
- United States
- Language:
- English
Similar Records
Cladding embrittlement during postulated loss-of-coolant accidents.
Alloying effect of Nb and Sn on the Zirconium alloy fuel claddings behavior at high temperature oxidation in steam - 6.1 STP-2016-0072