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Steam Oxidation Evaluation of Fe–Cr Alloys for Accident Tolerant Nuclear Fuel Cladding

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Abstract

New nuclear fuel cladding materials are being evaluated that can withstand steam environments ≥1200 °C for short (≤4 h) periods in case of a beyond design basis accident. This study focused on commercial and model Fe–Cr alloys, where there is considerable experience in fabricating and joining. Exposures in 1 bar steam and air for 4 h at 800–1300 °C showed that the commercial Fe–Cr alloys were very sensitive to composition and only Fe-25.8%Cr-1%Mo formed a protective chromia scale at 1200 and 1300 °C in steam. A model Fe-22.5%Cr + Mn,Si,Y alloy also formed a protective scale at 1200 °C in steam. Analytical transmission electron microscopy of the reaction products revealed that (1) nominally equiaxed Cr2O3 formed at 1000–1200 °C; (2) at 1000 °C, there was a Mn inner and outer layer but at 1100 and 1200 °C only an outer layer was observed; (3) an amorphous SiO2 inner layer was observed at 1000 and 1100 °C, but the SiO2 was crystalline on the 22.5%Cr model alloy at 1200 °C, which was confirmed by electron and X-ray diffraction; and (4) Fe was found throughout the Cr2O3 formed on alloys without Mn at 1200 °C in steam and air, Fe-rich oxide near the gas interface and Fe-rich metal precipitates near the metal–oxide interface. A few Fe-rich precipitates were detected in oxides formed at 1100 °C and none at 1000 °C. The incorporation of Fe and crystallization of SiO2 at 1200 °C may be detrimental to the formation of a protective chromia scale in steam at ≥1200 °C for this application and explain why such high Cr contents are needed for protective behavior.

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Acknowledgements

The experimental work was conducted by M. Howell, M. Stephens, T. Lowe, D. Coffey, T. Jordan and E. Cakmak. S. Dryepondt and K. Terrani provided useful comments on the manuscript. This research was funded by the U.S. Department of Energy’s Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program. The FEI Talos F200X STEM was used as part of the Nuclear Science User Facility.

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  1. Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN, 37831-6156, USA

    B. A. Pint & K. A. Unocic

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Pint, B.A., Unocic, K.A. Steam Oxidation Evaluation of Fe–Cr Alloys for Accident Tolerant Nuclear Fuel Cladding. Oxid Met 87, 515–526 (2017). https://doi.org/10.1007/s11085-017-9754-0

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