Burst and oxidation behavior of Cr-coated Zirlo during simulated LOCA testing

S. B. Bell; T. Graening; A. Evans; P. Kelly; B. A. Pint; K. A. Kane

doi:10.1016/j.jnucmat.2022.153679

Burst and oxidation behavior of Cr-coated Zirlo during simulated LOCA testing

S. B. Bell, T. Graening, A. Evans, P. Kelly, B. A. Pint, K. A. Kane

Research output: Contribution to journal › Article › peer-review

14 Scopus citations

Abstract

Cr-coated Zr-alloys are a near-term cladding concept to improve reactor safety during accident scenarios. Burst and steam oxidation behavior of bare and Cr-coated Zirlo claddings were examined under simulated loss-of-coolant accident conditions. The 4.4 µm coating had no substantial effect on ballooning or opening geometry but did increase burst temperatures at higher pressures. The coating reduced steam oxidation of the cladding compared to bare specimens, but in regions of high strain, the coating developed through-cracks allowing rapid underlying zirconia formation.

Original language	English
Article number	153679
Journal	Journal of Nuclear Materials
Volume	564
DOIs	https://doi.org/10.1016/j.jnucmat.2022.153679
State	Published - Jun 2022

Funding

Notice: This manuscript has been authored by UT-Battelle, LLC under Contract No. DE-AC05–00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ). The authors would like to thank B. Johnston for conducting the burst testing presented in this study, as well as M. Ridley and B. Garrison for technical review. This work was performed as a part of the Advanced Fuels Campaign directed by the United States Department of Energy Office of Nuclear Energy. The study was part-funded under the ?46 m Advanced Fuel Cycle Programme as part of the Department for Business, Energy and Industrial Strategy's (BEIS) ?505 m Energy Innovation Programme. The authors would also like to thank Westinghouse Electric Sweden AB for providing samples of Zirlo tubing for testing.

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abstract = "Cr-coated Zr-alloys are a near-term cladding concept to improve reactor safety during accident scenarios. Burst and steam oxidation behavior of bare and Cr-coated Zirlo claddings were examined under simulated loss-of-coolant accident conditions. The 4.4 µm coating had no substantial effect on ballooning or opening geometry but did increase burst temperatures at higher pressures. The coating reduced steam oxidation of the cladding compared to bare specimens, but in regions of high strain, the coating developed through-cracks allowing rapid underlying zirconia formation.",

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AU - Kane, K. A.

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AB - Cr-coated Zr-alloys are a near-term cladding concept to improve reactor safety during accident scenarios. Burst and steam oxidation behavior of bare and Cr-coated Zirlo claddings were examined under simulated loss-of-coolant accident conditions. The 4.4 µm coating had no substantial effect on ballooning or opening geometry but did increase burst temperatures at higher pressures. The coating reduced steam oxidation of the cladding compared to bare specimens, but in regions of high strain, the coating developed through-cracks allowing rapid underlying zirconia formation.

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Burst and oxidation behavior of Cr-coated Zirlo during simulated LOCA testing

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