Post-irradiation examinations of UO2 composites as part of the Accident Tolerant Fuels Campaign

Cappia, Fabiola; Harp, Jason M.; McCoy, Kevin

doi:10.1016/j.jnucmat.2019.01.050

Post-irradiation examinations of ${UO}_{2}$ composites as part of the Accident Tolerant Fuels Campaign

Journal Article · Tue Feb 05 00:00:00 EST 2019 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2019.01.050· OSTI ID:1498271

^[1]; ^[1]; McCoy, Kevin ^[2]

Idaho National Lab. (INL), Idaho Falls, ID (United States), Characterization and Advanced Post Irradiation Examination Division
Framatome Inc., Lynchburg, VA, Fuel Division

Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO₂-SiC) and diamond (UO₂-diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction under representative light water reactor power conditions. Both non-destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO₂-SiC pellets revealed enhanced cracking when compared to a UO₂ rodlet also irradiated under similar conditions. Additionally instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300°C. The microstructure of the diamond-added UO₂ was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature. The postirradiation examination results cast doubt on the suitability of these additives to improve UO₂ fuel performance in a way that would lead to enhanced accident tolerance.

View Accepted Manuscript (DOE)

Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE); USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)

Grant/Contract Number:: AC07-05ID14517

OSTI ID:: 1498271

Alternate ID(s):: OSTI ID: 1636093
OSTI ID: 22886719

Report Number(s):: INL/JOU--18-51148-Rev000

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: C Vol. 517; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Cited By (1)

Advanced Postirradiation Characterization of Nuclear Fuels Using Pulsed Neutrons Vogel, Sven C.; Bourke, Mark A. M.; Craft, Aaron E. JOM, Vol. 72, Issue 1 https://doi.org/10.1007/s11837-019-03849-2	journal	November 2019

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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
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