Electron microscopy characterization of the fuel-cladding interaction in medium burnup annular fast reactor MOX

Cappia, F.; Winston, A.; Miller, B.; Kombaiah, B.; Teng, F.; Murray, D.; Frazer, D.; Harp, J. M.

doi:10.1016/j.jnucmat.2021.152922

Electron microscopy characterization of the fuel-cladding interaction in medium burnup annular fast reactor MOX

Journal Article · Sun Mar 07 00:00:00 EST 2021 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2021.152922· OSTI ID:1782045

^[1]; Winston, A. ^[1]; Miller, B. ^[1]; Kombaiah, B. ^[1]; Teng, F. ^[1]; Murray, D. ^[1]; ^[1]; ^[2]

Idaho National Lab. (INL), Idaho Falls, ID (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

In this work, we present electron microscopy data focused on the fuel-cladding interaction layer in annular fast reactor MOX with HT-9 cladding at medium burnup. In agreement with previous literature data, the volatile fission products Cs, Te and I have migrated radially into the extreme fuel periphery and partially interacted with the cladding. The accumulation of Cs has occurred in the outermost rim of the fuel pellet, where grain recrystallization has also been observed. Significant amounts of Pd have been found in the interaction zone, particularly in the sample taken from the upper half of the fissile column where the cladding temperatures are higher. At this axial location, Cr has been enriched at the cladding inner surface and diffused into the fuel. Furthermore, chromium remains mainly in metallic form, but locally formed oxides. The fission products Cs, Te and I are found, with variable composition, in form of nanocrystalline regions dispersed in the metallic Cr-rich layer. The morphology and chemical characteristics of the layer suggest a non-oxidative corrosion mechanism as principal cladding degradation phenomenon occurring in this sample, with local onset of Cr oxidation within the nanocrystalline precipitates.

View Accepted Manuscript (DOE)

Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1782045

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 551; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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