Post-irradiation characterization of a high burnup mixed oxide fuel rod with minor actinides

Frazer, David; Cappia, F.; Harp, J. M.; Medvedev, P. G.; McClellan, K. J.; Voit, S. L.; Giglio, J.; Jädernäs, D.; Hosemann, P.

doi:10.1016/j.jnucmat.2022.153545

Title: Post-irradiation characterization of a high burnup mixed oxide fuel rod with minor actinides

Journal Article · 29 January 2022 · Journal of Nuclear Materials

DOI: https://doi.org/10.1016/j.jnucmat.2022.153545 · OSTI ID:1959325

Frazer, David ^[1]; Cappia, F. ^[2]; Harp, J. M. ^[2]; Medvedev, P. G. ^[2]; McClellan, K. J. ^[3]; Voit, S. L. ^[3]; Giglio, J. ^[2]; Jädernäs, D. ^[2]; Hosemann, P. ^[4]

University of California, Berkeley, CA (United States); Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
University of California, Berkeley, CA (United States); Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA (United States)

The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Here, it was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.

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Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States); Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA (United States)

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

Grant/Contract Number:: AC02-05CH11231; AC07-05ID14517

OSTI ID:: 1959325

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 562; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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