An evaluation of tri-valent oxide (Cr2O3) as a grain enlarging dopant for UO2 nuclear fuels fabricated under reducing environment

JOURNAL OF NUCLEAR MATERIALS

卷:553

Article

Silva, Chinthaka M.^1,2  Hunt, Rodney D.¹  Holliday, Kiel S.² 

[1] Oak Ridge Natl Lab, One Bethel Valley Rd, Oak Ridge, TN USA

[2] Lawrence Livermore Natl Lab, Livermore, CA 94550 USA

关键词: Nuclear fuel;    Doping UO2;    Cr2O3-doped UO2;    Lattice parameter;    Crystallite size;    Microstrain;   

DOI  :  10.1016/j.jnucmat.2021.153053

来源: Elsevier

【 摘 要 】

A study was performed to evaluate the microstructure and crystallography of nominally 500-2000 Cr2O3-doped UO2 fabricated in a temperature range of 1150-1750 degrees C under reducing experimental conditions. An increase in grain size of the samples was observed with the increase in heat treating temperature as expected. For a given sintering temperature (1700-1750 degrees C), an increase in the grain size was also observed with the increase in Cr2O3 concentration up to a value of 1000-1200 wppm. A decrease in fission gas release as a function of grain size was estimated for the Cr2O3-doped UO2 samples assuming specified post-irradiation annealing conditions. A nearly linear decrease was obtained in the lattice parameter of the Cr2O3-doped UO2 fcc phase with the increase in Cr2O3 concentration, especially up to a nominal value of 1000 wppm. The lattice parameter decrease was also persistent with the increase in the average grain size as a result of addition of Cr2O3 into the UO2 lattice. An increase in the crystallite size and a decrease in the microstrain of the fcc phase were observed with the increase in the average grain size of the samples, indicating a higher crystallinity of the Cr2O3-doped samples than that of the undoped UO2 sample. (C) 2021 Elsevier B.V. All rights reserved.

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