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The defect chemistry of UO2±x from atomistic simulations

Research output: Contribution to journalJournal article

E-pub ahead of print
<mark>Journal publication date</mark>1/03/2018
<mark>Journal</mark>Journal of Nuclear Materials
<mark>State</mark>E-pub ahead of print
Early online date1/03/18
<mark>Original language</mark>English


Control of the defect chemistry in UO2±x is important for manipulating nuclear fuel properties and fuel performance. For example, the uranium vacancy concentration is critical for fission gas release and sintering, while all oxygen and uranium defects are known to strongly influence thermal conductivity. Here the point defect concentrations in thermal equilibrium are predicted using defect energies from density functional theory (DFT) and vibrational entropies calculated using empirical potentials. Electrons and holes have been treated in a similar fashion to other charged defects allowing for structural relaxation around the localized electronic defects. Predictions are made for the defect concentrations and non-stoichiometry of UO2±x as a function of oxygen partial pressure and temperature. If vibrational entropy is omitted, oxygen interstitials are predicted to be the dominant mechanism of excess oxygen accommodation over only a small temperature range (1265 K–1350 K), in contrast to experimental observation. Conversely, if vibrational entropy is included oxygen interstitials dominate from 1165 K to 1680 K (Busker potential) or from 1275 K to 1630 K (CRG potential). Below these temperature ranges excess oxygen is predicted to be accommodated by uranium vacancies, while above them the system is hypo-stoichiometric with oxygen deficiency accommodated by oxygen vacancies. Our results are discussed in the context of oxygen clustering, formation of U4O9, and issues for fuel behavior. In particular, the variation of the uranium vacancy concentrations as a function of temperature and oxygen partial pressure will underpin future studies into fission gas diffusivity and broaden the understanding of UO2±x sintering.

Bibliographic note

This is the author’s version of a work that was accepted for publication in Journal of Nuclear Materials. Changes resulting from the publishing process, such as peer review, editing, corrections, structural formatting, and other quality control mechanisms may not be reflected in this document. Changes may have been made to this work since it was submitted for publication. A definitive version was subsequently published in Journal of Nuclear Materials, ???, ?, 2018 DOI: 10.1016/j.jnucmat.2018.02.034