Atomistic modeling of intrinsic and radiation-enhanced fission gas (Xe) diffusion in UO 2 ± x: Implications for nuclear fuel performance modeling

D. A. Andersson, P. Garcia, X. Y. Liu, G. Pastore, M. Tonks, P. Millett, B. Dorado, D. R. Gaston, D. Andrs, R. L. Williamson, R. C. Martineau, B. P. Uberuaga, C. R. Stanek

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Based on density functional theory (DFT) and empirical potential calculations, the diffusivity of fission gas atoms (Xe) in UO2 nuclear fuel has been calculated for a range of non-stoichiometry (i.e. UO2 ±x), under both out-of-pile (no irradiation) and in-pile (irradiation) conditions. This was achieved by first deriving expressions for the activation energy that account for the type of trap site that the fission gas atoms occupy, which includes the corresponding type of mobile cluster, the charge state of these defects and the chemistry acting as boundary condition. In the next step DFT calculations were used to estimate migration barriers and internal energy contributions to the thermodynamic properties and calculations based on empirical potentials were used to estimate defect formation and migration entropies (i.e. pre-exponentials). The diffusivities calculated for out-of-pile conditions as function of the UO2±x non-stoichiometry were used to validate the accuracy of the diffusion models and the DFT calculations against available experimental data. The Xe diffusivity is predicted to depend strongly on the UO2±x non-stoichiometry due to a combination of changes in the preferred Xe trap site and in the concentration of uranium vacancies enabling Xe diffusion, which is consistent with experiments. After establishing the validity of the modeling approach, it was used for studying Xe diffusion under in-pile conditions, for which experimental data is very scarce. The radiation-enhanced Xe diffusivity is compared to existing empirical models. Finally, the predicted fission gas diffusion rates were implemented in the BISON fuel performance code and fission gas release from a Risø fuel rod irradiation experiment was simulated.

Original languageEnglish (US)
Pages (from-to)225-242
Number of pages18
JournalJournal of Nuclear Materials
Issue number1-3
StatePublished - Aug 2014

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Materials Science(all)
  • Nuclear Energy and Engineering


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