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Journal Articles Journal of Nuclear Materials Year : 2017

Oxygen diffusion model of the mixed (U,Pu)O2 ± x: Assessment and application

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Abstract

The uranium-plutonium (U,Pu)O2 ± x mixed oxide (MOX) is used as a nuclear fuel in some light water reactors and considered for future reactor generations. To gain insight into fuel restructuring, which occurs during the fuel lifetime as well as possible accident scenarios understanding of the thermodynamic and kinetic behavior is crucial. A comprehensive evaluation of thermo-kinetic properties is incorporated in a computational CALPHAD type model. The present DICTRA based model describes oxygen diffusion across the whole range of plutonium, uranium and oxygen compositions and temperatures by incorporating vacancy and interstitial migration pathways for oxygen. The self and chemical diffusion coefficients are assessed for the binary UO2 ± x and PuO2 − x systems and the description is extended to the ternary mixed oxide (U,Pu)O2 ± x by extrapolation. A simulation to validate the applicability of this model is considered.
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Dates and versions

cea-02381332 , version 1 (26-11-2019)

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Emily Moore, Christine Guéneau, Jean-Paul Crocombette. Oxygen diffusion model of the mixed (U,Pu)O2 ± x: Assessment and application. Journal of Nuclear Materials, 2017, 485, pp.216-230. ⟨10.1016/j.jnucmat.2016.12.026⟩. ⟨cea-02381332⟩
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