On the oxidation state of UO2 nuclear fuel at a burn-up of around 100 MWd/kgHM

Abstract Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local Δ G ¯ ( O 2 ) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the Δ G ¯ ( O 2 ) of the fuel increased from −370 kJ mol−1 at r/r0 = 0.1 to −293 kJ mol−1 at r/r0 = 0.975. Thus, the Δ G ¯ ( O 2 ) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

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