Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate sample...

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Bibliographic Details
Published in:Journal of nuclear materials 2014-09, Vol.452 (1-3), p.533-547
Main Authors: Zweifel, T., Valot, Ch, Pontillon, Y., Lamontagne, J., Vermersch, A., Barrallier, L., Blay, T., Petry, W., Palancher, H.
Format: Article
Language:English
Subjects:
Engineering Sciences
Materials
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Summary:U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding.
ISSN:0022-3115
1873-4820
DOI:10.1016/j.jnucmat.2014.05.052