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

Article dans une revue avec comité de lecture
Auteur
ZWEIFEL, T.
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
VALOT, Ch.
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
PONTILLON, Y.
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
LAMONTAGNE, J.
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
VERMERSCH, A.
BLAY, T.
PETRY, W.
PALANCHER, H.
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
BARRALLIER, Laurent
211915 Mechanics surfaces and materials processing [MSMP]

URI
http://hdl.handle.net/10985/8337
DOI
10.1016/j.jnucmat.2014.05.052
Date
2014
Journal
Journal of Nuclear Materials

Résumé

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.

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