Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel
Article dans une revue avec comité de lecture
Author
ZWEIFEL, T.
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
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)]
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)]
92792 CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) [CEA-DES (ex-DEN)]
Date
2014Journal
Journal of Nuclear MaterialsAbstract
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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