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

T. Zweifel (1) , Ch. Valot (1) , Yves Pontillon (1) , J. Lamontagne (1) , A. Vermersch (2) , Laurent Barrallier (2) , T. Blay (1) , W. Petry (3) , H. Palancher (1)

1 CEA-DES (ex-DEN) - CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire)
2 MSMP - Mechanics surfaces and materials processing
3 TUM - Technische Universität Munchen - Technical University Munich - Université Technique de Munich

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.

Mots clés: Combustible nucléaire; Microstructure

Domaines: Matériaux

T. Zweifel, Ch. Valot, Yves Pontillon, J. Lamontagne, A. Vermersch, et al.. Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel. Journal of Nuclear Materials, 2014, 452 (1-3), pp.533-547. ⟨10.1016/j.jnucmat.2014.05.052⟩. ⟨hal-01021884⟩