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 degrees 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 degrees C and 670 degrees C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 degrees C FGs are released from IDL/matrix interfaces. The second peak at 670 degrees C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding. (C) 2014 Elsevier B.V. All rights reserved.