TEM characterization of irradiated U-7Mo/Mg dispersion fuel

Abstract This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3 , 7.4 × 10 14 f/cm 3 /s and 123 °C, and 5.5 × 10 21 f/cm 3 , 11.0 × 10 14 f/cm 3 /s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3 Mg 2 and Al 12 Mg 17 along with precipitates of MgO, Mg 2 Si and FeAl 5.3 . No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.