12Cr ODS steel samples were prepared by mechanical alloying of the metal powders with 20-30 nm Y2O3 particles followed by isostatic pressing, hot rolling and final heat treatment. Evolutions of oxide particles such as YTaO4 and YCrO3 after each fabrication step were investigated by using TEM with EDS. Crystallographic correlation between oxide particles and the matrix was investigated in a HIPped sample, and interactions between dislocations and oxide particles were observed in hot rolled or heat treated sample. Size distributions of oxide particles were measured by carbon replica samples and it was found that coarsening of oxide particles from 9 to 12 nm occurred during hot rolling process. Additional isothermal annealing at 1250 ˚C revealed that phase transformation of oxide particles from monoclinic YTaO4 to face centered cubic Y3TaO7 was observed.

Over the course of use, both in-service and during storage, fuel claddings for nuclear reactors undergo complex changes that can drastically change their material properties. Exposures to irradiation, temperature changes, and stresses, as well as contact with coolant, storage pool, and dry storage environments, may induce microstructural changes, such as formation of radiation defects, precipitate dissolution, and chemical segregation, that can ultimately result in failure of the cladding if pushed beyond its limit. In order to predict the performance of cladding in-service and during storage, understanding of the dominant processes related to these changes and their consequences is essential. In situ transmission electron microscopy (TEM) allows dynamic observation, at the nanoscale, of microstructural changes under a range of stimuli, making it an excellent tool for deepening our understanding of microstructural evolution in claddings. This proceeding presents details of the new in situ ion irradiation TEM and in situ gas cell TEM capabilities developed at Sandia National Laboratories. In addition, it will present the initial results from both systems investigating radiation tolerance of potential Generation IV cladding materials and understanding degradation mechanisms in Zr-based claddings of importance for dry storage.

Aqueous corrosion of zirconium alloys has become the major factor limiting prolonged fuel campaigns in nuclear plant. Studies using SEM, TEM and electrochemical impedance measurements have been interpreted as showing a dense inner-most oxide layer, and an increased thickness of the layer has been correlated to a better corrosion resistance. Many authors have reported that an ‘intermediate layer’ at the metal oxide interface has a complex structure or/and stochiometry different to that of both the bulk oxide and bulk metal, sometimes claimed to be a suboxide phase. Diffraction evidence has suggested the presence of both cubic ZrO and rhombohedral Zr3O phases, and compositional analysis has revealed similar variations in local oxygen stoichiometry.

We have carried out a systematic investigation of the structure and chemistry of the metal/oxide interface in samples of commercial ZIRLO corroded for times up to 180 days. We have developed new experimental techniques for the study of these interfaces both by Electron Energy Loss Spectroscopy (EELS) analysis in the Transmission Electron Microscope (TEM) and by Atom Probe Tomography (APT), and exactly the same samples have been investigated by both techniques. Our results show the development of a clearly defined suboxide layer of stoichiometry close to ZrO, and the subsequent disappearance of this layer at the first of the characteristic ‘breakaway’ transitions in the oxidation kinetics. We can correlate this behaviour with changes in the structure of the oxide layer, and particularly the development of interconnected porosity that links the corroding interface with the aqueous environment. Using high resolution SIMS analysis of isotopically spiked samples we demonstrate the penetration of the oxidising species through these porous outer oxide layers.