Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding
materials for light water reactor (LWR) fuels. These materials display a very good combination of
properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced
hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last
couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve
fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to
improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to
facilitate the development of such novel materials, it is very important to achieve not only engineering
control, but also a scientific understanding of the underlying material degradation mechanisms, both in
working conditions and in storage of used nuclear fuel.
This report strives to contribute to these efforts by constructing the thermodynamic models of both
alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was
placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of
oxidation. To that end, computational thermodynamics calculations were conducted concurrently with
first-principles atomistic modeling.