Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in nuclear industry as cladding materials for BWRs and PWRs, respectively. Over more than 60 years these materials displayed 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 [1-3]. However, over the last couple of years energetic efforts were undertaken to improve their oxidation resistance during off-normal temperature excursions, as well as to further improve upon the already achieved levels of mechanical behavior and reduced hydrogen uptake [1-3]. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of spent nuclear fuel. This paper 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.
ASJC Scopus subject areas
- Nuclear and High Energy Physics
- Materials Science(all)
- Nuclear Energy and Engineering