Accident-Tolerant Fuel Cladding and Zirconium Alloy Corrosion
This research community focuses on the materials science of nuclear fuel cladding, specifically developing zirconium alloys and protective coatings that resist oxidation and corrosion in water-cooled reactors.
The work centers on zirconium alloys, such as Zircaloy-4, and alternative FeCrAl alloys designed for accident-tolerant fuel applications. Key research themes include high-temperature steam oxidation, corrosion behavior in primary water, and the development of chromium-based coatings to enhance durability. The community also investigates the mechanical properties and microstructural evolution of these materials under reactor conditions, including the impact of hydrogen uptake and fretting corrosion. Recent efforts emphasize understanding degradation mechanisms at the grain boundary level and optimizing coating parameters to prevent failure during long-term service.
The largest share of the community's output is found in zirconium research, accounting for 3.0% of all zirconium research, with 4,608 papers in this group. Hydrogen research follows with 1,657 papers, representing 0.3% of that element's total research.
The community comprises 9,133 papers, with the most frequent publication venues being the Journal of Nuclear Materials, Nuclear Engineering and Design, and Corrosion Science.
Recent papers continue to focus on the degradation mechanisms of chromium-coated zirconium alloys during steam oxidation and fretting corrosion in simulated reactor water, including studies on grain boundary diffusion and coating delamination.