Accident-Tolerant Fuel Cladding and Zirconium Alloy Degradation
This community investigates the materials that encase nuclear fuel rods, focusing on how zirconium alloys and their protective coatings withstand high-temperature steam, radiation, and corrosion in water reactors.
The work centers on zirconium alloys, specifically their behavior under steam oxidation and hydrogen uptake. A major strand of research examines accident-tolerant fuel cladding, including chromium coatings and iron-chromium-aluminum alloys designed to resist degradation during reactor accidents. The community also studies the mechanics of zirconium hydride formation and fracture, using molecular dynamics and finite element simulations to model how these materials fail under irradiation and thermal stress. Applications are strictly tied to pressurized water reactor safety, with a focus on extending the service life of fuel cladding tubes and understanding the microstructural changes that lead to corrosion or embrittlement.
The largest share of the community's output is found in zirconium research, accounting for 7.7% of all zirconium studies, and 3,001 papers here. This element leads both in share and in absolute paper count within the group.
The community comprises 4,931 papers, published primarily in the Journal of Nuclear Materials, Corrosion Science, and Nuclear Engineering and Design.
Recent work continues to focus on the performance of chromium-coated zirconium alloys and FeCrAl coatings under high-temperature steam conditions. New studies also examine the effects of radiation damage on zirconium microstructures and the mechanical behavior of hydride precipitates in single-crystal zirconium.
Papers behind this description
- Autoclave grid-to-rod fretting corrosion behaviors of the Zr alloy fuel cladding with and without Cr coating through advanced characterization — Corrosion Science, 2024 — doi:10.1016/j.corsci.2024.112379
- Review on chromium coated zirconium alloy accident tolerant fuel cladding — Journal of Alloys and Compounds, 2021 — doi:10.1016/j.jallcom.2021.162450
- High-temperature steam corrosion performance of FeCrAl alloys after long-term service in PWR condition: A reliable evaluation and multiscale microstructural characterization — Corrosion Science, 2026 — doi:10.1016/j.corsci.2026.113747
- Investigating zirconium alloy corrosion with advanced experimental techniques: A review — Journal of Nuclear Materials, 2023 — doi:10.1016/j.jnucmat.2023.154586
- Simulation of fracture behaviors in hydrogenated zirconium alloys using a crystal plasticity coupled phase-field fracture model — International Journal of Plasticity, 2025 — doi:10.1016/j.ijplas.2025.104304
- High temperature steam oxidation of chromium-coated zirconium-based alloys: Kinetics and process — Corrosion Science, 2020 — doi:10.1016/j.corsci.2020.108537
- An Overview of Mechanisms of the Degradation of Promising ATF Cladding Materials During Oxidation at High Temperatures — High Temperature Corrosion of Materials, 2024 — doi:10.1007/s11085-024-10229-y
- Steam oxidation of Cr-coated zirconium alloy claddings at 1200 °C: Kinetics transition and failure mechanism of Cr coatings — Journal of Nuclear Materials, 2023 — doi:10.1016/j.jnucmat.2023.154684
- Influence of coating parameters on oxidation behavior of Cr-coated zirconium alloy for accident tolerant fuel claddings — Corrosion Science, 2022 — doi:10.1016/j.corsci.2022.110359
- Isothermal experiments on steam oxidation of magnetron-sputtered chromium-coated zirconium alloy cladding at 1200 ℃ — Corrosion Science, 2022 — doi:10.1016/j.corsci.2022.110544
- Mechanisms of Hydride Nucleation, Growth, Reorientation, and Embrittlement in Zirconium: A Review — Materials, 2023 — doi:10.3390/ma16062419
- The interactions of deformation twins, zirconium hydrides, and microcracks — International Journal of Plasticity, 2024 — doi:10.1016/j.ijplas.2024.104149
- Superior corrosion resistance and crack tolerance of FeCrAl coatings on Zr alloy under hydrogenated and oxygenated PWR primary water — Corrosion Science, 2025 — doi:10.1016/j.corsci.2025.113492
- Cross-sectional helium irradiation revealing interface-controlled bubble evolution in Cr/CrAlSiN multilayer coatings on zirconium alloys — Journal of Materials Science & Technology, 2026 — doi:10.1016/j.jmst.2026.02.034
- Hetero-deformation mechanisms and GND-associated plastic strain partitioning across single/polycrystal hetero-boundary in nickel-based superalloys — International Journal of Plasticity, 2026 — doi:10.1016/j.ijplas.2026.104636
- Temperature effects on radiation damage in HCP-zirconium: A molecular dynamics study using a fine-tuned machine-learned potential — Journal of Nuclear Materials, 2025 — doi:10.1016/j.jnucmat.2025.156025
- Post-ballooning and burst steam oxidation of accident tolerant zirconium alloy cladding with cracked chromium coating — Journal of Nuclear Materials, 2025 — doi:10.1016/j.jnucmat.2025.156095
- A novel high-yttrium aluminosilicate glass-ceramic for fabricating dense SiCf/SiC composite joints with enhanced hydrothermal corrosion resistance — Journal of the European Ceramic Society, 2025 — doi:10.1016/j.jeurceramsoc.2025.118062