The primary circuit of pressurized water reactor nuclear power plants operates in a high-temperature and high-pressure aqueous environment. Commercial zirconium alloys, the dominant fuel cladding materials in current use, possess desirable properties including radiation damage resistance, low neutron absorption cross-section, and good corrosion resistance under high-temperature and high-pressure water conditions. This work aims to improve the accident tolerance of nuclear-grade materials by doping specific alloying elements into zirconium alloys. Combining first-principles calculations, molecular dynamics simulations, and high-temperature high-pressure aqueous corrosion experiments, the present study investigates the microstructural evolution and oxide film development of the composite material before and after oxidation in high-temperature and high-pressure water. The damage features of zirconium-based alloys and their oxide films under harsh service conditions are characterized, and the failure mechanisms at the medium-substrate-coating interface are systematically explored. The results reveal the preferential growth of corrosion products along specific crystallographic planes of the alloy oxide film in high-temperature and high-pressure water, providing theoretical support for predicting oxide film growth behavior. Quantitative evaluation of charge transfer is performed to analyze the atomic-scale interactions of water molecules and solution ions with alloy coatings and oxide layers. The corrosion mechanism is clarified at the atomic level, offering data support for the development of high-performance nuclear power materials.

nuclear power plant; high-temperature and high-pressure water; oxidation corrosion; molecular simulation