The stability of metastable t-ZrO₂ instead of its fraction in an oxide film is a key factor for the protectiveness of the oxide film and corrosion resistance of Zr alloys, and the stability of t-ZrO₂ is influenced by alloying elements such as Nb and Sn.

In a fusion environment, tungsten, a plasma-facing material in a reactor, is subject to the irradiation of high-energy neutrons, generating a large amount of displacement damage and transmutation products (such as rhenium, Re). We studied the evolution of defects under irradiation in W and W-Re systems using the density functional theory (DFT) and rate theory (RT) method. The results indicate that the evolution of irradiation defects is mainly affected by the irradiation dose, dose rate, and temperature. During irradiation, loops form first in W, followed by the generation of voids, which are due to the different migration energies of point defects. Higher dose rates result in a higher density and larger size of defects in tungsten. Higher temperatures cause a decrease in void density and an increase in size. The results obtained at 600 °C were in good agreement with the reported TEM data. In W-Re alloys, it is indicated that the formation of loops is delayed because Re suppresses the nucleation of loops. The dynamic introduction of Re in W stabilizes the growth of defects compared to W-Re alloys, suggesting that transmuting elements have less detrimental effects on irradiation than alloying. As defect densities and sizes were quantified under different irradiation conditions, the results provide data for the multi-scale simulation of the radiation damage and thermal/mechanical properties in plasma-facing materials under fusion conditions.

Accurate evaluation of the thermal conductivity of UO₂ with defects is very significant for optimizing fuel performance and enhancing the safety design of reactors. We employed a method that combines the Boltzmann transport equation with DFT+U to calculate the thermal conductivity of UO₂ containing fission products and irradiation-induced point defects. Our investigation reveals that the thermal conductivity of UO₂ is influenced by defect concentration, defect type, and temperature. Fission products and irradiation defects result in a decrease in thermal conductivity, but they have markedly different impacts on phonon scattering mechanisms. Metal cations tend to scatter low-frequency phonons (less than 5.8 THz), while the fission gas xenon scatters both low-frequency and high-frequency phonons (greater than 5.8 THz), depending on its occupancy at lattice sites. Uranium vacancies scatter low-frequency phonons, while oxygen vacancies scatter high-frequency phonons. When uranium and oxygen vacancies coexist, they scatter phonons across the entire frequency spectrum, which further results in a significant reduction in the thermal conductivity of UO₂. Our calculated results align well with experimental data across a wide temperature range and provide fundamental insights into the heat transfer mechanisms in irradiated UO₂. These findings are essential for establishing a thermal conductivity database for UO₂ under various irradiation conditions and benefit the development of advanced high-performance UO₂ fuel.

Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO₂/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO₂ core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.

The microstructure evolutions of cascade annealing in α-Zr crystals at $T=300\mathrm{K}$, 400 K, and 500 K for energy of 2, 6 and 10 keV with initially driving directions $⟨0001⟩$ and $⟨01\overline{1}0⟩$ of the primary knocked atoms are investigated by using molecular dynamics simulations. The results show that the relaxation time (τ_r) of cascades can play the role of an effective parameter for describing radiation damages during molecular dynamics simulations. τ_r of cascade with the larger surface area but with the same volume is much smaller, which promotes faster recombination of defects during cascade annealing. Energy is a crucial factor in the formation of cascade displacements of atoms and appearing of defects compared with temperature and direction of movement for initially knocking atoms.

The effects of Nb concentration and temperature on the generalized stacking fault energy (GSFE) of basal, prismatic I, pyramidal I and II plane for Zr-Nb alloys are investigated by molecular dynamics simulations (MD). The stable and unstable SFEs of different slip systems show no significant change with the increasing Nb concentration (0, 0.5, 1.0, 1.5, 2.0, and 2.5 at.%) in Zr-Nb alloys at 0 K. Basal, pyramidal I and II planes slip of Zr-Nb alloys prefer to deform by full dislocation with the temperature increases. Additionally, plastic deformation anisotropy of Zr-Nb alloy is improved with the increasing temperature using both embedded atom method (EAM) and angular-dependent potentials (ADP). The present work provides a theoretical basis for understanding enhanced plasticity of Zr-Nb alloys under finite temperature.