Rongjian Pan

U3Si2 has attracted considerable interest as a nuclear fuel because of its high uranium density and favorable thermophysical properties. In U3Si2/Al dispersion fuels, however, the behavior of fission gases, particularly xenon (Xe), remains insufficiently understood, especially regarding the role of Al in mediating Xe stability and migration. In this work, first-principles calculations are combined with experimental observations to clarify how Al affects Xe behavior in U3Si2. The results show that Xe has extremely low solubility in pristine U3Si2 and exhibits an unfavorable tendency for interstitial clustering, which can be attributed to its large atomic size and the strong covalent Si-Si network in the host lattice. Al plays a dual role in this process. On the one hand, it facilitates local Xe mobility by reducing the effective size of Xe through charge transfer. On the other hand, Al preferentially occupies vacancy-type defects, thereby reducing the available trapping sites for Xe accumulation. Although the presence of Al lowers Xe migration barriers locally, the overall Xe behavior in U3Si2 remains strongly constrained by the crystal structure and defect environment. These findings provide atomistic insight into the role of Al in governing Xe behavior in U3Si2, and may contribute to the understanding and design of swelling-resistant U3Si2-based fuels.

U3Si2 is a promising accident-tolerant fuel, and its use in dispersion fuel form with an aluminum (Al) matrix is widely deployed in test and research reactors. However, in such U3Si2/Al dispersion fuels, managing fission gases, particularly xenon (Xe), remains a critical challenge for long-term stability, and the role of the Al matrix in mediating Xe behavior is still poorly understood. This study combines first-principles calculations and experimental validation to elucidate the atomic-scale interactions between Xe and Al in U3Si2. We reveal that Xe exhibits intrinsically low solubility and negligible interstitial clustering in the pristine U3Si2, attributable to its large atomic size and the pervasive covalent Si–Si network. Interestingly, Al plays a dual role. It enhances Xe diffusion by reducing its effective atomic size via charge transfer, while simultaneously suppressing bubble nucleation in the irradiated region by pre-occupying vacancy sites. Although Xe diffusion is enhanced within Al-rich regions, the overall migration through fuel particles remains limited, consistent with the swelling resistance observed in U3Si2. These findings provide a theoretical basis for predicting Xe behavior in U3Si2/Al dispersion fuel systems and reveal key mechanisms for designing swelling-resistant nuclear fuels.

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.

Accurately evaluate the thermal conductivity of UO2 with defects is very significant for the superior safety design of reactors and optimize fuel performance. The DFT+U combining with the phonon Boltzmann transport equation to obtain applied to calculate electronic structure, phonon spectrum and the thermal conductivity of UO2 with fission products and irradiation point defects. The calculated thermal conductivity of UO2 has a good agreement with the experimental data over a wide temperature range. Our investigation indicates that the thermal conductivity of UO2 is a function of defect concentration, defect type, and temperature. Fission products and irradiation defects lead to a decrease of thermal conductivity, but show distinctions on scattering mechanisms for phonons. Metal cations scatter low-frequency phonons, while fission gas xenon not only scatter low-frequency phonons but also high-frequency phonons depending on the lattice site it occupied. U vacancies scatter low-frequency phonons, O vacancies scatter high-frequency phonons, while U and O vacancies jointly scatter full-frequency phonons which will further reducing thermal conductivity of UO2. Our findings provides a fundamental insight to heat transfer in irradiated UO2 which is crucial for the development of high-performance UO2 fuel.

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.

Co-adsorption of multi-components in ZrCo-based hydrogen storage materials can lead to a number of synergistic effects, such as the modification of adsorption sites, and further worsen the hydrogen storage capability. In this work, we explore the co-adsorption of H and CO on the ZrCo(110) surface and find that the molecular CO can be adsorbed on the clean alloy surface and thus decrease the hydrogen storage ability of the alloy. Moreover, CO occupies the adsorption site of H and therefore prevents the adsorption and diffusion into the interior of the lattice. Fortunately, the Hf dopant reduces the number of adsorption sites of the CO molecule and inhibits the formation of carbides to a certain extent. In addition, the partial density of states (PDOS) result shows that there is almost no interaction between the s orbital of H and the s orbital of Co on the pure surface of pre-adsorbed CO, while on the Hf-doped surface of pre-adsorbed CO, the s orbital of H overlapped greatly with the s orbital of Co, indicating that Hf doping inhibits CO toxicity in the interaction between H and the surface. Hence, the doping of Hf has the effect of giving resistance to CO toxicity and is conducive to the adsorption of H.

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.