Jiantao Qin

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.

Zirconium alloys are subjected to a fast neutron flux in nuclear reactors, inducing the creation of a large number of point defects, both vacancy and self-interstitial. These point defects then diffuse and can be trapped by their different sinks or can cluster to form larger defects, such as vacancy and interstitial clusters. In this work, the trapping capability of small-vacancy clusters (two/three vacancies, V₂/V₃) in the α-Zr doped with alloying elements (Sn, Fe, Cr, and Nb) has been investigated by first-principle calculations. Calculation results show that for the supercells of α-Zr containing 142-zirconium atoms with the two-vacancy cluster, alloying elements of Sn and Nb in the second vacant site (V2) and Cr in the first vacant site (V1) are more easily trapped by two vacancies, respectively. However, the two sites are both captured more easily by two vacancies for Fe in the supercells of α-Zr containing 142-zirconium atoms inside due to the similar value of the Fermi level. For the supercells of α-Zr containing 141-zirconium atoms with the three-vacancy cluster, the alloying element of Sn in the third vacant site (V’3), Fe in the first vacant site (V’1), and Cr and Nb in the second vacant site (V’2) are more easily trapped by three vacancies, respectively.

Ferritic/martensitic (F/M) steels whose matrix is Fe-Cr are important candidate materials for fuel cladding of fast reactors, and they have excellent irradiation-swelling resistance. However, the mechanism of irradiation-swelling of F/M steels is still unclear. We use a first-principles method to reveal the influence of irradiation defects, i.e., Frenkel pair including atomic vacancy and self-interstitial atom, on the change of lattice volume of Fe-13Cr lattice. It is found that vacancy causes lattice contraction, while a self-interstitial atom causes lattice expansion. The overall effect of a Frenkel pair on the change of lattice volume is lattice expansion, leading to swelling of the alloy. Furthermore, the diffusion properties of point defects in Fe-13Cr are investigated. Based on the diffusion barriers of the vacancies and interstitial atoms, we find that the defects in Fe-13Cr drain out to surfaces/grain boundaries more efficiently than those in pure α-Fe do. Therefore, the faster diffusion of defects in Fe-13Cr is one of important factors for good swelling resistance of Fe-13Cr compared to pure α-Fe.

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.