氮化铀热中子截面的第一性原理计算

Nuclear design and neutronic analysis of thermal neutron reactor need high reliable thermal neutron cross sections. Uranium mononitride (UN) is a candidate fuel material for advanced power reactor with its better thermodynamics and accident tolerance. However, in thermal neutron region, reliable thermal neutron scattering cross sections are lacked for UN, which is disadvantageous to reactor physics simulations. The scattering law of the UN fuel material may impact the thermal neutron spectrum and criticality of the reactor systems. Neutron cross sections in thermal range are correlated with energy, temperature, physical and chemical properties of the scattering medium, reflecting the phonon spectra of material itself. In this paper, based on the ab initio method of quantum mechanics, phonon density of states in UN are calculated by VASP/PHONON code, and used for integral to obtain UN heat capacity at a constant volume. Adopting this new phonon density of states, NJOY/LEAPR code is used to generate S(α, β) data by thermal neutron scattering theory and NJOY/THERMR utilizes these data to produce thermal scattering matrix in order to investigate thermal kernel effect of UN. Subsequently, thermal neutron scattering cross sections of UN are generated with NJOY code system. Comparison with uranium dioxide (UO ₂ ) in the traditional PWR is done. Results indicate that optimized lattice parameter are in good agreement with the database; the optical modes are well separated from the acoustic modes compared with UO ₂ ; heat capacity at a constant volume is consistent with experimental value; the inelastic and elastic cross sections of ²³⁸ U in UN are lower than those of ²³⁸ U in UO ₂ . N in UN only deals with incoherent part in elastic cross sections. As the temperature increases, elastic cross sections of UN decrease while inelastic ones increase, and cross sections approach to free atom cross section at high energies. Considering the limitations of ¹⁴ N, the scattering law and inelastic scattering cross sections are also under investigation using ¹⁵ N in UN compound. This paper's conclusion fulfill the vacancy of thermal neutron scattering cross sections of UN, which laid a foundation for systematic study on the neutronics properties of UN fuel in the light water reactors as well as for the design of new neutron moderators and neutron filer.