Nuclear fuel cladding made of zirconium alloys constitute the first safety barrier in pressurized water reactors. The microstructure of these alloys not only controls mechanical properties, but also phenomenon such as corrosion or growth under irradiation. Enabling a more flexible use of nuclear energy in the mix while maintaining the structural integrity of fuel cladding under both operating and accidental conditions, we must understand the detailed mechanisms of microstructure evolution under irradiation. Numerous studies point toward the center part played by Nb in such microstructural evolution. For instance, diffusion flux coupling between solutes (Nb) and point defect created by irradiation gives rise to local Nb segregation, as well as precipitates which are not seen in non-irradiated samples. Atomic scale modeling brings in information that complements that obtained from experimental observations, allowing to confirm or disprove the evolution scenarios found in the literature. The aim of this Ph.D. work is to use the tools which have been developed to study irradiation effects in ferritic steels, and apply them to Zr alloys, with a focus on radiation induced segregation. Electronic structure calculations in the density functional theory approximation will be used to study the interactions between niobium atoms and point defects. From this data, we are able to compute transport coefficients, from which we can discuss quantitatively solute/point defect flux coupling and radiation induced segregation effects.