The time-dependent radius of the central void and the extents of the columnar grain, the equiaxed grain, and the unaffected grain regions of a typical oxide cylindrical fuel rod in a fast reactor at constant power level are determined. The temperature distributions in the fuel element are obtained. A model postulated to analyze and calculate the irradiation swelling and fission-gas release for oxide fuels of fast reactors is developed. The mechanical analysis is based on the thermal and radiation dilatations and on an elastoplastic approach for the Prandtl-Reuss material. An iteration method of successive approximation is used to compute the stresses and strains developed in the fuel elements. The computed results are shown by curves for the unsteady-state fuel restructuring of the fuel element.