Abstract In order to analytically investigate irradiation behavior of metallic fast reactor fuels, the authors have developed the ALFUS (ALoyed Fuel Unified Simulator) code. The ALFUS can mechanistically simulate gas release and deformation behavior of the uranium-zirconium alloy fuel. The stress-strain analysis model into which anisotropic strain due to cavitation at the grain and/or phase boundary in the a-uranium phase has been introduced simulates anisotropic deformation of the uranium-zirconium alloy fuel. The models included in the ALFUS are thought reasonable and consistent with knowledge obtained from the irradiation test results. When the fuel slug swells out and comes into contact with the cladding, compressive stress is produced in the slug and decreases volume of the open pore if it has been formed. This is an essential process to keep fuel-cladding mechanical interaction (FCMI) small in case of lower smear density fuel. Analyses with the ALFUS indicate a significant level of FCMI in case of higher than 80% smear density fuel.