Performance of two distinct fuel systems, U-7Mo fuel in Zircaloy (Zry-4) cladding and U-10Mo fuel in aluminum (Al6061-O) cladding was studied. First, a mini plate with Zry-4 cladding from a previous irradiation experiment was evaluated via finite element analysis (FEA). By using the same plate geometry and irradiation conditions, another plate consisting of U-10Mo fuel and Al6061-O cladding was simulated. The results were then comparatively evaluated to explore the feasibility of employing Zry-4 as an alternative cladding. Simulations indicated the Zircaloy cladding plate would operate roughly 50 °C hotter as compared with the Al alloy cladding plate. Larger deformations in the thickness direction for the plate with Zry cladding were noted. It was observed that the postfabrication stresses in the fuel would be relieved quickly in the reactor, regardless of cladding type. Although the fuel stresses would still develop at reactor shutdown, the fuel would be stress-free during the entire irradiation period for both cladding types. At shutdown, the plate with Zry cladding would have higher stresses due to higher operating temperatures. Similarly, the stresses after shutdown are higher in the foil core for the plates with Zry cladding. The Al cladding plate would have higher plastic strains as compared with the Zry cladding plate. The Zry cladding plate is significantly stiffer, causing higher stresses in the fuel zone and at the interface. Overall, employing Zry as an alternate cladding is not expected to produce a more favorable thermomechanical performance as compared to the performance of an Al alloy cladding plate.