Fuel cladding and encapsulation materials are fundamental to the structural integrity, thermal management, and neutronic performance of nuclear fuel assemblies. This study proposes the use of a hybrid composite material based on Zircaloy-4 reinforced with silicon carbide (SiC) for the encapsulation of UO₂ fuel pellets, aiming to enhance both the thermal and mechanical properties of the fuel system while maintaining favorable neutronic characteristics. The proposed Zircaloy-4/SiC composite is evaluated and compared with conventional metallic cladding materials, such as standard Zircaloy-4 and stainless steel, as well as ceramic encapsulants like stabilized zirconia (ZrO₂).The analysis considers key parameters including thermal conductivity, neutron absorption cross-section under normal and transient operating conditions. Simulations conducted using the SCALE code system assess the impact of the composite on reactivity, heat distribution, and fuel temperature profiles. Preliminary results indicate that the inclusion of SiC enhances the high-temperature performance of the cladding, while the Zircaloy-4 matrix preserves the low neutron absorption desirable for maintaining core reactivity. When compared to zirconia, the Zircaloy-4/SiC composite offers superior thermal conductivity and reduced swelling under irradiation, albeit with slightly higher neutron absorption. Nonetheless, the composite exhibits a balanced profile that combines the structural advantages of ceramics with the neutronic compatibility of metallic alloys. These findings support the viability of metal-matrix composite encapsulation as a promising pathway for accident-tolerant fuel (ATF) designs in advanced reactor systems, including SMRs and Generation IV concepts. Further experimental validation is recommended to confirm fabrication feasibility and in-reactor behavior.