Silicon carbide and nanostructured ferritic alloy (SiC-NFA) composites have the potential to maintain the outstanding high temperature corrosion and irradiation resistance and enhance the mechanical integrity for nuclear cladding. However, the formation of detrimental silicide phases due to reaction between SiC and NFA remains a major challenge. By introducing a carbon interfacial barrier on NFA (C@NFA), SiC-C@NFA composites are investigated to reduce the reaction between SiC and NFA. In a similar way, the effect of chromium carbide (Cr3C2) interfacial barrier on SiC (Cr3C2@SiC) is also presented for Cr3C2@SiC-NFA composites. Both the coatings were successful in suppressing silicide formation. However, despite the presence of coatings, SiC was fully consumed during spark plasma sintering process. TEM and EBSD investigations revealed that spark plasma sintered SiC-C@NFA and Cr3C2@SiC-NFA formed varying amounts of different carbides such as (Fe,Cr)7C3, (Ti,W)C and graphite phases in their microstructure. Detailed microstructural examinations after long term thermal treatment at 1000oC on the microstructure of Cr3C2@SiC-NFA showed precipitation of new (Fe,Cr)7C3, (Ti,W)C carbides and also the growth of existing and new carbides. The results were successfully explained using ThermoCalc precipitation and coarsening simulations respectively. The oxidation resistance of 5, 15 and 25 vol% SiC@NFA and Cr3C2@SiC-NFA composites at 500-1000oC temperature under air+45%water vapor containing atmosphere is investigated. Oxidation temperature effects on surface morphologies, scale characteristics, and cross-sectional microstructures were investigated and analyzed using XRD and SEM. SiC-C@NFA showed reduced weight gain but also showed considerable internal oxidation. Cr3C2@SiC-NFA composites showed a reduction in weight gain with the increasing volume fraction of Cr3C2@SiC (5, 15 and 25) without any indication of internal oxidation in the microstructure. 25 vol% SiC-C@NFA and 25 vol% Cr3C2@SiC-NFA showed over 90% and 97% increase in oxidation resistance (in terms of weight gain) as compared to NFA. The results were explained using the fundamental understanding of the oxidation process and ThermoCalc/DICTRA simulations. Finally, the irradiation performance of SiC-C@NFA and Cr3C2@SiC-NFA composites was assessed in comparison with NFA using state-of-the-art TEM equipped with in-situ ion irradiation capability. Kr++ ions with 1 MeV energy was used for irradiation experiments. The effect of ion irradiation was recorded after particular dose levels (0-10 dpa) at 300oC and 450oC temperatures. NFA sample showed heavy dislocation damage at both 300oC and 450oC increasing gradually with dose levels (0-10 dpa). Cr3C2@SiC-NFA showed similar behavior as NFA at 300oC. However, at 450oC, Cr3C2@SiC-NFA showed remarkably low dislocation loop density and loop size as compared to NFA. At 300oC, microstructures of NFA and Cr3C2@SiC-NFA show predominantly 1/2<111> type dislocation loops. At 450oC, NFA showed predominantly <100> type loops, however, Cr3C2@SiC-NFA composite was still predominant in ½<111> loops. The possible reasons for this interesting behavior were discussed based on the large surface sink effects and enhanced interstitial-vacancy recombination at higher temperatures. The molecular dynamics simulations did not show considerable difference in formation energies of ½<111> and <100> loops for NFA and Cr3C2@SiC-NFA composites. The additional Si element in the SiC-NFA sample could have been an important factor in determining the dominant loop types. SiC-C@NFA composites showed heavy dislocation damage during irradiation at 300oC. At 450oC, SiC-C@NFA showed high dislocation damage in thicker regions. Thinner regions near the edge of TEM samples were largely free from dislocation loops. The precipitation and growth of new (Ti,W)C carbides were observed at 450oC with increasing irradiation dose. (Fe,Cr)7C3 precipitates were largely free from any dislocation damage. Some Kr bubbles were observed inside (Fe,Cr)7C3 precipitates and at the interface between α-ferrite matrix and carbides ((Fe,Cr)7C3, (Ti,W)C). The results were discussed using the fundamental understanding of irradiation and ThermoCalc simulations. / Doctor of Philosophy / With the United Nations describing climate change as 'the most systematic threat to humankind', there is a serious need to control the world's carbon emissions. The ever increasing global energy needs can be fulfilled by the development of clean energy technologies. Nuclear power is an attractive option as it can produce low cost electricity on a large scale with greenhouse gas emissions per kilowatt-hour equivalent to wind, hydropower and solar. The problem with nuclear power is its vulnerability to potentially disastrous accidents. Traditionally, fuel claddings, rods which encase nuclear fuel (e.g. UO2), are made using zirconium based alloys. Under 'loss of coolant accident (LOCA) scenarios' zirconium reacts with high temperature steam to produce large amounts of hydrogen which can explode. The risks associated with accidents can be greatly reduced by the development of new accident tolerant materials. Nanostructured ferritic alloys (NFA) and silicon carbide (SiC) are long considered are leading candidates for replacing zirconium alloys for fuel cladding applications. In this dissertation, a novel composite of SiC and NFA was fabricated using spark plasma sintering (SPS) technology. Chromium carbide (Cr3C2) and carbon (C) coatings were employed on SiC and NFA powder particles respectively to act as reaction barrier between SiC and NFA. Microstructural evolution after spark plasma sintering was studied using advanced characterization tools such as scanning electron microscopy (SEM), electron backscattered diffraction (EBSD), transmission electron microscopy (TEM) and energy dispersive spectroscopy (EDS) techniques. The results revealed that the Cr3C2 and C coatings successfully suppressed the formation of detrimental reaction products such as iron silicide. However, some reaction products such as (Fe,Cr)7C3 and (Ti,W)C carbides and graphite retained in the microstructure. This novel composite material was subjected to high temperature oxidation under a water vapor environment to study its performance under the simulated reactor environment. The degradation of the material due to high temperature irradiation was studied using state-of-the-art TEM equipped with in-situ ion irradiation capabilities. The results revealed excellent oxidation and irradiation resistance in SiC-NFA composites as compared to NFA. The results were discussed based on fundamental theories and thermodynamic simulations using ThermoCalc software. The findings of this dissertation imply a great potential for SiC-NFA based composites for future reactor material designs.