The development of fast reactor systems capable of burning recycled transuranic (TRU) isotopes has been underway for decades at various levels of activity. These systems could significantly alleviate nuclear waste storage liabilities by consuming the long-lived isotopes of plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm). The fabrication of metal fuel alloys by melt casting pins containing the volatile elements Am and Np has been a major challenge due to their low vapor pressures; initial trials demonstrated significant losses during the casting process. A low temperature hot extrusion process was explored as a potential method to fabricate uranium-zirconium fuel alloys containing the TRU isotopes. The advantage of extrusion is that metal powders may be mixed and enclosed in process canisters to produce the desired composition and contain volatile components. Uranium powder was produced for the extrusion process by utilizing a hydride-dehydride process that was developed in conjunction with uranium alloy sintering studies. The extrusions occurred at 600 degrees C and utilized a hydraulic press capable of 450,000 N (50 tons) of force. Magnesium (Mg) metal was used as a surrogate metal for Pu and Am because of its low melting point (648 degrees C) and relatively high vapor pressure (0.2 atm at 725 degrees C). Samples containing U, Zr, and Mg powder were prepared in an inert atmosphere glovebox using copper canisters and extruded at 600 degrees C. The successful products of the extrusion method were characterized using thermal analysis with a differential scanning calorimeter as well as image and x-ray analysis utilizing an electron microprobe. The analysis showed that upon fabrication the matrix of the extruded metal alloy is completely heterogeneous with no mixing of the metal particle constituents. Further heat treating upon this alloy allows these different materials to interdiffuse and form mixed uraniumz-irconium phases with varying types of microstructures. Image and x-ray analysis showed that the magnesium surrogate present in a sample was retained with little evidence of losses due to vaporization.

There is great interest in understanding the contamination of stainless steel by radionuclides across the nuclear fuel cycle. Through study of uptake mechanisms, contaminant localisation and process conditions that affect uptake, decontamination strategies can be tailored to remove built-up radioactive species. This study focusses on understanding stainless steel contamination by inactive lanthanides and radioactive actinide species (U, Np, Pu and Am) in acidic media. Through depth profiling, contamination has been seen to localise on the interface of the oxide layer and the bulk metal (at depths of up to 100 nm) indicating a potentially tenacious contamination mechanism. Furthermore, contaminant was observed at greater depths within the material (up to ~300 nm), suggesting penetration beyond the passive layer in to the bulk of the material. Long term immersion studies (up to 9 months) in 12 M HNO3 have also been undertaken to investigate the effect of surface corrosion on contaminant uptake. After 3 months the surface had undergone intergranular corrosion and grain droppage was observed. Further surface analysis revealed localisation of contaminants within the steel grain boundaries and vacancies. Once again, this may necessitate a more aggressive decontamination strategy. Conditions have been identified that enhance uptake of transuranic contaminants. Highest levels of uptake were seen in polished steel samples immersed in 4 M HNO3. The Np-237, Pu-239 and Am-241 contaminated samples showed surface concentrations of up to 1.2x107, 9.4x105 and 1x109 Bq/m2, respectively. In the case of Np contamination of stainless steel, microfocus X-ray absorption spectroscopy has shown the surface-mediated reduction of Np(V) leading to Np(IV) adsorption.

The nuclear waste currently generated in the United States is stored in spent fuel pools and dry casks throughout the country awaiting a permanent disposal solution. One efficient solution would be to remove the actinides from the waste and transmute these isotopes in a fast spectrum reactor. Currently this technology is unavailable on a commercial scale and a considerable amount of research and development is still required. An alternate solution is to reprocess and recycle the used fuel in thermal reactors, creating new fuel while reducing the amount of waste and its impact to the environment. This thesis examines the possibility of multi-recycling the transuranics (Pu, Np, Am, and Cm) in a standard pressurized water reactor (PWR). Two types of recycling strategies will be examined: one where Pu, Np, and Am are recycled (TRU-Cm) and a second where the previous isotopes as well as Cm are recycled (TRU+Cm). To offset the hardened neutron spectrum that results from the inclusion of the transuranics, a smaller fuel pin is employed to provide additional moderation. Computer simulations are used to model the in-reactor physics and long-term isotopic decay. Each fuel type is assessed based on the required U-235 enrichment, void coefficient, transuranic production/destruction, and radiotoxicity reduction as compared to a UOX and MOX assembly. It is found that the most beneficial recycling strategy is the one where all of the transuranics are recycled. The inclusion of Cm reduces the required U-235 enrichment, compared to the other multi-recycled fuel and, after a significant number of recycles, can result in the required enrichment to decrease. This fuel type also maintains a negative void coefficient for each recycle. The void coefficient of the fuel type without Cm becomes positive after the third cycle. The transmutation destruction of the two multi-recycled assemblies is less than that of a MOX assembly, but the transmutation efficiency of the multi-recycled assemblies exceeds the MOX assemblies. The radiotoxicity of both multi-recycled assemblies is significantly lower than the UOX and MOX with the TRU+Cm fuel being the lowest. When Curium is recycled only 28,000 years are required for the radiotoxicity of the waste to reach that of natural Uranium and when Cm is not recycled, the amount of time increases to 57,000 years.

The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation. The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation.