Energy Sustainability

To address the control and production of tritium in the fusion fuel cycle, this thesis focuses on: (1) the impact of defects on tritium permeation barriers, and (2) disposal methods for waste streams resulting from tritium production processes. Tritium permeation in a fusion reactor can be reduced by decreasing solubility within the permeation barrier. To gain insights into the impacts of defects such as vacancies and amorphous structures on permeation, solubility calculations were carried out using first-principle methods. Solubility calculations find that defect sites such as silicon vacancies and non-stoichiometric carbon-rich amorphous regions are the greatest contributors to solubility. Several tritium production processes result in waste streams that may benefit from reprocessing of used nuclear fuel using pyroprocessing technologies. For managing pyroprocessing salt waste streams, a dehalogenation process followed by additions of iron oxide and 5 wt.% sodium borosilicate glass is proposed. This approach enhances the chemical durability of iron-phosphate waste forms, making them more suitable for immobilizing radionuclides from pyroprocessing waste streams.