Abstract Title
Corrosion of Composite Uranium Nitride Fuels
Abstract
Uranium mononitride (UN) has been identified as a possible accident tolerant fuel in nuclear reactors based on its high uranium density, thermal conductivity and low fission-gas release. Pure UN samples have been shown in studies to react with water at the operating temperatures of light water reactors, which make up the majority of reactors in the United States. Composite UN-UO2 fuels might be optimized for corrosion resistance in these conditions. An autoclave was re-engineered for work with radioactive materials by creating safeguards to prevent radioactive material release. UN was prepared from elemental uranium using a hydride-dehydride-nitride thermal synthesis prior to mixing with up to 10 wt% UO2. UN-UO2 composites were tested by placing samples in the water-filled autoclave at 320°C and 9 MPa. Pellets were characterized for weight change, surface hydration, and grain boundary deterioration using a sensitive digital balance, optical microscopy and electron microscopy. Corrosion products were identified using energy dispersive X-ray spectroscopy and X-ray diffraction. The amount of leached uranium in solution was measured using inductive coupled plasma mass spectroscopy.
Corrosion of Composite Uranium Nitride Fuels
Uranium mononitride (UN) has been identified as a possible accident tolerant fuel in nuclear reactors based on its high uranium density, thermal conductivity and low fission-gas release. Pure UN samples have been shown in studies to react with water at the operating temperatures of light water reactors, which make up the majority of reactors in the United States. Composite UN-UO2 fuels might be optimized for corrosion resistance in these conditions. An autoclave was re-engineered for work with radioactive materials by creating safeguards to prevent radioactive material release. UN was prepared from elemental uranium using a hydride-dehydride-nitride thermal synthesis prior to mixing with up to 10 wt% UO2. UN-UO2 composites were tested by placing samples in the water-filled autoclave at 320°C and 9 MPa. Pellets were characterized for weight change, surface hydration, and grain boundary deterioration using a sensitive digital balance, optical microscopy and electron microscopy. Corrosion products were identified using energy dispersive X-ray spectroscopy and X-ray diffraction. The amount of leached uranium in solution was measured using inductive coupled plasma mass spectroscopy.