Manufacture of a UO2-Based Nuclear Fuel with Improved Thermal Conductivity with the Addition of BeO

• DOI: 10.1007/s40553-017-0108-2
• Publication Date: 2017-04-10T18:33:43Z
• Authors: Chad B. Garcia; Ryan A. Brito; Luis H. Ortega; James P. Malone; Sean M. McDeavitt
• Keywords: Beryllium oxide; Nuclear Fuel

The low thermal conductivity of oxide nuclear fuels is a performance-limiting parameter. Enhancing this property may provide a contribution toward establishing accident-tolerant fuel forms. In this study, the thermal conductivity of UO2 was increased through the fabrication of ceramic-ceramic composite forms with UO2 containing a continuous BeO matrix. Fuel with a higher thermal conductivity will have reduced thermal gradients and lower centerline temperatures in the fuel pin. Lower operational temperatures will reduce fission gas release and reduce fuel restructuring. Additions of BeO were made to UO2 fuel pellets in 2.5, 5, 7.5, and 10 vol pct concentrations with the goals of establishing reliable lab-scale processing procedures, minimizing porosity, and maximizing thermal conductivity. The microstructure was characterized with electron probe microanalysis, and the thermal properties were assessed by light flash analysis and differential scanning calorimetry. Reliable, high-density samples were prepared using compaction pressure between 200 and 225 MPa and sintering times between 4 and 6 hours. It was found that the thermal conductivity of UO2 improved approximately 10 pct for each 1 vol pct BeO added over the measured temperature range 298.15 K to 523.15 K (25 °C to 250 °C) with the maximum observed improvement being \(\sim \)100 pct, or doubled, at 10 vol pct BeO.