Characterization of Zirconium Carbide Pellets for Nuclear Propulsion

For decades humans have had visions of traveling to Mars. With a traditional chemical rocket, the trip to Mars would take nine months, however a spacecraft powered by nuclear thermal propulsion using uranium carbide-based fuels could reach Mars in two months due to its powerful and efficient thrusts.

Design and testing of nuclear engines will require unique fuel configurations. Advanced manufacturing techniques are the most viable options for these applications in order to create the complex geometries necessary for efficient use, in addition to the fact that they produce minimal waste, compared to traditional manufacturing methods. After manufacturing, parts undergo sintering to produce parts that are > 90% dense.

For this project, zirconium carbide (ZrC) has been selected as a surrogate of uranium carbide because of the similar thermal properties as well as reduce hazards. Using ZrC powder that is pressed into pellets, we have been exploring further sintering temperatures and times to produce a (> 90%) theoretical density and analyze the results with XRD, SEM, and element analysis. This study focused on varying sintering times and temperature to optimize final sample density. With this information gathered, it will further aid advanced manufacturing methods of fuel from for space exploration.