|About this Abstract
||2022 TMS Annual Meeting & Exhibition
||Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
||Fabrication and Thermophysical Properties of (U,Zr)C; A High Uranium Density Fuel Candidate for Nuclear Thermal Propulsion Reactors
||Erofili Kardoulaki, Brian Taylor, Jhonathan Rosales, Tim Coons, Darrin Byler, Ken McClellan
|On-Site Speaker (Planned)
Advanced fuels with high melting points (>3000 K), high thermal conductivity, and ability to sustain exposure to corrosive environments are of interest for advanced nuclear reactors for terrestrial and space applications. These include advanced microreactor concepts for powering remote communities as well as space nuclear propulsion reactors to support manned Mars missions. One of the primary fuel candidates for these reactors is UN, however, adverse reaction of UN with hot hydrogen can prove detrimental. Another advanced fuel of interest is a mixed carbide such as (U,Zr)C, which can provide the same benefits as UN but has an even higher melting point and could resist hot hydrogen corrosion. In this work we have synthesized high purity (U,Zr)C via a carbothermic reduction route and have fabricated high density pellets via spark plasma sintering. Characterization of the pellets and measured thermophysical properties are presented here, confirming significant benefits of this fuel.
||Nuclear Materials, Ceramics, Characterization