Pre-Conceptual Development and Characterization of an Extruded Graphite Composite Fuel for the Treat Reactor

Abstract

To explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative Convert program is exploring options to replace the existing highly enriched uranium core with low enriched uranium (LEU) core. To construct a new LEU core, fabrication processes similar to those used for the original core must be identified and developed. Initially, graphite matrix fuel blocks were either uniaxially pressed or extruded following historic routes; however, the project expanded to explore methods to increase the graphite content of the fuel blocks and modern resins. Materials properties relevant to fuel performance including density and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed. LA-UR-14-27588