Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

Conference · Sun Mar 01 04:00:00 EST 2015

DOI:https://doi.org/10.1002/9781119093466.ch144· OSTI ID:1186757

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

In an effort 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 (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

• Research Organization: Idaho National Laboratory (INL)
• Sponsoring Organization: DOE - NE
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1186757
• Report Number(s): INL/CON-14-32449
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
LEU Fuel
Graphite Composite
TREAT