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Properties of Ceramic and Cermet Fuels for Sodium Graphite Reactors

Technical Report ·
DOI:https://doi.org/10.2172/4175587· OSTI ID:4175587
 [1];  [1]
  1. North American Aviation, Inc., Canoga Park, CA (United States). Atomics International Div.
Nuclear materials of interest as potential reactor fuels for sodium graphite reactors are reviewed to select those which appear most feasible for high-temperature, long-burnup application. Fuel properties such as melting point, thermal neutron absorption cross section, uranium content, chemical and physical properties, and fabrication details are presented. Other factors such as expected burnup capabilities, required enrichments, and conversion ratios were compared. A program is currently in progress to evaluate promising nuclear fuels. Fuel materials under consideration include the uranium compounds: uranium dioxide, uranium monocarbide, borides, sulfides, aluminide, nitrides, silicides, and phosphides. Various cermets are under consideration. These include dispersions of uranium compounds in matrices of uranium, uranium alloys, thorium, and thorium alloys. Included among the uranium alloy matrices are binary and ternary combinations of uranium with niobium, molybdenum, and zirconium.
Research Organization:
North American Aviation, Inc., Canoga Park, CA (United States). Atomics International Div.
Sponsoring Organization:
US Atomic Energy Commission (AEC)
DOE Contract Number:
AT(11-1)-GEN-8
NSA Number:
NSA-14-018165
OSTI ID:
4175587
Report Number(s):
NAA-SR--3880
Country of Publication:
United States
Language:
English