THE DEVELOPMENT OF URANIUM CARBIDE AS A NUCLEAR FUEL. Third Annual Report, September 1, 1961 to October 31, 1962

Technical Report · Tue Jan 15 04:00:00 EST 1963

DOI:https://doi.org/10.2172/4721743· OSTI ID:4721743

Crane, J; Kalish, H S; Litton, F B

= 9 6 < ? < 0 t and fabrication method on irradiation stability, thermal conductivity, and hot hardness of uranium carbide were determined. Hypostoichiometric and stoichiometric uranium carbides prepared by both powder metallurgy and skull casting and hyperstoichiometric cast carbide were tested. The preparation of 12% enriched uranium carbide specimens for irradiation testing was completed. Sintered specimens were 98% of theoretical density for hypostoichiometric uranium carbide and 92 to 93% of theoretical for stoichiometric uranium carbide. All cast specimens were above 98.7% of theoretical density. Five different specimens, 4.4 and 4.8 450 deg C in a 0.1 wt% carbon, cast material and sintered material, and 5.2 450 deg C in a 0.1 wt% carbon, cast uranium carbide, were canned separately in niobium-1 wt% Zr and inserted into each of three capsules. The fuel specimens in capsule UNC-1-2 were contained in a sodium bond. This capsule was removed after 15,000 MW-d/ton U burnup in the MTR. The specimens in capsule UNC-1-3 were canned using an interference fit between cladding and fuel. This capsule was removed from the MTR after 14,440 MW-d/ton U burnup. Fuel surface temperatures ran in the range of 650 to 870 un. Concent 85% C and center-line temperatures were calculated at about 1050 to 1100 un. Concent 85% C. Specimens were prepared for thermal expansion, thermal conductivity, and hot hardness measurements. These properties are being determined from room temperature to 1000 un. Concent 85% C for multiple specimens of each test condition. Thermal expansion of uranium carbide from room temperature to 1000 un. Concent 85% C was found to be 12.4 x 10/sup -6/ per un. Concent 85% C for 4.4 wt% carbon sintered material, and 11.5 x 10/sup - 6/ per un. Concent 85% C for 4.7 wt% carbon cast material, Additional data on thermal expansion are currently being obtained. Out-of-pile compatibility tests were run at 815 un. Concent 85% C for 260 hr for hypo and hyperstoichiometric uranium carbide with niobium-1 wt% Zr cladding and with sodium as the bonding material. The hynostoichiometric material was prepared by powder metallurgy; the hyperstoichiometric carbide was made by skull casting. These tests showed some surface damage of the uranium carbide at both levels of carbon content. This damage was in the form of oxidation of free uranium in the case of the low carbon material and appeared as small cracks and chipping in the case of hyperstoichiometric material. Evidence of carburization to a depth of 0.0001 in. was noticed at the surface of the cladding used with the hyperstoichiometric uranium carbide. The evidence was not conclusive and no decarburization of the fuel was observed. Exposure tests of cast uranium carbide specimens with as-cast as well as ground surfaces and at three levels of carbon content, 4.5, 4.7, and 5.3 wt% were performed. Satisfactory resistance to corrosion was observed for 4.5 and 4.7 wt% carbon material exposed to laboratory air for about 10 months, but the 5.3 wt% carbon specimens, both as-cast and ground surfaces, showed transverse cracking after five months. All specimens stored in vacuo showed some surface attack and slight pitting starting after five months storage but no cracking was observed. Carbon content of these specimens did not appear to have any effect on corrosion behavior. Additional information was gained on the reproducibility of the skull casting method during preparation of castings for property studies. This information indicated that stoichiometric UC castings containing less than 200 ppm oxygen and 100 ppm nitrogen could be consistently made to 4.8 450 deg C in a 0.1 wt% carbon using 4.6 wt% carbon feed material obtained by solid state reaction of UO/sub 2/ with graphite. A homogeneous skull of 4.8 wt% carbon was required. Similar reproducibility of casting carbon content was achieved at other carbon levels. Methods for minimizing certain casting defects were also found. The density of sintered uranium carbide was found to be strongly dependent upon carbon content. Carbon powder (4.3 to 4.4 wt%) was

• Research Organization: United Nuclear Corp. Development Div., White Plains, N.Y.
• Sponsoring Organization: USDOE
• NSA Number: NSA-17-023873
• OSTI ID: 4721743
• Report Number(s): UNC-5048
• Country of Publication: United States
• Language: English

Related Subjects

AIR
BONDING
BURNUP
CARBON
CASTING
CORROSION
CRACKS
DEFECTS
DENSITY
EXPANSION
FABRICATION
FUEL CANS
FUELS
GRAPHITE
HARDNESS
HIGH TEMPERATURE
IRRADIATION
METALS, CERAMICS, AND OTHER MATERIALS
MTR
NIOBIUM ALLOYS
OXIDATION
POWDER METALLURGY
QUANTITY RATIO
REACTORS
REPORT
SINTERED MATERIALS
SINTERING
STABILITY
SURFACES
THERMAL CONDUCTIVITY
URANIUM CARBIDES
URANIUM DIOXIDE
VACUUM
VOLATILITY
ZIRCONIUM ALLOYS