CARBIDE FUEL DEVELOPMENT. Phase II Report, Period of September 15, 1959 to September 15, 1960

Technical Report · Tue Nov 01 04:00:00 EST 1960

OSTI ID:4072676

Bolta, C; Strasser, A; Doherty, P; Saulino, F; Taylor, K

Uranium monocarbide powder having a good purity (average UC content of about 97%) was obtained by carbon reduction of UO/sub 2/. The chief impurities were UO/sub 2/ and UC/sub 2/. In addition, traces of nitrogen (less than 0.1%) were usually present. A preliminary study of the production of UC by the reaction of carbon with ammonium diuranate indicated that a good product can probably be obtained by this method. Pellets of UC with natural uranium were fabricated by cold pressing at 16,000 psi and sintering at 1850 deg C for 1 hr. Average densities of 12.7 g/cm/sup 3/ (93% of theoretical) and maximum densities of 13.0 g/cm/sup 3/ (95% of theoretical) were achieved. Small panticle size, about 1 mu , was found to be of prime importance in achieving high densities. Varying cold forming pressure from 15,000 to 40,000 psi and sintering at temperatures higher than 1850 deg C did not improve sintered densities. Approximately 1/2% of a temporary binder, Carbowax 6000, gave the best results. The structure of the pellets consisted of UC and a small amount of UC2. Pellets of UC with 24% enriched uranium were fabricated for irradiation tests by the same methods. Average densities of 12.8 g/cm/sup 3/ (94% of theoretical) and maximum densities of 13.1 g/cm/sup 3/ (96% of theoretical) were achieved. The structure of the pellets consisted of UC and indications of free uranium metal. The deliberate addition of uranium metal in amounts up to 10% of the UC powder did not greatly increase the final density of the pellets. Hot pressing of UC powder did not resuit in better densities than had been obtained by cold pressing and sintering. Simultaneous synthesis and fabrication of UC was tried by hot pressing uranium metal powder and carbon. Densities of 13.0 to 13.3 g/cm/sup 3/ were obtained. The pellets contained a major UC phase and a nonuniform distribution of free uranium metal. UC was held in contact with various cladding materials at 820 deg C up to 4000 hr, to study possible fuel-cladding interactions. There was no reaction or barely detectable reaction with niobium and 2 1/2 Cr-1 Mo steel. A slight reaction occurred with type 304 SS, and a considerable reaction occurred with Inconel-X, Zircaloy-2, and beryilium. The type 304, 2 1/4 Cr-1 Mo steels and niobium appeared to be satisfactory cladding materials for UC under the conditions of the test. The dissolution study of a few UC pellets containing simulated fission products (fisside) was completed. High uranium recovery was achieved. Eight irradiation capsules are planned. Variables to be investigated are burnup (2, 4, and 8% of all fuel atoms), maaimum fuel temperature (980 and 1370 deg C), power (14 and 20 kw/ft), effect of cladding coefficient of expansion on maximum fuel temperature (niobium and stainless steel). A measure of effective in-pile thermal conductivity and gas release will be attempted. The Carborundum Company plutonium facility and the NDA plutonium facility were completed. (auth)

• Research Organization: Nuclear Development Corp. of America, White Plains, N.Y.; Carborundum Co., Niagara Falls, N.Y.
• NSA Number: NSA-15-012469
• OSTI ID: 4072676
• Report Number(s): NDA-2145-6
• Country of Publication: United States
• Language: English

Related Subjects

AMMONIUM COMPOUNDS
ANISOTROPY
BONDING
CARBON
CARBOWAX
COLD WORKING
DENSITY
ENRICHMENT
FABRICATION
GRAIN SIZE
HEAT TREATMENTS
IMPURITIES
IRRADIATION
MATERIALS TESTING
METALS
MICROSTRUCTURE
NITROGEN
PELLETS
POWDERS
PRESSURE
PRODUCTION
REACTOR TECHNOLOGY
REDUCTION
SINTERING
TEMPERATURE
URANATES
URANIUM
URANIUM CARBIDES
URANIUM DIOXIDE