CARBIDE FUEL DEVELOPMENT. Phase III Report, September 15, 1960-September 15, 1961

Technical Report · Sat Sep 30 04:00:00 EDT 1961

OSTI ID:4803712

Sheridan, W; Strasser, A; Anderson, J; Taylor, K

UC--PuC solid solution powders containing only minor amounts of impurities were obtained by reacting a mixture of UO/sub 2/ and PuO/sub 2/ with carbon or by heating together a mixture of UC, Pu/sub 2/C/sub 3/, and U. UC-PuC pellets with the desired high densities of 95% of theoretical were obtained by cold pressing and sintering, using 0.1 wt% nickel as a sintering aid. Without nickel, the densities were much lower for the same sintering conditions. Prior to experiments with solid solution carbides, essentially stoichiometric UC powder was prepared in the plutonium facility by the carbon reduction of UO/sub 2/ in a helium atmosphere. Attempts to prepare stoichiometric PuC by the carbon reduction of PuO/sub 2/ failed, the product consisting of a mixture of plutonium carbides and oxides. Pu/sub 2/C/sub 3/ was readily obtained by reacting stoichiometric amounts of PuO/sub 2/ and carbon. Additional evaluation of test samples, using x-ray diffraction and a microprobe analyzer, confirmed that type 304 stainless steel, 2 1/4 Cr-1 Mo steel, and niobium are satisfactory UC cladding materials at 820 f C, for 4000 hr, while Inconel-X, Zircaloy-2, and beryllium are not satisfactony. The irradiation of two clad UC specimens was completed. The UC pellets were made by cold pressing and sintering UC powder, which in turn was made by reacting UO/sub 2/ with carbon. Irradiation conditions were 18,500 Mwd/tonne average burnup, 760 f C average central fuel temperature, and 315 w/cm average heat generation rate. In-pile measurements of the combined fuel plus helium gap conductivity gave an average value of 0.02 g-cal/sec-cm/sup 2/- f C/cm. The UC conductivity was similar to out-of-pile values for arc cast UC, if reasonable helium gap conductances were assumed. A glove box for high temperature property measurements was designed and is being assembled. A conceptual design of the facility for examining irradiated plutonium was made. (auth)

• Research Organization: United Nuclear Corp. Development Div., White Plains, N.Y.; Carborundum, Co., Niagara Falls, N.Y.
• NSA Number: NSA-16-006686
• OSTI ID: 4803712
• Report Number(s): NDA-2165-5
• Country of Publication: United States
• Language: English

Related Subjects

ALUMINUM ALLOYS
BERYLLIUM
CARBIDES
CARBON
CHEMICAL REACTIONS
CHROMIUM ALLOYS
COBALT ALLOYS
COLD WORKING
COMPACTING
DENSITY
DIFFRACTION
FABRICATION
FUEL CANS
FUELS
GLOVEBOXES
HEATING
HELIUM
HIGH TEMPERATURE
IMPURITIES
INCONEL ALLOYS
IRRADIATION
LABORATORY EQUIPMENT
MEASURED VALUES
METALS, CERAMICS, AND OTHER MATERIALS
MIXING
MOLYBDENUM ALLOYS
NICKEL
NICKEL ALLOYS
NIOBIUM
NIOBIUM ALLOYS
PELLETS
PLANNING
PLUTONIUM
PLUTONIUM CARBIDES
PLUTONIUM OXIDES
POWDERS
PREPARATION
PRESSING
PRESSURE
RADIATION PROTECTION
REDUCTION
REMOTE HANDLING
SINTERING
SOLID SOLUTIONS
STAINLESS STEEL-304
STAINLESS STEELS
THERMAL CONDUCTIVITY
TITANIUM ALLOYS
URANI