DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS

Technical Report · Thu Nov 19 00:00:00 EST 1959

DOI:https://doi.org/10.2172/4214694· OSTI ID:4214694

Thurber, W C; Beaver, R J

The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl/sub 4/ or UAl/sub 3/ compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm/sup 3/), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze the plates into fuel assemblies. Corrosion tests of the fuel material were conducted in 20 and 60 deg C water to determine the integrity of the fuel material in the event of sin inadventent cladding failure. In addition, specimens were prepared to evaluate penformance under extensive irradiation Prior to studying the uranium carbide-aluminum system, methods for preparing the carbides were investigated. Are melting uranium and carnon was satisfactory for obtaining small quantities of various carbides. Later, reaction of graphite with UO/sub 2/ was successfully employed in the preparation of large quantities of UC/sub 2/, Studies of the chemical compatibility of cold-pressed compacts containing 50 wt% uranium carbide dispersed in aluminum revealed a marked trend toward stebifity as the carbon content of the uranium carbide increased from 446 to 9.20% C. Severe volume increases occurred in monocarbide dispersions with attendant formation of large quantities of the uranium-allumnim inter-metallic compounds. Dicarbide dispersions, on the other band, exhibited negligible reaction with aluminum after extended periods at 580 and 620 deg C. However, it was demonstrated that hydrogen can promote a reaction in UC/sub 2/-Al compacts. The hydrogen appears to reduce the UC/sub 2/ to UC which can subsequently react with aluminum producing the previously noted deleterious effects. A growth study at 605 deg C of composite fuel plates containing 59 wt.% UC/sub 2/ revealed insignificant changes within processing periods envisioned for fuel element processing. However, plate elongations as high as 2.5% were observed after 100 hr at this temperature. Severe blistering which occurred on fuel plates fabricated in the initial stages of the investigation was attributed to gaseous hydrocarbons, and the condition was ellminated by vacuum degasification of cold-pressed compacts. With the exception of the degasification requirement, procedures for manufacturing UC- bearing fuel elements were identical to those specified for the Geneva Conference Reactor fuel elements. Dispersions of uranium dicarbide corroded catastrophically in 20 and 60 deg C water, thus limiting the application of this material However, spocimens were prepared and insented in the MTR to evaluate the irradiation behavior of this fuel because of its potential application in onganic- cooled reactors. (auth)

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE
• DOE Contract Number: W-7405-ENG-26
• NSA Number: NSA-14-003208
• OSTI ID: 4214694
• Report Number(s): ORNL-2618
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-60
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ALUMINUM
ALUMINUM ALLOYS
CHEMICAL REACTIONS
CORROSION
DEGASSING
DENSITY
DISPERSIONS
FABRICATION
FAILURES
FUEL CANS
FUEL ELEMENTS
GASES
GRAPHITE
HIGH TEMPERATURE
HOMOGENEOUS REACTORS
HYDROCARBONS
HYDROGEN
INTERMETALLIC COMPOUNDS
IRRADIATION
MATERIALS TESTING
MELTING
PERFORMANCE
PLATES
POWDERS
QUANTITY RATIO
RESEARCH REACTORS
STABILITY
TESTING
URANIUM ALLOYS
URANIUM CARBIDES
URANIUM OXIDES
VOLUME