ANNUAL PROGRESS REPORT ON FUEL ELEMENT DEVELOPMENT FOR FY 1962

Technical Report · Wed Aug 15 00:00:00 EDT 1962

DOI:https://doi.org/10.2172/4772687· OSTI ID:4772687

Gibson, G W; Francis, W C

Activities in a project aimed at the improvement of fuel elements for high flux test reactors are reported. The investigation of new fuel compositions, distributions, and geometries is being undertaken to increase fuel life, to improve the flux distribution, and to provide a means of safely reaching higher reactor operating power and power density in these reactors. The effects of nuclear irradiation on the fuel and structural materials is being studied to predict the performance of these materials in more advanced reactor designs. A summary of the past year's progress is given and the fabrication and irradiation of samples containing up to 50 wt % U--Al alloys, cermets of UO/sub 2/, U/sub 3/O/ sub 8/, UC, UN, U/sub 3/Si, and Al, clad in various Al an d Be--Al materials is described. The use of ThO/sub 2/ and Th cores, the addition of BeO to cermet cores and high density fuei cores of U--Al intermetailics produced by powder metallurgy techniques were studied during the year. High strength APM claddings involving Al/sub 2/O/sub 3/ contents from 8 to 10% were tested and indicate the need for improved quality control of the APM material. Duplex claddings involving burnable poison layers and APM clad with corrosion resistant X8001 showed promise where special properties are desired. The results of the work continue to demonstrate the excellent radiation stability of U--Al fuels even after long irradiation exposure at elevated temperatures. Tests up to 350 deg F and after 50% burnup of the U/sup 235/ in U--Al alloys, show no appreciable dimensional or microstructure changes. UO/sub 2/ and U/sub 3/O/sub 8/ react with Al under radiation to form UAl/sub 4/. Tensile tests of these fuels at ambient temperatures show appreciable loss in ductility with irradiation; several compositions actually exhibiting zero ductility. Irradiation at temperatures up to 200 deg F of cold-worked and of heat-treated Al does not destroy the pre- irradiation hardness and strength of these materials. Computer optimization of fuel element geometries from the standpoint of heat transfer, hydraulics, and strength resulted in the design and fabrication of a 32-plate fuel element. Hydraulic tests produced favorable results and the element is ready for MTR testing. Future work stressing materials development will be directed toward extending U--Al fuels to use at 400 to 800 deg F. Continued studies on graded fuels, Be damage, and the Th--U/sup 233/ system are also planned. Tensile testing will be extended to higher temperatures in pre- and post-irradiation measurements and the study of the effect on cold worked and tempered materials of elevated temperature-radiation exposures will be continued. (auth)

• Research Organization: Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
• Sponsoring Organization: USDOE
• DOE Contract Number: AT(10-1)-205
• NSA Number: NSA-17-004238
• OSTI ID: 4772687
• Report Number(s): IDO-16799
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-63
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ALUMINUM
ALUMINUM ALLOYS
ALUMINUM OXIDES
BERYLLIUM
BERYLLIUM ALLOYS
BERYLLIUM OXIDES
BUILDING MATERIALS
BURNUP
CERMETS
CHEMICAL REACTIONS
COLD WORKING
COMPUTERS
CONFIGURATION
CORROSION PROTECTION
DENSITY
DISTRIBUTION
DUCTILITY
EFFICIENCY
EXPANSION
FABRICATION
FUEL CANS
FUEL ELEMENTS
FUELS
HARDNESS
HEAT TRANSFER
HEAT TREATMENTS
HIGH TEMPERATURE
HYDRAULICS
INSPECTION
INTERMETALLIC COMPOUNDS
IRRADIATION
LAYERS
LOSSES
MATERIALS TESTING
MECHANICAL PROPERTIES
MELTING
METALLURGY
MTR
NEUTRON FLUX
OPTIMIZATION
PERFORMANCE
PLANNING
PLATES
POISONING
POWDER METALLURGY
POWER
PREPARATION
RADIATION CHEMISTRY
R