Skip to main content
OSTI.GOVU.S. Department of Energy Office of Scientific and Technical Information
U.S. Department of Energy
Office of Scientific and Technical Information
 

DEVELOPMENT OF FUELED GRAPHITE CONTAINING PYROLYTIC-CARBON COATED CARBIDE PARTICLES FOR NONPURGED, GAS-COOLED REACTOR SYSTEMS

Technical Report ·
OSTI ID:4110310
; ;

From American Nuclear Society Meeting, New York, Nov. 1963. Progress is reported in several areas of development of fueled graphite containing coated particles for nonpurged gas-cooled reactor systems. The sol-gel process has been modified for making spherical particles of both thorium-uranium carbide and thorium-uranium oxide suitable for coating. Equipment has been assembled and methods have been developed for deposition of pyrolytic-carbon coatings under well-controlled conditions. Damage to coated particles during fabrication into a graphite matrix depends on the molding pressure and the volumetric content of coated particles. Vendor-supplied coated particles and fueled graphite spheres have been evaluated extensively in both inand out-of-reactor tests. Duplex- and triplex-coated particles have excellent fission-gas retention at 2050 deg F to burnups of l5 at.%, and one triplex sample has also shown good retention at 2500 deg F to 20 at.% burnup. Fueled graphite spheres containing coated particles have good irradiation performance, but the fission-gas release rates are somewhat higher than for unsupported coated particles. Fueled graphite spheres react with water vapor about as rapidiy as do Speer Mod-2 and ATJ grades of graphite. The diffusion rates in pyrolytic carbon are the same for uranium, thorium, and protactinium. The diffusion rates in the direction parallel to the deposition plane are much higher than those in the perpendicular direction. (auth)

Research Organization:
Oak Ridge National Lab., Tenn.
DOE Contract Number:
W-7405-ENG-26
NSA Number:
NSA-18-012596
OSTI ID:
4110310
Report Number(s):
CONF-187-44
Country of Publication:
United States
Language:
English

Similar Records

FISSION-GAS RELEASE FROM PYROLYTIC-CARBON-COATED FUEL PARTICLES DURING IRRADIATION
Journal Article · Sat Feb 29 23:00:00 EST 1964 · Nuclear Science and Engineering (U.S.) · OSTI ID:4076634
Fuel element development program for the Pebble Bed Reactor. Final Report
Technical Report · Sat Apr 29 23:00:00 EST 1961 · OSTI ID:4804887
GRAPHITE-MATRIX NUCLEAR FUEL SYSTEMS FOR THE PEACH BOTTOM HTGR
Technical Report · Tue Feb 19 23:00:00 EST 1963 · OSTI ID:4675616

Related Subjects

ABSORPTION
BURNUP
CARBIDES
CARBON
CASTING
CHEMICAL REACTIONS
COATING
CONFIGURATION
CONTROL
DEGASSING
DIFFUSION
FABRICATION
FAILURES
FISSION PRODUCTS
FUELS
GAS COOLANT
GRAPHITE
GRAPHITE ATJ
HIGH TEMPERATURE
IRRADIATION
LAYERS
MACHINE PARTS
MATERIALS TESTING
METALS, CERAMICS, AND OTHER MATERIALS
PERFORMANCE
PLANNING
PRESSURE
PROTACTINIUM
PYROLYSIS
REACTORS
SAMPLING
SOL-GEL PROCESS
SPHERES
THORIUM
THORIUM CARBIDES
THORIUM OXIDES
URANIUM
URANIUM CARBIDES
URANIUM OXIDES
VAPORS
VOLUME
WATER