Separation of silicon carbide-coated fertile and fissile particles by gas classification
The separation of /sup 235/U and /sup 233/U in the reprocessing of HTGR fuels is a key feature of the feed-breed fuel cycle concept. This is attained in the Fort St. Vrain (FSV) reactor by coating the fissile (Th-/sup 235/U) particles and the fertile (Th-/sup 233/U) particles separately with silicon carbide (SiC) layers to contain the fission products and to protect the kernels from burning in the head-end reprocessing steps. Pneumatic (gas) classification based on size and density differences is the reference process for separating the SiC-coated particles into fissile and fertile streams for subsequent handling. Terminal velocities have been calculated for the +- 2 sigma ranges of particle sizes and densities for ''Fissile B''--''Fertile A'' particles used in the FSV reactor. Because of overlapping particle fractions, a continuous pneumatic separator appears infeasible; however, a batch separation process can be envisioned. Changing the gas from air to CO/sub 2/ and/or the temperature to 300/sup 0/C results in less than 10 percent change in calculated terminal velocities. Recently reported work in gas classification is discussed in light of the theoretical calculations. The pneumatic separation of fissile and fertile particles needs more study, specifically with regard to (1) measuring the recoveries and separation efficiencies of actual fissile and fertile fractions in the tests of the pneumatic classifiers; and (2) improving the contactor design or flowsheet to avoid apparent flow separation or flooding problems at the feed point when using the feed rates required for the pilot plant.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- W-7405-ENG-26
- OSTI ID:
- 7345989
- Report Number(s):
- ORNL/TM-5091; TRN: 76-018688
- Country of Publication:
- United States
- Language:
- English
Similar Records
Delayed neutron method for measurement of fissile/fertile content of samples ranging from environmental to irradiated fuel
Delayed neutron method for measurement of fissile/fertile content of samples ranging from environmental to irradiated fuel
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
COATED FUEL PARTICLES
SORTING
FERTILE MATERIALS
FISSILE MATERIALS
SPENT FUELS
HEAD END PROCESSES
PNEUMATICS
THORIUM CARBIDES
URANIUM 233
URANIUM 235
URANIUM CARBIDES
VRAIN REACTOR
ACTINIDE COMPOUNDS
ACTINIDE ISOTOPES
ACTINIDE NUCLEI
ALPHA DECAY RADIOISOTOPES
CARBIDES
CARBON COMPOUNDS
ENERGY SOURCES
ENRICHED URANIUM REACTORS
EVEN-ODD NUCLEI
FISSIONABLE MATERIALS
FLUID MECHANICS
FUEL PARTICLES
FUELS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HEAVY NUCLEI
HELIUM COOLED REACTORS
HTGR TYPE REACTORS
ISOMERIC TRANSITION ISOTOPES
ISOTOPES
MECHANICS
MINUTES LIVING RADIOISOTOPES
NUCLEAR FUELS
NUCLEI
POWER REACTORS
RADIOISOTOPES
REACTOR MATERIALS
REACTORS
THORIUM COMPOUNDS
URANIUM COMPOUNDS
URANIUM ISOTOPES
YEARS LIVING RADIOISOTOPES
050800* - Nuclear Fuels- Spent Fuels Reprocessing
210802 - Nuclear Power Plants- Economics- Fuel Cycle