Analysis of in-situ tritium recovery from solid fusion-reactor blankets

Conference · Tue Jan 01 00:00:00 EST 1980

OSTI ID:6329019

Smith, D L; Clemmer, R G; Jankus, V Z; Rest, J

The proposed concept for in-situ tritium recovery from the STARFIRE blanket involves circulation of a low pressure (approx. 0.05 MPa) helium through formed channels in the highly porous solid breeding material. Tritium generated within the grains must diffuse to the grain boundaries, migrate through the grain boundaries to the particle surface and then percolate through the packed bed to the helium purge channel. Highly porous ..cap alpha..-LiAlO/sub 2/ with a bimodal pore distribution is proposed for the breeding material to facilitate the tritium release.

• Research Organization: Argonne National Lab., IL (USA)
• DOE Contract Number: W-31-109-ENG-38
• OSTI ID: 6329019
• Report Number(s): CONF-801011-90; ON: DE81023919; TRN: 81-013042
• Resource Relation: Conference: 4. ANS topical meeting on the technology of controlled nuclear fusion, King of Prussia, PA, USA, 14 Oct 1980
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
BREEDING BLANKETS
TRITIUM RECOVERY
STARFIRE TOKAMAK
ALUMINIUM OXIDES
GRAIN BOUNDARIES
HELIUM
LITHIUM OXIDES
ALKALI METAL COMPOUNDS
ALUMINIUM COMPOUNDS
CHALCOGENIDES
CRYSTAL STRUCTURE
ELEMENTS
FLUIDS
GASES
LITHIUM COMPOUNDS
MICROSTRUCTURE
NONMETALS
OXIDES
OXYGEN COMPOUNDS
RARE GASES
REACTOR COMPONENTS
RECOVERY
THERMONUCLEAR REACTORS
TOKAMAK TYPE REACTORS
700206* - Fusion Power Plant Technology- Environmental Aspects