Cost of processing fuel from a molten salt, fusion/fission, hybrid reactor blanket

Conference · Sun Sep 01 04:00:00 EDT 1985 · Fusion Technol.; (United States)

OSTI ID:5705344

Watson, J S; Brashears, D E; Grimes, W R

A conceptual flowsheet was prepared for continuous processing of molten salt used as the blanket material for breeding tritium and fissile material (/sup 233/U) in a fusion/fission hybrid reactor. The salt, which has a melting point of about 530/sup 0/C, was 70 mol % LiF, 12 mol % BeF/sub 2/, and 18 mol % ThF/sub 4/. The hybrid reactor generates 3000 MWe, and the blanket contains 65 m/sup 3/ of the breeding salt.

• Research Organization: Oak Ridge National Laboratory, Oak Ridge, TN
• OSTI ID: 5705344
• Report Number(s): CONF-850405-
• Conference Information: Journal Name: Fusion Technol.; (United States) Journal Volume: 8:2
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
700205* -- Fusion Power Plant Technology-- Fuel
Heating
& Injection Systems
700209 -- Fusion Power Plant Technology-- Component Development & Materials Testing
ACTINIDE COMPOUNDS
ACTINIDE ISOTOPES
ACTINIDE NUCLEI
ALKALI METAL COMPOUNDS
ALKALINE EARTH METAL COMPOUNDS
ALPHA DECAY RADIOISOTOPES
BERYLLIUM COMPOUNDS
BERYLLIUM FLUORIDES
BETA DECAY RADIOISOTOPES
BETA-MINUS DECAY RADIOISOTOPES
BREEDING
BREEDING BLANKETS
CHEMICAL COMPOSITION
COST ESTIMATION
DIAGRAMS
EVEN-ODD NUCLEI
FEASIBILITY STUDIES
FLOWSHEETS
FLUORIDES
FLUORINE COMPOUNDS
HALIDES
HALOGEN COMPOUNDS
HEAVY NUCLEI
HYBRID REACTORS
HYDROGEN ISOTOPES
ISOTOPES
LIGHT NUCLEI
LITHIUM COMPOUNDS
LITHIUM FLUORIDES
LITHIUM HALIDES
MATERIALS
MOLTEN SALTS
NUCLEAR FUEL CONVERSION
NUCLEI
ODD-EVEN NUCLEI
POWER RANGE 1-10 GW
RADIOISOTOPES
REACTOR COMPONENTS
SALTS
THERMONUCLEAR REACTOR MATERIALS
THORIUM COMPOUNDS
THORIUM FLUORIDES
TRITIUM
URANIUM 233
URANIUM ISOTOPES
YEARS LIVING RADIOISOTOPES