An outboard shield design for TIBER-II with potential for tritium self-sufficiency: Appendix A-1

Conference · Thu Jan 01 00:00:00 EST 1987

OSTI ID:5757365

Sawan, M E; Sviatoslavsky, I N

The outboard breeding shield design for TIBER-II is described. The design allows for tritium self-sufficiency without compromising magnet protection, design simplicity, and the testing mission of the device. The shield consists of a beryllium pebble front zone backed by a steel pebble zone. The shield is cooled by an aqueous solution containing 16 g LiNO/sub 3/ per 100 cm/sup 3/. A double first wall is used to insure uniform cooling and minimize pressure. The design pressure for the outboard shield is 0.19 MPa and the coolant temperature is less than 75/sup 0/C. 6 refs., 5 figs., 3 tabs.

• Research Organization: Wisconsin Univ., Madison (USA). Fusion Technology Inst.
• DOE Contract Number: FG02-87ER52140
• OSTI ID: 5757365
• Report Number(s): CONF-871007-49; ON: DE88002459
• Resource Relation: Conference: 12. symposium on fusion engineering, Monterey, CA, USA, 12 Oct 1987; Other Information: Paper copy only, copy does not permit microfiche production
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
SHIELDING MATERIALS
DESIGN
TIBER-X TOKAMAK
BERYLLIUM
BREEDING BLANKETS
DIVERTORS
FIRST WALL
GRAPHITE
HYDRAULICS
LITHIUM COMPOUNDS
THERMAL ANALYSIS
TRITIUM
ALKALI METAL COMPOUNDS
ALKALINE EARTH METALS
BETA DECAY RADIOISOTOPES
BETA-MINUS DECAY RADIOISOTOPES
CARBON
ELEMENTAL MINERALS
ELEMENTS
FLUID MECHANICS
HYDROGEN ISOTOPES
ISOTOPES
LIGHT NUCLEI
MATERIALS
MECHANICS
METALS
MINERALS
NONMETALS
NUCLEI
ODD-EVEN NUCLEI
RADIOISOTOPES
REACTOR COMPONENTS
THERMONUCLEAR REACTOR WALLS
THERMONUCLEAR REACTORS
TOKAMAK TYPE REACTORS
YEARS LIVING RADIOISOTOPES
700201* - Fusion Power Plant Technology- Blanket Engineering
700209 - Fusion Power Plant Technology- Component Development & Materials Testing