MATERIAL BUCKLING MEASUREMENTS FOR GRAPHITE MODERATED, URANIUM CARBIDE- FUELED LATTICES

Technical Report · Wed May 22 00:00:00 EDT 1963

OSTI ID:4699844

Johnson, E J

Material buckling values for a series of graphitemoderated uranium carbide-fueled lattice s were obtained by exponential experiment using the AE-6 thermal column as the neutron source. Three lattice pitches with centerto-center fuel spacings of 9.02, 11.015, and 12.056 in. were examined. The unrt cell of the lattices was hexagonally shaped with a central control channel surrounded by 120 deg segments of six fuel elements located at the corners of the hexagon. The fuel consisted of 0.5-in.-diam. pins of 3 wt.% enriched uranium carbide arranged in an 18-pin cluster to mockup the element proposed for a 500 Mwe power reactor design. The fuel process tubes also contained solid sodium to mockup core cooiant. The central control channel can be changed to simulate either the wet or dry control concepts. The measurements were compared with theoretical values obtained from four-group theory with fast-flux weighting and thermal pin disadvantage factors. There is considerable disagreement between the two sets of results. Critical experiments performed for the 11-in. spacing only tend to bear out the validity of the exponential measurements. (auth)

• Research Organization: Atomics International. Div. of North American Aviation, Inc., Canoga Park, Calif.
• DOE Contract Number: AT(11-1)-GEN-8
• NSA Number: NSA-17-031267
• OSTI ID: 4699844
• Report Number(s): NAA-SR-Memo-8522
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-63
• Country of Publication: United States
• Language: English

Related Subjects

PHYSICS
AE-6
BUCKLING
CONFIGURATION
CONTROL ELEMENTS
CONTROL SYSTEMS
COOLANTS
CRITICAL ASSEMBLIES
CRITICALITY
ERRORS
EXPONENTIAL PILES
FAST NEUTRONS
FUEL ELEMENTS
FUELS
GRAPHITE MODERATOR
GROUP THEORY
MEASURED VALUES
MOCKUP
MODERATORS
NEUTRON FLUX
NEUTRON SOURCES
PLANNING
POWER PLANTS
REACTOR CORE
REACTORS
RODS
SODIUM
SOLIDS
SPACERS
THERMAL COLUMN
THERMAL NEUTRONS
TUBES
URANIUM CARBIDES