Rod bundle critical heat flux at low pressure

Conference · Sun Jan 01 00:00:00 EST 1978

OSTI ID:6542404

Wadkins, R P; Ambrosek, R G; Young, M W

A study of critical heat flux (CHF) in a close-packed bundle of electrically heated rods which simulated the closely spaced nuclear fuel rods of the Power Burst Facility (PBF) was performed. The study examined the effects that close rod spacing and rod bowing would have on CHF at low, near-atmospheric, pressure conditions similar to PBF conditions. The PBF nuclear reactor, which is used in the Nuclear Regulatory Thermal Fuels Behavior Program, has an open vessel and a driver core with forced upward flow through the close-packed rods. The core design power is 40 MW. An effort was undertaken to design a reload core with a steady state power level of 50 MW.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 6542404
• Report Number(s): CONF-781022-21; TRN: 79-001981
• Resource Relation: Conference: Meeting on nuclear power reactor safety, Brussels, Belgium, 16 Oct 1978
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PBF REACTOR
CRITICAL HEAT FLUX
FUEL ELEMENT CLUSTERS
SIMULATION
FUEL ASSEMBLIES
HEAT FLUX
PULSED REACTORS
REACTORS
TANK TYPE REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors