In-core flow velocity profiles during the first fuel cycle at Hatch-1: inferred from neutron noise. Final report

Technical Report · Thu Oct 01 04:00:00 EDT 1981

OSTI ID:5910137

March-Leuba, J; Sweeney, F J; King, W T; Renier, J A; Wood, R T

In-core neutron noise data recorded during the first fuel cycle at Hatch-1 were used to infer in-core steam velocity and the presence of boiling in the flow bypass region between fuel boxes. A comparison of velocities thus inferred with calculated fuel-bundle-corner velocities at one local power range monitor (LPRM) position and reactor operating conditions shows that calculated fuel-bundle-corner steam velocities agree with the experimentally inferred velocities. It is likely that bypass boiling occurred at 79% power and 81% core flow at all instrument tube locations when bypass inlet orifices were plugged. No bypass boiling was observed at core elevations lower than the C-LPRM detectors.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 5910137
• Report Number(s): EPRI-NP-2083; ON: DE82003073
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
BOILING
BWR TYPE REACTORS
DATA ACQUISITION
DATA ANALYSIS
ENRICHED URANIUM REACTORS
FLUID FLOW
HATCH-1 REACTOR
PHASE TRANSFORMATIONS
POWER REACTORS
REACTOR COMPONENTS
REACTOR CORES
REACTOR NOISE
REACTORS
STEAM
VELOCITY
WATER COOLED REACTORS
WATER MODERATED REACTORS