Partial blockages in LMFBR fuel assemblies

Journal Article · Thu Jan 01 04:00:00 EST 1976 · Nucl. Saf., v. 17, no. 1, pp. 19-32

OSTI ID:4068760

Fontana, M H

Experimental and analytical data on the effects of partial blockages in simulated liquid-metal-cooled fast breeder reactor rod bundles are reviewed and the results presented. Experiments performed in the Fuel Failure Mockup at Oak Ridge National Laboratory with 13- and 24-subchannel inlet blockages in 19-rod sodium-cooled electrically heated rod bundles indicate that excessive temperatures do not occur as a result of the blockages. Similar experiments with nonheat-generating blockages of 6 central subchannels and 14 edge subchannels in the heated zone of the rod bundle indicate acceptable local temperature increases at operating conditions. Experiments with water mockups show complex flow patterns in the wake zone behind blockages. Estimates of local convective heat transfer in the wake zone were made by measuring mass interchange between the recirculating flow zone and the free stream by salt-injection techniques. Generalizations obtained from the water mockups were used to predict temperatures in sodium-cooled rod bundles. Estimates indicate that large blockages (approximately 3 in. in diameter) would be required to cause sodium boiling in full-scale reactors. (auth)

• Research Organization: Oak Ridge National Lab., TN
• NSA Number: NSA-33-022334
• OSTI ID: 4068760
• Journal Information: Nucl. Saf., v. 17, no. 1, pp. 19-32, Journal Name: Nucl. Saf., v. 17, no. 1, pp. 19-32; ISSN NUSAA
• Country of Publication: United States
• Language: English

Related Subjects

*FLOW BLOCKAGE-- TEMPERATURE DISTRIBUTION
*FUEL ASSEMBLIES-- FLOW BLOCKAGE
*LMFBR TYPE REACTORS-- FUEL ASSEMBLIES
220900* --Nuclear Reactor Technology--Reactor Safety
DATA
FLUID FLOW
HEAT TRANSFER
N77500 --Reactors--Power Reactors
Breeding
N77900* --Reactors--Reactor Safety & Environmental Aspects
REACTOR SAFETY
SIMULATION
TEST FACILITIES