Upper-plenum flow-distribution study for the SRS (Savannah River site) L reactor using the TRAC-PF1/MOD2 code
Conference
·
· Transactions of the American Nuclear Society; (USA)
OSTI ID:6021882
- Los Alamos National Lab., NM (USA)
During a hypothetical loss-of-coolant accident (LOCA) in the L reactor at the Savannah River site (SRS), a mixture of air and water was recirculated to a 4.88-m (16-ft)-diam, 0.22-m (8 3/4-in.)-high plenum above the fuel assemblies. This plenum distributes coolant to the 600 fuel and target assemblies, which have downflow. The height of the coolant at each of the 600 positions in the plenum determines the flow to that assembly. The ability to predict this level is critical to an accurate analysis of a LOCA in the L-reactor geometry. The transient reactor analysis code (TRAC-PF1/MOD2) is uniquely suited to analyzing this type of problem because of its three-dimensional modeling capabilities. In 1985, six hydraulic tests were conducted by the Savannah River Laboratory to determine the L-reactor flow distributions during normal and accident conditions. This paper presents a comparison of the levels calculated by TRAC and the measured data.
- OSTI ID:
- 6021882
- Report Number(s):
- CONF-900608--
- Conference Information:
- Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 61
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACCURACY
COMPUTER CODES
ENERGY TRANSFER
FLUID MECHANICS
FUEL ASSEMBLIES
GEOMETRY
HEAT EXCHANGERS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
L REACTOR
LANL
LOSS OF COOLANT
MATHEMATICS
MECHANICS
NATIONAL ORGANIZATIONS
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SPECIAL PRODUCTION REACTORS
T CODES
THREE-DIMENSIONAL CALCULATIONS
TRANSIENTS
US DOE
US ORGANIZATIONS
VOID FRACTION
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACCURACY
COMPUTER CODES
ENERGY TRANSFER
FLUID MECHANICS
FUEL ASSEMBLIES
GEOMETRY
HEAT EXCHANGERS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
L REACTOR
LANL
LOSS OF COOLANT
MATHEMATICS
MECHANICS
NATIONAL ORGANIZATIONS
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SPECIAL PRODUCTION REACTORS
T CODES
THREE-DIMENSIONAL CALCULATIONS
TRANSIENTS
US DOE
US ORGANIZATIONS
VOID FRACTION