Post-test analysis of dryout test 7B' of the W-1 Sodium Loop Safety Facility Experiment with the SABRE-2P code. [LMFBR]

Conference · Thu Jan 01 00:00:00 EST 1981

OSTI ID:6070151

Rose, S D; Dearing, J F

An understanding of conditions that may cause sodium boiling and boiling propagation that may lead to dryout and fuel failure is crucial in liquid-metal fast-breeder reactor safety. In this study, the SABRE-2P subchannel analysis code has been used to analyze the ultimate transient of the in-core W-1 Sodium Loop Safety Facility experiment. This code has a 3-D simple nondynamic boiling model which is able to predict the flow instability which caused dryout. In other analyses dryout has been predicted for out-of-core test bundles and so this study provides additional confirmation of the model.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 6070151
• Report Number(s): CONF-810804-13; ON: DE81028565; TRN: 81-016043
• Resource Relation: Conference: 20. national heat transfer conference, Milwaukee, WI, USA, 2 Aug 1981
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LMFBR TYPE REACTORS
LOSS OF FLOW
DRYOUT
BOILING
COMPUTER CALCULATIONS
COMPUTER CODES
ETR REACTOR
FUEL ELEMENT CLUSTERS
REACTOR SAFETY EXPERIMENTS
S CODES
SODIUM
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ALKALI METALS
BREEDER REACTORS
ELEMENTS
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EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FUEL ASSEMBLIES
IRRADIATION REACTORS
ISOTOPE PRODUCTION REACTORS
LIQUID METAL COOLED REACTORS
METALS
PHASE TRANSFORMATIONS
REACTOR ACCIDENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
TANK TYPE REACTORS
TEST REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
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