Interpretation of experimental results from the CORA core melt progression experiments

Conference · Tue Jan 01 00:00:00 EST 1991

OSTI ID:6174809

Hohorst, J K; Allison, C M

Data obtained from the CORA bundle heatup and melting experiments, performed at Kernforschungszentrum, Karlsruhe, Germany, are being analyzed at the Idaho National Engineering Laboratory. The analysis is being performed as part of a systematic review of core melt progression experiments for the United States Nuclear Regulatory Commission to (a) develop an improved understanding of important phenomena occurring during a severe accident, (b) to validate existing severe accident models, and (c) where necessary, develop improved models. An assessment of the variations in damage progression behavior because of variations in test parameters (a) bundle design and size, (b) system pressure, (c) slow cooling of the damaged bundles in argon versus rapid quenching in water, and (d) bundle inlet temperatures and flow rates is provided in the paper. The influence of uncertainties in important test conditions is also discussed. Specific results presented include (a) bundle temperature, (b) the onset and movement of the oxidation front within the bundle, (c) fuel rod ballooning and rod failure, and (d) melt relocation and associated material interactions between bundle components and structures. 12 refs., 16 figs., 2 tabs.

• Research Organization: EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6174809
• Report Number(s): EGG-M-91247; CONF-911107-42; ON: DE92003271
• Resource Relation: Conference: Winter meeting of the American Nuclear Society (ANS), San Francisco, CA (United States), 10-15 Nov 1991
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
FUEL RODS
MELTDOWN
PWR TYPE REACTORS
BOUNDARY CONDITIONS
CORIUM
FAILURES
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
OXIDATION
REACTOR SAFETY
S CODES
TEMPERATURE DISTRIBUTION
ACCIDENTS
CHEMICAL REACTIONS
COMPUTER CODES
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
MECHANICS
POWER REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled