Results of the DF-4 BWR (boiling water reactor) control blade-channel box test
The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.
- Research Organization:
- Sandia National Labs., Albuquerque, NM (USA)
- Sponsoring Organization:
- USNRC
- DOE Contract Number:
- AC04-76DP00789
- OSTI ID:
- 6531370
- Report Number(s):
- SAND-90-2716C; CONF-9010185-7; ON: DE91002423
- Resource Relation:
- Conference: 18. water reactor safety information meeting, Gaithersburg, MD (USA), 22-24 Oct 1990
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
BWR TYPE REACTORS
REACTOR ACCIDENTS
CONTROL ELEMENTS
MELTING
REACTOR CORES
OXIDATION
BORON CARBIDES
COPPER OXIDES
FIELD TESTS
FUEL RODS
HYDROGEN
LOSS OF COOLANT
LOW TEMPERATURE
REACTOR SAFETY
SPENT FUEL CASKS
STAINLESS STEELS
STEAM
THERMOCOUPLES
URANIUM OXIDES
ZIRCALOY
ZIRCONIUM
ACCIDENTS
ACTINIDE COMPOUNDS
ALLOYS
BORON COMPOUNDS
CARBIDES
CARBON COMPOUNDS
CASKS
CHALCOGENIDES
CHEMICAL REACTIONS
CONTAINERS
COPPER COMPOUNDS
ELEMENTS
FUEL ELEMENTS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
MEASURING INSTRUMENTS
METALS
NONMETALS
OXIDES
OXYGEN COMPOUNDS
PHASE TRANSFORMATIONS
REACTOR COMPONENTS
REACTORS
SAFETY
STEELS
TESTING
TRANSITION ELEMENT COMPOUNDS
TRANSITION ELEMENTS
URANIUM COMPOUNDS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210100* - Power Reactors
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220900 - Nuclear Reactor Technology- Reactor Safety