Fuel Rod Material Behavior During Test PCM-1 [PWR]
- Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States)
The analysis is presented of the fuel rod materials behavior based on the postirradiation examination of the Test PCM-1 fuel rod from the Power-Cooling-Mismatch (PCM) Test Series. The test objective was to evaluate the behavior of a single pressurized water reactor (PWR) type fuel rod subjected to film boiling operation at high power following rod failure. The failure mechanisms and subsequent breakup of the fuel and cladding are discussed. The fuel rod cladding temperature profile is determined by metallographic examination of cladding microstructures and calculations based on kinetic correlations of the cladding external surface reaction layers with the duration of film boiling. Cladding-coolant and cladding-fuel interactions are investigated by metallographic and microprobe examination and chemical analysis of the cladding. Fuel restructuring and chemical changes are also addressed.
- Research Organization:
- Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USNRC; USDOE
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 5944414
- Report Number(s):
- NUREG/CR--0757; TREE--1333
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
CHALCOGENIDES
CRYSTAL STRUCTURE
ENERGY TRANSFER
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
MICROSTRUCTURE
OXIDES
OXYGEN COMPOUNDS
PERFORMANCE TESTING
POWER-COOLING-MISMATCH ACCIDENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TESTING
THERMAL STRESSES
TIN ALLOYS
TRANSITION ELEMENT COMPOUNDS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
ZIRCONIUM COMPOUNDS
ZIRCONIUM OXIDES
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
CHALCOGENIDES
CRYSTAL STRUCTURE
ENERGY TRANSFER
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
MICROSTRUCTURE
OXIDES
OXYGEN COMPOUNDS
PERFORMANCE TESTING
POWER-COOLING-MISMATCH ACCIDENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TESTING
THERMAL STRESSES
TIN ALLOYS
TRANSITION ELEMENT COMPOUNDS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
ZIRCONIUM COMPOUNDS
ZIRCONIUM OXIDES