Fuel rod material behavior during Test PCM-1. [PWR]
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 Laboratory (INL), Idaho Falls, ID (United States)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 5944414
- Report Number(s):
- NUREG/CR-0757; TREE-1333; TRN: 79-018411
- Country of Publication:
- United States
- Language:
- English
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Fuel rod behavior during test PCM-2. [PWR]
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL CANS
MICROSTRUCTURE
POWER-COOLING-MISMATCH ACCIDENTS
FUEL ELEMENT FAILURE
THERMAL STRESSES
PWR TYPE REACTORS
FUEL RODS
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
PERFORMANCE TESTING
REACTOR SAFETY
ZIRCALOY
ZIRCONIUM OXIDES
ACCIDENTS
ALLOYS
CHALCOGENIDES
CRYSTAL STRUCTURE
ENERGY TRANSFER
FUEL ELEMENTS
OXIDES
OXYGEN COMPOUNDS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
STRESSES
TESTING
TIN ALLOYS
TRANSITION ELEMENT COMPOUNDS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
ZIRCONIUM COMPOUNDS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled