Spent fuel resistance to internally produced cladding degradation
These tests were conducted over a narrow temperature range considerably above anticipated disposal conditions and utilized only one set of rods from a single reactor. The following conclusions are made: The measured cladding strain was sufficiently large so that failure mechanism verification by inducing breaches in unmodified rods heated to elevated temperatures for short periods of time does not appear to be practical based on a stress rupture mechanism. At the elevated test temperatures, though, Blackburn's formulization based on stress rupture gives very conservative estimates of breach times. In addition to the high cladding strain, the fuel exhibited no additional gas release or axial fission product migration at 482/sup 0/C. The nondestructive examination gave no additional indication of internal deterioration of the fuel rod.
- Research Organization:
- Westinghouse Hanford Co., Richland, WA (USA); Battelle Columbus Labs., OH (USA)
- DOE Contract Number:
- AC14-76FF02170
- OSTI ID:
- 5201336
- Report Number(s):
- HEDL-SA-2138; CONF-8005107-1
- Country of Publication:
- United States
- Language:
- English
Similar Records
High temperature postirradiation materials performance of spent pressurized water reactor fuel rods under dry storage conditions
High temperature post-irradiation performance of spent pressurized-water-reactor fuel rods under dry-storage conditions
Related Subjects
Handling
& Storage
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACCIDENTS
DEFORMATION
ENRICHED URANIUM REACTORS
FISSION PRODUCT RELEASE
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
HIGH TEMPERATURE
MANAGEMENT
MATERIALS TESTING
NONDESTRUCTIVE TESTING
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SIMULATION
SPENT FUEL STORAGE
TESTING
THERMAL REACTORS
TURKEY POINT-3 REACTOR
UNDERGROUND DISPOSAL
WASTE DISPOSAL
WASTE MANAGEMENT
WATER COOLED REACTORS
WATER MODERATED REACTORS