High temperature post-irradiation performance of spent pressurized-water-reactor fuel rods under dry-storage conditions
Post-irradiation studies on failure mechanisms of well characterized PWR rods were conducted for up to a year at 482, 510 and 571/sup 0/C in unlimited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn's analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding which decreases the internal rod pressures. The cladding creep also contributes to radial cracks, through the external oxide and internal FCCI layers, which propagated into and arrested in an oxygen stabilized ..cap alpha..-Zircaloy layer. There were no signs of either additional cladding hydriding, stress-corrosion cracking or fuel pellet degradation. Using the Larson-Miller formulization, a conservative maximum storage temperature of 400/sup 0/C is recommended to ensure a 1000-year cladding lifetime. This accounts for crack propagation and assumes annealing of the irradiation-hardened cladding.
- Research Organization:
- Hanford Engineering Development Lab., Richland, WA (USA); Battelle Columbus Labs., OH (USA)
- DOE Contract Number:
- AC06-76FF02170
- OSTI ID:
- 5720265
- Report Number(s):
- HEDL-SA-2484-FP; ON: DE82003780
- 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
Low-temperature rupture behavior of Zircaloy clad pressurized water reactor spent fuel rods under dry storage conditions
Low-temperature rupture behavior of Zircaloy-clad pressurized water reactor spent fuel rods under dry storage conditions
Journal Article
·
Wed Mar 31 23:00:00 EST 1982
· Nucl. Technol.; (United States)
·
OSTI ID:6885211
Low-temperature rupture behavior of Zircaloy clad pressurized water reactor spent fuel rods under dry storage conditions
Technical Report
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:5244914
Low-temperature rupture behavior of Zircaloy-clad pressurized water reactor spent fuel rods under dry storage conditions
Journal Article
·
Mon Oct 01 00:00:00 EDT 1984
· Nucl. Technol.; (United States)
·
OSTI ID:5944989
Related Subjects
050900* -- Nuclear Fuels-- Transport
Handling
& Storage
052002 -- Nuclear Fuels-- Waste Disposal & Storage
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
ACCIDENTS
ALLOYS
CRACK PROPAGATION
CREEP
ENRICHED URANIUM REACTORS
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
MECHANICAL PROPERTIES
PERFORMANCE
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SPENT FUEL STORAGE
STORAGE
THERMAL REACTORS
TIN ALLOYS
TURKEY POINT-3 REACTOR
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
Handling
& Storage
052002 -- Nuclear Fuels-- Waste Disposal & Storage
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
ACCIDENTS
ALLOYS
CRACK PROPAGATION
CREEP
ENRICHED URANIUM REACTORS
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
MECHANICAL PROPERTIES
PERFORMANCE
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SPENT FUEL STORAGE
STORAGE
THERMAL REACTORS
TIN ALLOYS
TURKEY POINT-3 REACTOR
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS