Low-temperature rupture behavior of Zircaloy-clad pressurized water reactor spent fuel rods under dry storage conditions
Creep rupture studies on five well-characterized Zircaloy-clad pressurized water reactor spent fuel rods, which were pressurized to a hoop stress of about145 MPa, were conducted for up to 2101 h at 323/sup 0/C. The conditions were chosen for limited annealing of in-reactor irradiation hardening. No cladding breaches occurred, although significant hydride agglomeration and reorientation took place in rods that cooled under stress. Observations are interpreted in terms of a conservatively modified Larson-Miller curve to provide a lower bound on permissible maximum dry-storage temperatures, assuming creep rupture as the life-limiting mechanism. If hydride reorientation can be ruled out during dry storage, 305/sup 0/C is a conservative lower bound, based on the creep-rupture mechanism, for the maximum storage temperature of rods with irradiation-hardened cladding to ensure a 100-yr cladding lifetime in an inert atmosphere. An oxidizing atmosphere reduced the lower bound on the maximum permissible storage temperature by about5/sup 0/C. While this lower bound is based on whole-rod data, other types of data on spent fuel behavior in dry storage might support a higher limit. This isothermal temperature limit does not take credit for the decreasing rod temperature during dry storage. High-temperature tests based on creep rupture as the limiting mechanism indicate that storage at temperatures between 400 and 440/sup 0/C may be feasible for rods that are annealed.
- Research Organization:
- Westinghouse Hanford Company, P.O. Box 1970, Mail Stop W/A-40, Richland, Washington 99352
- OSTI ID:
- 5944989
- Journal Information:
- Nucl. Technol.; (United States), Vol. 67:1
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
FUEL RODS
FRACTURE MECHANICS
PRESSURE EFFECTS
PWR TYPE REACTORS
CLADDING
CONTROLLED ATMOSPHERES
CREEP
DESTRUCTIVE TESTING
PRESSURIZING
RUPTURES
SPENT FUEL ELEMENTS
STRESS ANALYSIS
TEMPERATURE EFFECTS
ZIRCALOY
ALLOYS
ATMOSPHERES
DEPOSITION
FAILURES
FUEL ELEMENTS
MATERIALS TESTING
MECHANICAL PROPERTIES
MECHANICS
REACTOR COMPONENTS
REACTORS
SURFACE COATING
TESTING
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210200* - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled