Low-temperature rupture behavior of Zircaloy clad pressurized water reactor spent fuel rods under dry storage conditions

Technical Report · Sat Jan 01 04:00:00 EST 1983

OSTI ID:5244914

Einziger, R E; Kohli, R

Creep rupture studies on five well-characterized Zircaloy clad pressurized water reactor spent fuel rods, which were pressurized to a hoop stress of approximately 145 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-year cladding lifetime in an inert atmosphere. An oxidizing atmosphere reduces the lower bound on the maximum permissible storage temperature by approx. 5/sup 0/C. While high-temperature tests based on creep rupture as the limiting mechanism indicate that storage at temperatures between 400/sup 0/C and 440/sup 0/C may be feasible for rods which are annealed, tests to study rod performance in the 305/sup 0/ to 400/sup 0/C temperature range have not been conducted. 37 references, 10 figures, 7 tables.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA); Battelle Columbus Labs., OH (USA)
• DOE Contract Number: AC06-76FF02170
• OSTI ID: 5244914
• Report Number(s): HEDL-7400; ON: DE84007170
• Country of Publication: United States
• Language: English

Related Subjects

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36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
360106 -- Metals & Alloys-- Radiation Effects
ALLOYS
CREEP
DATA
DRY STORAGE
EXPERIMENTAL DATA
FAILURES
FUEL ELEMENTS
HARDENING
HARDNESS
HYDRIDES
HYDROGEN COMPOUNDS
INFORMATION
MECHANICAL PROPERTIES
NUMERICAL DATA
PHYSICAL RADIATION EFFECTS
PWR TYPE REACTORS
RADIATION EFFECTS
RADIATION HARDENING
REACTOR COMPONENTS
REACTORS
RUPTURES
SPENT FUEL ELEMENTS
SPENT FUEL STORAGE
STORAGE
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS