Deformation and fracture characteristics of spent Zircaloy fuel cladding
For a better understanding of Zircaloy fuel-rod failure by the pellet-cladding interaction (PCI) phenomenon, a mechanistic study of deformation and fracture behavior of spent power reactor fuel cladding under simulated PCI conditions was conducted. Zircaloy-2 cladding specimens, obtained from fuel assemblies of operating power reactors, were deformed to fracture at 325/sup 0/C by internal gas pressurization in the absence of fission product simulants. Fracture characteristics and microstructures were examined via SEM, TEM, and HVEM. Numerous dislocation tangles and cell structures, observed in TEM specimens of cladding tubes that failed in a ductile manner, were consistent with SEM observations of a limited number of dimples characteristic of microvoid coalescence. A number of brittle-type failures were produced without the influence of fission product simulants. The brittle cracks occurred near the areas compressed by the Swagelok fittings of the internally pressurized tube and propagated from the outer to the inner surface. Since the outer surface was isolated and maintained under a flowing stream of pure helium, it is unlikely that the brittle-type failure was influenced by any fission product traces. SEM fractography of the brittle-type failure revealed a large area of transgranular pseudocleavage with limited areas of ductile fluting, which were similar in appearance to the surfaces produced by in-reactor PCI-type failures. A TEM evaluation of the cladding in the vicinity of the through-wall crack revealed numerous locations that contained an extensive amount of second-phase precipitate (Zr/sub 3/O). We believe that the brittle-type failures of the irradiated spent fuel cladding in the stress rupture experiments are associated with segregation of oxygen, which leads to the formation of the order structure, an immobilization of dislocations, and minimal plastic deformation in the material.
- Research Organization:
- Argonne National Lab., IL (USA)
- DOE Contract Number:
- W-31-109-ENG-38
- OSTI ID:
- 6436507
- Report Number(s):
- CONF-820662-5; ON: DE83007936
- Resource Relation:
- Conference: 6. international conference on zirconium in the nuclear industry, Vancouver, BC, Canada, 28 Jun 1982; Other Information: Portions are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ZIRCALOY 2
DEFORMATION
FRACTURES
PHYSICAL RADIATION EFFECTS
BIG ROCK POINT REACTOR
SPENT FUEL ELEMENTS
ALLOYS
BWR TYPE REACTORS
CHROMIUM ADDITIONS
CHROMIUM ALLOYS
ENRICHED URANIUM REACTORS
FAILURES
FUEL ELEMENTS
IRON ADDITIONS
IRON ALLOYS
POWER REACTORS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
THERMAL REACTORS
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
360103* - Metals & Alloys- Mechanical Properties
360106 - Metals & Alloys- Radiation Effects
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled