Creep-Rupture Study of Annealed Zircaloy 4: Stress and Temperature Effects

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dc.contributor.advisor Gerd Duscher, Committee Member en_US
dc.contributor.advisor Mohamed A. Bourham, Committee Member en_US
dc.contributor.advisor K. Linga Murty, Committee Chair en_US
dc.contributor.author Marple, Brian Wesley en_US
dc.date.accessioned 2010-04-02T18:18:35Z
dc.date.available 2010-04-02T18:18:35Z
dc.date.issued 2005-11-22 en_US
dc.identifier.other etd-11212005-141740 en_US
dc.identifier.uri http://www.lib.ncsu.edu/resolver/1840.16/2897
dc.description.abstract Zircaloys are widely used as fuel rod cladding in light water reactors (LWRs) because of their low cross-section for absorption of thermal neutrons. Currently, the United States does not permit reprocessing of spent fuel so the primary barrier for the spent fuel in a repository will be the fuel rod cladding. Due to the decay heat of the spent fuel, creep rupture is considered to be the primary cause of failure in spent fuel cladding over the long period of time that it will be stored. A fundamental understanding of the creep mechanisms in Zircaloys is crucial to accurately predicting the integrity of the fuel cladding over long periods of time. Zirconium has a hexagonally close-packed crystal structure and because of this, exhibits creep anisotropy that is affected not only by the texture, but also by temperature, stress, and loading. Since the stress imposed on the spent fuel during long-term storage will be relatively low compared to service conditions, the low stress creep behavior must be characterized and mechanistically understood to avoid non-conservative estimates based on in-pile creep data. In addition, loading of spent fuel in a repository will be due to the internal pressure generated by fission product gasses and from the inert gas introduced at the time of fuel fabrication. This work focuses on the creep rupture behavior and microstructural characterization of annealed Zircaloy-4 at temperatures ranging from 250°C-600°C and stresses from 27 MPa-350 MPa. Typically, fuel assemblies that have been fabricated from Zircaloy-4 are not in the annealed condition. Instead, they are cold-worked and stress relieved (CWSR). Since low stress creep rupture testing would take years at low temperatures, high temperatures are used to observe the effects of low stress in a reasonable amount of time. At such high temperatures, the grain structure of the CWSR material would change drastically. Therefore the material was annealed prior to testing to avoid this complication. Testing on unirradiated material will yield higher strain rates because of irradiation hardening. Therefore, estimates based on unirradiated creep rupture data would be conservative. Prior to testing, optical metallographs were taken to characterize the grain structure. A limited texture study was performed to evaluate the texture coefficients for each direction in the rod. Transmission electron microscopy (TEM) was also performed to characterize the initial dislocation microstructure. After testing, diametric measurements were taken and the strain rate determined. From data at various stresses, the activation energy was derived along with equations predictive of rupture such as the Larson-Miller parameter and the Monkman-Grant relationship. Specimens of interest were selected for optical metallography and to obtain TEM micrographs of the dislocation microstructure. The activation energy deduced was in excellent agreement of that for self-diffusion. Optical metallography showed slight grain elongation in samples tested at high stresses while grains remained equiaxed at low stresses. TEM showed significant sub-grain formation at low stresses and random dislocation organization at higher stresses. en_US
dc.rights I hereby certify that, if appropriate, I have obtained and attached hereto a written permission statement from the owner(s) of each third party copyrighted matter to be included in my thesis, dissertation, or project report, allowing distribution as specified below. I certify that the version I submitted is the same as that approved by my advisory committee. I hereby grant to NC State University or its agents the non-exclusive license to archive and make accessible, under the conditions specified below, my thesis, dissertation, or project report in whole or in part in all forms of media, now or hereafter known. I retain all other ownership rights to the copyright of the thesis, dissertation or project report. I also retain the right to use in future works (such as articles or books) all or part of this thesis, dissertation, or project report. en_US
dc.subject Rupture en_US
dc.subject Zircaloy en_US
dc.subject Creep en_US
dc.title Creep-Rupture Study of Annealed Zircaloy 4: Stress and Temperature Effects en_US
dc.degree.name MS en_US
dc.degree.level thesis en_US
dc.degree.discipline Nuclear Engineering en_US


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