An Analysis of the Rupture Behavior of Pressurized Fast Reactor Cladding Tubes Subjected to Thermal Transients

[+] Author and Article Information
J. M. Kramer, R. J. DiMelfi

Argonne National Laboratory, Argonne, Ill. 60439

J. Eng. Mater. Technol 101(3), 293-298 (Jul 01, 1979) (6 pages) doi:10.1115/1.3443690 History: Received September 01, 1978; Online August 17, 2010


The rupture behavior of 20 percent cold-worked type 316 stainless steel fast reactor fuel cladding, subjected to thermal transients typical of hypothetical accident conditions, is studied by considering the response of a thin-walled cylinder loaded by constant internal pressure. The high-stress low-temperature failure behavior is analyzed using a correlation from low temperature tensile properties. The low-stress high-temperature regime is shown to be described by a combined creep-deformation crack-growth-law formulation including annealing effects via grain growth. Failure temperatures and failure ductilities calculated using these models, compare favorably with experiment. It is also shown how the models can be extended to explain the observed reduction in failure temperature and failure ductility of cladding tubes that have been exposed to irradiation and/or corrosive environments.

Copyright © 1979 by ASME
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