Post Irradiation Examination of SiC Tube Subjected to Simultaneous Irradiation and Radial High Heat Flux
- Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)
- General Atomics, 3550 General Atomics Ct., San Diego, CA 92121 (United States)
Silicon carbide (SiC) based composite is considered among leading candidate materials for accident tolerant fuel (ATF) cladding in light water reactors (LWRs) because of its outstanding irradiation and oxidation resistance, as well as low neutron absorption. Under normal operation conditions, the synergism between neutron irradiation and a high radial heat flux though the cladding thickness is expected to produce significant stress. This stress is attributed to a significant through-thickness temperature gradient and the resulting differential swelling strain due the highly temperature-dependent point-defect swelling of SiC. Although modeling of the in-pile stress state of SiC cladding has been conducted using finite element analysis, the analysis is challenging because of the complex stress state that depends upon swelling, irradiation creep, and thermal conductivity, all of which are both temperature and dose-dependent. These integral models have not yet been verified by experiments. In order to experimentally validate the multi-physics thermo-mechanical models of ATF SiC cladding during operation, SiC tube specimens have been irradiated under a high radial heat flux. This paper reports preliminary results from the post-irradiation examination (PIE) in on-going project supported by the U.S. Department of Energy's Nuclear Science User Facility program. The objectives of the PIE are (1) determination of the irradiation temperature distribution within the tube specimens, and (2) evaluation of the state and magnitude of the stress within the tube specimens arising from the temperature gradient. This paper mainly presents findings for objective (1)
- OSTI ID:
- 23047432
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 116; Conference: 2017 Annual Meeting of the American Nuclear Society, San Francisco, CA (United States), 11-15 Jun 2017; Other Information: Country of input: France; 7 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ABSORPTION
ACCIDENT-TOLERANT NUCLEAR FUELS
CLADDING
CREEP
FINITE ELEMENT METHOD
HEAT FLUX
IRRADIATION
NEUTRONS
OXIDATION
POINT DEFECTS
POST-IRRADIATION EXAMINATION
SILICON CARBIDES
STEADY-STATE CONDITIONS
SWELLING
TEMPERATURE DEPENDENCE
TEMPERATURE GRADIENTS
THERMAL CONDUCTIVITY
TUBES
WATER COOLED REACTORS
WATER MODERATED REACTORS