Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

Koyanagi, Takaaki; Katoh, Yutai

doi:10.1016/j.jnucmat.2017.07.007

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

Journal Article · Tue Jul 04 00:00:00 EDT 2017 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2017.07.007· OSTI ID:1376450

Koyanagi, Takaaki ^[1]; Katoh, Yutai ^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this paper examined SiC/SiC composites following neutron irradiation at 230–340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. Finally, this study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

View Accepted Manuscript (DOE)

Research Organization:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5); USDOE Office of Science (SC)

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1376450

Alternate ID(s):: OSTI ID: 1550415

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 494; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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