skip to main content

DOE PAGESDOE PAGES

This content will become publicly available on December 8, 2018

Title: Dimensional stability and anisotropy of SiC and SiC-based composites in transition swelling regime

Swelling, or volumetric expansion, is an inevitable consequence of the atomic displacement damage in crystalline silicon carbide (SiC) caused by energetic neutron irradiation. Because of its steep temperature and dose dependence, understanding swelling is essential for designing SiC-based components for nuclear applications. Here in this study, swelling behaviors of monolithic CVD SiC and nuclear grade SiC fiber – SiC matrix (SiC/SiC) composites were accurately determined, supported by the irradiation temperature determination for individual samples, following neutron irradiation within the lower transition swelling temperature regime. Slightly anisotropic swelling behaviors were found for the SiC/SiC samples and attributed primarily to the combined effects of the pre-existing microcracking, fiber architecture, and specimen dimension. A semi-empirical model of SiC swelling was calibrated and presented. Finally, implications of the refined model to selected swelling-related issues for SiC-based nuclar reactor components are discussed.
Authors:
ORCiD logo [1] ; ORCiD logo [2] ; ORCiD logo [2] ;  [3] ;  [4]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States)
  2. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  3. Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
  4. Electric Power Research Inst., Charlotte, NC (United States)
Publication Date:
Grant/Contract Number:
AC05-00OR22725
Type:
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 499; Journal Issue: C; Journal ID: ISSN 0022-3115
Publisher:
Elsevier
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE), Nuclear Reactor Technologies (NE-7)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS
OSTI Identifier:
1414701