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Article
Correlation Between the Microstructure and Mechanical Properties of Irradiated Fe-9Cr ODS
2014 Annual Meeting - Transactions of the American Nuclear Society and Embedded Topical Meeting: Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors, NSFM 2014
  • M. Swenson, Boise State University
  • C. Dolph, Boise State University
  • J. Wharry, Boise State University
Document Type
Conference Proceeding
Publication Date
1-1-2014
Abstract
The growing global demand for energy will increasingly call upon fusion reactors and Generation IV nuclear fission reactors to supply safe and reliable energy worldwide. Ferritic/martensitic (F/M) alloys are leading candidates for structural components in these reactors because of their high strength, dimensional stability, and low activation. In novel reactor concepts, these materials will be subject to extreme operating conditions, accumulating doses of irradiation up to a few hundred displacements per atom (dpa) at temperatures as high as 600°C. Oxide dispersion strengthened (ODS) F/M alloys containing a dispersion of Y-Ti-0 nanoclusters have been developed to operate at even higher temperatures.
Copyright Statement

This document was originally published by the American Nuclear Society in the 2014 Annual Meeting - Transactions of the American Nuclear Society and Embedded Topical Meeting: Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors, NSFM 2014. Copyright 2014 by the American Nuclear Society, La Grange Park, Illinois.

Citation Information
Swenson, M.; Dolph, C.; and Wharry, J.. (2014). "Correlation Between the Microstructure and Mechanical Properties of Irradiated Fe-9Cr ODS". 2014 Annual Meeting - Transactions of the American Nuclear Society and Embedded Topical Meeting: Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors, NSFM 2014, 110, 421-424.