To understand the response of Alloy 617 to long-time exposure conditions and to determine the supplementary data needs for structural components in Gen IV nuclear reactors, literature of aging and aging effects in the alloy was reviewed. Most of the reviewed data were produced in connection with the international research effort supporting high temperature gas-cooled reactor projects in the 1970s and 1980s. Topics considered included microstructural changes, hardness, tensile properties, toughness, creep-rupture, fatigue, and crack growth. It became clear that, for the long-time very high-temperature conditions of the Gen IV reactors, a significant effort would be needed to fully understand and characterize property changes. Several topics for further research were recommended.
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e-mail: renw@ornl.gov
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April 2009
Technology Review
A Review Paper on Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications1
Weiju Ren,
Weiju Ren
Materials Science and Technology Division,
e-mail: renw@ornl.gov
Oak Ridge National Laboratory
, MS-6155, Building 4500-S, Oak Ridge, TN 37831
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Robert Swimdeman
Robert Swimdeman
Search for other works by this author on:
Weiju Ren
Materials Science and Technology Division,
Oak Ridge National Laboratory
, MS-6155, Building 4500-S, Oak Ridge, TN 37831e-mail: renw@ornl.gov
Robert Swimdeman
J. Pressure Vessel Technol. Apr 2009, 131(2): 024002 (15 pages)
Published Online: December 30, 2008
Article history
Received:
October 5, 2006
Revised:
April 25, 2007
Published:
December 30, 2008
Citation
Ren, W., and Swimdeman, R. (December 30, 2008). "A Review Paper on Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications." ASME. J. Pressure Vessel Technol. April 2009; 131(2): 024002. https://doi.org/10.1115/1.2967885
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