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.
A Review Paper on Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications1
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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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