A REVIEW OF AGING EFFECTS IN ALLOY 617 FOR GEN IV NUCLEAR REACTOR APPLICATIONS
- ORNL
The literature was reviewed of aging and aging effects in Alloy 617 to determine the supplementary data needed to understand the response of the alloy to long-time exposure conditions being considered for structural components in Gen IV nuclear reactors. Most of the data were produced in connection with the international research effort supporting High Temperature Gas-Cooled Reactor (HTGR) 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.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 1093003
- Resource Relation:
- Conference: 2006 ASME Pressure Vessels and Piping Division Conference, Vancouver, British Columbia, Can, Canada, 20060723, 20060727
- Country of Publication:
- United States
- Language:
- English
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