A REVIEW PAPER ON AGING EFFECTS IN ALLOY 617 FOR GEN IVNUCLEAR REACTOR APPLICATIONS
- ORNL
To understand the response of Alloy 617 to long-time exposure conditions and determine the supplementary data needs for structural components in Gen IV nuclear reactors, the 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 (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:
- 964696
- Journal Information:
- Journal of Pressure Vessel Technology, Vol. 131, Issue 2; ISSN 0094-9930
- Country of Publication:
- United States
- Language:
- English
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