Preliminary Results From High Temperature Scoping Irradiation Experiments Of Selected Gen IV Structural Metallic Materials
- ORNL
The performance of Generation IV reactors as a class will be determined by the behavior of advanced engineering materials. In the case of materials utilized for reactor internals and pressure vessels, the effects of irradiation are major issues. The environmental conditions for most of the Gen IV reactors are generally beyond present day reactor technology, especially as regards the combinations of operating temperatures, reactor coolant characteristics, and neutron spectra. In some of the applications, the conditions lay well beyond advanced research programs in radiation effects on materials. Therefore, new experimental data as well as analytical predictions of expected behavior of candidate materials at conditions for which there are no experimental data will be required. In the Gen IV Materials Program cross-cutting task, plans are being developed and irradiations and testing are being carried out to address the issues described above. This paper provides preliminary results for the first series of scoping irradiation experiments with selected metallic alloys, some of which are considered candidate materials for current Gen IV reactor applications, while others are considered as potential future candidate materials. The material classes represented are (1) nickel-base alloys (alloy 800H and Inconel 617; (2) advanced oxide-dispersion strengthened steels (14WT and 14YWT); and (3) commercial ferritic-martensitic steels (9Cr-1MoV). The results presented are from tensile tests using small flat tensile specimens (SS-3) in both the unirradiated and irradiated conditions. Specimens were irradiated in so-called rabbit capsules in the High-Flux Isotope Reactor (HFIR) at temperatures from 550 to 750 C and to irradiation doses from about 1.28 to 1.61 dpa. For the preliminary results from the first phase of this study, the annealed 9Cr-1MoV shows small amounts of irradiation-induced hardening. For the Alloy 800H, however, the hardening resulting from the 580 C irradiation was significant, with increases in yield and ultimate strengths on the order of 50 to 100%. Results from the 660 C irradiation also show hardening, but with extremely low tensile elongations when tested at 700 C. For the ODS 14WT and 14YWT materials, the overall results do not indicate significant effects of irradiation at this relatively low exposure. The Inconel 617 will be tested in the second phase of testing.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 1019323
- Resource Relation:
- Conference: 2007 International Congress on Advances in Nuclear Power Plants, Nice, France, 20070513, 20070513
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ALLOYS
COOLANTS
HARDENING
INCOLOY 800H
INCONEL 617
IRRADIATION
NEUTRON SPECTRA
NUCLEAR POWER PLANTS
PERFORMANCE
PRESSURE VESSELS
REACTORS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTOR TECHNOLOGY
RESEARCH PROGRAMS
STEELS
TESTING
ULTIMATE STRENGTH