Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I [electronic resource].

Published: Rockville, Md. : U.S. Nuclear Regulatory Commission, 2010.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
Additional Creators: Argonne National Laboratory, U.S. Nuclear Regulatory Commission, and United States. Department of Energy. Office of Scientific and Technical Information

Access Online

www.osti.gov

Availability

Finding items...

Restrictions on Access

Free-to-read Unrestricted online access

Summary

Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to ≈25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

Report Numbers

E 1.99:anl/09-32
anl/09-32

Subject(s)

Other Subject(s)

Note

Published through SciTech Connect.
02/16/2010.
"anl/09-32"
Chen, Y.; Yang, Y.; Allen, T. R.; Chopra, O. K.; Shack, W. J.; Soppet, W. K.; Univ. of Wisconsin at Madison.

Funding Information

DE-AC02-06CH11357

View MARC record | catkey: 13801918