skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel

Abstract

Austenitic stainless steels (SS) have been used as core component materials for light water reactors. As reactors age, however, the material tends to suffer from degradation primarily resulting from irradiation-assisted stress corrosion cracking (IASCC) as well as intergranular stress corrosion cracking (IGSCC). Neutron-irradiated, thermally sensitized Type 304 (UNS S30400) stainless steels (SS) were examined by slow strain rate (SSR) stress corrosion cracking (SCC) tests in 290 C water of 0.2 ppm dissolved oxygen concentration (DO) and by SSR tensile tests in 290 C inert gas environment. Neutron fluences ranged from 4 x 10{sup 22} n/m{sup 2} to 3 x 10{sup 25} n/m{sup 2} (energy [E] > 1 MeV). percent intergranular (%IG) cracking, which has been used as an intergranular (IG) cracking susceptibility indicator in the SSR SCC tests, changes anomalously with neutron fluence in spite of the strain-to-failure rate decreasing with an increase of neutron fluence. Apparently, %IG is a misleading indicator for the irradiated, thermally sensitized Type 304 SS and for the irradiated, nonsensitized SS when IG cracking susceptibility is compared at different neutron fluences, test temperatures, DO, and strain rates. These test parameters may affect deformation and fracture behaviors of the irradiated SS during the SSR SCC tests,more » resulting in changing %IG, which is given by the ratio of the total IG cracking area to the entire fracture surface area. It is suggested that strain-to-IG crack initiation for the irradiated, thermally sensitized SS and for the irradiated, nonsensitized SS is the alternative indicator in the SSR SCC tests. An engineering expedient to determine the IG crack initiation strain is given by a deviating point on superposed stress-strain curves in inert gas and in oxygenated water. The strain-to-IG crack initiation becomes smaller with an increase of neutron fluence and DO. The SSR tensile tests in inert gas are needed to obtain strain-to-IG crack initiation in addition to the SSR SCC tests in water.« less

Authors:
; ; ;
Publication Date:
Research Org.:
Komae Research Lab., Tokyo (JP)
OSTI Identifier:
20075963
Resource Type:
Journal Article
Journal Name:
Corrosion (Houston)
Additional Journal Information:
Journal Volume: 56; Journal Issue: 5; Other Information: PBD: May 2000; Journal ID: ISSN 0010-9312
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; STRESS CORROSION; TENSILE PROPERTIES; RADIATION EFFECTS; STAINLESS STEEL-304; WATER MODERATED REACTORS; NEUTRONS; STRAIN RATE; CRACKS; TEMPERATURE RANGE 0400-1000 K

Citation Formats

Onchi, T., Hide, K., Mayuzumi, M., and Hoshiya, T. Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel. United States: N. p., 2000. Web. doi:10.5006/1.3280549.
Onchi, T., Hide, K., Mayuzumi, M., & Hoshiya, T. Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel. United States. doi:10.5006/1.3280549.
Onchi, T., Hide, K., Mayuzumi, M., and Hoshiya, T. Mon . "Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel". United States. doi:10.5006/1.3280549.
@article{osti_20075963,
title = {Intergranular stress corrosion cracking susceptibility of neutron-irradiated, thermally sensitized type 304 stainless steel},
author = {Onchi, T. and Hide, K. and Mayuzumi, M. and Hoshiya, T.},
abstractNote = {Austenitic stainless steels (SS) have been used as core component materials for light water reactors. As reactors age, however, the material tends to suffer from degradation primarily resulting from irradiation-assisted stress corrosion cracking (IASCC) as well as intergranular stress corrosion cracking (IGSCC). Neutron-irradiated, thermally sensitized Type 304 (UNS S30400) stainless steels (SS) were examined by slow strain rate (SSR) stress corrosion cracking (SCC) tests in 290 C water of 0.2 ppm dissolved oxygen concentration (DO) and by SSR tensile tests in 290 C inert gas environment. Neutron fluences ranged from 4 x 10{sup 22} n/m{sup 2} to 3 x 10{sup 25} n/m{sup 2} (energy [E] > 1 MeV). percent intergranular (%IG) cracking, which has been used as an intergranular (IG) cracking susceptibility indicator in the SSR SCC tests, changes anomalously with neutron fluence in spite of the strain-to-failure rate decreasing with an increase of neutron fluence. Apparently, %IG is a misleading indicator for the irradiated, thermally sensitized Type 304 SS and for the irradiated, nonsensitized SS when IG cracking susceptibility is compared at different neutron fluences, test temperatures, DO, and strain rates. These test parameters may affect deformation and fracture behaviors of the irradiated SS during the SSR SCC tests, resulting in changing %IG, which is given by the ratio of the total IG cracking area to the entire fracture surface area. It is suggested that strain-to-IG crack initiation for the irradiated, thermally sensitized SS and for the irradiated, nonsensitized SS is the alternative indicator in the SSR SCC tests. An engineering expedient to determine the IG crack initiation strain is given by a deviating point on superposed stress-strain curves in inert gas and in oxygenated water. The strain-to-IG crack initiation becomes smaller with an increase of neutron fluence and DO. The SSR tensile tests in inert gas are needed to obtain strain-to-IG crack initiation in addition to the SSR SCC tests in water.},
doi = {10.5006/1.3280549},
journal = {Corrosion (Houston)},
issn = {0010-9312},
number = 5,
volume = 56,
place = {United States},
year = {2000},
month = {5}
}