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Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel

Abstract

Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the  More>>
Authors:
"NONE"
Publication Date:
Aug 15, 2012
Product Type:
Technical Report
Report Number:
JNES-RE-2012-0001
Resource Relation:
Other Information: 1 ref., 8 figs., 3 tabs.; Related Information: In: Annual safety research report, JFY 2011| Japan Nuclear Energy Safety Organization, Tokyo (Japan)| 604 p.
Subject:
36 MATERIALS SCIENCE; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; CRACK PROPAGATION; CRACKING; DUCTILITY; EVALUATION; IRRADIATION; NEUTRONS; SHROUDS; STAINLESS STEEL-304; STRESS CORROSION
OSTI ID:
22321045
Research Organizations:
Japan Nuclear Energy Safety Organization, Nuclear Energy System Safety Division, Tokyo (Japan)
Country of Origin:
Japan
Language:
Japanese
Other Identifying Numbers:
TRN: JP1500845027247
Availability:
Available from the Internet at URL http://www.nsr.go.jp/archive/jnes/content/000123375.pdf
Submitting Site:
INIS
Size:
page(s) 197-202
Announcement Date:
Apr 06, 2015

Citation Formats

Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel. Japan: N. p., 2012. Web.
Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel. Japan.
2012. "Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel." Japan.
@misc{etde_22321045,
title = {Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel}
abstractNote = {Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)}
place = {Japan}
year = {2012}
month = {Aug}
}