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Title: Investigation of the protection potential against IASCC

Abstract

Environmentally induced cracking of components exposed to the radiation and high-temperature aqueous environment of a light water reactor (LWR) is known as irradiation-assisted stress corrosion cracking (IASCC). In boiling water reactors (BWRs) annealed austenitic stainless steels have a threshold level of irradiation damage. Above a fluence level of 5 x 10{sup 20}n/cm{sup 2}, these alloys are susceptible to cracking in the normal BWR core environment. Below this threshold value, laboratory stress corrosion tests of irradiated material have not indicated cracking over a range of corrosion potentials. Recent electrochemical measurements in the core bypass region of several operating BWRs have indicated that the corrosion potentials of Type 304 stainless steel can be as high as 0.250 {+-} 0.025 V (SHE). The potential depends on the specific measurement location. Testing indicates that at high potentials Type 304 stainless steel irradiated above the threshold fluence suffers IASCC, but, if the potential can be decreased, the phenomenon does not occur. An effective environmental modification for eliminating cracking of sensitized Type 304 stainless steel in BWR piping systems is the addition of hydrogen to the reactor feedwater. Hydrogen injection results in a decrease in the electrochemical potential (ECP). Below -0.230 V(SHE), pipe cracking known asmore » intergranular stress corrosion cracking (IGSCC) is effectively mitigated. The identification of a threshold potential for IGSCC was established by performing laboratory stress corrosion tests on thermally sensitized Type 304 stainless steel using the constant extension rate technique (CERT) at controlled ECPs. Similar logic was applied to the experimental program described in this paper.« less

Authors:
;  [1];  [2]
  1. General Electric Co., Pleasanton, CA (United States)
  2. Electric Power Research Institute, Palo Alto, CA (United States)
Publication Date:
OSTI Identifier:
48129
Report Number(s):
CONF-910808-
TRN: 95:011563
Resource Type:
Conference
Resource Relation:
Conference: 5. international symposium on environmental degradation of materials in nuclear power systems - water reactors, Monterey, CA (United States), 25-29 Aug 1991; Other Information: PBD: 1992; Related Information: Is Part Of Proceedings of the fifth international symposium on environmental degradation of materials in nuclear power systems - water reactors; PB: 995 p.
Country of Publication:
United States
Language:
English
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 36 MATERIALS SCIENCE; STAINLESS STEEL-304; STRESS CORROSION; INTERGRANULAR CORROSION; TENSILE PROPERTIES; PHYSICAL RADIATION EFFECTS; BWR TYPE REACTORS; WATER CHEMISTRY; TESTING

Citation Formats

Indig, M E, Wozadlo, G P, and Nelson, J L. Investigation of the protection potential against IASCC. United States: N. p., 1992. Web.
Indig, M E, Wozadlo, G P, & Nelson, J L. Investigation of the protection potential against IASCC. United States.
Indig, M E, Wozadlo, G P, and Nelson, J L. Thu . "Investigation of the protection potential against IASCC". United States.
@article{osti_48129,
title = {Investigation of the protection potential against IASCC},
author = {Indig, M E and Wozadlo, G P and Nelson, J L},
abstractNote = {Environmentally induced cracking of components exposed to the radiation and high-temperature aqueous environment of a light water reactor (LWR) is known as irradiation-assisted stress corrosion cracking (IASCC). In boiling water reactors (BWRs) annealed austenitic stainless steels have a threshold level of irradiation damage. Above a fluence level of 5 x 10{sup 20}n/cm{sup 2}, these alloys are susceptible to cracking in the normal BWR core environment. Below this threshold value, laboratory stress corrosion tests of irradiated material have not indicated cracking over a range of corrosion potentials. Recent electrochemical measurements in the core bypass region of several operating BWRs have indicated that the corrosion potentials of Type 304 stainless steel can be as high as 0.250 {+-} 0.025 V (SHE). The potential depends on the specific measurement location. Testing indicates that at high potentials Type 304 stainless steel irradiated above the threshold fluence suffers IASCC, but, if the potential can be decreased, the phenomenon does not occur. An effective environmental modification for eliminating cracking of sensitized Type 304 stainless steel in BWR piping systems is the addition of hydrogen to the reactor feedwater. Hydrogen injection results in a decrease in the electrochemical potential (ECP). Below -0.230 V(SHE), pipe cracking known as intergranular stress corrosion cracking (IGSCC) is effectively mitigated. The identification of a threshold potential for IGSCC was established by performing laboratory stress corrosion tests on thermally sensitized Type 304 stainless steel using the constant extension rate technique (CERT) at controlled ECPs. Similar logic was applied to the experimental program described in this paper.},
doi = {},
url = {https://www.osti.gov/biblio/48129}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1992},
month = {12}
}

Conference:
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