Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

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This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to ... continued below

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107 p.

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Chopra, O. K.; Chung, H. M. & Gavenda, D. J. October 1997.

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Description

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

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107 p.

Notes

INIS; OSTI as TI98000789

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  • Other Information: PBD: Oct 1997

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  • Other: TI98000789
  • Report No.: NUREG/CR--4667-Vol.23
  • Report No.: ANL--97/10
  • Grant Number: W-31109-ENG-38
  • DOI: 10.2172/541938 | External Link
  • Office of Scientific & Technical Information Report Number: 541938
  • Archival Resource Key: ark:/67531/metadc689607

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  • October 1997

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  • Aug. 14, 2015, 8:43 a.m.

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  • April 8, 2016, 8:05 p.m.

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Chopra, O. K.; Chung, H. M. & Gavenda, D. J. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996, report, October 1997; Washington D.C.. (digital.library.unt.edu/ark:/67531/metadc689607/: accessed August 22, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.