Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24 | Zendy

O.K. Chopra | Zendy; H.M. Chung | Zendy; E.E. Gruber | Zendy

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Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24

Author(s) -

O.K. Chopra,

H.M. Chung,

E.E. Gruber

Publication year - 1998

Publication title -

osti oai (u.s. department of energy office of scientific and technical information)

Language(s) - English

Resource type - Reports

DOI - 10.2172/595659

Subject(s) - stress corrosion cracking , materials science , austenitic stainless steel , boiling water reactor , light water reactor , piping , cracking , carbon steel , metallurgy , ultimate tensile strength , pressurized water reactor , boiling , alloy , volume (thermodynamics) , austenite , strain rate , corrosion , nuclear engineering , composite material , environmental science , microstructure , chemistry , physics , organic chemistry , quantum mechanics , environmental engineering , engineering

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. 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 Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (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 several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments

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