The submitted manuscript has been crcatcd by the Unlversity of Chicago
This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use wc, uld not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not ntx_essarilystate or reflect those of the United States Government or any agency thereof. The submitted manuscript has been authored by a contractor of the U. S, Government under contract No. W-31-109-ENG-38. Accordingly, the U. S. Government retains a nonaxclu$ive, royalty-free license to publish or reproduce the published form of thin contribution, or allow others to do 0o, for U. S. Government purpoms. August 1992 l_l_b']i_'i.i _,_,_ [J_ I_ Nr.JV I 19". 2 Presented at the Sixteenth International Symposium on Effects of Radiation on Materials. June 21-25, 1992. Denver. Colorado.
Previous Documents in SeriesEnvironmentally Assisted Cracking in Light Water Reactors Semiannual Report Apd-September 1985, NUREG/CR-4667 Vol. I, ANL-86-31 (June 1986. Reactors Semiannual Report October 1985-March 1986, NUREG/CR-4667 Vol. 11, ANG86-37 (September 1987). Apd-September 1986, NUREG/CR-4667 Vol. 111, ANL-87-37 (September 1987. Environmentally Assisted Cracking in Light Water Environmentally Assisted Cracking i n Light Water Reactors Semiannual Report Environmentally Assisted Cracking in Light Water Reactors Semiannual Report October 1986-March 1987. NUREG/CR-4667 Vol. IV, ANG87-4 1 (December 1987). Apd-September 1987, NUREG/CR-4667 Vol. V, ANL-88-32 (June 1988. October 1987-March 1988, NUREG/CR-4667 Vol. 6, ANG89/10 (August 1989). Apd-September 1988, NUREG/CR-4667 Vol. 7, ANG89/40 (March 1990). Environmentally Assisted Cracking in Light Water Reactors Semiannual Report Environmentally Assisted Cracking in Light Water Reactors Semiannual Report Environmentally Assisted Cracking i n Light Water Reactors Semiannual Report Environmentally Assisted Cracking in AbstractThis report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolved-oxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrobhemical potential on susceptibility to intergranular cracking of high-and commercial-purity Type 304 SS specimens from controlblade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289°C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials. 2.3. 4.5. 6.7.8. 9.10. 11.12.13.14. 15.16. 17. .3 .4 . .6 . .8 . .10 . .12 .13 . Fig. 44(B Executive Summary Fatigue of Fenitic Piping and Pressure Vessel SteelsPlain carbon and low-alloy steels are used extensively in steam supply systems of pressurized and boiling water nuclear reactors (PWRs and BWRs) as piping and pressure vessel materials. Fatigue tests are being conducted on AlOG-Gr B carbon steel and A533-Gr B and A302-Gr B low-alloy steels in water and in air at 288°C to establish the effects of material and loading variables on fatigue life. The results indicate only a marginal effect of lo...
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