Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

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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 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel ... continued below

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Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J. et al. October 1, 1999.

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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 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

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  • Report No.: ANL-99/11
  • Grant Number: DE-AC02-06CH11357
  • DOI: 10.2172/1012531 | External Link
  • Office of Scientific & Technical Information Report Number: 1012531
  • Archival Resource Key: ark:/67531/metadc840265

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  • October 1, 1999

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  • May 19, 2016, 3:16 p.m.

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  • May 23, 2016, 2:44 p.m.

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Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J. et al. Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998., report, October 1, 1999; United States. (digital.library.unt.edu/ark:/67531/metadc840265/: accessed December 16, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.