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A Review of Fatigue and SCC Crack Growth Rate Relationships for Ferritic and Stainless Steels and Ni-Cr-Fe Materials in BWR Environment

[+] Author Affiliations
Hardayal S. Mehta

GE Energy Nuclear, Sunol, CA

Paper No. PVP2006-ICPVT-11-93853, pp. 175-190; 16 pages
doi:10.1115/PVP2006-ICPVT-11-93853
From:
  • ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference
  • Volume 1: Codes and Standards
  • Vancouver, BC, Canada, July 23–27, 2006
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4752-7 | eISBN: 0-7918-3782-3
  • Copyright © 2006 by ASME

abstract

When in-service inspection of a nuclear plant component reveals the presence of cracking, an engineering evaluation (typically called a justification for continued operation, or JCO) is required to demonstrate the structural suitability for continued operation. A key element in such a flaw evaluation is the projected crack growth over the period when the cracked component will be reinspected. The crack growth is expected to be a combination of stress corrosion cracking (SCC) and corrosion fatigue. The ASME Section XI Code is in the process of developing a full range of SCC and corrosion fatigue crack growth rate relationships (CGRs) for stainless steel and Ni-Cr-Fe materials. The objective of this paper is to summarize several available SCC and fatigue crack growth rate relationships for these materials exposed to boiling water reactor (BWR) water environments. For completeness, low alloy steel SCC and corrosion fatigue CGRs in BWR water environment are also briefly reviewed. Two example evaluations are provided that used some of these CGRs in developing the JCOs for BWR components. A detailed comparison of these CGRs along with a review of the underlying data will be part of a future effort undertaken by the ASME Section XI Task Group.

Copyright © 2006 by ASME

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