Identification of PWR Stainless Steel Piping Safety Significant Locations Susceptible to Stress Corrosion Cracking

  • R. HoslerEmail author
  • A. Kulp
  • P. Stevenson
  • S. Petro
Conference paper
Part of the The Minerals, Metals & Materials Series book series (MMMS)


Stress corrosion cracking (SCC) of stainless steel was originally considered only an issue with boiling water reactors (BWRs), but operating experience has shown that this phenomenon also occurs in pressurized water reactors (PWRs), such as in off-chemistry locations of stagnant branch connection piping. In this paper, the safety significant stainless steel piping locations susceptible to SCC are identified for three representative PWR plants (Plant A [Babcock and Wilcox-designed], Plant B [Westinghouse-designed], and Plant C [Combustion Engineering-designed]). For the purpose of this paper, “safety significant” is defined as having a high consequence of failure as determined by the plant’s risk-informed in-service inspection (RI-ISI) program. Weld locations are considered susceptible to SCC when the water is stagnant and ≥200 °F during steady-state reactor operation. The results of this work will be used to develop guidance for selection of welds to inspect when addressing currently existing inspection requirements.


SCC susceptibility Stainless steel Primary piping 


  1. 1.
    NRC Information Notice 2011–04, Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking In Stainless Steel Piping In Pressurized Water Reactors. (ML103410363)Google Scholar
  2. 2.
    PWR Owners Group Materials Subcommittee Interim Strategy for Identifying Outside Diameter Initiated Stress Corrosion Cracking (ODSCC) of Stainless Steel Systems, NEI 03–08 Good Practice RecommendationGoogle Scholar
  3. 3.
    R. Hosler, S. Fyfitch, H. Malikowski, and G. Ilevbare, Review of Stress Corrosion Cracking of Pressure Boundary Stainless Steel in Pressurized Water Reactors and the Need for Long-term Industry Guidance, In 16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2013Google Scholar
  4. 4.
    Materials Reliability Program, Stress Corrosion Cracking of Stainless Steel Components in Primary Water Circuit Environments of Pressurized Water Reactors (MRP-236) (Palo Alto, CA, EPRI, 2007), p. 1015540Google Scholar
  5. 5.
    Materials Reliability Program, Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines—Supplemental Guidance (MRP-146S) (Palo Alto, CA, EPRI, 2009), p. 1018330Google Scholar

Copyright information

© The Minerals, Metals & Materials Society 2019

Authors and Affiliations

  1. 1.AREVAParisFrance
  2. 2.WestinghouseMonroevilleUSA
  3. 3.AEPColumbusUSA

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