Effects of LWR coolant environments on fatigue S-N curves for carbon and low-alloy steels Metadata

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Title

  • Main Title Effects of LWR coolant environments on fatigue S-N curves for carbon and low-alloy steels

Creator

  • Author: Chopra, O. K.
    Creator Type: Personal
  • Author: Shack, W. J.
    Creator Type: Personal

Contributor

  • Sponsor: U.S. Nuclear Regulatory Commission
    Contributor Type: Organization
    Contributor Info: Nuclear Regulatory Commission, Washington, DC (United States)

Publisher

  • Name: Argonne National Laboratory
    Place of Publication: Illinois
    Additional Info: Argonne National Lab., IL (United States)

Date

  • Creation: 1996-06-01

Language

  • English

Description

  • Content Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figure I-90 of Appendix I to Section III of the Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data indicate significant decreases in fatigue lives of carbon and low-alloy steels in LWR environments when five conditions are satisfied simultaneously: applied strain range, temperature, dissolved oxygen in the water, and S content of the steel are above minimum threshold levels, and loading strain rate is below a threshold value. Only moderate decrease in fatigue life is observed when any one of these conditions is not satisfied. This paper presents several methods that have been proposed for evaluating the effects of LWR coolant environments on fatigue S-N curves for carbon and low-alloy steels. Estimations of fatigue lives under actual loading histories are discussed.
  • Physical Description: 15 p.

Subject

  • Keyword: Crack Propagation
  • Keyword: Carbon Steels
  • STI Subject Categories: 36 Materials Science
  • Keyword: Oxidation
  • Keyword: Coolants
  • Keyword: Reactor Cooling Systems
  • Keyword: Temperature Dependence
  • Keyword: Low Alloy Steels
  • Keyword: Pressure Vessels
  • Keyword: Reactor Materials
  • STI Subject Categories: 22 Nuclear Reactor Technology
  • Keyword: Statistical Models
  • Keyword: Fatigue
  • Keyword: Nuclear Power Plants
  • Keyword: Strains
  • Keyword: Diagrams

Source

  • Conference: American Society of Mechanical Engineers (ASME) pressure vessels and piping conference, Montreal (Canada), 21-26 Jul 1996

Collection

  • Name: Office of Scientific & Technical Information Technical Reports
    Code: OSTI

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Article

Format

  • Text

Identifier

  • Other: DE96011156
  • Report No.: ANL/ET/CP--89687
  • Report No.: CONF-960706--16
  • Grant Number: W-31-109-ENG-38
  • Office of Scientific & Technical Information Report Number: 267512
  • Archival Resource Key: ark:/67531/metadc666519

Note

  • Display Note: INIS; OSTI as DE96011156
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