Structural Integrity of Water Reactor Pressure Boundary Components.

Abstract

This report describes progress in a continuing program to characterize material properties performance with respect to structural integrity of light water reactor pressure boundary components. Progress under fracture mechanics highlights J-R curve trends from low upper shelf A533-B weld deposits irradiated under the HSST program. Fatigue crack growth rates are being determined for a variety of pressure vessel and piping steels in simulated nuclear coolant environments. Three regions of crack growth behavior which have been associated with classical stress corrosion cracking and corrosion fatigue now have been clearly defined for reactor vessel steels. A theory of the influence of dissolved oxygen content in the fatigue crack growth in simulated PWR coolant is proposed. Work in radiation sensitivity describes recent progress in radiation studies involving reactor vessel steels in a coordinated IAEA program. Also reported is a notch ductility characterization of A508-2 forging steel with irradiation. (Author)

Open PDF

Document Details

Document Type
Technical Report
Publication Date
Feb 20, 1981
Accession Number
ADA095388

Entities

People

  • Frank J. Loss

Organizations

  • United States Naval Research Laboratory

Tags

Communities of Interest

  • Energy and Power Technologies
  • Ground and Sea Platforms

DTIC Thesaurus Topics

  • Chemistry
  • Crack Propagation
  • Cracks
  • Fracture (Mechanics)
  • J Integrals
  • Materials
  • Mechanics
  • Nuclear Energy
  • Nuclear Reactors
  • Plastic Explosives
  • Pressure Vessels
  • Radiation
  • Sine Waves
  • Stress Corrosion
  • Stress Corrosion Cracking
  • Test And Evaluation
  • Waveforms

Readers

  • Materials Science (Mechanical Engineering).
  • Materials Science and Engineering.
  • Thermal Physics or Thermal Science.