Structural integrity of water reactor pressure boundary components. Quarterly progress report Apr-Jun 80

Technical Report · Fri Feb 20 04:00:00 EST 1981

OSTI ID:5915451

Loss, F J

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.

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 5915451
• Report Number(s): NUREG/CR-1783
• Country of Publication: United States
• Language: English

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