Structural integrity of water reactor pressure boundary components. Annual report, Fiscal Year 1979

Technical Report · Mon Dec 31 04:00:00 EST 1979

OSTI ID:5653069

Loss, F J ^[1]

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1. ed.

This report describes research progress for Fiscal Year 1979 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 investigations includes the first J-R curves from irradiated A533-B weld deposit. A dynamic finite element analysis was also performed to verify the NRL experimental procedure for dynamic fracture toughness, K/sub Id/. Work in corrosion fatigue has investigated the effects of waveform and temperature on cyclic crack growth in reactor vessel steels; a hydrogen embrittlement model has been proposed. Research in radiation sensitivity has characterized the notch ductility of vessel steels at low fluence. Also investigated was the postirradiation notch ductility of vessel steels in a coordinated IAEA program. The effects of postirradiation annealing and reirradiation are described in terms of Charpy V-notch ductility and J-R curves. In addition, a survey of embrittlement recovery by postirradiation heat treatment has been prepared. Abstracts of reports prepared uner this program in FY 79 are also included.

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 5653069
• Report Number(s): NUREG/CR-1128; NRL-Memo-4122
• Country of Publication: United States
• Language: English

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