Structural integrity of water reactor pressure boundary components. Progress report ending 31 August 1977

Technical Report · Sun Jan 01 04:00:00 EST 1978

OSTI ID:6433258

Loss, F J

This report describes research progress in a continuing program to characterize materials properties performance with respect to structural integrity of light water reactor pressure boundary components. Progress for this reporting period is summarized in the following areas: evaluation of critical factors in crack growth rate studies in a pressurized water reactor environment; irradiation and postirradiation (annealing) heat treatment study of the toughness of pressure vessel steels having a low upper shelf level; exploratory investigation of notch ductility changes in A508-2 forgings; compositional variables affecting upper shelf toughness; and investigation of warm prestress phenomenon as a means to limit crack extension in a vessel during a loss-of-coolant accident. (Author)

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 6433258
• Report Number(s): AD-A-052553; NRL-MR-3700
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ALLOYS
CONTAINERS
CRACKS
DUCTILITY
FRACTURE PROPERTIES
IRON ALLOYS
IRON BASE ALLOYS
MECHANICAL PROPERTIES
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
RADIATION EFFECTS
REACTORS
STEELS
TENSILE PROPERTIES
WATER COOLED REACTORS