Structural integrity of water reactor pressure boundary components. Quarterly progress report, April-Jun 1979

Technical Report · Fri Sep 28 04:00:00 EDT 1979

OSTI ID:5576990

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: (a) postirradiation fracture toughness of IAEA coordinated program steels, (b) Charpy V-notch ductility trends of submerged arc weld deposit with cyclic postirradiation annealing and reirradiation treatments, (c) J-R curve characterization of irradiated submerged arc weld deposits, and (d) evaluation of critical factors in cyclic crack growth rate studies in a pressurized water reactor environment.

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 5576990
• Report Number(s): AD-A-075349
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
ALLOYS
ANNEALING
CONTAINERS
COOLING SYSTEMS
CRACKS
CRYSTAL STRUCTURE
DUCTILITY
FATIGUE
FRACTOGRAPHY
FRACTURE PROPERTIES
HEAT TREATMENTS
IRON ALLOYS
IRON BASE ALLOYS
IRRADIATION
JOINTS
LEVELS
MECHANICAL PROPERTIES
MICROSTRUCTURE
PRESSURE VESSELS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
STEELS
TENSILE PROPERTIES
UNDERWATER
WATER COOLED REACTORS
WELDED JOINTS