Structural integrity of water reactor pressure boundary components. Quarterly progress report Jan-Mar 80

Technical Report · Fri Aug 01 04:00:00 EDT 1980

OSTI ID:6880285

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 describes J-R curve trends from a low shelf A302-B steel and includes a comparison of R curves by the multispecimen and single specimen compliance procedures. Fatigue crack growth rates are being determined for a variety of pressure vessel and piping steels in simulated nuclear coolant environments. Static load cracking in this environment has been observed in bolt-loaded specimens taken from weld heat-affected zones. Work in radiation sensitivity and postirradiation properties recovery has defined tensile property changes under cyclic annealing and reirradiation treatments. Recent progress is described in radiation studies involving reactor vessel steels in a coordinated IAEA program. Also reported are notch ductility tests of reference steels of the NRC light water reactor, pressure vessel irradiation dosimetry program.

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 6880285
• Report Number(s): AD-A-088227
• 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
CHEMICAL REACTIONS
CONTAINERS
CORROSION
CRACK PROPAGATION
FATIGUE
FRACTURE PROPERTIES
IRON ALLOYS
IRON BASE ALLOYS
MECHANICAL PROPERTIES
PRESSURE VESSELS
REACTORS
STEELS
STRESS CORROSION
WATER COOLED REACTORS