Structural integrity of water reactor pressure boundary components. Quarterly progress report Oct-Dec 79

Technical Report · Thu Mar 20 04:00:00 EST 1980

OSTI ID:6885663

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 investigations includes the first J-R curve trends from A533-B weld deposit irradiated under the HSST program. A new experimental procedure was developed for the testing of 0.5T-CT specimens by the single specimen compliance technique. Fatigue crack growth rates are being determined for a variety of pressure vessel and piping steels in simulated nuclear coolant environments. New results of cyclic crack growth on several pressure vessel steels are presented along with the results of the first test of irradiated A533-B steel, tested in the high-temperature pressurized water environment. Work in radiation sensitivity and postirradiation properties recovery has embarked on phase 2 of the irradiation-anneal-reirradiation (IAR) program. Investigations were continued on the postirradiation notch sensitivity of reactor vessel steels in a coordinated IAEA program.

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 6885663
• Report Number(s): AD-A-084553
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
36 MATERIALS SCIENCE
360106 -- Metals & Alloys-- Radiation Effects
ALLOYS
CONTAINERS
CRACK PROPAGATION
FAILURES
FRACTURES
IRON ALLOYS
IRON BASE ALLOYS
MECHANICS
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
RADIATION EFFECTS
REACTORS
STEEL-ASTM-A533
STEELS
WATER COOLED REACTORS