Structural Integrity of Water Reactor Pressure Boundary Components.
Abstract
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.
Document Details
- Document Type
- Technical Report
- Publication Date
- Mar 20, 1980
- Accession Number
- ADA084553
Entities
People
- Frank J. Loss
Organizations
- United States Naval Research Laboratory