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PRESSURE VESSEL AND PIPING CODES

A Technical Basis for Characterizing Flaws Detected by Preservice and Inservice Examinations of Nuclear Power Plant Components

[+] Author and Article Information
R. R. Maccary

U. S. Nuclear Regulatory Commission, Washington, D.C.

J. Pressure Vessel Technol 97(4), 322-326 (Nov 01, 1975) (5 pages) doi:10.1115/1.3454316 History: Online October 25, 2010

Abstract

The nondestructive examination procedures specified by the rules of construction of the ASME Boiler and Pressure Vessel Code—Section III, “Nuclear Power Plant Components” require techniques whose flaw detection capabilities are well within the practical limits established for acceptable workmanship and quality of fabrication. The rules of the ASME Section XI, “Inservice Inspection of Nuclear Reactor Coolant Systems”, impose an additional series of examinations. Material or fabrication flaws detected during a preservice examination as well as flaws developed during service must be evaluated to establish the acceptability of the component for initial and continued service. These examination requirements have introduced the need to characterize the flaws detected by the examinations and to set “allowable flaw indication standards.” The principles of fracture mechanics provide an engineering tool which predicts the behavior of materials containing flaws under service loadings. These principles form the underlying basis upon which the allowable flaw indication standards of ASME Section XI were formulated. The development of new rules governing flaw indication characterization and allowable flaw indications standards, as specified in the ASME Code, Section XI, are reviewed.

Copyright © 1975 by ASME
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