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.
Skip Nav Destination
Article navigation
November 1975
Pressure Vessel And Piping Codes
A Technical Basis for Characterizing Flaws Detected by Preservice and Inservice Examinations of Nuclear Power Plant Components
R. R. Maccary
R. R. Maccary
U. S. Nuclear Regulatory Commission, Washington, D.C.
Search for other works by this author on:
R. R. Maccary
U. S. Nuclear Regulatory Commission, Washington, D.C.
J. Pressure Vessel Technol. Nov 1975, 97(4): 322-326 (5 pages)
Published Online: November 1, 1975
Article history
Online:
October 25, 2010
Citation
Maccary, R. R. (November 1, 1975). "A Technical Basis for Characterizing Flaws Detected by Preservice and Inservice Examinations of Nuclear Power Plant Components." ASME. J. Pressure Vessel Technol. November 1975; 97(4): 322–326. https://doi.org/10.1115/1.3454316
Download citation file:
Get Email Alerts
Cited By
Lateral and Transverse Stiffness Requirements for Supports of Pipeline Systems Conveying Fluids
J. Pressure Vessel Technol
Master Curve Evaluation Using the Fracture Toughness Data at Low Temperature of T − T0 < −50 °C
J. Pressure Vessel Technol (June 2025)
Investigation of the Dynamic Response of Soil/Steel Composite Structures of Vacuum Explosion Containment Vessels
J. Pressure Vessel Technol (June 2025)
Related Articles
ASME Boiler and Pressure Vessel Code Roadmap for Compact Heat Exchangers in High Temperature Reactors
ASME J of Nuclear Rad Sci (October,2020)
Overview of the Impact of Ultrasonic Examination Performance Demonstration on the ASME Boiler and Pressure Vessel Code
J. Pressure Vessel Technol (August,2002)
Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation
J. Pressure Vessel Technol (February,2001)
An Ability to Adapt and Change
Mechanical Engineering (November,2014)
Related Proceedings Papers
Related Chapters
Subsection NF—Supports
Companion Guide to the ASME Boiler & Pressure Vessel Codes, Volume 1 Sixth Edition
Part 2, Section II—Materials and Specifications
Companion Guide to the ASME Boiler & Pressure Vessel Code, Volume 1, Second Edition
Part 2, Section II—Materials and Specifications
Companion Guide to the ASME Boiler and Pressure Vessel Code, Volume 1, Third Edition