This paper describes industry programs to manage structural degradation and to justify continued operation of nuclear components when unexpected degradation has been encountered due to design materials and/or operational problems. Other issues have been related to operation of components beyond their original design life in cases where there is no evidence of fatigue crack initiation or other forms of structural degradation. Data from plant operating experience have been applied in combination with inservice inspections and degradation management programs to ensure that the degradation mechanisms do not adversely impact plant safety. Probabilistic fracture mechanics calculations are presented to demonstrate how component failure probabilities can be managed through augmented inservice inspection programs.
Skip Nav Destination
Article navigation
February 2001
Technical Papers
Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation
Stephen R. Gosselin
Stephen R. Gosselin
Search for other works by this author on:
Fredric A. Simonen
Stephen R. Gosselin
Contributed by the Pressure Vessels and Piping Division for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received by the PVP Division, January 2000; revised manuscript received October 23, 2000. Editor: S. Y. Zamrik.
J. Pressure Vessel Technol. Feb 2001, 123(1): 58-64 (7 pages)
Published Online: October 23, 2000
Article history
Received:
January 1, 2000
Revised:
October 23, 2000
Citation
Simonen, F. A., and Gosselin, S. R. (October 23, 2000). "Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation ." ASME. J. Pressure Vessel Technol. February 2001; 123(1): 58–64. https://doi.org/10.1115/1.1344237
Download citation file:
Get Email Alerts
Cited By
Experimental Research on Thermal-Oxidative Aging Performance of Polyethylene Pipe Under Hydrostatic Pressure
J. Pressure Vessel Technol
The upper bound of the buckling stress of axially compressed carbon steel circular cylindrical shells
J. Pressure Vessel Technol
Dynamics Modeling and Analysis of Small-Diameter Pipeline Inspection Gauge during Passing Through Elbow
J. Pressure Vessel Technol
Prestressing Estimation for Multilayer Clamping High Pressure Vessel Laminates
J. Pressure Vessel Technol (October 2024)
Related Articles
Japanese Activities Concerning Nuclear Codes and Standards—Part II
J. Pressure Vessel Technol (February,2006)
Guest Editorial
J. Pressure Vessel Technol (February,2004)
Japanese Activities Concerning Nuclear Codes and Standards—Part I
J. Pressure Vessel Technol (February,2006)
Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics (PVP2010-25195)
J. Pressure Vessel Technol (February,2013)
Related Proceedings Papers
Related Chapters
Iwe and Iwl
Companion Guide to the ASME Boiler and Pressure Vessel Code, Volume 2, Third Edition
Section XI: Rules for Inservice Inspection and Tests of Nuclear Power Plant Components
Online Companion Guide to the ASME Boiler & Pressure Vessel Codes
Overview of Section XI Stipulations
Companion Guide to the ASME Boiler and Pressure Vessel Code, Volume 2, Third Edition