“Aging and Reactor Water Effects on Fatigue Life” W. J. O’Donnell, J. S. Porowski, E. J. Hampton, M. L. Badlani, G. H. Weidenhamer, D. P. Jones, J. S. Abel, and B. Tomkins, presented at International Nuclear Power Plant Aging Symposium, Washington, D. C., August 30, 1988, Proceedings published by USNRC March 1989, NUREG/CP-0100.
Methods of including aging effects and reactor water enhanced crack propagation rates in Codified S-N fatigue life assessment curves are presented and illustrated. Such methods are essential because it is not feasible to produce experimentally based S-N life evaluation curves for all of the relevant cyclic rate and environmental conditions of interest within available finite research funding. Reactor water environmental effects are known to accelerate fatigue crack growth rates in reactor pressure vessel and piping materials.
Recently developed advanced elastic-plastic fracture mechanics technology is used herein as a means of correcting S-N fatigue life evaluation curves for measuring environmental crack growth rates effects. As an important illustration, ASME Code Section XI reactor water crack growth rate curves are used to generate revised new Section III and VIII fatigue design curves for A106 reactor piping. Reactor water effects on the fatigue life are found to be quite significant, and their inclusion in the S-N curves greatly improves the technical basis for assessing the residual component life which meets ASME Code safety margins for cumulative fatigue.
Company President, Bill, Sr. began his career in the Naval Nuclear Program at Westinghouse/ Bettis. He is active on the ASME Subcommittee on Design, and serves as a Contributing Member of the ASME (BPV III) Working Group on Fatigue Strength. He has also co-authored numerous papers on topics including design, fatigue and fracture.
O’Donnell Consulting performs Engineering Design and Analysis to ASME Code including B&PV Section VIII Division 1 and 2 – as well as the Power & Process Piping Codes.
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