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Development of the Optimized Fatigue Analysis Procedure to Evaluate Design Life for APR1400

[+] Author Affiliations
Jaegon Lee, Taesoon Kim, Chankook Moon

Korea Hydro & Nuclear Power Co., Ltd., Daejeon, South Korea

Kwanghan Lee

GNEC Inc., Daejeon, South Korea

Paper No. ICONE16-48659, pp. 665-670; 6 pages
doi:10.1115/ICONE16-48659
From:
  • 16th International Conference on Nuclear Engineering
  • Volume 1: Plant Operations, Maintenance, Installations and Life Cycle; Component Reliability and Materials Issues; Advanced Applications of Nuclear Technology; Codes, Standards, Licensing and Regulatory Issues
  • Orlando, Florida, USA, May 11–15, 2008
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 0-7918-4814-0 | eISBN: 0-7918-3820-X
  • Copyright © 2008 by ASME

abstract

Fatigue is one of the most important failure mechanisms to assess integrity and design life of nuclear power plants. Fatigue analysis procedure and the standard fatigue design curve (S-N curve) for the class 1 components are given in ASME code section III NB. However, the existing ASME fatigue design curve does not address the effects of light water reactor coolant environment. The life time of ALWRs is designed for 60 years, and recently the plant life time of currently operating NPPs has been extended 20 years more. If we assess the integrity and design life of major components by fatigue analysis considering environmental factor and S-N curve, the estimated fatigue usage factor will not meet the criterion. In this study, detailed fatigue analysis using three dimensional models were performed to develop the optimized fatigue analysis procedure and their results were compared with other references. The locations considered are the pressurizer surge line, the CVCS charging inlet nozzle and the steam generator economizer nozzle of the advanced power reactor 1400 (APR1400).

Copyright © 2008 by ASME

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