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Nuclear PWR 3-Loop Plants: Reactor Coolant Circuit Strategic Evaluation for Life Management of Primary Circuit Components

[+] Author Affiliations
Georges Bezdikian

EDF, Saint Denis, France

Paper No. PVP2007-26322, pp. 57-66; 10 pages
doi:10.1115/PVP2007-26322
From:
  • ASME 2007 Pressure Vessels and Piping Conference
  • Volume 7: Operations, Applications and Components
  • San Antonio, Texas, USA, July 22–26, 2007
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4285-1 | eISBN: 0-7918-3804-8
  • Copyright © 2007 by ASME

abstract

The French utility has organized the life management program of Nuclear Plants in function of several actions of expertises of knowledge on the long term experience feedback and the maintenance program for life. This program is a strategic stake based on technical point of view considering the aging assessment of the key components on reactor coolant circuit components — elbows, laterals — of the plant, combining the economic aspects, the life management of each components. The actual life evaluation is the results of prediction of life assessment from important program of expertises for the 3-loop PWR and 4-loop PWR plants in operation. For all of inlet and outlet Steam generators elbows and other elbows on coolant circuit, it was assessed the toughness characteristics and prediction to maintain components in operation for 40 years and 60 years. To optimize the strategic in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of replacement of several elbows that 60 years life management will be difficult and the association with steam generators replacements to chose the right period to replace some of them. This assessment is performed considering: • the life evaluation of Steam generators on the plants and alternative maintenance actions, • the large database from cast reactor coolant component expertised after removed from nuclear power plants, • the identification of degradation for different components and prediction criteria proposed, periodic maintenance and volume of expertises.

Copyright © 2007 by ASME

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