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Aging Management Strategies for Pressurized Water Reactor Vessel Internals

[+] Author Affiliations
Timothy J. Griesbach

ATI Consulting, Dublin, CA

Robert E. Nickell

Consultant, Poway, CA

H. T. Tang

EPRI, Palo Alto, CA

Jeff D. Gilreath, III

Duke Energy Company, Charlotte, NC

Paper No. PVP2004-3055, pp. 37-41; 5 pages
doi:10.1115/PVP2004-3055
From:
  • ASME/JSME 2004 Pressure Vessels and Piping Conference
  • Storage Tank Integrity and Materials Evaluation
  • San Diego, California, USA, July 25–29, 2004
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4685-7
  • Copyright © 2004 by ASME

abstract

Management of materials aging effects, such as loss of material, reduction in fracture toughness, or cracking, depends upon the demonstrated capability to detect, evaluate, and potentially correct conditions that could affect function of the internals during the license renewal term. License renewal applicants in their submittals to NRC have identified the general elements of aging management programs for Pressurized Water Reactor (PWR) internals, including the use of inservice inspection and monitoring with the possibility of enhancement or augmentation if a relevant condition is discovered. As plants near the license renewal term, plant-specific aging management programs will be implemented focusing on those regions most susceptible to aging degradation. A framework for the implementation of an aging management program is proposed in this paper. This proposed framework is based on current available research results and state of knowledge and utilizes inspections and flaw tolerance evaluations to manage the degradation issues. The important elements of this framework include: • The screening of components for susceptibility to the aging mechanisms, • Performing functionality analyses of the components with representative material toughness properties under PWR conditions, • Evaluating flaw tolerance of lead components or regions of greatest susceptibility to cracking, loss of toughness, or swelling, and • Using focused inspections to demonstrate no loss of integrity in the lead components or regions of the vessel internals. The EPRI Material Reliability Program (MRP) Reactor Internals Issue Task Group (RI-ITG) is actively working to develop the data and methods to quantify an understanding of aging and potential degradation of reactor vessel internals, to develop materials/components performance criteria, and to provide utilities tools for extending plant operations. Under this MRP Program, the technical basis for the framework will be documented. Then, based on that technical basis, PWR internals inspection and flaw evaluation guidelines will be developed for plants to manage reactor internals aging and associated potential degradation.

Copyright © 2004 by ASME

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