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Application of Aging Management Strategies for Reactor Vessel Internals and Core Support Structures

[+] Author Affiliations
Tamás R. Liszkai, Matthew Snyder, Steve Fyfitch, Hongqing Xu, Hasan Charkas

AREVA NP Inc., Lynchburg, VA

Paper No. PVP2010-26137, pp. 277-286; 10 pages
doi:10.1115/PVP2010-26137
From:
  • ASME 2010 Pressure Vessels and Piping Division/K-PVP Conference
  • ASME 2010 Pressure Vessels and Piping Conference: Volume 7
  • Bellevue, Washington, USA, July 18–22, 2010
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 978-0-7918-4926-2 | eISBN: 978-0-7918-3878-5
  • Copyright © 2010 by ASME

abstract

The Materials Reliability Program (MRP) Reactor Internals Focus Group (RI-FG) developed Pressurized Water Reactor (PWR) Internals Inspection and Evaluation (I&E) Guidelines under the sponsorship of the Electric Power Research Institute (EPRI). The I&E guidelines summarized in MRP-227 [1], provide a generic basis for U.S. utilities to develop their Aging Management Program (AMP) for managing the long-term aging degradation of PWR reactor internals including the existing and extended license periods. A number of internals structural bolts in the Babcock & Wilcox (B&W) design PWRs are fabricated from high-strength alloys such as Alloy A-286 or Alloy X-750. The materials in general, and bolts in particular, are known to be susceptible to stress corrosion cracking (SCC) based on past operating experience. The Upper and Lower Core Barrel (UCB and LCB) bolts have a core support function and have been generically categorized as Primary components for inspection in the I&E Guidelines. The remaining Alloy A-286 and Alloy X-750 structural bolts are in the Expansion category. Per 10CFR54, all U.S. PWRs are required to establish a unit-specific AMP for the extended license period in accordance with the ten elements of an effective AMP outlined in the Generic Aging Lessons Learned (GALL, NUREG-1801 Rev. 01, [2]) report published by the U.S. Nuclear Regulatory Commission (NRC). The goal of this paper is to provide an overview of the work performed by AREVA NP Inc. to support the development of the MRP I&E guidelines and unit-specific AMP for UCB and LCB bolts. A review of Alloy A-286 and Alloy X-750 bolts in the B&W design PWR is provided including the degradation mechanism, operating and inspection experience, replacement, and autoclave and in-reactor test results. The latest UT inspection technique used to characterize the extent of flaws is also discussed. Acceptance criteria for evaluating degraded conditions in UCB and LCB bolts were developed in accordance with the requirements of the ASME Section III, Subsection-NG core support structures requirements. In addition to Code compliance, special limits were established to limit the change in the core support structure stiffness. The acceptance criteria enable utilities to rapidly disposition UT inspection findings during an outage within 48 hours. In order to support the objectives of an efficient AMP for the UCB and LCB bolts, three-dimensional finite element models were prepared capable of evaluating all potential failure scenarios. These models enable accurate representation of flange flexibility and redistribution of loads due to deficient bolts. Prior to an outage, hypothetical patterns of bolt failures could be evaluated to support pre-outage planning and contingency preparation. During an outage, these models are used to disposition inspection results and help operability assessment of continued operation, and re-inspection requirement to ensure continued safety and integrity of the reactor vessel internals. Based on the existing work performed, future improvement and expansion of analytical capability is outlined in the last section of this paper. In conclusion, AREVA NP Inc. has demonstrated an effective use of a multi-disciplined approach using structural analyses, operating experience, material evaluations, and non-destructive examination (NDE) to fulfill both the development and implementation of unit-specific aging management commitments as required by MRP-227 for the current and extended license periods.

Copyright © 2010 by ASME
Topics: Reactor vessels

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