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Proposed Generic Communication; Fatigue Analysis of Nuclear Power Plant Components

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Nuclear Regulatory Commission.


Notice of opportunity for public comment.


The U.S. Nuclear Regulatory Commission (NRC) is proposing to issue a regulatory issue summary (RIS) to inform licensees of an analysis Start Printed Page 24095methodology used to demonstrate compliance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) fatigue acceptance criteria that could be nonconservative if not correctly applied.

This Federal Register notice is available through the NRC's Agencywide Documents Access and Management System (ADAMS) under accession number ML081080562.


Comment period expires June 16, 2008. Comments submitted after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except for comments received on or before this date.


Submit written comments to the Chief, Rulemaking, Directives and Editing Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Mail Stop T6-D59, Washington, DC 20555-0001, and cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to NRC Headquarters, 11545 Rockville Pike (Room T-6D59), Rockville, Maryland, between 7:30 a.m. and 4:15 p.m. on Federal workdays.

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John R. Fair at 301-415-2759 or by e-mail at

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NRC Regulatory Issue Summary 2008-XX

Fatigue Analysis of Nuclear Power Plant Components


All holders of operating licenses for nuclear power reactors, except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.


The U.S. Nuclear Regulatory Commission (NRC) is issuing this regulatory issue summary (RIS) to inform licensees of an analysis methodology used to demonstrate compliance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) fatigue acceptance criteria that could be nonconservative if not correctly applied.

Background Information

Title 10 of the Code of Federal Regulations (10 CFR) Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” requires that applicants for license renewal perform an evaluation of time-limited aging analyses relevant to structures, systems, and components within the scope of license renewal. The fatigue analysis of the reactor coolant pressure boundary components is an issue that involves time-limited assumptions. In addition, the staff has provided guidance in NUREG-1800, Rev. 1, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants,” issued September 2005. NUREG-1800, Rev. 1, specifies that the effects of the reactor water environment on fatigue life be evaluated for a sample of components to provide assurance that cracking because of fatigue will not occur during the period of extended operation. Since the reactor water environment has a significant impact on the fatigue life of components, many license renewal applicants have performed supplemental detailed analyses to demonstrate acceptable fatigue life for these components.

10 CFR 50.55a, “Codes and Standards,” specifies the ASME Code requirements for operating reactors. Some operating facilities may have performed supplemental detailed analysis of components because of new loading conditions identified after the plant began operation.

Summary of Issue

The staff identified a concern regarding the methodology used by some license renewal applicants to demonstrate the ability of nuclear power plant components to withstand the cyclic loads associated with plant transient operations for the period of extended operation. This particular analysis methodology involves the use of the Green's function to calculate the fatigue usage during plant transient operations such as startups and shutdowns.

The Green's function approach involves performing a detailed stress analysis of a component to calculate its response to a step change in temperature. This detailed analysis is used to establish an influence function, which is subsequently used to calculate the stresses caused by the actual plant temperature transients. This methodology has been used to perform fatigue calculations and as input for on-line fatigue monitoring programs. The Green's function methodology is not in question. The concern involves a simplified input for applying the Green's function in which only one value of stress is used for the evaluation of the actual plant transients. The detailed stress analysis requires consideration of six stress components, as discussed in ASME Code, Section III, Subsection NB, Subarticle NB-3200. Simplification of the analysis to consider only one value of the stress may provide acceptable results for some applications; however, it also requires a great deal of judgment by the analyst to ensure that the simplification still provides a conservative result.

The staff has requested that recent license renewal applicants that have used this simplified Green's function methodology perform confirmatory analyses to demonstrate that the simplified Green's function analyses provide acceptable results. The confirmatory analyses retain all six stress components. To date, the confirmatory analysis of one component, a boiling-water reactor feedwater nozzle, indicated that the simplified input for the Green's function did not produce conservative results in the nozzle bore area when compared to the detailed analysis. However, the confirmatory analysis still demonstrated that the nozzle had acceptable fatigue usage.

Licensees may have also used the simplified Green's function methodology in operating plant fatigue evaluations for the current license term. For plants with renewed licenses, the staff is considering additional regulatory actions if the simplified Green's function methodology was used.

Backfit Discussion

This RIS informs addressees of a potential nonconservative calculation methodology and reminds them that the ASME Code fatigue analysis should be performed properly. For license renewal, metal fatigue is evaluated as a time-limited aging analysis in accordance with 10 CFR 54.21(c). The associated staff review guidance appears in Section 4.3, “Metal Fatigue Analysis,” of NUREG-1800, Rev. 1. For operating reactors, the ASME Code requirements appear in 10 CFR 50.55a. This RIS does not impose a new or different regulatory staff position. It requires no action or written response and, therefore, is not a backfit under 10 CFR 50.109, “Backfitting.” Consequently, the NRC staff did not perform a backfit analysis.

Federal Register Notification

To be done after the public comment period.

Congressional Review Act

The NRC has determined that this RIS is not a rule as designated by the Congressional Review Act (5 U.S.C. 801-808) and, therefore, is not subject to the Act. Start Printed Page 24096

Paperwork Reduction Act Statement

This RIS does not contain information collection requirements that are subject to the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid Office of Management and Budget control number.


Please direct any questions about this matter to the technical contacts listed below: Michael J. Case, Director, Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation.

Technical Contacts: Kenneth C. Chang, NRR, E-mail:, 301-415-1913. John R. Fair, NRR, 301-415-2759, E-mail:


The NRC' s generic communications may be found on the NRC public Web site,, under Electronic Reading Room/Document Collections.

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Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible electronically from the Agencywide Documents Access and Management System (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site,​NRC/​ADAMS/​index.html. If you do not have access to ADAMS or if you have problems in accessing the documents in ADAMS, contact the NRC Public Document Room (PDR) reference staff at 1-800-397-4209 or 301-415-4737 or by e-mail to

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Dated at Rockville, Maryland, this 23rd day of April 2008.

For the Nuclear Regulatory Commission.

Martin C. Murphy,

Chief, Generic Communications Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation.

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[FR Doc. E8-9451 Filed 4-30-08; 8:45 am]