RIS 2008-10, 2008/04/11-Vermont Yankee July 2008 Evidentiary Hearing - Intervenor Exhibit NEC-JH 23, Nrc, NRC Regulatory Issue Summary 2008-10 Fatigue Analysis of Nuclear Power Plan Components (April 11, 2008)

2008/04/11-Vermont Yankee July 2008 Evidentiary Hearing - Intervenor Exhibit NEC-JH_23, Nrc, NRC Regulatory Issue Summary 2008-10 Fatigue Analysis of Nuclear Power Plan Components (April 11, 2008)

ML082340969
Person / Time
Site:	File:NorthStar Vermont Yankee icon.png
Issue date:	04/11/2008
Revision:	0
From:	Case M J Office of Nuclear Reactor Regulation
To:	NRC/SECY/RAS
SECY RAS
References
06-849-03-LR, 50-271-LR, Entergy-Intervenor-NEC-JH_23, RAS M-195 RIS-08-010
Download: ML082340969 (4)

NEC-JH_23 April 11, 2008NRC REGULATORY ISSUE SUMMARY 2008-10FATIGUE ANALYSIS OF NUCLEAR POWER PLANT COMPONENTS

ADDRESSEES

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

INTENT

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 theAmerican 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 ofOperating Licenses for Nuclear Power Plants," requires that applicants for license renewalperform an evaluation of time-limited aging analyses relevant to structures, systems, andcomponents within the scope of license renewal. The fatigue analysis of the reactor coolantpressure 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 ofLicense Renewal Applications for Nuclear Power Plants," issued September 2005.NUREG-1 800, Rev. 1, specifies that the effects of the reactor water environment on fatigue lifebe evaluated for a sample of components to provide assurance that cracking because of fatiguewill not occur during the period of extended operation. Since the reactor water environment hasa significant impact on the fatigue life of components, many license renewal applicants haveperformed supplemental detailed analyses to demonstrate acceptable fatigue life for thesecomponents.10 CFR 50.55a, "Codes and Standards," specifies the ASME Code requirements for operatingreactors. Some operating facilities may have performed supplemental detailed analysis ofcomponents because of new loading conditions identified after the plant began operation.IML08DOogPL5USNRCAugust 12, 2008 (11:00am)OFFICE OF SECRETARYRULEMAKINGS ANDADJUDICATIONS STAFFOFFEREDby: ApplicantI/censeeclnporNR9Sta -ItIENTIFIE on261 .Acdw~~ahm REJECTED NPW4-07-pý,J"-,o2g-'

SUMMARY OF ISSUE

The staff identified a concern regarding the methodology used by some license renewalapplicants to demonstrate the ability of nuclear power plant components to withstand the cyclicloads associated with plant transient operations for the period of extended operation. Thisparticular analysis methodology involves the use of the Green's function to calculate the fatigueusage during plant transient operations such as startups and shutdowns.The Green's function approach involves per-forming a detailed stress analysis of a component tocalculate its response to a step change in temperature. This detailed analysis is used toestablish an influence function, which is subsequently used to calculate the stresses caused bythe actual plant temperature transients. This methodology has been used to perform fatiguecalculations and as input for on-line fatigue monitoring programs. The Green's functionmethodology is not in question. The concern involves a simplified input for applying the Green'sfunction 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 inASMVE Code,Section III, Subsection NB, Subarticle NB-3200. Simplification of the analysis toconsider 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 simplificationstill provides a conservative result.The staff has requested that recent license renewal applicants that have used this simplifiedGreen's function methodology per-form confirmatory analyses to demonstrate that the simplifiedGreen's function analyses provide acceptable results. The confirmatory analyses retain all sixstress components. To date, the confirmatory analysis of one component, a boiling-waterreactor feedwater nozzle, indicated that the simplified input for the Green's function did notproduce conservative results in the nozzle bore area when compared to the detailed analysis.However, the confirmatory analysis still demonstrated that the nozzle had acceptable fatigueusage.Licensees may have also used the simplified Green's function methodology in operating plantfatigue evaluations for the current license term. For plants with renewed licenses, the staff isconsidering additional regulatory actions if the simplified Green's function methodology wasuse

BACKFIT DISCUSSION

This RIS informs addressees of a potential nonconservative calculation methodology andreminds them that the ASME Code fatigue analysis should be performed properly. For licenserenewal, metal fatigue is evaluated as a time-limited aging analysis in accordance with10 CFR 54.21(c). The associated staff review guidance appears in Section 4.3, "Metal FatigueAnalysis," of NUREG-1800, Rev. 1. For operating reactors, the ASME Code requirementsappear 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

A notice of opportunity for public comment on this RIS was not published in the Federal Registerbecause the RIS is informational.

CONGRESSIONAL REVIEW ACT

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

PAPERWORK REDUCTION ACT STATEMENT

This RIS does not contain information collection requirements that are subject to therequirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).Public Protection NotificationThe NRC may not conduct or sponsor, and a person is not required to respond to, a request forinformation or an information collection requirement unless the requesting document displays acurrently valid Office of Management and Budget control number..1

CONTACT

Please direct any questions about this matter to the technical contacts listed below.IRA/Michael J. Case, DirectorDivision of Policy and RulemakingOffice of Nuclear Reactor RegulationTechnical Contacts: Kenneth C. Chang, NRR301-415-1913E-mail: kxc2t)Nrc..qovJohn R. Fair, NRR301-415-2759E-mail: irf(@nrc.qovNote: The NRC's generic communications may be found on the NRC public Web site,http://www.nrc.qov, under Electronic Reading Room/Document Collections.