ML082340969
| ML082340969 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/11/2008 |
| From: | Michael Case 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) | |
See also: RIS 2008-10
Text
NEC-JH_23
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001
April 11, 2008
NRC REGULATORY ISSUE SUMMARY 2008-10
FATIGUE ANALYSIS OF NUCLEAR POWER PLANT COMPONENTS
ADDRESSEES
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.
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 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-1 800, 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.
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August 12, 2008 (11:00am)
OFFICE OF SECRETARY
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Page 2 of 4
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 per-forming 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
ASMVE 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 per-form 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.
Page 3 of 4
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
A notice of opportunity for public comment on this RIS was not published in the Federal Register
because the RIS is informational.
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.
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.
.1
Page 4 of 4
CONTACT
Please direct any questions about this matter to the technical contacts listed below.
IRA/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contacts:
Kenneth C. Chang, NRR
301-415-1913
E-mail: kxc2t)Nrc..qov
John R. Fair, NRR
301-415-2759
E-mail: irf(@nrc.qov
Note: The NRC's generic communications may be found on the NRC public Web site,
http://www.nrc.qov, under Electronic Reading Room/Document Collections.