[Federal Register Volume 73, Number 85 (Thursday, May 1, 2008)]
[Notices]
[Pages 24094-24096]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-9451]
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NUCLEAR REGULATORY COMMISSION
Proposed Generic Communication; Fatigue Analysis of Nuclear Power
Plant Components
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of opportunity for public comment.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
issue a regulatory issue summary (RIS) to inform licensees of an
analysis
[[Page 24095]]
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.
This Federal Register notice is available through the NRC's
Agencywide Documents Access and Management System (ADAMS) under
accession number ML081080562.
DATES: 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.
ADDRESSES: 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.
FOR FURTHER INFORMATION, CONTACT: John R. Fair at 301-415-2759 or by e-
mail at [email protected].
SUPPLEMENTARY INFORMATION:
NRC Regulatory Issue Summary 2008-XX
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-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.
[[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.
Contact
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:
[email protected], 301-415-1913. John R. Fair, NRR, 301-415-2759,
E-mail: [email protected].
Note: The NRC' s generic communications may be found on the NRC
public Web site, http://www.nrc.gov, under Electronic Reading Room/
Document Collections.
End of Draft Regulatory Issue Summary
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, http://www.nrc.gov/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
[email protected].
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.
[FR Doc. E8-9451 Filed 4-30-08; 8:45 am]
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