[Federal Register Volume 76, Number 190 (Friday, September 30, 2011)]
[Notices]
[Pages 60939-60941]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-25242]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0229]


Metal Fatigue Analysis Performed by Computer Software

AGENCY: Nuclear Regulatory Commission.

ACTION: Regulatory issue summary; request for comment.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to 
issue a regulatory issue summary (RIS) to remind its addressees of the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (ASME Code) requirements in accordance with Title 10 of the 
Code of Federal Regulations (10 CFR) 50.55a, ``Codes and Standards,'' 
and of the quality assurance (QA) requirements for design control in 
accordance with Appendix B, ``Quality Assurance Criteria for Nuclear 
Power Plants and Fuel Reprocessing Plants,'' to 10 CFR Part 50. 
Specifically, this RIS informs addressees of the NRC's findings from 
license renewal and new reactor audits on applicants' analyses and 
methodologies using the computer software package, 
WESTEMSTM, to demonstrate compliance with Section III, 
``Rules for Construction of Nuclear Facility Components,'' of the ASME 
Code.

DATES: Submit comments by October 31, 2011. Comments received after 
this date will be considered if it is practical to do so, but the NRC 
is able to assure consideration only for comments received on or before 
this date.

ADDRESSES: Please include Docket ID NRC-2011-0229 in the subject line 
of your comments. For additional instructions on submitting comments 
and instructions on accessing documents related to this action, see 
``Submitting Comments and Accessing Information'' in the SUPPLEMENTARY 
INFORMATION section of this document. You may submit comments by any 
one of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-
2011-0229. Address questions about NRC dockets to Carol Gallagher, 
telephone: 301-492-3668; e-mail: [email protected].
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.

SUPPLEMENTARY INFORMATION: 

Submitting Comments and Accessing Information

    Comments submitted in writing or in electronic form will be posted 
on the NRC Web site and on the Federal rulemaking Web site, http://www.regulations.gov. Because your comments will not be edited to remove 
any identifying or contact information, the NRC cautions you against 
including any information in your submission that you do not want to be 
publicly disclosed.
    The NRC requests that any party soliciting or aggregating comments 
received from other persons for submission to the NRC inform those 
persons that the NRC will not edit their comments to remove any 
identifying or contact information, and therefore, they should not 
include any information in their comments that they do not want 
publicly disclosed.
    You can access publicly available documents related to this 
document using the following methods:
     NRC's Public Document Room (PDR): The public may examine 
and have copied, for a fee, publicly available documents at the NRC's 
PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland 20852.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): Publicly available documents created or received at the NRC 
are available online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain entry into ADAMS, 
which provides text and image files of the NRC's public documents. If 
you do not have access to ADAMS or if there are problems in accessing 
the documents located in ADAMS, contact the NRC's PDR reference staff 
at 1-800-397-4209, 301-415-4737, or by e-mail to [email protected]. 
The draft RIS is available electronically under ADAMS Accession Number 
ML11252A520.
     Federal Rulemaking Web site: Public comments and 
supporting materials related to this notice can be found at http://www.regulations.gov by

[[Page 60940]]

searching on Docket ID NRC-2011-0229.

FOR FURTHER INFORMATION CONTACT: On Yee, Office of Nuclear Reactor 
Regulation, Division of License Renewal, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, telephone: 301-415-1905, e-mail: 
[email protected].

Draft NRC Regulatory Issue Summary 2011-Xxxx; Metal Fatigue Analysis 
Performed by Computer Software

Addressees

    All holders of, and applicants for, a power reactor operating 
license or construction permit under Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50, ``Domestic Licensing of Production and 
Utilization Facilities,'' except those that have permanently ceased 
operations and have certified that fuel has been permanently removed 
from the reactor vessel.
    All holders of, and applicants for, a power reactor early site 
permit, combined license, standard design certification, standard 
design approval, or manufacturing license under 10 CFR Part 52, 
``Licenses, Certifications, and Approvals for Nuclear Power Plants.''

Intent

    The U.S. Nuclear Regulatory Commission (NRC) is issuing this 
regulatory issue summary (RIS) to remind addressees of the American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 
(ASME Code) requirements in accordance with 10 CFR 50.55a, ``Codes and 
Standards,'' and of the quality assurance (QA) requirements for design 
control in accordance with Appendix B, ``Quality Assurance Criteria for 
Nuclear Power Plants and Fuel Reprocessing Plants,'' to 10 CFR part 50. 
Specifically, this RIS informs addressees of the NRC's findings from 
license renewal and new reactor audits on applicants' analyses and 
methodologies using the computer software package, WESTEMS\TM\, to 
demonstrate compliance with Section III, ``Rules for Construction of 
Nuclear Facility Components,'' of the ASME Code. The NRC expects 
addressees to review this RIS for applicability to their facilities and 
to consider actions as appropriate. This RIS requires no action or 
written response from addressees.

Background Information

    Section 54.21 of 10 CFR, ``Contents of Application--Technical 
Information,'' requires applicants for license renewal to perform an 
evaluation of time-limited aging analyses relevant to structures, 
systems, and components within the scope of license renewal. In most 
cases, fatigue analyses of the reactor coolant pressure boundary 
components involve time-limited assumptions. In addition, the staff has 
provided guidance in NUREG-1800, ``Standard Review Plan for Review of 
License Renewal Applications for Nuclear Power Plants,'' Revision 2, 
issued December 2010, which recommends that the effects of the reactor 
water environment on fatigue life be evaluated for a sample of 
components to provide assurance that cracking due to fatigue will not 
occur during the period of extended operation. Because 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.
    Regulatory Guide 1.28, ``Quality Assurance Program Criteria (Design 
and Construction),'' describes methods that the NRC considers 
acceptable for complying with the requirements in Appendix B to 10 CFR 
part 50 for establishing and implementing a QA program for the design 
and construction of nuclear power plants and fuel reprocessing plants.
    The regulations at 10 CFR 50.55a specify the ASME Code 
requirements. In particular, 10 CFR 50.55a(c) requires, in part, that 
components of the reactor coolant pressure boundary must meet the 
requirements for Class 1 components in Section III of the ASME Code, 
with limited exceptions specified in 10 CFR 50.55a(c)(2)(4). Some 
operating facilities may have performed a supplemental detailed fatigue 
analysis of components because of new operating conditions identified 
after the plant began operation.

Summary of Issue

    The staff has identified concerns about the computer software 
package, WESTEMS\TM\, that is used to demonstrate the ability of 
nuclear power plant components to withstand the cyclic loads associated 
with plant transient operations. This particular computer software 
package involves the use of computer code developed to calculate 
fatigue usage during plant transient operations such as startups and 
shutdowns, as discussed in ASME Code, Section III, Subsection NB, 
Subarticles NB-3200, ``Design By Analysis,'' and NB-3600, ``Piping 
Design.''
    The staff identified these concerns with the computer software 
package during the review of the AP1000 design certification 
application, and they are described in the staff's safety evaluation 
report (Agencywide Documents Access and Management System (ADAMS) 
Accession No. ML103430502) and its related audit report (ADAMS 
Accession No. ML110250634). One such concern was that the methodology 
used by this computer software package to determine the peak stress 
intensity range time history in fatigue calculations uses the algebraic 
summation of three orthogonal moment vectors. This algebraic summation 
methodology is not consistent with ASME Code, Section III, Subsection 
NB, Subarticle NB-3650, ``Analysis of Piping Products,'' which states 
that resultant moments from different load sets shall not be used in 
calculating the moment range (i.e., this algebraic summation 
methodology is not an accurate representation of the moment range). 
Therefore, the use of this practice could provide results that are not 
accurate. The staff also identified a concern in which, under certain 
circumstances, the use of this computer software package requires the 
user to manually modify peak and valley times/stresses during 
intermediate calculations in the software. Although this method of 
analyst intervention could provide acceptable results in some cases, 
reliance on the user's engineering judgment and ability to modify peak 
and valley times/stresses, without control and documentation, could 
produce results that are not predictable, repeatable, or conservative. 
Because of these concerns, the applicant for the AP1000 design 
certification elected to remove the use of this computer software 
package from its design certification document, such that it is not 
used in the design for the AP1000, as documented in ADAMS Accession No. 
ML102770329.
    License renewal applicants have attempted to use this computer 
software package to demonstrate acceptable fatigue calculations for 
plant operation during the period of extended operation. As a result of 
the concerns described above, the staff asked a license renewal 
applicant that has used this computer software package to perform an 
evaluation to demonstrate that the package provides acceptable results 
and to assess the impact of these identified concerns on the license 
renewal applicant's fatigue calculations (ADAMS Accession No. 
ML102810194). The staff conducted an audit to (1) review this 
evaluation, (2) address the user's ability to manually modify peak and 
valley times/stresses, and (3) address the aforementioned concern

[[Page 60941]]

with the algebraic summation of three orthogonal moment vectors.
    At the conclusion of the audit, the staff determined, as described 
in its audit report (ADAMS Accession No. ML110871243), that the license 
renewal applicant's use of this computer software package demonstrated 
(1) that it produced calculations of stresses and cumulative usage 
factors that are consistent with the methodology in ASME Code, Section 
III, Subsection NB, Subarticle NB-3200, (2) that the analyst's judgment 
in manually modifying peak and valley times/stresses in these 
calculations was reasonable and can be appropriately justified and 
documented, though justification of any user intervention should be 
documented, (3) that this applicant did not use this software to 
perform fatigue calculations as described in ASME Code, Section III, 
Subsection NB, Subarticle NB-3600, and (4) future use of this software 
should be accompanied by an acceptable demonstration that it performs 
fatigue calculations in accordance with ASME Code, Section III, 
Subsection NB, Subarticle NB-3600.
    This license renewal applicant performed evaluations on two of its 
components: A pressurized water reactor (PWR) pressurizer surge nozzle 
and a PWR safety injection boron injection tank nozzle. When 
considering the effects of the reactor water environment on fatigue 
life, these evaluations indicated a cumulative usage factor that was 
less than the ASME Code design limit of 1.0, provided that there was 
sufficient and clear records of justification for analyst intervention.
    The staff acknowledges that addressees may have used, or will make 
use of, other computer software packages in performing ASME Code 
fatigue calculations. Thus, the NRC encourages addressees to review the 
documents discussed above and to consider actions, as appropriate, to 
ensure compliance with the requirements for ASME Code fatigue 
calculations and QA programs, as described in 10 CFR 50.55a and 
Appendix B to 10 CFR part 50, respectively.

Backfit Discussion

    This RIS informs addressees of potential concerns with the use of 
computer software packages to perform ASME Code fatigue calculations 
and reminds them that they should perform these calculations in 
accordance with ASME Code requirements. The regulations at 10 CFR 
50.55a specify the ASME Code requirements. Regulatory Guide 1.28 
describes methods for establishing and implementing a QA program for 
the design and construction of nuclear power plants. For license 
renewal, metal fatigue is evaluated as a time-limited aging analysis in 
accordance with 10 CFR 54.21(c). Section 4.3, ``Metal Fatigue,'' of 
NUREG-1800 provides the associated staff review guidance. 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.

Paperwork Reduction Act Statement

    This RIS does not contain any information collections and, 
therefore, is not subject to the requirements of the Paperwork 
Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing collection 
requirements under 10 CFR Part 54 were approved by the Office of 
Management and Budget, control number 3150-0155.
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 
contact listed below:

Timothy J. McGinty, Director, Division of Policy and Rulemaking, Office 
of Nuclear Reactor Regulation.
Laura A. Dudes, Director, Division of Construction Inspection and 
Operational Programs, Office of New Reactors.

    Technical Contact: On Yee, NRR, 301-415-1905. E-mail: 
[email protected].

    Note:  NRC generic communications may be found on the NRC public 
Web site, http://www.nrc.gov, under NRC Library/Document 
Collections.


END OF DRAFT REGULATORY ISSUE SUMMARY

    Dated at Rockville, Maryland this 22nd day of September 2011.

    For the Nuclear Regulatory Commission.
Melanie A. Galloway,
Acting Director, Division of License Renewal, Office of Nuclear Reactor 
Regulation.
[FR Doc. 2011-25242 Filed 9-29-11; 8:45 am]
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