ML12163A271

Lr - FW: Draft Request for Information

ML12163A271
Person / Time
Site:	Limerick
Issue date:	05/01/2012
From:	Office of Nuclear Reactor Regulation
To:	Division of License Renewal
References
Download: ML12163A271 (5)

Text

LimerickNPEm Resource From: Kuntz, Robert Sent: Tuesday, May 01, 2012 2:19 PM To: Anthony Z. Roisman; gfettus@nrdc.org Cc: Smith, Maxwell; Kanatas, Catherine

Subject:

FW: DRAFT Request for Information Attachments: Limericick DRAFT RAIs RE APP J and BWRVIP.docx From: Kuntz, Robert Sent: Tuesday, March 06, 2012 6:53 AM To: 'Christopher.Wilson2@exeloncorp.com'

Subject:

DRAFT Request for Information

Chris, Attached is a DRAFT request for information related to the Limerick license renewal request. If Exelon would like clarification on the attached DRAFT RAI let me know and I will set up a teleconference for the staff to provide clarity.

Robert Kuntz Sr. Project Manager NRR/ADRO/DLR/RPB1 (301) 415-3733 robert.kuntz@nrc.gov 1

Hearing Identifier: Limerick_LR_NonPublic Email Number: 226 Mail Envelope Properties (Robert.Kuntz@nrc.gov20120501141900)

Subject:

FW: DRAFT Request for Information Sent Date: 5/1/2012 2:19:12 PM Received Date: 5/1/2012 2:19:00 PM From: Kuntz, Robert Created By: Robert.Kuntz@nrc.gov Recipients:

"Smith, Maxwell" <Maxwell.Smith@nrc.gov>

Tracking Status: None "Kanatas, Catherine" <Catherine.Kanatas@nrc.gov>

Tracking Status: None "Anthony Z. Roisman" <aroisman@nationallegalscholars.com>

Tracking Status: None "gfettus@nrdc.org" <gfettus@nrdc.org>

Tracking Status: None Post Office:

Files Size Date & Time MESSAGE 524 5/1/2012 2:19:00 PM Limericick DRAFT RAIs RE APP J and BWRVIP.docx 29696 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Recipients Received:

DRAI B.2.1.33-2

Background

GALL Report Aging Management Program (AMP) XI.S4, "10 CFR Part 50 Appendix J" program, in detection of aging effects, program element states that while the calculation of leakage rates and satisfactory performance of containment leakage rate testing demonstrates the leak-tightness and structural integrity of the containment, it does not by itself provide information that would indicate that aging degradation has initiated or that the capacity of the containment may have been reduced. The NRC through two Information Notices (INs) identified conditions that could impact leak tightness and aging degradation of the containment boundary pressure-retaining systems and components (SCs) by line vibrations and other external loadings.

Information Notices, IN 2005-23 Vibration Induced Degradation of Butterfly Valves and IN 2006-15 Vibration Induced Degradation and Failure of Safety-Related Valves, have been issued to plant's describing circumstances which aside from valve malfunctioning and leakage could also include accelerated aging, degradation, cracking, and loss of function in various systems penetrating the containment, including seals and gaskets the condition of which are to be monitored by the 10 CFR Part 50, Appendix J program.

Issue During the audit a search of the plants operating experience database indicated that Limerick Generating Stations (LGS), Unit 2 main steam isolation valve (MSIV) experienced vibration and or shuddering. Although the staff noted that the issue was resolved, the staff discussed with the applicant that such vibrations within proximity to the containment boundary could damage pressure-retaining components. The staff referenced the above INs and further noted to the applicant that effects of vibration could impact the integrity of SCs associated with Type B and C tests, including the inspection intervals of such components.

Request Discuss how IN 2005-23 and IN 2006-15 have been dispositioned and how the 10 CFR Part 50, Appendix J program will account for the recommendations in those INs for potentially affected SCs that could potentially compromise the containment pressure boundary integrity during the period of extended operation.

DRAI B.2.1.33-3

Background

GALL Report AMP XI.S4, 10 CFR Part 50, Appendix J, in its scope of program, program element states that the scope of the program includes all containment boundary pressure-retaining SCs. The Updated Final Safety Analysis Report (UFSAR) lists the containment components (penetrations and valves) subject to Type B or C testing as required by 10 CFR Part 50, Appendix J. The UFSAR also states that the Technical Requirements Manual (TRM) contains the plants testing requirements. The TRM as well has a list of the components subject to 10 CFR Part 50, Appendix J testing.

Issue During the audit the staff noted a condition report which stated that there are discrepancies between the UFSAR and the TRM documentation on implementing procedures and testing for the 10 CFR Part 50 Appendix J testing. Although these differences in testing procedures are being tracked by the applicant, it is unclear to the staff for the scope of program, program element, which document, the UFSAR or the TRM, the applicant will use for testing of SCs during the period of extended operation to meet the recommendations for the 10 CFR Part 50, Appendix J program.

Request

1. State which document, UFSAR or the TRM, will be used for testing of SCs during the period of extended operation to meet the scope of program, program element of 10 CFR Part 50, Appendix J program.

2. Update the LRA B.2.1.33 10 CFR Part 50, Appendix J program to indicate the document to be followed during the implementation of the 10 CFR Part 50 Appendix J program testing.

DRAI 4.1-2 Background and Issue:

10 CFR 54.21(c) indicates that license renewal applicants must include a list of time-limited aging analyses (TLAA), as defined in 10 CFR 54.3 and the TLAA must be dispositioned in accordance with 10 CFR 54.21(c)(1). The response to RAI BWRVIP-1, in a letter dated February 15, 2012, included a new Appendix C to the LGS LRA to address action items in all applicable BWRVIP reports credited for aging management.

License Renewal Action Item No. 14 for BWRVIP-74-A states:

Components that have indications that have been previously analytically evaluated in accordance with sub-section IWB-3600 of Section XI to the ASME Code until the end of the 40-year service period shall be re-evaluated for the 60-year service period corresponding to the LR term.

A commitment (Commitment No. 47) was provided in response to Action Item No. 14, to re-evaluate the flaw in the LGS, Unit 1 reactor pressure vessel (RPV) nozzle to safe-end weld VRR-1RD-1A-N2H in accordance with ASME Code Section XI, subsection IWB-3600 for the 60-year service period corresponding to the license renewal term.

The response did not include a justification of why this analysis was not identified as a TLAA in the LRA in accordance with 10 CFR 54.21(c)(1). In addition, the commitment to perform the analysis at a later date does not demonstrate an adequate evaluation of the TLAA.

Request:

Clarify how the flaw evaluation of the LGS, Unit 1 RPV nozzle to safe-end weld VRR-1RD-1A-N2H compares to the six criteria for TLAAs in 10 CFR 54.3, and justify whether or not the flaw evaluation should be identified as a TLAA for the LRA under the TLAA identification

requirements of 10 CFR 54.21(c)(1). If the analysis needs to be identified as a TLAA, provide necessary information and LRA revision to support the TLAA disposition.