ML21265A184

Correction to Issuance of Amendment to Adopt Technical Specification Task Force (TSTF) 510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection

ML21265A184
Person / Time
Site:	North Anna
Issue date:	12/06/2021
From:	Geoffrey Miller Plant Licensing Branch II
To:	Stoddard D Virginia Electric & Power Co (VEPCO)
Miller G
References
TAC ME9156, TAC ME9157
Download: ML21265A184 (4)

Text

December 6, 2021 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 - CORRECTION TO ISSUANCE OF AMENDMENT TO ADOPT TECHNICAL SPECIFICATION TASK FORCE (TSTF) 510, REVISION 2, REVISION TO STEAM GENERATOR PROGRAM INSPECION FREQUENCIES AND TUBE SAMPLE SELECTION (TAC NOS. ME9156 AND ME9157)

Dear Mr. Stoddard:

On January 28, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13018A140), the U.S. Nuclear Regulatory Commission (NRC) issued Amendment Nos. 269 and 250 to Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station (North Anna), Unit Nos. 1 and 2, respectively. The amendments revised the North Anna and Surry Technical Specifications requirements as described in TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection.

During the review of your current application for TSTF-577, the NRC recognized that it issued a TS page with those amendments that did not incorporate all of the changes you had requested.

The NRC staff has determined that the error was made inadvertently and has corrected the error. The corrections do not change any of the conclusions associated with the issuance of Amendment Nos. 269 and 250 to Renewed Facility Operating License Nos.

NPF-4 and NPF-7 for North Anna, Unit Nos. 1 and 2, respectively, and do not affect the no significant hazards consideration published in the Federal Register on October 2, 2012 (77 FR 60155). Accordingly, Enclosure 1 to this letter provides the corrected page.

D. Stoddard If you have any questions, please contact me at 301-415-4032, or via e-mail at Ed.Miller@nrc.gov.

Sincerely,

/RA/

G. Edward Miller, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339

Enclosure:

1. Revised TS page cc: Listserv

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program

a. (continued) inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm for all SGs.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.

North Anna Units 1 and 2 5.5-6 Amendments 269/250

ML21265A184 OFFICE NRR/DORL/ NRR/DORL/ NRR/DORL/ NRR/DORL/

LPL2-1/PM LPL2-1/LA LPL2-1/BC LPL2-1/PM NAME GEMiller KGoldstein MMarkley GEMiller DATE 12/2/21 09/24/2021 12/3/21 12/6/21