ML23153A173

Correction to Amendment Nos. 346 & 286 Millstone, 294 & 277 North Anna, 311 & 311 Surry, and 225 Summer to Revise Technical Specifications to Adopt TSTF-554,Rev Reactor Coolant Leakage Requirement

ML23153A173
Person / Time
Site:	Millstone, Surry, North Anna
Issue date:	06/16/2023
From:	Geoffrey Miller Plant Licensing Branch II
To:	Stoddard D Dominion Nuclear
Miller G
References
EPID L-2022-LLA-0078
Download: ML23153A173 (1)

Text

June 16, 2023 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NOS. 2 AND 3, NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, SURRY POWER STATION, UNIT NOS. 1 AND 2, AND VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 CORRECTION TO AMENDMENT NO(S). 346 AND 286 (MILLSTONE), 294 AND 277 (NORTH ANNA), 311 AND 311 (SURRY), AND 225 (SUMMER) TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-554, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS (EPID L-2022-LLA-0078)

Dear Mr. Stoddard:

On May 1, 2022 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML23072A089), the U.S. Nuclear Regulatory Commission (NRC) issued amendments to the Dominion fleet. The amendments revised the respective Technical Specifications (TSs) consistent with Technical Specification Task Force (TSTF) Traveler TSTF-554, Revise Reactor Coolant Leakage Requirements, to revise the TS definition of leakage, clarify the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified.

Subsequently, Dominion Energy identified some errors in issued pages. Specifically, the camera-ready pages provided for Millstone Units 2 and 3 had minor typographical errors.

Additionally, the NRC staff incorrectly reflected the licensee name on the license page for Summer. As evaluated below, the NRC staff is reissuing the affected pages.

For Millstone 2, the provided page 3/4 4-9 inadvertently changed the word valve to value in Action 3.4.6.2.a. The NRC staff confirmed that the marked-up pages provided with the amendment properly indicated the word valve. Therefore, consistent with the NRC staff guidance dated January 16, 1997 (ML103260096), based on the NRCs policy established by SECY-96-238, Guidance for Correction of Technical Specification Typographical Errors, dated November 19, 1996, this error can be corrected by a letter to the licensee from the NRC staff.

For Millstone 3, the provided page 3/4 4-22 inadvertently added the word the before the word in Action 3.4.6.2.c. The NRC staff confirmed that the marked-up pages provided with the amendment did not include the word the. Therefore, consistent with NRC staff guidance dated January 16, 1997, based on the NRCs policy established by SECY-96-238, this error can be corrected by a letter to the licensee from the NRC staff.

D. Stoddard For Summer, license page 3 inadvertently reflected the licensee as SCE&G where it should have been DESC. This error was made by the NRC staff in when it generated the amended license page. Therefore, consistent with NRC staff guidance dated January 16, 1997, based on the NRCs policy established by SECY-96-238, this error can be corrected by a letter to the licensee from the NRC staff.

Accordingly, the corrected pages are enclosed with this letter. The corrections do not change any of the conclusions associated with the issuance of the amendments, and do not affect the no significant hazards consideration published in the Federal Register on August 9, 2022 (87 FR 48516).

If you have any questions, please contact me at 301-415-2481, or via email 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. 50-336, 50-423, 50-338, 50-339, 50-280, 50-281, and 50-395

Enclosure:

Corrected TS Pages cc: Listserv

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System Operational LEAKAGE shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 75 GPD primary to secondary LEAKAGE through any one steam generator, and

d. 10 GPM IDENTIFIED LEAKAGE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With PRESSURE BOUNDARY LEAKAGE, isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. With any RCS operational LEAKAGE not within limits for reasons other than PRESSURE BOUNDARY LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

c. With ACTION and associated completion time not met, or primary to secondary LEAKAGE not within limits, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 NOTES

1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance at the frequency specified in the Surveillance Frequency Control Program.

MILLSTONE - UNIT 2 3/4 4-9 Amendment No. 25, 37, 82, 85, 101, 121, 138, 215, 228, 299, 324, 346

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary LEAKAGE through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE,

e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2250 20 psia, and f.* 0.5 gpm LEAKAGE per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2250 20 psia from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With PRESSURE BOUNDARY LEAKAGE, isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activiated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With primary to secondary LEAKAGE not within limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any RCS operational LEAKAGE not within limits for reasons other than PRESSURE BOUNDARY LEAKAGE, primary to secondary LEAKAGE, or LEAKAGE from Reactor Coolant System Pressure Isolation Valves, reduce LEAKAGE to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• This requirement does not apply to Pressure Isolation Valves in the Residual Heat Removal flow path when in, or during the transition to or from, the shutdown cooling mode of operation.

MILLSTONE - UNIT 3 3/4 4-22 Amendment No. 209, 238, 286

(3) DESC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage amounts required for reactor operation, as described in the Final Safety Analysis Report, as amended through Amendment No. 33; (4) DESC, pursuant to the Act and 10 CFR Part 30, 40 and 70 to receive, possess and use at any time byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) DESC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus of components; and (6) DESC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as m[a]y be produced by the operation of the facility.

C. This renewed license shall be deemed to contain, and is subject to, the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level DESC is authorized to operate the facility at reactor core power levels not in excess of 2900 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to this renewed license.

The preoccupation tests, startup tests and other items identified in Attachment 1 to this renewed license shall be completed as specified.

Attachment 1 is hereby incorporated into this renewed license.

(2) Technical Specifications and Environmental Protection Plant The Technical Specifications contained in Appendix A, as revised through Amendment No. 225, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. Dominion Energy South Carolina, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed Facility Operating License No. NPF-12 Amendment No. 225

ML23153A173 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DORL/LPL1/LA NRR/DORL/LPL2-1/BC NAME GEMiller KGoldstein (RButler for) KEntz MMarkley DATE 6/6/2023 6/5/2023 6/6/2023 6/16/2023