ML22146A027

Application to Revise Technical Specifications to Adopt TSTF-554, Revise Reactor Coolant.

ML22146A027
Person / Time
Site:	Millstone, Summer, Surry, North Anna
Issue date:	05/25/2022
From:	Lawrence D Dominion Energy Nuclear Connecticut, Dominion Energy South Carolina, Dominion Energy Virginia
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
21-147, TSTF-554
Download: ML22146A027 (82)

Text

Dominion Energy Services, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 DominionEnergy.com May 25, 2022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

10 CFR 50.90 Serial No.:

21-147 NRA/SS RO Docket Nos.: 50-336/423 50-280/281 50-338/339 50-395 License Nos NPF-49 DPR-65 DPR-32/37 NPF-4/7 NPF-12 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

DOMINION ENERGY SOUTH CAROLINA, INC.

MILLSTONE POWER STATION UNITS 2 AND 3 SURRY POWER STATION UNITS 1 AND 2 NORTH ANNA POWER STATION UNITS 1 AND 2 V. C. SUMMER NUCLEAR STATION UNIT 1 APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-554, "REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS" Pursuant to 10 CFR 50.90 Dominion Energy Connecticut, Inc., Dominion Energy Virginia, and Dominion Energy South Carolina, Inc., (hereafter "the Licensees") are submitting a request for amendments to the Technical Specifications (TS) for Millstone Power Station (MPS) Units 2 and 3, Surry Power Station (SPS) Units 1 and 2, North Anna Power Station (NAPS) Units 1 and 2, and V. C. Summer Nuclear Station (VCS)

Unit 1, respectively.

The licensees request adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the MPS Units 2 and 3, SPS Units 1 and 2, NAPS Units 1 and 2, and VCS Unit 1 TS. The proposed amendment revises the Technical Specification (TS) definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified. The change is requested as part of the Consolidated Line Item Improvement Process (CLI IP).

The enclosure provides a description and assessment of the proposed changes. provides the existing TS pages marked to show the proposed changes. provides revised (clean) TS pages. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only. The Facility

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281 /338/339/395 Page 2 of 4 Safety Review Committee at each station has reviewed and concurred with the determinations herein.

Approval of the proposed amendment is requested by April 30, 2023. Once approved, the amendment shall be implemented within 30 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State Officials.

If you have any questions regarding this submittal, please contact Mr. Shayan Sinha at (804) 273-4687.

Sincerely, Douglas C. Law nee Vice-President - Nuclear Engineering and Fleet Support

Enclosure:

Description and Assessment Attachments:

1. Proposed Technical Specifications Changes (Mark-Up)

2. Revised Technical Specifications Pages

3. Proposed Technical Specification Bases Changes (Mark-Up)

Commitments made in this letter: None COMMONWEAL TH OF VIRGINIA

)

)

COUNTY OF HENRICO

)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Douglas C. Lawrence, who is Vice President -

Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc., Virginia Electric and Power Company, and Dominion Energy South Carolina, Inc.

He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this,ZSffi day of t1 o-\\,\\ j My Commission Expires: Q~o+ 31, Zo?.3.

, 2022.

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 Page 3 of 4 cc:

U.S. Nuclear Regulatory Commission, Region I Regional Administrator 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 U.S. Nuclear Regulatory Commission, Region II Regional Administrator Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector Millstone Power Station NRC Senior Resident Inspector Surry Power Station NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector V.C. Summer Nuclear Station Mr. G. Edward Miller NRC Senior Project Manager - North Anna Units 1 and 2 and V. C. Summer U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 Mr. R. V. Guzman NRC Senior Project Manager - Millstone Units 2 and 3 U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 C-2 11555 Rockville Pike Rockville, MD 20852-2738 Mr. L. John Klos NRC Project Manager - Surry Units 1 and 2 U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738

Old Dominion Electric Cooperative Electronically Distributed Director, Radiation Division Serial No.: 21-147 Docket NOS.: 50-336/423/280/281 /338/339/395 Page 4 of 4 Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 State Health Commissioner Virginia Department of Health James Madison Building - 7th Floor 109 Governor Street; Room 730 Richmond, Virginia 23219 Ms. Anuradha Nair-Gimmi Bureau of Environmental Health SeNices South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Mr. G. J. Lindamood Santee Cooper - Nuclear Coordinator 1 Riverwood Drive Moncks Corner, SC 29461

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 Enclosure DESCRIPTION AND ASSESSMENT Dominion Energy Connecticut, Inc. (DENC)

Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC)

Millstone Power Station Units 2 and 3 North Anna Power Station Units 1 and 2 Surry Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 Enclosure, Page 1 of 4 DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Dominion Energy Connecticut, Inc., (DENC), Virginia Electric and Power Company (Dominion Energy Virginia), and Dominion Energy South Carolina, Inc. (DESC), (hereafter the Licensees) request adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the Millstone Power Station (MPS) Units 2 and 3, Surry Power Station (SPS) Units 1 and 2, North Anna Power Station (NAPS)

Units 1 and 2, and V. C. Summer Nuclear Station (VCS) Unit 1 Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified.

2.0 ASSESSMENT

2.1 Applicability of Safety Evaluation The licensees have reviewed the safety evaluation for TSTF-554 provided to the Technical Specifications Task Force in a letter dated December 18, 2020. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-554. The licensees have concluded that the justifications presented in TSTF-554 and the safety evaluation prepared by the NRC staff are applicable to MPS Units 2 and 3, SPS Units 1 and 2, NAPS Units 1 and 2, and VCS Unit 1 and justify this amendment for the incorporation of the changes into the respective plants TS.

2.2 Variations The Licensees are proposing the following variation from the TS changes described in TSTF-554 or the applicable parts of the NRC staff safety evaluation, dated December 18, 2020:

SPS Units 1 and 2, MPS Units 2 and 3 and VCS Unit 1 utilize different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-554 was based. Specifically:

The SPS TS LEAKAGE Definition is numbered 1.0.X, whereas the TSTF-554 LEAKAGE Definition is numbered 1.1. The SPS TS numbering for RCS Leakage is 3.1.C (titled RCS Operational LEAKAGE), whereas the TSTF-554 numbering for RCS Leakage is 3.4.13. The SPS TS Bases for RCS Operational Leakage is discussed in the TS 3.1.C and 4.13 Bases, whereas TSTF-554 Bases for RCS Leakage is discussed in B 3.4.13.

The MPS Unit 2 TS for LEAKAGE Definition is numbered 1.14, whereas the TSTF-554 LEAKAGE Definition is numbered 1.1. The MPS Unit 2 TS numbering for RCS Leakage is 3.4.6.2 (titled Reactor Coolant System Operational LEAKAGE), whereas the TSTF-554 numbering for RCS

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 Enclosure, Page 2 of 4 Leakage is 3.4.13. The MPS Unit 2 TS Bases for RCS Leakage is discussed in Bases 3/4.4.6.2, whereas the TSTF-554 Bases for RCS Leakage is discussed in B 3.4.13.

The MPS Unit 3 TS for LEAKAGE Definition is numbered 1.16, whereas the TSTF-554 LEAKAGE Definition is numbered 1.1. The MPS Unit 3 TS numbering for RCS Leakage is 3.4.6.2 (titled Reactor Coolant System Operational LEAKAGE), whereas the TSTF-554 numbering for RCS Leakage is 3.4.13. The MPS Unit 3 TS Bases for RCS Leakage is discussed in Bases 3/4.4.6.2, whereas the TSTF-554 Bases for RCS Leakage is discussed in B 3.4.13.

The VCS Unit 1 TS for IDENTIFIED LEAKAGE and PRESSURE BOUNDARY LEAKAGE Definitions are numbered 1.15 and 1.21, respectively, whereas the TSTF-554 LEAKAGE Definition is numbered 1.1. The VCS Unit 1 TS numbering for RCS Leakage is 3.4.6.2 (titled Reactor Coolant System Operational Leakage), whereas the TSTF-554 numbering for RCS Leakage is 3.4.13. The VCS Unit 1 TS Bases for RCS Leakage is discussed in Bases 3/4.4.6.2, whereas the TSTF-554 Bases for the RCS Leakage is discussed in B 3.4.13.

These differences are administrative and do not affect the applicability of TSTF-554 to the SPS Units 1 and 2, MPS Units 2 and 3 and VCS Unit 1 TS.

In the Standard Technical Specifications on which TSTF-554 is based, all leakage definitions appear as part of a single defined term, "LEAKAGE." In the VCS Unit 1 TS, the terms related to leakage are separate definitions consistent with NUREG-0452, Revision 4 (Standard Technical Specifications for Westinghouse Pressurized Reactors, Fall 1981). However, the changes to the affected defined terms "IDENTIFIED LEAKAGE" and "PRESSURE BOUNDARY LEAKAGE" are the same as the changes made to those terms in TSTF-554.

For NAPS Units 1 and 2, no variations from the TS changes described in TSTF-554 or the applicable parts of the NRC staff safety evaluation dated December 18, 2020, are proposed. One typographical correction is being included in the NAPS Unit 1 and 2 TS Bases to change an incorrect reference from GDC [Generic Design Criterion] 3 to GDC

30.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis The licensees request adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the Millstone Power Station (MPS) Units 2 and 3, Surry Power Station (SPS)

Units 1 and 2, North Anna Power Station (NAPS) Units 1 and 2, and V. C. Summer Nuclear Station (VCS) Unit 1 Technical Specifications (TS). The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 Enclosure, Page 3 of 4 boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified.

The licensees have evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified.

The proposed change revises the definition of pressure boundary leakage. Pressure boundary leakage is a precursor to some accidents previously evaluated. The proposed change expands the definition of pressure boundary leakage by eliminating the qualification that pressure boundary leakage must be from a "nonisolable" flaw. A new TS Action is created which requires isolation of the pressure boundary flaw from the Reactor Coolant System. This new action provides assurance that the flaw will not result in any accident previously evaluated.

Pressure boundary leakage, and the actions taken when pressure boundary leakage is detected, is not assumed in the mitigation of any accident previously evaluated. As a result, the consequences of any accident previously evaluated are unaffected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified. The proposed change does not alter the design function or operation of the RCS. The proposed change does not alter the ability of the RCS to perform its design function. Since pressure boundary leakage is an evaluated accident, the proposed change does not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 Enclosure, Page 4 of 4

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed amendment revises the TS definition of "Leakage," clarifies the requirements when pressure boundary leakage is detected, and adds a Required Action when pressure boundary leakage is identified. The proposed change does not affect the initial assumptions, margins, or controlling values used in any accident analysis. The amount of leakage allowed from the RCS is not increased. The proposed change does not affect any design basis or safety limit or any Limiting Condition for Operation.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the licensees conclude that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 PROPOSED TECHNICAL SPECIFICATIONS CHANGES (MARK-UP)

Dominion Energy Connecticut, Inc. (DENC)

Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC)

Millstone Power Station Units 2 and 3 North Anna Power Station Units 1 and 2 Surry Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

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-FOR INFORMATION ONLY-Serial No.21-147, Page 1 of 18

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-FOR INFORMATION ONLY-Serial No.21-147, Page 2 of 18 INSERT A

INSERT A

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 />.

Serial No.21-147, Page 3 of 18

MILLSTONE - UNIT 3 DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microCurie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

1.12 DELETED ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

1.14 DELETED FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

LEAKAGE 1.16 LEAKAGE shall be:

1.16.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either to not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE,; or October 27, 2008 Amendment No. 84, 87, 126, 187, 216, 220, 238, 246 1-3

-FOR INFORMATION ONLY-Serial No.21-147, Page 4 of 18

MILLSTONE - UNIT 3 DEFINITIONS c.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

1.16.3 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in a RCS component body, pipe wall, or vessel wall,. LEAKAGE past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE, and 1.16.4 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all LEAKAGE which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

MASTER RELAY TEST 1.17 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include continuity check of each associated slave relay.

MEMBER(S) OF THE PUBLIC 1.18 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

The term REAL MEMBER OF THE PUBLIC means an individual who is exposed to existing dose pathways at one particular location.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

May 31, 2007 Amendment No. 238 1-4

-FOR INFORMATION ONLY-Serial No.21-147, Page 5 of 18

MILLSTONE - UNIT 3 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 primary to secondary LEAKAGE not within limits or any PRESSURE BOUNDARY LEAKAGE, 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 />.

b.

With any RCS operational LEAKAGE not within limits, other than PRESSURE BOUNDARY LEAKAGE, LEAKAGE from Reactor Coolant System Pressure Isolation Valves or primary to secondary LEAKAGE, reduce the leakage rate 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 STANDBY 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 />.

c.

With any Reactor Coolant System Pressure Isolation Valve LEAKAGE greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY 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.

May 31, 2007 Amendment No. 209, 238 3/4 4-22

-FOR INFORMATION ONLY-Serial No.21-147, Page 6 of 18 INSERT A d.

INSERT A 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 /> or be in at least HOT STANDBY 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 STANDBY 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 />.

Serial No.21-147, Page 7 of 18

North Anna Units 1 and 2 1.1-4 Amendments 258/239 Definitions 1.1 1.1 Definitions LEAKAGE (continued)

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either to not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLEOPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, (continued)

FOR INFORMATION ONLY Serial No.21-147, Page 8 of 18

RCS Operational LEAKAGE 3.4.13 North Anna Units 1 and 2 3.4.13-1 Amendments 248/228 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE;

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B A. RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B A.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C B. Required Action and associated Completion Time of Condition A not met.

OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

C B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND C B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> FOR INFORMATION ONLY INSERT A Serial No.21-147, Page 9 of 18

INSERT A CONDITION REQUIRED ACTION COMPLETION TIME A. Pressure boundary LEAKAGE exists.

A.1 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.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Serial No.21-147, Page 10 of 18

W.

STAGGERED TEST BASIS A staggered test basis shall consist of:

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

X.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either to not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE,;

or 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);.

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE, and c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

Amendment Nos. 251 and 250 TS 1.0-8 03-29-07 FOR INFORMATION ONLY Serial No.21-147, Page 11 of 18 LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

TS Markup for Surry Units 1 and 2

C.

RCS Operational LEAKAGE Applicability The following specifications are applicable to RCS operational LEAKAGE whenever Tavg (average RCS temperature) exceeds 200°F (200 degrees Fahrenheit).

Specifications 1.

RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE, b.

1 gpm unidentified LEAKAGE, c.

10 gpm identified LEAKAGE, and d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

2.3.a. If RCS operational LEAKAGE is not within the limits of 3.1.C.1 for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within the specified limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

If the LEAKAGE is not reduced to within the specified limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

3 4.

If RCS pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within the limit specified in 3.1.C.1.d, the unit shall be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

TS 3.1-13 11-05-09 Amendment Nos. 267 and 250 FOR INFORMATION ONLY Serial No.21-147, Page 12 of 18 2.a. If pressure boundary LEAKAGE exists, 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. If the pressure boundary LEAKAGE is not isolated as specified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

TS Markup for Surry Units 1 and 2

4.

Detected or suspected leakage from the Reactor Coolant System shall be investigated and evaluated. At least two means shall be available to detect reactor coolant system leakage. One of these means must depend on the detection of radionuclides in the containment.

5.a. Prior to going critical all primary coolant system pressure isolation valves listed below shall be functional as a pressure isolation device, except as specified in 3.1.C.5.b. Valve leakage shall not exceed the amounts indicated.

b.

If Specification 3.1.C.5.a cannot be met, an orderly shutdown shall be initiated and the reactor shall be in HOT SHUTDOWN 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 />.

Notes (a) 1.

Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4.

Leakage rates greater than 5.0 gpm are considered unacceptable.

Unit 1 Unit 2 Max. Allowable Leakage (see note (a) below)

Loop A, Cold Leg 1-SI-79, 1-SI-241 2-SI-79, 2-SI-241 5.0 gpm for each valve Loop B, Cold Leg 1-SI-82, 1-SI-242 2-SI-82, 2-SI-242 Loop C, Cold Leg 1-SI-85, 1-SI-243 2-SI-85, 2-SI-243 TS 3.1-13a 03-29-07 Amendment Nos. 251 and 250 FOR INFORMATION ONLY Serial No.21-147, Page 13 of 18 5

6 6

6 TS Markup for Surry Units 1 and 2

DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either to not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

SUMMER - UNIT 1 1-3 Amendment No. 146, 179 Serial No.21-147, Page 14 of 18

DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

SUMMER - UNIT 1 1-4 Amendment No. 104, 117, 179 Serial No.21-147, Page 15 of 18

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 from the Reactor Coolant System, and e.

33 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 20 psig.

f.

The leakage rate specified for each Reactor Coolant System Pressure Isolation Valve in Table 3.4-1 at a Reactor Coolant System pressure of 2235 20 psig.

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

a.

With any PRESSURE BOUNDARY LEAKAGE or with primary-to-secondary leakage not within limit, 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 />.

b.

With any operational Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, primary-to-secondary leakage, and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate 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 STANDBY 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 />.

c.

With any Reactor Coolant System Pressure Isolation Valve Leakage greater than the limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY 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 />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 The Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a.

Monitoring the reactor building atmosphere (gaseous or particulate) radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SUMMER - UNIT 1 3/4 4-19 Amendment No. 154, 165, 179 Serial No.21-147, Page 16 of 18 INSERT A d.

INSERT A

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 /> or be in at least HOT STANDBY 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 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 STANDBY 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 />.

Serial No.21-147, Page 17 of 18

SUMMER - UNIT 1 3/4 4-20 Amendment No. 179, 213 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.

Monitoring the reactor building sump inventory at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 +/- 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

d.

Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.(1) This requirement is not applicable to primary-to-secondary leakage.

e.

Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit.

a.

During startup following each refueling outage, which may be extended to a performance-based frequency not to exceed 3 refueling outages (to a maximum of 60 months) following 2 consecutive satisfactory tests.

b.

Prior to returning the valve to service following maintenance repair or replacement work on the valve.

c.

Prior to entering MODE 2 following valve actuation due to automatic or manual action or flow through the valve for valves denoted on Table 3.4-1 by an asterisk*.

d.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.6.2.3 Primary-to-secondary leakage shall be verified 150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.(1)

(1) Not required to be performed/completed 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.

Serial No.21-147, Page 18 of 18 For Information Only (no changes on page)

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 REVISED TECHNICAL SPECIFICATIONS PAGES Dominion Energy Connecticut, Inc. (DENC)

Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC)

Millstone Power Station Units 2 and 3 North Anna Power Station Units 1 and 2 Surry Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

MILLSTONE - UNIT 2 DEFINITIONS CORE ALTERATION 1.12 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

1.14.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals, and 1.14.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of Leakage Detection Systems; or c.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

1.14.3 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE, and 1.14.4 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all LEAKAGE which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

Amendment No. 38, 263, 280, 299, 1-3 Serial No.21-147, Page 1 of 22

MILLSTONE - UNIT 2 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 value, 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.

Amendment No. 25, 37, 82, 85, 101, 121, 138, 215, 228, 299, 324, 3/4 4-9 Serial No.21-147, Page 2 of 22

MILLSTONE - UNIT 3 DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microCurie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

1.12 DELETED ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

1.14 DELETED FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

LEAKAGE 1.16 LEAKAGE shall be:

1.16.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of Leakage Detection Systems; or Amendment No. 84, 87, 126, 187, 216, 220, 238, 246, 1-3 Serial No.21-147, Page 3 of 22

MILLSTONE - UNIT 3 DEFINITIONS c.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

1.16.3 PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be LEAKAGE (except primary to secondary LEAKAGE) through a fault in a RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE, and 1.16.4 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all LEAKAGE which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

MASTER RELAY TEST 1.17 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include continuity check of each associated slave relay.

MEMBER(S) OF THE PUBLIC 1.18 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

The term REAL MEMBER OF THE PUBLIC means an individual who is exposed to existing dose pathways at one particular location.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

Amendment No. 238, 1-4 Serial No.21-147, Page 4 of 22

MILLSTONE - UNIT 3 DEFINITIONS OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

1.22 DELETED PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3650 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

Amendment No. 60, 187, 188, 238,

242, 1-5 Serial No.21-147, Page 5 of 22

MILLSTONE - UNIT 3 DEFINITIONS SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.31 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

1.37 DELETED Amendment No. 238, 1-6 Serial No.21-147, Page 6 of 22

MILLSTONE - UNIT 3 DEFINITIONS UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTING 1.39 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

1.40 Deleted 1.41 Deleted CORE OPERATING LIMITS REPORT (COLR) 1.42 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Unit Operation within these operating limits is addressed in individual specifications.

1.43 Deleted 1.44 Deleted Amendment No. 39, 50, 60, 72, 100, 189, 268, 273, 1-7 Serial No.21-147, Page 7 of 22

MILLSTONE - UNIT 3 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 the 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.

Amendment No. 209, 238, 3/4 4-22 Serial No.21-147, Page 8 of 22

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION (Continued) d.

With any Reactor Coolant System Pressure Isolation Valve LEAKAGE greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY 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 />.

4.4.6.2.1 Reactor Coolant System operational LEAKAGE shall be demonstrated to be within each of the above limits by:

a.

Deleted b.

Deleted c.

Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2250 20 psia at the frequency specified in the Surveillance Frequency Control Program with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; 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.

d.

Performance of a Reactor Coolant System water inventory balance at the frequency specified in the Surveillance Frequency Control Program; NOTE 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.

e.

Verification that primary to secondary LEAKAGE is 150 gallons per day through any one Steam Generator at the frequency specified in the Surveillance Frequency Control Program, and; f.

Monitoring the Reactor Head Flange Leakoff System at the frequency specified in the Surveillance Frequency Control Program.

Amendment No. 100, 133, 174, 206, 209, 238, 258, 3/4 4-23 Serial No.21-147, Page 9 of 22

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.2.2(1)(2)Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying LEAKAGE to be within its limit:

a.

At the frequency specified in the Surveillance Frequency Control Program, b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, c.

Deleted d.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve, and e.

When tested pursuant to Specification 4.0.5.

(1) The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

(2) This surveillance is not required to be performed on Reactor Coolant System Pressure Isolation Valves located in the RHR flow path when in, or during the transition to or from, the shutdown cooling mode of operation.

Amendment No. 238, 258, 3/4 4-23a Serial No.21-147, Page 10 of 22

North Anna Units 1 and 2 1.1-4 Amendments Definitions 1.1 1.1 Definitions LEAKAGE (continued)

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLEOPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency (continued)

Serial No.21-147, Page 11 of 22

Definitions 1.1 1.1 Definitions North Anna Units 1 and 2 1.1-5 Amendments 258/239 OPERABLEOPERABILITY (continued) electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT RATIO (QPTR)

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2940 MWt.

REACTOR TRIP SYSTEM (RTS) RESPONSE TIME The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

Serial No.21-147, Page 12 of 22

North Anna Units 1 and 2 1.1-6 Amendments 258/239 Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)

A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the (continued)

Serial No.21-147, Page 13 of 22

North Anna Units 1 and 2 1.1-7 Amendments 258/239 Definitions 1.1 1.1 Definitions TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)

(Continued) required setpoint within the necessary accuracy.

The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

Serial No.21-147, Page 14 of 22

RCS Operational LEAKAGE 3.4.13 North Anna Units 1 and 2 3.4.13-1 Amendments 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE;

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 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.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B.

RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Serial No.21-147, Page 15 of 22

RCS Operational LEAKAGE 3.4.13 North Anna Units 1 and 2 3.4.13-2 Amendments C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND C.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

---

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.

In accordance with the Surveillance Frequency Control Program SR 3.4.13.2

---

NOTE--------------------

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.

Verify primary to secondary LEAKAGE is

< 150 gallons per day through any one SG.

In accordance with the Surveillance Frequency Control Program ACTIONS Serial No.21-147, Page 16 of 22

W.

STAGGERED TEST BASIS A staggered test basis shall consist of:

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

X.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems, or; 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE).

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE, and c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

Amendment Nos.

TS 1.0-8 Serial No.21-147, Page 17 of 22

C.

RCS Operational LEAKAGE Applicability The following specifications are applicable to RCS operational LEAKAGE whenever Tavg (average RCS temperature) exceeds 200°F (200 degrees Fahrenheit).

Specifications 1.

RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE, b.

1 gpm unidentified LEAKAGE, c.

10 gpm identified LEAKAGE, and d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

2.a. If pressure boundary LEAKAGE exists, 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.

If the pressure boundary LEAKAGE is not isolated as specified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

3.a. If RCS operational LEAKAGE is not within the limits of 3.1.C.1 for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within the specified limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

If the LEAKAGE is not reduced to within the specified limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

4.

If primary to secondary LEAKAGE is not within the limit specified in 3.1.C.1.d, the unit shall be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

TS 3.1-13 Amendment Nos.

Serial No.21-147, Page 18 of 22

5.

Detected or suspected leakage from the Reactor Coolant System shall be investigated and evaluated. At least two means shall be available to detect reactor coolant system leakage. One of these means must depend on the detection of radionuclides in the containment.

6.a. Prior to going critical all primary coolant system pressure isolation valves listed below shall be functional as a pressure isolation device, except as specified in 3.1.C.6.b. Valve leakage shall not exceed the amounts indicated.

b.

If Specification 3.1.C.6.a cannot be met, an orderly shutdown shall be initiated and the reactor shall be in HOT SHUTDOWN 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 />.

Notes (a) 1.

Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4.

Leakage rates greater than 5.0 gpm are considered unacceptable.

Unit 1 Unit 2 Max. Allowable Leakage (see note (a) below)

Loop A, Cold Leg 1-SI-79, 1-SI-241 2-SI-79, 2-SI-241 5.0 gpm for each valve Loop B, Cold Leg 1-SI-82, 1-SI-242 2-SI-82, 2-SI-242 Loop C, Cold Leg 1-SI-85, 1-SI-243 2-SI-85, 2-SI-243 TS 3.1-13a Amendment Nos.

Serial No.21-147, Page 19 of 22

SUMMER - UNIT 1 1-3 Amendment No. 146,179,221 DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC or the components have been evaluated in accordance with an NRC approved methodology.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b.

Leakage into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems, or

c.

Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

Serial No.21-147, Page 20 of 22

SUMMER - UNIT 1 1-4 Amendment No. 104,117,179 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

Serial No.21-147, Page 21 of 22

SUMMER - UNIT 1 3/4 4-19 Amendment No. 154,165,179 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 from the Reactor Coolant System, and

e.

33 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 +/-

20 psig.

f.

The leakage rate specified for each Reactor Coolant System Pressure Isolation Valve in Table 3.4-1 at a Reactor Coolant System pressure of 2235 +/- 20 psig.

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 /> or be in at least HOT STANDBY 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 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 STANDBY 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 />.

d.

With any Reactor Coolant System Pressure Isolation Valve Leakage greater than the limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY 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 />.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 The Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a.

Monitoring the reactor building atmosphere (gaseous or particulate) radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Serial No.21-147, Page 22 of 22

Serial No.: 21-147 Docket Nos.: 50-336/423/280/281/338/339/395 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP)

Dominion Energy Connecticut, Inc. (DENC)

Dominion Energy Virginia Dominion Energy South Carolina, Inc. (DESC)

Millstone Power Station Units 2 and 3 North Anna Power Station Units 1 and 2 Surry Power Station Units 1 and 2 V. C. Summer Nuclear Station Unit 1

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-FOR INFORMATION ONLY-INSERT B Serial No.21-147, Page 1 of 30

INSERT B Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

Serial No.21-147, Page 2 of 30

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-FOR INFORMATION ONLY-INSERT C Serial No.21-147, Page 3 of 30

INSERT C

a.

If PRESSURE BOUNDARY LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by 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 />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in IDENTIFIED LEAKAGE or UNIDENTIFIED LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

Serial No.21-147, Page 4 of 30

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-FOR INFORMATION ONLY-Serial No.21-147, Page 5 of 30

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM BASES 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS (Continued)

In this case, CVLKR2 and CVLKR3I monitor flow rate by comparing level indications on the containment drains sump when Pumps P10, P2A, P2B and P1 are not running.

2.

The plant process computer calculates a leakage rate every 30 seconds when 3DAS-P10, 3DAS-P1, 3DAS-P2A and 3DAS-P2B indicate stop. This leakage rate is displayed via computer points CVLKR3I and CVLKR2 when 3DAS-P10 is off and when the unidentified leakage sump is overflowing to the containment drains sump. When one of these pumps does run, the leakage rate calculation is stopped and resumes 10 minutes after all pumps stop. If it cannot provide value of the leakage rate within any 54 minute interval, two computer point alarms (CVDASP2NC, UNDNT LKG RT NOT CALC and CVDASP2NC, SMP 3 LKG RT NT CALC) are generated which alerts the Operator that UNIDENTIFIED LEAKAGE cannot be determined.

3.

This monitoring system is not seismic Category I, but is expected to remain OPERABLE during an OBE. If the monitoring system is not OPERABLE following a seismic event, the appropriate ACTION according to Technical Specifications will be taken.

4.

Two priority computer alarms (CVLKR2 and CVLKR3I) are generated if the calculated leakage rate is greater than a value specified on the Priority Alarm Point Log. This alarm value should be set to alert the Operators to a possible RCS leak rate in excess of the Technical Specification maximum allowed UNIDENTIFIED LEAKAGE. The alarm value may be set at one gallon per minute or less above the rate of IDENTIFIED LEAKAGE, from the reactor coolant or auxiliary systems, into the containment drains sump. The rate of IDENTIFIED LEAKAGE may be determined by either measurement or by analysis. If the Priority Alarm Point Log is adjusted, the high leakage rate alarm will be bounded by the IDENTIFIED LEAKAGE rate and the low leakage rate alarm will be set to notify the operator that a decrease in leakage may require the high leakage rate alarm to be reset. The priority alarm setpoint shall be no greater than 2 gallons per minute. This ensures that the IDENTIFIED LEAKAGE will not mask a small increase in UNIDENTIFIED LEAKAGE that is of concern. The 2 gallons per minute limit is also within the containment drains sump level monitoring system alarm operating range which has a maximum setpoint of 2.5 gallons per minute.

5.

To convert containment drains sump run times to a leakage rate, refer to procedure SP3670.1 for guidance on the conversion method.

3/4.4.6.2 OPERATIONAL LEAKAGE LCO RCS operational LEAKAGE shall be limited to:

a. PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is prohibited allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further LBDCR No. 07-MP3-032 August 8, 2007 Amendment No.

B 3/4 4-4c

-FOR INFORMATION ONLY-Serial No.21-147, Page 6 of 30

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

RCPB deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b. UNIDENTIFIED LEAKAGE One gallon per minute (gpm) of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Primary to Secondary LEAKAGE through Any One Steam Generator (SG)

The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Reference 4).

The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary LEAKAGE through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational LEAKAGE rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

d. IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of UNIDENTIFIED LEAKAGE and is well within the capability of the RCS makeup system. IDENTIFIED LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (CONTROLLED LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

e. CONTROLLED LEAKAGE The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2250 psia. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the safety analyses.

A limit of 40 gpm is placed on CONTROLLED LEAKAGE.

f. RCS Pressure Isolation Valve LEAKAGE The specified allowable leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure.

LBDCR No. 07-MP3-032 August 8, 2007 Amendment No. 209, B 3/4 4-4d

-FOR INFORMATION ONLY-INSERT B Serial No.21-147, Page 7 of 30

INSERT B Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

Serial No.21-147, Page 8 of 30

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.6.2.f, RCS Pressure Isolation Valve (PIV) Leakage, measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS b c., c d.

UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE or RCS pressure isolation valve LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

a b., b c., c d.

If any PRESSURE BOUNDARY LEAKAGE cannot be isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> exists, or primary to secondary LEAKAGE is not within limits, or if a UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, or RCS pressure isolation valve LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. The reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In COLD SHUTDOWN, the pressure stresses acting on the reactor coolant pressure boundary are much lower, and further deterioration is much less likely.

LBDCR No. 06-MP3-005 May 25, 2006 Amendment No. 209, B 3/4 4-4e

-FOR INFORMATION ONLY-INSERT C Serial No.21-147, Page 9 of 30

INSERT C

a.

If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by 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 />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

Serial No.21-147, Page 10 of 30

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REQUIREMENTS 4.4.6.2.1.c CONTROLLED LEAKAGE is determined under a set of reference conditions, listed below:

a.

One Charging Pump in operation.

b.

RCS pressure at 2250 +/- 20 psia.

By limiting CONTROLLED LEAKAGE to 40 gpm during normal operation, it can be assured that during an SI with only one charging pump injecting, RCP seal injection flow will continue to remain less than 80 gpm as assumed in the accident analysis. When the seal injection throttle valves are set with a normal charging lineup, the throttle valve position bounds conditions where higher charging header pressures could exist. Therefore, conditions which create higher charging header pressures such as an isolated charging line, or two pumps in service are bounded by the single pump-normal system lineup surveillance configuration. Basic accident analysis assumptions are that 80 gpm flow is provided to the seals by a single pump in a runout condition.

4.4.6.2.1.d Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the reactor coolant pressure boundary is maintained. PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The Surveillance is modified by two Notes.

Note 1 states that this SR is 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 establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper water inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

LBDCR No. 06-MP3-005 May 25, 2006 Amendment No.

B 3/4 4-4f

-FOR INFORMATION ONLY-Serial No.21-147, Page 11 of 30

MILLSTONE - UNIT 3 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in RCS LCO 3.4.6.1, Leakage Detection Systems.

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The surveillance frequency is controlled under the Surveillance Frequency Control Program.

4.4.6.2.1.e This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.5, Steam Generator Tube Integrity, should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the surveillance is 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. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The surveillance frequency is controlled under the Surveillance Frequency Control Program. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 5).

4.4.6.2.2 The Surveillance Requirements for RCS pressure isolation valves provide assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

LBDCR 12-MP3-010 September 20, 2012 Amendment No.

B 3/4 4-4g

-FOR INFORMATION ONLY-Serial No.21-147, Page 12 of 30

North Anna Units 1 and 2 B 3.4.13-1 Revision 0 RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

General Design Criteria 3 30(Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

FOR INFORMATION ONLY Serial No.21-147, Page 13 of 30

RCS Operational LEAKAGE B 3.4.13 BASES North Anna Units 1 and 2 B 3.4.13-3 Revision 28 LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No P pressure boundary LEAKAGE is prohibited allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further RCPB deterioration, resulting in higher LEAKAGE.

Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System.

Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage (continued)

FOR INFORMATION ONLY INSERT B Serial No.21-147, Page 14 of 30

INSERT B Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

Serial No.21-147, Page 15 of 30

North Anna Units 1 and 2 B 3.4.13-4 Revision 28 RCS Operational LEAKAGE B 3.4.13 BASES LCO

d. Primary to Secondary LEAKAGE through Any One SG (continued) rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.14, RCS Pressure Isolation Valve (PIV) Leakage, measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS A.1 B.1 Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

C B.1 and C B.2 If any pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limit, or if unidentified LEAKAGE, or identified LEAKAGE, cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

The reactor must be brought to MODE 3 (Continued)

FOR INFORMATION ONLY within any of the Required Actions and associated Completion Times cannot be met, INSERT C Serial No.21-147, Page 16 of 30

INSERT C If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by 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 />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

Serial No.21-147, Page 17 of 30

RCS Operational LEAKAGE B 3.4.13 BASES North Anna Units 1 and 2 B 3.4.13-5 Revision 28 ACTIONS C B.1 and C B.2 (continued) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

SURVEILLANCE REQUIREMENTS SR 3.4.13.1 Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The surveillance is modified by two Notes. Note 1 states that this SR is 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 establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and (continued)

FOR INFORMATION ONLY Serial No.21-147, Page 18 of 30

North Anna Units 1 and 2 B 3.4.13-6 Revision 46 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS SR 3.4.13.1 (continued) the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

These leakage detection systems are specified in LCO 3.4.15, RCS Leakage Detection Instrumentation.

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG.

Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.20, Steam Generator Tube Integrity, should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note, which states that the Surveillance is 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. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

FOR INFORMATION ONLY Serial No.21-147, Page 19 of 30

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES - Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is 1 gpm or increases to 1 gpm as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a main steam line break (MSLB) accident. Other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The UFSAR (Ref. 2) analysis for SGTR assumes the contaminated secondary fluid is released via power operated relief valves or safety valves. The source term in the primary system coolant is transported to the affected (ruptured) steam generator by the break flow. The affected steam generator discharges steam to the environment for 30 minutes until the generator is manually isolated. The 1 gpm primary to secondary LEAKAGE transports the source term to the unaffected steam generators. Releases continue through the unaffected steam generators until the Residual Heat Removal System is placed in service.

The MSLB is less limiting for site radiation releases than the SGTR. The safety analysis for the MSLB accident assumes 1 gpm total primary to secondary LEAKAGE, including 500 gpd leakage into the faulted generator. The dose consequences resulting from the MSLB and the SGTR accidents are within the limits defined in the plant licensing basis.

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LIMITING CONDITIONS FOR OPERATION - RCS operational LEAKAGE shall be limited to:

a.

Pressure Boundary LEAKAGE No Ppressure boundary LEAKAGE is prohibited allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further RCPB deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

TS 3.1-14a 11-05-09 Amendment Nos. 267 and 250 FOR INFORMATION ONLY Serial No.21-147, Page 20 of 30 TS Bases Markup for Surry Units 1 and 2

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary. Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

c. Primary to Secondary LEAKAGE through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 3). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY - In REACTOR OPERATION conditions where Tavg exceeds 200°F, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In COLD SHUTDOWN and REFUELING SHUTDOWN, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.1.C.56 measures leakage through each individual pressure isolation valve (PIV) and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leaktight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

TS 3.1-14b 11-05-09 Amendment Nos. 267 and 250 FOR INFORMATION ONLY Serial No.21-147, Page 21 of 30 TS Bases Markup for Surry Units 1 and 2

ACTIONS 3.1.C.2.a If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by 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 />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

3.1.C.2 3.a Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This completion time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

3.1.C.2.b, 3.1.C.3.b, and 3.1.C.3 4 If any pressure boundary LEAKAGE cannot be isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> exists, or primary to secondary LEAKAGE is not within limits, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. The reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed completion times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In COLD SHUTDOWN, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

REFERENCES 1.

UFSAR, Chapter 4, Surry Units 1 and 2.

2.

UFSAR, Chapter 14, Surry Units 1 and 2.

3.

NEI 97-06, Steam Generator Program Guidelines.

4.

EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines.

TS 3.1-15 03-29-07 Amendment Nos. 251 and 250 FOR INFORMATION ONLY Serial No.21-147, Page 22 of 30 TS Bases Markup for Surry Units 1 and 2

4.13 RCS OPERATIONAL LEAKAGE Applicability The following specifications are applicable to RCS operational LEAKAGE whenever Tavg (average RCS temperature) exceeds 200°F (200 degrees Fahrenheit).

Objective To verify that RCS operational LEAKAGE is maintained within the allowable limits, the following surveillances shall be performed at the frequencies specified in the Surveillance Frequency Control Program.

Specifications A.

Verify RCS operational LEAKAGE is within the limits specified in TS 3.1.C by performance of RCS water inventory balance.1, 2 B.

Verify primary to secondary LEAKAGE is 150 gallons per day through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.1 Notes:

1.

Not required to be completed 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.

BASES SURVEILLANCE REQUIREMENTS (SR)

SR 4.13.A Verifying RCS LEAKAGE to be within the Limiting Condition for Operation (LCO) limits ensures the integrity of the reactor coolant pressure boundary (RCPB) is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions (stable pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The surveillance is modified by two notes.

Note 1 states that this SR is not required to be completed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable unit conditions are established.

TS 4.13-1 04-17-12 Amendment Nos. 277 and 277 FOR INFORMATION ONLY Serial No.21-147, Page 23 of 30 TS Bases Markup for Surry Units 1 and 2

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in the TS 3.1.C Bases.

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 4.13.B This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.1.H, Steam Generator Tube Integrity, should be evaluated.

The 150 gallons per day limit is measured at room temperature as described in Reference 4. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG.

If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG. The surveillance is modified by a Note, which states that the Surveillance is 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. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRl guidelines (Ref. 4). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

TS 4.13-2 04-29-11 Amendment Nos. 273 and 272 FOR INFORMATION ONLY Serial No.21-147, Page 24 of 30 TS Bases Markup for Surry Units 1 and 2

REACTOR COOLANT SYSTEM SUMMER - UNIT 1 B 3/4 4-4e Amendment No. BRN-07-001, BRN-12-001 BASES OPERATIONAL LEAKAGE (Continued)

Limiting Condition for Operation (LCO)

Reactor Coolant System operational leakage shall be limited to:

a.

PRESSURE BOUNDARY LEAKAGE NoPRESSURE BOUNDARY LEAKAGE is prohibited allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further RCPB deterioration, resulting in higher leakage. Violation of this LCO could result in continued degradation of the Reactor Coolant Pressure Boundary. Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFED LEAKAGE One gallon per minute (gpm) of UNIDENTIFED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the Reactor Coolant Pressure Boundary, if the leakage is from the pressure boundary.

c.

Primary-to-Secondary Leakage Through Any One Steam Generator The limit of 150 gallons per day (gpd) per steam generator is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines (Reference 1). The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, The RCS operational primary-to-secondary leakage through any one steam generator shall be limited to 150 gpd. The limit is based on operating experience with steam generator tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

d.

IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFED LEAKAGE and is well with the capability of the Reactor Coolant System Makeup System. IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or CONTROLLED LEAKAGE. Violation of this LCO could result in continued degradation of a component or system.

e.

CONTROLLED LEAKAGE The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 33 gpm with the modulating valve in the supply line fully open at a nominal RCS reassure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analysis.

INSERT B Serial No.21-147, Page 25 of 30

INSERTB Separating the sources of leakage (i.e., leakage from an identified source versus leakage from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action.

Serial No.21-147, Page 26 of 30

REACTOR COOLANT SYSTEM SUMMER - UNIT 1 B 3/4 4-4f Amendment No. BRN-07-001, BRN-12-001 BASES OPERATIONAL LEAKAGE (Continued) f.

Reactor Coolant System Pressure Isolation Valve Leakage 10CFR50.2, 10CFR50.55a(c), and GDC 55 of 10CFR50, Appendix A define RCS PIVs as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB) which separate the high pressure RCS from an attached low pressure system.

During their service lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV leakage LCO allows leakage through these valves in amounts that do not compromise safety.

The Reactor Coolant System Pressure Isolation Valve (PIV) Leakage limit applies to each individual valve. Leakage through both series PIVs in a line must be included as part of IDENTIFIED LEAKAGE governed by LCO 3.4.6.2, Reactor Coolant System Operational Leakage. This is true during operation only when the loss of RCS mass through two series valves is determined by water inventory balance (SR 4.4.6.2.1.d). A known component of the identified leakage before operation begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing. Leakage measured through one PIV in a line is not RCS operational LEAKAGE if the other PIV is leaktight.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low-pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting system are degraded or degrading. Excessive PIV leakage could lead to overpressure of the low-pressure piping or components, potentially resulting in a loss of coolant accident (LOCA) outside of containment.

The PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The NRC, through NUREG-1431, has endorsed this PIV leakage rate limit.

The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

Leakage from the RCS Pressure Isolation Valves may be identified by surveillance testing performed during plant heatup or cooldown above 2000 psig and may be adjusted to obtain the leakage value at 2235 +/- 20 psig using calculation guidance provided by the ASME OM Code.

Applicability In MODES 1, 2, 3, and 4, the potential for Reactor Coolant Pressure Boundary leakage is greatest when the Reactor Coolant System is pressurized.

In MODES 5 and 6, leakage limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for leakage.

Serial No.21-147, Page 27 of 30

REACTOR COOLANT SYSTEM SUMMER - UNIT 1 B 3/4 4-4g Amendment BRN-07-001, BRN-12-001 BASES OPERATIONAL LEAKAGE (Continued)

Actions

a.

If any PRESSURE BOUNDARY LEAKAGE exists, or primary-to-secondary leakage is not within limit, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This ACTION reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the Reactor Coolant Pressure Boundary are much lower, and further deterioration is much less likely.

b.

Any operational leakage, excluding PRESSURE BOUNDARY LEAKAGE and primary-to-secondary leakage, in excess of the LCO limits must be reduced to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactor must be shut down. This ACTION is necessary to prevent further deterioration of the Reactor Coolant Pressure Boundary.

c.

With PIV leakage in excess of the limit, the high pressure portion of the affected system must be isolated 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 standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 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 ACTION is necessary to prevent over pressurization of low pressure systems, and the potential for intersystem LOCA.

Surveillance Requirements 4.4.6.2.1 Verifying Reactor Coolant System leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary is maintained.

PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance of a Reactor Coolant System water inventory balance.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure. Therefore, the Surveillance is modified by a note. The note states that this Surveillance Requirement is 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.

For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and Reactor Coolant Pump seal injection and return flows.

INSERT C

a. & b.

c.

d.

limits cannot be isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Serial No.21-147, Page 28 of 30

INSERTC

a.

If PRESSURE BOUNDARY LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by 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 />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal leakage past the isolation device is acceptable as it will limit RCS leakage and is included in IDENTIFIED LEAKAGE or UNIDENTIFIED LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.

Serial No.21-147, Page 29 of 30

REACTOR COOLANT SYSTEM SUMMER - UNIT 1 B 3/4 4-4h Amendment No. BRN-07-001, RN-12-001, 213 BASES OPERATIONAL LEAKAGE (Continued)

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor containment atmosphere radioactivity and containment sump level. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in LCO 3.4.6.1, Reactor Coolant System, Leakage Detection Systems.

Part (d) notes that this SR is not applicable to primary-to-secondary leakage because leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The 72-hour frequency is a reasonable interval to trend leakage and recognizes the importance of early leakage detection in the prevention of accidents.

4.4.6.2.2 This Surveillance Requirement verifies RCS Pressure Isolation Valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Pressure Isolation Valve (PIV) testing is being conducted on a performance-based testing interval.

Performance-based is defined as a PIV which demonstrates good performance for two consecutive outages and may have its test interval extended to every third refueling outage (to a maximum of 60 months). A PIV, which has demonstrated leakage less than its TS allowable leakage limit for two consecutive outages is classified as a good performer. In the event of a PIV leakage test failure, PIV testing would require the component to return to the initial interval of every refueling outage, or two years until good performance is re-established.

Those PIVs listed in TS Table 3.4-1 will be demonstrated OPERABLE by verifying their leakage rates are within TS allowable leakage limits via testing performed at a frequency based on their good performance, ranging from every refueling outage to every third refueling outage (to a maximum of 60 months).

4.4.6.2.3 This Surveillance Requirement verifies that primary-to-secondary leakage is less than or equal to 150 gpd through any one steam generator. Satisfying the primary-to-secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If this Surveillance Requirement is not met, compliance with LCO 3.4.5 should be evaluated. The 150-gpd limit is measured at room temperature as described in Reference 2. The operational leakage rate limit applies to leakage through any one steam generator. If it is not practical to assign the leakage to an individual steam generator, all the primary-to-secondary leakage should be conservatively assumed to be from one steam generator.

The Surveillance Requirement is modified by a note, which states that the Surveillance is 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. For Reactor Coolant System primary-to-secondary leakage determination, steady state is defined as stable Reactor Coolant System pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and reactor coolant pump seal injection and return flows.

The frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary-to-secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents. The primary-to-secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 2).

References 1.

NEI 97-06, Steam Generator Program Guidelines 2.

EPRI TR-104788, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines Serial No.21-147, Page 30 of 30