ML17152A218

Southern Nuclear Fleet - Issuance of Amendments Regarding the Adoption of TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing (CAC Nos. MF8176 - MF8181)

ML17152A218
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	06/30/2017
From:	Michael Orenak Plant Licensing Branch II
To:	Hutto J Southern Nuclear Operating Co
Orenak M, 415-3229
References
CAC MF8176, CAC MF8177, CAC MF8178, CAC MF8179, CAC MF8180, CAC MF8181
Download: ML17152A218 (78)

Text

Mr. James J. Hutto Regulatory Affairs Director UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 30, 2017 Southern Nuclear Operating Company, Inc.

P.O. Box 1295 /Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2; VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2; AND EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING THE ADOPTION OF TSTF-545, REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING" (CAC NOS. MF8176, MF8177, MF8178, MF8179, MF8180, AND MF8181)

Dear Mr. Hutto:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 212 to the Joseph M. Farley Nuclear Plant (FNP) Unit 1, Renewed Facility Operating License No. NPF-2; Amendment No. 209 to FNP, Unit 2, Renewed Facility Operating License No. NPF-8; Amendment No. 187 to the Vogtle Electric Generating Plant (VEGP), Unit 1, Renewed Facility Operating License NPF-68; Amendment No. 170 to VEGP, Unit 2, Renewed Facility Operating License NPF-81; Amendment No. 286 to the Edwin I. Hatch Nuclear Plant (HNP), Unit No. 1, Renewed Facility Operating License DPR-57; and Amendment No. 231 to HNP, Unit No. 2, Renewed Facility Operating License NPF-5.

The amendments are in response to your application dated July 28, 2016. The amendments consist of modifications consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing," dated October 21, 2015.

J. A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket Nos. 50-348, 50-364, 50-424, 50-425, 50-321, and 50-366

Enclosures:

1. Amendment No. 212 to NPF-2

2. Amendment No. 209 to NPF-8

3. Amendment No. 187 to NPF-68

4. Amendment No. 170 to NPF-81

5. Amendment No. 286 to DPR-57

6. Amendment No. 231 to NPF-5

7. Safety Evaluation cc w/enclosures: Distribution via Listserv Michael D. Orenak, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 212 Renewed License No. NPF-2

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 1, Renewed Facility Operating License No. NPF-2, filed by Southern Nuclear Operating Company, Inc. (the licensee), dated July 28, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: June 30, 201 7 FOR THE NUCLEAR REGULATORY COMMISSION

~?~

Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 209 Renewed License No. NPF-8

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility Operating License No. NPF-8, filed by Southern Nuclear Operating Company, Inc. (the licensee), dated July 28, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 209, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: June 30, 201 7 FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NOS. 212 AND 209 JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-2 AND NPF-8 DOCKET NOS. 50-348 AND 50-364 Replace the following pages of the License and Appendix "A" Technical Specifications (TSs) with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages License License No. NPF-2, page 4 License No. NPF-8, page 3 TSs 1.1-3 3.4.10-2 3.4.12-4 3.5.2-2 3.6.3-6 3.6.6-3 3.7.1-2 3.7.2-2 3.7.3-2 3.7.5-3 5.5-5 Insert Pages License License No. NPF-2, page 4 License No. NPF-8, page 3 TSs 1.1-3 3.4.10-2 3.4.12-4 3.5.2-2 3.6.3-6 3.6.6-3 3.7.1-2 3.7.2-2 3.7.3-2 3.7.5-3 3.7.5-4 5.5-5 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3)

Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the Issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

a. Southern Nuclear shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation.

b. Deleted per Amendment 13

c.

Deleted per Amendment 2

d. Deleted per Amendment 2

e. Deleted per Amendment 152 Deleted per Amendment 2

f.

Deleted per Amendment 158

g. Southern Nuclear shall maintain a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

Farley - Unit 1

1) Identification of a sampling schedule for the critical parameters and control points for these parameters;

2) Identification of the procedures used to quantify parameters that are critical to control points;

3) Identification of process sampling points;

4) A procedure for the recording and management of data;

5) Procedures defining corrective actions for off control point chemistry conditions; and Renewed License No. NPF-2 Amendment No. 212 (2)

Alabama Power Company, pursuant to Section 103 of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess but not operate the facility at the designated location in Houston County, Alabama in accordance with the procedures and limitations set forth in this renewed license.

(3)

Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproducts, source or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2775 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 209, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

(3)

Delete per Amendment 144

( 4)

Delete Per Amendment 149 (5)

Delete per Amend 144 Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 209

1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Farley Units 1 and 2 Definitions 1.1 The 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.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

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 not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3.

Reactor Coolant System (RCS) LEAKAGE thrqugh a steam generator (SG) to the Secondary System;

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; 1.1-3 (continued)

Amendment No. 212 (Unit 1)

Amendment No. 209 (Unit 2)

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within+/- 1%.

FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Farley Units 1 and 2 3.4.10-2 Amendment No.212 (Unit 1)

Amendment No. 209 (Unit 2)

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.12.1 Verify a maximum of one charging pump is capable of injecting into the RCS when one or more RCS cold legs is s 180°F.

SR 3.4.12.2 Verify a maximum of two charging pumps are capable of injecting into the RCS when all RCS cold legs are> 180°F.

SR 3.4.12.3 Verify each accumulator is isolated.

SR 3.4.12.4 Verify RHR suction isolation valves are open for each required RHR suction relief valve.

SR 3.4.12.5

---

N()TE--------------------------------

Only required to be met when complying with LCO 3.4.12.b.

Verify RCS vent ;;::: 2.85 square inches open.

SR 3.4.12.6 Verify each required RHR suction relief valve setpoint.

LTOP System 3.4.12 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM AND In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.4.12-4 Amendment No. 212 (Unit 1}

Amendment No. 209 (Unit 2}

SURVEILLANCE REQUIREMENTS SR 3.5.2.1 SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.4 SURVEILLANCE

---

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

Only required to be performed for valves 8132A and 81328 when Centrifugal Charging Pump A is inoperable.

Verify the following valves are in the listed position with power to the valve operator removed.

Number 8884, 8886 8132A,8132B 8889 Position Closed Open Closed Function Centrifugal Charging Pump to RCS Hot Leg Centrifugal Charging Pump discharge isolation RHR to RCS Hot Leg Injection

---

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

Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

ECCS -

Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.5.2-2 Amendment No.212 (Unit 1)

Amendment No.209 (Unit 2)

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SR 3.6.3.3 SR 3.6.3.4 SR 3.6.3.5 SR 3.6.3.6 SURVEILLANCE

---

N()TES--------------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

The blind flange on the fuel transfer canal flange is only required to be verified closed after each draining of the canal.

Verify each containment isolation manual valve and blind flange that is located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Verify the isolation time of each automatic power operated containment isolation valve in the INSERVICE TESTING PR()GRAM is within limits.

Perform leakage rate testing for containment penetrations containing containment purge valves with resilient seals.

Verify each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

FREQUENCY Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program AND Within 92 days after opening the valve In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.6.3-6 Amendment No. 212 (Unit 1)

Amendment No. 209 (Unit 2)

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.2 Operate each required containment cooling train fan In accordance with unit for 2: 15 minutes.

the Surveillance Frequency Control Program SR 3.6.6.3 Verify each containment cooling train cooling water In accordance with flow rate is 2: 1600 gpm.

the Surveillance Frequency Control Program SR 3.6.6.4 Verify each containment spray pump's developed In accordance with head at the flow test point is greater than or equal to the INSERVICE the required developed head.

TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment spray valve in the In accordance with flow path that is not locked, sealed, or otherwise the Surveillance secured in position, actuates to the correct position Frequency Control on an actual or simulated actuation signal.

Program SR 3.6.6.6 Verify each containment spray pump starts In accordance with automatically on an actual or simulated actuation the Surveillance signal.

Frequency Control Program SR 3.6.6.7 Verify each containment cooling train starts In accordance with automatically on an actual or simulated actuation the Surveillance signal.

Frequency Control Program SR 3.6.6.8 Verify each spray nozzle is unobstructed.

In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.6.6-3 Amendment No. 212 (Unit 1)

Amendment No. 209 (Unit 2)

ACTIONS CONDITION REQUIRED ACTION

8.

(continued) 8.2

---

NOTE:------------

Only required in MODE 1.

Reduce the Power Range Neutron Flux-High reactor trip setpoint to less than or equal to the Maximum Allowable % RTP specified in Table 3.7.1-1 for the number of.

OPERABLE MSSVs.

C.

Required Action and C. 1 Be in MODE 3.

associated Completion Time not met.

AND OR C.2 Bein MODE4.

One or more steam generators with ;::. 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SR 3.7.1.1 SURVEILLANCE

---

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

On ly required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per Table 3.7.1-2 in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift setting shall be within +/-1%.

MSSVs 3.7.1 COMPLETION TIME 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 6 hours 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Farley Units 1 and 2 3.7.1-2 Amendment No.212 (Unit 1)

Amendment No. 209 (Unit 2)

ACTIONS CONDITION REQUIRED ACTION E.

One or more steam lines E.1 Verify one MSIV closed in with two MSIVs affected steam line.

inoperable in MODE 2 or 3.

F.

Required Action and F.1 Be in MODE 3.

associated Completion Time of Condition D or E AND not met.

F.2 Be in MODE 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE MS IVs 3.7.2 COMPLETION TIME 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND Once per 7 days thereafter 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours FREQUENCY SR 3.7.2.1

---

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

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is s 7 seconds.

In accordance with the INSERVICE TESTING PROGRAM Farley Units 1 and 2 3.7.2-2 Amendment No. 212 (Unit 1)

Amendment No. 209 (Unit 2)

Main FW Stop Valves and MFRVs and Associated Bypass Valves 3.7.3 ACTIONS CONDITION REQUIRED ACTION C.

One or more MFRV C.1 Close or isolate bypass bypass valves valve.

inoperable.

AND C.2 Verify bypass valve is closed or isolated.

D.

Two valves in the same D.1 Isolate affected flow path.

flow path inoperable.

E.

Required Action and E.1 Bein MODE 3.

associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SR 3.7.3.1 SURVEILLANCE Verify the closure time of each Main FW Stop Valve, MFRV, and associated bypass valve is in accordance with the time requirement in the INSERVICE TESTING PROGRAM.

COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Once per 7 days 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6 hours FREQUENCY In accordance with the INSERVICE TESTING PROGRAM.

Farley Units 1 and 2 3.7.3-2 Amendment No.212 (Unit 1)

Amendment No. 209 (Unit 2)

SURVEILLANCE REQUIREMENTS SR 3.7.5.2 SR 3.7.5.3 SR 3.7.5.4 SURVEILLANCE

---

N()TE---------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after ;::::. 1005 psig in the steam generator.

Verify the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head.

---

N()TE---------------------------------

AFW train(s) may be considered ()PERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

---

N()TES---------------------------------

1.

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after ;::::. 1005 psig in the steam generator.

2.

AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

Verify each AFW pump starts automatically on an actual or simulated actuation signal.

AFWSystem 3.7.5 FREQUENCY In accordance with the INSERVICE TESTING PR()GRAM.

In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Farley Units 1 and 2 3.7.5-3 Amendment No. 212 (Unit 1)

Amendment No. 209 (Unit 2)

AFWSystem 3.7.5 SURVEILLANCE REQUIREMENTS SR 3.7.5.5 SURVEILLANCE FREQUENCY Verify the turbine driven AFW pump steam admission In accordance valves open when air is supplied from their respective with the air accumulators.

Surveillance Frequency Control Program Farley Units 1 and 2 3.7.5-4 Amendment No. 212 (Unit 1)

Amendment No. 209 (Unit 2)

5.5 Programs and Manuals Programs and Manuals 5.5 5.5. 7 Reactor Coolant Pump Flywheel Inspection Program (continued)

b.

A surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the disassembled flywheel.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program.

5.5.8 Not Used 5.5.9 Steam Generator (SG> Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a.

Provisions tor condition monitoring assessments. Condition monitoring Farley Units 1 and 2 5.5-5 (continued)

Amendment No.212 (Unit 1)

Amendment No.209 (Unit 2)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DAL TON, GEORGIA DOCKET NO. 50-424 VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 187 Renewed License No. NPF-68

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia (the owners), dated July 28, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-68 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 187, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: June 30, 20 I 7 FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DAL TON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 170 Renewed License No. NPF-81

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia (the owners), dated July 28, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 1 O CFR Chapter I;

8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-81 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 170, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: June 30, 2017 FOR THE NUCLEAR REGULATORY COMMISSION

~<~

Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NOS. 187 AND 170 VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. NPF-68 AND NPF-81 DOCKET NOS. 50-424 AND 50-425 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages License License No. NPF-68, page 4 License No. NPF-81, page 3 TSs 1.1-3 3.4.10-2 3.4.14-3 3.5.2-2 3.6.3-5 3.6.6-2 3.7.1-2 3.7.2-2 3.7.3-2 3.7.9-3 5.5-6 Insert Pages License License No. NPF-68, page 4 License No. NPF-81, page 3 TSs 1.1-3 3.4.10-2 3.4.14-3 3.5.2-2 3.6.3-5 3.6.6-2 3.7.1-2 3.7.2-2 3.7.3-2 3.7.9-3 5.5-6 (1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 187, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4)

(5)

(6)

(7)

(8)

(9)

(10)

Deleted Deleted Deleted Deleted Deleted Deleted Mitigation Strategl'. License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training and response personnel (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for Implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy Renewed Operating License NPF-68 Amendment No. 187 (2)

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, pursuant to the Act and 10 CFR Part 50, to possess but not operate the facility at the designated location in Burke County, Georgia, in accordance with the procedures and limitations set forth in this license; (3)

Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as my be produced by the operation of the facility authorized herein.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 170 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 74. The SRs listed below shall be Renewed Operating License NPF-81 Amendment No. 170

1. 1 Definitions (continued)

E-AVERAGE DISINTEGRATION ENERGY ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Vogtle Units 1 and 2 E shall be the average (weighted in proportion to Definitions 1.1 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 > 14 minutes, making up at least 95% of the total noniodine activity in the coolant.

The 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.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

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 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);

1.1-3 (continued)

Amendment No. 187 (Unit 1)

Amendment No. 170 (Unit 2)

SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within+/- 1%.

Vogtle Units 1 and 2 3.4.10-2 Pressurizer Safety Valves 3.4.10 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Amendment No. 187 (Unit 1)

Amendment No. 170 (Unit 2)

SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SURVEILLANCE

---

N()TES----------------------------

1.

Not required to be performed in M()DES 3 and 4.

2.

Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3.

RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is equivalent to ~ 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure

~ 2215 psig and ~ 2255 psig.

RCS PIV Leakage 3.4.14 FREQUENCY In accordance with the INSERVICE TESTING PR()GRAM, and 18 months Prior to entering M()DE 2 whenever the unit has been in M()DE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months (except for valves HV-8701A/B and HV-8702A/B) continued Vogtle Units 1 and 2 3.4.14-3 Amendment No.187 (Unit 1)

Amendment No.170 (Unit 2)

SURVEILLANCE REQUIREMENTS SR 3.5.2.1 Valve Number HV-8835 HV-8840 HV-8813 HV-8806 HV-8802A, B HV-8809A, B SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 SURVEILLANCE Verify the following valves are in the listed position with the power lockout switches in the lockout position.

Valve Function SI Pump Cold Leg lnj.

RHR Pump Hot Leg lnj.

SI Pump Mini Flow lsol.

SI Pump Suction from RWST SI Pump Hot Leg lnj.

RHR Pump Cold Leg lnj.

Valve Position OPEN CLOSED OPEN OPEN CLOSED OPEN

---

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

Not required to be met for system vent flow paths opened under administrative control Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Verify ECCS locations susceptible to gas accumulation are sufficiently filled with water.

Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position actuates to the correct position on an actual or simulated actuation signal.

ECCS - Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with*

the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program (continued)

Vogtle Units 1 and 2 3.5.2-2 Amendment No.187 (Unit 1)

Amendment No. 170 (Unit 2)

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.3.4 SR 3.6.3.5 SR 3.6.3.6 SR 3.6.3.7 SURVEILLANCE

---

N()TES----------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

The fuel transfer tube blind flange is only required to be verified closed once after refueling prior to entering M()DE 4 from M()DE 5.

Verify each containment isolation manual valve and blind flange that is located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Verify the isolation time of each automatic power operated containment isolation valve is within limits.

Perform leakage rate testing for containment purge valves with resilient seals.

Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

FREQUENCY Prior to entering M()DE 4 from M()DE 5 if not performed within the previous 92 days In accordance with the INSERVICE TESTING PR()GRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Vogtle Units 1 and 2 3.6.3-5 Amendment No. 187 (Unit 1)

Amendment No. 170 (Unit 2)

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1

---

NC>TE-------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each containment spray manual, power In accordance with operated, and automatic valve in the flow path the Surveillance that is not locked, sealed, or otherwise secured in Frequency Control position is in the correct position.

Program SR 3.6.6.2 C>perate each containment cooling train fan unit In accordance with for ~ 15 minutes.

the Surveillance Frequency Control Program SR 3.6.6.3 Verify each pair of containment fan coolers In accordance with cooling water flow rate is ;;::: 1359 gpm.

the Surveillance Frequency Control Program SR 3.6.6.4 Verify each containment spray pump's developed In accordance with head at the flow test point is greater than or equal the INSERVICE to the required developed head.

TESTING PRC>G RAM SR 3.6.6.5 Verify each automatic containment spray valve in In accordance with the flow path that is not locked, sealed, or the Surveillance otherwise secured in position actuates to the Frequency Control correct position on an actual or simulated Program actuation signal.

SR 3.6.6.6 Verify each containment spray pump starts In accordance with automatically on an actual or simulated actuation the Surveillance signal.

Frequency Control Program (continued)

Vogtle Units 1 and 2 3.6.6-2 Amendment No. 187 (Unit 1)

Amendment No. 170 (Unit 2)

ACTIONS (continued)

CONDITION REQUIRED ACTION

8.

Required Action and 8.1 Be in MODE 3.

associated Completion Time not met.

AND OR 8.2 Bein MODE4.

One or more steam generators (SG) with four or more MSSVs per SG inoperable.

SURVEILLANCE REQUIREMENTS SR 3.7.1.1 SURVEILLANCE

---

1\\JOTE---------------~-------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per Table 3.7.1-2 in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within +/- 1 %.

MSSVs 3.7.1 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Vogtle Units 1 and 2 3.7.1-2 Amendment f\\Jo.187 (Unit 1)

Amendment No.170 {Unit 2)

ACTIONS (continued)

CONDITION REQUIRED ACTION D.

One or more steam lines D. 1 Verify one MSIV system with one MSIV system closed in affected steam inoperable in MODE 2 or line.

3.

E.

One or more steam lines E.1 Verify one MSIV system with two MSIV systems closed in affected steam inoperable in MODE 2 or line.

3.

F.

Required Action and F.1 Bein MODE3.

associated Completion Time of Condition D or AND E not met.

F.2 Bein MODE4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE MS IVs 3.7.2 COMPLETION TIME 7 days AND Once per 7 days thereafter.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND Once per 7 days thereafter 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours FREQUENCY SR 3.7.2.1

---

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

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV system is

5 seconds on an actual or simulated actuation signal.

Vogtle Units 1 and 2 3.7.2-2 In accordance with the INSERVICE TESTING PROGRAM Amendment No. 187 (Unit 1)

Amendment No. 170 (Unit 2)

MFIVs and MFRVs and Associated Bypass Valves 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION C.

One or more MFRV or C.1 Close or isolate bypass MFIV bypass valves valve.

inoperable.

AND C.2 Verify bypass valve is closed or isolated.

D.

Both isolation systems D.1 Isolate affected feedwater inoperable in one or line.

more feedwater lines.

E.

Required Action and E.1 Be in MODE 3.

associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SR 3.7.3.1 SURVEILLANCE

---

~------------------NOTE:-----------------------------

Only required to be performed in MODE 1.

Verify the closure time of each MFIV, MFRV, and associated bypass valve is ~ 5 seconds on an actual or simulated actuation signal.

COMPLETION TIME 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Once per 7 days 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6 hours FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Vogtle Units 1 and 2 3.7.3-2 Amendment No.187 (Unit 1)

Amendment No.170 (Unit 2)

SURVEILLANCE REQUIREMENTS SR 3.7.9.1 SR 3.7.9.2 SR 3.7.9.3 SR 3.7.9.4 SR 3.7.9.5 SURVEILLANCE Verify water level of NSCW basin is?:: 80.25 ft.

Verify water temperature of NSCW basin is

~ 90°F.

Operate each required NSCW cooling tower fan for?:: 15 minutes.

Verify NSCW basin transfer pump operation.

Verify ambient wet-bulb temperature is within the three fan/spray cell region of Figure 3.7.9-1 when one NSCW tower fan/spray cell is out-of-service and daily high temperature (dry-bulb) is forecasted to be> 48°F.

FREQUENCY UHS 3.7.9 In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Vogtle Units 1 and 2 3.7.9-3 Amendment No. 187 (Unit 1)

Amendment No.170 (Unit 2)

5.5 Programs and Manuals 5.5.8 Not Used Vogtle Units 1 and 2 5.5-6 Programs and Manuals 5.5 (continued)

Amendment No.187 (Unit 1)

Amendment No. 170 (Unit 2)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DAL TON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 286 Renewed License No. DPR-57

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility) Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia (the owners), dated July 28, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 1 O CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix 8), as revised through Amendment No. 286, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance.

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: June 30, 201 7 FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 286 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages License 4

TSs 1.1-3 3.1-20 3.4-6 3.5-4 3.5-8 3.6-11 3.6-12 3.6-18 3.6-26 5.0-10 Insert Pages License 4

TSs 1.1-3 3.1-20 3.4-6 3.5-4 3.5-8 3.6-11 3.6-12 3.6-18 3.6-26 5.0-10 for sample analysis or instrumentation calibration, or associated with radioactive apparatus or components; (6)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

(C)

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at steady sate reactor core power levels not in excess of 2804 megawatts thermal.

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Plan (Appendix 8), as revised through Amendment No. 286 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.

(3)

Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained in the updated Fire Hazards Analysis and Fire Protection Program for the Edwin I. Hatch Nuclear Plant, Units 1 and 2, which was originally submitted by letter dated July 22, 1986. Southern Nuclear may make changes to the fire protection program without prior Commission approval only if the changes Renewed License No. DPR-57 Amendment No. 286

Definitions 1.1 1.1 Definitions (continued)

END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT)

SYSTEM RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE LINEAR HEAT GENERATION RATE LOGIC SYSTEM FUNCTIONAL TEST HATCH UNIT 1 The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval from initial signal generation by the associated turbine stop valve limit switch or from when the turbine control valve hydraulic control oil pressure drops below the pressure switch setpoint to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2.

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

b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;

c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d.

Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation in an arbitrary length of fuel rod, usually six inches. It is the integral of the heat flux over the heat transfer area associated with the unit length.

A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

1.1-3 (continued)

Amendment No. 286

SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.7.7 SR 3.1.7.8 SR 3.1.7.9 SR 3.1.7.10 HATCH UNIT 1 SURVEILLANCE Verify each pump develops a flow rate ~ 41.2 gpm at a discharge pressure ~ 1232 psig.

Verify flow through one SLC subsystem from pump into reactor pressure vessel.

Verify all heat traced piping between storage tank and pump suction is unblocked.

Verify sodium pentaborate enrichment is 2: 60.0 atom percent B-10.

3.1-20 SLC System 3.1.7 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pump suction piping temperature is restored within the Region A limits of Figure 3.1.7-2 Prior to addition to SLC tank Amendment No. 286

SURVEILLANCE REQUIREMENTS SR 3.4.3.1 HATCH UNIT 1 SURVEILLANCE Verify the safety function lift setpoints of the S/RVs are as follows:

Number of S/RVs 11 Setpoint

.(Q§jgl 1150 +/- 34.5 Following testing, lift settings shall be within+/- 1 %.

3.4-6 S/RVs 3.4.3 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Amendment No. 286

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.6 SR 3.5.1.7 SR 3.5.1.8 HAl"CH UNIT 1 SURVEILLANCE

---

NOl"E-------------------------------

Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4

> 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Verify each recirculation pump discharge valve cycles through one complete cycle of full travel or is de-energized in the closed position.

Verify the following ECCS pumps develop the specified flow rate against a system head corresponding to the specified reactor pressure.

SYSTEM HEAD CORRESPONDING NO. OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs LPCI

~ 4250 gpm 1

~ 17,000 gpm 2

~ 113 psig

~ 20 psig

---

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 reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1058 psig and

~ 920 psig, the HPCI pump can develop a flow rate ;;:: 4250 gpm against a system head corresponding to reactor pressure.

3.5-4 ECCS - Operating 3.5.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program (continued)

Amendment No. 286

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 SR 3.5.2.6 HATCH UNIT 1 SURVEILLANCE Verify, for each required ECCS injection/ spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

---

N()TES-----------------------------

1. ()ne LPCI subsystem may be considered

()PERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

2. Not required to be met for system vent flowpaths opened under administrative control.

Verify each required ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure.

SYSTEM HEAD CORRESPONDING NO.OF TOA REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs

<!:4250 gpm 1

113 psig LPCI

7700 gpm 1

20 psig

---

N()TE--------------------------------

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.

3.5-8 ECCS - Shutdown 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PR()GRAM In accordance with the Surveillance Frequency Control Program Amendment No. 286

SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.2 SR 3.6.1.3.3 SR 3.6.1.3.4 SR 3.6.1.3.5 HATCH UNIT 1 SURVEILLANCE

---

N()TES---------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

---

N()TES----------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge.

Verify the isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.

3.6-11 PC I Vs 3.6.1.3 FREQUENCY In accordance with the Surveillance Frequency Control Program Prior to entering M()DE 2 or 3 from M()DE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING*

PR()GRAM (continued)

Amendment No. 286

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.6.1.3.6 Verify the isolation time of each MSIV is

~ 3 seconds and s 5 seconds.

SR 3.6.1.3.7 Verify each automatic PCIV, excluding EFCVs, actuates to the isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify each reactor instrumentation line EFCV (of a representative sample) actuates to restrict flow to within limits.

SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP system.

SR 3.6.1.3.10 Verify combined MSIV leakage rate for all four main steam lines is s 100 scfh when tested at

~ 28.0 psig and < 50.8 psig.

OR Verify combined MSIV leakage rate for all four main steam lines is s 144 scfh when tested at

~ 50.8 psig.

SR 3.6.1.3.11 Deleted SR 3.6.1.3.12 Cycle each 18 inch excess flow isolation damper to the fully closed and fully open position.

SR 3.6.1.3.13 Verify the combined leakage rate for all secondary containment bypass leakage paths is s 0.02 La when pressurized to;=: Pa.

HATCH UNIT 1 3.6-12 PC I Vs 3.6.1.3 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Primary Containment Leakage Rate Testing Program In accordance with the Surveillance Frequency Control Program In accordance with the Primary Containment Leakage Rate Testing Program Amendment No. 286

Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION E.

Two lines with one or more E.1 Restore all vacuum reactor building-to-breakers in one line to suppression chamber OPERABLE status.

vacuum breakers inoperable for opening.

F.

Required Action and F.1 Be in MODE 3.

Associated Completion Time of Condition A, B, or AND E not met.

F.2 Be in MODE4.

SURVEILLANCE REQUIREMENTS SR 3.6.1.7.1 SR 3.6.1.7.2 SR 3.6.1.7.3 HATCH UNIT 1 SURVEILLANCE

---

NOTES---------------------------

1.

Not required to be met for vacuum breakers that are open during Surveillances.

2.

Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed.

Perform a functional test of each vacuum breaker.

Verify the opening setpoint of each vacuum breaker is s 0.5 psid.

3.6-18 COMPLETION TIME 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 12 hours 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Amendment No. 286

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.2.3.2 SR 3.6.2.3.3 HATCH UNIT 1 SURVEILLANCE Verify RHR suppression pool cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water.

Verify each required RHR pump develops a flow rate 2: 7700 gpm through the associated heat exchanger while operating in the suppression pool cooling mode.

3.6-26 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM Amendment No. 286

5.5 Programs and Manuals (continued}

5.5.5 Component Cyclic or Transient Limit Programs and Manuals 5.5 This program provides controls to track FSAR Section 4.2, cyclic and transient occurrences, to ensure that reactor coolant pressure boundary components are maintained within the design limits.

5.5.6 Not Used 5.5.7 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF} filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, Sections C.5.c and C.5.d, and in accordance with Regulatory Guide 1.52, Revision 2.

(continued}

HATCH UNIT 1 5.0-10 Amendment No. 286

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DAL TON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 231 Renewed License No. NPF-5

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself; Georgia Power Company; Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; and City of Dalton, Georgia (the owners), dated July 28, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-5 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 231, are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance.

Attachment:

Changes to the Operating License and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: June 30, 20 I 7

ATTACHMENT TO LICENSE AMENDMENT NO. 231 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages License 4

TSs 1.1-3 3.1-19 3.4-6 3.5-4 3.5-9 3.6-11 3.6-12 3.6-18 3.6-26 5.0-10 Insert Pages License 4

TSs 1.1-3 3.1-19 3.4-6 3.5-4 3.5-9 3.6-11 3.6-12 3.6-18 3.6-26 5.0-10 (6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

(C)

This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions2 specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady sate reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No. 231 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission.

(a) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained 2 The original licensee authorized to possess, use, and operate the facility with Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

Renewed License No. NPF-5 Amendment No. 231

Definitions 1.1 1.1 Definitions (continued)

EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE

TIME END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT)

SYSTEM RESPONSE TIME INSERVICE TESTING PROGRAM ISOLATION SYSTEM RESPONSE TIME LEAKAGE HATCH UNIT2 The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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.

The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval from initial signal generation by the associated turbine stop valve limit switch or from when the turbine control valve hydraulic control oil pressure drops below the pressure switch setpoint to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. 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.

LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2.

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

b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; (continued) 1.1-3 Amendment No. 231

SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.7.6 SR 3.1.7.7 SR 3.1.7.8 SR 3.1.7.9 SR 3.1.7.10 HATCH UNIT 2 SURVEILLANCE Verify each SLC subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

• Verify each pump develops a flow rate ~ 41.2 gpm at a discharge pressure ~ 1232 psig.

Verify flow through one SLC subsystem from pump into reactor pressure vessel.

Verify all heat traced piping between storage tank and pump suction is unblocked.

Verify sodium pentaborate enrichment is

~ 60.0 atom percent B-10.

3.1-19 SLC System 3.1.7 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the IN SERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pump suction piping temperature is restored within the Region A limits of Figure 3. 1. 7-2 Prior to addition to SLC tank Amendment No. 231

SURVEILLANCE REQUIREMENTS SR 3.4.3.1 HATCH UNIT2 SURVEILLANCE Verify the safety function lift setpoints of the S/RVs are as follows:

Number of S/RVs 11 Setpoint

.{Q§ig1 1150 +/- 34.5 Following testing, lift settings shall be within +/- 1 %.

3.4-6 S/RVs 3.4.3 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Amendment No. 231

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.6 SR 3.5.1.7 SR 3.5.1.8 HATCH UNIT 2 SURVEILLANCE

---

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

Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4

> 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Verify each recirculation pump discharge valve cycles through one complete cycle of full travel or is de-energized in the closed position.

Verify the following ECCS pumps develop the specified flow rate against a system head corresponding to the specified reactor pressure.

SYSTEM HEAD CORRESPONDING NO. OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs LPCI

~4250 gpm 1

C!: 17,000 gpm 2

C!: 113 psig 2!: 20 psig

---

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 reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressures 1058 psig and

~ 920 psig, the HPCI pump can develop a flow rate~ 4250 gpm against a system head corresponding to reactor pressure.

3.5-4 ECCS - Operating 3.5.1 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program (continued)

Amendment No. 231

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 SR 3.5.2.6 HATCH UNIT 2 SURVEILLANCE Verify, for each required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

---

N()TES-------------------------------

1. One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

2. Not required to be met for system vent flowpaths opened under administrative control.

Verify each required ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure.

SYSTEM HEAD CORRESPONDING NO. OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS

a

4250 gpm 1

a

113 psig LPCI

~ 7700 gpm 1

~ 20 psig

---

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

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray subsystem actuates on an actual or simulated automatic initiation signal.

3.5-9 ECCS - Shutdown 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Amendment No. 231

SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.2 SR 3.6.1.3.3 SR 3.6.1.3.4 SR 3.6.1.3.5 HATCH UNIT 2 SURVEILLANCE

---

N{)TES---------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

---

N{)TES----------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge.

Verify the isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.

3.6-11 PC I Vs 3.6.1.3 FREQUENCY In accordance with the Surveillance Frequency Control Program Prior to entering M()DE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM (continued)

Amendment No. 231

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.6.1.3.6 Verify the isolation time of each MSIV is

~ 3 seconds and s 5 seconds.

SR 3.6.1.3.7 Verify each automatic PCIV, excluding EFCVs, actuates to the isolation position on an actual or simulated isolation signal.

SR 3.6.1.3.8 Verify each reactor instrumentation line EFCV (of a representative sample) actuates to restrict flow to within limits.

SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP system.

SR 3.6.1.3.10 Verify the combined leakage rate for all secondary containment bypass leakage paths is s 0.02 La when pressurized to~ Pa.

SR 3.6.1.3.11 Verify combined MSIV leakage rate for all four main steam lines is s 100 scfh when tested at ~

28.8 psig and< 47.3 psig.

OR Verify combined MSIV leakage rate for all four main steam lines is s 144 scfh when tested at~

47.3 psig.

SR 3.6.1.3.12 Deleted SR 3.6.1.3.13 Cycle each 18 inch excess flow isolation damper to the fully closed and fully open position.

HATCH UNIT2 3.6-12 PC IVs 3.6.1.3 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Primary Containment Leakage Rate Testing Program In accordance with the Primary Containment Leakage Rate Testing Program In accordance with the Surveillance Frequency Control Program Amendment No. 231

Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION F.

Required Action and Associated Completion Time of Condition A, B, or E not met.

F.1 Be in MODE 3.

AND F.2 Bein MODE 4.

SURVEILLANCE REQUIREMENTS SR 3.6.1.7.1 SR 3.6.1. 7.2 SR 3.6.1.7.3 HATCH UNIT 2 SURVEILLANCE

---

NOTES---------------------------

1.

Not required to be met for vacuum breakers that are open during Surveillances.

2.

Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed.

Perform a functional test of each vacuum breaker.

Verify the opening setpoint of each vacuum breaker is :S 0.5 psid.

3.6-18 COMPLETION TIME 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Amendment No. 231

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.2.3.2 SR 3.6.2.3.3 HATCH UNIT 2 SURVEILLANCE Verify RHR suppression pool cooling subsystem locations susceptible to gas accumulation are sufficiently filled with water.

Verify each required RHR pump develops a flow rate <!: 7700 gpm through the associated heat exchanger while operating in the suppression pool cooling mode.

3.6-26 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM Amendment No.231

5.5 Programs and Manuals (continued) 5.5.5 Component Cyclic or Transient Limit Programs and Manuals 5.5 This program provides controls to track FSAR Section 5.2, cyclic and transient occurrences, to ensure that reactor coolant pressure boundary components are maintained within the design limits.

5.5.6 Not Used 5.5.7 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, Sections C.5.c and C.5.d, and in accordance with Regulatory Guide 1.52, Revision 2.

(continued)

HATCH UNIT 2 5.0-10 Amendment No. 231

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1AND2 AMENDMENT NO. 212 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AMENDMENT NO. 209 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 AND VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 AMENDMENT NO. 187 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-68 AMENDMENT NO. 170 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-81 AND EDWIN I. HATCH, UNIT NOS. 1 AND 2 AMENDMENT NO. 286 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57 AMENDMENT NO. 231 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

DOCKET NOS. 50-348, 50-364, 50-424, 50-425, 50-321, AND 50-366

1.0 INTRODUCTION

By application dated July 28, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16214A252), Southern Nuclear Operating Company (SNC, the licensee), requested changes to the technical specifications (TSs) for the Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Vogtle Electric Generating Plant (VEGP), Units 1 and 2, and the Edwin I. Hatch Nuclear Plant (HNP), Unit Nos. 1 and 2. Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015 (ADAMS Accession No. ML15294A555).

The licensee's proposed changes delete the FNP and VEGP TS 5.5.8, "lnservice Testing Program," and HNP TS 5.5.6, "lnservice Testing [IST] Program]," and adds a new defined term, "INSERVICE TESTING PROGRAM," to the FNP, VEGP, and HNP TSs. All existing references to the "lnservice Testing Program", in the FNP, VEGP, and HNP TS SRs, along with references to the "IST Program", and "lnservice Testing Plan" in the FNP TS SRs, are replaced with "INSERVICE TESTING PROGRAM" so that the SRs refer to the new definition in lieu of the deleted program.

The licensee's letter dated July 28, 2016, also included a request to use American Society of Mechanical Engineers (ASME) Code Case OMN-20, "lnservice Test Frequency," as an alternative to certain ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) requirements at FNP and VEGP for the current 10-year lnservice Testing (IST) program interval, which is the fourth interval at FNP and the third interval at VEGP.

Because the VEGP fourth 10-year IST program interval began on June 1, 2017, SNC submitted a separate letter on July 28, 2016 (ADAMS Accession No. ML16210A460), requesting a similar alternative for the fourth 10-year interval. The U.S. Nuclear Regulatory Commission (NRC) staff considered these requests separately from the proposed license amendment, and authorized the licensee's use of this alternative for the FNP fourth interval and the VEGP third and fourth interval by letter dated October 14, 2016 (ADAMS Accession No. ML16264A321).

Additionally, by letter dated May 4, 2015 (ADAMS Accession No. ML15124A904), SNC requested to use OMN-20 as an alternative to ASME OM Code requirements at HNP for the fifth 10-year IST program interval, which started on January 1, 2016. The NRC staff authorized the licensee's use of this alternative for the HNP fifth interval by letter dated December 30, 2015 (ADAMS Accession No. ML15310A406).

2.0 REGULATORY EVALUATION

2.1 Description of lnservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The ASME OM Code provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 1 O of the Code of Federal Regulations (10 CFR), paragraph 50.55a(f), "lnservice testing requirements," requires that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The facility's TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation in 10 CFR 50.55a(f)(5)(ii) states, in part, "If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program." TSTF-545, Revision 3, provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a, "Codes and standards," and the TSs. TSTF-545, Revision 3, proposes elimination of the lnservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the lnservice Testing Program. The elimination of the lnservice Testing Program from the TSs could cause uncertainty regarding the correct application of these SRs. Therefore, TSTF-545, Revision 3, also proposes adding a new definition, "INSERVICE TESTING PROGRAM," to the TSs, which would be defined as "the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." TSTF-545, Revision 3, proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

The NRC approved TSTF-545, Revision 3, by letter dated December 11, 2015 (ADAMS Package Accession No. ML15317A071), and published a notice of availability in the Federal Register (FR) on March 28, 2016 (81 FR 17208).

2.2 Proposed Technical Specifications Changes The licensee requested to delete TS 5.5.8 from the Administrative Controls section of the FNP and VEGP TSs, and delete TS 5.5.6 from the Administrative Controls section of the HNP TSs, and replace it with the word "Not Used."

The FNP and VEGP TS 5.5.8 currently states:

This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a.

Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days

b.

The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c.

The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d.

Nothing in the ASME ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

The HNP, Unit Nos. 1 and 2, TS 5.5.6 currently states:

This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports.

e.

Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follows:

ASME Boiler and Pressure Vessel Code and Applicable Addenda Terminology for lnservice Testing Activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Yearly or annually Required Frequencies for Performing lnservice Testing Activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 366 days

f.

The provisions of SR 3.0.2 are applicable to the frequencies for performing inservice testing activities;

g.

The provisions of SR 3.0.3 are applicable to inservice testing activities; and

h.

Nothing in the ASME ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

SR 3.0.2 allows an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, SR 3.0.3 allows the licensee to delay declaring the associated limiting condition for operation not met in order to perform the missed surveillance. The licensee did not request changes to SR 3.0.2 or SR 3.0.3.

The licensee requested to revise the Definitions section of the TSs by adding the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The licensee also requested that all existing occurrences of "lnservice Testing Program" in the FNP, VEGP, and HNP TS SRs, along with the occurrences of "IST Program", and "lnservice Testing Plan" in the FNP TS SRs, be replaced with "INSERVICE TESTING PROGRAM," so that the SRs refer to the new definition in lieu of the deleted program.

2.3 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes:

Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Section 50.36(c)(3) of 10 CFR states that "[s]urveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." Section 50.36(c)(5) of 1 O CFR states that "[a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner."

The NRC staff's guidance for review of the TSs is in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition," Chapter 16, "Technical Specifications," Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the staff has prepared improved STSs for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. The licensee's proposed amendment is based on TSTF-545, Revision 3, which is an NRG-approved change to the improved STSs.

The NRC staff's review includes consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved STSs, as modified by NRG-approved TSTF travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.

lnservice Testing Pursuant to 10 CFR 50.54, "Conditions of licenses," the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 1 O CFR 50.55a(f). The regulations in 10 CFR 50.55a(f) state, in part:

(f) lnservice testing requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV [Boiler and Pressure Vessel] Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions.... [referring to 10 CFR 50.55a(f)(1) through (f)(6)]....

The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements were suitable for incorporation into the NRC's rules.

The regulation in 10 CFR 50.55a(f)(5)(ii) states, in part:

(ii) /ST program update: Conflicting /ST Code requirements with technical specifications. If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program....

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and lnservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants," Final Report, October 2013 (ADAMS Accession No. ML13295A020) provides guidance for the inservice testing of pumps and valves.

NUREG-0800, Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints," Revision 3, March 2007 (ADAMS Accession No. ML070720041 ), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps and valves.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's application to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5)

(i.e., provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). In making its determination as to whether to amend the license, the NRC staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54.

Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

3.1 Deletion of the lnservice Testing Program from the TSs TS 5.5.8 for FNP and VEGP, and TS 5.5.6 for HNP, requires the licensee to have an inservice testing program that provides controls for inservice testing of ASME Code Class 1, 2, and 3 components (i.e., pumps and valves). Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54.

Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained below, it is not necessary to have additional administrative controls in the TSs relating to the inservice testing program to assure operation of the facility in a safe manner.

Consideration of TS 5.5.8.a for FNP and VEGP, and TS 5.5.6.a for HNP The ASME OM Code requires testing to normally be performed within certain time periods. In TS 5.5.8.a for FNP and VEGP, and TS 5.5.6 for HNP, inservice testing frequencies are more precise than those specified in the ASME OM Code and applicable addenda (e.g., "at least once per 31 days" contrasted with "monthly"). However, the NRC staff determined that the more precise inservice testing frequencies are not necessary to assure operation of the facility in a safe manner and that the deletion of FNP and VEGP TS 5.5.8.a and HNP TS 5.5.6.a is acceptable.

Consideration of TS 5.5.8.b for FNP and VEGP, and TS 5.5.6.b for HNP In TS 5.5.8.b for FNP and VEGP, and TS 5.5.6.b for HNP, the licensee is allowed to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.8.a for FNP and VEGP, and TS 5.5.6.a for HNP. Similar to TS 5.5.8.b for FNP and VEGP, and TS 5.5.6.b for HNP, the NRC authorization of ASME Code Case OMN-20, "lnservice Test Frequency," by letter dated October 14, 2016, for FNP and VEGP, and December 30, 2015 for HNP, also permits the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent.

The NRC staff determined that the TS 5.5.8.b for FNP and VEGP, and TS 5.5.6.b for HNP, allowance to extend inservice testing intervals is not needed to assure operation of the facility in a safe manner. The deletion of TS 5.5.8.b for FNP and VEGP, and TS 5.5.6.b for HNP, does not impact the licensee's ability to extend inservice testing intervals using Code Case OMN-20, as authorized by the NRC. Therefore, the NRC staff determined that deletion of TS 5.5.8.b for FNP and VEGP, and TS 5.5.6.b for HNP, is acceptable.

Consideration of TS 5. 5. 8. c for FNP and VEGP, and TS 5. 5. 6. c for HNP In TS 5.5.8.c for FNP and VEGP, and TS 5.5.6.c for HNP, the licensee is allowed to use SR 3.0.3 when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 3.0.3 allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 3.0.3 for inservice tests is limited to those inservice tests required by an SR. In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test. Deletion of TS 5.5.8.c for FNP and VEGP, and TS 5.5.6.c for HNP, does not change any of these requirements, and SR 3.0.3 will continue to apply to those inservice tests required by SRs. Based on the above, the NRC staff determined that deletion of TS 5.5.8.c for FNP and VEGP, and TS 5.5.6.c for HNP, is acceptable.

Consideration of TS 5.5.8.d for FNP and VEGP, and TS 5.5.6.d for HNP In TS 5.5.8.d for FNP and VEGP, and TS 5.5.6.c for HNP, the TS states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. However, the regulations in 10 CFR 50.55a(f)(5)(ii) address what to do if a revised inservice testing program for a facility conflicts with the TSs for the facility; they require the licensee to apply for an amendment to the TSs to conform the TSs to the revised program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for a TS stating how to address conflicts between the TSs and the inservice testing program because the regulations specify how conflicts must be resolved. Therefore, the NRC staff finds that the deletion of TS 5.5.8.d for FNP and VEGP, and TS 5.5.6.d for HNP, is acceptable.

Conclusion Regarding Deletion of TS 5.5.8 for FNP and VEGP, and TS 5.5.6 for HNP The NRC staff determined that the requirements currently in TS 5.5.8 for FNP and VEGP, and TS 5.5.6 for HNP, are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the NRC staff concludes that deletion of TS 5.5.8 from the FNP and VEGP TSs, and deletion of TS 5.5.6 from the HNP TS, is acceptable because they are not required by 10 CFR 50.36(c)(5).

3.2 Definition of INSERVICE TESTING PROGRAM and Revision to SRs The licensee proposes to revise the TS Definitions section to include the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The proposed definition of the INSERVICE TESTING PROGRAM is consistent with the definition in TSTF-545, Revision 3. The definition is acceptable to the NRC staff because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).

The licensee requested that all existing references to the "lnservice Testing Program" in the FNP, VEGP, and HNP SRs, along with the references to "IST Program", and "lnservice Testing Plan" in the FNP SRs, be revised to "INSERVICE TESTING PROGRAM" to reference the new TS defined term in lieu of the deleted program. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in SRs with the new definition.

The NRC staff verified that for each FNP, VEGP, and HNP SR reference to the "lnservice Testing Program,", and references to "IST Program", and "lnservice Testing Plan" in the FNP SRs, the licensee proposed to change the reference to "INSERVICE TESTING PROGRAM."

The proposed change does not alter how the SR testing is performed. However, the inservice testing frequencies could change because the TSs will no longer include the more precise test frequencies in TS 5.5.8.a for FNP and VEGP, and TS 5.5.6.a for HNP. As discussed in Section 3.1 of this safety evaluation, the NRC staff determined that the TSs do not need to include the more precise testing frequencies currently in TS 5.5.8.a for FNP and VEGP, and TS 5.5.6.a for HNP. Based on its review, the NRC staff determined that revising the SRs to refer to the new definition is acceptable because these SRs will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The NRC staff also determined that, with the proposed changes that allow less-precise testing frequencies, 1 O CFR 50.36(c)(3) will continue to be met because the SRs will continue to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

3.3 Deviations from TSTF-545 In its application.dated July 28, 2016, the licensee identified the following deviations from TSTF-545, Revision 3:

1. FNP, Units 1 and 2, HNP, Units 1 and 2, and VEGP, Units 1 and 2, TS utilize different numbering than the Standard Technical Specifications on which TSTF-545 was based. Specifically, FNP, Units 1 and 2, and VEGP, Units 1 and 2, differ from TSTF-545 in that while the numbering for the section to be deleted (i.e. 5.5.8) is consistent with the applicable Standard Technical Specifications, the proposed revision labels TS 5.5.8 [are noted] as "Not Used" and so maintains the existing numbering for subsequent sections in the TS and corresponding Bases.

2. HNP, Units 1 and 2, TS differ in numbering from TSTF-545 in that the numbering for the section to be deleted (i.e. 5.5.6) differs from the applicable Standard Technical Specifications numbering (5.5.7); in this case, since the proposed revision labels TS 5.5.6 [are noted] as "Not Used" and maintains the existing numbering for the subsequent sections, the proposed revision will result in the numbering of the subsequent sections of the TS and corresponding Bases becoming consistent with TSTF-545.

3. Also, variant terminology such as "lnservice Testing Plan," "lnservice Test Program," or "IST Program" is sometimes used in place of "lnservice Testing Program" in the existing FNP, [Units 1 and 2], and VEGP, Units 1 and 2, TS and Bases, including usage in surveillance requirements (SR) statements. These terms are replaced with "INSERVICE TESTING PROGRAM" in the proposed license amendments.

The NRC staff finds that the proposed deviations are editorial in nature and the licensee's proposed TS changes remain consistent with the intent of TSTF-545, Revision 3. Therefore, the NRC staff finds that the licensee's proposed TS changes are acceptable.

3.4 Correction of Typographical Error in HNP, Unit 1, TS 3.6.1.7, Condition F HNP, Unit No. 1, Amendment No. 281 (ADAMS Accession No. ML16257A724) was issued on December 19, 2016. The amendment was issued with TS 3.6.1.7, Condition F, stating (with underline added for emphasis),

Required Action and Associated Completion Time or Condition A, B, or E not met.

In its application for the amendment dated December 15, 2015 (ADAMS Accession No. ML15351A023), the licensee requested that Condition F state (with underline added for emphasis),

Required Action and Associated Completion Time of Condition A, B, or E not met.

The issuance of this TS with "or instead of "of' in Condition F was a typographical error and had no effect on plant operations, plant safety, or the safety evaluation associated with Amendment No. 281. The revised TS page 3.6-18 being issued with this amendment corrects this typographical error.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia and Alabama State officials were notified of the proposed issuance of the amendment on May 31, 2017. The NRC staff confirmed that the Alabama State official on May 31, 2017, and the Georgia State official on June 12, 2017, respectively, had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the FR on September 27, 2016 (81 FR 66309, 81 FR 66310, and 81 FR 66311). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's 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.

Principal Contributors: Blake Purnell, NRR Caroline Tilton, NRR John Huang, NRR Michael Orenak, NRR Date: June 30, 2017

ML17152A218

• via email OFFICE NRR/DORL/LPL2-1 /PM NRR/DORL/LPL2-1 /LA NRR/DSS/STSB/BC(A)*

NAME MOrenak KGoldstein JWhitman DATE 06/14/2017 06/27/17 06/22/2017 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1 /PM NAME MMarkley Orenak DATE 06/30/2017 06/30/2017