Text

Issuance of Amendments to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application
	ML18172A172
Person / Time
Site:	Oconee, Mcguire, Catawba, Harris, Robinson, McGuire
Issue date:	08/15/2018
From:	Dennis Galvin; Plant Licensing Branch II
To:	Capps C; Duke Energy Carolinas
	Galvin, D, 415-6256
References
	EPID L-2017-LLA-0377
	Download: ML18172A172 (113)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 15, 2018 Mr. Steven Capps Senior Vice President Nuclear Corporate Duke Energy Corporation 526 South Church Street, EC-07H Charlotte, NC 28202

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2; MCGUIRE NUCLEAR STATION, UNITS 1 AND 2; OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3; SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1; AND H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENTS TO ADOPT TSTF-545, REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING" (EPID L-2017-LLA-0377)

Dear Mr. Capps:

The U.S. Nuclear Regulatory Commission (NRC) has issued the following enclosed amendments: Amendment Nos. 299 and 295 to Renewed Facility Operating License Nos. NPF-35 and NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively; Amendment Nos. 309 and 288 to Renewed Facility Operating License Nos. NPF-9 and NPF-17 for the McGuire Nuclear Station, Units 1 and 2, respectively; Amendment Nos. 409, 411, and 410 to Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55 for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, respectively; Amendment No. 166 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1; and Amendment No. 259 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2.

The amendments revise the technical specifications (TSs) in response to the Duke Energy Carolinas, LLC and Duke Energy Progress, LLC application dated November 7, 2017. The amendments revise the TSs for each of these facilities based on Technical Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing."

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

S. Capps If you have any questions, please contact me at (301) 415-6256 or by e-mail at Dennis.Galvin@nrc.gov.

Sincerely, Dennis J. Galvin, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413, 50-414, 50-369, 50-370, 50-269, 50-270, 50-287, 50-400, and 50-261

Enclosures:

1. Amendment No. 299 to NPF-35

2. Amendment No. 295 to NPF-52

3. Amendment No. 309 to NPF-9

4. Amendment No. 288 to NPF-17

5. Amendment No. 409 to DPR-38

6. Amendment No. 411 to DPR-47

7. Amendment No. 410 to DPR-55

8. Amendment No. 166 to NPF-63

9. Amendment No. 259 to DPR-23

10. Safety Evaluation cc: Mr. Robert T. Simril Mr. Thomas Ray Site Vice President Site Vice President Catawba Nuclear Station McGuire Nuclear Station Duke Energy Carolinas, LLC Duke Energy Carolinas, LLC 4800 Concord Road 12700 Hagers Ferry Road York, SC 29745 Huntersville, NC 28078-8985 Mr. Ed Burchfield, Jr. Mr. Ernest J. Kapopoulos, Jr.

Site Vice President Site Vice President Oconee Nuclear Station H. B. Robinson Steam Electric Plant Duke Energy Carolinas, LLC Duke Energy Progress, LLC 7800 Rochester Highway 3581 West Entrance Road, RNPA01 Seneca, SC 29672-0752 Hartsville, SC 29550 Ms. Tanya Hamilton Site Vice President Duke Energy Progress, LLC Shearon Harris Nuclear Power Plant, Unit 1 5413 Shearon Harris Road, M/C HNP01 New Hill, NC 27562-0165 Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 299 Renewed License No. NPF-35

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

A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-35 filed by the Duke Energy Carolinas, LLC, acting for itself, and North Carolina Electric Membership Corporation (licensees), dated November 7, 2017, 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.

Enclosure 1

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-35 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 299, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC 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.

FOR THE NUCLEAR REGULATORY COMMISSION

~g'Ju~

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-35 and Technical Specifications Date of Issuance: August 15, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 295 Renewed License No. NPF-52

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

A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-52 filed by the Duke Energy Carolinas, LLC, acting for itself, North Carolina Municipal Power Agency No. 1 and Piedmont Municipal Power Agency (licensees), dated November 7, 2017, 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.

Enclosure 2

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-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 295, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC 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.

FOR THE NUCLEAR REGULATORY COMMISSION D -;\1\10-/

~. 't::f;l)~ . .-

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-52 and the Technical Specifications Date of Issuance: August 1 5, 2018

ATTACHMENT TO CATAWBA NUCLEAR STATION, UNITS 1 AND 2 LICENSE AMENDMENT NO. 299 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 295 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses 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 License License NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 TSs TSs 1.1-3 1.1-3 3.4.10-2 3.4.10-2 3.4.14-3 3.4.14-3 3.5.2-2 3.5.2-2 3.6.3-6 3.6.3-6 3.6.6-2 3.6.6-2 3.7.1-2 3.7.1-2 3.7.2-2 3.7.2-2 3.7.3-2 3.7.3-2 3.7.5-3 3.7.5-3 5.5-6 5.5-6

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 299, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. NPF-35 Amendment 299

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 295, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),

the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. NPF-52 Amendment No. 295

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT Xe-133 DOSE EQUIVALENT Xe-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT Xe-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE(ESF)RESPONSE when the monitored parameter exceeds its ESF actuation TIME 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.

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

(continued)

Catawba Units 1 and 2 1.1-3 Amendment Nos. 299 I 295

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift settings shall be ~ TESTING 2460 psig and 5- 2510 psig. PROGRAM Catawba Units 1 and 2 3.4.10-2 Amendment Nos. 299 I 295

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 -------------------------------NOTES---------------------------------

1. Not required to be performed in MODES 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. In accordance with the INSERVICE TESTING Verify leakage from each RCS PIV is equivalent to~ 0.5 PROGRAM, and gpm per nominal inch of valve size up to a maximum of 5 in accordance with gpm at an RCS pressure~ 2215 psig and ~ 2255 psig. the Surveillance Frequency Control Program Prior to entering MODE2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve (continued)

Catawba Units 1 and 2 3.4.14-3 Amendment Nos. 299 I 295

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with In accordance with power to the valve operator removed. the Surveillance Frequency Control Number Position Function Program Nl162A Open SI Cold Leg Injection Nl121A Closed SI Hot Leg Injection Nl152B Closed SI Hot Leg Injection Nl183B Closed RHR Hot Leg Injection Nl173A Open RHR Cold Leg Injection Nl178B Open RHR Cold Leg Injection Nl100B Open SI Pump Suction from RWST Nl147B Open SI Pump Mini-Flow SR 3. 5. 2. 2 ----------------------------------NO TE---------------------------------

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

Verify each ECCS manual, power operated, and In accordance with automatic valve in the flow path, that is not locked, the Surveillance sealed, or otherwise secured in position, is in the correct Frequency Control position. Program SR 3.5.2.3 Verify ECCS locations susceptible to gas accumulation In accordance with are sufficiently filled with water. the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM (continued)

Catawba Units 1 and 2 3.5.2-2 Amendment Nos. 299 / 295

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 6. 3. 4 ------------------------------NO TE------------------------------------

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

Verify each containment isolation manual valve and blind Prior to entering flange that is located inside containment or annulus and MODE 4 from not locked, sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed, except for performed within containment isolation valves that are open under the previous administrative controls. 92 days SR 3.6.3.5 Verify the isolation time of automatic power operated In accordance with containment isolation valve is within limits. the INSERVICE TESTING PROGRAM SR 3.6.3.6 Perform leakage rate testing for Containment Purge In accordance with System, Hydrogen Purge System, and Containment Air the Containment Release and Addition System valves with resilient seals. Leakage Rate Testing Program SR 3.6.3.7 Verify each automatic containment isolation valve that is In accordance with not locked, sealed or otherwise secured in position, the Surveillance actuates to the isolation position on an actual or Frequency Control simulated actuation signal. Program (continued)

Catawba Units 1 and 2 3.6.3-6 Amendment Nos. 299 I 295

Containment Spray System 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.2 Verify each containment spray pump's developed head at In accordance with the flow test point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM SR 3.6.6.3 Deleted.

SR 3.6.6.4 Deleted.

SR 3.6.6.5 Verify that each spray pump is de-energized and In accordance with prevented from starting upon receipt of a terminate signal the Surveillance and is allowed to manually start upon receipt of a start Frequency Control permissive from the Containment Pressure Control Program System (CPCS).

SR 3.6.6.6 Verify that each spray pump discharge valve closes or is In accordance with prevented from opening upon receipt of a terminate the Surveillance signal and is allowed to manually open upon receipt of a Frequency Control start permissive from the Containment Pressure Control Program System (CPCS).

SR 3.6.6.7 Verify each spray nozzle is unobstructed. Following activities which could result in nozzle blockage SR 3.6.6.8 Verify containment spray locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.6-2 Amendment Nos. 299 / 295

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7 .1.1 ----------------------------------NOTE--------------------------------

Only required to be performed prior to entry into.

MODE 2.

Verify each required MSSV lift setpoint per Table 3. 7 .1-2 In accordance with in accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift setting shall be within TESTING

+/-.1%. PROGRAM Catawba Units 1 and 2 3.7.1-2 Amendment Nos. 299 I 295

MSIVs 3.7.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and 0.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C not AND met.

0.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7. 2. 1 ----------------------------------NOTE--------------------------------

On ly required to be performed prior to entry into MODE 2.

Verify closure time of each MSIV is within limits on an In accordance with actual or simulated actuation signal. the INSERVICE TESTING PROGRAM Catawba Units 1 and 2 3.7.2-2 Amendment Nos. 299 I 295

MFIVs, MFCVs, Associated Bypass Valves and Tempering Valves 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more MFIV or C.1 Close or isolate bypass 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months MFCV bypass valves valve.

inoperable.

AND C.2 Verify bypass valve is Once per closed or isolated. 7 days D. Two valves in the same D. 1 Isolate affected flow path. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow path or the tempering valve inoperable.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify the closure time of each MFIV, MFCV, their In accordance with associated bypass valve, and the tempering valve is within the INSERVICE limits on an actual or simulated actuation signal. TESTING PROGRAM Catawba Units 1 and 2 3.7.3-2 Amendment Nos. 299 / 295

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7 .5.1 --------------------------------------NOTE----------------------------

Not applicable to automatic valves when THERMAL POWER is~ 10% RTP.

Verify each AFW manual, power operated, and automatic In accordance valve in each water flow path, and in both steam supply with the flow paths to the steam turbine driven pump, that is not Surveillance locked, sealed, or otherwise secured in position, is in the Frequency correct position. Control Program SR 3.7.5.2 --------------------------------------NOTE----------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after.::_ 600 psig in the steam generator.

Verify the developed head of each AFW pump at the flow In accordance test point is greater than or equal to the required with the developed head. INSERVICE TESTING PROGRAM SR 3. 7. 5. 3 --------------------------------------NOTE----------------------------

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not locked, In accordance sealed, or otherwise secured in position, actuates to the with the correct position on an actual or simulated actuation Surveillance signal. Frequency Control Program (continued)

Catawba Units 1 and 2 3.7.5-3 Amendment Nos. 299 I 295

Programs and Manuals 5.5 5.5 Programs and Manuals (continued}

5.5.8 lnservice Testing Program (Deleted)

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

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 for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the (continued)

Catawba Units 1 and 2 5.5-6 Amendment Nos. 299 I 295

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 309 Renewed License No. NPF-9

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

A The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-9, filed by the Duke Energy Carolinas, LLC (licensee), dated November 7, 2017, 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.

Enclosure 3

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-9 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 309, are hereby incorporated into this renewed operating license. The licensee 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.

FOR THE NUCLEAR REGULATORY COMMISSION

~.~~

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-9 and the Technical Specifications Date of Issuance: August 15, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 288 Renewed License No. NPF-17

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

A. The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-17, filed by the Duke Energy Carolinas, LLC (the licensee), dated November 7, 2017, 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.

Enclosure 4

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-17 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 288 are hereby incorporated into this renewed operating license. The licensee 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.

FOR THE NUCLEAR REGULATORY COMMISSION

~bl)~

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-17 and the Technical Specifications Date of Issuance: August 15, 2018

ATTACHMENT TO MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 LICENSE AMENDMENT NO. 309 RENEWED FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND LICENSE AMENDMENT NO. 288 RENEWED FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following pages of the Renewed Facility Operating Licenses 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 License License NPF-9, page 3 NPF-9, page 3 NPF-17, page 3 NPF-17, page 3 TSs TSs 1.1-3 1.1-3 3.4.10-2 3.4.10-2 3.4.14-3 3.4.14-3 3.5.2-3 3.5.2-3 3.6.3-6 3.6.3-6 3.6.6-2 3.6.6-2 3.7.1-2 3.7.1-2 3.7.2-2 3.7.2-2 3.7.3-2 3.7.3-2 3.7.5-3 3.7.5-3 5.5-6 5.5-6

(4) 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; (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2, and; (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.

C. This renewed operating 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 incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal ( 100% ).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 309, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50. 71 (e )( 4 ), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Renewed License No. NPF-9 Amendment No. 309

(4) 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; (5) Pursuant to the Act and 10 CFR Parts, 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as my be produced by the operation of McGuire Nuclear Station, Units 1 and 2; and, (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such by product material as may be produced by the Duke Training and Technology Center.

C. This renewed operating 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 thereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3469 megawatts thermal ( 100% ).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 288 are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.

Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59, and otherwise complies with the requirements in that section.

Renewed License No. NPF-17 Amendment No. 288

Definitions 1.1 1.1 Definitions (continued)

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE(ESF)RESPONSE when the monitored parameter exceeds its ESF actuation TIME 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.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.SSa(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

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

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

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

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

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

(continued)

McGuire Units 1 and 2 1.1-3 Amendment Nos. 309 / 288

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift settings shall be ~ TESTING 2460 psig and ~ 251 O psig. PROGRAM McGuire Units 1 and 2 3.4.10-2 Amendment Nos. 309 / 288

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ------------------------------NOTE------------------------------------

1. Not required to be performed in MODES 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 In accordance with gpm per nominal inch of valve size up to a maximum of 5 the INSERVICE gpm at an RCS pressure~ 2215 psig and~ 2255 psig. TESTING PROGRAM, and in accordance with the Surveillance Frequency Control Program Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve (continued)

McGuire Units 1 and 2 3.4.14-3 Amendment Nos. 309 I 288

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued}

SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is In accordance with not locked, sealed, or otherwise secured in position, the Surveillance actuates to the correct position on an actual or simulated Frequency Control actuation signal. Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual In accordance with or simulated actuation signal. the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each In accordance with position stop is in the correct position. the Surveillance Frequency Control Centrifugal Charging Safety Injection Program Pump Injection Throttle Pump Throttle Valve Number Valve Number Nl480 Nl488 Nl481 Nl489 Nl482 Nl490 Nl483 Nl491 SR 3.5.2.8 Verify, by visual inspection, that the ECCS containment In accordance sump strainer assembly and the associated enclosure are with the not restricted by debris and show no evidence of structural Surveillance distress or abnormal corrosion. Frequency Control Program McGuire Units 1 and 2 3.5.2-3 Amendment Nos. 309 I 288

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 6. 3. 4 -------------- *------------------NOTE---------------------------------

Va Ives and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind Prior to entering flange that is located inside containment or annulus and MODE 4from not locked, sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed, except for performed within containment isolation valves that are open under the previous administrative controls. 92 days SR 3.6.3.5 Verify the isolation time of automatic power operated In accordance with containment isolation valve is within limits. the INSERVICE TESTING PROGRAM In accordance with SR 3.6.3.6 Perform leakage rate testing for containment purge lower the Containment and upper compartment and incore Instrument room Leakage Rate valves with resilient seals. Testing Program SR 3.6.3.7 Verify each automatic containment isolation valve that is In accordance with not locked, sealed or otherwise secured in position, the Surveillance actuates to the isolation position on an actual or Frequency control simulated actuation signal. Program (continued)

McGuire Units 1 and 2 3.6.3-6 Amendment Nos. 309 I 288

Containment Spray System 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.2 Verify each containment spray pump's developed head at In accordance with the flow test point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM SR 3.6.6.3 Not Used Not Used SR 3.6.6.4 Not Used Not Used SR 3.6.6.5 Verify that each spray pump is de-energized and In accordance with prevented from starting upon receipt of a terminate signal the Surveillance and is allowed to manually start upon receipt of a start Frequency Control permissive from the Containment Pressure Control Program System (CPCS).

SR 3.6.6.6 Verify that each spray pump discharge valve closes or is In accordance with prevented from opening upon receipt of a terminate the Surveillance signal and is allowed to manually open upon receipt of a Frequency Control start permissive from the Containment Pressure Control Program System (CPCS).

SR 3.6.6.7 Verify each spray nozzle is unobstructed. Following activities which could result in nozzle blockage SR 3.6.6.8 Verify containment spray locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency control Program McGuire Units 1 and 2 3.6.6-2 Amendment Nos. 309 I 288

MSSVs 3.7.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

8.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators with less than two MSSVs OPERABLE.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7.1.1 --------------------------------NOTE----------------------------------

Only required to be performed prior to entry into MODE 2.

Verify each required MSSV lift setpoint per Table 3. 7.1-2 In accordance with in accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift setting shall be within TESTING

+/-.1%. PROGRAM McGuire Units 1 and 2 3.7.1-2 Amendment Nos. 309 / 288

MSIVs 3.7.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C not AND met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 --------------------------------NOTE---------------------------------- In accordance with Only required to be performed prior to entry into MODE 2. the INSERVICE TESTING PROGRAM Verify closure time of each MSIV is ~ 8.0 seconds on an actual or simulated actuation signal.

McGuire Units 1 and 2 3.7.2-2 Amendment Nos. 309 I 288

MFIVs, MFCVs, MFCV's Bypass Valves, and MFW/AFW NBVs 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more MFCV's C. 1 Close or isolate MFCV's 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months bypass valves or bypass valve or MFW/AFW MFW/AFW NBVs NBV.

inoperable.

AND C.2 Verify MFCV's bypass Once per valve or MFW/AFW NBV is 7 days closed or isolated.

D. Two valves in the same D.1 Isolate affected flow path. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow path inoperable.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify the closure time of each MFIV, MFCV, MFCVs In accordance with bypass valve, and MFW/AFW NBV is :::; 10 seconds on an the INSERVICE actual or simulated actuation signal. TESTING PROGRAM McGuire Units 1 and 2 3.7.3-2 Amendment Nos. 309 / 288

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7. 5. 1 --------------------------------NO TE----------------------------------

Not applicable to automatic valves when THERMAL POWER is ~ 10% RTP.

Verify each AFW manual, power operated, and automatic In accordance valve in each water flow path, and in both steam supply with the flow paths to the steam turbine driven pump, that is not Surveillance locked, sealed, or otherwise secured in position, is in the Frequency correct position. Control Program SR 3. 7. 5. 2 --------------------------------NO TE----------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ~ 900 psig in the steam generator.

Verify the developed head of each AFW pump at the flow In accordance test point is greater than or equal to the required with the developed head. IN SERVICE TESTING PROGRAM SR 3. 7. 5. 3 --------------------------------NO TE----------------------------------

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not locked, In accordance sealed, or otherwise secured in position, actuates to the with the correct position on an actual or simulated actuation Surveillance signal. Frequency Control Program (continued)

McGuire Units 1 and 2 3.7.5-3 Amendment Nos. 309 / 288

5.5 Programs and Manuals (continued) 5.5.8 ,lnservice Testing Program (Deleted)

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

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 for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

( continued)

McGuire Units 1 and 2 5.5-6 Amendment Nos. 309 / 288

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 DUKE ENERGY CAROLINAS. LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 409 Renewed License No. DPR-38

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

A. The application for amendment to the Oconee Nuclear Station, Unit No. 1 (the facility), Renewed Facility Operating License No. DPR-38, filed by the Duke Energy Carolinas, LLC (the licensee), dated November 7, 2017, 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.

Enclosure 5

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and Paragraph 3.B of Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 409 are hereby incorporated in the license. The licensee 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.

FOR THE NUCLEAR REGULATORY COMMISSION

~~%

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-38 and the Technical Specifications Date of Issuance: August 15, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 411 Renewed License No. DPR-47

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

A. The application for amendment to the Oconee Nuclear Station, Unit No. 2 (the facility), Renewed Facility Operating License No. DPR-47, filed by the Duke Energy Carolinas, LLC (the licensee), dated November 7, 2017, 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.

Enclosure 6

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and Paragraph 3.8 of Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 411 are hereby incorporated in the license. The licensee 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.

FOR THE NUCLEAR REGULATORY COMMISSION

(; ,!\\&-/

l\d)o~-*

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-47 and the Technical Specifications Date of Issuance: August 15, 201 8

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 410 Renewed License No. DPR-55

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

A. The application for amendment to the Oconee Nuclear Station, Unit No. 3 (the facility), Renewed Facility Operating License No. DPR-55, filed by the Duke Energy Carolinas, LLC (the licensee), dated November 7, 2017, 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.

Enclosure 7

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and Paragraph 3.8 of Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 410 are hereby incorporated in the license. The licensee 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.

FOR THE NUCLEAR REGULATORY COMMISSION 0X ~

l 0\/~7.

\J 't:§J/ .

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-55 and the Technical Specifications Date of Issuance: August 15, 2018

ATTACHMENT TO OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3 LICENSE AMENDMENT NO. 409 RENEWED FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND LICENSE AMENDMENT NO. 411 RENEWED FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND LICENSE AMENDMENT NO. 410 RENEWED FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert DPR-38, page 3 DPR-38, page 3 DPR-47, page 3 DPR-47, page 3 DPR-55, page 3 DPR-55, page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert Remove Insert 1.1-3 1.1-3 3.7.5-4 3.7.5-4 3.4.10-2 3.4.10-2 3.7.10-2 3.7.10-2 3.5.2-5 3.5.2-5 3.7.10-3 3.7.10-3 3.5.3-3 3.5.3-3 3.7.19-1 3.7.19-1 3.6.3-5 3.6.3-5 3.7.19-3 3.7.19-3 3.6.5-4 3.6.5-4 3.10.1-5 3.10.1-5 3.7.1-1 3.7.1-1 5.0-12 5.0-12 3.7.3-2 3.7.3-2 5.0-13 5.0-13

A Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 409 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-38 Amendment No. 409

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 411 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ,r1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-47 Amendment No. 411

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 410 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. This license is subject to the following antitrust conditions:

Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.

Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1J1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.

1. As used herein:

(a) "Bulk Power" means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.

(b) "Neighboring Entity" means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and transmission of electricity which meets each of Renewed License No. DPR-55 Amendment No. 410

Definitions 1.1 1.1 Definitions (continued)

CONTROL RODS CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations.

CORE ALTERATION CORE AL TERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

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

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of the Environmental Protection Agency (EPA) Federal Guidance Report No. 11.

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

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

OCONEE UNITS 1, 2, & 3 1.1-3 Amendment Nos. 409, 411, & 410

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

OCONEE UNITS 1, 2, & 3 3.4.10-2 Amendment Nos. 409, 411, & 410

HPI 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify each HPI pump's developed head at the In accordance with the test flow point is greater than or equal to the INSERVICE TESTING required developed head. PROGRAM SR 3.5.2.4 Verify each HPI automatic valve in the flow In accordance with the path that is not locked, sealed, or otherwise Surveillance Frequency secured in position, actuates to the correct Control Program position on an actual or simulated actuation signal.

SR 3.5.2.5 Verify each HPI pump starts automatically on In accordance with the an actual or simulated actuation signal. Surveillance Frequency Control Program SR 3.5.2.6 Verify, by visual inspection, each HPI train In accordance with the reactor building sump suction inlet is not Surveillance Frequency restricted by debris and suction inlet strainers Control Program show no evidence of structural distress or abnormal corrosion.

SR 3.5.2.7 Cycle each HPI discharge crossover valve and In accordance with the LPI-HPI flow path discharge valve. Surveillance Frequency Control Program OCONEE UNITS 1, 2, & 3 3.5.2-5 Amendment Nos. 409, 411, & 410

LPI 3.5.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.3.2 Verify LPI locations susceptible to gas In accordance with the accumulation are sufficiently filled with water. Surveillance Frequency Control Program SR 3.5.3.3 Verify each LPI pump's developed head at the In accordance with the

test flow point is greater than or equal to the INSERVICE TESTING required developed head. PROGRAM SR 3.5.3.4 Verify each LPI automatic valve in the flow In accordance with the path that is not locked, sealed, or otherwise Surveillance Frequency secured in position, actuates to the correct Control Program position on an actual or simulated actuation signal.

SR 3.5.3.5 Verify each LPI pump starts automatically on In accordance with the an actual or simulated actuation signal. Surveillance Frequency Control Program SR 3.5.3.6 Verify, by visual inspection, each LPI train In accordance with the reactor building sump suction inlet is not Surveillance Frequency restricted by debris and suction inlet strainers Control Program show no evidence of structural distress or abnormal corrosion.

OCONEE UNITS 1, 2, & 3 3.5.3-3 Amendment Nos. 409,411, & 410

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.4. Verify the isolation time of each automatic In accordance with the power operated containment isolation valve is INSERVICE TESTING within limits. PROGRAM SR 3.6.3.5 Verify each automatic containment isolation In accordance with the valve that is not locked, sealed, or otherwise Surveillance Frequency secured in position, actuates to the isolation Control Program position on an actual or simulated actuation signal.

OCONEE UNITS 1, 2, & 3 3.6.3-5 Amendment Nos. 409, 411, & 410

Reactor Building Spray and Cooling Systems 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 ---------------------------N()TE-----------------------

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

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

SR 3.6.5.2 ()perate each required reactor building In accordance with the cooling train fan unit for ;;:: 15 minutes. Surveillance Frequency Control Program SR 3.6.5.3 Verify each required reactor building spray In accordance with the pump's developed head at the flow test point INSERVICE TESTING is greater than or equal to the required PROGRAM developed head.

SR 3.6.5.4 Verify that the containment heat removal In accordance with the capability is sufficient to maintain post Surveillance Frequency accident conditions within design limits. Control Program (continued)

()C()NEE UNITS 1, 2, & 3 3.6.5-4 Amendment Nos. 409, 411, & 410

MSRVs 3.7.1

3. 7 PLANT SYSTEMS 3.7.1 Main Steam Relief Valves (MSRVs)

LCO 3.7.1 Eight MSRVs shall be OPERABLE on each main steam line.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more MSRVs A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND A.2 Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 ------------------------NOTE---------------------------

Only required to be performed in MODES 1 and 2.

Verify each MSRV lift setpoint in accordance In accordance with the with the INSERVICE TESTING PROGRAM. INSERVICE TESTING PROGRAM OCONEE UNITS 1, 2, & 3 3.7.1-1 Amendment Nos. 409, 411, & 410

MFCVs and SFCVs 3.7.3 ACTIONS (continued}

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 -------------------------NOTE--------------------------

Only required to be performed in MODES 1 and 2.

Verify the closure time of each MFCV and In accordance with the SFCV is ::;; 25 seconds on an actual or INSERVICE TESTING simulated actuation signal. PROGRAM OCONEE UNITS 1, 2, & 3 3.7.3-2 Amendment Nos. 409, 411, & 410

EFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each EFW manual, and non-automatic In accordance with the power operated valve in each water flow path Surveillance Frequency and in the steam supply flow path to the Control Program turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2 Verify the developed head of each EFW pump In accordance with the at the flow test point is greater than or equal INSERVICE TESTING to the required developed head. PROGRAM SR 3.7.5.3 --------------------------NOTE-------------------------

Not required to be met in MODES 3 and 4.

Verify each EFW automatic valve that is not In accordance with the locked, sealed, or otherwise secured in Surveillance Frequency position, actuates to the correct position on an Control Program actual or simulated actuation signal.

SR 3.7.5.4 --------------------------NOTE--------------------------

Not required to be met in MODES 3 and 4.

Verify each EFW pump starts automatically In accordance with the on an actual or simulated actuation signal. Surveillance Frequency Control Program SR 3.7.5.5 Verify proper alignment of the required EFW Prior to entering MODE 2 flow paths by verifying valve alignment from whenever unit has been the upper surge tank to each steam in MODE 5 or 6 for > 30 generator. days OCONEE UNITS 1, 2, & 3 3.7.5-4 Amendment Nos. 409, 411, & 410

PSW System 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Verify the required PSW battery terminal voltage In accordance with the is greater than or equal to the minimum Surveillance Frequency established float voltage. Control Program.

SR 3.7.10.2 Verify the required Keowee Hydroelectric Station In accordance with the power supply can be aligned to and power the Surveillance Frequency PSW electrical system. Control Program.

SR 3.7.10.3 Verify developed head of PSW primary and In accordance with the booster pumps at flow test point is greater than or INSERVICE TESTING equal to the required developed head. PROGRAM.

SR 3.7.10.4 Verify PSW battery capacity of the required In accordance with the battery is adequate to supply, and maintain in Surveillance Frequency OPERABLE status, required emergency loads for Control Program.

the design duty cycle when subjected to a battery service test.

SR 3.7.10.5 Verify the required PSW battery charger supplies In accordance with the

~ 300 amps at greater than or equal to the Surveillance Frequency minimum established float voltage for > 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Control Program.

OR Verify the required battery charger can recharge the battery to the fully charged state within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding PSW event discharge state.

SR 3.7.10.6 ----------------------------N()TE-----------------------------

Both HPI pump motors are individually tested although only one (1) HPI pump motor is required to support PSW system OPERABILITY.

Verify that the required PSW switchgear and In accordance with the transfer switches can be aligned and power both Surveillance Frequency the "A" and "B" HPI pump motors. Control Program.

SR 3.7.10.7 Perform functional test of required power transfer In accordance with the switches used for pressurizer heaters, PSW Surveillance Frequency control, electrical panels, vital l&C chargers, and Control Program.

valves.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10-2 Amendment Nos. 409, 411, & 410

PSW System 3.7.10 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.10.8 ----------------------------N()TE-----------------------------

Cooling water flow to the HPI pump motors are individually tested although only flow to the HPI pump motor aligned to PSW power is required to support PSW system ()PERABILITY.

Verify PSW booster pump and valves can provide In accordance with the adequate cooling water flow to HPI pump motor INSERVICE TESTING coolers. PROGRAM.

SR 3.7.10.9 Verify developed head of PSW portable pump at In accordance with the the flow test point is greater than or equal to Surveillance Frequency required developed head. Control Program.

SR 3.7.10.10 Verify the required PSW valves are tested in In accordance with the accordance with the INSERVICE TESTING INSERVICE TESTING PR()GRAM. PROGRAM.

SR 3.7.10.11 Perform CHANNEL CHECK for each required In accordance with the PSW instrument channel. Surveillance Frequency Control Program.

SR 3.7.10.12 Perform CHANNEL CALIBRATl()N for each In accordance with the required PSW instrument channel. Surveillance Frequency Control Program.

SR 3.7.10.13 Verify for the required PSW battery that the cells, In accordance with the cell plates and racks show no visual indication of Surveillance Frequency physical damage or abnormal deterioration that Control Program.

could degrade battery performance.

()C()NEE UNITS 1, 2, & 3 3.7.10-3 Amendment Nos. 409, 411, & 410

SFPC Purification System Isolation from BWST 3.7.19

3. 7 Plant Systems

3. 7 .19 Spent Fuel Pool Cooling (SFPC) Purification System Isolation from Borated Water Storage Tank (BWST)

LCO 3. 7 .19 a. Two SFPC Purification System BWST automatic isolation valves shall be OPERABLE.

b. SFPC Purification System branch line manual valves shall be closed and meet INSERVICE TESTING PROGRAM leakage requirements.

APPLICABILITY: MODES 1, 2, 3 and 4 when the SFPC Purification System is not isolated from the BWST ACTIONS

~----NOTES------------------------------------------------------

1. SFPC Purification System flow path from the BWST may be unisolated intermittently under administrative controls.

2. Separate Condition entry allowed for each SFPC Purification System branch line manual valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. One automatic isolation A. 1 Isolate the flow path by 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months valve inoperable. use of at least one closed and de-activated automatic valve, one closed and de-activated non-automatic power operated valve, closed manual valve, or blind flange.

AND A.2 Verify the flow path is Once per 31 days isolated.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.19-1 Amendment Nos. 409, 411, & 410

SFPC Purification System Isolation from BWST 3.7.19 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.19.1 Verify SFPC Purification System branch line manual In accordance with valves that are not locked, sealed, or otherwise the Surveillance secured in position are closed. Frequency Control Program SR 3.7.19.2 Verify SFPC Purification System branch line manual In accordance with valves meet INSERVICE TESTING PROGRAM the INSERVICE Leakage Requirements. TESTING PROGRAM SR 3.7.19.3 Verify SFPC Purification System BWST automatic In accordance with isolation valves are OPERABLE in accordance with the INSERVICE the INSERVICE TESTING PROGRAM. TESTING PROGRAM SR 3.7.19.4 Verify each SFPC Purification System BWST In accordance with automatic isolation valve that is not locked, sealed, or the Surveillance otherwise secured in position, actuates to the isolation Frequency Control position on an actual or simulated actuation signal. Program OCONEE UNITS 1, 2, & 3 3.7.19-3 Amendment Nos. 409, 411, & 410

SSF 3.10.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.10.1.11 Verify for required SSF battery that the cell to In accordance with the cell and terminal connections are clean, tight Surveillance Frequency and coated with anti-corrosion material. Control Program SR 3.10.1.12 Verify battery capacity of required battery is In accordance with the adequate to supply, and maintain in Surveillance Frequency OPERABLE status, the required maximum Control Program loads for the design duty cycle when subjected to a battery service test.

SR 3.10.1.13 Perform CHANNEL CALIBRATION for each In accordance with the required SSF instrument channel. Surveillance Frequency Control Program SR 3.10.1.14 Verify OPERABILITY OF SSF valves in In accordance with the accordance with the INSERVICE TESTING INSERVICE TESTING PROGRAM. PROGRAM SR 3.10.1.15 ----------------------NOTE-----------------------------

Not applicable to the SSF submersible pump.

Verify the developed head of each required In accordance with the SSF pump at the flow test point is greater INSERVICE TESTING than or equal to the required developed head. PROGRAM SR 3.10.1.16 Verify the developed head of the SSF In accordance with the submersible pump at the flow test point is Surveillance Frequency greater than or equal to the required Control Program developed head.

OCONEE UNITS 1, 2, & 3 3.10.1-5 Amendment Nos. 409, 411, & 410

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, as amended by relief granted in accordance with 10 CFR 50.55a(a)(3).

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.8 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for inspection of each reactor coolant pump flywheel.

At approximately three-year intervals, the bore and keyway of each reactor coolant pump flywheel shall be subjected to an inplace, volumetric examination.

Whenever maintenance or repair activities necessitate flywheel removal, a surface examination of exposed surfaces and a complete volumetric examination shall be performed if the interval measured from the previous such inspection is greater than 6 2/3 years. The interval may be extended up to one year to permit inspections to coincide with a planned outage.

5.5.9 lnservice Testing Program (Deleted)

NOTE: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

OCONEE UNITS 1, 2, & 3 5.0-12 Amendment Nos. 409, 411, & 410

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 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 for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

OCONEE UNITS 1, 2, & 3 5.0-13 Amendment Nos. 409, 411, & 410

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 166 Renewed License No. NPF-63

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

A. The application for amendment by Duke Energy Progress, LLC (the licensee),

dated November 7, 2017, 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 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.

Enclosure 8

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-63 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix 8, both of which are attached hereto, as revised through Amendment No. 166, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE NUCLEAR REGULATORY COMMISSION

~,~?

Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-63 and the Technical Specifications Date of Issuance: August 1 5, 2018

ATTACHMENT TO LICENSE AMENDMENT NO. 166 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following page of the Renewed Facility Operating License with the revised page.

The revised page is identified by amendment number and contains a line in the margin indicating the area of change.

Remove Insert NPF-63, Page 4 NPF-63, Page 4 Replace the following pages of the Appendix A Technical Specifications (TS) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert i i 1-3 1-3 3/4 1-9 3/4 1-9 3/41-10 3/4 1-10 3/4 4-8 3/4 4-8 3/4 4-9 3/4 4-9 3/4 4-12 3/4 4-12 3/4 5-5 3/4 5-5 3/4 6-11 3/4 6-11 3/4 6-15 3/4 6-15 3/4 6-32 3/4 6-32 3/4 7-1 3/4 7-1 3/4 7-9 3/4 7-9 3/4 7-30 3/4 7-30 6-19g 6-19g

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 I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1) Maximum Power Level Duke Energy Progress, LLC, is authorized to operate the facility at reactor Core power levels not in excess of 2948 megawatts thermal (100 percent rated core power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 166, are hereby incorporated into this license. Duke Energy Progress, LLC. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Antitrust Conditions Duke Energy Progress, LLC. shall comply with the antitrust conditions delineated in Appendix C to this license.

(4) Initial Startup Test Program (Section 14) 1 Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(5) Steam Generator Tube Rupture (Section 15. 6. 3)

Prior to startup following the first refueling outage, Carolina Power & Light Company* shall submit for NRC review and receive approval if a steam generator tube rupture analysis, including the assumed operator actions, which demonstrates that the consequences of the design basis steam generator tube rupture event for the Shearon Harris Nuclear Power Plant are less than the acceptance criteria specified in the Standard Review Plan, NUREG-0800, at 15.6.3 Subparts II (1) and (2) for calculated doses from radiological releases. In preparing their analysis Carolina Power &

Light Company* will not assume that operators will complete corrective actions within the first thirty minutes after a steam generator tube rupture.

1 The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

On April 29, 2013, the name "Carolina Power & Light Company" (CP&L) was changed to "Duke Energy Progress, Inc." On August 1, 2015, the name "Duke Energy Progress, Inc." was changed to "Duke Energy Progress, LLC."

Renewed License No. NPF-63 Amendment No. 166

INDEX 1.0 DEFINITIONS SECTION PAGE 1.1 ACTION ......................................................................................................... 1-1 1.2 ACTUATION LOGIC TEST ............................................................................ 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST ................................................. 1-1 1.4 AXIAL FLUX DIFFERENCE ........................................................................... 1-1 1.5 CHANNEL CALIBRATION ............................................................................. 1-1 1.6 CHANNEL CHECK ........................................................................................ 1-1 1.7 CONTAINMENT INTEGRITY ........................................................................ 1-2 1.8 CONTROLLED LEAKAGE ............................................................................. 1-2 1.9 CORE AL TERATION ..................................................................................... 1-2 1.9a CORE OPERATING LIMITS REPORT .......................................................... 1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST .................................................. 1-2 1.11 DOSE EQUIVALENT l-131 .......................................................................... 1-2a 1.12 E -AVERAGE DISINTEGRATION ENERGY ................................................ 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME ............................. 1-3 1.14 EXCLUSION AREA BOUNDARY .................................................................. 1-3 1.15 FREQUENCY NOTATION ............................................................................. 1-3 1.16 (DELETED) .................................................................................................... 1-3 1.17 IDENTIFIED LEAKAGE ................................................................................. 1-3 1.17a INSERVICE TESTING PROGRAM ............................................................... 1-3 1.18 MASTER RELAY TEST ................................................................................. 1-4 1.19 MEMBER(S) OF THE PUBLIC ...................................................................... 1-4 1.20 OFFS ITE DOSE CALCULATION MANUAL .................................................. 1-4 1.21 OPERABLE - OPERABILITY ......................................................................... 1-4 1.22 OPERATIONAL MODE - MODE. ................................................................... 1-4 1.23 PHYSICS TESTS ........................................................................................... 1-4 1.24 PRESSURE BOUNDARY LEAKAGE ............................................................ 1-4 1.25 PROCESS CONTROL PROGRAM ............................................................... 1-5 1.26 PURGE - PURGING ...................................................................................... 1-5 1.27 QUADRANT POWER TILT RATI0 ................................................................ 1-5 1.28 RATED THERMAL POWER .......................................................................... 1-5 1.29 REACTOR TRIP SYSTEM RESPONSE TIME .............................................. 1-5 1.30 REPORTABLE EVENT .................................................................................. 1-5 1.31 SHUTDOWN MARGIN .................................................................................. 1-5 SHEARON HARRIS - UNIT 1 Amendment No. 166

DEFINITIONS E -AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration (MeV/d) for isotopes, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

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

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

EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the land is not controlled by the licensee to limit access.

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

1.16 (DELETED)

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

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

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

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).

INSERVICE TESTING PROGRAM 1.17a The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

SHEARON HARRIS - UNIT 1 1-3 Amendment No. 166

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging/safety injection pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 4*, 5*11, and 6*#.

ACTION:

With no charging/safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging/safety injection pump shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the reactor coolant system and reactor coolant pump seals, that a differential pressure across the pump of greater than or equal to 2446 psid is developed when tested pursuant to the INSERVICE TESTING PROGRAM.

4.1.2.3.2 All charging/safety injection pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable** by verifying that each pump's motor circuit breaker is secured in the open position prior to the temperature of one or more of the RCS cold legs decreasing below 325°F and at least once per 31 days thereafter, except when the reactor vessel head is removed.

A maximum of one charging/safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 325°F and the reactor vessel head is in place.

- An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator or by a manual isolation valve secured in the closed position.

For periods of no more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , when swapping pumps, it is permitted that there be no OPERABLE charging/safety injection pump. No CORE AL TE RATIONS or positive reactivity changes are permitted during this time.

SHEARON HARRIS - UNIT 1 3/4 1-9 Amendment No. 166

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging/safety injection pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With only one charging/safety injection pump OPERABLE, restore at least two charging/safety injection pumps to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging/safety injection pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

The 'A' Train charging/safety pump is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the 'A' Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the 'A' Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the 'A' Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in the HNP LAR submittal correspondence letter HNP-16-056.

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging/safety injection pumps shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the Reactor Coolant System and reactor coolant pump seals, that a differential pressure across each pump of greater than or equal to 2446 psid is developed when tested pursuant to the INSERVICE TESTING PROGRAM.

SHEARON HARRIS - UNIT 1 3/4 1-10 Amendment No. 166

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig +/- 1%.

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SHEARON HARRIS - UNIT 1 3/4 4-8 Amendment No. 166

REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig +/- 1%.*

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SHEARON HARRIS - UNIT 1 3/4 4-9 Amendment No. 166

REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of the INSERVICE TESTING PROGRAM, each PORV shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by:

a. Performing a CHANNEL CALIBRATION of the actuation instrumentation, and

b. Operating the valve through one complete cycle of full travel during MODES 3 or 4, prior to going to 325°F.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.

4.4.4.3 The accumulator for the safety-related PORVs shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by isolating the normal air and nitrogen supplies and operating the valves through a complete cycle of full travel.

SHEARON HARRIS - UNIT 1 3/4 4-12 Amendment No. 166

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying automatic interlock action of the RHR system from the Reactor Coolant System by ensuring that with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened.

2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on safety injection actuation test signal and on safety injection switchover to containment sump from an RWST Lo-Lo level test signal, and

2. Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:

a) Charging/safety injection pump, b) RHR pump.

f. By verifying that each of the following pumps develops the required differential pressure when tested pursuant to the INSERVICE TESTING PROGRAM:

1. Charging/safety injection pump (Refer to Specification 4.1.2.4)

2. RHR pump.::: 100 psid at a flow rate of at least 3663 gpm.

g. By verifying that the locking mechanism is in place and locked for the following High Head ECCS throttle valves:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and

2. At the frequency specified in the Surveillance Frequency Control Program.

SHEARON HARRIS - UNIT 1 3/4 5-5 Amendment No. 166

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and transferring suction to the containment sump.

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

ACTION:

With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Refer also to Specification 3.6.2.3 Action.

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

- The 'A' Train Containment Spray System is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the 'A' Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the 'A' Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the 'A' Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position*;

b. By verifying that, on an indicated recirculation flow of at least 1832 gpm, each pump develops a differential pressure of greater than or equal to 186 psi when tested pursuant to the INSERVICE TESTING PROGRAM;

c. At the frequency specified in the Surveillance Frequency Control Program by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation test signal and

2. Verifying that each spray pump starts automatically on a containment spray actuation test signal.

3. Verifying that, coincident with an indication of containment spray pump running, each automatic valve from the sump and RWST actuates to its appropriate position following an RWST Lo-Lo test signal.

d. At the frequency specified in the Surveillance Frequency Control Program by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

e. At the frequency specified in the Surveillance Frequency Control Program by verifying that containment spray locations susceptible to gas accumulation are sufficiently filled with water.

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

SHEARON HARRIS - UNIT 1 3/4 6-11 Amendment No. 166

CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program by:

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and

c. Verifying that on a Containment Ventilation Isolation test signal, each normal, preentry purge makeup and exhaust, and containment vacuum relief valve actuates to its isolation position, and

d. Verifying that, on a Safety Injection "S" test signal, each containment isolation valve receiving an "S" signal actuates to its isolation position, and

e. Verifying that, on a Main Steam Isolation test signal, each main steam isolation valve actuates to its isolation position, and

f. Verifying that, on a Main Feedwater Isolation test signal, each feedwater isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit specified in the Technical Specification Equipment List Program, plant procedure PLP-106, when tested pursuant to the INSERVICE TESTING PROGRAM.

SHEARON HARRIS- UNIT 1 3/4 6-15 Amendment No. 166

CONTAINMENT SYSTEMS 3/4.6.5 VACUUM RELIEF SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5 The containment vacuum relief system shall be OPERABLE with an Actuation Setpoint of equal to or less negative than -2.5 inches water gauge differential pressure (containment pressure less atmospheric pressure)

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

ACTION:

With one containment vacuum relief system inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.5 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

SHEARON HARRIS - UNIT 1 3/4 6-32 Amendment No. 166

3/4. 7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION

3. 7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one or more main steam line Code safety valves inoperable, operation may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

SHEARON HARRIS - UNIT 1 3/4 7-1 Amendment No. 166

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE.

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

ACTION:

MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2, 3, and 4:

With one MSIV inoperable, subsequent operation in MODE 2, 3, or 4 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.

SHEARON HARRIS - UNIT 1 3/4 7-9 Amendment No. 166

PLANT SYSTEMS 3/4.7.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.13 At least two independent Essential Services Chilled Water System loops shall be OPERABLE.

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

ACTION:

With only one Essential Services Chilled Water System loop OPERABLE, restore at least two loops to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

The 'A' Train Essential Services Chilled Water System loop is allowed to be inoperable for a total of 14 days only to allow for the implementation of design improvements on the 'A' Train ESW pump. The 14 days will be taken one time no later than October 29, 2016. During the period in which the 'A' Train ESW pump supply from the Auxiliary Reservoir or Main Reservoir is not available, Normal Service Water will remain available and in service to supply the 'A' Train ESW equipment loads until the system is ready for post maintenance testing. Allowance of the extended Completion Time is contingent on meeting the Compensatory Measures and Conditions described in HNP LAR submittal correspondence letter HNP-16-056.

SURVEILLANCE REQUIREMENTS 4.7.13 The Essential Services Chilled Water System shall be demonstrated OPERABLE by:

a. Performance of surveillances as required by the INSERVICE TESTING PROGRAM, and

b. At the frequency specified in the Surveillance Frequency Control Program by demonstrating that:

1. Non-essential portions of the system are automatically isolated upon receipt of a Safety Injection actuation signal, and

2. The system starts automatically on a Safety Injection actuation signal.

SHEARON HARRIS - UNIT 1 3/4 7-30 Amendment No. 166

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

m. lnservice Testing Program (Deleted)

Note: See Section 1.17a for the definition of INSERVICE TESTING PROGRAM.

SHEARON HARRIS - UNIT 1 6-19g Amendment No. 166

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 259 Renewed License No. DPR-23

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

A. The application for amendment by Duke Energy Progress, LLC (the licensee),

dated November 7, 2017, 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 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.

Enclosure 9

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and Paragraph 3.8. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 259 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION Booma Venkataraman, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-23 and the Technical Specifications Date of Issuance: August 15, 2018

ATTACHMENT TO LICENSE AMENDMENT NO. 259 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Insert DPR-23, page 3 DPR-23, page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-3 1.1-3 1.1-4 1.1-4 3.4-24 3.4-24 3.4-39 3.4-39 3.5-6 3.5-6 3.6-12 3.6-12 3.6-17 3.6-17 3.6-21 3.6-21 3.7-3 3.7-3 3.7-7 3.7-7 3.7-9 3.7-9 5.0-11 5.0-11

3 D. 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 and equipment calibration or associated with radioactive apparatus or components; E. 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 operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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:

A. Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 259 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

( 1) For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Renewed Facility Operating License No. DPR-23 Amendment No. 259

Definitions 1.1 1.1 Definitions E-AVERAGE iodines, with half lives > 15 minutes, making up DISINTEGRATION ENERGY at least 95% of the total noniodine activity in (continued) the coolant.

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

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or return), 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);

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

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

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

(continued)

HBRSEP Unit No. 2 1.1-3 Amendment No. 259

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

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

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

a. Described in Chapter 14, Initial Test Program of the Updated Final Safety Analysis Report (UFSAR);

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

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2339 MWt.

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

(continued)

HBRSEP Unit No. 2 1.1-4 Amendment No. 259

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE in In accordance with accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift settings shall be TESTING within+/- 1%. PROGRAM HBRSEP Unit No. 2 3.4-24 Amendment No. 259

RCS PIVs 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 --------------------------------N()TES---------------------------

1. Not required to be performed in MODES 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 less than or In accordance with equal to an equivalent of 5 gpm at an RCS pressure the INSERVICE

~ 2235 psig, and verify the margin between the TESTING results of the previous leak rate test and the 5 gpm PROGRAM and limit has not been reduced by ~ 50% for valves with 24 months leakage rates> 1.0 gpm.

Prior to entering MODE2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months (continued)

HBRSEP Unit No. 2 3.4-39 Amendment No. 259

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM SR 3.5.2.4 Verify each ECCS automatic valve in the flow path 24 months that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.5 Verify each ECCS pump starts automatically on an 24 months actual or simulated actuation signal.

SR 3.5.2.6 Verify, by visual inspection, the ECCS containment 24 months sump suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion.

(continued)

HBRSEP Unit No. 2 3.5-6 Amendment No. 259

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.4 Verify the isolation time of each automatic power In accordance operated containment isolation valve is within limits. with the INSERVICE TESTING PROGRAM SR 3.6.3.5 Verify each automatic containment isolation valve 24 months that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

SR 3.6.3.6 Verify each 42 inch inboard containment purge valve 24 months is blocked to restrict the valve from opening > 70°.

HBRSEP Unit No. 2 3.6-12 Amendment No. 259

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.2 Operate each containment cooling train fan unit for 31 days

~ 15 minutes.

SR 3.6.6.3 Verify cooling water flow rate to each cooling unit is 31 days

~ 750 gpm.

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 24 months flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.6.6 Verify each containment spray pump starts 24 months automatically on an actual or simulated actuation signal.

SR 3.6.6.7 Verify each containment cooling train starts 24 months automatically on an actual or simulated actuation signal.

SR 3.6.6.8 Verify each spray nozzle is unobstructed. Following activities which could result in nozzle blockage HBRSEP Unit No. 2 3.6-17 Amendment No. 259

Isolation Valve Seal Water System 3.6.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.8.3 Verify the opening time of each air operated In accordance with header injection valve is within limits. the INSERVICE TESTING PROGRAM SR 3.6.8.4 Verify each automatic valve in the IVSW System 24 months actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.8.5 Verify the IVSW dedicated nitrogen bottles will 24 months pressurize the IVSW tank to ~ 46.2 psig.

SR 3.6.8.6 Verify total IVSW seal header flow rate is 24 months s 124 cc/minute HBRSEP Unit No. 2 3.6-21 Amendment No. 259

MSSVs 3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 -----------------------------N()TE------------~------------------

()nly required to be performed in M()DES 1 and 2.

Verify each required MSSV lift setpoint per Table In accordance with 3.7.1-2 in accordance with the INSERVICE TESTING the INSERVICE PROGRAM. Following testing, lift setting shall be TESTING within ,:!:1%. PR()GRAM HBRSEP Unit No. 2 3.7-3 Amendment No. 259

MSIVs 3.7.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C not met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 -------------------------NOTE----------~----------~------------

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is within limits on an In accordance with actual or simulated actuation signal. the INSERVICE TESTING PROGRAM HBRSEP Unit No. 2 3.7-7 Amendment No. 259

MFIVs, MFRVs, and Bypass Valves 3.7.3 ACTIONS (continued}

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more bypass C.1 Close or isolate bypass 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months valves inoperable. valve.

AND C.2 Verify bypass valve is Once per 7 days closed or isolated.

D Two valves in the same D.1 Isolate affected flow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow path inoperable. path.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify the closure time of each MFRV and bypass In accordance with valve is within limits on an actual or simulated the INSERVICE actuation signal. TESTING PROGRAM SR 3.7.3.2 Verify the closure time of each MFIV is within limits on In accordance with an actual or simulated actuation signal. the INSERVICE TESTING PROGRAM HBRSEP Unit No. 2 3.7-9 Amendment No. 259

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program provides controls for the inspection of each reactor coolant pump flywheel in accordance with the lnservice Inspection Program.

5.5.8 lnservice Testing Program (Deleted)

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

(continued)

HBRSEP Unit No. 2 5.0-11 Amendment No. 259

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 299 AND 295 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-35 AND NPF-52 AMENDMENT NOS. 309 AND 288 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-9 AND NPF-17 AMENDMENT NOS. 409,411, AND 410 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-38, DPR-47, AND DPR-55 AMENDMENT NO. 166 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 AMENDMENT NO. 259 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DUKE ENERGY CAROLINAS, LLC AND DUKE ENERGY PROGRESS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413, 50-414 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369, 50-370 OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, 50-287 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

By application dated November 7, 2017 (Reference 1), Duke Energy Progress, LLC, and Duke Energy Carolinas, LLC (the licensee) requested changes to the Technical Specifications (TSs) for Catawba Nuclear Station, Units 1 and 2 (Catawba); McGuire Nuclear Station, Units 1 and 2 (McGuire); Oconee Nuclear Station, Unit Nos. 1, 2, and 3 (Oconee); Shearon Harris Nuclear Power Plant, Unit 1 (Shearon Harris); and H.B. Robinson Steam Electric Plant, Unit No. 2 Enclosure 10

(Robinson). 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 [1ST] Program Removal & [and] Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing," dated October 21, 2015 (Reference 2).

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 system, structure, or component after first electrical generation by nuclear heat. The American Society of Mechanical Engineers Operation and Maintenance of Nuclear Power Plants (ASME OM Code) provides requirements for inservice testing (1ST) of certain components in light-water nuclear (LWR) power plants. The ASME OM Code identifies the components subject to testing (i.e., pumps, valves, pressure relief devices, and dynamic restraints), methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and record keeping. Paragraph 50.55a(f), "lnservice testing requirements," of Title 10 of the Code of Federal Regulations ( 10 CFR) requires that 1ST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The TSs also prescribe 1ST requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation under 10 CFR 50.55a(f)(5)(ii) states that if a revised inservice test program for a facility conflicts with the TSs, the licensee must apply to the Commission for amendment of the TSs to conform the TSs to the revised program. Revision 3 of TSTF-545 provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a and the TSs. Revision 3 of TSTF-545 proposes elimination of the 1ST Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the 1ST Program. The elimination of the 1ST Program from the TSs could cause confusion about the correct application of these surveillance requirements. 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). Revision 3 of TSTF-545 proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

By letter dated December 11, 2015 (Reference 3), the U.S. Nuclear Regulatory Commission (NRC) found the changes to the Standard Technical Specifications (STSs) proposed in TSTF-545, Revision 3 to be suitable for incorporation into the STSs and announced that licensees could request amending their licenses to adopt TSTF-545, Revision 3. The December 15, 2015, letter also made available a model safety evaluation (SE). The NRC published a notice of availability of TSTF-545, Revision 3 in the Federal Register (FR) on March 28, 2016 (81 FR 17208). This SE incorporates clarifications, additional justifications, and formatting changes to the TSTF-545 model SE, consistent with other SEs for issued TSTF-545 related amendments.

2.2 Proposed Technical Specifications Changes The licensee requested to revise the plants' TSs by deleting the 1ST program from the Administrative Controls TS sections for Catawba (TS 5.5.8), McGuire (TS 5.5.8), Oconee

(TS 5.5.9), Shearon Harris (TS 6.8.4.m), and Robinson (TS 5.5.8), as follows, with proposed deletions shown as stricken text and additions as bolded text.

Catawba:

lnservice Testing Program (Deleted)

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

This program provides controls for inservise testing of ASME Code Class 1, 2, and a components including applicable supports. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Poi.\ler Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inseF\*ice inservice testing activities testing activities

'.'Veekly At least once per 7 days Monthly At least once per 31 days Quarterly or every a months At least once per Q2 days Semiannually or every 6 months At least once per 184 days Every Q months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less for performing inservice testing activities;

c. The provisions of SR a.a.a are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

McGuire:

lnservice Testing Program (Deleted}

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

This program provides controls f.or inservise testing of ASME Gode Glass 1, 2, and 3 components including applicable supports. The program shall include the f.ollowing:

a. Testing Frequencies applicable to the ASME Gode f.or Operation and Maintenance of Nuclear Pov,er Plants (ASME OM Gode) and applicable Addenda are as f.ollows:

ASME OM Gode and applicable Required Frequencies for Addenda terminology f.or performing insorvico testing insorvico testing activities activities Weekly ,'\t least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 day-s Every Q months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to tho above required Frequencies and other normal and accelerated Frequencies specified as 2 years or loss in tho lnsorvico Testing Program f.or performing inservico testing activities;

c. Tho provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler OM Gode shall be construed to supersede the requirements of any TS.

Oconee:

lnservice Testing Program (Deleted)

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

This program provides controls for inservice testing of ASME Code Class 1, 2, a

and pumps and valves:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days a

Quarterly or every months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities;

c. The provisions of SR a.a.a are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

Shearon Harris:

lnservice Testing Program (Deleted)

Note: See Section 1.17a for the definition of INSERVICE TESTING PROGRAM.

This program provides oontrols for inservioe testing of ASME Code Class 1, 2, a

and oomponents, The program shall inolude the following:

1) Testing frequenoies speoified in the ASME Code for Operations and Maintenanoe of Nuolear Pmver Plants and applioable Addenda as follows:

ASME Code for Operation and Maintenanoe of Nuolear Power Plants and applioable Addenda Required Frequenoies for terminology for inservioe testing performing inservioe aotivities testing aotivities

\/Veekly At least onoe per 7 days Monthly At least onoe J:IOF a1 days a

Quarterly or every months At least onoe per 92 days Semiannually or every 6 months At least onoe per 184 days Every 9 months At least onoe per 276 days Yearly or annually At least onoe per J66 days Biennially or every 2 years At least onoe per 7J 1 days

2) The provisions of SR 4.0.2 are applioable to the abo*.<<e required Frequonoies and to other normal and aooolerated Frequenoies speoified as 2 years or less in tho lnservioe Testing Program for performing insorvioe testing aotivities.

J) The provisions of SR 4.0.a are applioable to inservioe testing aotivitios. and

4) ~Jothing in tho ASME Code for Operation and Maintonanoo of Nuolear Pov,er Plants shall be oonstrued to supersede the requirements of any Teohnioal Speoifioation

Robinson:

lnservice Testing Program (Deleted)

Note: See Section 1.1 for the definition of INSERVICE TESTING PROGRAM.

This program proi.*ides oontrols for inservioe testing of ASME Code Class 1, 2, and 3 pumps and 'Jalvos. Tho program shall inoludo tho following:

a. Testing frequonoios spooifiod in Sootion XI of tho ASME Boiler and Pressure Vessel Code and applioablo Addenda as follows:

ASME Boiler and Pressure Vessel Code and applioablo Required Froquonoios for Addenda terminology for performing insorvioo insorvioo testing aotivitios testing aotivitios

'.'Vookly At least onoo per 7 days Monthly l\t least onoo per 31 days Quarterly or every a months A-t least onoo per 92 days Semiannually or every 6 months At least onoo per 184 days E¥Ory 9 months At least onoo per 276 days Yearly or annually A-t least onoo per 366 days Biennially or every 2 years At least onoo per 731 days

b. Tho provisions of SR 3.0.2 are applioablo to tho above required Froquonoios for performing insorvioo testing aotivitios;

o. Tho provisions of SR a.a.a are applioablo to insorvioo testing aotivitios; and

d. Nothing in tho ASME Boiler and Pressure Vessel Code shall be oonstruod to supersede tho requirements of any TS.

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 plants' TSs be replaced with "INSERVICE TESTING PROGRAM," so that it refers to the new definition in lieu of the deleted program.

For Shearon Harris, the licensee proposed conforming changes to the TSs' index pages denoting the addition of the new definition.

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. Paragraph 50.36(c)(3) states that SRs 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. Paragraph 50.36(c)(5) states that administrative 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 Edition," Chapter 16,

Technical Specifications," Revision 3, March 2010 (Reference 4). As described therein, as part of the regulatory standardization effort, the NRC 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 amendments are based on TSTF-545, Revision 3, which is an NRG-approved change to the improved STSs. The NRC staffs review included consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. The NRC staff gives special attention 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.

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 1ST of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f).

Section 50.55a(f) states in-part that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel (BPV) Code and ASME OM Code as specified in the paragraph, and that each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions of 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 1ST program requirements were suitable for incorporation into the NRC's rules.

Paragraph 50.55(a)(f){5)(ii) of 10 CFR states in part that if a revised inservice test program for a facility conflicts with the TSs for the facility, the licensee must apply for an amendment of the TSs to conform the TSs to the revised program.

Revision 2 of NUREG-1482, "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 (Reference 5),

provides guidance for the 1ST 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 (Reference 6), provides guidance and acceptance criteria for the NRC staff's review of the 1ST 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 regulations, guidance, and licensing information discussed in Section 2.3 of this SE. 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). The NRC staff also considered whether the TSs, as amended, would 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 per 10 CFR 50.36(c)(3). 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 Technical Specifications For Catawba, McGuire, Oconee, Shearon Harris, and Robinson, the 1ST program TSs are TS 5.5.8, TS 5.5.8, TS 5.5.9, TS 6.8.4.m, and TS 5.5.8, respectively, which are in the Administrative Controls section of the TSs. The 1ST program TSs have requirements for 1ST 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 1ST of pumps and valves. Therefore, requiring the licensee to have an 1ST 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 1ST program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained further in this SE, it is not necessary to have additional administrative controls in the TSs for these plants relating to the 1ST program to assure operation of the facility in a safe manner.

Deletion of Technical Specification lnservice Testing Program Frequency Descriptions The ASME OM Code requires testing to normally be performed within certain time periods. The plants' current TS 1ST program proposed to be deleted set 1ST frequencies more precisely 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 1ST frequencies are not necessary to assure operation of the facility in a safe manner.

Therefore, the NRC staff finds these proposed changes acceptable.

Deletion of Surveillance Requirement 3.0.2/4.0.2 Provisions from lnservice Testing Program Technical Specifications The plants' current TS 1ST program proposed to be deleted allows the licensee to use SR 3.0.2 (SR 4.0.2 for Shearon Harris) to extend, by up to 25 percent, the interval between 1ST activities, as required by the plants' 1ST TSs. Similar to these TSs, Code Case OMN-20, incorporated by reference in 10 CFR 50.55a, also permits the licensee to extend the 1ST intervals specified in

the ASME OM Code by up to 25 percent. The NRC staff determined that the TS allowance to extend 1ST intervals is not needed to assure operation of the facility in a safe manner.

Therefore, the NRC staff determined that deletion of these TSs is acceptable. Therefore, the NRC staff finds these proposed changes acceptable.

Deletion of Surveillance Requirement 3.0.3/4.0.3 Provisions from lnservice Testing Program Technical Specifications The plants' current TS 1ST program proposed to be deleted allows the licensee to use SR 3.0.3 (SR 4.0.3 for Shearon Harris) when it discovers that an SR associated with an inservice test was not performed within its specified frequency. Surveillance Requirement 3.0.3 (or SR 4.0.3) 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 (or SR 4.0.3) 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 this TS does not change any of these requirements, and SR 3.0.3 (or SR 4.0.3) will continue to apply to those inservice tests required by SRs.

Therefore, the NRC staff finds the deletion of these TSs is acceptable.

Deletion of Catawba TS 5.5.Bd, McGuire TS 5.5.Bd, Oconee TS 5.5.9d, Shearon Harris TS 6.8.4.m4, and Robinson TS 5.5d.8.

Catawba TS 5.5.8d, McGuire TS 5.5.8d, Oconee TS 5.5.9d, Shearon Harris TS 6.8.4.m4, and Robinson TS 5.5d.8 state that nothing in the ASME OM (or BPV) 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 1ST program for a facility conflicts with the TSs for the facility.

These regulations 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 1ST program because the regulations specify how conflicts must be resolved. Therefore, the NRC staff finds the deletion of these TSs acceptable.

Conclusion Regarding Deletion of lnservice Testing Program Technical Specifications The NRC staff determined that the requirements currently in the 1ST program TSs are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the NRC staff concludes that deletion of the 1ST program TSs from the licensee's TSs is acceptable because the 1ST program TSs are not required by 10 CFR 50.36(c)(5).

3.2 Definition of INSERVICE TESTING PROGRAM and Revision to Surveillance Requirements The licensee proposed 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 is consistent with the definition in TSTF-545, Revision 3. The NRC staff finds the definition acceptable because it correctly refers to the 1ST requirements in 10 CFR 50.55a(f).

The licensee requested that all existing references to the "lnservice Testing Program" in the TSs be revised to "INSERVICE TESTING PROGRAM" to reference the new TS defined term in lieu

of the deleted 1ST program TSs. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in TSs with the new definition. The NRC staff verified that for each reference to the "lnservice Testing Program," 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 1ST frequencies could change because the plants' TSs will no longer include the more precise test frequencies. As discussed in Section 3.1 of this SE, the NRC staff determined that the TSs do not need to include the more precise testing frequencies currently in the TSs. 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, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue 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.

3.3 Conforming Changes and Variations from TSTF-545 The NRC staff evaluated the following conforming changes and variations from TSTF-545, Revision 3, not previously addressed in this SE.

a. The TSs for Shearon Harris have not been converted to the improved STSs on which TSTF-545, Revision 3 is based. As a result, the numbering, format, and content of these TSs vary from TSTF-545, Revision 3. In addition, all the plants' TSs use different numbering than the improved STSs, on which TSTF-545, Revision 3 is based. The NRC staff finds that the licensee's proposed differences in numbering, format, and content are editorial in nature and that the licensee's proposed TS changes are consistent with the intent of TSTF-545, Revision 3. Therefore, the NRC staff finds that the licensee's proposed TS changes are acceptable.

b. The index for Shearon Harris TSs is included as part of the TSs. Therefore, the licensee included conforming changes to the index resulting from the addition of the new definition. The NRC staff finds that the proposed differences are editorial in nature and that the licensee's proposed TS changes are consistent with the intent of TSTF-545, Revision 3. Therefore, the NRC staff finds that the licensee's proposed TS changes are acceptable.

c. The 1ST program in the TSs for Robinson refers to testing frequencies specified in Section XI of the ASME BPV Code, which varies from TSTF-545, Revision 3.

The Code of record for Robinson's 1ST is the ASME OM Code, and 10 CFR 50.55a(f) requires the 1ST program to meet the ASME OM Code.

Therefore, deletion of this TSs does not create new requirements for this plant.

As discussed in Section 3.1 of this SE, the NRC staff finds the proposed deletion of the 1ST program TSs, which include these references to Section XI of the ASME BPV Code, acceptable.

d. The licensee proposed to replace the content of 1ST program TSs for all plants with the word, "Deleted," and retain the existing numbering sequence. The licensee also added a note referring to the new 1ST definition. The NRC staff finds that these proposed changes are editorial in nature and consistent with the

intent of TSTF-545, Revision 3. Therefore, the NRC staff finds that the licensee's proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the NRC staff notified officials from the States of North Carolina and South Carolina on June 21, 2018, of the proposed issuance of the amendments. Each State's official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

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

The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on the finding published in the Federal Register on January 16, 2018 (83 FR 2227). Accordingly, the amendments meet 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Henderson, K., Duke Energy Progress, LLC and Duke Energy Carolinas, LLC, letter to U.S. Nuclear Regulatory Commission, "Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, 'TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing,"' dated November 7, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17312A362).

2. Technical Specifications Task Force, Standard Technical Specifications 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 (ADAMS Accession No. ML15294A555).

3. Hsueh, K., U.S. Nuclear Regulatory Commission, letter to Technical Specifications Task Force, "Final Model Safety Evaluation of Technical Specifications Traveler TSTF-545, Revision 3, 'TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing' (TAC No. MF3349)," dated December 11, 2015 (ADAMS Package Accession No. ML15317A071 ).

4. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 16, "Technical Specifications," Revision 3, March 201 O (ADAMS Accession No. ML100351425).

5. U.S. Nuclear Regulatory Commission, NUREG-1482, "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,"

Revision 2, October 2013 (ADAMS Accession No. ML13295A020).

6. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,"

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 ).

Principal Contributors: Caroline Tilton Da~: August 15, 2018

S. Capps

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2; MCGUIRE NUCLEAR STATION, UNITS 1 AND 2; OCONEE NUCLEAR STATION, UNIT NOS. 1, 2, AND 3; SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1; AND H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENTS TO ADOPT TSTF-545, REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING" (EPID L-2017-LLA-0377) DATED AUGUST 15, 2018 DISTRIBUTION:

PUBLIC PM File Copy RidsACRS_MailCTR Resource RidsNrrDorlLpl2-1 Resource RidsNrrDorlLpl2-2 Resource RidsNrrDssStsb Resource RidsNrrLABClayton Resource RidsNrrLAJBurkhardt Resource RidsNrrLAKGoldstein Resource RidsNrrPMCatawba Resource RidsNrrPMShearonHarris Resource RidsNrrPMMcGuire Resource RidsNrrPMOconee Resource RidsNrrPMRobinson Resource RidsRgn2MailCenter Resource CTilton, NRR ADAMS Access1on No.: ML18172A172

bymemo ML18170A301 ** b1yema1*1 OFFICE NRR/DORL/LPL2-2/PM N RR/DORL/LPL2-2/LA NRR/DSS/STSB/BC*

NAME DGalvin BClayton (JBurkhardt for) VCusumano DATE 7/30/18 7/25/18 6/20/18 OFFICE OGC** NRR/DORL/LPL2-2/BC(A) NRR/DORL/LPL2-2/PM NAME BHarris (NLO) BVenkataraman DGalvin DATE 8/9/18 8/15/18 8/15/18 OFFICIAL RECORD COPY

v t e Letter Sequence Approval
EPID:L-2017-LLA-0377, TSTF-545 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2017-12-07 [Table View]

ML18172A172

Text

SUBJECT:

Dear Mr. Capps:

Enclosures:

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1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

SUBJECT:

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