Text

Issuance of Amendments to Revise the Technical Specification for Engineered Safety Feature Actuation System Instrumentation
	ML21224A101
Person / Time
Site:	Mcguire, Catawba, Harris, McGuire
Issue date:	10/07/2021
From:	Tanya Hood; Plant Licensing Branch II
To:	Snider S; Duke Energy Corp
	T Hood, NRR/DORL 301-415-1387
References
	EPID L-2020-LLA-0262, RA-20-0207
	Download: ML21224A101 (43)
	v • d • e

October 7, 2021 Mr. Steven Snider Vice President - Nuclear Engineering Nuclear Corporate Duke Energy Corporation 526 South Church Street, EC-07H Charlotte, NC 28202

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2; SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1; MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS TO REVISE THE TECHNICAL SPECIFICATION FOR ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (EPID L-2020-LLA-0262)

Dear Mr. Snider:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the following enclosed Amendment Nos. 310 and 306 to Renewed Facility Operating License Nos. NPF-35 and NPF-52 for the Catawba Nuclear Station, Units 1 and 2 (CNS), respectively; Amendment No. 186 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1; and Amendment Nos. 320 and 299 to Renewed Facility Operating License Nos. NPF-9 and NPF-17 for the McGuire Nuclear Station, Units 1 and 2, respectively; in response to your application dated December 3, 2020.

The amendments revise the Technical Specifications (TS) for the Engineered Safety Feature Actuation System Instrumentation by adding a footnote to identify the enabled functions and the applicable MODES for the Reactor Trip, P-4 interlock function. The revision removes the turbine trip function of the P-4 interlock in MODE 3 from the existing TS. In addition, only for CNS, the amendments remove the P-4 interlock steam dump function in MODES 1, 2, and 3.

A copy of the NRC staffs Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

S. Snider If you have any questions, please contact me at (301) 415-1387 or by e-mail at Tanya.Hood@nrc.gov.

Sincerely,

/RA/

Tanya E. Hood, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413, 50-414, 50-369, 50-370, and 50-400

Enclosures:

1. Amendment No. 310 to NPF-35

2. Amendment No. 306 to NPF-52

3. Amendment No. 186 to NPR-63

4. Amendment No. 320 to NPF-9

5. Amendment No. 299 to NPF-17

6. Safety Evaluation cc:

Mr. Robert T. Simril Mr. Thomas Ray Site Vice President 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 Ms. Kim Maza Site Vice President Shearon Harris Nuclear Power Plant 5413 Shearon Harris Rd, M/C HNP01 New Hill, NC 27562-9300 Additional Distribution via Listserv

DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 310 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 December 3, 2020, 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 310, 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 Digitally signed by David J.

David J. Wrona Date: 2021.10.07 18:49:17 Wrona -04'00' David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 7, 2021

DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 306 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 December 3, 2020, 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 306, 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 Digitally signed by David J. David J. Wrona Date: 2021.10.07 Wrona 18:49:52 -04'00' David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 7, 2021

ATTACHMENT TO CATAWBA NUCLEAR STATION, UNITS 1 AND 2 LICENSE AMENDMENT NO. 310 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 306 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 Insert License License NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 TSs TSs 3.3.2-18 3.3.2-18

(2) TECHNICAL SPECIFICATIONS The Technical Specifications contained in Appendix A, as revised through Amendment No. 310 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 No. 310

(2) TECHNICAL SPECIFICATIONS The Technical Specifications contained in Appendix A, as revised through Amendment No. 306, 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. 306

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

7. Automatic Switchover to Containment Sump

a. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

b. Refueling Water 1,2,3,4 4 N SR 3.3.2.1 91.9 inches 95 inches Storage Tank SR 3.3.2.7(a)(b)

(RWST) Level - SR 3.3.2.9(a)(b)

Low SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8. ESFAS Interlocks

a. Reactor 1,2,3(h) 1 per train, F SR 3.3.2.8 NA NA Trip, P-4 2 trains

b. Pressurizer 1,2,3 3 O SR 3.3.2.5 1944 and 1955 psig Pressure, P-11 SR 3.3.2.9 1966 psig

c. Tavg - Low Low, 1,2,3 1 per loop O SR 3.3.2.5 550°F 553°F P-12 SR 3.3.2.9

9. Containment Pressure Control System

a. Start Permissive 1,2,3,4 4 per train P SR 3.3.2.1 1.0 psid 0.9 psid SR 3.3.2.7 SR 3.3.2.9

b. Termination 1,2,3,4 4 per train P SR 3.3.2.1 0.25 psid 0.35 psid SR 3.3.2.7 SR 3.3.2.9

10. Nuclear Service 1,2,3,4 3 per pit Q,R SR 3.3.2.1 El. 555.4 ft El. 557.5 ft Water Suction SR 3.3.2.9 Transfer - Low Pit SR 3.3.2.11 Level SR 3.3.2.12 (a) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR.

(h) The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2

Isolate MFW with coincident low Tavg - MODES 1, 2 and 3

Prevent reactuation of SI after a manual reset of SI - MODES 1, 2, and 3

Prevent opening of MFIVs if closed on SI or SG Water Level - High High - MODES 1, 2, and 3 Catawba Units 1 and 2 3.3.2-18 Amendment Nos. 310/306

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 186 Renewed License No. NPF-63

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

A. The application for amendment to the Shearon Harris Nuclear Power Plant, Unit 1 (the facility), Renewed Facility Operating License No. NPF-63 by Duke Energy Progress, LLC (the licensee), dated December 3, 2020, 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 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-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 B, both of which are attached hereto, as revised through Amendment No. 186, 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 Digitally signed by David J.

David J. Wrona Date: 2021.10.07 18:50:27 Wrona -04'00' David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 7, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 186 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 Pages Insert Pages 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 Pages Insert Pages 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3/27 3/4 3-27a 3/4 3/27a 3/4 3-48 3/4 3-48 3/4 3-49 3/4 3-49

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. 186, 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.

1The 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 of 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. 186

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. OF CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT CHANNELS TO TRIP OPERABLE MODES ACTION

8. Containment Spray Switch-over to Containment Sump (Continued)

b. RWST--Low Low See Item 7.b. above for all RWST--Low Low initiating functions and requirements.

Coincident With Containment Spray See Item 2 above for all Containment Spray initiating functions and requirements.

9. Loss-of-Offsite Power

a. 6.9 kV Emergency Bus--Undervoltage 3/bus 2/bus 2/bus 1, 2, 3, 4 15a Primary

b. 6.9 kV Emergency Bus--Undervoltage 3/bus 2/bus 2/bus 1, 2, 3, 4 15a Secondary

10. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, P-11 3 2 2 1, 2, 3 20 Not P-11 3 2 2 1, 2, 3 20

b. Low-Low Tavg, P-12 3 2 2 1, 2, 3 20

c. Reactor Trip, P-4 2 2 2 1, 2, 3 ## 22

d. Steam Generator Water Level, P-14 See Item 5.b. above for all P-14 initiating functions and requirements.

SHEARON HARRIS - UNIT 1 3/4 3-25 Amendment No. 186

TABLE 3.3-3 (Continued)

TABLE NOTATIONS

Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock)

Setpoint.

- During CORE ALTERATIONS or movement of irradiated fuel in containment, refer to Specification 3.9.9.

- - Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

1. The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2

Isolate Main Feedwater with coincident low Tavg - MODES 1, 2, and 3

Prevent reactuation of Safety Injection after a manual reset of Safety Injection - MODES 1, 2, and 3

Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3 ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in accordance with the Risk-Informed Completion Time Program 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 ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15a - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With less than the minimum channels OPERABLE, operation may proceed provided the minimum number of channels is restored within one hour, otherwise declare the affected diesel generator inoperable. When performing surveillance testing of either primary or secondary undervoltage relays, the redundant emergency bus and associated primary and secondary relays shall be OPERABLE.

ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.

SHEARON HARRIS - UNIT 1 3/4 3-26 Amendment No. 186

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 17 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the Containment Purge Makeup and Exhaust Isolation valves are maintained closed while in MODES 1, 2, 3 and 4 (refer to Specification 3.6.1.7). For MODE 6, refer to Specification 3.9.4.

ACTION 18 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or in accordance with the Risk-Informed Completion Time Program 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 .

ACTION 19 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in accordance with the Risk-Informed Completion Time Program, and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.2.1.

ACTION 20 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 21 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in accordance with the Risk-Informed Completion Time Program 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 at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 22 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 .

ACTION 23 - With the number of OPERABLE channels less than the Total Number of Channels, declare the associated equipment inoperable and take the appropriate ACTION required in accordance with the specific equipment specification.

ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in accordance with the Risk-Informed Completion Time Program 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 ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

SHEARON HARRIS - UNIT 1 3/4 3-27 Amendment No. 186

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 25 - During CORE ALTERATIONS or movement of irradiated fuel within containment, comply with the ACTION statement of Specification 3.9.9.

ACTION 26 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.2.1.

ACTION 27 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or in accordance with the Risk-Informed Completion Time Program 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 .

SHEARON HARRIS - UNIT 1 3/4 3-27a Amendment No. 186

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE MASTER SLAVE WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

10. Engineered Safety Features Actuation System Interlocks (Continued)

c. Reactor Trip, P-4 N.A. N.A. N.A. SFCP N.A. N.A. N.A. 1, 2, 3 ##

d. Steam Generator Water See Item 5.b., above for P-14 Surveillance Requirements.

Level, P-14 SHEARON HARRIS - UNIT 1 3/4 3-48 Amendment No. 186

TABLE 4.3-2 (Continued)

TABLE NOTATION (1) Each train shall be tested at the frequency specified in the Surveillance Frequency Control Program.

(2) The Surveillance Requirements of Specification 4.9.9 apply during CORE ALTERATIONS or movement of irradiated fuel in containment.

(3) Except for relays K601, K602, K603, K608, K610, K615, K616, K617, K622, K636, K739, K740 and K741 which shall be tested at the frequency specified in the Surveillance Frequency Control Program and during each COLD SHUTDOWN exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months unless they have been tested within the previous 92 days.

(4) The Steam Line Isolation-Safety Injection (Block-Reset) switches enable the Negative Steam Line Pressure Rate--High signal (item 4.e) when used below the P-11 setpoint.

Verify proper operation of these switches each time they are used.

Setpoint verification not required.

During CORE ALTERATIONS or movement of irradiated fuel in containment.

- Trip Function automatically blocked above P-11 and may be blocked below P-11 when safety injection or low steamline pressure is not blocked.

1. The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2

Isolate Main Feedwater with coincident low Tavg - MODES 1, 2, and 3

Prevent reactuation of Safety Injection after a manual reset of Safety Injection -

MODES 1, 2, and 3

Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3 SHEARON HARRIS - UNIT 1 3/4 3-49 Amendment No. 186

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 320 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 December 3, 2020, 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-9 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 320, 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 Digitally signed by David J. David J. Wrona Date: 2021.10.07 Wrona 18:51:04 -04'00' David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 7, 2021

DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 299 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 December 3, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-17 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 299, 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 Digitally signed by David J. David J. Wrona Date: 2021.10.07 Wrona 18:51:30 -04'00' David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: October 7, 2021

ATTACHMENT TO MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 LICENSE AMENDMENT NO. 320 RENEWED FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND LICENSE AMENDMENT NO. 299 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 Insert License License NPF-9, page 3 NPF-9, page 3 NPF-17, page 3 NPF-17, page 3 TSs TSs 3.3.2-15 3.3.2-15

(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. 320, 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. 320

(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 Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or 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. 299, 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. 299

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

8. ESFAS Interlocks

a. Reactor Trip, 1,2,3(f) 1 per train, F SR 3.3.2.7 NA NA P-4 2 trains

b. Pressurizer 1,2,3 3 Q SR 3.3.2.5 < 1965 psig 1955 psig Pressure, P-11 SR 3.3.2.8

c. Tavg - Low Low, 1,2,3 1 per loop Q SR 3.3.2.5 > 551°F 553°F P-12 SR 3.3.2.8

9. Containment 1,2,3,4 4 per train, R SR 3.3.2.1 Refer to Note Refer to Note Pressure Control 2 trains SR 3.3.2.3 1 on Page 1 on page System SR 3.3.2.8 3.3.2-14 3.3.2-14 (f) The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2

Isolate MFW with coincident low Tavg - MODES 1, 2, and 3

Prevent reactuation of SI after a manual reset of SI - MODES 1, 2, and 3

Prevent opening of MFIVs if closed on SI or SG Water Level - High High - MODES 1, 2, and 3 McGuire Units 1 and 2 3.3.2-15 Amendment Nos. 320/299

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 310 AND 306 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-35 AND NPF-52 AMENDMENT NO. 186 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 AMENDMENT NOS. 320 AND 299 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-9 AND NPF-17 DUKE ENERGY CAROLINAS, LLC AND DUKE ENERGY PROGRESS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated December 3, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20338A264), Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC (Duke Energy, the licensee), submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for amendments to the Technical Specifications (TS) for the Catawba Nuclear Station, Units 1 and 2 (CNS); McGuire Nuclear Station, Units 1 and 2 (MNS); and Shearon Harris Nuclear Power Plant, Unit 1 (HNP).

The proposed amendments would revise the TS for the Engineered Safety Feature Actuation System Instrumentation (ESFAS), by adding a footnote to identify the enabled functions and the applicable MODES for the Reactor Trip, P-4 interlock function. The P-4 interlock is currently required by the TS in MODES 1 (Power Operation), 2 (Startup), and 3 (Hot Standby). The proposed amendments would remove the turbine trip function of the P-4 interlock in MODE 3 from the existing TS. In addition, only for CNS, the proposed amendments would remove the steam dump function of the P-4 interlock in MODES 1, 2, and 3.

In Attachment 2 of the LAR, the licensee submitted proposed TS Bases changes that correspond to the proposed TS changes. The proposed TS Bases changes are for information only and the licensee will revise them in accordance with each Units associated TS Bases Control Program (i.e., TS 5.5.14 for CNS and MNS, TS 6.8.4.n for HNP.)

Enclosure 6

2.0 REGULATORY EVALUATION

2.1 System Description The following system design and operation discussion is applicable to CNS, MNS, and HNP.

Differences from common design and operation are noted as applicable to one or more plants.

Based on the preselected values of the unit parameters, the ESFAS initiates necessary safety systems in order to protect core design limits and the Reactor Coolant System (RCS) pressure boundary, and to mitigate accidents. The ESFAS is segmented into three distinct but separate modules; field transmitters or process sensors and instrumentation, signal processing units, and solid state protection system (SSPS) including input, logic, and output bays.

This amendment request focuses on the SSPS segment of the ESFAS. The SSPS initiates the proper unit shutdown or engineered safety feature (ESF) actuation in accordance with the defined logic and based on the bistable outputs from the signal process control and protection system. The SSPS equipment is used for the decision logic processing of outputs from the signal processing equipment bistables. To meet the redundancy requirements, two trains of SSPS are provided to perform the same functions. If one train is taken out of service for maintenance or test purposes, the second train will provide ESF actuation for the unit. If both trains are taken out of service or placed in test, a reactor trip will result.

The SSPS performs the decision logic for most ESF equipment actuation; generates the electrical output signals that initiate the required actuation; and provides the status, permissive, and annunciator output signals to the control room. The bistable outputs from the signal processing equipment are sensed by the SSPS equipment and combined into logic matrixes that represent combinations indicative of various transients. If a required logic matrix combination is completed, the system will send actuation signals via master and slave relays to those components whose aggregate function best serves to alleviate the condition and restore the unit to a safe condition.

The required channels of ESFAS instrumentation provide unit protection in the event of analyzed design-basis accidents. However, to allow some flexibility in unit operations, several interlocks are included as part of the ESFAS architecture. These interlocks permit the operator to block some signals, automatically enable other signals, prevent some actions from occurring, and initiate other actions to occur. The interlock Functions in TS back up manual actions to ensure bypassable functions are in operation under the conditions assumed in the safety analyses. One of these interlocks, and the subject of this amendment request, is the Reactor Trip, P-4 ESFAS interlock. The P-4 interlock is enabled when a reactor trip breaker and its associated bypass breaker are open. In the LAR, the licensee stated that the functions of the P-4 interlock are the following:

Trip the main turbine; Isolate Main Feedwater (MFW) with coincident low Tavg; Prevent reactuation of Safety Injection (SI) after a manual reset of SI; Prevent opening of the MFW isolation valves if they were closed on SI or High Steam Generator Water Level signal; and For CNS, transfers the steam dump from the load rejection controller to the unit trip controller.

A plant uses a protection system to initiate a unit shutdown or an engineered safety feature actuation in accordance with the plant design. The reactor trip P-4 interlock is one of the ESFAS instrumentation functions. The P-4 interlock function initiates on a reactor trip breaker opening. The P-4 interlock is currently required by the TS in MODES 1, 2, and 3. The licensee is proposing to delete the application of P-4 interlock to trip the turbine in MODE 3 for all three sites based on the very low steam flow in MODE 3 which is used for turbine shell warming. The licensee stated in its letter dated December 3, 2020, that the steam flow is low until the turbine is placed online in MODE 1. The licensee stated that until the turbine is placed online, secondary heat removal is accomplished by a combination of steam flow through steam dumps and steam generator blowdown to the condenser.

2.2 Reason for Proposed Change As the licensee stated in its letter dated December 3, 2020:

Duke Energy is proposing changes to the ESFAS Instrumentation TS for CNS, MNS and HNP to prevent unnecessary trips of the main turbine during turbine warm up that adversely impacts startup and shutdown evolutions. In MODE 3, surveillance testing activities such as rod drop testing, utilize the closing and opening of reactor trip breakers. The block of a turbine trip on a reactor trip is only applied presently in MODES 4 or below. The proposed change would allow the block of the turbine trip on a reactor trip in MODE 3.

2.3 Description of Proposed TS Changes The licensee is proposing to modify the TS to identify the enabled functions and the applicable MODES for the Reactor Trip, P-4 interlock function by adding a new footnote to:

CNS TS 3.3.2, Table 3.3.2-1 Engineered Safety Feature Actuation System Instrumentation, Function 8.a. Reactor Trip, P-4, MNS TS 3.3.2, Table 3.3.2-1 Engineered Safety Feature Actuation System Instrumentation, Function 8.a. Reactor Trip, P-4, HNP TS 3/4.3.2, Table 3.3-3 Engineered Safety Features Actuation System Instrumentation, Functional Unit 10.c. Reactor Trip, P-4, and HNP TS 3/4.3.2, Table 4.3-2 Engineered Safety Features Actuation System Instrumentation Surveillance Requirements, Channel Functional Unit 10.c. Reactor Trip, P-4.

2.3.1 CNS Proposed TS Changes The current CNS TS 3.3.2, Table 3.3.2-1, Function 8.a. Reactor Trip, P-4 is applicable to MODES 1, 2, and 3, without exception.

The proposed change to the TS would add footnote (h) to CNS TS 3.3.2, Table 3.3.2-1, Function 8.a. Reactor Trip, P-4 as follows:

(h) The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate MFW with coincident low Tavg - MODES 1, 2 and 3 Prevent reactuation of SI after a manual reset of SI - MODES 1, 2, and 3

Prevent opening of MFIVs if closed on SI or SG Water Level - High High - MODES 1, 2, and 3 2.3.2 MNS Proposed TS Changes The current MNS TS 3.3.2, Table 3.3.2-1, Function 8.a. Reactor Trip, P-4, is applicable to MODES 1, 2, and 3, without exception.

The proposed change to the TS would add footnote (f) to MNS TS 3.3.2, Table 3.3.2-1, Function 8.a. Reactor Trip, P-4 as follows:

(f) The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate MFW with coincident low Tavg - MODES 1, 2 and 3 Prevent reactuation of SI after a manual reset of SI - MODES 1, 2, and 3 Prevent opening of MFIVs if closed on SI or SG Water Level - High High - MODES 1, 2, and 3 2.3.3 HNP Proposed TS Changes The current HNP TS 3/4.3.2, Table 3.3-3, Functional Unit 10.c. Reactor Trip, P-4, is applicable to MODES 1, 2, and 3, without exception. Also, Table 4.3-2, Channel Functional Unit 10.c.

Reactor Trip, P-4 surveillance is required for MODES 1, 2, and 3, without exception.

The proposed change to the TS would add a footnote "##" to the HNP APPLICABLE MODES portion of Functional Unit 10.c. of TS 3/4.3.2, Table 3.3-3, and to HNP MODES FOR WHICH SURVEILLANCE IS REQUIRED portion of Channel Functional Unit 10.c. of TS 3/4.3.2, Table 4.3-2. The proposed footnote identifies the enabled functions and applicable MODES for the P-4 interlock function. The proposed footnote states:

1. The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate Main Feedwater with coincident low T avg - MODES 1, 2, and 3 Prevent re-actuation of Safety Injection after a manual reset of Safety Injection - MODES 1, 2, and 3 Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3 2.4 Regulatory Requirements and Guidance The Commission's regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, Technical specifications. This regulation requires that the TSs include items in five categories: (1) safety limits, limiting safety system settings, and limiting control setting, (2) limiting conditions for operation (LCOs), (3) surveillance requirements, (4) design features, and (5) administrative controls.

The regulation in 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether an LCO is required to be included in TSs. These criteria are:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

General Design Criterion (GDC) 13, Instrumentation and Control, of Appendix A to 10 CFR Part 50, requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems, and that appropriate controls be provided to maintain these variables and systems within prescribed operating ranges.

GDC 20, Protection system functions, of Appendix A to 10 CFR Part 50, requires that the protection system be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDC 21, Protection system reliability and testability, of Appendix A to 10 CFR Part 50, requires, in part, that the protection system be designed and tested for high functional reliability.

GDC 50, Containment design basis, of Appendix A to 10 CFR Part 50 states, in part, that the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.

NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 6.2.1.2, Subcompartment Analysis (ADAMS Accession No. ML070620009), Section 6.2.1.3, Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (LOCAs) (ADAMS Accession No. ML053560191), and Section 6.2.1.4, Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures (ADAMS Accession No. ML070620010), dated March 2007, provides review guidance for the postulated effects of pipe breaks.

3.0 TECHNICAL EVALUATION

3.1 Main Turbine Trip 3.1.1 LOCA Mass & Energy GDC 50 requires that the containment structure and containment heat removal system shall be designed to accommodate the calculated pressure and temperature conditions resulting from a LOCA. The calculated containment pressure and temperature conditions are resultant from the mass and energy (M&E) releases to the containment in a design basis accident. The review guidance for the M&E releases are in SRP Sections 6.2.1.2, 6.2.1.3, and 6.2.1.4. The NRC staff reviewed the effects on the M&E releases to the containment resulting from the proposed TS changes which remove the turbine trip function of the P-4 interlock in MODE 3 from the existing ESFAS TS.

The NRC staff reviewed the LAR Sections 3.1.21a and 3.1.21b; the CNS and MNS Updated Final Safety Analysis Report (UFSAR), Chapter 6, Sections 6.2.1.3.2 (CNS Units 1 and 2, Chapter 6, Revision (Rev.) 21, Engineered Safety Features, (ADAMS Accession No. ML20106E917); MNS Units 1 and 2, Chapter 6, Rev. 22, Engineered Safety Features, (ADAMS Accession No. ML20309A729), and HNP Unit 1, UFSAR, Chapter 6, Section 6.2.1.3, Engineered Safety Features, Amendment No. 63 (ADAMS Accession No. ML20147A023),

regarding the long-term M&E release analyses for postulated LOCAs that assume the turbine trips following the reactor trip. In HNP UFSAR Chapter 6.2.1.3 LOCA M&E release analyses, initial system conditions are chosen to maximize the available energy in the system. Core power is assumed to be 102% of the rated thermal power. Steam generator parameters are based on 100% full power (FP). These conditions result in conservatively high M&E releases.

The licensee indicated that none of the conditions, models or methodology used to analyze the long-term LOCA M&E releases will become more limiting by implementing the proposed P-4 TS change.

Based on the NRC staffs review, it concludes that the current design basis long-term LOCA M&E release analyses documented in the CNS, MNS, and HNP UFSARs would remain valid, because (1) the proposed TS changes of removing MODE 3 from applicable MODES for ESFAS reactor trip P-4 interlock function to trip main turbine does not affect the M&E releases to the containment in the current design bases, and (2) the M&E releases in the current design basis LOCA are assumed to occur in MODE 1, which contains more M&E inventory for releases and are more limiting than the LOCA to occur in MODE 3.

Further, the NRC staff reviewed Sections 3.1.21a and 3.1.21b of the LAR, CNS and MNS UFSAR Sections 6.2.1.3.1, and HNP UFSAR Section 6.2.1.3 regarding short-term M&E release analyses for postulated LOCAs. The results of the UFSAR Section 6.2.1.3.1.1 analyses are used as input to the subcompartment analyses which include a short pressure pulse in lower containment (less than 5 seconds) that accompanies a LOCA.

Based on the staffs review, the NRC staff found that due to the short duration of the postulated release, the P-4 interlock turbine trip function has no impact on the short-term M&E release analyses. Therefore, the NRC staff concludes that the current design basis short-term LOCA M&E releases analyses documented in the CNS, MNS, and HNP UFSARs would remain valid.

3.1.1.1 Main Steam Line Break Inside Containment The NRC staff reviewed Sections 3.1.22a and 3.1.22b of the LAR, CNS and MNS UFSAR, Chapter 6, Sections 6.2.1.4, and HNP UFSAR, Chapter 6, Section 6.2.1.4 regarding the main steam line break (MSLB) M&E release analysis for breaks inside containment, which assumes that the turbine trip is coincident with reactor trip resulting from a SI signal due to high containment pressure. The licensee stated that all the initial and boundary conditions maximize the steam release out the break, rather than to the turbine, and that the turbine trip on reactor trip is not credited in the analysis as a mitigating function. The licensee also stated that the closure of the main steam isolation valves occurs on high-high containment pressure closely following the reactor trip, after which a turbine trip would have no impact on containment response.

Based on the staffs review, the NRC staff found that the P-4 interlock function has no impact on the MSLB M&E release analysis and that the current design basis M&E release analyses documented in the CNS and MNS UFSAR Sections 6.2.1.4 and HNP UFSAR Section 6.2.1.4 would remain valid, because the turbine trip on reactor trip is not credited in the MSLB M&E release calculation and the steam line isolation occurs following the reactor trip such that turbine trip would have no impact on M&E releases nor on containment pressure and temperature responses.

In summary, based on the NRC staffs review described above in 3.1.1 and 3.1.1.1, the staff concludes that the current licensing basis LOCA and MSLB analyses relating to M&E releases, containment pressure and temperature limits, and GDC 50 would remain unchanged by these amendments.

3.1.1.2 Steam Flow The NRC staff reviewed CNS Units 1 and 2 UFSAR, Chapter 7, Rev. 21, Instrumentation and Controls, (ADAMS Accession No. ML20106E920); MNS Units 1 and 2, Chapter 7, Rev. 22, Instrumentation and Controls, (ADAMS Accession No. ML20309A756); and HNP Unit 1, Amendment 63, Chapter 7, Instrumentation and Controls, (ADAMS Accession No. ML20147A024). Each USFAR, Section 7.2.1.1.1, Functional Performance Requirements, describes the Reactor Protection System initiating a turbine trip signal whenever a reactor trip occurs in order to prevent insertion of positive reactivity, which could result from an overcooling of the RCS.

The NRC staff noted that turbine trip interlock P-4 is required in MODE 1, when the main turbine is in service. The licensee does not propose to delete the turbine trip when interlock P-4 is active in MODES 1 and 2. In the LAR, the licensee stated that in MODE 3 the main turbine is on a turning gear and not in service, steam flow is aligned to steam dumps, and steam flow entering the turbine for the purpose of warming is considered negligible. Therefore, the licensee is proposing that the applicability of interlock P-4 MODE 3 can be excluded from the TS.

The NRC staff reviewed the licensee's proposal to exclude the turbine trip to the plant trip controller as a function from being required in MODE 3 as part of the P-4 interlock. The staff agrees that the steam flow entering the turbine for warming purposes during MODE 3 can be considered negligible. Therefore, the NRC staff concludes that the licensee's proposal to exclude the turbine trip P-4 interlock function from CNS, MNS, and HNP during MODE 3 is acceptable.

3.1.1.3 Events that are Limiting in MODE 1 The NRC staff reviewed the CNS Units 1 and 2 UFSAR, Chapter 15, Rev. 21, Accident Analysis, (ADAMS Accession No. ML20106E946); MNS Units 1 and 2 Chapter 15, Rev. 22, Accident Analysis, (ADAMS Accession No. ML20309A752); and HNP Unit 1, Chapter 15, Amendment No. 63, Accident Analysis, (ADAMS Accession No. ML20147A032) to determine if the removal of the P-4 interlock in MODE 3 would cause the event to be limiting in MODE 3.

For events that are currently limiting from MODE 1 initial conditions, the NRC staffs review determined that the removal of the interlock in MODE 3 would not cause the event to be limiting from MODE 3 initial conditions. As shown in Table 1 of this safety evaluation (SE), the effective multiplication factor (Keff) is more subcritical in MODE 3 compared to MODE 1 and there is no thermal power produced in MODE 3 while in MODE 1 the power is greater than 5%.

Additionally, for events that are limiting in MODE 1 because of reactivity increases, these initial conditions are maximized to result in conservatively high reactivity increases. These differences in MODE 1 are substantial and since the steam flow through the turbine is negligible in MODE 3, removing the turbine trip function would also have a negligible impact on the RCS cooldown and subsequent reactivity increases. Therefore, the NRC staff concludes that all events that are currently limiting in MODE 1 would remain limiting with the removal of the P-4 interlock function in MODE 3 and GDC 21 regarding systems being designed and tested for high functional reliability remains applicable.

Table 1: MODES of Operation for CNS, MNS, and HNP MODE Reactivity Condition, Keff Rated Thermal Power

1. Power Operation 0.99 > 5%

2. Startup 0.99 5%

3. Hot Standby < 0.99 0%

3.1.1.4 Events that result in an Increase in Heat Removal by the Secondary System For events that were not limiting in MODE 1, the NRC staff reviewed each event group to determine if the removal of the P-4 interlock in MODE 3 would cause the event to be limiting in MODE 3. The licensee discusses each event that falls into this category in Sections 3.1.1 through 3.1.4 of its LAR.

As described in Section 3.1.1 of the LAR, the excessive feedwater flow event is analyzed, in part, at MODE 2. The licensee credited a larger shutdown margin in MODE 3 compared to MODE 2 to determine that the event would remain limiting in MODE 2. The NRC staff confirmed this by comparison of the effective multiplication factor between the two MODES. As seen in Table 1 of this SE, more positive reactivity would be needed in the MODE 3 event for it to become limiting compared to MODE 2. Since the steam flow through the turbine is very low in MODE 3, the removal of the P-4 interlock would not cause a significant enough increase in reactivity to become limiting relative to the event in MODE 2. Therefore, the NRC staff finds that the event in MODE 2 would remain limiting.

Section 3.1.2 of the LAR describes the excessive load increases analyses. This event is limiting in MODE 1; therefore, the event will remain limiting in MODE 1 per the justification in the Events that are Limiting in MODE 1 section of this SE.

Section 3.1.3 of the LAR describes the steam releases because of an inadvertent opening of a steam generator relief or safety valve. These events are bounded by steam system breaks which are discussed in LAR Section 3.1.4.

Section 3.1.4 of the LAR describes steam release caused by a rupture of a main steam line.

These events are analyzed, in part, in MODE 2 and some consideration is given to the event in MODE 3. In MODE 3, there is a potential that the SI actuation signals are blocked, preventing borated water from entering the core during the event which is a concern relative to recriticality.

However, the licensee currently administratively controls boron in the RCS when in MODE 3 or when SI is blocked to prevent a recriticality event. When analyzed in MODE 2, to conservatively maximize reactivity, the turbine trip function is not modeled which aligns it with the removal of the P-4 interlock in MODE 3. Since the turbine trip is not modeled in MODE 2 and the licensee does not intend to change the administrative controls on boron in the RCS in MODE 3 or when SI is blocked, the NRC staff finds that the event will remain bounding in MODE 2 with the removal of the P-4 interlock in MODE 3.

3.1.1.5 Events that Result in a Decrease in Heat Removal by the Secondary System The licensee discusses each event that falls into this category in Section 3.1.5 of the LAR.

These events consider the consequences of decreased heat removal in the secondary system and are analyzed to protect RCS integrity and are driven by power mismatches between primary and secondary sides of the RCS. All of these events are limiting in MODE 1.

For the loss of external electrical load, turbine trip, inadvertent closure of the main steam isolation valve, and loss of condenser vacuum the action that limits these events is the turbine trip from MODE 1. Therefore, the NRC staff finds that the event would remain bounding in MODE 1 with the removal of the P-4 interlock in MODE 3.

For the loss of offsite power, loss of normal feedwater events, and feedwater system pipe breaks, an early turbine trip is more limiting since it reduces the heat removal from the secondary side and increases the power mismatch. Therefore, the NRC staff finds that the event would remain bounding in MODE 1 with the removal of the P-4 interlock in MODE 3 and hence, acceptable .

3.1.1.6 Events that Result in a Decrease in Reactor Coolant Flow The licensee discusses each event that falls into this category in Section 3.1.6 of the LAR.

Each of these events consider reductions in RCS flow and are analyzed to protect against departure from nucleate boiling ratio (DNBR) and are, in part, driven by high core power. These events are limiting in MODE 1. Therefore, the NRC staff concludes that these events will remain limiting in MODE 1 per the justification in Section 3.1.1.3 of this SE.

3.1.1.7 Events Caused by the Addition of Positive Reactivity to the RCS The licensee discusses each event that falls into this category in Section 3.1.7 through 3.1.13 of the LAR.

Section 3.1.7 of the LAR describes the uncontrolled rod cluster control assembly (RCCA) bank withdrawal from a subcritical or lower power startup condition. These events are analyzed at MODE 2 conditions but do not currently credit the turbine trip during the event. The NRC staff concludes that the event would still be limiting in MODE 2 with the removal of the P-4 interlock in MODE 3 because the turbine trip is not credited in MODE 2 but the core is critical (i.e., Keff 0.99) which will increase the consequences of the event.

Section 3.1.8 of the LAR describes the RCCA bank withdrawal at power (i.e., MODE 1).

Because by definition this event is analyzed at power, the NRC staff concludes that the removal of the P-4 interlock in MODE 3 will not impact this event.

Section 3.1.9 of the LAR describes events as a result of RCCA misoperation. For RCCAs that are dropped or misaligned, the consequences of these events are DNBR. These events are limiting in MODE 1; therefore, the event will remain limiting in MODE 1 per the justification in the Events that are Limiting in MODE 1 section of this SE. For a withdrawal of a single RCCA event, the analysis does not credit the turbine trip P-4 interlock function to mitigate the event.

Thus, the proposed change to remove the turbine trip P-4 interlock function in MODE 3 does not affect the RCCA misoperation events. Therefore, the NRC staff concludes that that the event in the UFSARs would still be limiting with the removal of the P-4 interlock in MODE 3.

Section 3.1.10 of the LAR describes the startup of an inactive reactor coolant pump at an incorrect temperature. The analysis does not credit the turbine trip to mitigate the event.

Therefore, the NRC staff concludes that the events stated in Chapter 15 of the UFSARs would still be limiting with the removal of the P-4 interlock in MODE 3 since its removal will not impact any of the assumptions in the current analyses.

Section 3.1.11 of the LAR describes the chemical and volume control system (CVCS) malfunctions that result in a decrease in the boron concentration in the reactor coolant. This event is analyzed in MODES 1 through 5, however, the analyses do not credit the P-4 interlock turbine trip on the reactor trip. Therefore, the NRC staff concludes that the event in the UFSARs would still be limiting with the removal of the P-4 interlock in MODE 3 since its removal will not impact any of the assumptions in the current analyses.

Section 3.1.12 of the LAR describes the inadvertent loading and operation of a fuel assembly in an improper position. The analyses do not credit the P-4 interlock turbine trip on the reactor trip.

Therefore, the NRC staff concludes that the event in the UFSARs would still be limiting with the removal of the P-4 interlock in MODE 3 since its removal will not impact any of the assumptions in the current analyses.

Section 3.1.13 of the LAR describes the spectrum of RCCA ejection accidents. The analyses do not credit the P-4 interlock turbine trip on the reactor trip. Therefore, the NRC staff concludes that the event in the UFSARs would still be limiting with the removal of the P-4 interlock in MODE 3, since its removal will not impact any of the assumptions in the current analyses.

3.1.1.8 Events Caused by the Addition of Inventory to the RCS The licensee discusses each event that falls into this category in Sections 3.1.14 and 3.1.15 of the LAR.

Section 3.1.14 of the LAR discusses the inadvertent emergency core cooling system (ECCS) actuation during power operation that could overfill the pressurizer and damage the RCS.

Because the definition of this event indicates that it is analyzed at power (i.e., MODE 1), the NRC staff concludes that removal of the P-4 interlock in MODE 3 will not impact this event.

Section 3.1.15 of the LAR discusses CVCS malfunctions that increase reactor coolant inventory.

The consequences of this event are bounded by the CVCS malfunction that decreases boron concentration in the RCS and the inadvertent ECCS actuation during power operation that could

overfill the pressurizer. The NRC staff concludes that the event will remain bounding in MODE 2 with the removal of the P-4 interlock in MODE 3.

3.1.1.9 Events that Result in a Decrease of RCS Inventory The licensee discusses each event that falls into this category in Sections 3.1.16 through 3.1.19 of the LAR.

Each event discussed in Sections 3.1.16 through 3.1.18 of the LAR are events that are limiting in MODE 1. Therefore, the NRC staff concludes that removal of the interlock in MODE 3 would not cause the events to be limiting from MODE 3 initial conditions.

Section 3.1.19 of the LAR describes the large and small break LOCA analyses. Since the LOCA analyses do not credit the turbine trip functions of the P-4 interlock, the NRC staff finds that the removal of the turbine trip function of the P-4 interlock in MODE 3 will not impact this event.

3.1.1.10 Anticipated Transient Without Scram (ATWS)

The licensee discusses the ATWS event in Section 3.1.20 of the LAR. This event is only postulated in MODE 1. Therefore, the NRC staff finds that the removal of the P-4 interlock in MODE 3 will not impact this event.

3.1.2 Steam Dump Control (CNS Only)

In the LAR, the licensee states that the steam dump control system function listed in the current CNS TS Bases 3.3.2 is not required or credited in any plant safety analysis. In reviewing the criterion for inclusion into the TSs, the licensee states the instrumentation utilized to initiate transfer to the plant trip steam dump controller does not serve any of the functions delineated under Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The licensee states in its LAR that one of the functions of the P-4 interlock in the CNS TS is to transfer the steam dump from the load rejection controller to the unit trip controller. The licensee states that MNS and HNP steam controllers perform the same function, using different signals, without a notation in TS and that CNS Chapter 7, UFSAR Section 7.7, Control Systems Not Required for Safety, identifies Steam Dump Control as a plant control system not required for safety.

The licensee states in its LAR, that The instrumentation utilized to initiate transfer to the unit trip steam dump controller does not serve a primary protective function to warrant inclusion in the TS. The instrumentation does not serve to ensure that the plant is operated within the bounds of initial conditions assumed in design basis accident and transient analyses. Likewise, the transfer to the unit trip steam dump controller instrumentation does not serve as part of the primary success path of a safety analysis used to demonstrate that the consequences of these events are within the appropriate acceptance criteria.

The NRC staff reviewed the licensee's proposal to exclude the transfer of the steam dumps to the plant trip controller as a function from being required in MODE 3 as part of the P-4 interlock.

The NRC staff finds that the transfer of steam dumps to the plant controller upon a reactor trip is

not a safety function credited to mitigate a design basis event in the plants licensing basis.

Therefore, the staff concludes that the licensee's proposal to exclude the steam dump transfer function from CNS TS Table 3.3.2-1, Function 8.a, based on not meeting the criteria outlined in 10 CFR 50.36 (c)(2)(ii), is acceptable. The NRC staff notes that the P-4 interlock to trip the turbine in MODES 1 and 2 will continue to be in place.

The NRC staff also finds that the transfer of the steam dump from the load rejection controller to the unit trip controller of the P-4 interlock for CNS was not a primary success path to mitigate a design basis event and was not used to maintain the initial conditions for a design basis event and is acceptable to remove from the technical specifications. The staff concludes that all the current design basis analyses would not be impacted by the removal of this requirement.

Therefore, the NRC staff finds that the steam dump function of P-4 does not warrant inclusion in the TS and it can be excluded from CNS TS 3.3.2, Table 3.3.2-1, Function 8.a. Additionally, the staff finds that the removal of this requirement will not invalidate GDC 20, Protection system functions for CNS because this function does not meet the criterion of a protection system function 1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

3.1.3 Compliance with the 10 CFR 50.36(c)(2)(ii) Requirements The proposed addition of footnotes to the TSs would remove out of the TSs the following functions (referred to as the subject functions):

(1) The P-4 interlock requirement during MODE 3 for CNS, MNS, and HNP, and (2) The steam dump from the load rejection controller to the unit trip controller of the P-4 interlock for CNS in MODES 1, 2, and 3.

The NRC staff reviewed the proposed changes to the TS against the 10 CFR 50.36(c)(2)(ii) criteria and determined as follows.

(A) Criterion 1. The subject functions are not used to detect and indicate a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. The subject functions are not a process variable, design feature, or operating restriction that was an initial condition of a design basis accident or transient analysis.

(C) Criterion 3. No credit is taken for the subject functions in the analysis of transients and accidents for CNS, MNS, and HNP. The "subject functions" are not considered as part of primary success path related to the integrity of a fission product barrier. Therefore, the subject functions are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. The subject functions are not significant to public health and safety in that no credit is taken for the subject functions for consequence mitigation in applicable design basis accident or transient analysis. Therefore, the subject functions in

applicable MODES are not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Based on the above, the NRC staff finds that the existing LCO and related surveillance requirement associated with the subject functions do not satisfy any of the criteria in 10 CFR 50.36(c)(2)(ii). Additionally, the NRC staff finds that the proposed TS footnotes provide clarification of the applicability of the P-4 interlock requirements. Therefore, the NRC staff concludes that the proposed removal of the subject functions out of the TSs does not violate the 10 CFR 50.36(c)(2)(ii) requirements and is acceptable.

3.1.4 Evaluation of GDCs 13, 20, and 21 There are no instrument and control design changes associated with this LAR. The proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 out of the CNS, MNS and HNP TS is consistent with the CNS, MNS, and HNP designs and analyses and will continue to ensure proper actuation to satisfy the anticipatory trip function as evaluated by the NRC staff above. As such, there will be no degradation in the performance of, nor an increase in the number of challenges to, equipment assumed to function during an accident situation.

There will be no change to normal plant operating parameters or accident mitigation capabilities.

Based on the above, the NRC staff concludes that there is reasonable assurance that the requirement of GDC 13, regarding the need for appropriate controls to be provided for maintaining variables and systems can be operated within prescribed operating ranges, has been met.

Based on the above, the NRC staff concludes that there is reasonable assurance that the requirement of GDC 20, regarding the need for that the protection system be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety, has been met.

Based on the above, the NRC staff concludes that there is reasonable assurance that the requirement of GDC 21, regarding the need for the protection system to be designed for high functional reliability and inservice testability appropriate controls to be provided for maintaining variables and systems can be operated within prescribed operating ranges, has been met.

3.1.5 Technical Conclusion The proposed changes are acceptable to the NRC staff as evaluated above. Transfer of steam dump control from the load rejection controller to the unit trip controller is currently applicable to CNS units 1 and 2, and is not used for MNS and HNP. The staff reviewed the removal of the steam dump from the load rejection controller to the unit trip controller of the P-4 interlock for CNS in Modes 1, 2, and 3. The specific instrumentation requirements related to transfer to the unit trip steam dump controller are not required to be in the CNS TS based on the criteria in 10 CFR 50.36, GDC 13, GDC 20, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. This function is not credited for in the plant accident analysis and the applicability of GDC 21 continue to be effective without any changes. Based on above, the NRC staff finds the proposed TS changes to remove the turbine trip function of the P-4 interlock in MODE 3 from the existing ESFAS TS do not affect the M&E releases in the containment for the containment

pressure and temperature limits in the current licensing bases for containment analyses, which are consistent with SRP Sections 6.2.1.2, 6.2.1.3, and 6.2.1.4. Additionally, GDC 50 will continue to be satisfied. Therefore, the NRC staff finds that the revised TS would continue to meet the requirements of 10 CFR 50.36(c)(2)(ii) and that the revised TS would not impact the licensees compliance with GDC 13, 20, 21, and 50. Therefore, the staff finds the proposed TS changes acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina and South Carolina State officials were notified of the proposed issuance of the amendments on August 12, 2021 and August 24, 2021, respectively. On August 12, 2021, and August 24, 2021, respectively the State officials confirmed that the State of North Carolina and the State of South Carolina 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. 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 published in the Federal Register on March 23, 2021(86 FR 15501), and there has been no public comment on such finding. 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.

Principal Contributors: J. Ashcraft, NRR J. Borromeo, NRR C. Li, NRR M. Razzaque, NRR G. Singh, NRR Date: October 7, 2021

ML21224A101 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-2/LA NRR/DSS/SCPB/BC NAME SDevlin-Gill JKlos RButler BWittick DATE 08/03/2021 08/23/2021 08/17/2021 08/27/2021 OFFICE NRR/DSS/SNSB/BC NRR/DEX/EICB/BC NRR/DSS/STSB/(A) BC OGC - NLO NAME SKrepel MWaters NJordan JAzeizat DATE 06/17/2021 06/07/2021 09/8/2021 10/05/2021 OFFICE NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME DWrona THood DATE 10/06/2021 10/07/2021

v t e Letter Sequence Approval
EPID:L-2020-LLA-0262, TSTF-493 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other: ML22012A419
MONTHYEAR2021-01-06 [Table View]

ML21224A101

Text

SUBJECT:

Dear Mr. Snider:

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

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