ML21320A001

Issuance of Amendment No. 189 to Regarding Administrative Change to Reflect Development of a Technical Requirements Manual

ML21320A001
Person / Time
Site:	Harris
Issue date:	01/20/2022
From:	Michael Mahoney Plant Licensing Branch II
To:	Maza K Duke Energy Progress
Mahoney M
References
EPID L 2021 LLA 0108
Download: ML21320A001 (28)

Text

January 20, 2022 Ms. Kim Maza Site Vice President Shearon Harris Nuclear Power Plant Mail Code NHP01 5413 Shearon Harris Road New Hill, NC 27562-9300

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 189 REGARDING ADMINISTRATIVE CHANGE TO REFLECT DEVELOPMENT OF A TECHNICAL REQUIREMENTS MANUAL (EPID L-2021-LLA-0108)

Dear Ms. Maza:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 189 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). This amendment is in response to your application dated June 7, 2021.

The amendment revises the Technical Specifications (TS) to reflect the transition of the licensee-controlled plant procedure PLP-106, Technical Specification Equipment List Program, to a licensee-controlled Technical Requirements Manual (TRM). The proposed change is an administrative change with no impact on technical content.

K. Maza A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions regular monthly Federal Register notice.

Sincerely,

/RA/

Michael Mahoney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. Amendment No. 189 to NPF-63

2. Safety Evaluation cc: Listserv

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 189 Renewed License No. NPF-63

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

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

dated June 7, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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-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. 189, 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 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by David J. David J. Wrona Date: 2022.01.20 Wrona 10:21:01 -05'00' David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility License No. NPF-63 and Technical Specifications Date of Issuance: January 20, 2022

ATTACHMENT TO LICENSE AMENDMENT NO. 189 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 marginal line indicating the area of change:

Remove Insert Page 4 Page 4 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 2-10 2-10 3/4 3-1 3/4 3-1 3/4 3-9 3/4 3-9 3/4 3-17 3/4 3-17 3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 4-34 3/4 4-34 3/4 4-35 3/4 4-35 3/4 4-36 3/4 4-36 3/4 4-37 3/4 4-37 3/4 4-41 3/4 4-41 3/4 6-14 3/4 6-14 3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 3/4 7-19 3/4 7-19 3/4 7-24 3/4 7-24 3/4 8-19 3/4 8-19

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. 189, 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. 189

TABLE 2.2-1 (Continued)

TABLE NOTATIONS The values denoted with [*] are specified in the COLR.

NOTE 3: (Continued)

K6 = [*]/°F for T > T" and K6 = [*] for T T",

T = As defined in Note 1, T" = Reference Tavg at RATED THERMAL POWER ( [*]°F),

S = As defined in Note 1, and f2(I) = [*].

NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% of T span for T input, 1.35% of Tavg span for Tavg input; and 0.6% of I span for I input.

NOTE 5: The sensor error is: 1.3% of T span for T/Tavg temperature measurements; and 0.8% of T span for pressurizer pressure measurements.

NOTE 6: The sensor error (in % span of Steam Flow) is: 1.1% for steam flow; 1.8% for feedwater flow; and 2.4% for steam pressure.

NOTE 7: If the as-found channel setpoint is outside its predefined as-found tolerance, the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

NOTE 8: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpoint in Table 2.2-1 (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 NTSPs and the as-found and the as-left tolerances are specified in EGR-NGGC-0153, Engineering Instrument Setpoints. The as-found and as-left tolerances are specified in the Technical Requirements Manual.

SHEARON HARRIS - UNIT 1 2-10 Amendment No. 175, 189

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION: As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channels and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit, specified in the Technical Requirements Manual, at the frequency specified in the Surveillance Frequency Control Program.

SHEARON HARRIS - UNIT 1 3/4 3-1 Amendment No. 154, 189

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES This table is deleted from Technical Specifications.

The information in this table is controlled by the Technical Requirements Manual.

PAGE 3/4 3-10 HAS BEEN DELETED.

SHEARON HARRIS - UNIT 1 3/4 3-9 Amendment No. 25, 189

INSTRUMENTATION ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within its limit specified in the Technical Requirements Manual at the frequency specified in the Surveillance Frequency Control Program.

SHEARON HARRIS - UNIT 1 3/4 3-17 Amendment No. 154, 189

TABLE 3.3-4 (Continued)

TABLE NOTATIONS

• Time constants utilized in the lead-lag controller for Steam Line Pressure--Low are 50 seconds and 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

• • The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate--High is 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.

1. The indicated values are the effective, cumulative, rate-compensated pressure drops as seen by the comparator.

NOTE 1: If the as-found channel setpoint is outside its predefined as-found tolerance, the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

NOTE 2: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpoint in Table 3.3-4 (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 NTSPs and the as-found and the as-left tolerances are specified in EGR-NGGC-0153, Engineering Instrument Setpoints.

The as-found and as-left tolerances are specified in the Technical Requirements Manual.

SHEARON HARRIS - UNIT 1 3/4 3-36 Amendment No. 161, 189

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES This table is deleted from Technical Specifications.

The information in this table is controlled by the Technical Requirements Manual.

PAGES 3/4 3-38 THROUGH 3/4-40 HAVE BEEN DELETED.

SHEARON HARRIS - UNIT 1 3/4 3-37 Amendment No. 25, 189

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.2 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, and inservice leak and hydrostatic testing with:

a. A maximum heatup rate as shown on Table 4.4-6.

b. A maximum cooldown rate as shown on Table 4.4-6.

c. A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: MODES 4, 5, and 6 with reactor vessel head on.

ACTION:

With any of the pressure limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; if the pressure and temperature limit lines shown on Figure 3.4-2 and 3.4-3 were exceeded, perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or maintain the RCS Tavg and pressure at less than 200°F and 500 psig, respectively.

SURVEILLANCE REQUIREMENTS 4.4.9.2.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at the frequency specified in the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.2.2 Deleted from Technical Specifications. Refer to the Technical Requirements Manual.

SHEARON HARRIS - UNIT 1 3/4 4-34 Amendment No. 154, 189

FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 55 EFPY SHEARON HARRIS - UNIT 1 3/4 4-35 Amendment No. 183, 189

FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 55 EFPY SHEARON HARRIS - UNIT 1 3/4 4-36 Amendment No. 183, 189

TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM This table is deleted from Technical Specifications.

The information in this table is controlled by the Technical Requirements Manual.

SHEARON HARRIS - UNIT 1 3/4 4-37 Amendment No. 25, 189

500 LOW PORV HIGH PORV PORV SETPOINT (PSIG) 400 300 0 100 200 300 400 MEASURED RCS TEMPERATURE (°F)

RCS TEMP (°F) LOW PORV* (psig) HIGH PORV* (psig) 90 400 410 250 400 410 325 440 450

• VALUES BASED ON 55 EFPY REACTOR VESSEL DATA INSTRUMENT ERRORS ARE CONTROLLED BY THE TECHNICAL REQUIREMENTS MANUAL.

FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM SHEARON HARRIS - UNIT 1 3/4 4-41 Amendment No. 183, 189

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve specified in the Technical Requirements Manual shall be OPERABLE with isolation times less than or equal to required isolation times.

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

ACTION:

With one or more of the containment isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk-Informed Completion Time Program, or

b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk-Informed Completion Time Program by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk-Informed Completion Time Program by use of at least one closed manual valve or blind flange, or

d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.3.1 Each isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.

SHEARON HARRIS - UNIT 1 3/4 6-14 Amendment No. 184, 189

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 Requirements Manual when tested pursuant to the INSERVICE TESTING PROGRAM.

SHEARON HARRIS - UNIT 1 3/4 6-15 Amendment No. 166, 189

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES This table is deleted from Technical Specifications.

The information in this table is controlled by the Technical Requirements Manual.

PAGES 3/4 6-17 THROUGH 3/4 6-29 HAVE BEEN DELETED.

SHEARON HARRIS - UNIT 1 3/4 6-16 Amendment No. 25, 189

PLANT SYSTEMS 3/4.7.8 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.8 All snubbers shall be OPERABLE. The only snubbers excluded from the requirements are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.

ACTION:

With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per augmented inservice inspection program on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 4.7.8 Each snubber shall be demonstrated OPERABLE by performance of the augmented inservice inspection program specified in the Technical Requirements Manual.

PAGES 3/4 7-20 THROUGH 3/4 7-23 HAVE BEEN DELETED.

SHEARON HARRIS - UNIT 1 3/4 7-19 Amendment No. 25, 189

FIGURE 4.7-1 SAMPLE PLAN (2) FOR SNUBBER FUNCTIONAL TEST This figure is deleted from Technical Specifications and is controlled by the Technical Requirements Manual.

SHEARON HARRIS - UNIT 1 3/4 7-24 Amendment No. 25, 189

ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Specifications 3/4.8.4.1 and 3/4.8.4.2 have been deleted from Technical Specifications and relocated to the Technical Requirements Manual.

PAGES 3/4 8-20 THROUGH 3/4 8-43 HAVE BEEN DELETED.

Pages 3/4 8-20, 3/4 8-21, 3/4 8-39, and 3/4 8-40 by Amendment No. 182.

Pages 3/4 8-22 through 3/4 8-38B and 3/4 8-41 through 3/4 8-43 by Amendment No. 13.

SHEARON HARRIS - UNIT 1 3/4 8-19 Amendment No. 182, 189

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 189 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By application dated June 7, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21158A131), Duke Energy Progress, LLC (the licensee), requested changes to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris). The license amendment request (LAR) proposes to revise the TS to reflect the transition of the licensee-controlled plant procedure PLP-106, Technical Specification Equipment List Program, to a licensee-controlled Technical Requirements Manual (TRM). The proposed change is an administrative change with no impact on technical content.

2.0 REGULATORY EVALUATION

2.1 Background

The Harris TS are based upon the format and content of the NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors," series. As a result, the Harris TS numbers and associated Bases numbers differ from those contained in the improved Standard Technical Specifications for Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5, dated September 2021 (ADAMS Accession Nos. ML21259A155 and ML21259A159, respectively).

The licensee has not converted the Harris TS to the Improved Standard TS structure and has continued to maintain a licensee-controlled document in the form of plant procedure PLP-106, Technical Specification Equipment List Program, to capture certain relocated TS content including equipment lists, figures, and surveillance programs. PLP-106 is a document incorporated by reference into the Harris Final Safety Analysis Report (FSAR) and is subject to the update and reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.71(e) and change controls of 10 CFR 50.59.

Enclosure 2

2.2 Description of Changes For the following Harris TS, the licensee proposes to replace all references to plant procedure PLP-106 with references to the TRM:

TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, Note 8 TS 3/4.3.1, Reactor Trip System Instrumentation TS Table 3.3-2, Reactor Trip System Instrumentation Response Times TS 3/4.3.2, Engineered Safety Features Actuation System Instrumentation TS Table 3.3-4, Engineered Safety Features Actuation System Instrumentation Trip Setpoints, Note 2 TS Table 3.3-5, Engineered Safety Features Response Times TS 3/4.4.9, Pressure/Temperature Limits TS Figure 3.4-2, Reactor Coolant System Cooldown Limitations - Applicable Up to 55 EFPY TS Figure 3.4-3, Reactor Coolant System Heatup Limitations - Applicable Up to 55 EFPY TS Table 4.4-5, Reactor Vessel Material Surveillance Program TS Figure 3.4-4, Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System TS 3/4.6.3, Containment Isolation Valves TS Table 3.6-1, Containment Isolation Valves TS 3/4.7.8, Snubbers TS Figure 4.7-1, Sample Plan (2) for Snubber Functional Test TS 3/4.8.4, Electrical Equipment Protective Devices 2.3 Applicable Regulatory Requirements and Guidance Regulations In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) Surveillance Requirements (SRs); (4) design features; and (5) administrative controls.

Guidance NRC Administrative Letter (AL) 96-04 (ADAMS Accession No. ML031110087) states An acceptable approach that several licensees have used is to incorporate the details of the relocated technical specification requirements into a manual, and then reference the manual in the FSAR.

3.0 TECHNICAL EVALUATION

Based on the NRC's Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, 58 FR 39132, July 22, 1993, certain requirements may be relocated from the TS to other licensee-controlled documents (such as a TRM). The requirements in a

TRM are considered as part of the licensing basis (as a part of the Updated Final Safety Analysis Report (UFSAR)) and are to be treated as such.

Similar to a TRM, the licensee previously developed Technical Specification Equipment List Program, plant procedure PLP-106, in an effort to centralize the requirements relocated from the TS and to ensure the necessary administrative controls are applied to these requirements.

The proposed change to reference a TRM in place of plant procedure PLP-106 does not adversely alter the current TS or introduce any new TS requirements. Additionally, there is no impact on plant operations or systems as a result of the proposed change.

The relocated TS content currently captured in PLP-106 will be maintained in the TRM.

Further, the licensee stated, As a TRM, the content previously contained in PLP-106 will continue to be incorporated by reference in the FSAR and subject to the update and reporting requirements of 10 CFR 50.71(e), with changes processed in accordance with 10 CFR 50.59.

Finally, when implemented, the licensee stated that the change will provide continuity and consistency throughout the Duke Energy fleet in the processing of changes to relocated TS content in licensee-controlled documents.

3.1 NRC Staff Conclusion

The NRC staff finds that the proposed change from a licensee-controlled plant procedure PLP-106 to a licensee-controlled TRM is an administrative change that does not substantively change TS requirements. As such, the NRC staff concludes the regulatory requirements of 10 CFR 50.36 continue to be met, and therefore, the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on December 2, 2021. The State of North Carolina official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20 and changes administrative requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration (86 FR 55012, dated October 5, 2021), 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) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to 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: M. Mahoney, NRR C. Ashley, NRR Date of Issuance: January 20, 2022

ML21320A001 OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DSS/STSB/BC NAME MMahoney RButler VCusumano DATE 12/01/2021 11/23/2021 12/01/2021 OFFICE OGC NLO DORL/LPL2-2/BC DORL/LPL2-2/PM NAME JAzeizat DWrona MMahoney DATE 12/13/2021 01/20/2022 01/20/2022