RA-23-0005, License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specifications

License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specifications

ML23151A724
Person / Time
Site:	Harris
Issue date:	05/31/2023
From:	Haaf T Duke Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-23-0005
Download: ML23151A724 (1)

Text

Thomas P. Haaf Site Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill, NC 27562-9300 984-229-2512 10 CFR 50.90 May 31, 2023 Serial: RA-23-0005 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Renewed License No. NPF-63

Subject:

License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specifications Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy),

hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment to the HNP TS will modify the HNP TS Surveillance Requirement (SR) 4.6.1.1 to eliminate the requirement to perform periodic position verification for containment penetrations that are maintained locked, sealed, or otherwise secured closed, as well as adopt TS Task Force (TSTF) Improved Standard TS (ISTS) Change Traveler No. 45 (TSTF-45-A), Revision 2, Exempt Verification of Containment Isolation Valves that are Not Locked, Sealed, or Otherwise Secured (ADAMS Accession No. ML040400137).

The proposed amendment will also revise HNP TS 3.3.3.5, Remote Shutdown System, to increase the completion time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance and remove the requirement to submit a Special Report. It will also relocate the content in Table 3.3-9, Remote Shutdown System, and Table 4.3-6, Remote Shutdown Monitoring Instrumentation Surveillance Requirements, in accordance with TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (ADAMS Accession No. ML040620072). Additionally, the proposed amendment will update HNP SR 4.3.1.1, Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirements, to address the application of the Surveillance Frequency Control Program (SFCP) to establish the Frequency for performance of the Analog Channel Operational Test (ACOT) of select Reactor Trip System (RTS) instrumentation.

Changes are also proposed to the Administrative Controls Section of the HNP TS to reflect current organizational titles as well as remove reporting requirements that are redundant to existing regulations. The proposed changes above reflect requirements consistent with those in Revision 5 of NUREG-1431, Standard Technical Specifications - Westinghouse Plants (ADAMS Accession No. ML21259A155).

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been concluded that the proposed changes involve no significant hazards consideration.

( ~ DUKE ENERGY

U.S. Nuclear Regulatory Commission Serial: RA-23-0005 Page 2 of 2 The enclosure to this license amendment request provides Duke Energy's evaluation of the proposed changes. In addition, Attachment 1 to the enclosure provides a copy of the existing TS pages marked with the proposed changes. Attachment 2 contains the existing TS Bases pages marked to show the proposed changes for information only. Changes to the TS Bases will be implemented in accordance with the TS Bases Control Program for each site upon implementation of the respective amendment.

Approval of the proposed license amendment is requested within twelve months of acceptance.

The amendment shall be implemented within 120 days from approval.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated North Carolina State Official.

This letter contains no regulatory commitments.

Please refer any questions regarding this submittal to Ryan Treadway, Director-Nuclear Fleet Licensing, at 980-373-5873.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on May 31, 2023.

Sincerely, Thomas P. Haaf Site Vice President Harris Nuclear Plant

Enclosure:

Evaluation of the Proposed Changes Attachments: 1. Proposed Technical Specification Changes (Mark-up)

2. Proposed Technical Specification Bases Changes (Mark-up) cc:

P. Boguszewski, Senior NRC Resident Inspector, HNP D. Crowley, Radioactive Materials Branch Manager, N.C. DHSR M. Mahoney, NRC Project Manager, HNP L. Dudes, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial: RA-23-0005 Enclosure ENCLOSURE EVALUATION OF THE PROPOSED CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 20 PAGES PLUS THE COVER

U.S. Nuclear Regulatory Commission Page 1 of 20 Serial: RA-23-0005 Evaluation of the Proposed Changes License Amendment Request to Align Certain Technical Specification Requirements with Industry Standards Provided in Improved Standard Technical Specifications 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy),

hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment to the HNP TS will modify the HNP TS Surveillance Requirement (SR) 4.6.1.1 to eliminate the requirement to perform periodic position verification for containment penetrations that are maintained locked, sealed, or otherwise secured closed, as well as adopt TS Task Force (TSTF) Improved Standard TS (ISTS) Change Traveler No. 45 (TSTF-45-A), Revision 2, Exempt Verification of Containment Isolation Valves that are Not Locked, Sealed, or Otherwise Secured (ADAMS Accession No. ML040400137).

The proposed amendment will also revise HNP TS 3.3.3.5, Remote Shutdown System, to increase the completion time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance and remove the requirement to submit a Special Report. It will also relocate the content in Table 3.3-9, Remote Shutdown System, and Table 4.3-6, Remote Shutdown Monitoring Instrumentation Surveillance Requirements, in accordance with TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (ADAMS Accession No. ML040620072). Additionally, the proposed amendment will update HNP SR 4.3.1.1, Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirements, to address the application of the Surveillance Frequency Control Program (SFCP) to establish the Frequency for performance of the Analog Channel Operational Test (ACOT) of select Reactor Trip System (RTS) instrumentation.

Changes are also proposed to the Administrative Controls Section of the HNP TS to reflect current organizational titles as well as remove reporting requirements that are redundant to existing regulations. The proposed changes above reflect requirements consistent with those in Revision 5 of NUREG-1431, Standard Technical Specifications - Westinghouse Plants (ADAMS Accession No. ML21259A155).

2.0 DETAILED DESCRIPTION

2.1 Background

System Design and Operation Containment Isolation System Per Section 6.2.4 of the HNP Updated Final Safety Analysis Report (FSAR), the Containment Isolation System consists of the valves and actuators required to isolate the Containment following a loss-of-coolant accident (LOCA), steam line rupture, or fuel handling accident inside the Containment. In general, the Containment Isolation System closes fluid penetrations that support those systems not required for emergency operation. Fluid penetrations supporting Engineered Safety Features (ESF) Systems have remote manual isolation valves which may be closed from the Control Room, if necessary. Automatic isolation valves close upon receipt of an isolation signal from a sensor. All power operated isolation valves have position indication in the Control Room.

U.S. Nuclear Regulatory Commission Page 2 of 20 Serial: RA-23-0005 Each containment isolation valve is designed to ensure its performance under all anticipated environmental conditions including maximum differential pressure, seismic occurrences, steam-laden atmosphere, high temperature, and high humidity. The valve types utilized for containment isolation service are designs which provide rapid closure and near zero leakage.

Therefore, essentially no leakage is anticipated through the containment isolation valves when in closed position. Verification that actual leakage rates from the Containment are within design limits is provided by periodic leakage rate testing in accordance with 10 CFR 50, Appendix J.

RTS Instrumentation Based on the values of selected unit parameters, the RTS initiates a unit shutdown to protect against violating the core fuel design limits and RCS pressure boundary during anticipated operational occurrences, as well as assist the Engineered Safety Features Systems in mitigating accidents. The protection and monitoring systems have been designed to assure safe operation of the reactor, which is achieved by specifying limiting safety system settings in terms of parameters directly monitored by the RTS, as well as specifying limiting conditions for operation (LCOs) on other reactor system parameters and equipment performance.

As defined in HNP TS 1.3, an Analog Channel Operational Test (ACOT) shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify the operability of alarm, interlock and/or trip functions. The ACOT includes any necessary adjustments of the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.

Remote Shutdown System While the Control Room is designed to be available at all times, in the event continued occupancy of the Control Room is impossible, a separate auxiliary control panel (ACP) will be used to achieve and maintain hot standby (Mode 3). Furthermore, the reactor can be brought to cold shutdown (Mode 5) from the Control Room, and if necessary, from outside the Control Room. If temporary evacuation of the Control Room is required, the operator can establish and maintain the station in safe shutdown condition from outside the Control Room from the ACP, Auxiliary Transfer Panels (ATPs), and essential local control stations. The prime intent of the ACP and ATPs is to enable the operators to achieve and maintain hot standby condition.

The operability of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from these locations outside of the Control Room. This capability is required in the event Control Room habitability is lost and is consistent with General Design Criterion (GDC) 19 of 10 CFR Part 50. The monitoring instrumentation essential and/or desirable to support this function is identified in HNP FSAR Table 7.4.1-1.

Additionally, Remote Shutdown System operability ensures that a fire will not preclude achieving safe shutdown. The Remote Shutdown System instrumentation, control, and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is consistent with GDC 3, 10 CFR 50.48(a) and 10 CFR 50.48(c).

U.S. Nuclear Regulatory Commission Page 3 of 20 Serial: RA-23-0005 Surveillance Frequency Control Program With the implementation of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b (ADAMS Accession No. ML090850642) by HNP, most periodic frequencies of TS surveillances were relocated to a licensee-controlled program, the SFCP, along with the inclusion of the requirements for the new program in the Administrative Controls Section of the TS. License Amendment No. 154, by letter dated November 29, 2016 (ADAMS Accession No. ML16200A285), addressed the NRCs approval of HNPs implementation of TSTF-425.

The SFCP describes the requirements for the program to control changes to the relocated surveillance frequencies, establishing Nuclear Energy Institute (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, April 2007 (ADAMS Accession No. ML071360456), as the basis for making any changes to the surveillance frequencies once they are relocated out of the TS. The NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing per letter dated September 19, 2007 (ADAMS Accession No. ML072570267). The technical methodology provided in NEI 04-10 uses a risk-informed, performance-based approach for establishment of surveillance frequencies that is consistent with the philosophy of Regulatory Guide 1.174. Specifically, the use of Probabilistic Risk Assessment (PRA) methods is employed to determine the risk impact of the revised intervals.

Additionally, TSTF-425 established that all surveillance frequencies can be relocated except:

frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);

frequencies that are purely event-driven (e.g., each time the control rod is withdrawn to the full out position);

frequencies that are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power reaching 95% RTP [Rated Thermal Power]); and frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., drywell to suppression chamber differential pressure decrease).

Special Reports Regulatory Guide (RG) 1.16, Revision 1, "Reporting of Operating Information," was published by the NRC in October 1973 to provide an acceptable basis for meeting the reporting requirements of the facility operating license. It provided a description of each of the periodic reports, including annual reports and the Startup Report, that licensees are required to submit to demonstrate compliance with the TS reporting requirements. The NRC withdrew RG 1.16 in August 2009 via the Federal Register (74 FR 40244) on the basis that it was no longer needed since TS reporting requirements are contained in 10 CFR 50, as well as other parts of 10 CFR Chapter 1.

U.S. Nuclear Regulatory Commission Page 4 of 20 Serial: RA-23-0005 Additionally, the results of an NRC information gathering assessment that were provided in Generic Letter (GL) 97-02, "Revised Contents of the Monthly Operating Report," identified the existence of duplicative reporting and determined that some reports could be reduced in scope or eliminated.

2.2 Current TS Requirements The HNP TS are based upon the format and content of Revision 4 of the NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors" (ADAMS Accession No. ML102590431). As a result, the HNP TS surveillance numbers and associated Bases numbers differ from those contained in NUREG-1431.

TS SR 4.6.1.1 The HNP TS SR 4.6.1.1 for demonstrating containment integrity states:

4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that all penetrations*# not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3;

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c. By performing required visual examinations and leakage rate testing, except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.

With the following associated Notes:

• Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

Valves CP-B3, CP-B7, and CM-B5 may be verified at the frequency specified in the Surveillance Frequency Control Program by manual remote keylock switch position.

TS 3.3.3.5 Limiting Condition for Operation (LCO) 3.3.3.5.a for the Remote Shutdown System states:

3.3.3.5.a The Remote Shutdown System monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

The associated actions for TS 3.3.3.5 are as follows:

U.S. Nuclear Regulatory Commission Page 5 of 20 Serial: RA-23-0005 ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With the number of OPERABLE remote shutdown monitoring channels less than the Total Number of Channels required by Table 3.3-9, restore the inoperable channels to OPERABLE status within 60 days or submit a Special Report in accordance with Specification 6.9.2 within 14 additional days.

c. With one or more inoperable Remote Shutdown System transfer switches, power, or control circuits required by 3.3.3.5.b, restore the inoperable switch(s)/circuit(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Lastly, SR 4.3.3.5.1 states:

4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

TS SR 4.3.1.1 HNP TS SR 4.3.1.1 for addressing operability of Reactor Trip System instrumentation channels and interlocks states:

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.

The scope of this LAR specifically addresses the frequency of performing the Analog Channel Operational Test (ACOT) for the following Functional Units:

2.b.

Power Range, Neutron Flux - Low Setpoint;

5.

Intermediate Range, Neutron Flux; and

6.

Source Range, Neutron Flux The current frequency is specified as prior to each reactor startup, if not performed in the previous 31 days.

TS Section 6.0 - Administrative Controls HNP Administrative Control TS 6.6.1 addresses actions to be taken by the site for reportable events. Specifically, it states:

6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

U.S. Nuclear Regulatory Commission Page 6 of 20 Serial: RA-23-0005

b. Each REPORTABLE EVENT shall be reviewed by the On-Site Review Committee (ORC), and the results of this review shall be submitted to the Manager - Nuclear Assessment Section and the Vice President - Harris Nuclear Plant.

Additionally, the HNP TS currently utilize the positional title Superintendent-Shift Operations in numerous places throughout the Administrative Controls portion of the TS, specifically related to the individual responsible for the control room command function. These sections include:

TS 6.1 Responsibility TS 6.2 Organization TS 6.12 High Radiation Area 2.3 Reason for the Proposed Change TS SR 4.6.1.1 The proposed change to TS SR 4.6.1.1 is based on TSTF-45-A, Revision 2, which was approved generically for the Westinghouse Standard Technical Specifications, NUREG-1431, per "HNP-99-113, Informs That CP&L Proposes to Provide Response to NRC 990414 RAI Re [[generic letter" contains a listed "[" character as part of the property label and has therefore been classified as invalid., Pressure-Locking & Thermal-Binding of SR Power-Operated Gate Valves, by 990930|letter dated July 26, 1999]] (ADAMS Accession No. ML9907300113). The SRs were intended to ensure the position of valves that could be inadvertently repositioned. However, containment isolation valves that are locked, sealed, or otherwise secured are verified to be in the correct position upon being locked, sealed, or otherwise secured. Incorporating this particular change will provide dose savings by allowing containment isolation devices to be exempted from verification for applicable surveillance requirements and verified through administrative means for the related required actions.

Additionally, the proposed change is consistent with other HNP SRs to verify the position of valves, including: SR 4.1.2.1 (Boration Systems valves) SR 4.5.2, (Emergency Core Cooling System valves), SR 4.7.1.2.1 (Auxiliary Feedwater System valves), SR 4.6.2.1.c (Containment Spray System valves), SR 4.6.2.2.c (Spray Additive System valves), SR 4.7.3 (Component Cooling Water System valves), SR 4.7.4 (Emergency Service Water System valves), SR 4.7.6.d (Control Room Emergency Filtration System valves) and SR 4.7.13.b (Essential Services Chilled Water System valves).

TS 3.3.3.5 The current limitation of restoring inoperable transfer switches, power, or control circuits within 7 days is more restrictive than the allowance of other TS which were modified to allow the use of a Risk-Informed Completion Time (RICT) per License Amendment No. 184, as issued by letter dated April 2, 2021 (ADAMS Accession No. ML21047A314). This has the potential impact of effectively limiting a RICT to 7 days, rather than the potential full backstop limit of 30 days.

Changing the LCO time for TS 3.3.3.5 to 30 days will allow future RICTs to be limited by the applicable LCO (e.g., TS 3.8.1.1.b.3 for an inoperable Emergency Diesel Generator) instead of the LCO for the Remote Shutdown System, as well as align with the limit in the current Standard Technical Specifications for Westinghouse PWRs (i.e., NUREG-1431).

The proposed change to TS 3.3.3.5 will also eliminate a redundant reporting requirement that is no longer considered warranted, as consistent with the Federal Register Notice that withdrew

U.S. Nuclear Regulatory Commission Page 7 of 20 Serial: RA-23-0005 RG 1.16, the results provided in GL 97-02, and the guidance provided in NUREG-1431.

Reporting would continue to be based on the NRC's regulatory requirements as prescribed in 10 CFR Part 50.

Lastly, HNP TS Tables 3.3-9 and 4.3-6 are proposed for relocation to the HNP Technical Requirements Manual (TRM) and the SFCP, respectively. This proposed change aligns with the logic provided in NRC-approved TSTF-266-A, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (ADAMS Accession No. ML040620072), which established that listing specific instrumentation and controls is unnecessary and could lead to needless expenditure of licensee and NRC resources to process changes for information that can be adequately controlled by the licensee.

TS SR 4.3.1.1 The NRC previously approved of the adoption of TSTF-425 and the establishment of a SFCP for use by HNP. While TSTF-425 included the application of the SFCP to the corresponding ISTS SR in NUREG-1431 for performance of the Channel Operational Test (COT) for Power Range Neutron Flux - Low, Intermediate Range Neutron Flux, and Source Range Neutron Flux (i.e., ISTS SR 3.3.1.8 for Functions 2.b, 4, and 5, respectively), the HNP LAR requesting the adoption of TSTF-425 did not identify the performance of the ACOT for these Functions of HNP TS Table 4.3-1 as being applicable (i.e., Functional Units 2.b, 5, and 6, respectively). The HNP TS SR Frequency for these Functional Units is currently required to be performed prior to each reactor startup, if not performed in the previous 31 days. Application of the SFCP to these Functional Units and the performance of the ACOT is in accordance with NRC-approved TSTF-425 as previously adopted by HNP and will allow for the elimination of unnecessary performances of ACOT on these RTS instrumentation.

TS Section 6.0 - Administrative Controls The intent of the proposed change to HNP TS 6.6.1 is to eliminate redundant reporting requirements. Reporting would continue to be based on the NRC's regulatory requirements as prescribed in 10 CFR Part 50.

Additionally, the change to staff position title Superintendent - Shift Operations to Shift Manager throughout Section 6.0 of the HNP TS is to more accurately reflect current HNP staff nomenclature and responsibilities. The proposed changes are administrative in nature.

2.4 Description of the Proposed Change TS SR 4.6.1.1 The proposed changes to the applicable portions of TS SR 4.6.1.1 are as follows and are consistent with the wording in NUREG-1431 ISTS SRs 3.6.3.3 and 3.6.3.4, with consideration of the different TS format:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that all penetrations*# not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their closed positions,

U.S. Nuclear Regulatory Commission Page 8 of 20 Serial: RA-23-0005 each containment isolation manual valve and blind flange#* that is located outside containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed, except as provided in Table 3.6-1 of Specification 3.6.3;

d. Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days by verifying each containment isolation manual valve and blind flange*

that is located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except as provided in Specification 3.6.3.

• Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

New SR 4.6.1.1.d is proposed to replace the existing

• note and would reflect the SR for similar devices located inside containment. In addition, a new note is proposed that will allow verification by use of administrative means of the valves and blind flanges that are located in high-radiation areas. In this regard, the amendment would adopt the allowances afforded to ISTS SRs 3.6.3.3 and 3.6.3.4 per NRC-approved TSTF-45-A.

TS 3.3.3.5 The proposed changes to HNP TS 3.3.3.5 include relocating TS Table 3.3-9 from TS to the TRM, relocating TS Table 4.3-6 to the SFCP, deleting the special reporting requirement in Action b, extending the 7-day completion time to 30 days, and including a requirement to be in 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 /> of not meeting Action a. These changes are consistent with NUREG-1431 ISTS 3.3.4.

LIMITING CONDITION FOR OPERATION 3.3.3.5.a The Remote Shutdown System monitoring instrumentation channels shown in Table 3.3-9 specified in the Technical Requirements Manual shall be OPERABLE.

ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel(s) to OPERABLE status within 7 days 30 days, or be in 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 HOT SHUTDOWN within the next 12 following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the number of OPERABLE remote shutdown monitoring channels less than the Total Number of Channels required by Table 3.3-9, restore the inoperable channels to OPERABLE status within 60 days or submit a Special Report in accordance with Specification 6.9.2 within 14 additional days. DELETED.

c. With one or more inoperable Remote Shutdown System transfer switches, power, or control circuits required by 3.3.3.5.b, restore the inoperable switch(s)/circuit(s) to

U.S. Nuclear Regulatory Commission Page 9 of 20 Serial: RA-23-0005 OPERABLE status within 7 days 30 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6 specified in the Surveillance Frequency Control Program.

TS SR 4.3.1.1 The proposed change to HNP TS SR 4.3.1.1 addresses updating Table 4.3-1 to add a new Table Notation that will take into account the SFCP for establishing the surveillance frequency of the ACOT prior to reactor startup for the following Functional Units:

2.b.

Power Range, Neutron Flux - Low Setpoint;

5.

Intermediate Range, Neutron Flux; and

6.

Source Range, Neutron Flux Specifically, these Functional Units will no longer reflect application of Table Notation 1 (i.e., If not performed in previous 31 days), but instead new Table Notation 17, which will state: If not performed within the Frequency specified in the Surveillance Frequency Control Program. An additional change is proposed to Table 4.3-1 to include a frequency requirement to perform the ACOT for these Functional Units in accordance with the SFCP. This is in addition to the requirement to perform prior to reactor startup (as modified by new Table Notation 17). These changes are consistent with the frequencies associated with corresponding NUREG-1431 ISTS SR 3.3.1.8. An editorial change is also proposed to condense the Table Notation content on TS Pages 3/4 3-14, 3/4 3-14a and 3/4 3-15 to improve readability and remove extraneous pages.

TS Section 6.0 - Administrative Controls The proposed change to HNP TS 6.6.1 will delete the content for Reportable Events entirely.

Specifically, it will reflect the following:

6.6.1 Deleted. The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the On-Site Review Committee (ORC), and the results of this review shall be submitted to the Manager - Nuclear Assessment Section and the Vice President - Harris Nuclear Plant.

Additionally, all instances of the positional title Superintendent - Shift Operations throughout TS Section 6.0 will be updated to reflect Shift Manager, as consistent with current HNP staffing titles.

U.S. Nuclear Regulatory Commission Page 10 of 20 Serial: RA-23-0005

3.0 TECHNICAL EVALUATION

TS SR 4.6.1.1 The containment isolation valves are not being modified by this proposed amendment. HNP TS Definition 1.7 states that containment integrity shall exist when all penetrations required to be closed during accident conditions are either: 1) capable of being closed by an operable containment automatic isolation valve system; or 2) closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in TS 3.6.3.

HNP TS SR 4.6.1.1 provides the requirements that must be met to ensure containment integrity is demonstrated.

Primary containment integrity ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

The proposed change will specifically exclude verification of normally locked, sealed, or otherwise secured closed valves, blind flanges, and deactivated automatic valves, as these components are already in their required accident position. The TS SR will instead reflect the wording included in NUREG-1431 ISTS SR 3.6.3.3 and SR 3.6.3.4 with consideration for the difference in TS format. Both ISTS SR 3.6.3.3 and SR 3.6.3.4 reflect the adoption of TSTF A, as approved by the NRC per "HNP-99-113, Informs That CP&L Proposes to Provide Response to NRC 990414 RAI Re [[generic letter" contains a listed "[" character as part of the property label and has therefore been classified as invalid., Pressure-Locking & Thermal-Binding of SR Power-Operated Gate Valves, by 990930|letter dated July 26, 1999]] (ADAMS Accession No. ML19067A141).

Per TSTF-45-A, the proposed change is consistent with the valve position verification requirement for valves that have a function during an accident in other system TSs. For instance, HNP TS SR 4.5.2.b requires that each Emergency Core Cooling System subsystem be demonstrated operable, in part, by performance of a valve alignment of each valve that is in the flow path that is not locked, sealed, or otherwise secured in position. Similarly, HNP TS SR 4.7.1.2.1.b for the Auxiliary Feedwater System and HNP TS SR 4.7.4.a for the Emergency Service Water System also require demonstration of operability, in part, by verifying that each valve in the flow path of the system that is not locked, sealed, or otherwise secured in position, is in its correct position.

Administrative controls are in place that govern position verification for locked, sealed, or otherwise secured valves and blind flanges. In the case of automatic valves deactivated in accordance with HNP TS 3.6.3, Containment Isolation Valves, Action b, procedural controls are in place that track both the valve number and the associated deactivated power supply, ensuring movement of the valve is prohibited. As such, there is a very low probability that unacceptable alignment can occur.

Lastly, the addition of a note allowing an alternate (e.g., administrative) means of verifying valves located in high radiation areas aligns with the requirements of HNP TS 6.12, High Radiation Area, which already restricts access to these areas. The likelihood that valves located in high radiation areas will be misaligned after they have been verified to be locked closed is very small.

U.S. Nuclear Regulatory Commission Page 11 of 20 Serial: RA-23-0005 TS 3.3.3.5 Duke Energy is proposing changes to HNP TS 3.3.3.5 to reflect requirements consistent with those in NUREG-1431 for the Remote Shutdown System. One such change will increase the completion time from 7 days to 30 days for inoperable Remote Shutdown System monitoring instrumentation in ACTION a and ACTION c. This proposed increase is based on operating experience as well as the low probability of an event occurring that would require the evacuation of the Control Room. This justification aligns with the industry justification for the 30-day completion time reflected in the Bases for NUREG-1431 ISTS 3.3.4 and is further supported in the model safety evaluation associated with TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4B (ADAMS Accession No. ML18267A259),

which states that the completion times in the current TSs (i.e., NUREG-1431, including ISTS 3.3.4) were established using experiential data, risk insights, and engineering judgement.

ACTION a is also proposed to be modified to include a requirement that the unit be in 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 /> instead of the current be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, as this action is consistent with a controlled plant shutdown. The unit would still be required to be in Hot Shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Both of these times are reasonable for reaching the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Currently, HNP TS 3.3.3.5 ACTION b requires submitting a special report to the NRC if the number of operable remote shutdown monitoring channels is less than the total number of channels required by TS Table 3.3-9 for greater than 60 days. As part of this license amendment request, Duke Energy is proposing the elimination of this reporting requirement, as consistent with NUREG-1431 ISTS 3.3.4, Remote Shutdown System, which does not include a similar reporting requirement.

The report itself is not needed to ensure operation of the facility in a safe manner. The Actions associated with HNP TS 3.3.3.5 ensure continued safe operation by establishing and maintaining measures that ensure the Remote Shutdown System monitoring instrumentation channels are restored to operable status within the required limits. The report only provides information after a 60-day period in the event the instrumentation is not returned to operable status. It neither seeks approval from the NRC nor ensures safe operation of the facility during or after the 60 days provided to submit the report. As such, elimination of the report is appropriate on the basis that the change is consistent with NUREG-1431 and the report is not necessary to ensure operation of the facility in a safe manner.

A change is also proposed to relocate the tables of instrumentation that correspond to TS 3.3.3.5 as aligned with the guidance provided in TSTF-266-A, Revision 3. These tables list the specific instruments and respective channels necessary for each function provided by the Remote Shutdown System. The listing of this specific instrumentation in the TS is unnecessary to provide adequate assurance that the functions can be performed. The GDC 19 requirement is that the remote shutdown capability be provided. The updated LCO for TS 3.3.3.5 will reference the TRM for the relocated list of monitoring instrumentation channels currently provided in Table 3.3-9. This is a deviation from the approved guidance in TSTF-266-A, which relocates the respective table to the TS Bases, where changes can be administered under the provisions of the TS Bases Control Program. However, changes to the TRM are made under the same administrative provisions and regulated by 10 CFR 50.59. Additionally, the TRM is a document

U.S. Nuclear Regulatory Commission Page 12 of 20 Serial: RA-23-0005 incorporated by reference in the HNP FSAR. As such, changes to the TRM implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). Furthermore, the content of TS Table 4.3-6 indicates that the frequency of the required Channel Checks and Channel Calibrations are in accordance with the SFCP. This table already exists in the document controlled by TS 6.8.4.p, Surveillance Frequency Control Program.

With the removal of this table from TS 3.3.3.5, Duke Energy proposes a modification to SR 4.3.3.5.1 to instead point to the SFCP for the list of frequencies.

TS SR 4.3.1.1 Duke Energy is proposing the addition of a new Note to HNP TS Table 4.3-1 that will be applicable specifically to the ACOT frequency for power range (low setpoint), intermediate range, and source range neutron flux instrumentation. The current frequency is modified by a Note that exempts performance of the surveillance prior to startup if it has been performed within the previous 31 days. The new Note will reflect that performance of the ACOT is only required prior to startup if not performed within the Frequency specified in the SFCP. Duke Energy is also proposing to apply an additional SR Frequency for ACOT performance for the Power Range (low setpoint) and Intermediate Range neutron flux instrumentation that will be in accordance with the SFCP. The Source Range neutron flux instrumentation already has an additional ACOT Frequency in accordance with SFCP. Application of the SFCP in this manner is consistent with TSTF-425, Revision 3, as well as NUREG-1431 ISTS SR 3.3.1.8 for performance of COT.

Initially, the HNP application for adoption of TSTF-425 dated August 18, 2015 (ADAMS Accession No. ML15236A265), as supplemented by letters dated September 29, 2015 (ADAMS Accession No. ML15272A443), February 5, 2016 (ADAMS Accession No. ML16036A091), April 28, 2016 (ADAMS Accession No. ML16119A326) and May 19, 2016 (ADAMS Accession No. ML16141A048), did not address the application of the SFCP to the frequency of ACOT performance prior to reactor startup for Power Range (low setpoint), Intermediate Range, and Source Range neutron flux RTS instrumentation.

As mentioned in Section 2.1 above, TSTF-425 established that all surveillance frequencies can be relocated except:

frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);

frequencies that are purely event-driven (e.g., each time the control rod is withdrawn to the full out position);

frequencies that are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power reaching 95% RTP [Rated Thermal Power]); and frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g.,

drywell to suppression chamber differential pressure decrease).

While the current HNP TS SR frequency for this instrumentation has an event-driven component (i.e., prior to each reactor startup), the Note that modifies it does not. As evidenced by the inclusion of NUREG-1431 ISTS SR 3.3.1.8 in the scope of TSTF-425, and the reference

U.S. Nuclear Regulatory Commission Page 13 of 20 Serial: RA-23-0005 to the SFCP in the Note modifying the prior to reactor startup Frequency for ISTS SR 3.3.1.8, it is acceptable to extend the provisions of TSTF-425 to the HNP TS SR frequency for the aforementioned instrumentation. The modified frequency will not impact the ability of the SR to continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation (LCOs) will be met.

The Safety Evaluation associated with HNP License Amendment 154 and the adoption of TSTF-425 states:

The FR [Federal Register] notice published on July 6, 2009 (74 FR 31996), which announced the availability of TSTF-425, states that the addition of the SFCP to the TSs provides the necessary administrative controls to require that surveillance frequencies relocated to the SFCP are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met. The FR notice also states that changes to surveillance frequencies in the SFCP are made using the methodology contained in NEI 04-10, including qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, and recommended monitoring of structures, systems, and components (SSCs), and are required to be documented.

Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants" (i.e., the Maintenance Rule), and 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. Such failures can result in the licensee increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10 provides for monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs.

While TSTF-425 transferred control of frequencies for existing surveillances to the SFCP, it did not add, delete, or modify the content of the surveillance actions themselves. The proposed amendment relocates only existing fixed periodic surveillance frequency content for the existing identified ACOT surveillances in the HNP TS.

Additionally, the proposed amendment incorporates a new SR Frequency for ACOT performance for the Power Range (low setpoint) and Intermediate Range neutron flux instrumentation that is in accordance with the SFCP. The current surveillance frequency only requires performance prior to each reactor startup. Since a SFCP Frequency does not currently exist for this HNP RTS instrumentation, Duke Energy is proposing a starting Frequency of 184 days in accordance with the corresponding Frequencies provided in NUREG-1431 that were established in accordance with the Westinghouse Owners Group licensing topical report WCAP-15376-P, Revision 0, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times (ADAMS Accession No. ML003770874), which provided the justification for extending these ACOT surveillance test intervals for this RTS instrumentation from quarterly (92 days) to semi-annual (184 days), prior to the adoption of TSTF-425 and the introduction of the SFCP. These changes to the surveillance test intervals were previously incorporated in NUREG-1431 per TSTF-411, Revision 1, Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P) (ADAMS Accession No. ML022470164).

The approach used in this program is consistent with Regulatory Guides 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the

U.S. Nuclear Regulatory Commission Page 14 of 20 Serial: RA-23-0005 Current Licensing Bases, and 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications. The approach addresses the impact on defense-in-depth and the impact on safety margins, as well as an evaluation of the impact on risk. Going forward, these frequencies will be managed in accordance with the SFCP, as governed by HNP TS 6.8.4.p, which ensures that SRs specified in the TS are performed at intervals sufficient to assure the associated LCOs are met. Changes to the Frequencies will be made in accordance with NEI 04-10, Revision 1.

TS Section 6.0 - Administrative Controls HNP TS 6.6.1 currently requires the submission of a report to the NRC pursuant to the requirements of Section 50.73 to 10 CFR Part 50. Duke Energy proposes the deletion of this reporting requirement on the basis that the change is consistent with the NUREG-1431 ISTS and the report is a duplicative requirement. While the requirement to report is not contained within the Administrative Controls Section of NUREG-1431, the requirements in 10 CFR 50.73 adequately address the notification of the Commission and submission of a Licensee Event Report (LER) within the required 60 days after discovery of a Reportable Event.

In addition to containing requirements redundant to current regulations, the content in HNP TS 6.6.1 is addressed by the Duke Energy Quality Assurance Program Description (QAPD), as referenced in Chapter 17 of the HNP UFSAR (ADAMS Accession No. ML21319A386). This document contains select relocated administrative controls from the TS, as consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. Specifically, Duke Energy QAPD Section 17.3.4.6, Reportable Event Action, states:

17.3.4.6 Reportable Event Action Procedures are established to assure events are reviewed and notifications and reports are made as required by Regulations including, but not limited to, 10 CFR Part 21, 10 CFR 50.72, and 10 CFR 50.73.

These procedures require for significant incidents occurring during operation where a safety limit is exceeded, or which could otherwise be related to the nuclear safety of the station, the Site executive is notified, the event is investigated, and a report prepared.

These reports:

a) Contain a summary description of the circumstances and information relating to the subject incident.

b) Contain an evaluation of the effects of the incident.

c) Describe corrective action taken or recommended as a result of the incident.

d) Describe, analyze and evaluate any significant nuclear safety related implications of the incident.

SSCs categorized as Safety-Related, Low Safety Significant (RISC-3) in accordance with 10CFR50.69 and the site license are no longer subject to the requirements of this document. These 50.69 LSS SSCs are no longer subject to the requirements of 10 CFR 50 Appendix B, 10 CFR Part 21, 10 CFR 50.72, 10 CFR 50.73 and other regulations as noted in the rule.

U.S. Nuclear Regulatory Commission Page 15 of 20 Serial: RA-23-0005 The reporting requirement does not satisfy the criteria of 10 CFR 50.36 for inclusion in TS as a limiting condition for operation and is adequately controlled by other regulations. Furthermore, future changes to the Duke Energy QAPD are controlled by compliance changes to 10 CFR 50.54(a). As such, elimination of this Administrative Control TS is acceptable due to the actions being proposed for elimination being adequately addressed by both regulation and the Duke Energy QAPD.

Regarding the proposed changes to staff title Superintendent - Shift Operations to Shift Manager throughout Section 6.0 of the HNP TS, the change itself is in name only. The actual function of these individuals is not being altered. The changes are consistent with the guidance of ISTS (NUREG-1431) and do not adversely impact the minimum staffing levels required by TS. The proposed changes continue to ensure that no individual is assigned functions that will result in conflicting roles during design basis, fire, security or other events.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Guidance 10 CFR 50.36, Technical specifications The NRC's regulatory requirements related to the content of the TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." This regulation requires that the TS include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation (LCO), (3) surveillance requirements, (4) design features, and (5) administrative controls.

Per 10 CFR 50.36(c)(2)(ii), a TS LCO must be established for each item meeting one or more of the following criteria:

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.

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.

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.

Criterion 4:

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Per 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

U.S. Nuclear Regulatory Commission Page 16 of 20 Serial: RA-23-0005 Per 10 CFR 50.36(c)(5), administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.

Appendix A to Part 50, General Design Criteria (GDC) for Nuclear Power Plants GDC 16, Containment design, requires that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 19, Control room, requires that a Control Room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

It also requires that adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Furthermore, equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

GDC 53, Provisions for containment testing and inspection, requires that the reactor containment shall be designed to permit (1) appropriate periodic inspection of all important area, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak-tightness of penetrations which have resilient seals and expansion bellows.

GDC 54, Piping systems penetrating containment, requires that piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. These piping systems are to be designed with a capability to periodically test the operability of the isolation valves and associated apparatus and determine whether valve leakage is within acceptable limits.

GDC 55, Reactor coolant pressure boundary penetrating containment, requires, in part, that each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves unless it can be demonstrated that the containment isolation provisions for a specific class of lines are acceptable on some other defined basis.

GDC 56, Primary containment isolation, requires, in part, that each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves unless it can be demonstrated that the containment isolation provisions for a specific class of lines are acceptable on some other defined basis.

U.S. Nuclear Regulatory Commission Page 17 of 20 Serial: RA-23-0005 GDC 57, Closed system isolation valves, requires, in part, that each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation.

The proposed changes do not impact the design of the plant; therefore, these requirements remain satisfied.

Conclusion Duke Energy has evaluated the proposed changes against the applicable regulatory requirements described above. Based on this evaluation, there is reasonable assurance that the health and safety of the public will remain unaffected following the approval of these proposed changes.

4.2 Precedents TS SR 4.6.1.1 The NRC issued an amendment to Arkansas Nuclear One, Unit No. 2, per letter dated December 18, 2006 (ADAMS Accession No. ML062910435), that addressed modifications to the TS SR 4.6.1.1 regarding containment isolation valve position verification. Specifically, the amendment eliminated the requirement to verify containment isolation valves that are maintained locked, sealed, or otherwise secured closed from the monthly position verification. It also added a new SR to replace the existing note and address the similar devices located inside containment, as well as added a new note to allow verification of the valves and blind flanges that are located in high-radiation areas by administrative means.

TS 3.3.3.5 The NRC issued amendments to the South Texas Project, Units 1 and 2, per letter dated August 20, 2004 (ADAMS Accession No. ML042370841), to revise TSs for the Remote Shutdown System to reflect requirements consistent with those in NUREG-1431, Standard Technical Specifications - Westinghouse Plants. The changes increased the completion time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance and relocated the description of the required components to a licensee-controlled document.

TS SR 4.3.1.1 Many sites that were issued license amendments to implement TSTF-425, Revision 3, proposed and obtained NRC approval of a Note in the Frequency column for corresponding ISTS SR 3.3.1.8 (COT) that stated the Surveillance was only required when not performed within the Frequency specified in the SFCP. Recent examples include:

U.S. Nuclear Regulatory Commission Page 18 of 20 Serial: RA-23-0005 Watts Bar Nuclear Plant, Units 1 and 2, Amendment No. 132 to Facility Operating License No. NPF-90 and Amendment No. 36 to Facility Operating License No. NPF-96 per letter dated February 28, 2020 (ADAMS Accession No. ML20028F733)

H. B. Robinson Steam Electric Plant, Unit No. 2, Amendment No. 265 to Renewed Facility Operating License No. DPR-23 per letter dated August 15, 2019 (ADAMS Accession No. ML19158A307)

Wolf Creek Generating Station, Unit 1, Amendment No. 227 to Renewed Facility Operating License No. NPF-42 per letter dated April 8, 2021 (ADAMS Accession No. ML21053A117).

TS Section 6.0 - Administrative Controls By letter dated April 24, 2003 (ADAMS Accession No. ML031140670), the NRC issued amendments to the South Texas Project, Units 1 and 2, approving revisions to specific requirements of TS Section 6.0 to be consistent with NUREG-1431. Two changes in particular that were approved for these amendments include removing the Reportable Events TS and updating specific management titles to reflect more generic title positions.

Furthermore, by letter dated February 12, 2015 (ADAMS Accession No. ML15002A324), the NRC issued license amendments to Catawba Nuclear Station, Units 1 and 2, McGuire Nuclear Station, Units 1 and 2, and Oconee Nuclear Stations, Units 1, 2, and 3, to address administrative and editorial changes to update organizational titles to more closely reflect the current respective organizations.

4.3 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed amendment to the HNP TS will modify the HNP TS Surveillance Requirement (SR) 4.6.1.1 to eliminate the requirement to perform periodic position verification for containment penetrations that are maintained locked, sealed, or otherwise secured closed, as well as adopt TS Task Force (TSTF) Improved Standard TS (ISTS) Change Traveler No. 45 (TSTF-45-A), Revision 2, Exempt Verification of Containment Isolation Valves that are Not Locked, Sealed, or Otherwise Secured (ADAMS Accession No. ML040400137). The proposed amendment will also revise HNP TS 3.3.3.5, Remote Shutdown System, to increase the completion time for inoperable Remote Shutdown System components to a time that is more consistent with their safety significance and remove the requirement to submit a Special Report.

It will also relocate the content in Table 3.3-9, Remote Shutdown System, and Table 4.3-6, Remote Shutdown Monitoring Instrumentation Surveillance Requirements, in accordance with TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of Instrumentation and Controls (ADAMS Accession No. ML040620072). Additionally, the proposed amendment will update HNP SR 4.3.1.1, Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirements, to address the application of the Surveillance Frequency Control Program (SFCP) to establish the Frequency for performance of the Analog Channel Operational Test (ACOT) of select Reactor Trip System (RTS) instrumentation. Changes are also proposed to the Administrative Controls Section of the HNP TS to reflect current organizational titles as well as remove reporting requirements that are redundant to existing regulations. The proposed changes above reflect requirements consistent with those in Revision 5 of NUREG-1431,

U.S. Nuclear Regulatory Commission Page 19 of 20 Serial: RA-23-0005 Standard Technical Specifications - Westinghouse Plants (ADAMS Accession No. ML21259A155).

Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

(1)

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

Response: No.

Duke Energy is proposing a change that will exempt containment isolation valves (CIVs) that are locked, sealed, or otherwise secured in position from the periodic verification of valve position required by SR 4.6.1.1. The exempted valves are verified to be in the correct position upon being locked, sealed, or secured. Therefore, the valves are in the condition assumed in the accident analysis, in which case the proposed change will not affect the initiators or mitigation of any accident previously evaluated.

Duke Energy is also proposing to revise HNP TS 3.3.3.5 to increase the completion time for inoperable Remote Shutdown System components, remove the requirement to submit a Special Report, and relocate the content in TS Tables 3.3-9 and 4.3-6 to licensee-controlled documents.

Changes are also proposed to amend HNP TS Table 4.3-1 to apply the SFCP to establish the Frequency of ACOT performance for select RTS instrumentation and apply administrative changes to HNP TS Section 6.0 to reflect current organizational titles and remove reporting requirements redundant to existing regulations. These proposed changes do not make changes to the physical plant or analytical methods.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, or components from performing their intended function to mitigate the consequences on an initiating event with the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational or public radiation exposure.

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

(2)

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

Response: No.

The proposed changes neither install or remove any plant equipment, nor alter the design, physical configuration, or mode of operation of any plant structure, system, or component.

U.S. Nuclear Regulatory Commission Page 20 of 20 Serial: RA-23-0005 The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the Updated FSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.

Specifically, no new hardware is being added to the plant as part of the proposed changes, no existing equipment design or function is being modified, and no significant changes in operations are being introduced. No new equipment performance burdens are imposed.

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

(3)

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

Response: No.

The proposed changes in this license amendment request do not alter the design, configuration, operation, or function of any plant system, structure, or component. The ability of any operable structure, system, or component to perform its designated safety function is unaffected by these changes. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. They do not alter any safety analysis assumptions, initial conditions, or results of any accident analyses. The proposed changes will not result in plant operation in a configuration outside the design basis.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above evaluation, Duke Energy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

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

5.0 ENVIRONMENTAL CONSIDERATION

S Duke Energy has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined by 10 CFR 20, or it would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs be prepared in connection with the proposed amendment.

U.S. Nuclear Regulatory Commission Serial: RA-23-0005 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

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S/U(17),

SFCP S/U(17),

SFCP 17

SHEARON HARRIS - UNIT 1 3/4 3-14 Amendment No. 154 TABLE 4.3-1 (Continued)

TABLE NOTATIONS When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1)

If not performed in previous 31 days.

(2)

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3)

Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

Detector plateau curves shall be obtained, and evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7)

Each train shall be tested at the frequency specified in the Surveillance Frequency Control Program.

(8)

Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(9)

Setpoint verification is not applicable.

(10)

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the reactor trip breakers.

Included for information only. No changes proposed to technical content on this page.

Included for information only. No changes proposed to technical content on this page. Editorial change will relocate content to prior page.

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(11)

CHANNEL CALIBRATION shall include the RTD response time.

(12)

Verify that appropriate signals reach the undervoltage and shunt trip relays, for both the main and bypass breakers, from the manual reactor trip switch.

SHEARON HARRIS - UNIT 1 3/4 3-14a Amendment No. 4-:3

ADD:

(17)

If not performed within the Frequency specified in the Surveillance Frequency Control Program.

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(13) Remote manual shunt trip prior to placing breaker in service.

(14) Automatic Llndervoltage trip.

(15) Not used.

(16) The MODES specified for these channels in Table 4.3-2 are more restrictive and. therefore. applicable.

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

SHEARON HARRIS - UNIT 1 3/4 3-63 Amendment No. 179 INSTRUMENTATION REMOTE SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.5.a The Remote Shutdown System monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

3.3.3.5.b All transfer switches, Auxiliary Control Panel Controls and Auxiliary Transfer Panel Controls for the OPERABILITY of those components required by the SHNPP Safe Shutdown Analysis to (1) remove decay heat via auxiliary feedwater flow and steam generator power-operated relief valve flow from steam generators A and B, (2) control RCS inventory through the normal charging flow path, (3) control RCS pressure, (4) control reactivity, and (5) remove decay heat via the RHR system shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With the number of OPERABLE remote shutdown monitoring channels less than the Total Number of Channels required by Table 3.3-9, restore the inoperable channels to OPERABLE status within 60 days or submit a Special Report in accordance with Specification 6.9.2 within 14 additional days.

c.

With one or more inoperable Remote Shutdown System transfer switches, power, or control circuits required by 3.3.3.5.b, restore the inoperable switch(s)/circuit(s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

4.3.3.5.2 Each Remote Shutdown System transfer switch, power and control circuit and control switch required by 3.3.3.5.b, shall be demonstrated OPERABLE at the frequency specified in the Surveillance Frequency Control Program.

30 days specified in the Surveillance Frequency Control Program.

DELETED.

30 days Add:

Pages 3/4 3-64 through 3/4 3-65 deleted by Amendment No. XXX specified in the Technical Requirements Manual 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

~..._______.

Relocated to the Technical Requirements Manual

c

)>

o

o 1-t l/l C

z 1-t

-i w

.J:I>

w m

.J:I>

l.

C TAOLE 3.3-9 REMOTE SIIUTDOWN SYS 1 tH READOUT LOCATION

1.

ant System Hot-Leg Temperature ACP*

2.

stem Cold-Leg Temperature ACP*

3.

4.

5.

6.

7.

8.

9.

Pressurizer Pressure Pressurizer Level Steam Generator Pressure Steam Generator Water Level--Wfde Rang (Note 1)

Residual *Heat Removal Flow Auxiliary Feedwater Flow (Note 1)

Condensate Storage Tank Level ACP*

ACP*

ACP*

ACP*

ACP*

ACP*

10.

11.

Reactor Coolant System Pressure-Wide Range ACP*

Wide-Range Flux Monitor (SR Indicator)

ACP*

12.

Charging Header Flow

13.

a.

Auxiliary Feedwater Turbine Steam Inlet--Pump Discharge 6P or

b.

Auxiliary Feedwater Turbine Speed

14.

Boric Acid Tank Level

• ACP = Auxiliary Control Panel

• • SSA= Safe Shutdown Analysis ACP*

ACP*

ACP*

ACP*

TOTAL NO.

OF CIIANNELS 2

2 2

2 1/Steam Generator 1/Steam Generator 2

I/Steam Generator 2

I 1

I 1

C HINIHUH CHANNELS OPERABLE 2

2 1-SSA Channel**

I-SSA Channel**

1/Steam Generator I/Steam Generator I (Note 2)

N.A. (Note 3)

I-SSA Channel"'"'

I-SSA Channel"'"'

I-SSA Channel"'~

I-SSA Channel"'"'

I-SSA Channel*"'

Note 1 - Steam Generators A&D Only Nol.c? ? - tum Trc1ln R Only Hot.e 3 - Steam Generator Water Level

SHEARON HARRIS - UNIT 1 3/4 3-65 Amendment No. 154 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION

1.

Reactor Coolant System Hot-Leg Temperature SFCP SFCP

2.

Reactor Coolant System Cold-Leg Temperature SFCP SFCP

3.

Pressurizer Pressure SFCP SFCP

4.

Pressurizer Level SFCP SFCP

5.

Steam Generator Pressure SFCP SFCP

6.

Steam Generator Water Level--Wide Range SFCP SFCP

7.

Residual Heat Removal Flow SFCP SFCP

8.

Auxiliary Feedwater Flow SFCP SFCP

9.

Condensate Storage Tank Level SFCP SFCP

10.

Reactor Coolant System Pressure--Wide Range SFCP SFCP

11.

Wide-Range Flux Monitor (SR Indicator)

SFCP SFCP

12.

Charging Header Flow SFCP SFCP

13.

a. Auxiliary Feedwater Turbine Steam Inlet--Pump Discharge P SFCP SFCP

b. Auxiliary Feedwater Turbine Speed SFCP SFCP

14.

Boric Acid Tank Level SFCP SFCP

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

Valves CP-B3, CP-B7, and CM-B5 may be verified at the frequency specified in the Surveillance Frequency Control Program by manual remote keylock switch position.

SHEARON HARRIS - UNIT 1 3/4 6-1 Amendment No. 181 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At the frequency specified in the Surveillance Frequency Control Program by verifying that all penetrations*# not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3;

b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c.

By performing required visual examinations and leakage rate testing, except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.

Insert A Insert A:

At the frequency specified in the Surveillance Frequency Control Program by verifying that each containment isolation manual valve and blind flange#* that is located outside containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed, except as provided in Specification 3.6.3.

Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days by verifying each containment isolation manual valve and blind flange* that is located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except as provided in Specification 3.6.3.

d.

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

SHEARON HARRIS - UNIT 1 6-1 Amendment No. 

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Superintendent-Shift Operations (or, during his absence from the control room, a designated individual) shall be responsible for the control room command function.

6.2 ORGANIZATION 6.2.1 Onsite And Offsite Organization An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be documented in the FSAR and updated in accordance with 10 CFR 50.71(e).

b.

There shall be an individual executive position (corporate officer) in the offsite organization having corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

c.

There shall be an individual management position in the onsite organization having responsibility for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant.

d.

Although the individuals who train the operating staff and those who carry out the quality assurance functions may report to the appropriate manager onsite, they shall have sufficient organizational freedom to be independent from operating pressures.

e.

Although health physics individuals may report to any appropriate manager onsite, for matters relating to radiological health and safety of employees and the public, the health physics manager shall have direct access to that onsite individual having responsibility for overall unit management. Health physics personnel shall have the authority to cease any work activity when worker safety is jeopardized or in the event of unnecessary personnel radiation exposures.

Shift Manager

Shift Manager ADMINISTRA~IVE CONTROLS UNIT STAFF 6.2.2 The unit organization shall be subject to the following :

a.

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1 :

b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

In addition. while the unit is in MODE 1.

2. 3. or 4. at least one licensed Senior Operator shall be in the control room:

c.

An individual qualified as a Radiation Control Technician* shall be on site when fuel is in the reactor;

d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation:

e.

The Manager-Operations shall meet one of the following :

(1)

Hold a Senior Operator License or.

(2)

Have held a Senior Operator License for a similar unit. or (3)

Have been certified for equivalent senior operator knowledge for a similar unit.

If the Manager-Operations does not hold a Senior Reactor Operator License. an off-shift who reports directly to the Manager-Operat1 and holds a Senior Reactor Operator License shall be designate to supervise shift work and licensed activities.

The Radiation Control Technician composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. in order to accommodate unexpected absence. provided immediate action is taken to fill the req~ired positions.

SHEARON HARRIS - UNIT 1 6-la Amendment No.

9-9

Shift Manager SM Shift Manager Shift Manager SM Shift Manager TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION POSITION MODE 1, 2, 3, or 4 MODE 5 or 6

£-Se 1

1 SRO 1

None RO 2

I AO 2

1 STA None

~

Unit I SRO RO AO STA Individual with a Senior Operator license on Unit 1 Individual with an Operator license on Unit 1 Auxiliary Operator - license not required Shift Technical Advisor The shift crew composition may be one less than the m1n1mum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the SuperiAtenden~

ns from the control room while the unit is in MODE 1, 2, 3, or 4, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function.

During any absence of the from the control room while the unit is in MODE 5 or 6, an individu ith a valid Senior Operator license or Operator license shall be designated t ssume the control room command function.

nned in MODES 1, 2, 3, and 4 unless the or the individual with a Senior Operator license meets the qualifications for the STA as required by the NRC.

SHEARON HARRIS - UNIT 1 6-5 Amendment No.

83

Shift Manager ADMINISTRATIVE CONTROLS 6.2.3 DELETED 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 shall provide advisory technical support to the :'ritiee-t'4fl~Aft1~~fH-'tt--Hflie-t"-it-f:-'Htt9i-s in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

The Shift Technical Advisor shall have a baccalaureate degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 Deleted SHEARON HARRIS - UNIT 1 6-6 Amendment No.

8a

SHEARON HARRIS - UNIT 1 6-16 Amendment No. 

ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the On-Site Review Committee (ORC), and the results of this review shall be submitted to the Manager - Nuclear Assessment Section and the Vice President - Harris Nuclear Plant.

6.7 SAFETY LIMIT VIOLATION 6.7.1 Deleted.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No.

82-33; c.

Security Plan implementation; d.

Emergency Plan implementation; e.

PROCESS CONTROL PROGRAM implementation; f.

OFFSITE DOSE CALCULATION MANUAL implementation; Deleted.

Shift Manager ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)

5.

Be under the surveillance. as specified in the RWP or equivalent. while in the area, of an individual qualified in radiation protection procedures. equipped with a radiation monitoring and indicating device; who is responsible for controlling personnel radiation exposure within the area.

e.

Except for individuals qualified in radiation protection procedures.

or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination.

knowledge. and pre-job briefing does not require documentation prior to entry.

6.12.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a.

Each accessible entryway to such an area shall be conspicuously posted as a locked high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and in addition:

1.

All such door and gate keys shall administrative control of the ~~'-'l-fl'H>ftfl4ilm::--

~t44'---Hfif>i~-'l-f,l'\\"s or the Radiation Control Supervisor or designated representative: and

2. Doors and gates shall remain locked or guarded except during periods of personnel or eqiupment entry or exit.

b.

Access to, and activities in. each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c.

Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to. exit from, and work in such areas.

d.

Each individual or group entering such an area shall :

1. Possess an alarming dosimeter with an appropriate alarm setpoint : or

2.

Possess a radiation monitoring device that continuously transmits dose rate and cumulative dose information to a temote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area; or SHEARON HARRIS - UNIT 1 6-26a Amendment No. t-25

U.S. Nuclear Regulatory Commission Serial: RA-23-0005 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP)

(FOR INFORMATION ONLY)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 2 PAGES PLUS THE COVER

SHEARON HARRIS - UNIT 1 B 3/4 6-1 Amendment No. 181 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La, during performance of the periodic test, to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates is consistent with the requirements of the Containment Leakage Rate Testing Program for Type A, B, and C tests.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

Action statement "a" has been modified by a note. The note allows use of the air lock for entry and exit for seven days under administrative controls if both air locks have an inoperable door.

This seven day restriction begins when a door in the second air lock is discovered to be inoperable. Containment entry may be required to perform Technical Specification surveillances and actions, as well as other activities on equipment inside containment that are required by Technical Specifications (TS) or other activities that support TS required equipment. In addition, containment entry may be required to perform repairs on vital plant equipment, which if not repaired, could lead to a plant transient or a reactor trip. This note is not intended to preclude performing other activities (i.e., non-TS required activities or repairs on non-vital plant equipment) if the containment is entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probability of an event that could pressurize containment during the short time that an OPERABLE door is expected to be open.

Insert A

INSERT A SR 4.6.1.1.a requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. A Note addresses valves and blind flanges located in high radiation areas and allows them to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.

SR 4.6.1.1.d requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. A Note addresses valves and blind flanges located in high radiation areas and allows them to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.