ML21035A132

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Issuance of Amendment No. 182 Regarding Technical Specifications Related to Accident Monitoring Instrumentation, and Refueling Operations Instrumentation
ML21035A132
Person / Time
Site: Harris 
Issue date: 03/16/2021
From: Michael Mahoney
Plant Licensing Branch II
To: Maza K
Duke Energy Progress
Mahoney M
References
EPID L-2020-LLA-0045
Download: ML21035A132 (31)
v • d • e


Text

March 16, 2021 Ms. Kim Maza Site Vice President Shearon Harris Nuclear Power Plant 5413 Shearon Harris Road Mail Code NHP01 New Hill, NC 27562-9300

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 182 REGARDING TECHNICAL SPECIFICATIONS RELATED TO ACCIDENT MONITORING INSTRUMENTATION, AND REFUELING OPERATIONS INSTRUMENTATION (EPID L-2020-LLA-0045)

Dear Ms. Maza:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 182 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1. This amendment is in response to your application dated March 12, 2020, as supplemented by letters dated June 22, 2020 and October 2, 2020.

The amendment revises Technical Specification (TS) 3.3.3.6, Accident Monitoring Instrumentation, to change the allowed outage time for inoperable post-accident monitoring (PAM) instrumentation in Action Statements a and b, replace the shutdown requirement in Action Statement a, for inoperable PAM instruments when the minimum required channels are operable, with a requirement to submit a Special Report to the NRC within 14 days of exceeding the completion time, delete Action Statements d and e, and add a Note that allows a separate entry for each instrument function.

The amendment also revises TS 3.9.2, Instrumentation, to remove the audible indication requirement in Mode 6, as well as relocate the requirements for electrical equipment protective devices in TS 3.8.4.1, Containment Penetration Conductor Overcurrent Protective Devices, and TS 3.8.4.2, Motor Operated Valves Thermal Overload Protection, from TSs to licensee-controlled procedure PLP-106, Technical Specification Equipment List Program. Additionally, the amendment also revises the Note in TS 3.9.2 to allow for the substitution of Wide Range Neutron Flux Monitors for both of the Source Range Neutron Flux Monitors required to be operable while in Mode 6.

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

Sincerely,

/RA/

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

Enclosures:

1. Amendment No. 182 to NPF-63
2. Safety Evaluation cc: Listserv

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 182 Renewed License No. NPF-63

1.

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

A.

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

dated March 12, 2020, as supplemented by letters dated June 22, 2020 and October 2, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-63 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 182, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility License No. NPF-63 and Technical Specifications Date of Issuance: March 16, 2021 Undine S.

Shoop Digitally signed by Undine S. Shoop Date: 2021.03.16 09:50:21 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 182 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following page of the Renewed Facility Operating License with the revised page.

The revised page is identified by amendment number and contains a marginal line indicating the area of change:

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

Remove Insert xi xi xv xv 3/4 3-66 3/4 3-66 3/4 3-67 3/4 3-67 3/4 8-19 3/4 8-19 3/4 8-20 3/4 8-21 3/4 8-39 3/4 8-40 3/4 9-3 3/4 9-3 Renewed License No. NPF-63 Amendment No. 182 C.

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

(1)

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

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 182, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

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

(4)

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

(5)

Steam Generator Tube Rupture (Section 15.6.3)

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

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

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

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

INDEX SHEARON HARRIS - UNIT 1 xi Amendment No. 182 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8.4 (DELETED).............................................................................................3/4 8-19 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION................................................................... 3/4 9-1 TABLE 3.9-1 ADMINISTRATIVE CONTROLS TO PREVENT DILUTION DURING REFUELING............................................................................................ 3/4 9-2 3/4.9.2 INSTRUMENTATION.............................................................................. 3/4 9-3 3/4.9.3 (DELETED).............................................................................................. 3/4 9-4 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................................ 3/4 9-5 3/4.9.5 (DELETED).............................................................................................. 3/4 9-6 3/4.9.6 (DELETED).............................................................................................. 3/4 9-7 3/4.9.7 (DELETED).............................................................................................. 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level..................................................................................... 3/4 9-9 Low Water Level......................................................................................3/4 9-10 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM..........................3/4 9-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL....................................................3/4 9-12 3/4.9.11 WATER LEVEL - NEW AND SPENT FUEL POOLS...............................3/4 9-13 3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM........3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN............................................................................3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.....................................................................................................3/4 10-2 3/4.10.3 PHYSI T

CS ESTS......................................................................................3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS................................................................3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN..................................3/4 10-5

INDEX SHEARON HARRIS - UNIT 1 xv Amendment No. 182 BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE........................................................................................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.............. B 3/4 7-2 3/4.7.3 COMPONENT COOLING WATER SYSTEM................................................. B 3/4 7-3 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM.................................................. B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK.................................................................................. B 3/4 7-3 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM............................ B 3/4 7-3 3/4.7.7 REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM........B 3/4 7-3a 3/4.7.8 SNUBBERS.................................................................................................... B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION.......................................................... B 3/4 7-5 3/4.7.10 (DELETED)..................................................................................................... B 3/4 7-5 3/4.7.11 (DELETED)..................................................................................................... B 3/4 7-5 3/4.7.12 (DELETED)..................................................................................................... B 3/4 7-5 3/4 7.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM................................... B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, AND 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION..................................................................... B 3/4 8-1 3/4.8.4 (DELETED)..................................................................................................... B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.......................................................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION..................................................................................... B 3/4 9-1 3/4.9.3 (DELETED)..................................................................................................... B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS............................................... B 3/4 9-1 3/4.9.5 (DELETED)..................................................................................................... B 3/4 9-2 3/4.9.6 (DELETED)..................................................................................................... B 3/4 9-2 3/4.9.7 (DELETED)..................................................................................................... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.................... B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM................................. B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND NEW AND SPENT FUEL POOLS................................................................. B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM............. B 3/4 9-4

SHEARON HARRIS - UNIT 1 3/4 3-66 Amendment No. 182 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

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

A separate ACTION entry is allowed for each INSTRUMENT listed in Table 3.3-10.

a.

With the number of OPERABLE accident monitoring instrumentation channels less than the Total Required Number of Channels requirements shown in Table 3.3-10 restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a Special Report pursuant to Specification 6.9.2 within the following 14 days from the time the action is required. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channel(s) to operable status.

b.

With the number of OPERABLE accident monitoring instrumentation channels, except the radiation monitors, the Pressurizer Safety Valve Position Indicator, or the Reactor Coolant System Subcooling Margin Monitor, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days 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 at least 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 />.

c.

With the number of OPERABLE accident monitoring instrumentation channels for the radiation monitor(s), the Pressurizer Safety Valve Position Indicator*, or the Reactor Coolant System Subcooling Margin Monitor#, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore the inoperable channel(s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 14 days, that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channel(s) to OPERABLE status.

d.

DELETED.

e.

DELETED.

The alternate method shall be a check of safety valve piping temperatures and evaluation to determine position.

The alternate method shall be the initiation of the backup method as required by Specification 6.8.4.d.

SHEARON HARRIS - UNIT 1 3/4 3-67 Amendment No. 182 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-7.

SHEARON HARRIS - UNIT 1 3/4 8-19 Amendment No. 182 ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Specifications 3/4.8.4.1 and 3/4.8.4.2 have been deleted from Technical Specifications and relocated to plant procedure PLP-106.

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

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

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

  • Wide Range Neutron Flux Monitors may be substituted for Source Range Neutron Flux Monitors provided the two required OPERABLE monitors (Source Range Neutron Flux Monitors and/or Wide Range Neutron Flux Monitors) are located on opposite sides of the core.

SHEARON HARRIS - UNIT 1 3/4 9-3 Amendment No. 182 REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors* shall be OPERABLE, each with continuous visual indication in the control room.

APPLICABILITY: MODE 6.

ACTION:

a.

With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b.

With both of the above required monitors inoperable or not operating, in addition to Action a. above, immediately initiate actions to restore one source range neutron flux monitor to OPERABLE status and determine the boron concentration of the Reactor Coolant System within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

SURVEILLANCE REQUIREMENTS 4.9.2 Each neutron flux monitor shall be demonstrated OPERABLE by performance of:

a.

A CHANNEL CHECK at the frequency specified in the Surveillance Frequency Control Program, b.

A CHANNEL CALIBRATION at the frequency specified in the Surveillance Frequency Control Program.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 182 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By application dated March 12, 2020 (Reference 1), as supplemented by letters dated June 22, 2020 and October 2, 2020 (References 2 and 3, respectively), Duke Energy Progress, LLC (the licensee), requested changes to the technical specifications (TSs) for the Shearon Harris Nuclear Power Plant (Harris or HNP), Unit 1. The license amendment request (LAR) proposes changes to TS 3.3.3.6, Accident Monitoring Instrumentation, to revise the allowed outage time for inoperable post-accident monitoring (PAM) instrumentation in Action Statements a and b, replace the shutdown requirement in Action Statement a, for inoperable PAM instruments when the minimum required channels are operable, with a requirement to submit a Special Report to the NRC within 14 days of exceeding the completion time, delete Action Statements d and e, and add a Note that allows a separate entry for each instrument function.

The licensee is also proposing to revise TS 3.9.2, Instrumentation, to remove the audible indication requirement in Mode 6, as well as relocate the requirements for electrical equipment protective devices in TS 3.8.4.1, Containment Penetration Conductor Overcurrent Protective Devices, and TS 3.8.4.2, Motor Operated Valves Thermal Overload Protection, from TSs to licensee-controlled procedure PLP-106, Technical Specification Equipment List Program.

Additionally, the licensee is proposing a revision to the Note in TS 3.9.2 to allow for the substitution of Wide Range Neutron Flux Monitors for both of the Source Range Neutron Flux Monitors required to be operable while in Mode 6.

The supplements dated June 22, 2020 and October 2, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs initial proposed no significant hazards consideration determination as published in the Federal Register on August 25, 2020 (85 FR 52370).

2.0 REGULATORY EVALUATION

2.1 System Descriptions Accident Monitoring Instrumentation The Post-Accident Monitoring System is designed to monitor plant variables during and following an accident. It continuously displays the appropriate monitored variables and enables the operator to perform manual safety functions and to determine the effect of these actions.

Nuclear Instrumentation In Chapter 7.2.1.1.6, Analog System, of the Updated Final Safety Analysis Report (Reference 4), the licensee describes the Nuclear Instrumentation at HNP, in part:

The primary function of nuclear instrumentation is to protect the reactor by monitoring the neutron flux and generating appropriate trips and alarms for various phases of reactor operating and shutdown conditions. It also provides a secondary control function and indicates reactor status during startup and power operation. The Nuclear Instrumentation System uses information from three separate types of instrumentation channels to provide three discrete protection levels. Each range of instrumentation (source, intermediate, and power) provides the necessary overpower reactor trip protection required during operation in that range. The overlap of instrument ranges provides reliable continuous protection beginning with source level through the intermediate and low power level.

NUREG-1431, Bases, Section B 3.9.3 (Reference 5), describes an audible alarm for the source range neutron monitoring instrumentation:

The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition.

Two OPERABLE source range neutron flux monitors are required to provide a signal to alert the operator to unexpected changes in core reactivity such as with a boron dilution accident or an improperly loaded fuel assembly. The audible count rate from the source range neutron flux monitors provides prompt and definite indication of any boron dilution.

With no audible [alarm] [count rate] OPERABLE, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN.

AC (Alternative Current) Power Systems According to the HNP Final Safety Analysis Report (FSAR), Section 8.3.1, AC Power Systems, (Reference 6), AC Power Systems, modular type penetrations are used for all electrical conductors passing through the containment wall. High-voltage penetrations are equipped with bushing type terminations, and low-voltage power control modules are provided with pigtails. Safety-related penetrations are protected with primary and backup overcurrent protective devices. The penetration circuit protection is designed, so that the primary and back-up disconnecting devices can each limit the maximum thermal energy resulting from current flow at the penetration to a value less than that required for thermal damage to the penetration conductor. The penetrations are designed to withstand (without loss of mechanical integrity) the maximum short circuit current that could occur due to a through fault for a period long enough to allow back-up circuit protection to operate, assuming a failure of the primary protection device.

Thermal overload protection is provided for all safety-related motor-operated valves (MOVs).

Electrical circuits bypass the thermal overload protection devices of all safety-related MOVs under accident conditions. This gives enough time for the motor, if overloaded but not stalled, to complete the valve closure (or opening).

The requirements for the containment penetration conductor overcurrent protective devices and MOVs overload protective devices are provided in existing HNP TS 3/4 8.4, Electrical Equipment Protective Devices Containment Penetration Conductor Overcurrent Protective Devices.

2.2 Description of Changes The licensee proposed to revise HNP Technical Specifications, as follows:

TS Index - 3.0/4.0 Limiting Conditions for Operation and Surveillance Requirements TS Index Page xi currently states, in part, as follows:

SECTION PAGE 3/4 8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices.3/4 8-19 TABLE 3.8-1 DELETED3/4 8-21 Motor-Operated Valves Thermal Overload Protection..3/4 8-39 Table 3.8.2 DELETED3/4 8-40 TS Index Page xi will state, in part, as follows:

SECTION PAGE 3/4 8.4 DELETED...3/4 8-19 TS Index - 3.0/4.0 Bases TS Index Page xv currently states, in part, as follows:

SECTION PAGE 3/4 8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICESB 3/4 8-3 TS Index Page xv will state, in part, as follows:

SECTION PAGE 3/4 8.4 DELETEDB 3/4 8-3 TS 3.3.3.6 - Accident Monitoring Instrumentation TS LCO 3.3.3.6 currently states, in part, as follows:

a. With the number of OPERABLE accident monitoring instrumentation channels, except In Core Thermocouples and Reactor Vessel Level, less than the Total Required Number of Channels requirements shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days, 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 at least 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 />.
b. With the number of OPERABLE accident monitoring instrumentation channels, except the radiation monitors, the Pressurizer Safety Valve Position Indicator, the Reactor Coolant System Subcooling Margin Monitor, In Core Thermocouples or Reactor Vessel Level, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 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 at least 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 />.
d. With the number of OPERABLE accident monitoring instrumentation channels for In Core Thermocouples or Reactor Vessel Level less than the total required number of channels shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 30 days or submit a Special Report, pursuant to specification 6.9.2, within the following 14 days from the time the action is required. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to operable status.
e. With the number of OPERABLE accident monitoring instrument channels for In Core Thermocouples or Reactor Vessel Level less than the minimum channels OPERABLE requirement of Table 3.3-10, either restore one channel to OPERABLE status within 7 days or be in at least HOT STANDBY in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least 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 />.

TS LCO 3.3.3.6 will state, in part, as follows:

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Required Number of Channels requirements shown in Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 30 days, or submit a Special Report pursuant to specification 6.9.2 within the following 14 days from the time the action is required. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channel(s) to operable status.
b. With the number of OPERABLE accident monitoring instrumentation channels, except the radiation monitors, the Pressurizer Safety Valve Position Indicator, or the Reactor Coolant System Subcooling Margin Monitor, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel(s) to OPERABLE status within 7 days 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 at least 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 />.
d. DELETED.
e. DELETED.

Additionally, the licensee proposed to add a NOTE to TS LCO 3.3.3.6, prior to the action statements, as follows:

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

A separate ACTION entry is allowed for each INSTRUMENT listed in Table 3.3-10.

TS 3.8.4, Electrical Equipment Protective Devices - Containment Penetration Conductor Overcurrent Protective Devices The licensee proposed to delete current TS Section 3/4.8.4, which contains limiting condition for operation (LCO) 3.8.4.1, Surveillance Requirement (SR) 4.8.4.1, LCO 3.8.4.2, and SR 4.8.4.2.

The licensee proposed to delete these LCOs and SRs from TSs and relocate them to the licensee-controlled plant procedure PLP-106, Technical Specification Equipment List Program.

TS 3.9.2 - Instrumentation TS 3.9.2 TS LCO, currently states, as follows:

As a minimum, two Source Range Neutron Flux Monitors* shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

TS 3.9.2 TS LCO, will state, as follows:

As a minimum, two Source Range Neutron Flux Monitors* shall be OPERABLE, each with continuous visual indication in the control room.

Additionally, the licensee proposed to modify the Note at the bottom on TS 3/4 9-3 (from the

The Note currently states, as follows:

  • A Wide Range Neutron Flux Monitor may be substituted for one of the Source Range Neutron Flux Monitors provided the OPERABLE Source Range Neutron Flux Monitor is capable of providing audible indication in the containment and in the control room.

The Note will state, as follows:

  • Wide Range Neutron Flux Monitors may be substituted for Source Range Neutron Flux Monitors provided the two required OPERABLE monitors (Source Range Neutron Flux Monitors and/or Wide Range Neutron Flux Monitors) are located on opposite sides of the core.

2.3 Applicable Regulatory Requirements and Guidance Regulations The NRC staff identified the following regulatory requirements as being applicable to the proposed amendment:

The NRC's regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." This regulation requires that the TSs include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in plant TSs.

Section 50.36(c)(2)(i) of 10 CFR states, in part:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow remedial action permitted by the technical specifications until the condition can be met.

Section 50.36(c)(2)(ii) of 10 CFR states that a technical specification LCO of a nuclear reactor must be established for each item meeting one or more of the following criteria:

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

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

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

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

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

Regulatory Guidance The NRC staff considered the following guidance for its review:

NUREG-1431, Volume 1, Revision 4, "Standard Technical Specifications, Westinghouse Plants," (Reference 7).

Regulatory Guide (RG) 1.97, Revision 3, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," (Reference 8).

3.0 TECHNICAL EVALUATION

3.1 TS 3.3.3.6 - Accident Monitoring Instrumentation In order to ensure the availability of information on selected plant parameters to monitor and assess these variables following an accident, HNP TS 3.3.3.6 places operability requirements for accident monitoring channels in Power Operation, Startup and Hot Standby (Modes 1, 2 and 3). HNPs TS Table 3.3-10, Accident Monitoring Instrumentation, lists all the instruments for which HNP TS 3.3.3.6 is applicable and the requirement for the total number of channels, and the minimum channels operable for each of these instruments.

Action Statement a HNP TS 3.3.3.6 Action Statement a allows 7 days for the restoration of inoperable channel(s) when the number of operable PAM instrumentation channels is less than the total required per TS Table 3.3-10. If the PAM instrumentation channel(s) cannot be restored to operable status within the time allowed, a plant shutdown is required.

As part of this amendment request, the licensee proposes to revise Action Statement a to allow 30 days instead of 7 days for the restoration of the inoperable channel(s). Also, the licensee proposes to replace the shutdown requirement with a requirement to submit to the NRC within 14 days of exceeding the completion time, a Special Report discussing the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrument to operable status. In Section 2.4 of the LAR, the licensee stated that the revision of Action Statement a completion time is consistent with the completion time in LCO 3.3.3 of the Improved Standard Technical Specifications (ISTS), and the replacement of the shutdown requirement with a Special Report is consistent with the required action in LCO 3.3.3 of the ISTS.

During the development of the ISTS (NUREG -1431), it was determined that the 7-day completion time was overly restrictive. NUREG-1431, TS 3.3.3 (HNPs TS 3.3.3.6 equivalent) extended the completion time for one inoperable instrument channel from 7 days to 30 days.

NUREG-1431, Bases Section B 3.3.3, states:

The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

NUREG-1431, TS 3.3.3 requires a written report to be submitted, per NUREG-1431 TS 5.6.5, to the NRC that discuses an alternate method of monitoring, the cause of the inoperability, and the plans for restoration of the channel(s). NUREG-1431, Bases Section B 3.3.3, states:

This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability and given the likelihood of unit conditions that would require information provided by this instrumentation.

The NRC staff reviewed HNPs TS 3.3.3.6 and NUREG-1431, TS 3.3.3 and found the following similarities:

The LCO for both NUREG-1431 ISTS and HNPs TS specifies the operability of the PAM instrumentation channels ((NUREG-1431 functions) listed in Table 3.3-10 (NUREG-1431 Table 3.3.3-1). These instrumentation monitors Type A and Category I variables as defined in RG 1.97.

The applicability for both NUREG-1431 ISTS and HNPs TS is Modes 1, 2 and 3.

The surveillance requirements (SRs) for both NUREG-1431 ISTS and HNPs TS require the performance of Channel Check and Channel Calibration of each of the PAM instrumentation channels in accordance with the Surveillance Frequency Program (HNPs Table 4.3-7 Accident Monitoring Instrumentation Surveillance Requirements).

The NRC staff, therefore, agrees that the LCO and SRs are reasonably similar and consistent with the applicable ISTS in NUREG-1431, TS 3.3.3. Additionally, as stated above, the 30-day completion time is based on operating experience and takes into account the remaining operable channel, the passive nature of the instrument, and the low probability of an event requiring PAM instrumentation during this period. Also, the replacement of the shutdown requirement with a requirement to submit a special report is appropriate because alternative actions are identified, and the likelihood of an event requiring information from this instrumentation is low. Therefore, the NRC staff finds the licensees proposal to revise Action Statement a to allow 30 days instead of 7 days for the restoration of the inoperable channel(s) and to replace the shutdown requirement with a requirement to submit a special report, to be acceptable.

Action Statement b HNP TS 3.3.3.6 Action Statement b allows 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the restoration of inoperable channel(s) when the number of operable PAM instrumentation channels is less than the minimum required per TS Table 3.3-10. If the PAM instrumentation channel(s) cannot be restored to operable status within the time allowed, a plant shutdown is required.

As part of this amendment request, the licensee proposes to revise Action Statement b to allow 7 days instead of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the restoration of the inoperable channel(s). The licensee is not proposing to modify the shutdown requirement, which will remain part of Action Statement

b. In Section 2.4 of the LAR the licensee stated that the revision of the Action Statement b completion time is consistent with the required action in LCO 3.3.3 of the ISTS.

NUREG-1431, TS 3.3.3 extended the completion time for two inoperable instrument channels from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days. NUREG-1431, Bases Section B 3.3.3, states:

The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur.

Based on its review, the NRC staff agrees that the LCO and SRs are reasonably similar and consistent with the applicable ISTS in NUREG-1431, TS 3.3.3. Additionally, as stated above, the 7 days completion time is based on the relatively low probability of an event requiring PAM instrument operation and the shutdown requirement will not be modified. Therefore, the NRC staff finds the licensees proposal to revise Action Statement b to allow 7 days instead of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the restoration of the inoperable channel(s), to be acceptable.

Action Statements d and e In HNP license amendment 110, dated May 30, 2002 (Reference 9), the licensee added action statements d and e to HNPs TS 3.3.3.6. Action Statement d allows 30 days for the restoration of inoperable In-Core Thermocouples and Reactor Vessel Level instrumentation channel(s) when the number of operable instrumentation channels is less than the total required per TS Table 3.3-10, Accident Monitoring Instrumentation. If the instrumentation channel(s) cannot be restored to operable status within the time allowed, action statement d requires the submittal to the NRC of a Special Report discussing the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrument to operable status. Action Statement e allows 7 days for the restoration of inoperable In Core Thermocouples and Reactor Vessel Level instrumentation channel(s) when the number of operable instrumentation channels is less than the minimum required per TS Table 3.3-10. If the instrumentation channel(s) cannot be restored to operable status within the time allowed, action statement e requires a plant shutdown. As part of HNP license amendment 110, the licensee also excluded In-Core Thermocouples and Reactor Vessel Level instrumentation from the restoration time requirements of Action Statements a and b.

As part of the March 12, 2020, LAR, the licensee proposes to delete action statements d and e and to remove the exclusion of the In-Core Thermocouples and Reactor Vessel Level instrumentation from action statements a and b. In Section 2.4 of the LAR the licensee stated that In aligning with Actions a and b for ISTS 3.3.3, the actions specific to accident monitoring instrumentation channels for In-Core Thermocouples and Reactor Vessel Level (i.e., Actions d and e) are no longer necessary.

The NRC staff reviewed the proposed action statement a and b, and the current action statement d and e. Since the restoration time for In-Core Thermocouples and Reactor Vessel Level instrumentation has already been extended in the previous HNP license amendment 110, and the exclusion of the In-Core Thermocouples and Reactor Vessel Level instrumentation will be removed, the requirements of action statement a and b will be identical to the requirements of action statement d and e, respectively.

The NRC staff agrees that action statements d and e will no longer be necessary.

Therefore, the licensee's proposal to delete action statements d and e and to remove the exclusion of the In-Core Thermocouples and Reactor Vessel Level instrumentation from action statements a and b is acceptable.

New NOTE As part of the March 12, 2020, LAR, the licensee proposes to add a new note to HNPs TS 3.3.3.6. The new note would allow the action statement of LCO 3.3.3.6 to be entered independently for each instrument listed in TS Table 3.3-10. The allowed outage time for each inoperable channel of an instrument would be tracked separately, starting from the time the action was entered for that particular instrumentation channel. In Section 3.0 of the LAR the licensee stated that, This change clarifies the application of the actions for multiple inoperable instruments, and is consistent with the note in the ISTS.

NUREG-1431, TS 3.3.3 includes a Note to allow separate condition entry for each function (HNP instrument). NUREG-1431, Bases Section B 3.3.3, states:

A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

Based on its review, the NRC staff agrees that the HNP LCO and SRs are reasonably similar to and consistent with the applicable ISTS in NUREG-1431, TS 3.3.3. Additionally, as stated above, the new note will clarify the application of completion times rules. Therefore, the NRC staff finds the licensees proposal to add a note to HNPs TS 3.3.3.6 to allow a separate action entry for each instrument listed in TS Table 3.3-10, acceptable.

3.2 TS 3.9.2 - Instrumentation In order to ensure that the reactor remains subcritical during Refueling (Mode 6) and to prevent boron dilution, HNP TS 3.9.1, Boron Concentration places limitations on the boron concentration of the Reactor Coolant System (RCS), the refueling canal, and the refueling cavity. One method to preclude boron dilution events in PWRs is an analysis, which assumes a maximum unborated water flow and determines if there is adequate time for operator action to mitigate the event. Another method of precluding boron dilution events in PWRs during refueling is to implement an administrative process that ensures that all sources of unborated water are isolated from the reactor vessel. When the latter method is used, no further analyses are required. In the HNP UFSAR, Chapter 15, the licensee discusses the boron dilution event analysis and states that No analysis was performed for a boron dilution event in Mode 6, since administrative controls are in place to prevent an uncontrolled boron dilution while the unit is in the refueling mode. These administrative controls over the required valves to prevent boron dilution are included in LCO 3.9.1.b (Table 3.9-1).

HNP TS 3.9.2 requires a minimum of two Source Range Neutron Flux Monitors (SRNFM) to be operable while in Mode 6 in order to detect the changes to the reactivity condition of the core, which could result in a loss of the required shutdown margin. Both Neutron Flux Monitors are required to provide a continuous visual indication in the control room, while one is required to provide audible and visual indications in containment and in the control room.

As part of the March 20, 2020, LAR, licensee proposes to remove the requirement that one of the two Neutron Flux Monitors provide audible indication in the containment and control room.

In Section 2.3 of the LAR, the licensee stated that this change is in alignment with NUREG-1431, ISTS 3.9.3, Nuclear Instrumentation, for plants that isolate all boron dilution paths.

NUREG-1431, TS 3.9.3 (HNPs TS 3.9.2 equivalent) requires that two source range neutron flux monitors be operable and provides a bracketed option for a requirement for operability of one source range audible [alarm] [count rate]. NUREG-1431, Bases Section B 3.9.3, states:

Bracketed options are provided for source range OPERABILITY requirements to include audible alarm or count rate function. These options apply to plants that assume a boron dilution event that is mitigated by operator response to an audible indication. For plants that isolate all boron dilution paths (per LCO 3.9.2),

the source range OPERABILITY includes only a visual monitoring function.

The Neutron Flux Monitors audible alarm provides operators with a prompt and definite indication of a boron dilution event, for which quick detection assures enough time is available to manually isolate the unborated water source and stop the dilution prior to loss of shutdown margin. As discussed above, HNP LCO 3.9.1.b (Table 3.9-1) provides administrative controls that isolate all boron dilution paths during Mode 6, and are equivalent to NUREG-1431s LCO 3.9.2. Due to all boron dilution paths already being isolated in Mode 6, with a Technical Specification that administratively controls the isolation of these dilution paths, no further isolation actions by the operator would be needed to be taken, solely due to the existence of an audible alarm. Without the need for operator action in response to the audible alarm, the TS 3.9.2 requirement for one of the two Neutron Flux Monitors to provide audible indication in the containment and control room is not necessary, consistent with NUREG-1431, TS 3.9.3.

Therefore, the NRC staff finds the licensees proposal to remove the TS 3.9.2 requirement that one of the two Neutron Flux Monitors provide an audible indication in the containment and control room, acceptable.

In HNP license amendment 105, dated September 10, 2001 (Reference 10), the licensee added a TS 3.9.2 Note that allows for a Wide Range Neutron Flux Monitor (WRNFM) to be substituted for one of the SRNFMs as long as the remaining SRNFM can provide an audible indication. In connection to the removal of the audio indication requirement, the licensee proposes an expansion to this allowance, to modify the note to permit both SRNFMs to be substituted with WRNFMs provided that the Neutron Flux Monitors are located on opposite sides of the core. In Section 2.3 of the LAR the licensee stated that SRNFM inoperability has previously delayed refueling activities. During Mode 6 WRNFMs would be used if both SRNFMs were inoperable for maintenance activities. However, the WRNFMs do not have the audible alarm capability.

In the safety evaluation for HNP license amendment 105, the NRC staff discussed the differences between the SRNFMs and WRNFMs:

The WRNFM provides the same level of quality assurance, redundancy, and necessary display range as the SRNFM. Although the WRNFM detector has a neutron sensitivity of 1.0 counts per second/neutron flux(nv), compared to the SRNFM detector sensitivity of 10 cps/nv, the difference in sensitivity is acceptable since the purpose of the detectors is to monitor trends in neutron flux, which can be accomplished with the visible indication on the WRNFM channel. A WRNFM channel (source range indicator) is required to be operable per TS 3/4.3.3.5 Remote Shutdown System in Modes 1-3. The function of the WRNFM for TS 3/4.3.3.5 is to monitor the core reactivity in a shutdown condition, which is also the function of the SRNFM in Mode 6. Therefore, there are no potential adverse consequences of using a WRNFM in place of an SRNFM.

In its October 2, 2020, response to the NRC staffs requests for additional information (RAIs)

(Reference 11), the licensee provided further information on the performance of the WRNFM versus the SRNFM and stated that the indicator range for the WRNFMs starts at 10-1 cps

[counts per second] and overlaps the indicator range of the SRNFMs, which starts at 100 cps.

The licensee stated that it reviewed data from a previous refueling outage in order to compare the changes in neutron activity that both the SRNFM and the WRNFM were indicating. The licensee concluded that Based on the ability of the WRNFMs to detect changes in neutron activity that coincide with those detected by the SRNFMs during core reload, the WRNFMs have demonstrated that they are as capable as the SRNFMs to detect the onset of neutron activity events in the core during Mode 6.

Based on the information above, the NRC staff finds that the SRNFMs and the WRNFMs are both capable of detecting and indicating neutron flux levels over the shutdown range required during Mode 6 operation. As described above, the WRNFMs do not provide the audible count rate feature; however, as discussed above, the audible count rate in the containment and control room is not necessary in Mode 6, provided administrative controls are in place to ensure all sources of unborated water have been isolated. The licensee stated that continuous visual indication will be provided for both the SRNFMs and the WRNFMs in the containment and control room. Therefore, the NRC staff finds the licensee's proposal to modify the TS 3.9.2 Note to permit both SRNFMs to be substituted with WRNFMs, as acceptable.

3.3 TS 3/4.8.4 - Electrical Equipment Protective Devices - LCOs 3.8.4.1 and 3.8.4.2 LCO 3.8.4.1 addresses the containment penetration conductor overcurrent protective devices that minimize the damage from a fault in a component inside containment, or in cabling which penetrates containment. This prevents an electrical penetration from being damaged in such a way that the containment structure could be breached.

LCO 3.8.4.2 addresses the bypassing of the thermal overload protection for certain MOVs during accident conditions, minimizing the potential that the actuation of a thermal overload device could prevent a vital piece of equipment from performing its intended function.

10 CFR 50.36(c)(2)(ii) provides four criteria used to determine items that must have an LCO established in the TS. The NRC staff evaluated the deletion of HNP LCO 3.8.4.1 and LCO 3.8.4.2 to determine if the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs overload protection during accident conditions meet the four criteria for inclusion in the HNP TS.

Criterion 1 of 10 CFR 50.36(c)(2)(ii), addresses instrumentation installed to detect excessive reactor coolant system (RCS) leakage.

The containment penetration conductor overcurrent protective devices and the bypassing of the thermal overload protection for certain MOVs during accident conditions are not used to detect excessive RCS leakage, which is an indication of a significant degradation of the reactor coolant pressure boundary.

Therefore, the NRC staff finds that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection during accident conditions do not meet 10 CFR 50.36(c)(2)(ii), Criterion 1.

Criterion 2 of 10 CFR 50.36(c)(2)(ii), captures those process variables that have initial values assumed in the design basis accident and transient analyses and that are monitored and controlled during power operation. This criterion also includes active design features (e.g., high-pressure/low-pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients.

In Enclosure 1, Section 3.0, Technical Evaluation, of the March 12, 2020, LAR, the licensee states, in part:

Both the containment penetration conductor overcurrent protective devices and the bypassing of the MOV thermal overload protection for certain valves during accident situations help preserve the assumptions of the accident analysis by enhancing proper equipment operation. However, they are not process variables, design features, or operating restrictions that are initial conditions 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.

The NRC staff notes that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection are not relied upon in the HNP FSAR Chapter 15 design basis accident analyses. Therefore, the NRC staff finds that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection during accident conditions do not meet 10 CFR 50.36(c)(2)(ii), Criterion 2.

Criterion 3 of 10 CFR 50.36(c)(2)(ii), captures those structures, systems, and components that are part of the primary success path of the safety analysis (the actions required to mitigate the consequences of the design basis accidents and transients). The primary success path of a safety analysis consists of the combinations and sequences of equipment needed to operate so that the plant responses to the design basis accident and the transients limit the consequences of these events within the appropriate acceptance criteria. Also captured by this criterion are those support and actuation systems that are necessary in the primary success path, but this criterion does not include backup and diverse equipment.

In Enclosure 1, Section 3.0 of the March 12, 2020, LAR, the licensee states, in part:

The containment penetration conductor overcurrent protective devices are installed to minimize the damage from a fault in a component inside containment, or in conductors which penetrate containment. The MOV thermal overload protection is installed to provide equipment protection, where bypassing the thermal overload protection of certain MOVs during accident conditions minimizes the potential that the actuation of a thermal overload device could prevent a vital piece of equipment from performing its intended function.

However, neither the containment penetration conductor overcurrent protective devices nor the MOV thermal overload protection are structures, systems, or components that are part of the primary success path and which function or actuate 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.

The NRC staff notes that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection are not required to mitigate the consequences of design basis accidents and transients described in the HNP FSAR, and thus, are not part of the primary success path or of the support or actuation systems necessary in the primary success path. Therefore, the NRC staff finds that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection for certain valves during accident conditions do not meet 10 CFR 50.36(c)(2)(ii),

Criterion 3.

Criterion 4 of 10 CFR 50.36(c)(2)(ii), captures those structures, systems, and components that operating experience and probabilistic risk assessment has shown to be significant to the public health and safety.

In Enclosure 1, Section 3.0 of the March 12, 2020, LAR, the licensee states, in part:

The containment penetration conductor overcurrent protective devices and the MOV thermal overload protection are not structures, systems, or components that operating experience or probabilistic safety assessment has shown to be significant to the public health and safety. The Maintenance Rule (10 CFR 50.65) does not require these types of protections to be monitored for availability.

Additionally, a review of industry operating experience did not produce any examples where either protection has had a significant adverse effect on public health and safety.

Based on the above statements, the NRC staff finds that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection during accident conditions do not meet 10 CFR 50.36(c)(2)(ii), Criterion 4.

In summary, as discussed above, the containment penetration conductor overcurrent protective devices in LCO 3.8.4.1 and the bypassing of the thermal overload protection for certain MOVs under accident conditions in LCO 3.8.4.2 do not meet any of the criteria of 10 CFR 50.36(c)(2)(ii) for items that must have an LCO established in TS. In addition, the NUREG-143, Section 3.8, Electrical Power Systems, do not contain LCOs for containment penetration conductor overcurrent protective devices and bypassing of MOV thermal overload protection during accident conditions.

Therefore, the NRC staff concludes that the HNP TS Section 3/4.8.4, LCO 3.8.4.1 and LCO 3.8.4.2 requirements can be relocated from the TS since the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection under accident conditions are not required to have LCOs established in the HNP TS according to 10 CFR 50.36(c)(2)(ii).

3.4 TS 3/4.8.4 - Electrical Equipment Protective Devices - SRs 4.8.4.1 and 4.8.4.2 In accordance with 10 CFR 50.36(c)(3), TS SR 4.8.4.1 and SR 4.8.4.2 were provided to ensure, in part, that LCO 3.8.4.1 and LCO 3.8.4.2, respectively, will be met.

Since the licensee proposed to remove LCO 3.8.4.1 and LCO 3.8.4.2 from TS, as discussed above, the NRC staff concludes that their respective SRs, SR 4.8.4.1 and SR 4.8.4.2, are no longer needed to meet 10 CFR 50.36(c)(3), and as such, can be relocated from the TS.

3.5 TS Index For the TS Index, the licensee proposed corresponding changes to TS sections that will be deleted from HNP TS on Page xi (3/4.8.4, Tables 3.8-1 and 3.8-2) and Page xv (3/4.8.4). The NRC staff finds that these editorial changes clarify the contents of the HNP TSs, but do not affect any technical requirements. Therefore, the staff finds the changes to the TS Index are acceptable.

3.6 Technical Evaluation Summary The NRC staffs evaluation determined that the proposed TS 3.3.3.6 changes retain operability requirements for accident monitoring channels that will ensure the availability of information on selected plant parameters to monitor and assess these variables following an accident.

Furthermore, the NRC staff determined that the proposed TS 3.9.2 changes retain the limitations on the boron concentration of the RCS, the refueling canal, and the refueling cavity that will ensure the reactor remains subcritical during Mode 6 and that prevent boron dilution.

The revised HNP TS therefore continues to satisfy the criteria of 10 CFR 50.36(c)(2)(i).

The NRC staff determined that the containment penetration conductor overcurrent protective devices and the bypassing of certain MOVs thermal overload protection under accident conditions are not required to have LCOs established in the HNP TS according to 10 CFR 50.36(c)(2)(ii). The NRC staff concludes that the TS 3/4.8.4, LCO 3.8.4.1 and LCO 3.8.4.2 requirements can therefore be relocated from the TS to licensee-controlled procedure PLP-106.

Additionally, since the licensee proposed to remove LCO 3.8.4.1 and LCO 3.8.4.2 from TS, the NRC staff concludes that their respective SRs, SR 4.8.4.1 and SR 4.8.4.2, are no longer needed to meet 10 CFR 50.36(c)(3), and therefore, can be relocated from the TS to licensee-controlled procedure PLP-106.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on December 9, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20344A419). The State of North Carolina official responded on December 10, 2020, with no comments (ADAMS Accession No. ML20345A305).

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration in the Federal Register on August 25, 2020 (85 FR 52370), and there has been no public comment on such finding.

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

7.0 REFERENCES

1.

Maza, Kim E., Duke Energy Progress, LLC., letter to U.S. Nuclear Regulatory Commission, License Amendment Request to Revise Technical Specifications Related to Accident Monitoring Instrumentation, Refueling Operations Instrumentation, and Electrical Equipment Protective Devices, March 12, 2020 (ADAMS Accession No. ML20072M618).

2.

Maza, Kim E., Duke Energy Progress, LLC., letter to U.S. Nuclear Regulatory Commission, Supplement to License Amendment Request to Revise Technical Specifications Related to Accident Monitoring Instrumentation, Refueling Operations Instrumentation, and Electrical Equipment Protective Devices, June 22, 2020 (ADAMS Accession No. ML20174A639).

3.

Maza, Kim E., Duke Energy Progress, LLC., letter to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications Related to Accident Monitoring Instrumentation, Refueling Operations Instrumentation, and Electrical Equipment Protective Devices, October 2, 2020 (ADAMS Accession No. ML20276A317).

4.

Duke Energy, Shearon Harris Nuclear Power Plant, Unit 1 - Shearon Harris Nuclear Plant, Unit 1, Amendment 63 to Final Safety Analysis Report, Chapter 7, Instrumentation and Controls, (ADAMS Accession No. ML20147A024), Part of Submittal of Updated Final Safety Analysis Report (Amendment 63), Technical Specification Bases Revision, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes, May 15, 2020 (ADAMS Package No. ML20147A016).

5.

U.S. Nuclear Regulatory Commission, Standard Technical Specifications:

Westinghouse Plants - Bases, NUREG-1431, Volume 2, Revision 4, April 2012 (ADAMS Accession No. ML12100A228).

6.

Duke Energy, Shearon Harris Nuclear Power Plant, Unit 1 - Shearon Harris Nuclear Plant, Unit 1, Amendment 63 to Final Safety Analysis Report, Chapter 8, Electric Power, (ADAMS Accession No. ML20147A025), Part of Submittal of Updated Final Safety Analysis Report (Amendment 63), Technical Specification Bases Revision, Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes, May 15, 2020 (ADAMS Package No. ML20147A016).

7.

U.S. Nuclear Regulatory Commission, "Standard Technical Specifications:

Westinghouse Plants - Specifications," NUREG-1431, Volume 1, Revision 4, April 2012 (ADAMS Accession No. ML12100A222).

8.

U.S. Nuclear Regulatory Commission, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Regulatory Guide (RG) 1.97, Revision 3, May 1983 (ADAMS Accession No. ML003740282).

9.

Goshen, John M., U.S. Nuclear Regulatory Commission to Scarola, James, Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment RE: Emergency Change to Technical Specifications Regarding Post-Accident Monitoring Instrumentation, May 30, 2002 (ADAMS Package Accession No. ML021560583).

10.

Laufer, Richard J., U.S. Nuclear Regulatory Commission to Scarola, James, Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Use of Alternate Source Range Neutron Flux Monitors in Refueling Mode (TAC No. MB0783), September 10, 2001 (ADAMS Package Accession No. ML012770022).

11.

Electronic correspondence from Mahoney, Michael, U.S. Nuclear Regulatory Commission to Zaremba, Art, Duke Energy, Request for Additional Information -

Shearon Harris Nuclear Power Plant, Unit 1 - Revise TS Related to Accident Monitoring and RFO Instrumentation (EPID L-2020-LLA-0045), September 3, 2020 (ADAMS Accession No. ML20247J382).

Principal Contributors: S. Basturescu, NRR A. Foli, NRR M. Mahoney, NRR Date of Issuance: March 16, 2021

ML21035A132

  • by memorandum **via e-mail OFFICE DORL/LPL2-2/PM**

DORL/LPL2-2/LA**

DSS/EICB/BC*

NAME MMahoney BAbeywickrama MWaters DATE 02/08/2021 02/04/2021 10/28/2020 OFFICE DEX/EEOB/BC*

DSS/STSB/BC**

DSS/SNSB/BC**

NAME BTitus VCusumano SKrepel DATE 08/14/2020 01/19/2021 01/28/2021 OFFICE OGC - NLO NLO**

DORL/LPL2-2/BC**

DORL/LPL2-2/PM**

NAME AGhosh UShoop MMahoney DATE 02/24/2021 03/16/2021 03/16/2021

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