ML22041A010

V.C. Summer 1, Issuance of Amendment Nos. 283 (Millstone), 291 and 274 (North Anna), and 221 (Summer) to Revise TSs to Adopt TSTF-569 Revision of Response Time Testing Definition

ML22041A010
Person / Time
Site:	Millstone, Summer, North Anna
Issue date:	03/01/2022
From:	Geoffrey Miller Plant Licensing Branch II
To:	Stoddard D Dominion Energy Co
Miller G, NRR/DORL/LPL2-1, 415-2481
References
EPID L-2019-LLA-0186
Download: ML22041A010 (31)

Text

March 1, 2022 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION UNIT NO. 3, NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, AND VIRGIL C. SUMMER NUCLEAR STATION UNIT NO.

1 ISSUANCE OF AMENDMENT NO(S). 283 (MILLSTONE), 291 AND 274 (NORTH ANNA), AND 221 (SUMMER) TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-569, REVISION OF RESPONSE TIME TESTING DEFINITION (EPID L-2019-LLA-0186)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 283 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3, Amendment Nos. 291 and 274 to Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station (North Anna), Unit Nos. 1 and 2, respectively, and Amendment No. 221 to Renewed Facility Operating License No. for the Virgil C. Summer Nuclear Station, Unit No. 1. These amendments are in response to your application dated October 7, 2021.

The amendments revise the respective Technical Specifications (TSs) to adopt Technical Specification Task Force (TSTF) Traveler TSTF-569, Revise Response Time Testing Definition, to revise the TS definitions for the engineered safety feature response time and reactor trip system response time.

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

Sincerely,

/RA/

G. Edward Miller, Project Manager Plant Licensing Branch 2-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338, 50-339, 50-423, and 50-395

Enclosures:

1. Amendment No. 283 to NPF-49

2. Amendment No. 291 to NPF-4

2. Amendment No. 274 to NPF-7

4. Amendment No. 221 to NPF-12

5. Safety Evaluation cc: Listserv DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL DOCKET NO. 50-423 MILLSTONE POWER STATION UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 283 Renewed License No. NPF-49

1.

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

A.

The application for amendment by Dominion Energy Nuclear Connecticut, Inc.

(DENC, the licensee), dated October 7, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

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-49 are hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 283 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Operation

Attachment:

Changes to Renewed Facility Operating License No. NPF-49 and Technical Specifications Date of Issuance: March 1, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.03.01 13:28:06 -05'00'

ATTACHMENT TO MILLSTONE POWER STATION, UNIT NO. 3 LICENSE AMENDMENT NO. 283 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-49, page 4 NPF-49, page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1-3 1-3 1-5 1-5 (2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 283 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

DENC shall not take any action that would cause Dominion Energy, Inc.

or its parent companies to void, cancel, or diminish DENC's Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.

(4)

Immediately after the transfer of interests in MPS Unit No. 3 to DNC*, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC* would then hold, be at a level no less than the formula amount under 10 CFR 50.75.

(5)

The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC* is effected and thereafter is subject to the following:

(a)

The decommissioning trust agreement must be in a form acceptable to the NRC.

(b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Energy, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(c)

The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(d)

The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

• On May 12, 2017, the name Dominion Nuclear Connecticut, Inc. changed to Dominion Energy Nuclear Connecticut, Inc.

Renewed License No. NPF-49 Amendment No. 270-282 283

MILLSTONE - UNIT 3 DEFINITIONS DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microCurie/gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

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

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

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

LEAKAGE 1.16 LEAKAGE shall be:

1.16.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or Amendment No. 84, 87, 126, 187, 216, 220, 238, 246, 283 1-3

DEFINITIONS OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYS I CS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (l) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

1.22 DELETED PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors sha11 be used for computing the average.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3650 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

MILLSTONE - UNIT 3 1-5 Amendment No. 6<:f, -l-8-1, 8-8, >>,

1/4, 283

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 291 Renewed License No. NPF-4

1.

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

A.

The application for amendment by Virginia Electric and Power Company et al.,

(the licensee) dated October 7, 2021 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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 paragraph 2.C (2) of Renewed Facility Operating License No. NPF-4, as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A, as revised through Amendment No. 291, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Operation

Attachment:

Changes to Renewed Facility Operating License No. NPF-4 and Technical Specifications Date of Issuance: March 1, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.03.01 13:28:59 -05'00'

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 274 Renewed License No. NPF-7

1.

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

A.

The application for amendment by Virginia Electric and Power Company et al.,

(the licensee) dated October 7, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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 paragraph 2.C (2) of Renewed Facility Operating License No. NPF-7, as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 274, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Operation

Attachment:

Changes to Renewed Facility Operating License No. NPF-7 and Technical Specifications Date of Issuance: March 1, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.03.01 13:29:36 -05'00'

ATTACHMENT TO NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 LICENSE AMENDMENT NO. 291 RENEWED FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 LICENSE AMENDMENT NO. 273 RENEWED FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-4, page 3 NPF-4, page 3 NPF-7, page 3 NPF-7, page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1-3 1.1-3 1.1-5 1.1-5

NORTH ANNA - UNIT 1 Renewed License NPF-4 Amendment No. 291 (2) Pursuant to the Act and 10 CFR Part 70, VEPCO to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material, without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or component; and (5) Pursuant to the Act and 10 CFR Parts 30 and 70, VEPCO to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 VEPCO is authorized to operate the North Anna Power Station, Unit No. 1, at reactor core power levels not in excess of 2940 megawatts (thermal).

(2) Technical Specifications Technical Specifications contained in Appendix A, as revised through Amendment No. 291 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

NORTH ANNA - UNIT 2 Renewed License NPF-7 Amendment No. 274 (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material, without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or component; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations as 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 VEPCO is authorized to operate the facility at steady state reactor core power levels not in excess of 2940 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 274 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the insurance of the condition or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission:

a. If VEPCO plans to remove or to make significant changes in the normal operation of equipment that controls the amount of radioactivity in effluents from the North Anna Power Station, the

1.1 Definitions DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE North Anna Units 1 and 2 Definitions 1.1 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-13lm, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionucl ides in Air, Water, and Soil. 11 The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (continued) 1.1-3 Amendments 291/274

Definitions 1.1 1.1 Definitions North Anna Units 1 and 2 1.1-5 Amendments 291/274 OPERABLEOPERABILITY (continued) component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

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

These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR;

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

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT RATIO (QPTR)

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2940 MWt.

REACTOR TRIP SYSTEM (RTS) RESPONSE TIME The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.

SHUTDOWN MARGIN (SDM)

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

(continued)

DOMINION ENERGY SOUTH CAROLINA, INC.

SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 221 Renewed License No. NPF-12

1.

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

A.

The application for amendment to the Virgil C. Summer Nuclear Station, Unit No. 1 (the facility), Renewed Facility Operating License No. NPF-12, filed by the Dominion Energy South Carolina, Inc. (the licensee), dated October 7, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering public health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations as set forth in 10 CFR Chapter I; D.

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

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

2.

Accordingly, the license is hereby amended by a page change 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-12 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 221, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. Dominion Energy South Carolina, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: March 1, 2022 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2022.03.01 13:30:29 -05'00'

ATTACHMENT TO VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 LICENSE AMENDMENT NO. 221 RENEWED FACILITY OPERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-12, page 3 NPF-12, page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1-3 1-3 1-5 1-5

Renewed Facility Operating License No. NPF-12 Amendment No. 221 (3)

SCE&G, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage amounts required for reactor operation, as described in the Final Safety Analysis Report, as amended through Amendment No. 33; (4)

SCE&G, pursuant to the Act and 10 CFR Part 30, 40 and 70 to receive, possess and use at any time byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

SCE&G, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus of components; and (6)

SCE&G, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as m[a]y be produced by the operation of the facility.

C.

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

(1)

Maximum Power Level SCE&G is authorized to operate the facility at reactor core power levels not in excess of 2900 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to this renewed license.

The preoccupation tests, startup tests and other items identified in to this renewed license shall be completed as specified. is hereby incorporated into this renewed license.

(2)

Technical Specifications and Environmental Protection Plant The Technical Specifications contained in Appendix A, as revised through Amendment No. 221, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRG or the components have been evaluated in accordance with an NRG approved methodology.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

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

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

Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

SUMMER - UNIT 1 1-3 Amendment No. 146,179, 221

DEFINITIONS PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RA TIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2900 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRG, or the components have been evaluated in accordance with an NRG approved methodology.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50. 73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SLAVE RELAY TEST 1.29 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

1.30 Not Used SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

SUMMER - UNIT 1 1-5 Amendment No. 35, 104, 117, 133, 146, 221

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 291 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-4 AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-7 AMENDMENT NO. 283 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 AMENDMENT NO. 221 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-12 VIRGINIA ELECTRIC AND POWER COMPANY DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

DOMINION ENERGY SOUTH CAROLINA, INC.

NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 MILLSTONE POWER STATION, UNIT NO. 3 VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 DOCKET NOS. 50-338, 50-339. 50-423, AND 50-395

1.0 INTRODUCTION

By application dated October 7, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21280A328), Dominion Energy Connecticut, Inc., Virginia Electric and Power Company (Dominion Energy Virginia), and Dominion Energy South Carolina, Inc. (hereafter referred to as Dominion or the licensee) submitted a license amendment request (LAR) for the Millstone Power Station (Millstone or MPS) Unit 3, North Anna Power Station (North Anna or NAPS) Units 1 and 2, and Virgil C. Summer Nuclear Station (Summer or VCS)

Unit 1. The amendments would revise technical specification (TS) definitions for engineered safety feature (ESF) response time and reactor trip system (RTS) response time that are referenced in surveillance requirements (SRs), hereafter, referred to as response time testing (RTT).

The proposed changes are based on Technical Specifications Task Force (TSTF) traveler TSTF-569, Revision 2, Revise Response Time Testing Definition, dated June 25, 2019 (ADAMS Accession No. ML19176A034). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving TSTF-569, Revision 2, on August 14, 2019 (ADAMS Accession No. ML19176A191). The description of the generic changes and their justification are contained in these two documents.

2.0 REGULATORY EVALUATION

The licensee's request involves adding an option used to satisfy surveillance requirements related to RTT. As described in 10 CFR 50.36(c)(3):

Surveillance requirements 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.

Appendix A to 10 CFR Part 50 provides General Design Criteria (GDC) for nuclear power plants. Plant-specific design criteria are described in the plants Updated Final Safety Analysis Report (UFSAR).

GDC 13, Instrumentation and Control, as discussed in Section 3.1.9 in the North Anna UFSAR, 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 ensure 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.

GDC 21, Protection System Reliability and Testability, as discussed in Section 3.1.17 in the North Anna UFSAR, states:

The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.

Redundancy and independence designed into the protection system shall be sufficient to ensure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated.

The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

Additionally, the NRC staff considered the following guidance in its review of the LAR:

Regulatory Guide (RG) 1.118, Revision 3, Periodic Testing of Electric Power and Protection Systems, April 1995 (ADAMS Accession No. ML003739468), endorses the Institute of Electrical and Electronics Engineers, Inc. (IEEE) Std. 338-1987, IEEE Standard Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, which was approved on March 3, 1988, by the American National Standards Institute.

Branch Technical Position (BTP) 7-17, Guidance on Self-Test and Surveillance Test Provisions, August 23, 2016 (ADAMS Accession No. ML16019A316), states, in part:

Failures detected by hardware, software, and surveillance testing should be consistent with the failure detectability assumptions of the single-failure analysis and the failure modes and effects analysis.

3.0 TECHNICAL EVALUATION

3.1 Description of Response Time Testing The RTS for MPS, NAPS, and VCS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and the reactor coolant system (RCS) pressure boundary during anticipated operational occurrences and to assist the engineering safety feature actuation system (ESFAS) in mitigating accidents. The ESFAS initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary and to mitigate accidents.

The RTT verifies that the individual channel or train actuation response times are less than or equal to the maximum values assumed in the accident analyses. The RTT acceptance criteria are under licensee control. Individual component response times are not modeled in the accident analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment reaches the required functional state (e.g., control and shutdown rods fully inserted in the reactor core).

3.2 Proposed Changes to the Technical Specifications The MPS, NAPS, and VCS Limiting Condition for Operation (LCO) 3.3.2, requires the ESFAS instrumentation for each Function in TS Table 3.3-3 (for MPS and VCS) and Table 3.3.2-1 (for NAPS) "Engineered Safety Feature Actuation System Instrumentation" to be OPERABLE. To assure the LCO is met, surveillance requirement (SR) 4.3.2.2 (for MPS and VCS) and SR 3.3.2.9 (for NAPS), requires the licensee to verify that ESF RESPONSE TIMES are within limits. Similarly, MPS, NAPS, and VCS LCO 3.3.1 requires the RTS instrumentation for each Function in TS Table 3.3-1 (for MPS and VCSO and Table 3.3.1-1 (for NAPS) "Reactor Trip System Instrumentation" to be OPERABLE, and SR 4.3.1.2 (for MPS and VCS) and SR 3.3.1.16 (for NAPS) requires the licensee to verify that RTS RESPONSE TIMES are within limits. The licensee proposed to add a statement to Section 1.0 (for MPS and VCS) and Section 1.1 (for NAPS) of the TS definitions for ESF RESPONSE TIME and RTS RESPONSE TIME, which states acceptable means to measure each response time, and provide an alternative that may be used "[i]n lieu of measurement."

In its application, the licensee stated that it requests adoption of NRC-approved TSTF-569. The only revision of TSTF-569 that is NRC approved is Revision 2. As described in Section 1, Summary Description, of Revision 2 of TSTF-569:

The proposed change revises the definitions to eliminate the requirement for prior NRC review and approval of the response time verification of similar components, while retaining the requirement for the verification to be performed using the methodology contained in Attachment 1, titled, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing. The proposed change will permit licensees to verify the response time of similar component types using the methodology contained in Attachment 1, without obtaining prior NRC approval for each component.

Accordingly, as shown in the LAR, the request would add an additional alternative to the "in lieu of measurement" to measuring ESF RESPONSE TIME and RTS RESPONSE TIME. The additional alternative for ESF RESPONSE TIME would be "[i]n lieu of measurement, response time may be verified for selected components provided... the components have been evaluated in accordance with an NRC approved methodology." Similarly, for RTS RESPONSE TIME, "[i]n lieu of measurement, response time may be verified for selected components provided that...

the components have been evaluated in accordance with an NRC approved methodology."

The application stated that the licensee concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to MPS, NAPS, and VCS and provide the justification for the amendment request. The application identified the requested variations (e.g., identifying the site-specific TS numbers and being committed to the 1971 version of IEEE Standard 338).

3.2.1 Variations

1. The licensee stated that MPS, NAPS, and VCS are committed to the 1971 version of IEEE Standard 338, while TSTF-569 and the NRC staff safety evaluation refer to the 1977 and 1987 versions of IEEE Standard 338, "Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems," and the 2012 version titled, "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems."

2. The MPS TS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the MPS Unit 3 TS for RTS Instrumentation is 3.3.1, whereas the TSTF-569 numbering for RTS Instrumentation is 3.3.1.16. Also, the MPS Unit 3 TS for ESFAS Instrumentation is 3.3.2, whereas the TSTF-569 numbering for ESFAS Instrumentation is 3.3.10.

3. The NAPS TS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the NAPS TS for ESF Actuation System (ESFAS)

Instrumentation is 3.3.9, whereas the TSTF-569 numbering for ESFAS Instrumentation is 3.3.10.

4. The VCS TS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the VCS Unit 1 TS for RTS Instrumentation is 3.3.1, whereas the TSTF-569 numbering for RTS Instrumentation is 3.3.10. Additionally, the VCS Unit 1 TS for ESFAS Instrumentation is 3.3.2, whereas the TSTF-569 numbering for ESFAS Instrumentation is 3.3.10.

5. The VCS TS Bases utilize a different reference format than the Standard Technical Specifications on which TSTF-569 was based. Specifically, the VCS TS Bases do not itemize references in each bases section. Therefore, the TS Bases change identifies the documents rather than noting a reference.

6. The VCS TS Amendment 158 approved two additional upgraded 7300 process cards, for which WCAP-14036 was applicable. The VCS TS Bases were revised to include these additional cards.

3.3

NRC Staff Evaluation

3.3.1 Application of TSTF-569 The NRC staff reviewed the request by comparing the licensees proposal against the changes described in TSTF-569, Revision 2. The NRC staff compared MPS, NAPS, and VCS design and existing TS with the design and TS presumed in TSTF-569. As explained below, the NRC staff concluded that the design and licenses (including TS) were sufficient to justify the licensee's reliance on the staff's safety evaluation of TSTF-569 as justification for adopting TSTF-569 in the MPS, NAPS, and VCS licenses.

The TSTF-569 is designed to make changes to NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, April 2012, Volume 1, Specifications (ADAMS Accession No. ML12100A222), and Volume 2, Bases (ADAMS Accession No. ML12100A228).

The staff compared the technical specifications assumed in TSTF-569 with the current technical specifications for MPS, NAPS, and VCS. The staff did not identify any material differences in the relevant technical specifications, including the evaluation of the proposed variations discussted in section 3.3.2 of this Safety Evaluation.

The licensee is relying on the previous analyses of TSTF-569. For the reasons stated in the NRC staff's SE for TSTF-569, the staff found that the methodology contained in TSTF-569, Rev. 2, Attachment 1, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing" provides a consistent, clear, and concise framework for determining that replacement components will operate at a level equivalent to that of the components being replaced. As such, using that methodology will assure that the necessary quality of the components is maintained and that the limiting conditions for operation will be met. Accordingly, approving the incorporation of that methodology into the licensing basis, and amending the TS to allow usage of the approved methodology, coupled with approving the aspect of the license amendment request to use the methodology in TSTF-569, Rev. 2, Attachment 1, results in TS that meet 10 CFR 50.36(c)(3) by assuring that performing SR 4.3.1.2 and 4.3.2.2 (for MPS and NAPS) and SR 3.3.1.16 and 3.3.2.9 for NAPS while using the new "[i]n lieu of" option, will assure that associated aspects of LCO 3.3.1 and 3.3.2 will be met.

3.3.2 Variations 3.3.2.1 Variation 1 MPS, NAPS, and VCS are committed to the 1971 version of IEEE Standard 338, which does not discuss response time testing. TSTF-569 and the NRC staff safety evaluation refer to the 1977 and 1987 versions of IEEE Standard 338, "Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems," and the 2012 version titled, "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems." The NRC staff reviewed this difference and determined that it is acceptable because the methodologies contained in TSTF-569, Attachment 1, provide adequate criteria for ensuring that replacement component's degraded response time issues or failures would be captured, and, as a result, conformance with the IEEE 338 design criteria is not affected.

3.3.2.2 Variations 2 through 5 The MPS, NAPS, and VCS utilize different numbering than the Standard Technical Specifications on which TSTF-569 was based. The NRC staff reviewed the TS numbering differences and determined that this variation is acceptable because the different numbering continues to apply the same requirements and are editorial in nature, and, therefore, continue to meet the intent of TSTF-569.

3.3.2.3 Variation 6 VCS TS Bases reflect approval for two additional upgraded 7300 process cards, for which WCAP-14036 was applicable. The VCS Unit 1 TS Bases were revised to include these additional cards following approval of amendment No. 158 (ADAMS Accession No. ML020770614). The NRC safety for amendment No. 158 found that the bounding values in WCAP-14036 were valid for these cards as well. Therefore, the NRC staff finds this variation acceptable.

3.3.3 Conclusion Based on the preceding evaluation, the NRC staff concludes that the revised TSs proposed by the licensee will continue to meet 10 CFR 50.36(c)(3) and the GDC, as incorporated into the North Anna UFSAR. The NRC acknowledges the licensee-controlled conforming changes to the TS Bases.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Virginia official was notified of the proposed issuance of the amendments. On January 21, 2022, the state official confirmed that the Commonwealth had no comments.

In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendments. On February 3, 2022, the state official confirmed that the State of Connecticut had no comments.

In accordance with the Commissions regulations, the South Carolina state official was notified of the proposed issuance of the amendments. On January 31, 2022, the state official confirmed that the State of South Carolina had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on November 30, 2021 (86 FR 67987). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributor: T. Sweat, NRR Date: March 1, 2022

ML22041A010

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