Text

Issuance of Amendments Revise Technical Specifications to Adopt TSTF 554
	ML22104A222
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford
Issue date:	05/12/2022
From:	Siva Lingam; NRC/NRR/DORL/LPL4
To:	Gaston R; Entergy Services
	Lingam S, NRR/DORL/LPL4
References
	EPID L-2021-LLA-0183
	Download: ML22104A222 (47)
	v • d • e

May 12, 2022 Mr. Ron Gaston Vice President, Regulatory Assurance Entergy Services, LLC M-ECH-29 1340 Echelon Parkway Jackson, MS 39213

SUBJECT:

ARKANSAS NUCLEAR ONE, UNITS 1 AND 2; GRAND GULF NUCLEAR STATION, UNIT 1; RIVER BEND STATION, UNIT 1; AND WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENTS TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-554, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS (EPID L-2021-LLA-0183)

Dear Mr. Gaston:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued amendments consisting of changes to the Technical Specifications in response to your application dated October 6, 2021, for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively);

Grand Gulf Nuclear Station, Unit 1 (Grand Gulf); River Bend Station, Unit 1 (River Bend); and Waterford Steam Electric Station, Unit 3 (Waterford-3). The following amendments are enclosed:

Amendment No. 276 to Renewed Facility Operating License No. DPR-51 for ANO-1, Amendment No. 330 to Renewed Facility Operating License No. NPF-6 for ANO-2 Amendment No. 231 to Renewed Facility Operating License No. NPF-29 for Grand Gulf Amendment No. 210 to Renewed Facility Operating License No. NPF-47 for River Bend Amendment No. 265 to Renewed Facility Operating License No. NPF-38 for Waterford-3 The amendments are based on Technical Specifications Task Force (TSTF) Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, dated January 16, 2020, using the Consolidated Line Item Improvement Process. The NRC issued a Final Safety Evaluation approving TSTF-554 on December 18, 2020.

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

Sincerely,

/RA/

Siva P. Lingam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-313, 50-368, 50-416, 50-458, and 50-382

Enclosures:

1. Amendment No. 276 to DPR-51

2. Amendment No. 330 to NPF-6

3. Amendment No. 231 to NPF-29

4. Amendment No. 210 to NPF-47

5. Amendment No. 265 to NPF-38

6. Safety Evaluation cc: Listserv

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. DPR-51

1.

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

A.

The application for amendment by Entergy Operations, Inc. (EOI), dated October 6, 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. DPR-51 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-51 and the Technical Specifications Date of Issuance: May 12, 2022 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2022.05.12 09:06:52 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 Replace the following pages of Renewed Facility Operating License No. DPR-51 and 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.

Renewed Facility Operating License REMOVE INSERT 3

3 Technical Specifications REMOVE INSERT 1.1-3 1.1-3 1.1-4 1.1-4 3.4.13-1 3.4.13-1

(5)

EOI, 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 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6)

EOI, pursuant to the Act and 10 CFR Parts 30 and 70, 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 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 276, are hereby incorporated in the renewed license.

EOI shall operate the facility in accordance with the Technical Specifications.

(3)

Safety Analysis Report The licensees SAR supplement submitted pursuant to 10 CFR 54.21(d),

as revised on March 14, 2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.

(4)

Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan, as submitted on May 4, 2006.

Renewed License No. DPR-51 Amendment No. 276 Revised by letter dated July 18, 2007

Definitions 1.1 ANO-1 1.1-3 Amendment No. 215,224,243,257,270, 1.1 Definition (continued)

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion."

DOSE EQUIVALENT XE-133 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-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."

INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; or 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE).

276

Definitions 1.1 ANO-1 1.1-4 Amendment No. 215,243,270, 1.1 Definition (continued)

LEAKAGE (continued) b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE.

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE-OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, 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 the SAR; b.

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

Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

=

1 Quadrants all in Power Average Quadrant Core any in Power 100 QPT 276

RCS Operational LEAKAGE 3.4.13 ANO-1 3.4.13-1 Amendment No. 215,224, 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

1 gpm unidentified LEAKAGE; c.

10 gpm identified LEAKAGE; and d.

150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

RCS unidentified or identified LEAKAGE not within limits, except for primary to secondary LEAKAGE.

B.1 Reduce LEAKAGE to within limits.

18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months C.

Required Action and associated Completion Time not met.

OR Primary to secondary LEAKAGE not within limit.

C.1 Be in MODE 3.

AND C.2 Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours 276

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 330 Renewed License No. NPF-6

1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee), dated October 6, 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-6 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 330, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-6 and the Technical Specifications Date of Issuance: May 12, 2022 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2022.05.12 09:07:27 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 330 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368 Replace the following pages of Renewed Facility Operating License No. NPF-6 and 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.

Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 1-3 1-3 1-4 1-4 3/4 4-14 3/4 4-14

Renewed License No. NPF-6 Amendment No. 330 3

(4)

EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70 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; (5)

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

EOI, pursuant to the Act and 10 CFR Parts 30 and 70 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 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; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 3026 megawatts thermal. Prior to attaining this power level EOI shall comply with the conditions in Paragraph 2.C.(3).

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 330, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

Exemptive 2nd paragraph of 2.C.2 deleted per Amendment 20, 3/3/81.

(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 issuance of the renewed license 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.

2.C.(3)(a)

Deleted per Amendment 24, 6/19/81.

ARKANSAS - UNIT 2 1-3 Amendment No. 157,220,255,266, 319,322, DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - The injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b.

Bistable channels - The injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

c.

Digital computer channels - The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify OPERABILITY.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SHUTDOWN MARGIN 1.13 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 control element assemblies are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 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 to not interfere with the operation of leakage detection systems, or c.

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

330

ARKANSAS - UNIT 2 1-4 Amendment No. 157,255,266,282, 293,315, DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

AZIMUTHAL POWER TILT - Tq 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT I-131 1.18 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

DOSE EQUIVALENT XE-133 1.19 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-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.

1.20 Deleted FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

330

ARKANSAS - UNIT 2 3/4 4-14 Amendment No. 184,266,280,327, Order dated 4/20/81 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

150 gallons per day primary to secondary leakage through any one steam generator (SG),

d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.

Leakage as specified in Table 3.4.6-1 for those Reactor Coolant System Pressure Isolation Valves identified in Table 3.4.6-1.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

a.

With any primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and primary to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate (Note 1) the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two valves* in each high pressure line having a non-functional valve and be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

These valves may include check valves for which the leakage rate has been verified, manual valves or automatic valves. Manual and automatic valves shall be tagged as closed to preclude inadvertent valve opening.

Note 1: Enter applicable ACTION(s) for systems(s) made inoperable by an inoperable pressure isolation valve.

330

ENTERGY OPERATIONS, INC.

SYSTEM ENERGY RESOURCES, INC.

COOPERATIVE ENERGY, A MISSISSIPPI ELECTRIC COOPERATIVE ENTERGY MISSISSIPPI, LLC DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 231 Renewed License No. NPF-29

1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee), dated October 6, 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-29 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 231 are hereby incorporated into this renewed license. Entergy Operations, 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 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-29 and the Technical Specifications Date of Issuance: May 12, 2022 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2022.05.12 09:07:56 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 231 RENEWED FACILITY OPERATING LICENSE NO. NPF-29 GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 Replace the following pages of Renewed Facility Operating License No. NPF-29 and 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.

Renewed Facility Operating License Remove Insert Technical Specifications Remove Insert 1.0-4 1.0-4 3.4-12 3.4-12 3.4-13 3.4-13

4 Amendment No. 231 amended, are fully applicable to the lessors and any successors in interest to those lessors, as long as the renewed license of GGNS Unit 1 remains in effect.

(b)

SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54. In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility.

C.

The 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 Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 4408 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 231 are hereby incorporated into this renewed license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

During Cycle 19, GGNS will conduct monitoring of the Oscillation Power Range Monitor (OPRM). During this time, the OPRM Upscale function (Function 2.f of Technical Specification Table 3.3.1.1-1) will be disabled and operated in an indicate only mode and technical specification requirements will not apply to this function. During such time, Backup Stability Protection measures will be implemented via GGNS procedures to provide an alternate method to detect and suppress reactor core thermal hydraulic instability oscillations. Once monitoring has been successfully completed, the OPRM Upscale function will be enabled and technical specification requirements will be applied to the function; no further operating with this function in an indicate only mode will be conducted.

Definitions 1.1 GRAND GULF 1.0-4 Amendment No. 120, 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE; d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length RATE (LHGR) of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all TEST required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may (continued) 231

RCS Operational LEAKAGE 3.4.5 GRAND GULF 3.4-12 Amendment No. 120, 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Operational LEAKAGE LCO 3.4.5 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

5 gpm unidentified LEAKAGE; c.

30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; and d.

2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressure boundary LEAKAGE exists.

A.1 Isolate affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.

Unidentified LEAKAGE not within limit.

OR Total LEAKAGE not within limit.

B.1 Reduce LEAKAGE to within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued) 231

RCS Operational LEAKAGE 3.4.5 GRAND GULF 3.4-13 Amendment No. 120, 219, ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Unidentified LEAKAGE increase not within limit.

C.1 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D.

Required Action and associated Completion Time not met.

D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify RCS unidentified LEAKAGE, total LEAKAGE, and unidentified LEAKAGE increase are within limits.

In accordance with the Surveillance Frequency Control Program 231

ENTERGY LOUISIANA, LLC AND ENTERGY OPERATIONS, INC.

DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 210 Renewed License No. NPF-47

1.

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

A.

The application for amendment by Entergy Operations, Inc. (EOI), dated October 6, 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, as amended, 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-47 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. 210 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-47 and the Technical Specifications Date of Issuance: May 12, 2022 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2022.05.12 09:08:25 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 210 RENEWED FACILITY OPERATING LICENSE NO. NPF-47 RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458 Replace the following pages of Renewed Facility Operating License No. NPF-47 and 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.

Renewed Facility Operating License Remove Insert Technical Specifications Remove Insert 1.0-4 1.0-4 3.4-12 3.4-12 3.4-13 3.4-13

(2)

EOI, pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use and operate the facility at the above designated location in accordance with the procedures and limitations set forth in this renewed license; (3)

EOI, pursuant to Section 103 of the Act and 10 CFR Part 70, to receive, possess and to 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 Final Safety Analysis Report, as supplemented and amended; (4)

EOI, pursuant to Section 103 of the Act and 10 CFR Parts 30, 40 and 70, 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; (5)

EOI, pursuant to Section 103 of the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)

EOI, pursuant to Section 103 of the Act and 10 CFR Parts 30, 40 and 70, 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 set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and 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 EOI is authorized to operate the facility at reactor core power levels not in excess of 3091 megawatts thermal (100% rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 210 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 210

Definitions 1.1 RIVER BEND 1.0-4 Amendment No. 81, 1.1 Definitions (continued)

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 1.

LEAKAGE into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2.

LEAKAGE into the drywell atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems; b.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c.

Total LEAKAGE Sum of the identified and unidentified LEAKAGE; d.

Pressure Boundary LEAKAGE LEAKAGE through a fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

(continued) 210

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RCS Operational LEAKAGE 3.4.5 RIVER BEND 3.4-13 Amendment No. 81, 196, ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Unidentified LEAKAGE increase not within limit.

C.1 Verify source of unidentified LEAKAGE increase is not service sensitive type 304, type 316 austenitic stainless steel, or other intergranular stress corrosion cracking susceptible material.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D.

Required Action and associated Completion Time not met.

D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify RCS unidentified LEAKAGE, total LEAKAGE, and unidentified LEAKAGE increase are within limits.

In accordance with the Surveillance Frequency Control Program 210

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 265 Renewed License No. NPF-38

1.

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

A.

The application for amendment by Entergy Operations, Inc. (EOI),dated October 6, 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-38 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. 265, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI 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 from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-38 and the Technical Specifications Date of Issuance: May 12, 2022 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2022.05.12 09:08:57 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 265 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38 WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 Replace the following pages of Renewed Facility Operating License No. NPF-38 and 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.

Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 1-4 1-4 1-5 1-5 3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19

the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on 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 EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 265, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

Antitrust Conditions (a)

Entergy Louisiana, LLC shall comply with the antitrust license conditions in Appendix C to this renewed license.

(b)

Entergy Louisiana, LLC is responsible and accountable for the actions of its agents to the extent said agent's actions contravene the antitrust license conditions in Appendix C to this renewed license.

AMENDMENT NO. 265

WATERFORD - UNIT 3 1-4 AMENDMENT NO. 68, 84, 116, 204, DEFINITIONS IDENTIFIED LEAKAGE (Continued) b.

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

Reactor Coolant System leakage through a steam generator to the secondary system (primary to secondary leakage).

MEMBER(S) OF THE PUBLIC 1.15 MEMBER(S) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.16 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specification 6.9.1.7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.17 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.18 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.

265

WATERFORD - UNIT 3 1-5 AMENDMENT NO. 68, 204, DEFINITIONS PHYSICS TEST 1.19 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) 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.

PLANAR RADIAL PEAKING FACTOR - Fxy 1.20 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a fault in a Reactor Coolant System component body, pipe wall, or vessel wall. Leakage past seals, packing, and gaskets is not PRESSURE BOUNDARY LEAKAGE.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that process-ing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be 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.

265

WATERFORD - UNIT 3 3/4 4-18 AMENDMENT NO. 197,199, 204, 249, REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant System operational leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 gpm UNIDENTIFIED LEAKAGE, c.

75 gallons per day primary to secondary leakage, through any one steam generator (SG),

d.

10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.

1 gpm leakage at a Reactor Coolant System pressure of 2250 +/- 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With any primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With any Reactor Coolant System operational leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE, primary to secondary leakage, and leakage from Reactor Coolant System pressure isolation valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual or deactivated automatic valve, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

265

WATERFORD - UNIT 3 3/4 4-19 AMENDMENT NO. 96,107, 204, 249, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS NOTE:

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

4.4.5.2.1 Reactor Coolant System leakages, except for primary to secondary leakage, shall be demonstrated to be within each of the above limits by performance of a Reactor Coolant System water inventory balance in accordance with the Surveillance Frequency Control Program.

4.4.5.2.2 Primary to secondary leakage shall be verified to be 75 gallons per day through any one SG in accordance with the Surveillance Frequency Control Program.

4.4.5.2.3 Each Reactor Coolant System pressure isolation valve specified in Table 3.4-1, Section A and Section B, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a.

In accordance with the Surveillance Frequency Control Program, b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, c.

Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, d.

Following valve actuation for valves in Section B due to automatic or manual action or flow through the valve:

1.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and 2.

Within 31 days by verifying leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.5.2.4 Each Reactor Coolant System pressure isolation valve power-operated valve specified in Table 3.4-1, Section C, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a.

In accordance with the Surveillance Frequency Control Program, and b.

Prior to returning the valve to service following maintenance, repair, or replacement work on the valve.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

265

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 276 AND 330 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-51 AND NPF-6, AMENDMENT NO. 231 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-29, AMENDMENT NO. 210 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-47, AND AMENDMENT NO. 265 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 GRAND GULF NUCLEAR STATION, UNIT 1 RIVER BEND STATION, UNIT 1 WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NOS. 50-313, 50-368, 50-416, 50-458, AND 50-382

1.0 INTRODUCTION

By application dated October 6, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21279A231), Entergy Operations, Inc. (the licensee) submitted a license amendment request (LAR) to revise the Technical Specifications (TSs) for Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2, respectively), Grand Gulf Nuclear Station, Unit 1 (Grand Gulf), River Bend Station, Unit 1 (River Bend), and Waterford Steam Electric Station, Unit 3 (Waterford-3).

In the letter dated October 6, 2021, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed LAR under the Consolidated Line Item Improvement Process.

The proposed changes would revise the TSs related to reactor coolant system (RCS) operational leakage and the definition of the term LEAKAGE based on Technical Specifications Task Force (TSTF) Traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, dated January 16, 2020 (ML20016A233), and the associated NRC staff Final Safety Evaluation (SE) of TSTF-554 dated December 18, 2020 (ML20322A024).

The licensee has proposed variations from the TS changes described in TSTF-554, Revision 1.

The variations are described and evaluated in Section 3.2 of this SE.

1.1 RCS Description Components that contain or transport the coolant to or from the reactor core make up the RCS.

Materials can degrade as a result of the complex interaction of the materials, the stresses they encounter, and through operational wear or mechanical deterioration during normal and upset operating environments. Such material degradation could lead to leakage of reactor coolant into containment buildings.

RCS leakage falls under two main categoriesidentified leakage and unidentified leakage.

Identifying the sources of leakage is necessary for prompt identification of potentially adverse conditions, assessment of safety significance of the leakage, and quick corrective action.

A limited amount of leakage from the reactor coolant pressure boundary (RCPB) directly into the containment/drywell atmosphere is expected as the RCS and other connected systems cannot be made 100 percent leak tight. This leakage is detected, located, and isolated from the containment atmosphere so as to not interfere with measurement of unexpected RCS leakage detection.

Leakage from the RCPB inside the primary containment/drywell is detected by independently monitored parameters, such as sump level changes and containment/drywell gaseous and particulate radioactivity levels. Plant TSs identify at least two independent and diverse means and/or methods of detection. The primary means of quantifying significant leakage in the containment/drywell is the containment/drywell sump monitoring system. The containment atmosphere particulate and gaseous radioactivity monitors are sensitive to radioactivity in any RCS leakage but do not provide a reasonably accurate means of quantifying leakage.

The containment/drywell sump monitoring system monitors the liquid collected in the sump.

This liquid consists of leakage from the RCS, leakage from other systems inside primary containment (e.g., component cooling water), and condensation of steam released from the RCS or other high-temperature systems that is condensed by the containment/drywell coolers and directed to the sump. The containment sump instrumentation measures the rate of liquid accumulation in the sump, displays results in the main control room, and provides for an alarm for high rates of liquid accumulation. The rate of liquid accumulation may be determined by changes in measured level in the sump or by the time between periodic pump operation to drain the sump between known sump levels.

Gaseous and/or particulate primary containment atmospheric radioactivity monitors continuously monitor the containment atmosphere during reactor operation for indications of leakage. The RCS contains radioactivity that, when released to the primary containment, can be detected by the gaseous or particulate primary containment atmospheric radioactivity monitor. Radioactivity detection systems are included for monitoring particulate and/or gaseous activities because of their sensitivities and rapid responses to RCS leakage. Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products have been released from fuel elements. To enhance detection capability, radioactivity alarm settings are typically set to provide the most sensitive response without causing an excessive number of spurious alarms.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS leakage into the containment area is necessary. Separation of identified leakage from unidentified leakage provides quantitative information to the operators, allowing them to take corrective action should leakage occur that is detrimental to the safety of the unit and the public.

1.2 Proposed TS Changes to Adopt TSTF-554 In accordance with NRC staff-approved TSTF-554, the licensee proposed changes that would revise the TSs related to RCS operational leakage and the definition of the term LEAKAGE.

Specifically, the licensee proposed the following TS changes to adopt TSTF-554:

For each facility, the TS 1.1 identified LEAKAGE definition in subsection a.2 for ANO-1, Grand Gulf, and River Bend, and subsection 1.14b for ANO-2 and Waterford-3, would be revised to remove the exclusion of pressure boundary leakage from identified leakage by deleting either, and the phrase not to be pressure boundary LEAKAGE.

Pressure boundary leakage is currently defined in the TSs as leakage (except primary to secondary leakage in a pressurized-water reactor (PWR)) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. For each unit, the pressure boundary LEAKAGE definition as specified in TS 1.1c for ANO-1, TS 1.1d for Grand Gulf and River Bend, TS 1.16 for ANO-2, and TS 1.21 for Waterford-3 would be revised to delete the word nonisolable. The sentence, LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE, would be relocated from the Technical Specifications Bases and added to the definition.

Additionally, for each unit, the LEAKAGE definition would be revised by other editorial and punctuation changes to reflect the deletion and listed definitions.

The RCS operational LEAKAGE requirements in TS 3.4.5 for River Bend and Grand Gulf, TS 3.4.5.2 for Waterford-3, TS 3.4.13 for ANO-1, and TS 3.4.6.2 for ANO-2, would be revised as follows:

o For ANO-1, River Bend and Grand Gulf TSs identified above, a new Condition A to isolate the pressure boundary leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , would be added.

o For the Waterford-3 and ANO-2 TSs identified above, a new Action b would be added as follows:

With any PRESSURE BOUNDARY LEAKAGE not within limit, isolate affected component, pipe, or vessel from the Reactor Coolant System by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

o Existing Conditions A and B and the associated required actions in TSs identified above for ANO-1; Conditions A, B, and C for Grand Gulf and River Bend; and TS Actions b and c for ANO-2 and Waterford-3, would be renumbered.

o Existing Condition C in TSs identified above for ANO-1, Grand Gulf, and River Bend would be revised to delete the condition Pressure boundary LEAKAGE exists, because pressure boundary leakage would be addressed by the new Condition A. In addition, this Condition would be revised to be applicable when the required action and associated completion time of the other conditions are not met.

Action b in TSs identified above for ANO-2 and Waterford-3 currently excludes the Pressure Boundary Leakage and primary to secondary leakage, requirement, therefore no change to these facilities TSs is required.

1.3 Additional Proposed TS Changes The application identified certain variations from TSTF-554. Section 3.2 of this SE provides a description of the variations and the NRC staffs evaluation.

2.0 REGULATORY EVALUATION

The regulation under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c)(2) requires that TSs include limiting conditions for operation (LCOs). Per 10 CFR 50.36(c)(2)(i),

LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation also requires that [w]hen a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared standard technical specifications (STS) for each of the LWR nuclear designs.

Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the NRC Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ML12100A177 and ML12100A178, respectively); Standard Technical Specifications, Combustion Engineering Plants, NUREG-1432, Volume 1, Specifications, and Volume 2 Bases, Revision 4.0, dated April 2012 (ML12102A165 and ML12102A169, respectively); Standard Technical Specifications, General Electric BWR [Boiling Water Reactor]/6 Plants, NUREG-1434, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ML12104A195 and ML12104A196, respectively), as modified by NRC-approved travelers.

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes to Adopt TSTF-554 The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-554. The NRC staff finds that the licensees proposed changes to the ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford-3 TSs as described in Section 1.1 of this SE, are consistent with those found acceptable in TSTF-554. The NRC staff determined that the STS changes approved in TSTF-554 are applicable to ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford-3 TSs because ANO-1, ANO-2, and Waterford-3 are PWRs and

Grand Gulf and River Bend are BWRs, and, in accordance with the SRP Chapter 16.0, the NRC staff has approved the TSTF-554 changes for PWR and BWR designs.

The NRC staff finds that proposed changes to the TS definition for each facility clarify what constitutes pressure boundary leakage, and the source of leakage does not matter if the TSs have separate limits on pressure boundary leakage. The NRC staff also finds that the TSs for ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford-3 listed in the amendment correctly specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. Also, the NRC staff finds that the proposed changes to specific LCOs as listed in Section 1.2 of this SE are adequate remedial actions to be taken until each LCO can be met to provide adequate protection to the health and safety of the public. Thus, the proposed changes continue to meet the requirements of 10 CFR 50.36(c)(2)(i) as discussed in Section 3.0 of the NRC staffs Final SE for TSTF-554.

3.2 Technical Evaluation of Additional Proposed TS Changes 3.2.1 Editorial Changes The licensee noted that ANO-2, Grand Gulf, River Bend, and Waterford-3 TSs have different numbering and nomenclature than STS. The NRC staff finds that the different TS numbering and nomenclature changes are acceptable because they do not substantively alter TS requirements.

3.2.2 Other Variations ANO-1 TS Variation The application does not specify any variation for ANO-1 TS from TSTF-554.

ANO-2 TS Variation The licensee noted that ANO-2 TSs are based on NUREG-0212, Standard Technical Specifications, Combustion Engineering Plants. TSTF-554 is based on NUREG-1432 of the same title.

Prior to the development of STS, such as, NUREG-1432, NUREG-0212 was considered not a part of the standards but simply the latest approved TSs for a given style of plant, such as Waterford-3. The issuance of NUREG-1432, considered the requirements of NUREG-0212. As a result, the licensees proposed changes meet the intent of NUREG-1432 and do not affect the applicability of TSTF-554 to the Waterford-3 TS.

The licensee noted that in the STS on which TSTF-554 is based, all leakage definitions appear as part of the single defined term, LEAKAGE. In the ANO-2 TS, the terms related to leakage are separate definitions. However, the changes to the affected defined terms IDENTIFIED LEAKAGE and PRESSURE BOUNDARY LEAKAGE are the same as the changes made to those terms in TSTF-554.

The NRC staff reviewed the licensees variation and determined that the modified definitions are acceptable since the application of those definitions in the TS are in accordance with the approved TSTF-554.

The licensee stated that in the ANO-2 TS, the Operational Leakage specification is numbered 3.4.6.2. In the STS on which TSTF-554 is based, the TS is numbered 3.4.13.

Further, the licensee stated that in TSTF-554, a new Action A (TS 3.4.13) is created that is applicable when there is pressure boundary leakage. In the ANO-2 TS, existing Action a (TS 3.4.6.2) applies when there is pressure boundary leakage or primary to secondary leakage not within limit. The ANO-2 TS are revised such that Action a is applicable when primary to secondary leakage is not within limit and a new Action b is created that applies to the existence of pressure boundary leakage. The existing ANO-2 Actions b and c are renamed Actions c and d.

The NRC staff reviewed the licensees statement and determined that the variation has no effect on the applicability of the proposed change and is, therefore, acceptable.

The licensee stated that in the ANO-2 TS, shutdown requirements are included in each Action. The TSTF-554 shutdown actions in STS Condition C are incorporated into new ANO-2 Action b. The requirements are consistent with TSTF-554.

The NRC staff reviewed the licensees statement and determined that the variation is administrative, has no effect on the applicability of the proposed change and is, therefore, acceptable.

Grand Gulf and River Bend TSs Variation from TSTF-554 The current Grand Gulf and River Bend TS 3.4.5, RCS Operational Leakage, Action B differs slightly from the same Action in NUREG-1434, Revision 4 (i.e., the basis for TSTF-554). Specifically, the NUREG-1434 TS 3.4.5 Action B includes Required Actions B.1 and B.2, which are linked with an OR connector, each with a 4-hour Completion Time (CT):

B.1 Reduced LEAKAGE to within limit.

OR B.2 Verify source of unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

Failure to comply with either of the two Required Actions within the specified CT would require the unit to be placed in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

In contrast to NUREG-1434, Action B, the Grand Gulf and River Bend TSs only include NUREG-1434 Required Action B.2 as a specified action. This difference does not impact the overall intent of the Action. As stated in the NUREG-1434, TS Bases (Volume 2), an unidentified increase of greater than 2 gallons per minute in leakage within a 4-hour period is an indication of a potential flaw in the RCPB and must be quickly evaluated. Although the increase may not necessarily violate the absolute unidentified leakage limit, certain susceptible components must be determined not to be the source of the leakage increase within the required CT. Thus, current Action B in the Grand Gulf and River Bend TSs is consistent with the intent of the NUREG-1434 Action B. In addition, this difference does not affect the applicability of TSTF-554 to the Grand Gulf and River Bend TSs.

The NRC staff finds the variation acceptable because it continues to meet the requirements of 10 CFR 50.36(c)(2)(i), by providing remedial actions and shutting down the reactor if the remedial actions cannot be met.

Waterford-3 TS Variation from TSTF-554 The licensee stated that the Waterford-3 TS are based on NUREG-0212. TSTF-554 is based on NUREG-1432 of the same title.

Prior to the development of STS, such as, NUREG-1432, NUREG-0212 was considered not a part of the standards but simply the latest approved TSs for a given style of plant, such as Waterford-3. The issuance of NUREG-1432, considered the requirements of NUREG-0212. As a result, the licensees proposed changes meet the intent of NUREG-1432 and do not affect the applicability of TSTF-554 to the Waterford-3 TS.

The licensee noted that in the STS on which TSTF-554 is based, all leakage definitions appear as part of the single defined term, LEAKAGE. In the Waterford-3 TS, the terms related to leakage are separate definitions. However, the changes to the affected defined terms Identified Leakage and Pressure Boundary Leakage are the same as the changes made to those terms in TSTF-554.

The NRC staff reviewed the licensees variation and determined that the modified definitions are acceptable since the application of those definitions in the TS are in accordance with the approved TSTF-554.

The licensee noted that in TSTF-554, a new Action A (TS 3.4.13) is created that is applicable when there is pressure boundary leakage. In the Waterford-3 TS, existing Action a (TS 3.4.5.2) applies when there is pressure boundary leakage or primary to secondary leakage not within limit. The Waterford-3 TSs are revised such that Action a is applicable when primary to secondary leakage is not within limit and a new Action b is created that applies to the existence of pressure boundary leakage.

The existing Waterford-3 Actions b and c are renamed Actions c and d.

The NRC staff reviewed the licensees statement and determined that the variation has no effect on the applicability of the proposed change and is, therefore, acceptable.

The licensee stated that in the Waterford-3 TS, shutdown requirements are included in each Action. TSTF-554 shutdown actions in STS Condition C are incorporated into new Waterford Action b. The requirements are consistent with TSTF-554.

The NRC staff reviewed the licensees statement and determined that the variation is administrative, has no effect on the applicability of the proposed change and is, therefore, acceptable.

The licensee stated that an editorial change that is unrelated to this TSTF is being made to DEFINITION 1.23 PURGE - PURGING. A preexisting typographical error was identified at the end of the second line. The word concentration had been mis-typed as concentra-. The application states, Since we are working on this page

of TS, the typographical error is being corrected to concentration. Correcting this typographical error does not negatively affect the definition.

The NRC staff concurs with editorial change.

Based on the above, the NRC staff finds that the licensees proposed variations from TSTF-554 are reasonable and do not affect the applicability of TSTF-554 to TSs for ANO-2, Grand Gulf, River Bend, and Waterford-3. The NRC staff, therefore, finds the variations acceptable.

3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arkansas, Mississippi, and Louisiana State officials were notified of the proposed issuance of the amendments on April 12, 2022. The State officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

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

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions 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: R. Grover, NRR Date: May 12, 2022

ML22104A222

by email OFFICE NRR/DORL/LPL4/PM*

NRR/DORL/LPL4/LA*

NRR/DSS/STSB/BC*

NRR/DSS/SCPB/BC*

NAME SLingam PBlechman VCusumano BWittick DATE 4/13/2022 4/25/2022 4/8/2022 5/5/2022 OFFICE OGC NLO*

NRR/DORL/LPL4/BC*

NRR/DORL/LPL4/PM*

NAME RHarper JDixon-Herrity SLingam DATE 5/10/2022 5/12/2022 5/12/2022

v t e Letter Sequence Approval
EPID:L-2021-LLA-0183, TSTF 554 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2021-11-04 [Table View]

ML22104A222

Text

SUBJECT:

Dear Mr. Gaston:

Enclosures:

Attachment:

Attachment:

Attachment:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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