Text

Issuance of Amendments Related to Revision of Technical Specifications to Use Online Monitoring Methodology
	ML26007A265
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford
Issue date:	01/21/2026
From:	James Drake; Plant Licensing Branch IV
To:	Gustafson O; Entergy Services, Entergy Services
	Mahoney, M
References
	EPID L-2024-LLA-0160
	Download: ML26007A265 (0)
	v • d • e

January 21, 2026 Mr. Otto Gustafson 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 RELATED TO REVISION OF TECHNICAL SPECIFICATIONS TO USE ONLINE MONITORING METHODOLOGY (EPID L-2024-LLA-0160)

Dear Mr. Gustafson:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued amendments consisting of changes to the technical specifications (TSs) in response to Entergy Operations, Inc. (Entergy, the licensee) application dated December 4, 2024. The following amendments are enclosed:

Amendment Nos. 285 and 338 to Renewed Facility Operating License Nos. DPR-51 and NPF-6 for Arkansas Nuclear One, Units 1 and 2, respectively.

Amendment No. 240 to Renewed Facility Operating License No. NPF-29 for Grand Gulf Nuclear Station, Unit 1.

Amendment No. 219 to Renewed Facility Operating License No. NPF-47 for River Bend Station, Unit 1.

Amendment No. 277 to Renewed Facility Operating License No. NPF-38 for Waterford Steam Electric Station, Unit 3.

The proposed amendments revise certain definitions in each plants Definitions TS section and adds a new Online Monitoring Program in each plants Administrative Controls TS section.

Entergy proposes to use online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed amendments are based on the NRC-approved topical report Analysis and Measurement Services Corporation (AMS)-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters.

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/

Jason Drake, 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, 50-382

Enclosures:

1. Amendment No. 285 to DPR-51

2. Amendment No. 338 to NPF-6

3. Amendment No. 240 to NPF-29

4. Amendment No. 219 to NPF-47

5. Amendment No. 277 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. 285 Renewed License No. DPR-51

1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee), dated December 4, 2024, 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. 285, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.

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 Michael Mahoney, Acting 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: January 21, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.01.21 15:27:04 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 285 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 1.1-1 1.1-1 5.0-20a 5.0-20a 5.0-20b

(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. 285, 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. 285 Revised by letter dated July 18, 2007

Definitions 1.1 ANO-1 1.1-1 Amendment No. 215, 285 1.0 USE AND APPLICATION 1.1 Definitions

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ALLOWABLE THERMAL POWER ALLOWABLE THERMAL POWER shall be the maximum steady state reactor core heat transfer rate to the reactor coolant permitted by consideration of the number and configuration of reactor coolant pumps (RCPs) in operation.

AXIAL POWER IMBALANCE AXIAL POWER IMBALANCE shall be the power in the top half of the core, expressed as a percentage of RATED THERMAL POWER (RTP), minus the power in the bottom half of the core, expressed as a percentage of RTP.

AXIAL POWER SHAPING APSRs shall be the control components with part length RODS (APSRs) absorbers used to control the axial power distribution of the reactor core. The APSRs are positioned manually by the operator and are not trippable.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program).

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

Programs and Manuals 5.5 ANO-1 5.0-20a Amendment No. 281,283, 285 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals c.

When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1.

For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2.

For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the plant configuration change, whichever is less.

3.

Revising the RICT is not required If the plant configuration change would lower plant risk and would result in a longer RICT.

d.

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1.

Numerically accounting for the increased possibility of CCF in the RICT calculation; or 2.

Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e.

A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards using penalty factors, and configuration specific extreme winds and tornado hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f.

The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities."

g.

A report shall be submitted in accordance with Specification 5.6.8 before a newly developed method is used to calculate a RICT.

Programs and Manuals 5.5 ANO-1 5.0-20b Amendment No. 285 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.19 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.

The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (proprietary version). The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1) Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration

check,

2) Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3) Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4) Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 338 Renewed License No. NPF-6

1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee), dated December 4, 2024, 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. 338, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications

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 Michael Mahoney, Acting 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: January 21, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.01.21 15:27:50 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 338 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.

Operating License REMOVE INSERT 3

3 Technical Specifications REMOVE INSERT 1-2 1-2 1-5 1-5 6-18d

Renewed License No. NPF-6 Amendment No. 338 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. 338, 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 1-2 Amendment No. 154,157,319,334,338 DEFINITIONS CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when:

1.8.1 All penetrations required to be closed during accident conditions are either:

a.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or b.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.8.2 All equipment hatches are closed and sealed, 1.8.3 Each airlock is OPERABLE pursuant to Specification 3.6.1.3, 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.8.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or O-rings) is OPERABLE.

CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

ARKANSAS - UNIT 2 1-5 Amendment No. 24,60,157,193,239, 324,338 DEFINITIONS AXIAL SHAPE INDEX 1.22 The AXIAL SHAPE INDEX shall be the power generated in the lower half of the core less the power generated in the upper half of the core divided by the sum of these powers.

REACTOR TRIP SYSTEM RESPONSE TIME 1.23 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 electrical power is interrupted to the CEA drive mechanism. 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.24 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 methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

PHYSICS TESTS 1.25 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.

SOFTWARE 1.26 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation and procedures.

PLANAR RADIAL PEAKING FACTOR Fxy 1.27 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.

ARKANSAS - UNIT 2 6-18d Amendment No. 338 ADMINISTRATIVE CONTROLS 6.5.21 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.

The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (proprietary version). The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1) Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check,

2) Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3) Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4) Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 4.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

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. 240 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 December 4, 2024, 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. 240 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 Michael Mahoney, Acting 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: January 21, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.01.21 15:28:12 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 240 RENEWED FACILITY OPERATING LICENSE NO. NPF-29 GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 Replace the following page 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.

Facility Operating License REMOVE INSERT 4

4 Technical Specifications REMOVE INSERT 1.0-1 1.0-1 1.0-3a 1.0-3a 1.0-3b 1.0-5 1.0-5 5.0-16c 5.0-16c 5.0-16d 5.0-16d

4 Amendment No. 240 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. 240 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.

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ENTERGY LOUISIANA, LLC AND ENTERGY OPERATIONS, INC.

DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 Renewed License No. NPF-47

1.

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

A.

The application for amendment by Entergy Operations, Inc. (the licensee), dated December 4, 2024, 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. 219 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 Michael Mahoney, Acting 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: January 21, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.01.21 15:28:37 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 219 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.

Facility Operating License Remove Insert 3

3 Technical Specifications Remove Insert 1.0-1 1.0-1 1.0-3 1.0-3 1.0-3a 1.0-5 1.0-5 1.0-6 1.0-6 5.0-16c 5.0-16c 5.0-16d Amendment No. 219 (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.

(7)

EOI, pursuant to the Act and 10 CFR Part 30, 40, and 70 to receive, possess and use, in amounts as required, such byproduct and special nuclear materials as may be produced by the operation of 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, without restriction to chemical or physical form for the purposes of sample analysis, equipment calibration, or equipment repair.

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. 219 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.

Definitions 1.1 RIVER BEND 1.0-1 Amendment No. 81,207,215,219 1.0 USE AND APPLICATION 1.1 Definitions

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific planar height HEAT GENERATION RATE and is equal to the sum of the LHGRs for all the fuel rods in (APLHGR) the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

(continued)

Definitions 1.1 RIVER BEND 1.0-3 Amendment No. 81, 84, 193, 219 1.1 Definitions (continued)

DRAIN TIME (continued)

EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE TIME END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) SYSTEM RESPONSE TIME d.

No additional draining events occur; and e.

Realistic cross-sectional areas and drain rates are used.

A bounding DRAIN TIME may be used in lieu of a calculated value.

The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval from initial movement of the associated turbine stop valve or the turbine control valve to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

(continued)

Definitions 1.1 RIVER BEND 1.0-3a Amendment No. 219 1.1 Definitions (continued)

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

(continued)

Definitions 1.1 RIVER BEND 1.0-5 Amendment No. 81,114,129,219 1.1 Definitions (continued)

MAXIMUM FRACTION The MFLPD shall be the largest value of the fraction of OF LIMITING limiting power density in the core. The fraction of limiting POWER DENSITY (MFLPD) power density shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR)

RATIO (MCPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, 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, division, 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, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3091 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint TIME at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

(continued)

Definitions 1.1 RIVER BEND 1.0-6 Amendment No. 81, 180, 219 1.1 Definitions (continued)

SHUTDOWN MARGIN (SDM)

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a.

The reactor is xenon free; b.

The moderator temperature is 68°F; corresponding to the most reactive state; and c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME RESPONSE TIME consists of two components:

a.

The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and b.

The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.

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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

Programs and Manuals 5.5 RIVER BEND 5.0-16c Amendment No. 213,218,219 5.5 Programs and Manuals 5.5.16 Risk Informed Completion Time Program (continued) c.

When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1.

For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2.

For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e.,

not the RICT) or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the plant configuration change, whichever is less.

3.

Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d.

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1.

Numerically accounting for the increased possibility of CCF in the RICT calculation; or 2.

Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the functions(s) performed by the inoperable SSCs.

e.

A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f.

The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-informed Activities."

g.

A report shall be submitted in accordance with Specification 5.6.6 before a newly developed method is used to calculate a RICT.

(continued)

Programs and Manuals 5.5 RIVER BEND 5.0-16d Amendment No. 219 5.5 Programs and Manuals 5.5.17 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.

The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (proprietary version). The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1) Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration

check,

2) Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3) Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4) Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 277 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 December 4, 2024, 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 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. 277, 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 Michael Mahoney, Acting 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: January 21, 2026 MICHAEL MAHONEY Digitally signed by MICHAEL MAHONEY Date: 2026.01.21 15:29:02 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 277 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 4

4 Technical Specifications REMOVE INSERT 1-1 1-1 1-3 1-3 1-6 1-6 6-11 6-11

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. 277, 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. 277

WATERFORD - UNIT 3 1-1 AMENDMENT NO. 266, 277 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

AXIAL SHAPE INDEX 1.2 The AXIAL SHAPE INDEX shall be the power generated in the lower half of the core less the power generated in the upper half of the core divided by the sum of these powers.

AZIMUTHAL POWER TILT - Tq 1.3 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

WATERFORD - UNIT 3 1-3 AMENDMENT NO. 102,175,199,268,277 DEFINITIONS CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

COLR - CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT is the Waterford 3 specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Technical Specification 6.9.1.11. Plant operation within these operating limits is addressed in individual specifications.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP-30, Supplement to Part 1, Pages 192-212, Tables titled, "Committed Dose Equivalent in Target Organs or Tissue per Intake of Unit Activity."

- AVERAGE DISINTEGRATION ENERGY 1.11 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 noniodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES 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 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

WATERFORD - UNIT 3 1-6 AMENDMENT NO. 175,182,183,199, 268,277 DEFINITIONS RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3716 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.25 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 electrical power to the CEA drive mechanism is interrupted. The response time may be measured by 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

REPORTABLE EVENT 1.26 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRITY 1.27 SHIELD BUILDING INTEGRITY shall exist when:

a.

Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at least one door shall be

closed, b.

The shield building filtration system is in compliance with the requirements of Specification 3.6.6.1, and c.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE.

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

WATERFORD - UNIT 3 6-11 AMENDMENT NO. 270,277 Next Page is 6-14 ADMINISTRATIVE CONTROLS 6.5.20 Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.

The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (proprietary version). The program shall include the following elements:

a.

Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1) Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check,

2) Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance,

3) Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4) Documentation of the results of the online monitoring data analysis.

b.

Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c.

Performance of calibration checks for transmitters at the specified backstop frequencies.

d.

The provisions of Surveillance Requirement 4.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 285 AND 338 TO RENEWED FACILITY OPERATING LICENSE NOS.

DPR-51 AND NPF-6, AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-29, AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-47 AMENDMENT NO. 277 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 December 4, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24339B304), Energy Operations Inc. (Entergy, the licensee) requested changes to 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).

The proposed amendments would revise certain definitions in each plants Definitions TS section and adds a new Online Monitoring Program in each plants Administrative Controls TS section. Entergy proposes to use online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed amendments are based on the U.S. Nuclear Regulatory Commission (NRC, the Commission) approved topical report Analysis and Measurement Services Corporation (AMS)-TR-0720R2-A, Online Monitoring

Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (ML21235A493).

The NRC staff issued a safety evaluation (SE) approving the -A version of the AMS-TR-0720R2-A on August 11, 2021, Final Safety Evaluation for AMS [Analysis and Measurement Services] Online Monitoring Topical Report (package ML21179A060). The SE states, in part, that the NRC staff finds that implementation of an OLM program in accordance with the approved AMS OLM TR provides an acceptable alternative to periodic manual calibration surveillance requirements upon implementation of the application-specific action items Entergy has not proposed any deviations from the approved AMS-TR-0720R2-A.

The NRC staff conducted a virtual regulatory audit starting February 26, 2025, concluding with a call with the licensee on April 22, 2025, to examine the licensees non-docketed information on the proposed OLM methodologies. The NRC staff did not identify any need for additional information and issued an audit summary dated January 7, 2026 (ML25357A113).

2.0 REGULATORY EVALUATION

2.1

System Description

The licensee provided the following system description in its December 4, 2024, license amendment request (LAR):

The transmitters in the OLM program provide input to the Reactor Protection Systems (RPS), Engineered Safeguards Actuation System (ESAS) and Engineered Safety Feature Actuation Systems (ESFAS) and are used for Post Accident Monitoring (PAM), the Remote Shutdown System, and Low Temperature Overpressure Protection (LTOP).

The RPS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and RCS pressure boundary during anticipated operational occurrences and to assist the Engineered Safety Features Systems in mitigating accidents.

The ESFAS and related systems initiate 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 primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents.

The Remote Shutdown System provides the operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible.

The LTOP controls prevent RCS overpressure at low temperatures, so the integrity of the reactor coolant pressure boundary is not compromised by violating the pressure and temperature limits. LTOP provides the allowable combinations for pressure and temperature during cooldown, shutdown, and heatup to keep from violating the pressure and temperature limits.

Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plants. The changes will not impact how the plants operate. Entergy will use condition-based frequency to determine when transmitter calibrations are needed instead of performing calibrations based on a calendar frequency. Existing calibration methods will be used when it is determined that transmitter calibration is needed.

2.2 Requested Changes The licensee proposed TS changes, which are described below for each Entergy fleet unit.

Additions of text are shown in bolded, underlined text.

2.2.1 ANO-1 Definition Changes As described in the LAR, the licensee proposed changes to the following definition in ANO-1 TS 1.1 Definitions of TS 1.0, Use and Application Definitions.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

The ANO-1 TSs do not have any Definitions for RESPONSE TIME TESTING therefore, no additional changes are required in TS 1.1 Definitions.

2.2.2 ANO-2 Definition Changes As described in the LAR, the licensee proposed changes to the following definitions in ANO-2 TS 1.0, Definitions.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

REACTOR TRIP SYSTEM RESPONSE TIME 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 electrical power is interrupted to the CEA [control element assembly] drive mechanism. 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

ENGINEERED SAFETY FEATURE RESPONSE TIME The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF [engineered safety feature] 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

2.2.3 Grand Gulf Definition Changes As described in the LAR, the licensee proposed changes to the following definitions in Grand Gulf TS 1.1, Definitions.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

Also, on this page (i.e., TS 1.0-3a), in the header, a minor editorial change is proposed. TS 1.1 Definitions is being revised to read, 1.1 Definitions (continued).

END OF CYCLE RECIRCULATION PUMP TRIP (EOC RPT) SYSTEM RESPONSE TIME The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval from initial movement of the associated turbine stop valve or the turbine control valve to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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, except for the breaker arc suppression time, which is not measured but is validated to conform to the manufacturers design value. In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

Also, the following minor editorial changes are proposed to be made on TS page 1.0-5:

In the header, 1.1 Definitions is being adjusted to read, 1.1 Definitions (continued).

For the definition LOGIC SYSTEM FUNCTIONAL TEST which is continued from prior page. The word (continued) is located one line too low, which is being adjusted to be just after TEST to make the definition title cleaner.

In the footer, old amendment number 156 had an errant underline. This underline is being removed.

2.2.4 River Bend Definition Changes As described in the LAR, the licensee proposed changes to the following definitions in River Bend TS 1.1, Definitions.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) SYSTEM RESPONSE TIME The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval from initial movement of the associated turbine stop valve or the turbine control valve to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

TURBINE BYPASS SYSTEM RESPONSE TIME The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components:

a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and

b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.

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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

2.2.5 Waterford Definition Changes As described in the LAR, the licensee proposed changes to the following definitions in Waterford TS 1.0, Definitions.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST (excluding transmitters in the Online Monitoring Program). The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

ENGINEERED SAFETY FEATURES RESPONSE TIME The ENGINEERED SAFETY FEATURES 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 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

An additional minor editorial change is proposed to be made to TS page 1-3. There is an errant

mark that is being removed from below the ENGINEERED SAFETY FEATURES RESPONSE TIME definition.

REACTOR TRIP SYSTEM RESPONSE TIME 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 electrical power to the CEA drive mechanism is interrupted. The response time may be measured by 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 (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

2.2.6 Addition of new Online Monitoring Program (all plants)

The licensee proposed to add a new Online Monitoring Program TS as ANO-1 TS 5.5.19, ANO-2 TS 6.5.21, Grand Gulf TS 5.5.16, River Bend TS 5.5.17, and Waterford TS 6.5.20. The new Online Monitoring Program TS would state:

Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.

The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:

a. Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.

1) Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check.

2) Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance.

3) Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and

4) Documentation of the results of the online monitoring data analysis.

b. Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.

c. Performance of calibration checks for transmitters at the specified backstop frequencies.

d. The provisions of Surveillance Requirement 3.0.3 [Note, for Improved Technical Specifications (ITS) plants, ANO-1, Grand Gulf, and River Bend, 3.0.3 is used, for custom TS plants, ANO-2 and Waterford, 4.0.3 is used] are

applicable to the required calibration checks specified in items a.3, b, and c above.

In section 2.5.6 of the LAR the licensee describes for Waterford item d, above, that the change should apply to SR 4.0.3. However, in the LAR, the licensee provided TS markup pages for Waterford, item d, inadvertently referencing SR 3.0.3 instead of SR 4.0.3.

2.3 Regulatory Evaluation 2.3.1 Applicable Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(1)(ii)(A) requires, in part, that limiting safety system settings (LSSS) are settings for automatic protective devices related to those variables having significant safety functions. Where an LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded. It also requires that the licensee take appropriate action and notify the NRC if the licensee determines that an automatic safety system does not function as required. The licensee is then required to notify the Commission, review the matter and record the results of the review.

The regulation at 10 CFR 50.36(c)(3) states, 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.

The regulation at 10 CFR 50.36(c)(5) states, in part, that [a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, establishes the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission.

Both ANO-1 and ANO-2 construction permits were issued prior to 1971; However, though originally designed to comply with the pre-GDC published in July 1967, their Updated Final Safety Analysis Report (UFSAR) (as amended) address the NRC GDC published as Appendix A to 10 CFR Part 50 in July 1971. Grand Gulf, River Bend, and Waterford construction permits were issued after 1971 and their UFSAR documents conformity to GDC published as Appendix A to 10 CFR Part 50 in July 1971.

The NRC determined that the following GDC are applicable to this review:

GDC 13, Instrumentation and control, states that [i]nstrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

GDC 20, Protection system functions, states that [t]he protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

2.3.2 Applicable Regulatory Guidance NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, Branch Technical Position (BTP) 7-12, Guidance on Establishing and Maintaining Instrument Setpoints, Revision 6, August 2016 (ML16019A200)

NRCs Regulatory Guide (RG) 1.105, Revision 4, Setpoints for Safety-Related Instrumentation, February 2021 (ML20330A329). This RG describes an approach that is acceptable to the NRC staff to meet regulatory requirements to ensure that: (a) setpoints for safety-related instrumentation are established to protect nuclear power plant safety and analytical limits, and (b) the maintenance of instrument channels implementing these setpoints ensures they are functioning as required, consistent with the plant TSs.

RG 1.105 endorses American National Standards Institute (ANSI)/International Society of Automation (ISA) Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation. Among other things, the ANSI/ISA 67.04.01 standard provides criteria for assessing the performance of safety related instrument channels to ensure they remain capable of achieving their required safety functions in a reliable manner. This performance monitoring process requires the establishment of acceptable As-Found tolerance limits used to check whether an instrument channel is functioning as required, and the establishment of acceptable As-Left tolerance limits used to establish the maximum allowed deviation from the desired setpoint of the instrument channel and still be considered as within calibration.

3.0 TECHNICAL EVALUATION

3.1 Description of the OLM Program The proposed ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford OLM programs are based on the AMS OLM TR, AMS-TR-0720R2-A, which provides a methodology for performing OLM of the output signals of pressure and differential pressure transmitters. This methodology was developed by AMS to be used in nuclear power plants as an analytical tool to monitor sensor deviations from proper calibration conditions by monitoring performance of the sensor during plant operation between scheduled refueling outages. The purpose of this monitoring is to flag to plant personnel any sensor performance that has deviated sufficiently from ideal calibration conditions to warrant an engineering evaluation to determine whether a calibration is required to be performed for that channel at the next refueling outage.

3.2 Description and Evaluation of TS Changes The licensees submittal requested approval to implement its OLM program by revising appropriate sections in TS 1.0 USE AND APPLICATION, TS 1.1 Definitions, for ANO-1, River Bend, Grand Gulf and TS 1.0 Definitions for ANO-2, and Waterford. The licensee also adds a new section titled, Online Monitoring Program, for TS 5.5.16, TS 5.5.17, TS 5.5.19, TS 6.5.20, TS 6.5.21, for Grand Gulf, River Bend, ANO-1, Waterford and ANO-2 respectively.

The licensee proposes to use the OLM methodology presented in AMS-TR-0720R2-A as the technical basis to change from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM analysis results. Markup of the TS pages were provided in enclosure attachments 1 through 5 of the licensees submittal.

The regulation at 10 CFR 50.36(a)(1) states, in part, that: [a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

Accordingly, the licensee also submitted, for NRC staff information, the proposed TS Bases changes that correspond to the proposed TS changes. The NRC staff did not review the TS Bases.

3.3 OLM Noise Analysis Implementation The licensee provided, as part of its LAR, the steps to implement the noise analysis technique to assess dynamic failure modes of pressure transmitters. The information provided in sections 3.3.1 through 3.3.6 of the LAR is mapped to the steps found in section 11.3.3, Steps for Implementation of Noise Analysis Technique, of AMS-TR-0720R2-A. The NRC staff finds that the provided mapping and information related to the noise analysis implementation is consistent with the AMS OLM TR, AMS-TR-0720R2-A.

3.4 Use and Application, Definitions TS Section For ANO-1, ANO-2, Grand Gulf, River Bend and Waterford, the TS definition for the term CHANNEL CALIBRATION is being revised to account for the approved OLM methodologies.

The specific change allows transmitters that are included in the licensees OLM program to be excluded from the scope of instrumentation to be periodically calibrated within the frequencies established in their respective TS.

The NRC staff reviewed this proposed change considering the context of the OLM program.

This change is acceptable because the OLM processes would include an acceptable method for identifying performance issues as they occur and initiating corrective actions when pre-established OLM limits are exceeded. The corrective actions would also include performing instrument calibrations as necessary to restore instrument performance to within acceptable parameters. Data collected during OLM activities would be used to adjust OLM limits such that poorly performing instruments would be calibrated at greater frequencies to address any potential impact on long term plant performance.

For ANO-2 and Waterford, the TS definition for the terms ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME and REACTOR TRIP SYSTEM RESPONSE TIME are being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The previous exclusion from response time testing had been based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.

For Grand Gulf and River Bend, the TS definition for the terms EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME, END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) SYSTEM RESPONSE TIME, ISOLATION SYSTEM RESPONSE TIME, and REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME are being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The previous exclusion from response time testing had been based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.

For River Bend, the TS definition for the term TURBINE BYPASS RESPONSE TIME is being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The previous exclusion from response time testing had been based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.

The NRC staff finds these revised definitions to be acceptable because the OLM program will continue to monitor instrument performance and will be capable of detecting instrument degradation or failures that could affect response time performance. The previous definitions for these terms allowed exclusion from response time testing because instrument failures that affect time response would be detectable during the periodic calibration tests and channel check activities. Since the OLM program will retain the capability of detecting and correcting instrument degraded performance or fault conditions, the NRC staff considers this method to be an acceptable and approved methodology to support continued exception of these instruments from time response testing.

3.5 New Online Monitoring Program This new section provides a description of the AMS-based OLM program, including key elements. The new TS stipulates that the OLM program must be implemented in accordance with the NRC-approved AMS OLM TR, AMS-TR-0720R2-A. The NRC staff reviewed the TS description of the OLM program in the LAR and found that it is consistent with the program description provided in the approved AMS OLM TR, AMS-TR-0720R2-A. To verify that the ANO-1, ANO-2, Grand Gulf, River Bend and Waterford programs will be implemented in accordance with the NRC-approved TR, the NRC staff conducted an audit per audit plan dated February 26, 2025 (ML25056A269), with a supplemental audit item issued on February 28, 2025 (ML25351A239), and examined several plant specific reports that documented program implementation activities. These reports are described in the NRC staffs audit summary report dated January 7, 2026 (ML25357A113). The NRC staff audit confirmed that key elements including calculations of OLM limits, amenable transmitters to be included in the OLM program, backstop calculations, noise analysis implementation, maximum sampling rate calculations, OLM coverage of transmitter setpoints and range, drift monitoring, plant procedures for data retrieval, and analysis and capture of data for the OLM program would be implemented as described in AMS-TR-0720R2-A.

The NRC staff also reviewed Entergys responses to each of the Application Specific Action Items (ASAIs) that are contained in section 4.0 of the NRC SE for AMS OLM TR, AMS-TR-0720R2-A. These licensee responses are provided in section 3.4 of the LAR dated December 4, 2024. The NRC staff evaluated these ASAIs in section 3.6 of this SE. The NRC determined that all plant specific actions and the ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford OLM programs would be implemented in conformance with the approved AMS OLM TR, AMS-TR-0720R2-A.

3.6 AMS TR-0720R2-A - ASAIs 3.6.1 ASAI 1 - Evaluation and Proposed Mark-up of Existing Plant Technical Specifications ASAI 1:

When preparing a license amendment request to adopt OLM methods for establishing calibration frequency, licensees should consider markups that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance. Such TS changes would need to include appropriate markups of the TS tables describing limiting conditions for operation and surveillance requirements, the technical basis for the changes, and the administrative programs section.

The licensee provided markups of the applicable TSs that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance for transmitters that are included in the OLM program. Therefore, the NRC staff finds that ASAI 1 is met.

3.6.2 ASAI 2 - Identification of Calibration Error Source ASAI 2:

When determining whether an instrument can be included in the plant OLM program, the licensee shall evaluate calibration error source in order to account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system. Calibration errors identified through OLM should be attributed to the transmitter until testing can be performed on other support devices to correctly determine the source of calibration error and reallocate errors to these other loop components.

The NRC staff performed an audit of the ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford OLM program reports to verify that calibration error sources were being factored into account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system. The NRC staff confirmed that the OLM program attributes calibration errors to the transmitter unless testing is subsequently performed to determine and reallocate calibration error to other instrument loop components. Therefore, the NRC staff finds that ASAI 2 is met.

3.6.3 ASAI 3 - Response Time Test Elimination Basis ASAI 3:

If the plant has eliminated requirements for performing periodic RT [response time] testing of transmitters to be included in the OLM program, then the licensee shall perform an assessment of the basis for RT test elimination to determine if this basis will remain valid upon implementation of the OLM program and to determine if the RT test elimination will need to be changed to credit the OLM program rather than the periodic calibration test program.

The transmitters that are being incorporated into the OLM program were excluded from RT testing. The licensee, therefore, performed an assessment of the basis for RT testing exclusions and determined that the OLM program will continue to support exclusion from RT testing because the OLM methods will detect transmitter failures that would affect RT performance. The basis for this exclusion in the Use and Application, Definitions TS Section is evaluated in section 3.4 of this SE. Therefore, the NRC staff finds that ASAI 3 is met.

3.6.4 ASAI 4 - Use of Calibration Surveillance Interval Backstop ASAI 4:

In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe how they intend to apply backstop intervals as a means for mitigating the potential that a process group could be experiencing undetected common mode drift characteristics.

The NRC staff performed an audit, which included the backstop interval calculations performed for ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford, transmitters being incorporated into the proposed OLM program and confirmed that these calculations were performed in a manner consistent with the processes outlined in the approved AMS OLM TR for determining maximum calibration intervals. Therefore, the NRC staff finds that ASAI 4 is met.

3.6.5 ASAI 5 - Use of Criteria other than in AMS OLM TR for Establishing Transmitter Drift Flagging Limit ASAI 5:

In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe whether they intend to adopt the criteria within the AMS OLM TR for flagging transmitter drift or whether they plan to use a different methodology for determining this limit.

The NRC staff determined that the Entergy proposed OLM program is consistent with the AMS OLM TR, AMS-TR-0720R2-A, and therefore, a different methodology is not being employed.

Therefore, the NRC staff finds the criteria in ASAI 5 are met.

3.7 OLM Program Implementation The licensee stated in section 3.2 of its LAR, that the AMS Bridge and the AMS Calibration Reduction System software programs were developed under AMSs 10 CFR Part 50, Appendix B, compliant Quality Assurance (QA) program. The NRC staff conducted an inspection of AMS to review AMSs implementation of its QA program with respect to the design, testing, and error controls for the AMS Bridge and the AMS Calibration Reduction System software programs.

The NRC staff documented its inspection findings in an inspection report dated March 14, 2025, Nuclear Regulatory Commission Inspection Report of Analysis and Measurement Services No. 99902075/2025-201 (ML25071A181). As stated in this inspection report, the NRC staff determined that AMS is implementing its design control and test control program in accordance with the regulatory requirements of Criterion III, Design Control, and Criterion XI, Test Control, of Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50, for the AMS Bridge and the AMS Calibration Reduction System software programs. In addition, the NRC staff reviewed AMSs processes for controlling software errors and with exception of one minor procedural issue, the NRC staff determined that AMS is implementing its non-conforming materials, parts, or components program in accordance with the regulatory requirements of Criterion XV, Nonconforming Materials, Parts, or Components, of Appendix B to 10 CFR Part 50 for the AMS Bridge and the AMS Calibration Reduction System software programs.

3.8 Technical Summary The NRC staff finds that the licensees proposed implementation of the ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford OLM programs are consistent with the approved AMS OLM TR, AMS-TR -720R2-A. The NRC staff also finds the proposed revision to each units TS 1.0, TS 1.1 and proposed addition of Grand Gulf TS 5.5.16, River Bend TS 5.5.17, ANO-1 TS 5.5.19, Waterford TS 6.5.20, and ANO-2 TS 6.5.21, to be acceptable.

The NRC staff determined that implementation of the proposed OLM program for ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford will continue to support establishment and maintenance of LSSSs associated with the transmitters that are included in the program. These settings will continue to ensure that associated automatic protective actions will correct abnormal situations before safety limits are exceeded. Implementation of the OLM programs at ANO-1, ANO-2, Grand Gulf, River Bend, and Waterford would identify those protection system instrument channels that require recalibration, which helps to ensure that the licensee would take appropriate actions if the licensee determines that an automatic safety system does not function as required. The surveillance requirements relating to test, calibration, and inspection of these transmitters will also continue to ensure that the adequate quality of systems and components is maintained.

Therefore, the NRC staff finds that the requirements of 10 CFR 50.36(c)(1)(ii)(A), and 10 CFR 50.36(c)(3) will continue to be met. Further, 10 CFR 50.36(c)(5) is met by the addition of the new program to the licensees TSs. Additionally, the NRC staff finds that the licensees implementation of the OLM Program in accordance with approved TR AMS-TR-720R2-A will continue to meet the requirements of principal design criteria GDCs 13 and 20 for ANO-1, ANO-2, Grand Gulf, River Bend and Waterford, as documented in their respective UFSARs.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, State officials for Arkansas, Louisiana, and Mississippi were notified of the proposed issuance of the amendments on January 8, 2026. The State officials had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility components located within the restricted area as defined in 10 CFR Part 20 and change 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, published in the Federal Register on February 18, 2025 (90 FR 9739), 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 Contributors: GBlas Rodriquez, NRR FOBrien, NRR TSweat, NRR MMahoney, NRR Date: January 21, 2026

ML26007A265

by eConcurrence **via Email OFFICE NRR/DORL/LPL4/PM*

NRR/DORL/LPL4/LA*

NRR/DEX/EICB/BC**

NAME MMahoney PBlechman SDarbali (JZhao for)

DATE 1/7/2025 1/14/2026 1/5/2025 OFFICE NRR/DSS/STSB/BC*

NRR/DORL/LPL4/BC (A)*

NRR/DORL/LPL/PM*

NAME SMehta MMahoney JDrake DATE 1/15/2026 1/21/2026 1/21/2026

v t e Letter Sequence Approval
EPID:L-2024-LLA-0160, TSTF-493, TSTF-569 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval, Approval, Approval, Approval Other: ML25351A239
MONTHYEAR2025-04-23 [Table View]

ML26007A265

Text

SUBJECT:

Dear Mr. Gustafson:

Enclosures:

Attachment:

Attachment:

Attachment:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

System Description

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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