ML24128A042

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License Amendment Request to Remove Obsolete License Conditions
ML24128A042
Person / Time
Site: Grand Gulf, River Bend, Waterford  Entergy icon.png
Issue date: 05/07/2024
From: Couture P
Entergy Operations
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
CNRO2024-00004
Download: ML24128A042 (1)
v • d • e


Text

Phil Couture Senior Manager Fleet Regulatory Assurance 601-368-5102

Entergy Operations, Inc. 1340 Echelon Parkway, Jackson, MS 39213 CNRO2024-00004 10 CFR 50.90 May 7, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

License Amendment Request to Remove Obsolete License Conditions Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed Facility Operating License No. NPF-29 River Bend Station, Unit 1 NRC Docket No. 50-458 Renewed Facility Operating License No. NPF-47 Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38 In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 50 Section 50.90, "Application for amendment of license, construction permit, or early site permit," Entergy Operations Inc. (Entergy) hereby requests an amendment to the Renewed Facility Operating License Numbers NPF-29, NPF-47, and NPF-38 for Grand Gulf Nuclear Station Unit 1 (GGNS),

River Bend Station Unit 1 (RBS) and Waterford Steam Electric Station Unit 3 (WF3). The proposed amendment would remove License Condition 2.F, which requires the listed Entergy sites to report certain violations of Renewed Facility Operating License Section 2.C within twenty-four hours to the Nuclear Regulatory Commission (NRC) Operations Center via the Emergency Notification System with a written follow-up at a later date. This change is consistent with the notice published in the Federal Register (FR) on November 4, 2005, 70 FR 67202 (Reference 1) as part of the consolidated line-item improvement process (CLIIP).

The Enclosure to this letter provides a description and evaluation of the proposed change for each Entergy unit. Attachments 1, 2, and 3 provide the existing Renewed Facility Operating License pages marked up to show the proposed changes. Attachments 4, 5, and 6 provide revised (re-typed) pages.

Entergy requests approval of the proposed license amendment 13 months after the letter date.

The proposed changes would be implemented within 90 days of issuance of the amendment.

CNRO2024-00004 Page 2 of 3 There are no regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), a copy of this license amendment request, with enclosure, is being provided to the designated State Officials.

Should you have any questions or require additional information please contact me at 601-368-5102.

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

Executed on May 7, 2024.

Respectfully, Phil Couture PC/dlw

Enclosure:

Evaluation of the Proposed Change Attachments to

Enclosure:

1.

GGNS - Renewed Facility Operating License Page Markups

2.

RBS - Renewed Facility Operating License Page Markups

3.

WF3 - Renewed Facility Operating License Page Markups

4.

GGNS - Retyped Renewed Facility Operating License Page

5.

RBS - Retyped Renewed Facility Operating License Page

6.

WF3 - Retyped Renewed Facility Operating License Page Philip Couture Digitally signed by Philip Couture Date: 2024.05.07 07:28:38 -05'00'

CNRO2024-00004 Page 3 of 3 Reference

1. Federal Register - 70 FR 67202, "Notice of Availability of Model Application Concerning Elimination of Typical License Concition Requiring Reporting of Violations of Section 2.C op Operating License Using the Consolidated Linee Item Improvement Process [CLIIP],"

dated November 4, 2005.

cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - GGNS NRC Senior Resident Inspector - RBS NRC Senior Resident Inspector - WF3 NRC Project Manager - GGNS NRC Project Manager - RBS NRC Project Manager - WF3 NRC Project Manager - Fleet Designated State Official - Mississippi Designated State Official - Louisiana

Enclosure CNRO2024-00004 Evaluation of the Proposed Change

CNRO2024-00004 Enclosure Page 1 of 22 TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION............................................................................................... 2 2.0 DETAILED DESCRIPTION................................................................................................ 2 2.1.1 Current Operating Licensing Requirements............................................................ 2 2.1.1.1 GGNS Current Operating Licensing Requirements................................................ 2 2.1.1.2 RBS Current Operating Licensing Requirements.................................................... 9 2.1.1.3 WF3 Current Operating Licensing Requirements................................................. 10 2.1.2 Reason for the Proposed Change......................................................................... 17 2.1.3 Description of the Proposed Change.................................................................... 17

3.0 TECHNICAL EVALUATION

............................................................................................. 17 3.1 General Guidance from the CLIIP................................................................................ 17 3.2 Variations from the CLIIP............................................................................................. 18 3.2.1 Variations from the CLIIP - GGNS........................................................................ 18 3.2.2 Variations from the CLIIP - RBS........................................................................... 19 3.2.3 Variations from the CLIIP - WF3........................................................................... 19

4.0 REGULATORY EVALUATION

......................................................................................... 19 4.1 Applicable Regulatory Requirements/Criteria............................................................... 19 4.2 Precedent..................................................................................................................... 20 4.3 No Significant Hazards Consideration Analysis............................................................ 20 4.4 Conclusions.................................................................................................................. 21

5.0 ENVIRONMENTAL CONSIDERATION

........................................................................... 21

6.0 REFERENCES

................................................................................................................. 21 7.0 ATTACHMENTS.............................................................................................................. 22

CNRO2024-00004 Enclosure Page 2 of 22 EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 50 Section 50.90, "Application for amendment of license, construction permit, or early site permit," Entergy Operations Inc. (Entergy) proposes an amendment to the Renewed Facility Operating License Numbers NPF-29, NPF-47, and NPF-38 for Grand Gulf Nuclear Station Unit 1 (GGNS), River Bend Station Unit 1 (RBS) and Waterford Steam Electric Station Unit 3 (WF3). The proposed amendment would remove License Condition 2.F, which requires the listed Entergy sites to report certain violations of Renewed Facility Operating License Section 2.C within twenty-four hours to the Nuclear Regulatory Commission (NRC) Operations Center via the Emergency Notification System with a written follow-up at a later date. This change is consistent with the notice published in the Federal Register (FR) on November 4, 2005, 70 FR 67202 (Reference 2) as part of the consolidated line-item improvement process (CLIIP).

2.0 DETAILED DESCRIPTION 2.1.1 Current Operating Licensing Requirements The current requirements of the license condition are as follows. Two abbreviations used in several of the items below are, SER - Safety Evaluation Report and SSER - Supplement Safety Evaluation Report.

2.1.1.1 GGNS Current Operating Licensing Requirements Operating License Condition 2.F:

[Entergy Operations, Inc.] EOI shall report any violations of the requirements contained in Section 2, Items C. (1), C.(4) through C.(38) of this renewed license within twenty-four (24) hours. Initial notification shall be made in accordance with the provisions of 10 CFR 50.72 with written follow-up in accordance with the procedures described in 10 CFR 50.73(b), (c),

and (e).

The existing conditions in Section 2.C that are subject to the current reporting requirement consist of the following [(1), C.(4) through C.(38)]:

(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.

(4) Independent Verification of Staff Performance and Other Plant Activities (Section 13.4, SER, SSER #2)

(a) MP&L1 shall establish a subcommittee of the Corporate Safety Review Committee to review and evaluate the:

Note 1 - The original license authorized Mississippi Power & Light Company (MP&L) to operate the facility. Amendment 27 authorized SERI to operate the facility. Amendment 125 resulted in a name change for Mississippi Power

CNRO2024-00004 Enclosure Page 3 of 22

& Light Company (MP&L) to Entergy Mississippi, LLC., which was subsequently changed to Entergy Mississippi LLC.

1. Status and readiness of the plant and systems needed to support intended modes of operation and/or testing;
2. Readiness of personnel to conduct intended operation and testing;
3. Morale and attitudes of plant personnel that have a bearing on safe plant operation;
4. Past performance in plant operations and adherence to procedures and administrative controls;
5. Changes in current organization with regard to experience and qualifications of plant management and supervisory personnel since the last evaluation;
6. Results and effectiveness of the Plant Safety Review Committee (PSRC),
7. Status of plant as compared to other [Boiling Water Reactor] BWR startups based on the subcommittees knowledge and experience.

Reviews shall be conducted prior to exceeding 50 percent of full power and within 30 days following completion of the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> warranty run. The subcommittee shall be composed of a minimum of three professionals not employees of MP&L, with experience which will be responsive to the concerns presented above. In conducting these evaluations, the subcommittee shall conduct interviews of representatives of all levels of plant staff management. The subcommittee shall report directly to the Chairman of the Corporate Safety Review Committee and, in turn, MP&L shall submit the report of these reviews to NRC.

(b) The Plant Safety Review Committee shall review all Unit 1 Preoperational Testing and System Demonstration activities performed concurrently with the Unit 1 Startup Test Program to assure that the activity will not affect the safe performance of the portion of the Unit 1 Startup Program being performed. The review shall address, as a minimum, system interaction, span of control, staffing, procedures, security and health physics, with respect to performance of the activities concurrent with the portion of the Unit 1 Startup Program being performed.

(5) Deferred Preoperational Deficiencies MP&L shall satisfactorily resolve those deficiencies which were deferred from the preoperational testing program on a schedule that shall assure that the capability of a system required to be operable by Technical Specification is not degraded.

(6) Soil Structure Interaction (Section 3.7.1, SER, SSER #2)

Prior to startup following the first refueling outage, MP&L shall complete structural modifications, if required, as a result of the NRC staffs completion of its review of MP&L* responses.

(7) Seismic Instrumentation (Section 3.7.4, SER, SSER #2)

Prior to startup following the first refueling outage, the installation of triaxial strong motion accelerometers on reactor supports shall be completed.

CNRO2024-00004 Enclosure Page 4 of 22 (8) Masonry Walls (Section 3.8.3, SER, SSER #2)

Prior to startup following the first refueling outage, MP&L shall complete structural modifications, if required, as a result of the NRC staffs completion of its review of the MP&L response to IE Bulletin 80-11.

(9) Dynamic Testing (Section 3.9.2, SER, SSER #2, SSER #4, SSER #5)

MP&L shall conduct vibrational measurement and inspection programs during preoperational and initial startup testing in accordance with the guidelines of Regulatory Guide 1.20, "Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing," for prototype reactors. An evaluation report demonstrating satisfactory results shall be provided to the NRC for review and approval no later than 6 months after completion of the startup test program.

(10) Dynamic Qualification (3.10, SER, SSER #1, SSER #2, SSER #4, SSER #5)

(a) Prior to startup following the first refueling outage, MP&L shall complete any modifications or replacement of equipment found necessary as a result of the fatigue evaluation. In the interim, MP&L shall document the occurrence of every safety relief valve actuation into the suppression pool; the associated cumulative damage factors shall be calculated for typical representative equipment and kept up-to-date; and EOI shall report to NRC any malfunction of equipment that occurs due to any safety relief valve discharge.

(b) MP&L shall perform an in-situ test of the High Pressure Core Spray (HPCS) service water pump and evaluate the effects of flow induced vibration on the HPCS service water pump. This evaluation shall be provided to the NRC for review and approval.

Prior to startup following the first refueling outage, MP&L shall complete all modifications as a result of the NRC staffs review of the test results and evaluation.

(c) Prior to actual use in fuel handling operations, MP&L shall qualify the fuel-handling and auxiliary platform, in-vessel rack, and storage container for defective fuel.

(11) Environmental Qualification (Section 3.11, SER; SSER #1; Appendix H, SSER #2; SSER #5)

Prior to March 31, 1985, MP&L shall environmentally qualify all electrical equipment as required by 10 CFR 50.49.

(12) Surveillance of Control Blade (Section 4.2.3.14, SER)

Within 30 days after plant startup following the first refueling outage, System Energy Resources, LLC shall comply with items 1, 2 and3 of Bulletin No. 79-26 and submit a written response to NRC on item 3.

(13) Core Stability Analysis and Prohibition of Natural Circulation (Section 4.4.1, SER)

(a) Prior to startup following the first refueling outage, MP&L shall submit a new core stability analysis for operation beyond cycle 1.

(b) Natural circulation shall be prohibited as an operating mode.

(14) Loose Parts Monitoring (Section 4.4.1, SER)

Prior to startup following the first refueling outage, MP&L shall submit an evaluation of the Loose Parts Monitoring System to address conformance to R.G. 1.133, Rev. 1, dated May 1981.

CNRO2024-00004 Enclosure Page 5 of 22 (15) Scram Discharge Volume (Sections 4.6, SER)

Prior to startup following the first refueling outage, MP&L shall incorporate the following additional modifications into the scram discharge volume system:

(a) Redundant vent and drain valves, and (b) Diverse and redundant scram instrumentation for each instrumented volume, including both delta pressure sensors and float sensors.

(16) Containment Purge (Section 6.2.4, SSER #5)

(Deleted)

(17) Containment Pressure Boundary (Section 6.2.8, SER)

Prior to startup following the first refueling outage, MP&L shall replace the feedwater check valve disc with a disc made from a suitable material.

(18) Pressure Interlocks on Valves Interfacing at Low and High Pressure (Section 6.3.4, SSER #2)

Prior to startup following the first refueling outage, MP&L shall implement isolation protection against over pressurization of the low pressure emergency core cooling systems (RHR/LPCI and LPCS) at the high and low pressure interface containing a check valve and a closed motor-operated valve.

(19) IE Information Notice 79-22, Qualification of Control System (Section 7.8.C, SER, SSER #2)

Prior to startup following the first refueling outage, MP&L shall complete any design changes found necessary as a result of this review.

(20) Standby Service Water System (Section 9.2.1 SER, SSER #2)

(Deleted)

(21) Spent Fuel Pool Ventilation System (Section 9.4.2, SER, SSER #2)

If spent irradiated fuel is placed in the spent fuel pool prior to installation and operability of the safety related backup fuel pool cooling pump room coolers, the plant shall be placed in shutdown condition and remain shut down with the RHR System dedicated to the fuel pool cooling mode.

(22) Remote Shutdown Panel (Section 9.5.4.1, SER, SSER #2)

Prior to startup following the first refueling outage, MP&L shall install electrical isolation switches between the control room and the Division 1 remote shutdown panel.

(23) Fire Protection Program (Section 9.5.9, SER)

Replaced by Paragraph 2.C.(41).

(24) Interplant Communication Systems (Section 9.6.1.2, SER, SSER #2, SSER #4, SSER #5)

Tests of the communication systems used to mitigate the consequences of an event and attain a safe plant shutdown shall be completed during preoperational and startup tests.

An evaluation of the test results shall be provided for NRC review within 90 days after test completion. Any system modifications found necessary as a result of NRC review shall be completed prior to startup following the first refueling outage.

CNRO2024-00004 Enclosure Page 6 of 22 (25) Reliability of Diesel-Generators (Sections 8.3.1, 9.6.3 through 9.6.7, SER, SSER #2, SSER #4, SSER #6)

(a) Prior to startup following the first refueling outage, a heavy duty turbocharger gear drive assembly shall be installed on all EMD diesel-generators.

(26) Turbine Disc Integrity (Section 10.2.1, SER, SSER #1)

(DELETED)

(27) Circulating Water System (Section 10.4.5, SER)

EOI shall not fill the Unit 2 circulating water system (including the natural draft cooling tower basin) until Unit 1 flooding concerns related to this system are resolved to the satisfaction of the NRC staff.

(28) Advisor to the Vice President MP&L shall have on its nuclear operations staff, one or more corporate management officials or advisors (who may be either permanent employees or contracted consultants) who have substantial commercial nuclear power plant operating management experience and who will advise on all decisions affecting safe operation of the plant.

This requirement shall be in effect until the plant has accumulated at least 6 months at power levels above 90 percent of full power.

(29) Operating Shift Advisor (Section 13.1.2, SER)

At least one individual on each operating shift shall have substantive previous BWR operating experience, including startup and shutdown of a BWR and under conditions that one might expect to encounter during the initial startup and power escalation at Grand Gulf plant. This individual is not required to be licensed on Grand Gulf Unit 1 and need not be a MP&L employee, but as a minimum shall be retained on a contract basis to act as a consultant or advisor to the GGNS shift crew. Such an experienced person shall be assigned to each operating shift until the plant achieves and demonstrates full power operation.

(30) Training Instructors (Section 13.2, SER)

[DELETED]

(31) Initial Test Program (Section 14, SER)

MP&L shall conduct the post-fuel-loading initial test program (set forth in Section 14 of the Final Safety Analysis Report, as amended) without making any major modifications of this program unless such modifications have been identified and have received prior NRC approval. Major modifications are defined as:

(a) Elimination of any test identified in Section 14 of the Final Safety Analysis Report, as amended, as being essential; (b) Modification of test objectives, methods or acceptance criteria for any test identified in Section 14 of the Final Safety Analysis Report, as amended, as being essential; (c) Performance of any test at a power level different from that described in the program; and (d) Failure to complete any tests included in the described program (planned or scheduled for power levels up to the authorized power level).

CNRO2024-00004 Enclosure Page 7 of 22 (32) Deleted (33) NUREG-0737 Conditions (Section 22.2)

The following conditions shall be completed to the satisfaction of the NRC. These conditions reference the appropriate items in Section 22.2, "[Three Mile Island] TMI Action Plan Requirements for Applicants for Operating Licenses", in the Safety Evaluation Report and Supplements 1, 2, 3, 4, and 5 to NUREG-0831.

(a) Control Room Design Review (I.D.1, SER; Appendix E, SSER #2, SSER #4, SSER #5)

Prior to startup following the first refueling outage, SERI shall demonstrate the ability to maintain an "effective temperature" condition of 85°F or less in the remote shutdown panel (RSP) room for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with an ambient outdoor temperature of at least 95°F.

(b) Training During Low-Power Testing (I.G.1, SER)

Prior to restart following the first refueling outage, MP&L shall complete the additional training and testing related to TMI Action Plan I.G.1 as described in Section 2.3 of the MP&L submittal dated April 3, 1986.

(c) Deleted (d) Hydrogen Control (Section II.B.7, SER, SSER #2, SSER #3, SSER #4, SSER #5)

(1) During the first cycle of operation, MP&L shall maintain a suitable program of analysis and testing of the installed hydrogen ignition system. EOI shall submit to the NRC quarterly reports on the status of their research programs.

(i) EOI shall amend its research program on hydrogen control measures to include, but not be limited to, the following items:

1. Perform containment sensitivity analysis to determine the adequacy of the hydrogen control system for a spectrum of degraded core accidents including the determination of accident sequences for which equipment survivability is assured;
2. Research to investigate the conditions leading to and consequences resulting from hydrogen combustion in the wetwell and containment.

Testing shall be performed in a larger scale facility such as the one-quarter scale test facility proposed by MP&L; Facility such as the one-quarter scale test facility proposed by MP&L;

3. Research to investigate the conditions leading to and consequences resulting from hydrogen combustion in the drywell;
4. Confirmatory tests on thermal response of selected equipment exposed to hydrogen burns.

(ii) EOI shall perform feasibility studies to examine the options for enhancing equipment survivability for essential equipment located in the vicinity of the suppression pool or other regions subjected to severe environments. The options to be studied in such feasibility studies shall include thermal shielding, additional cooling, and relocation of essential equipment.

CNRO2024-00004 Enclosure Page 8 of 22 (2) (i) EOI shall complete its research program on hydrogen control to show that the hydrogen control system will perform its intended function in a manner that provides adequate safety margins. This research program shall be completed on a schedule which reflects the requirements of 10 CFR 50.44.

(ii) If it is determined that plant modifications are required to obtain NRC approval that an adequate hydrogen control system for Grand Gulf is installed, then these modifications shall be completed on a schedule which is approved by the NRC.

(e) Instrumentation for Detection of Inadequate Core Cooling (II.F.2, SER, SSER #2)

MPL shall submit a report addressing the analysis performed by the BWR Owners Group regarding additional instrumentation relative to inadequate core cooling and shall implement the staffs requirements after the completion of the staffs review of this report. These modifications shall be completed on a schedule acceptable to the staff.

(f) Modification of Automatic Depressurization System Logic - Feasibility for Increased Diversity for Some Event Sequences (II.K.3.18, SER, SSER, #2, SSER #4)

Prior to startup following the first refueling outage, MP&L shall provide, for NRC review, justification for the timer delay settings, revisions to the emergency procedures covering the use of the manual inhibit switch, proposed Technical Specification surveillance procedures for the timer and switch, and shall implement alternative logic modification (Option 4) of the automatic depressurization system.

Manual inhibit switch, proposed Technical Specification surveillance procedures for the timer and switch, and shall implement alternative logic modification (Option 4) of the automatic depressurization system.

(g) Qualification of [Automatic Depressurization System] ADS Accumulators (II.K.3.28, SSER #5)

Prior to startup following the first refueling outage, MP&L shall perform an integrated leak test on the ADS air system, perform sampling to establish instrument air quality, provide instrumentation to monitor ADS air receiver pressure, establish suitable surveillance procedures for the ADS air system and provide proposed changes to the Technical Specifications associated with the surveillance procedures.

(34) [Safety Relief Valves] SRV Test Program (Section A-39, Appendix C, SER, SSER #1, SSER #2)

During Cycle 1, an inplant SRV test program shall be carried out to confirm that the containment building response to SRV loads is acceptable. Results of these tests shall be provided to NRC no later than four months after test completion.

(35) Post- [Loss of Coolant Accident] LOCA Vacuum Breaker Position Indicators Prior to startup following the first refueling outage, MP&L shall install position indicators with redundant indication and alarm in the control room for the check valves associated with the drywell post-LOCA vacuum breakers.

(36) Emergency Response Facilities (Generic Letter 82-33, NUREG-0737 Supplement 1, SSER #5)

EOI shall complete the emergency response capabilities, as required by Attachment 1.

CNRO2024-00004 Enclosure Page 9 of 22 (37) Evaluation of Licensees Technical Specification Problem Sheets (Section 16.3, SSER #6)

Prior to startup following the first refueling outage, MP&L shall implement the following modifications:

(a) Include an emergency override of the test mode of the Division 3 HPCS diesel generator to permit response to emergency signals and to return the control of the diesel generator to the emergency standby mode. (Item No. 333, [Technical Specification] TS. 4.8.1.1.2d.12.b)

(b) Provide the second level undervoltage protection for Division 3 power supply (Item No. 373, [Technical Specification] T.S. Table 3.3.3-2)

(c) Incorporate a bypass or coincident logic in all Division 1 and 2 diesel generator protective trips, except for trips on diesel engine overspeed and generator differential current (Item No. 808, T.S. 4.8.1.1.2.d.16.d).

(38) Control Room Leak Rate (Section 6.2.6, SSER #6)

EOI shall operate Grand Gulf Unit 1 during Modes 1 through 3 with an allowable control room leak rate not to exceed 2000 [cubic feet per minute] cfm (not including ingress/egress leakage of 10 cfm).

2.1.1.2 RBS Current Operating Licensing Requirements Operating License Condition 2.F:

Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, EOI shall report any violations of the requirements contained in Section 2, Items C.(1);

C.(3) through (9); and C.(11) through (16) of this renewed license in the following manner:

initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written follow-up within 60 days in accordance with the procedures described in 10 CFR 50.73(b), (c) and (e).

The existing conditions in Section 2.C that are subject to the current reporting requirement consist of the following:

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.

(3) Antitrust Conditions

a. Entergy Louisiana, LLC shall comply with the antitrust conditions in Appendix C, attached hereto, which is hereby incorporated in this renewed license.
b. EOI shall not market or broker power or energy from River Bend Station, Unit 1.

Entergy Louisiana, LLC is responsible and accountable for the actions of its agent, EOI, to the extent said agents actions affect the marketing or brokering of power or energy from River Bend Station, Unit 1 and, in any way, contravene the antitrust conditions of this paragraph or Appendix C of this renewed license.

(4) DELETED (5) Mark III Related Issues (Section 6.2.1.9, SER and SSER 2)

CNRO2024-00004 Enclosure Page 10 of 22

a. EOI shall not use the residual heat removal system in the steam condensing mode without prior written approval of the staff.
b. DELETED (6) DELETED (7) DELETED (8) DELETED (9) DELETED (11) Operating Staff Experience Requirements (Section 13.1.2.1, SSER 2)

EOI shall have a licensed senior operator on each shift, while in Operating Condition 1, 2 and 3, who has had at least six months of hot operating experience on a plant comparable to River Bend Station, including at least six weeks at power levels greater than 20% of full power, and who has had startup and shutdown experience.

(12) DELETED (13) Partial Feedwater Heating (Section 15.1, SER)

During power operation, the facility shall not be operated with a feedwater heating capacity which would result in a rated thermal power feedwater temperature less than 326 °F.

(14) DELETED (15) DELETED (16) DELETED 2.1.1.3 WF3 Current Operating Licensing Requirements Operating License Condition 2.F:

Except as otherwise provided in the Technical Specifications or the Environmental Protection Plan, EOI shall report any violations of the requirements contained in Section 2.C of this renewed license in the following manner. Initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written follow-up within 30 days in accordance with the procedures described in 10 CFR 50.73(b),

(c) and (e).

The existing conditions in Section 2.C that are subject to the current reporting requirement consist of the following:

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. 270, and the Environmental Protection Plan contained in Appendix B, are hereby

CNRO2024-00004 Enclosure Page 11 of 22 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.

4. DELETED
5. Initial Inservice Inspection Program (Section 6.6, SSER 51)

By June 1, 1985, the licensees2 must submit an initial inservice inspection program for staff review and approval.

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

Note 2: The license originally authorized Entergy Louisiana, LLC to possess, use, and operate the facility. Consequently, certain historical references applicable to Entergy Louisiana, LLC as one of the "licensees" appear in these license conditions.]

6. Environmental Qualification (Section 3.11, SSER 8)

Prior to November 30, 1985, the licensees shall environmentally qualify all electrical equipment according to the provisions of 10 CFR 50.49.

7. Axial Fuel Growth (Section 4.2, SSER 5)

Prior to entering Startup (Mode 2) after each refueling, EOI shall either provide a report that demonstrates that the existing fuel element assemblies (FEA) have sufficient available shoulder gap clearance for at least the next cycle of operation, or identify to the NRC and implement a modified FEA design that has adequate shoulder gap clearance for at least the next cycle of operation. This requirement will apply until the NRC concurs that the shoulder gap clearance provided is adequate for the design life of the fuel.

8. Emergency Preparedness (Section 13.3, SSER 8)

In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

9. Fire Protection EOI shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated November 17, 2011 (and supplements dated January 26, 2012, September 27, 2012, October 16, 2012, May 16, 2013, June 26, 2013, December 18, 2013, June 11, 2014, March 12, 2015, April 10, 2015, May 14, 2015,

CNRO2024-00004 Enclosure Page 12 of 22 August 27, 2015, September 8, 2015, September 24, 2015, October 13, 2015, and January 18, 2016), and as approved in the safety evaluation dated June 27, 2016.

Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c),

and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, EOI may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in [National Fire Protection Association]

NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x10-7/year (yr) for [Core Damage Frequency] CDF and less than 1x10-8/yr for [Large Early Release Frequency] LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval (1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. EOI may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. EOI may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is

CNRO2024-00004 Enclosure Page 13 of 22 adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

  • "Fire Alarm and Detection Systems" (Section 3.8);
  • "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);
  • "Gaseous Fire Suppression Systems" (Section 3.10); and
  • "Passive Fire Protection Features" (Section 3.11).

This condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

(2) Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to EOI 's fire protection program that have been demonstrated to have no more than a minimal risk impact.

EOI may use its screening process as approved in the NRC safety evaluation dated June 27, 2016, to determine that certain fire protection program changes meet the minimal criterion. EOI shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) and (3) below, risk-informed changes to EOI 's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

(2) EOI shall implement the modifications to its facility, as described in Attachment S, Table S-1 "Plant Modifications" in Entergy Operations, Inc. letter W3F1-2016-0003, dated January 18, 2016, to complete the transition to full compliance with 10 CFR 50.48(c) by completion of the first refueling outage greater than 12 months following issuance of the license amendment.

EOI shall maintain appropriate compensatory measures in place until completion of the modification listed as S 1-5, (Installation of qualified1-hour ERFBS fire wrap barrier in Fire Area RAB 6). All other modifications listed in Table S-1 are either installed or have no associated compensatory measure.

(3) EOI shall implement the items listed in Attachment S, Table S-2, "Implementation Items," in Entergy Operations, Inc. letter W3F1-2016-0003, dated January 18, 2016, within 6 months following issuance of the license amendment.

10. Post-Fuel-Loading Initial Test Program (Section 14, SSER 10)

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

11. Emergency Response Capabilities (Section 22, SSER 8)

EOI shall comply with the requirements of Supplement 1 to NUREG-0737 for the conduct of a Detailed Control Room Design Review (DCRDR). Prior to May 1, 1985, EOI

CNRO2024-00004 Enclosure Page 14 of 22 shall submit for staff review and approval the DCRDR Summary Report, including a description of the process used in carrying out the function and task analysis performed as a part of both the DCRDR and the Procedures Generation Package efforts.

12. Reactor Coolant System (RCS) Depressurization Capability (Section 5.4.3, SSER 8)

By June 18, 1985, the licensees shall submit the results of confirmatory tests regarding the depressurization capability of the auxiliary pressurizer spray (APS) system. This information must demonstrate that the APS system can perform the necessary depressurization to meet the steam generator single-tube rupture accident acceptance criteria (SRP 15.6.3) with loop charging isolation valve failed open. Should the test results fail to demonstrate that the acceptance criteria are met, the licensees must provide for staff review and approval, justification for interim operation, and a schedule for corrective actions.

13. Response to Salem [Anticipated Transient Without SCRAM] ATWS Event (Section 7.2.9, SSER 8)

The licensees shall submit responses and implement the requirements of Generic Letter 83-28 on a schedule which is consistent with that given in the licensee's letter of May 30, 1984.

14. DELETED
15. Qualification of Personnel (Section 13.1.3, SSER 8)

EOI shall have on each shift operators who meet the requirements described in. Attachment 2 is hereby incorporated into this renewed license.

16. Operational [Quality Assurance] QA Enhancement Program (SSER 9)

The items listed below shall be completed on the scheduled indicated.

(a) Prior to completion of Phase Ill of the Waterford 3 startup test program, the licensees shall conduct a comprehensive audit of the Operational QA Program that will include a summary QA document of the Operational QA Program, the definition of responsibilities and interfaces, and guidance on the location of information on QA matters at all levels of concern.

(b) Prior to completion of Phase Ill of the Waterford 3 startup test program, the licensees shall supplement its existing QA training program to incorporate specific discussion of QA problems experienced during construction and how this experience applies to operational activities.

(c) Prior to completion of Phase Ill of the Waterford 3 startup test program, the licensees shall address each of the recommendations in the Task Force Support Group (TFSG) Limited Scope Audit Report of LP&L Operational Quality Assurance Program, dated December 4, 1984.

(d) Prior to completion of Phase Ill of the Waterford 3 startup test program, the licensees shall complete corrective actions related to the 23 NRC issues as identified in the LP&L responses.

17. Basemat EOI shall comply with its commitments to perform a basemat cracking surveillance program and additional confirmatory analyses of basemat structural strength as

CNRO2024-00004 Enclosure Page 15 of 22 described in its letter of February 25, 1985. Any significant change to this program shall be reviewed and approved by the NRC staff prior to its implementation.

18. Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and Guidance
2. Assessment of mutual aid fire fighting assets
3. Designated staging areas for equipment and materials
4. Command and control
5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:
1. Protection and use of personnel assets
2. Communications
3. Minimizing fire spread
4. Procedures for implementing integrated fire response strategy
5. Identification of readily-available pre-staged equipment
6. Training on integrated fire response strategy
7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:
1. Water spray scrubbing
2. Dose to onsite responders
19. Control Room Envelope Habitability Program Upon implementation of Amendment No. 218 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 6.5.17, in accordance with TS 6.5.17.c.(i), the assessment of CRE habitability as required by Specification 6.5.17.c.(ii), and the measurement of CRE pressure as required by Specification 6.5.17.d, shall be considered met. Following implementation:

(a) The first performance of SR 6.5.17, in accordance with Specification 6.5.17.c.(i),

shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 4.0.2, as measured from April 17, 2004, the date of the most recent successful tracer gas test, as stated in the October 8, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 6.5.17.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 4.0.2, as measured from April 17, 2004, the date of the most recent successful tracer gas test, as stated in the October 8, 2004 letter response to Generic Letter 003-01, or within

CNRO2024-00004 Enclosure Page 16 of 22 the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 6.5.17.d, shall be within 18 months, plus the 138 days allowed by SR 4.0.2, as measured from August 13, 2008, the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

20. Control Element Assembly Drop Time Curve Validation (Amendment 246)

Prior to Cycle 21 Mode 2 operation, the licensees shall verify the control element assembly drop time test data demonstrates faster control element assembly drop times than the drop time curve provided in Table 15.0-5 of the Final Safety Analysis Report, as amended.

21. License Renewal License Conditions (a) The information in the FSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the "Safety Evaluation Report Related to the License Renewal of Waterford Steam Electric Station Unit 3," are collectively the "License Renewal FSAR Supplement." This Supplement is henceforth part of the FSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, EOI may make changes to the programs, activities, and commitments described in this Supplement, provided the EOI evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, "Changes, Tests, and Experiments," and otherwise complies with the requirements in that section.

(b) The License Renewal FSAR Supplement, as defined in license condition 21(a) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

(1) EOI shall implement those new programs and enhancements to existing programs no later than 6 months before the PEO.

(2) EOI shall complete those activities by the 6 month date prior to the PEO or to the end of the last refueling outage before the PEO, whichever occurs later.

(3) EOI shall notify the NRC in writing within 30 days after having accomplished item (b)(1) above and include the status of those activities that have been or remain to be completed in item (b)(2) above.

22. 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants Entergy is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other

CNRO2024-00004 Enclosure Page 17 of 22 external hazards except seismic; and the alternative seismic approach as described in Entergy's submittal letter dated December 18, 2020, and all its subsequent associated supplements as specified in License Amendment No. 269 dated November 30, 2022.

Entergy will complete closure of the four Human Reliability Analysis (HRA) Finding level Facts and Observations (F&Os) identified as Finding Numbers HR 1-2, HR 7-1, HR 7-3, and HR 7-4, in Table A3-2 of Entergy letter to NRC, dated April 25, 2022, and in Table E2-2 of Entergy letter to NRC, dated May 16, 2022, using an accepted NRC process (Nuclear Energy Institute (NEI) Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13) prior to implementation of 10 CFR 50.69 and the risk-informed completion time (RICT) program.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

2.1.2 Reason for the Proposed Change Section 2.F of the Renewed Facility Operating Licenses requires violations of the specified conditions included in Section 2.C of the Renewed Facility Operating License to be reported to the NRC. The NRC's requirements for immediate notification with written follow-up requirements (Licensee Event Reports) of events at operating nuclear power plants are contained in the 10 CFR 50.72 regulation and the 10 CFR 50.73 regulation. A difference in reporting requirements can lead to variations in reporting.

An operating license improvement was announced in the Federal Register 70 FR 67202 on November 4, 2005 (Reference 2) to eliminate the license condition requiring reporting of violations of Section 2.C as part of the consolidated line item improvement process (CLIIP).

The model safety evaluation (SE) and a model no significant hazards consideration (NSHC) determination relating to the elimination of this license condition were issued in a Federal Register 70 FR 51098 notice on August 29, 2005 (Reference 1).

2.1.3 Description of the Proposed Change Consistent with the CLIIP Notice of Availability in 70 FR 67202 (Reference 2), the proposed amendment will delete Section 2.F of Renewed Facility Operating License Numbers NPF-29, NFP-47, and NPF-38 for GGNS, RBS, and WF3. In place of the current content outlined in Section 2.1.1 above, Section 2.F of the operating license will state "Deleted."

3.0 TECHNICAL EVALUATION

3.1 General Guidance from the CLIIP Entergy has reviewed the model SE published in 70 FR 51098, (Reference 1) as part of the CLIIP Notice of Opportunity to Comment and concluded that the justifications presented in the SE prepared by the NRC staff are applicable to the listed Entergy Sites. The model safety evaluation notes that this license condition is typically in Condition 2.[X]. For the Entergy sites, this is Condition 2.F. Per this notice, the NRC states:

For those cases where the current Renewed Facility Operating License requirement to report violations is also reportable in accordance with the regulations defined in

CNRO2024-00004 Enclosure Page 18 of 22 10 CFR 50.72 and 10 CFR 50.73, the NRC staff finds that the regulations adequately address this issue and the elimination of the duplicative requirement in the Renewed Facility Operating License is acceptable.

Some of the conditions addressed in Section 2.[C] of the Renewed Facility Operating License may address the maintenance of particular programs, administrative requirements, or other matters where a violation of the requirement would not result in a report to the NRC in accordance with 10 CFR 50.72 or 10 CFR 50.73. In most cases, there are requirements for reports to the NRC related to these conditions in other regulations, the specific license condition or technical specification, or an NRC-approved program document. In other cases, there are reports to other agencies or news releases that would prompt a report to the NRC (in accordance with 10 CFR 50.72(b)(2)(xi)). The NRC staff also assessed violations of administrative requirements that could be reportable under the current License Condition but that may not have a duplicative requirement in a regulation or other regulatory requirement. The NRC staff finds that the requirements to report such problems within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with written reports to follow using the licensee event reports (LER) process is not needed. The NRC staff is confident that the information related to such violations that is actually important to the NRCs regulatory functions would come to light in a time frame comparable to the 60-day LER requirements. The information would become available to the appropriate NRC staff through the inspection program, updates to program documents, resultant licensing actions, public announcements, or some other reliable mechanism.

Therefore, the elimination of Section 2.F of the listed Entergy Renewed Facility Operating Licenses will not result in a loss of information to the NRC that would adversely affect either its goal to protect public health and safety or its ability to carry out its various other regulatory responsibilities.

3.2 Variations from the CLIIP There are minor variations in the operating licenses for each site when compared to the CLIIP.

The variations in the operating license do not affect the meaning or applicability of the CLIIP to each site. The variations for each site are listed below.

3.2.1 Variations from the CLIIP - GGNS The subject License Condition being evaluated for deletion is denoted at 2.[X] in the CLIIP. For GGNS this License Condition is numbered as 2.F. This is administrative and does not impact the applicability of the CLIIP.

The wording of the GGNS License Condition 2.F is slightly different than the wording used within the model SE. GGNS License Condition 2.F requires reporting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> similar to the model SE. The GGNS license condition requires follow-up in accordance with 10 CFR 50.73(b), (c), and (e). The model SE condition species a 30-day follow-up in accordance with 10 CFR 50.73. The model SE also calls out changes to the reporting regulations in 10 CFR 50.72 and 50.73 becoming effective in January 2001 per 65 FR 63769 (Reference 3) and included extending the allowable reporting times for licensee event reports (LERs) from 30 days to 60 days. This is administrative and does not impact the applicability of the CLIIP.

The model SE specifies that Section 2.C in its entirety would be included in the reporting requirements of Condition 2.F. The GGNS License specifies a limited number of items

CNRO2024-00004 Enclosure Page 19 of 22 within 2.C that are applicable. The proposed license amendment aligns all reporting requirements with 10 CFR 50.72 and 10 CFR 50.73. The variation in wording does not affect the applicability of the CLIIP to GGNS.

3.2.2 Variations from the CLIIP - RBS The subject License Condition being evaluated for deletion is denoted at 2.[X] in the CLIIP. For RBS this License Condition is numbered as 2.F. This is administrative and does not impact the applicability of the CLIIP.

The wording of the RBS License Condition 2.F is slightly different than the wording used within the model SE. RBS License Condition 2.F requires reporting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> similar to the model SE. The RBS license condition requires follow-up within 60 days in accordance with 10 CFR 50.73(b), (c), and (e) while the model SE condition species a 30-day follow-up. The model SE also calls out changes to the reporting regulations in 10 CFR 50.72 and 50.73 becoming effective in January 2001 per 65 FR 63769 (Reference 3) and included extending the allowable reporting times for licensee event reports (LERs) from 30 days to 60 days. The RBS license condition 2.F follow-up time already aligns with the model SE. The variation in wording does not affect the applicability of the CLIIP to RBS.

The model SE specifies that Section 2.C in its entirety would be included in the reporting requirements of Condition 2.F. The RBS License specifies a limited number of items within 2.C that are applicable. The proposed license amendment aligns all reporting requirements with 10 CFR 50.72 and 10 CFR 50.73 3.2.3 Variations from the CLIIP - WF3 The subject License Condition being evaluated for deletion is denoted at 2.[X] in the CLIIP. For WF3 this License Condition is numbered as 2.F. This is administrative and does not impact the applicability of the CLIIP.

The wording of the WF3 License Condition 2.F is slightly different than the wording used within the model SE. WF3 License Condition 2.F requires reporting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> similar to the model SE. The WF3 license condition requires follow-up within 30 days in accordance with 10 CFR 50.73(b), (c), and (e). The model SE condition species a 30-day follow-up in accordance with 10 CFR 50.73. The model SE also calls out changes to the reporting regulations in 10 CFR 50.72 and 50.73 becoming effective in January 2001 per 65 FR 63769 (Reference 3) and included extending the allowable reporting times for licensee event reports (LERs) from 30 days to 60 days. Therefore, moving from a 30-day follow-up to a 60-day follow-up is consistent with the model SE.

This is administrative and does not impact the applicability of the CLIIP.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The reporting of emergencies, unplanned events or conditions, and other special cases is addressed within the reporting requirements in 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors," and 10 CFR 50.73, "Licensee event report system." Per 65 FR 63769 (Reference 3) changes to the reporting regulations in 10 CFR 50.72 and 50.73 became effective in January 2001.

CNRO2024-00004 Enclosure Page 20 of 22 The proposed change is consistent with the notice published in the Federal Register 70 FR 67202 (Reference 2) as part of the CLIIP.

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

4.2 Precedent By letter dated February 24, 2021 (Reference 4), Duke Energy submitted a License Amendment Request (LAR) for Shearon Harris Nuclear Power Plant that included a change to the Renewed Facility Operating License to delete License Condition 2.G (consistent with Entergy sites Condition 2.F). This change was requested as part of the consolidated line item improvement process and consistent with the model safety evaluation published in the Federal Register 70 FR 67202 (Reference 2). The NRC approved the Duke request by letter dated March 10, 2022 (Reference 5).

4.3 No Significant Hazards Consideration Analysis Entergy Operations Inc. (Entergy) proposes an amendment to the Renewed Facility Operating License Numbers NPF-29, NPF-47, and NPF-38 for Grand Gulf Nuclear Station Unit 1 (GGNS),

River Bend Station Unit 1 (RBS) and Waterford Steam Electric Station Unit 3 (WF3). The proposed amendment would remove License Condition 2.F which requires the listed Entergy sites to report certain violations of Operating License Section 2.C within twenty-four hours to the Nuclear Regulatory Commission (NRC) Operations Center via the Emergency Notification System with a written follow-up within the specified period of time.

Entergy has reviewed the proposed no significant hazards consideration determination published in 70 FR 51098 (Reference 1), as part of the CLIIP Notice of Opportunity - to Comment. Entergy has concluded that the proposed determination presented in the notice is applicable to GGNS, RBS and WF3.

As required by 10 CFR 50.91(a), Entergy's analysis of the issue of no significant hazards consideration is presented below:

1.

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

Response: No The proposed change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices.

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

2.

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

CNRO2024-00004 Enclosure Page 21 of 22 Response: No The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility.

Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

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

Response: No The proposed change deletes a reporting requirement. The change does not affect plant equipment or operating practices.

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

Based upon the reasoning presented above, Entergy concludes that the requested change involves no significant hazards consideration, as set forth in 10 CFR 50.92(c), "Issuance of Amendment."

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

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment changes recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Federal Register - 70 FR 51098, "Notice of Opportunity To Comment on Model Safety Evaluation on Elimination of Typical License Condition Requiring Reporting of Violations of Section 2.C of Operating License Using the Consolidated Line Item Improvement Process,"

dated August 29, 2005.

2. Federal Register - 70 FR 67202, "Notice of Availability of Model Application Concerning Elimination of Typical License Condition Requiring Reporting of Violations of Section 2.C op

CNRO2024-00004 Enclosure Page 22 of 22 Operating License Using the Consolidated Line Item Improvement Process [CLIIP]," dated November 4, 2005.

3. Federal Register - 65 FR 63769, "[Issuing of Final Rule for 10 CFR 50 and 72 Reporting Requirements,] dated October 25, 2000.
4. Duke Energy Letter to NRC, "[Shearon Harris] License Amendment Request to Remove Extraneous Content and Requirements from the Operating License and Technical Specifications," ML21055A819, dated February 24, 2021.
5. NRC Letter to Duke Energy, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 192 Regarding Removal of Extraneous Content and Requirements from the Renewed Facility Operating License and Technical Specifications, ML22020A007, dated March 10, 2022.

7.0 ATTACHMENTS

1.

GGNS - Renewed Facility Operating License Page Markups

2.

RBS - Renewed Facility Operating License Page Markups

3.

WF3 - Renewed Facility Operating License Page Markups

4.

GGNS - Retyped Renewed Facility Operating License Page

5.

RBS - Retyped Renewed Facility Operating License Page

6.

WF3 - Retyped Renewed Facility Operating License Page

Enclosure, Attachment 1 CNRO2024-00004 GGNS - Renewed Facility Operating License Page Markups (1 Page Follows)

Enclosure, Attachment 2 CNRO2024-00004 RBS - Renewed Facility Operating License Page Markups (1 Page Follows)

Enclosure, Attachment 3 CNRO2024-00004 WF3 - Renewed Facility Operating License Page Markups (1 Page Follows)

Enclosure, Attachment 4 CNRO2024-00004 GGNS - Retyped Renewed Facility Operating License Page (1 Page Follows)

Enclosure, Attachment 5 CNRO2024-00004 RBS - Retyped Renewed Facility Operating License Page (1 Page Follows)

Enclosure, Attachment 6 CNRO2024-00004 WF3 - Retyped Renewed Facility Operating License Page (1 Page Follows)

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