Text

Units, 1 and 2 - Issuance of Amendments to Revise Technical Specifications to Use Framatome Gaia Fuel (EPID L-2024-LLA-0072) (NON-PROPRIETARY)
	ML25153A008
Person / Time
Site:	Byron, Braidwood ; (NPF-037, NPF-066, NPF-072, NPF-077)
Issue date:	08/21/2025
From:	Scott Wall; Plant Licensing Branch III
To:	Rhoades D; Constellation Energy Generation
	Wall, S
References
	EPID L-2024-LLA-0072
	Download: ML25153A008 (1)
	v • d • e

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION August 21, 2025 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2; AND BYRON STATION, UNITS, 1 AND 2 - ISSUANCE OF AMENDMENTS TO REVISE TECHNICAL SPECIFICATIONS TO USE FRAMATOME GAIA FUEL (EPID L-2024-LLA-0072)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed amendments (listed below) in response to the Constellation Energy Generation, LLC (CEG) application dated May 28, 2024, as supplemented by letters dated April 21, 2025; May 1, 2025, and August 6, 2025:

1.

Amendment No. 242 to Renewed Facility Operating License No. NPF-72 and Amendment No. 242 to Renewed Facility Operating License No. NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood), respectively;

2.

Amendment No. 240 to Renewed Facility Operating License No. NPF-37 and Amendment No. 240 to Renewed Facility Operating License No. NPF-66 for Byron Station, Units 1 and 2 (Byron), respectively.

The proposed amendment would revise the Braidwood and Byron technical specifications (TSs) to allow loading of Framatome, Inc GAIA fuel with M5Framatome as a fuel cladding material. Since the GAIA fuel uses M5Framatome fuel rod cladding, the licensee included a Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46 and 10 CFR Part 50, Appendix K exemption request as a part of the license amendment request. The staff reviewed the exemption request in a separate safety evaluation (SE), dated August 21, 2025.

The NRC staff has determined that the related SE contains proprietary information pursuant to 10 CFR 2.390, Public inspections, exemptions, request for withholding. The proprietary information is indicated by bold text enclosed with ((double brackets)). The proprietary version of the SE is provided as enclosure 5. Accordingly, the NRC staff has also prepared a non-proprietary version of the SE, which is provided as enclosure 6.

to this letter contains proprietary information. When separated from, this document is DECONTROLLED.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION A copy of the NRC staffs Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Scott P. Wall, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.: STN 50-456, STN 50-457, STN 50-454, and STN 50-455

Enclosures:

1. Amendment No. 242 to NPF-72

2. Amendment No. 242 to NPF-77

3. Amendment No. 240 to NPF-37

4. Amendment No. 240 to NPF-66

5. Safety Evaluation (Proprietary)

6. Safety Evaluation (Non-Proprietary) cc: Listserv

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 242 Renewed License No. NPF-72

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC dated May 28, 2024, as supplemented by letters dated April 21, 2025, May 1, 2025, and August 6, 2025, 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 renewed 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-72 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 242 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of its date of issuance and shall be implemented during the defuel window in the Fall Refueling Outage 2025.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 21, 2025

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 242 Renewed License No. NPF-77

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC dated May 28, 2024, as supplemented by letters dated April 21, 2025, May 1, 2025, and August 6, 2025,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 renewed 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-77 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 242 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of its date of issuance and shall be implemented during the defuel window in the Spring Refueling Outage 2026.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 21, 2025

ATTACHMENT TO LICENSE AMENDMENT NOS. 242 AND 242 RENEWED FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Renewed Facility Operating Licenses and 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 Licenses REMOVE INSERT License NPF-72 License NPF-72 Page 3 Page 3 License NPF-77 License NPF-77 Page 3 Page 3 Technical Specifications REMOVE INSERT 2.0 - 1 2.0 - 1 2.0 - 2 2.0 - 2 3.2.1 - 1 3.2.1 - 1 3.2.1 - 2 3.2.1 - 2 3.2.1 - 4 3.2.1 - 4 3.2.1 - 5 3.2.1 - 5 3.5.1 - 2 3.5.1 - 2 3.5.4 - 2 3.5.4 - 2 4.0 - 1 4.0 - 1 5.5 - 20 5.5 - 20 5.6 - 3 5.6 - 3 5.6 - 4 5.6 - 4 5.6 - 5 5.6 - 5 5.6 - 6 5.6 - 6 5.6 - 7 5.6 - 7 5.6 - 8 5.6 - 8 5.6 - 9

(2)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)

Constellation Energy Generation, LLC, 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; (4)

Constellation Energy Generation, LLC, 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 (5)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 242 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-72 Amendment No. 242

(2)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)

Constellation Energy Generation, LLC, 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; (4)

Constellation Energy Generation, LLC, 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 (5)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 242 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-77 Amendment No. 242

SLs 2.0 BRAIDWOOD UNITS 1 & 2 2.0 1 Amendment 242 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded.

2.1.1.1 For Westinghouse fuel in MODE 1, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.24 for the WRB-2 DNB correlation for a thimble cell, 1.25 for the WRB-2 DNB correlation for a typical cell and 1.19 for the ABB-NV DNB correlation for a thimble cell and a typical cell.

2.1.1.2 For Westinghouse fuel in MODE 2, the DNBR shall be maintained 1.17 for the WRB-2 DNB correlation, and 1.13 for the ABB-NV DNB correlation and 1.18 for the WLOP DNB correlation.

2.1.1.3 For Westinghouse fuel in MODES 1 and 2, the peak fuel centerline temperature shall be maintained < 5080F decreasing by 58F per 10,000 MWD/MTU burnup.

2.1.1.4 For Framatome fuel, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.12 for the ORFEO-GAIA DNB correlation and 1.15 for the ORFEO-NMGRID DNB correlation.

2.1.1.5 For Framatome fuel, for UO2 fuel, the peak centerline temperature shall be maintained 5090 F, decreasing by 13.7 F per 10,000 MWD/MTU burnup.

2.1.1.6 For Framatome fuel, for UO2-Gd2O3 fuel, the peak centerline temperature shall be maintained 5090 F -

9360*E2 - 399.6*E, decreasing by 13.7 F per 10,000 MWD/MTU burnup where E is the gadolinia weight-fraction.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 psig.

(continued)

SLs 2.0 BRAIDWOOD UNITS 1 & 2 2.0 2 Amendment 242 2.0 SAFETY LIMITS (SLs) (continued) 2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 1 Amendment 242 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by FQ C(Z) and FQ V(Z), shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE---------

Required Action A.4 shall be completed whenever this Condition is entered.

FQ C(Z) not within limit.

A.1 Reduce THERMAL POWER 1% RTP for each 1% FQ C(Z) exceeds limit.

AND A.2 Reduce Power Range Neutron Flux-High trip setpoints 1%

for each 1% FQ C(Z) exceeds limit.

15 minutes after each FQ C(Z) determination 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FQ C(Z) determination AND A.3 Reduce Overpower T trip setpoints 1%

for each 1% FQ C(Z) exceeds limit.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FQ C(Z) determination AND A.4 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing THERMAL POWER above the limit of Required Action A.1 (continued)

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 2 Amendment 242 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. ---------NOTE---------

Required Action B.2 shall be completed whenever this Condition is entered.

B.1 Reduce AFD limits as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months FQ V(Z) not within limit.

B.2 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing AFD limits above the limits of Required Action B.1 C. Required Action and associated Completion Time not met.

C.1 Be in MODE 2.

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

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 4 Amendment 242 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2

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

If FQ V(Z) measurements indicate that either the maximum over z [FQ C(Z)/ K(Z)]

OR maximum over z [FQ V(Z)/ K(Z)]

has increased since the previous evaluation of FQ C(Z):

a. Increase FQ V(Z) by the greater of a factor of [1.02] or the appropriate factor specified in the COLR and reverify FQ V(Z) is within limits specified in the COLR; or

b. Repeat SR 3.2.1.2 once per 7 EFPD until either a. above is met or two successive flux maps indicate that the maximum over z [FQ C(Z)/ K(Z)]

and maximum over z [FQ V(Z)/ K(Z)]

has not increased.

(continued)

FQ(Z) 3.2.1 BRAIDWOOD UNITS 1 & 2 3.2.1 5 Amendment 242 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued)

Verify FQ V(Z) is within limit specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQ V(Z) was last verified AND In accordance with the Surveillance Frequency Control Program

Accumulators 3.5.1 BRAIDWOOD UNITS 1 & 2 3.5.1 2 Amendment 242 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water level in each accumulator is 31% and 63%.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is 602 psig and 647 psig.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each accumulator is 2600 ppm and 2900 ppm.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.5

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

Only required to be performed for affected accumulators after each solution volume increase of 10% of indicated level that is not the result of addition from the refueling water storage tank containing a boron concentration 2600 ppm and 2900 ppm.

Verify boron concentration in each accumulator is 2600 ppm and 2900 ppm.

Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SR 3.5.1.6 Verify power is removed from each accumulator isolation valve operator.

In accordance with the Surveillance Frequency Control Program

RWST 3.5.4 BRAIDWOOD UNITS 1 & 2 3.5.4 2 Amendment 242 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1

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

Only required to be performed when ambient air temperature is < 35F or > 100F.

Verify RWST borated water temperature is 35F and 100F.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.2

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

Only required to be performed when ambient air temperature is < 35F.

Verify RWST vent path temperature is 35F.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST borated water level is 89%.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.4 Verify RWST boron concentration is 2700 ppm and 2900 ppm.

In accordance with the Surveillance Frequency Control Program

Design Features 4.0 BRAIDWOOD UNITS 1 & 2 4.0 1 Amendment 242 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Site Location The site is located in Reed Township, approximately 20 mi (32 km) south-southwest of the city of Joliet in northern Illinois.

4.1.2 Exclusion Area Boundary (EAB)

The EAB shall not be less than 1591 ft (485 meters) from the outer containment wall.

4.1.3 Low Population Zone (LPZ)

The LPZ shall be a 1.125 mi (1811 meter) radius measured from the midpoint between the two reactors.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, Optimized ZIRLOTM, or M5Framatome clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium, hafnium, or a mixture of both types.

Programs and Manuals 5.5 BRAIDWOOD UNITS 1 & 2 5.5 20 Amendment 242 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42.8 psig for Unit 1 and 38.7 psig for Unit 2 The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Containment leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

< 0.60 La for the Type B and C tests and < 0.75 La for Type A tests; and

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 3 Amendment 242 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

SL 2.1.1, "Reactor Core SLs";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Moderator Temperature Coefficient";

LCO 3.1.4, "Rod Group Alignment Limits";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor

)

F

( N H

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation";

LCO 3.3.9, "Boron Dilution Protection System (BDPS)";

LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration";

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluations Methodology," July 1985.

2.

Not Used.

3.

NFSR-0016, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," July 1983.

4.

NFSR-0081, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," July 1990.

5.

Not used.

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 4 Amendment 242 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6.

WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

7.

Not Used.

8.

Not Used.

9.

WCAP-10216-P-A, Revision 1, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.

10.

WCAP-8745-P-A, "Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions,"

September 1986.

11.

WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.

12.

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995, (Westinghouse Proprietary).

13.

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM," July 2006, (Westinghouse Proprietary).

14.

ANP-10297P-A Revision 0, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results," February 2013.

15.

ANP-10297 Revision 0 Supplement 1P-A Revision 1, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results," December 2020.

16.

ANP-10311P-A Revision 1, "COBRA-FLX: A Core Thermal-Hydraulic Analysis Code," October 2017.

17.

ANP-10341P-A Revision 0, "The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations,"

September 2018.

18.

XN-75-32(P)(A) Supplements 1, 2, 3, & 4, "Computational Procedure for Evaluating Fuel Rod Bowing," February 1983.

19.

ANP-10342P-A Revision 0, "GAIA Fuel Assembly Mechanical Design," September 2019.

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 5 Amendment 242 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20.

ANP-10323P-A Revision 1, "GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors," November 2020.

21.

BAW-10227P-A Revision 2, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," January 2023.

22.

EMF-2103P-A Revision 3, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors,"

June 2016.

23.

EMF-2328(P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 2001.

24.

EMF-2328(P)(A) Revision 0 Supplement 1 (P)(A)

Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," December 2016.

25.

ANP-10349P-A Revision 0, "GALILEO Implementation in LOCA Methods," November 2021.

26.

ANP-10339P-A Revision 0, "ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology,"

October 2023.

27.

ANP-10338P-A Revision 0, "AREATM - ARCADIA Rod Ejection Accident," December 2017.

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met; and

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 6 Amendment 242 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System";

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

NRC letters dated January 21, 1998, "Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report,"

2.

NRC letter dated August 8, 2001, Issuance of Exemption from the requirements of 10 CFR 50.60 and Appendix G for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2,

3.

Westinghouse WCAP-16143, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,"

4.

NRC letter dated August 31, 2020, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2, Exemption from the Requirements of 10 CFR 50.61 and 10 CFR 50, Appendix G (EPID L-2019-LLE-0022)," and NRC letter dated September 18, 2020, "Braidwood Station, Units 1 and 2, and Byron Station Unit Nos. 1 and 2 -

Issuance of Amendment Nos. 217, 217, 221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-2019-LLA-0215)," and

5.

The PTLR will contain the complete identification for each of the TS referenced Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements); and

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 7 Amendment 242 5.6 Reporting Requirements 5.6.7 Post Accident Monitoring Report When a report is required by Condition C or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported in the Inservice Inspection Summary Report in accordance with 10 CFR 50.55a and ASME Section XI.

5.6.9 Steam Generator (SG) Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program.

The report shall include:

a.

The scope of inspections performed on each SG;

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c.

For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 8 Amendment 242 5.6 Reporting Requirements 5.6.9 Steam Generator (SG) Tube Inspection Report (continued)

d.

An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f.

The results of any SG secondary side inspections;

g.

For Unit 2, the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report;

h.

For Unit 2, the calculated accident induced leakage rate from the portion of the tubes below 14.01 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 3.11 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and

i.

For Unit 2, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT.

The report shall include:

a.

The PRA models upgraded to include newly developed methods;

b.

A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, Newly Developed Method Requirements and Peer Review;

Reporting Requirements 5.6 BRAIDWOOD UNITS 1 & 2 5.6 9 Amendment 242 5.6 Reporting Requirements 5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report (continued)

c.

Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d.

All changes to key assumptions related to newly developed methods or their implementations.

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 240 Renewed License No. NPF-37

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC dated May 28, 2024, as supplemented by letters dated April 21, 2025, May 1, 2025, and August 6, 2025, 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 renewed 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-37 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 240 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of its date of issuance and shall be implemented during the defuel window in the Spring Refueling Outage 2026.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 21, 2025

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 240 Renewed License No. NPF-66

1.

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

A.

The application for amendment by Constellation Energy Generation, LLC dated May 28, 2024, as supplemented by letters dated April 21, 2025, May 1, 2025, and August 6, 2025, 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-66 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 240, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of its date of issuance and shall be implemented during the defuel window in the Fall Refueling Outage 2026.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 21, 2025

ATTACHMENT TO LICENSE AMENDMENT NOS. 240 AND 240 RENEWED FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455 Replace the following pages of the Renewed Facility Operating Licenses and 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 Licenses REMOVE INSERT License NPF-37 License NPF-37 Page 3 Page 3 License NPF-66 License NPF-66 Page 3 Page 3 Technical Specifications REMOVE INSERT 2.0 - 1 2.0 - 1 2.0 - 2 2.0 - 2 3.2.1 - 1 3.2.1 - 1 3.2.1 - 2 3.2.1 - 2 3.2.1 - 4 3.2.1 - 4 3.2.1 - 5 3.2.1 - 5 3.5.1 - 2 3.5.1 - 2 3.5.4 - 2 3.5.4 - 2 4.0 - 1 4.0 - 1 5.5 - 20 5.5 - 20 5.6 - 3 5.6 - 3 5.6 - 4 5.6 - 4 5.6 - 5 5.6 - 5 5.6 - 6 5.6 - 6 5.6 - 7 5.6 - 7 5.6 - 8 5.6 - 8 5.6 - 9

(2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

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; (4)

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 (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 240 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Deleted.

(4)

Deleted.

Renewed License No. NPF-37 Amendment No. 240

(2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

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; (4)

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 (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No. 240, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Renewed License No. NPF-37, dated November 19, 2015, are hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Deleted.

Renewed License No. NPF-66 Amendment No. 240

SLs 2.0 BYRON UNITS 1 & 2 2.0 1 Amendment 240 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded.

2.1.1.1 For Westinghouse fuel in MODE 1, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.24 for the WRB-2 DNB correlation for a thimble cell, 1.25 for the WRB-2 DNB correlation for a typical cell and 1.19 for the ABB-NV DNB correlation for a thimble cell and a typical cell.

2.1.1.2 For Westinghouse fuel in MODE 2, the DNBR shall be maintained 1.17 for the WRB-2 DNB correlation, and 1.13 for the ABB-NV DNB correlation and 1.18 for the WLOP DNB correlation.

2.1.1.3 For Westinghouse fuel in MODES 1 and 2, the peak fuel centerline temperature shall be maintained 5080F, decreasing by 58F per 10,000 MWD/MTU burnup for all assemblies except for U72Y for Cycle 25, which decreases by 9°F per 10,000 MWD/MTU burnup.

2.1.1.4 For Framatome fuel, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.12 for the ORFEO-GAIA DNB correlation and 1.15 for the ORFEO-NMGRID DNB correlation.

2.1.1.5 For Framatome fuel, for UO2 fuel, the peak centerline temperature shall be maintained 5090 F, decreasing by 13.7 F per 10,000 MWD/MTU burnup.

2.1.1.6 For Framatome fuel, for UO2-Gd2O3 fuel, the peak centerline temperature shall be maintained 5090 F -

9360*E2 - 399.6*E, decreasing by 13.7 F per 10,000 MWD/MTU burnup where E is the gadolinia weight-fraction.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 psig.

(continued)

SLs 2.0 BYRON UNITS 1 & 2 2.0 2 Amendment 240 2.0 SAFETY LIMITS (SLs) (continued) 2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 1 Amendment 240 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by FQ C(Z) and FQ V(Z), shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE---------

Required Action A.4 shall be completed whenever this Condition is entered.

FQ C(Z) not within limit.

A.1 Reduce THERMAL POWER 1% RTP for each 1% FQ C(Z) exceeds limit.

AND A.2 Reduce Power Range Neutron Flux-High trip setpoints 1%

for each 1% FQ C(Z) exceeds limit.

15 minutes after each FQ C(Z) determination 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FQ C(Z) determination AND A.3 Reduce Overpower T trip setpoints 1%

for each 1% FQ C(Z) exceeds limit.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each FQ C(Z) determination AND A.4 Perform SR 3.2.1.1 and SR 3.2.1.2.

Prior to increasing THERMAL POWER above the limit of Required Action A.1 (continued)

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 2 Amendment 240 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. ---------NOTE---------

Required Action B.2 shall be completed whenever this Condition is entered.

FQ V(Z) not within limit.

B.1 Reduce AFD limits as specified in the COLR.

AND B.2 Perform SR 3.2.1.1 and SR 3.2.1.2.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Prior to increasing AFD limits above the limits of Required Action B.1 C. Required Action and associated Completion Time not met.

C.1 Be in MODE 2.

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

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 4 Amendment 240 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2

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

If FQ V(Z) measurements indicate that either the maximum over z [FQ C(Z) / K(Z)]

OR maximum over z [FQ V(Z)/ K(Z)]

has increased since the previous evaluation of FQ C(Z):

a. Increase FQ V(Z) by the greater of a factor of [1.02] or the appropriate factor specified in the COLR and reverify FQ V(Z) is within limits specified in the COLR; or

b. Repeat SR 3.2.1.2 once per 7 EFPD until either a. above is met or two successive flux maps indicate that the maximum over z [FQ C(Z)/ K(Z)]

and maximum over z [FQ V(Z)/ K(Z)]

has not increased.

(continued)

FQ(Z) 3.2.1 BYRON UNITS 1 & 2 3.2.1 5 Amendment 240 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued)

Verify FQ V(Z) is within limit specified in the COLR.

Prior to exceeding 75% RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQ V(Z) was last verified AND In accordance with the Surveillance Frequency Control Program

Accumulators 3.5.1 BYRON UNITS 1 & 2 3.5.1 2 Amendment 240 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water level in each accumulator is 31% and 63%.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is 602 psig and 647 psig.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each accumulator is 2600 ppm and 2900 ppm.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.5

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

Only required to be performed for affected accumulators after each solution volume increase of 10% of indicated level that is not the result of addition from the refueling water storage tank containing a boron concentration 2600 ppm and 2900 ppm.

Verify boron concentration in each accumulator is 2600 ppm and 2900 ppm.

Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SR 3.5.1.6 Verify power is removed from each accumulator isolation valve operator.

In accordance with the Surveillance Frequency Control Program

RWST 3.5.4 BYRON UNITS 1 & 2 3.5.4 2 Amendment 240 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1

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

Only required to be performed when ambient air temperature is < 35F or > 100F.

Verify RWST borated water temperature is 35F and 100F.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.2

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

Only required to be performed when ambient air temperature is < 35F.

Verify RWST vent path temperature is 35F.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST borated water level is 89%.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.4 Verify RWST boron concentration is 2700 ppm and 2900 ppm.

In accordance with the Surveillance Frequency Control Program

Design Features 4.0 BYRON UNITS 1 & 2 4.0 1 Amendment 240 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Site Location The site is located in Rockvale Township, approximately 3.73 mi (6 km) south-southwest of the city of Byron in northern Illinois.

4.1.2 Exclusion Area Boundary (EAB)

The EAB shall not be less than 1460 ft (445 meters) from the outer containment wall.

4.1.3 Low Population Zone (LPZ)

The LPZ shall be a 3.0 mi (4828 meter) radius measured from the midpoint between the two reactors.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, Optimized ZIRLOTM, or M5Framatome clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies (LTAs) that have not completed representative testing may be placed in nonlimiting core regions.

One LTA containing up to six Accident Tolerant Fuel (ATF) lead test rods may be placed in the Unit 2 reactor for evaluation.

This LTA may be loaded in a core location that will result in the LTA exceeding 62 GWd/MTU burnup at the end of Cycle 25. The LTA shall comply with fuel limits specified in the COLR and Technical Specifications under all operational conditions.

Programs and Manuals 5.5 BYRON UNITS 1 & 2 5.5 20 Amendment 240 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42.8 psig for Unit 1 and 38.7 psig for Unit 2 The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Containment leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests; and

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 3 Amendment 240 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

SL 2.1.1, "Reactor Core SLs";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Moderator Temperature Coefficient";

LCO 3.1.4, "Rod Group Alignment Limits";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor

)

F

( N H

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation";

LCO 3.3.9, "Boron Dilution Protection System (BDPS)";

LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration";

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluations Methodology," July 1985.

2.

Not Used.

3.

NFSR-0016, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," July 1983.

4.

NFSR-0081, "Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes," July 1990.

5.

Not used.

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 4 Amendment 240 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6.

WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.

7.

Not Used.

8.

Not Used.

9.

WCAP-10216-P-A, Revision 1, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.

10.

WCAP-8745-P-A, "Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions,"

September 1986.

11.

WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.

12.

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995, (Westinghouse Proprietary).

13.

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM," July 2006, (Westinghouse Proprietary).

14.

ANP-10297P-A Revision 0, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results," February 2013.

15.

ANP-10297 Revision 0 Supplement 1P-A Revision 1, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results," December 2020.

16.

ANP-10311P-A Revision 1, "COBRA-FLX: A Core Thermal-Hydraulic Analysis Code," October 2017.

17.

ANP-10341P-A Revision 0, "The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations,"

September 2018.

18.

XN-75-32(P)(A) Supplements 1, 2, 3, & 4, "Computational Procedure for Evaluating Fuel Rod Bowing," February 1983.

19.

ANP-10342P-A Revision 0, "GAIA Fuel Assembly Mechanical Design," September 2019.

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 5 Amendment 240 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

20.

ANP-10323P-A Revision 1, "GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors," November 2020.

21.

BAW-10227P-A Revision 2, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," January 2023.

22.

EMF-2103P-A Revision 3, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors,"

June 2016.

23.

EMF-2328(P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 2001.

24.

EMF-2328(P)(A) Revision 0 Supplement 1 (P)(A)

Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," December 2016.

25.

ANP-10349P-A Revision 0, "GALILEO Implementation in LOCA Methods," November 2021.

26.

ANP-10339P-A Revision 0, "ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology,"

October 2023.

27.

ANP-10338P-A Revision 0, "AREATM - ARCADIA Rod Ejection Accident," December 2017.

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met; and

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 6 Amendment 240 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a.

RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System";

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

NRC letter dated January 21, 1998, "Byron Station Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report,"

2.

NRC letter dated August 8, 2001, Issuance of Exemption from the requirements of 10 CFR 50.60 and Appendix G, for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2,

3.

Westinghouse WCAP-16143, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,"

4.

NRC letter dated August 31, 2020, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2, Exemption From the Requirements of 10 CFR 50.61 and 10 CFR 50, Appendix G (EPID L-2019-LLE-0022)," and NRC letter dated September 18, 2020, "Braidwood Station, Units 1 and 2, and Byron Station Unit Nos. 1 and 2 -

Issuance of Amendment Nos. 217, 217, 221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-2019-LLA-0215)," and

5.

The PTLR will contain the complete identification for each of the TS referenced Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements); and

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 7 Amendment 240 5.6 Reporting Requirements 5.6.7 Post Accident Monitoring Report When a report is required by Condition C or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported in the Inservice Inspection Summary Report in accordance with 10 CFR 50.55a and ASME Section XI.

5.6.9 Steam Generator (SG) Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program.

The report shall include:

a.

The scope of inspections performed on each SG;

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c.

For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 8 Amendment 240 5.6 Reporting Requirements 5.6.9 Steam Generator (SG) Tube Inspection Report (continued)

d.

An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f.

The results of any SG secondary side inspections;

g.

For Unit 2, the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report;

h.

For Unit 2, the calculated accident induced leakage rate from the portion of the tubes below 14.01 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 3.11 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and

i.

For Unit 2, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT.

The report shall include:

a.

The PRA models upgraded to include newly developed methods;

b.

A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, Newly Developed Method Requirements and Peer Review;

Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 9 Amendment 240 5.6 Reporting Requirements 5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report (continued)

c.

Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d.

All changes to key assumptions related to newly developed methods or their implementations.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION ENCLOSURE 6 (NON-PROPRIETARY)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72 AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77 AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37 AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 CONSTELLATION ENERGY GENERATION, LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, AND STN 50-455 Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.

Redacted information is identified by blank space enclosed within (( double brackets )).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72 AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77 AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37 AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 CONSTELLATION ENERGY GENERATION, LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-456, STN 50-457, STN 50-454, AND STN 50-455

1.0 PROPOSED CHANGE

By letter dated May 28, 2024 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML24149A125), as supplemented by letters dated April 21, 2025 (ML25111A257),May 1, 2025 (ML25121A230), and August 6, 2025 (ML25225A150), Constellation Energy Generation, LLC (CEG, or the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Units 1 and 2 (Byron). The licensee requested revisions to Technical Specifications (TS) to allow the use of Framatome GAIA fuel at Braidwood and Byron. Since the GAIA fuel uses M5Framatome fuel rod cladding, the licensee also included an exemption request from portions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, and 10 CFR Part 50, Appendix K, ECCS [Emergency Core Cooling System] Evaluation Models, as Attachment 18 to the LAR.

The NRC staff reviewed the exemption request in a separate safety evaluation (SE)

(ML25147A263).

From October 31, 2024, through April 25, 2025, the NRC staff conducted a regulatory audit to support its review of the amendment request, as discussed in the staffs audit plan dated November 25, 2024 (ML24299A255), and audit summary dated May 16, 2025 (ML25125A263).

The supplemental letters dated April 21, 2025, May 1, 2025, and August 6, 2025, provided additional information that clarified the application, did not expand the scope of the application

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on November 5, 2024 (89 FR 87898).

2.0 REGULATORY EVALUATION

2.1 Proposed TS Changes

The licensees proposed changes to TSs, limiting conditions for operation (LCOs), and surveillance requirements (SRs) are described below where additions are in bold text while deletions are in strikethrough text.

TS 2.1.1, Reactor Core SLs 2.1.1.1 For Westinghouse fuel inIn MODE 1, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.24 for the WRB-2 DNB correlation for a thimble cell, 1.25 for the WRB-2 DNB correlation for a typical cell and 1.19 for the ABB-NV DNB correlation for a thimble cell and a typical cell.

2.1.1.2 For Westinghouse fuel inIn MODE 2, the DNBR shall be maintained 1.17 for the WRB-2 DNB correlation, and 1.13 for the ABB-NV DNB correlation and 1.18 for the WLOP DNB correlation.

[Braidwood 2.1.1.3]

2.1.1.3 For Westinghouse fuel inIn MODES 1 and 2, the peak fuel centerline temperature shall be maintained < 5080°F decreasing by 58°F per 10,000 MWD/MTU burnup.

[Byron 2.1.1.3]

2.1.1.3 For Westinghouse fuel inIn MODES 1 and 2, the peak fuel centerline temperature shall be maintained < 5080°F decreasing by 58°F per 10,000 MWD/MTU burnup for all assemblies except for U72Y for Cycle 25, which decreases by 9°F per 10,000 MWD/MTU burnup.

2.1.1.4 For Framatome fuel, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.12 for the ORFEO-GAIA DNB correlation and 1.15 for the ORFEO-NMGRID DNB correlation.

2.1.1.5 For Framatome fuel, for UO2 fuel, the peak centerline temperature shall be maintained < 5090 °F, decreasing by 13.7 °F per 10,000 MWD/MTU burnup.

2.1.1.6 For Framatome fuel, for UO2-Gd2O3 fuel, the peak centerline temperature shall be maintained < 5090 °F - 9360*E2 - 399.6*E, decreasing by 13.7 °F per 10,000 MWD/MTU burnup where E is the gadolinia weight-fraction.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by FQC(Z) and FQWV(Z), shall be within the limit specified in the COLR.

CONDITION REQUIRED ACTION COMPLETION TIME B. FQWV(Z) not within limit.

B.1 Reduce THERMAL POWER as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.2 Reduce AFD limits as specified in the COLR.

AND 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B.3 Reduce Power Range Neutron Flux-High trip setpoints 1% for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action B.1.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.4 Reduce Overpower DT trip setpoints 1% for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action B.1.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.25 Perform SR 3.2.1.1 and SR 3.2.1.2 Prior to increasing THERMAL POWER and AFD limits above the limits of Required Actions B.1 and B.2

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION SURVEILLANCE FREQUENCY SR 3.2.1.2

NOTES------------------------------

1.

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

2.

If FQWV(Z) measurements indicate that either the maximum over z [FQC(Z) / K(Z)]

OR maximum over z [ZQWV(Z) / K(Z) ]

has increased since the previous evaluation of FQC(Z) or if FQW(Z) is expected to increase prior to the next evaluation of FQC(Z) :

a.

Increase FQWV(Z) by the appropriate factor specified in the COLR and reverify FQWV(Z) is within limits specified in the COLR; or

b.

Repeat SR 3.2.1.2 once per 7 EFPD until either a. above is met or two successive flux maps indicate that the maximum over z [FQC(Z) / K(Z)]

and maximum over z [FQWV(Z) / K(Z)]

has not increased.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued)

NOTES------------------------------

3.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after declaring PDMS inoperable.

Performance of SR 3.2.1.4 satisfies the initial performance of this SR after declaring PDMS inoperable.

Verify FQWV(Z) is within limit specified in the COLR.

Prior to exceeding 75%

RTP after each refueling AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQWV(Z) was last verified AND In accordance with the Surveillance Frequency Control Program SURVEILLANCE FREQUENCY SR 3.2.1.4

NOTES------------------------------

Only required to be performed when PDMS is OPERABLE.

Verify FQWV(Z) is within limit specified in the COLR.

In accordance with the Surveillance Frequence Control Program

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION TS 3.5.1, Accumulators SURVEILLANCE FREQUENCY SR 3.5.1.4 Verify boron concentration in each accumulator is 26002200 ppm and 29002400 ppm.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.5

NOTES------------------------------

Only required to be performed for affected accumulators after each solution volume increase of 10% of indicated level that is not the result of addition from the refueling water storage tank containing a boron concentration 26002200 ppm and 29002400 ppm.

Verify boron concentration in each accumulator is 26002200 ppm and 29002400 ppm.

Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months TS 3.5.4, Refueling Water Storage Tank (RWST)

SURVEILLANCE FREQUENCY SR 3.5.4.4 Verify RWST boron concentration is 27002300 ppm and 29002500 ppm.

In accordance with the Surveillance Frequency Control Program TS 4.2.1, Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLOTM, or M5Framatome clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.

Limited substitutions of zirconium alloy or stainless-steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

TS 5.5.16, Containment Leakage Rate Testing Program The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa, is 42.8 psig for Unit 1 and 38.738.4 psig for Unit 2

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION TS 5.6.5, Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

SL 2.1.1, Reactor Core SLs; LCO 3.1.1, SHUTDOWN MARGIN (SDM);

LCO 3.1.3, Moderator Temperature Coefficient; LCO 3.1.4, Rod Group Alignment Limits; LCO 3.1.5, Shutdown Bank Insertion Limits; LCO 3.1.6, Control Bank Insertion Limits; LCO 3.1.8, PHYSICS TESTS Exceptions - MODE 2; LCO 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

LCO 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FDHN);

LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFD);

LCO 3.2.5, Departure from Nucleate Boiling Ratio (DNBR);

LCO 3.3.1, Reactor Trip System (RTS) Instrumentation; LCO 3.3.9, Boron Dilution Protection System (BDPS);

LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and LCO 3.9.1, Boron Concentration;

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, Westinghouse Reload Safety Evaluations Methodology, July 1985.

2.

WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

3.

NFSR-0016, Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods, July 1983.

4.

NFSR-0081, Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes, July 1990.

5.

ComEd letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and ComEd application of the UET methodology addressed in Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Control Systems.

6.

WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.

7.

Not Used.

8.

Not Used.

9.

WCAP-10216-P-A, Revision 1, Relaxation of Constant Axial Offset Control -

FQ Surveillance Technical Specification, February 1994.

10. WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

11. WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999.

12. WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995, (Westinghouse Proprietary).

13. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, July 2006, (Westinghouse Proprietary).

14. ANP-10297P-A Revision 0, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, February 2013.

15. ANP-10297 Revision 0 Supplement 1P-A Revision 1, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, December 2020.

16. ANP-10311P-A Revision 1, COBRA-FLX: A Core Thermal-Hydraulic Analysis Code, October 2017.

17. ANP-10341P-A Revision 0, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, September 2018.

18. XN-75-32(P)(A) Supplements 1, 2, 3, & 4, Computational Procedure for Evaluating Fuel Rod Bowing, February 1983.

19. ANP-10342P-A Revision 0, GAIA Fuel Assembly Mechanical Design, September 2019.

20. ANP-10323P-A Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, November 2020.

21. BAW-10227P-A Revision 2, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, January 2023.

22. EMF-2103P-A Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016.

23. EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

24. EMF-2328(P)(A) Revision 0 Supplement 1(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, December 2016.

25. ANP-10349P-A Revision 0, GALILEO Implementation in LOCA Methods, November 2021.

26. ANP-10339P-A Revision 0, ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology, October 2023.

27. ANP-10338P-A Revision 0, AREA' - ARCADIA Rod Ejection Accident, December 2017.

2.2 Applicable Regulations and Guidance The NRC staff considered the following regulatory requirements and guidance during its review of the LAR.

Regulatory Requirements The regulations in 10 CFR 50.90 state that whenever a holder of an operating license desires to amend the license, including TSs in the license, an application for amendment must be filed with the Commission fully describing the changes desired. The regulations at 10 CFR 50.92(a) state that determinations on whether to grant an applied for license amendment are guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a) (regarding, among

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION other things, consideration of the operating procedures, the facility and equipment, the use of the facility, and other TSs, or the proposals) and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commission's regulations.

The regulations under 10 CFR 50.36, Technical specifications, provide regulatory requirements related to the content of TSs. Section 50.36(b) of 10 CFR requires that each license authorizing the operation of a facility will include TSs and that the TSs will be derived from the safety analysis. Section 50.36(c) of 10 CFR specifies the categories that are to be included in the TSs including (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls.

Sections 50.36(c)(1), (c)(4), and (c)(5) of 10 CFR require the following:

The regulation under 10 CFR 50.36(c)(1)(i)(A), Safety limits, limiting safety system settings and limiting control settings, states, in part, that [s]afety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

The regulation under 10 CFR 50.36(c)(2), Limiting conditions for operation, states, in part, [w]hen a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The regulation under 10 CFR 50.36(c)(4), Design features states that [d]esign features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.

The regulation under 10 CFR 50.36(c)(5), Administrative controls, 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 the operation of the facility in a safe manner.

Key regulatory requirements specified in 10 CFR 50.46(a) that are relevant to the proposed license amendment include:

Each pressurized light-water reactor (LWR) fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must perform analysis of core cooling performance under postulated loss-of-coolant accident (LOCA) conditions using an acceptable evaluation model (EM).

An acceptable LOCA EM must be used that either applies realistic methods with an explicit accounting for uncertainties or follows the prescriptive, conservative requirements of Appendix K to 10 CFR Part 50.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Core cooling performance must be analyzed for a number of postulated LOCAs of different sizes, locations, and other characteristics to ensure that the most severe event is calculated.

The following ECCS acceptance criteria of 10 CFR 50.46(b)(1) though (b)(5) state in part:

(1)

Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

(2)

Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3)

Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(4)

Coolable geometry. Calculated changes in-core geometry shall be such that the core remains amenable to cooling.

(5)

Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

The above acceptance criteria of 10 CFR 50.46(b)(1) through (5) are referred to as the peak cladding temperature (PCT) criterion, the maximum local oxidation (MLO) criterion, the hydrogen generation (or core wide oxidation (CWO)) criterion, the coolable geometry criterion, and the long-term cooling criterion respectively.

The regulations in 10 CFR 50.67, Accident source term, require, in part, in 10 CFR 50.67(b)(2) that the applicants analysis must demonstrate with reasonable assurance that: (i) an individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release would not receive a radiation dose in excess of 0.25 Sv (25 rem) [roentgen equivalent man] Total Effective Dose Equivalent (TEDE); (ii) an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE; and (iii) adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions, without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

The 10 CFR Part 50, Appendix A, General Design Criteria (GDCs) applicable to this LAR are as follows:

GDC 4, Environmental and dynamic effects design bases, states, in part, that structures, systems, and components important to safety shall be designed to

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION accommodate the effects of and to be compatible with the environmental conditions associated with postulated accidents, including loss-of-coolant accidents and shall be appropriately protected against dynamic effects, including the effects of discharging fluids, that may result from equipment failures.

GDC 10, Reactor design, states that [t]he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 11, Reactor inherent protection, states that [t]he reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

GDC 12, Suppression of reactor power oscillations, states that [t]he reactor core and associated coolant, control, and protection systems be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

GDC 16, Containment design, states that [r]eactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 19, Control room, states that [a] control room shall be provided from which actions can be taken to operate the [plant] safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

GDC 20, Protection system functions, states that [t]he protection system 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.

GDC 25, Protection system requirements for reactivity control malfunctions, states that

[t]he protection system be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

GDC 26, Reactivity control system redundancy and capability, states, in part, that two independent reactivity control systems of different design principles be provided, one of which can hold the reactor core subcritical under cold conditions.

GDC 27, Combined reactivity control system capability, states, in part, that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION GDC 35, Emergency core cooling, states:

A system to provide abundant emergency core cooling shall be provided.

The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC 38, Containment heat removal, states, in part that [a] system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.

GDC 50, Containment design basis, states, in part, that [t]he reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.

Appendix K to 10 CFR Part 50 establishes required and acceptable features of EMs for heat removal by the ECCS after the blowdown phase of a LOCA. It consists of the following two parts:

required and acceptable features of LOCA EMs and, documentation required for LOCA EMs.

The first part specifies modeling requirements and acceptable methods for simulating significant physical phenomena throughout all phases of a design basis LOCA event, including relevant heat sources, fuel rod performance, and thermal-hydraulic (T-H) behavior.

The second part specifies requirements for the documentation of LOCA EMs, including a complete description, a code listing, sensitivity studies, and comparisons against experimental data.

Regulatory Guidance The NRC staff relied on the following sections of NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP) in its review of this LAR:

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Section 3.7.1, Seismic Design Parameters, Revision 4, dated December 2014 (ML14198A460).

Section 4.2, Fuel System Design, Revision 3, March 2007 (ML070740002).

Section 4.4, Thermal and Hydraulic Design, Revision 2, dated March 2007 (ML07055060).

Chapter 15, Introduction - Transient and Accident Analyses, Revision 3, dated March 2007 (ML070710376).

Section 6.4, Control Room Habitability System, Revision 3, dated March 2007 (ML070550069).

Section 6.5.2, Containment Spray as a Fission Product Cleanup System, Revision 2, dated December 1988 (ML052340761).

Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, dated July 2000 (ML003734190).

Regulatory Guide (RG) 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, dated June 2020 (ML20055F490), details acceptable methods and procedures to use when analyzing a postulated control rod drop accident.

RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, Revision 1, dated March 2007 (ML070350028).

RG 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Revision 3, dated October 2012 (ML12220A043).

RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ML003716792).

RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, Revision 0, dated June 2003 (ML031530505)

NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, dated September 13, 2004 (ML042360586).

Regulatory Issue Summary (RIS) 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ML053460347).

NRC Information Notice (IN) 2012-09, Irradiation Effects on Fuel Assembly Spacer Grid Strength, dated June 28, 2012 (ML113470490).

2.3 Description Braidwood and Byron are four-loop Westinghouse pressurized water reactors (PWRs). They currently use Westinghouse VANTAGE+ fuel clad with Optimized ZIRLOTM. CEG is proposing to transition to Framatome GAIA fuel and related Framatome design methods. The GAIA fuel design consists of a 17x17 array with GAIA and intermediate GAIA mixing (IGM, spacer) grids, a lower high mechanical performance (HMP) grid, and an upper HMP grid. The fuel assembly includes a MONOBLOC guide tube design, M5Framatome fuel rod design, and a GRIP lower

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION nozzle. The fuel is standard UO2 fuel with 2, 4, 6, and 8 weight percent gadolinia rods included.

The GAIA fuel design is described in NRC-approved topical report (TR) ANP-10342P-A, Revision 0, GAIA Fuel Assembly Mechanical Design, September 2019 (ML19309D916 (public), and ML19309D917 (not publicly available; proprietary information)).

Note that M5Framatome is also referred to as M5 by the licensee and these two terms are used interchangeably throughout the various documents reviewed by NRC staff.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the regulations, guidance, and plant-specific design and licensing basis discussed in Section 2.2 of this safety evaluation. The NRC staff reviewed the licensees statements in the LAR, Attachments to the LAR, and the relevant sections of the Braidwood and Byron TS and Updated Final Safety Analysis Report (UFSAR). The staff reviewed the limitations and conditions (L&Cs) for all proposed topical reports to assure they are met satisfactorily.

The NRC staffs technical evaluation is organized into the following sections:

Section Topic 3.1 GAIA Fuel Assemblies 3.2 Core Source Term and Dose Consequences

3.3 Proposed TS Changes

3.4 Loss-of-Coolant Accident (LOCA) Analysis 3.5 Containment Integrity 3.6 Control Rod Ejection Accident (REA) 3.7 Non-LOCA Transients/Accidents 3.8 Co-Resident Fuel Considerations 3.9 Additional Items 3.1 GAIA Fuel Assemblies The fuel system design bases must reflect four objectives: (1) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs), (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained. To satisfy these objectives, acceptance criteria are needed for fuel system damage, fuel rod failure, and fuel coolability. The LAR evaluates mechanical design-related aspects of the fuel system design basis.

3.1.1 Mechanical Design The GAIA fuel mechanical design acceptance criteria are met up to the licensed peak UO2 fuel rod burnup of 62 Gigawatt days per Metric Ton Uranium (GWd/MTU).

The design criteria relating to fuel system damage should not be exceeded during normal operation including AOOs. Fuel rod failure should be precluded, and fuel damage criteria should ensure that fuel system dimensions remain within operational tolerances and that functional capabilities are not reduced below those presumed in the safety analysis. Each damage

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION mechanism listed in SRP Section 4.2 will be reviewed to confirm that the design criteria are not exceeded during normal operation for the GAIA design.

3.1.1.1 Normal Operation and AOO Component Stress The licensed method associated with the normal operation and AOO component stress analysis is in the GAIA Mechanical Design TR (Reference 7.1). The design criteria for stress are that the stress intensities for GAIA fuel assembly components shall be less than the stress limits based on the American Society of Mechanical Engineers (ASME) Code,Section III criteria (Reference 7.2), unless otherwise specified. These design criteria are consistent with the acceptance criteria of SRP Section 4.2; therefore, the stress criteria are acceptable for application to the GAIA fuel design.

The licensee used a deterministic method to obtain the most limiting stress value. The method provides the most conservative stress value for each fuel assembly component. Positive margin to the design criteria is shown for each of the fuel assembly components; therefore, the NRC staff concludes that the fuel assembly design satisfies the design criteria for design stress.

3.1.1.2 Shipping and Handling Component Stress Licensing criterion for the shipping and handling of component stress fuel damage mechanism is in the GAIA Mechanical Design TR (Reference 7.1). Stresses and/or loads associated with shipping and handling shall be less than limits based on Section III of the ASME Code (Reference 7.2) for all components, unless otherwise specified. The licensee evaluated all axial loads for shipping and handling for beginning of life (BOL) conditions and consider tension and compression. Shipping conditions are evaluated based on the slightly elevated shipping container temperatures for normal conditions of transport. The applicable shipping and handling load limits for GAIA fuel include (1) 6 g1 lateral acceleration, (2) 4 g maximum axial acceleration, and (3) 2.5 g axial handling. The associated component stresses are within the specified ASME Code limits, and the staff therefore concluded that the GAIA fuel assembly is structurally adequate for all shipping and handling loading conditions.

3.1.1.3 Fuel Rod Plenum Spring Stress Licensing criterion for normal operation, AOO, shipping, and handling fuel rod plenum spring fuel damage is in the GAIA Mechanical Design TR (Reference 7.1). Stresses and/or loads associated with normal operation, AOO, shipping, and handling shall be less than the limits based on Section III of the ASME Code for all components, unless otherwise specified. The licensee evaluated conditions include (1) 4 g maximum acceleration, (2) 2.5 g axial handling, and (3) sufficient preload to maintain contact during pellet densification. The associated fuel rod plenum string stresses are within the specified Code limits, and the staff therefore concluded that the GAIA fuel plenum spring stresses are acceptable.

3.1.1.4 General Component Fatigue For all components other than fuel rod cladding, the cumulative usage factor (CUF) shall be less than 1.0. The licensed method associated with the general component fatigue analysis is per Section 8.1.2.2 of the GAIA Mechanical Design TR (Reference 7.1). Due to the significant load-bearing function of the guide tubes (GTs), in conjunction with their relatively thin-walled 1 g is acceleration of earth's gravity which is approximately 32.2 feet per second2 (ft/s2)

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION construction, GT fatigue is considered the most limiting and bounds all other components of the fuel assembly with the exception of the holddown springs.

The licensee performed fatigue calculations in accordance with approved methods for the GAIA structural components, including fuel rods, GTs, and holddown springs. All component margins are positive, and the staff therefore concluded that the GAIA fuel assembly is structurally adequate for all normal operating conditions and AOOs.

3.1.1.5 Fretting Wear The design criteria for fretting are that the GAIA fuel assembly design shall be shown to have no failure due to fretting (Reference 7.1). This criterion is consistent with the acceptance criteria of SRP Section 4.2; therefore, the fretting criteria are acceptable for application to the GAIA fuel design.

The licensee performed extensive autoclave testing using expected end-of-life (EOL) condition for the GAIA fuel assemblies. Fretting wear and performance testing were performed at the HERMES P (Cadarache, France) and PHTF (Richland, Washington) flow test facilities. The 1000-hour endurance flow tests were performed and followed up by additional tests at the PETER loop (Erlangen Germany). Evaluations of this extensive testing showed that the GAIA fuel assembly is predicted to meet all criteria through EOL. Therefore, NRC staff finds that the licensee has demonstrated that the GAIA fuel has the ability to meet this criterion.

3.1.1.6 Fuel Rod and Fuel Assembly Growth Axial and lateral dimensional changes in the fuel rod and fuel assembly can occur due to irradiation growth, irradiation relaxation, creep, thermal expansion, etc. and can cause component-to-component or component-to-core interferences. These may lead to component failures and/or impacts on thermal-hydraulic limits, control rod insertion, and/or handling damage.

Fuel rod irradiation growth is addressed by providing sufficient clearance between the fuel rod and assembly top and bottom nozzles at EOL. Fuel assembly irradiation growth is addressed by providing sufficient clearance between the fuel assembly and reactor core plates at EOL. To assess fuel rod and fuel assembly growth, empirical models are used to compute the irradiation growth of the applicable components and the resulting changes are compared with the specified dimensions. This is in accordance with the criteria in the GAIA Mechanical Design topical report (Reference 7.1) previously approved by the NRC.

The licensee applied following methodologies and NRC-approved topical reports to assess the fuel rod and fuel assembly growth. The upper bound fuel rod growth and lower bound fuel assembly growth is used in conjunction with component manufacturing tolerances to determine the fuel rod shoulder gap margin. The upper bound fuel assembly growth is used in conjunction with component and core plate manufacturing tolerances to determine the fuel assembly gap margin. Limiting burnups and temperatures are considered.

The NRC-approved M5 fuel rod growth model in Reference 7.7 is used for the fuel rod growth bounds. The NRC-approved Q12 guide tube growth model in Reference 7.4 is used for the fuel assembly growth bounds.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff concludes that both fuel rod growth and fuel assembly growth criteria have been met.

3.1.1.7 Fuel Assembly Lift-off Framatomes licensing criterion for the fuel assembly lift-off fuel damage mechanism is per Section 8.1.7.1 of the GAIA Mechanical Design TR (Reference 7.1).

Section 5.1.5 of Attachment 13 to the proposed LAR summarizes the licensee's proposed approach for addressing fuel assembly lift-off. The design criteria for assembly lift-off are that during normal operating conditions and AOOs (with the exception of a pump over-speed transient), the GAIA fuel holddown springs maintain positive holddown margin (i.e., fuel assembly contact with the lower support plate). Assuming a pump over-speed transient, fuel assembly lift-off can occur, but the fuel assembly top and bottom nozzles maintain engagement with reactor internal pins, and the holddown springs maintain positive holddown margin after the event.

The fuel assembly lift-off methodology makes use of conventional open-literature equations to obtain a balance of forces on the fuel assembly in the vertical direction. The forces are due to fluid friction loss, buoyancy, momentum change, holddown spring force, and gravity. Forces due to friction losses are obtained through the use of loss coefficients derived from flow testing.

Holddown forces are obtained from testing. Other forces due to momentum and buoyancy are calculated based on the applicable fluid conditions. The evaluation includes the assessment of bounding operating conditions, component dimensional characteristics, and material characteristics. The Q12 GT growth model in Reference 7.4 is used to determine the fuel assembly growth bounds. Uncertainties are accounted for using a combination of deterministic and statistical methods.

In Section 3.3.1.9 of its SE on Reference 7.1, the NRC staff found that Framatome performs a combination of deterministic and statistically based analysis and can demonstrate that during all conditions considered, except for the pump over-speed transient, the fuel assembly lift-off criteria are met. During the pump over-speed transient, any lift-off will not be capable of disengaging the fuel assembly from core support plate alignment pins, and the holddown spring deflection is less than the worst-case normal operating cold-shutdown condition. Therefore, the NRC staff concluded in its SE on Reference 7.1 that for the GAIA fuel assembly design, the fuel assembly lift-off criteria are met.

For the present review, the NRC staff reviewed the submitted information in the proposed LAR concerning the licensees methodology and finds that the methods and acceptance criteria proposed are consistent with those generically approved in Reference 7.1. Positive holddown margin is maintained for all events except the pump over-speed transient, for which the lift is small and would not be expected to lead to misalignments or other adverse impacts on the fuel assemblies or core internals. Therefore, consistent with the approved methodology in Reference 7.1, the staff finds the proposed approach to be acceptable.

3.1.1.8 General Component Structural Deformation Earthquakes and postulated pipe breaks in the reactor coolant system would result in external forces on the fuel assembly. The fuel assembly is designed to withstand the structural loads from the Operating Basis Earthquake (OBE), Safe Shutdown Earthquake (SSE), and LOCA

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION events without loss of the capability to perform in a manner credited in its design basis for these events.

Licensing criteria for the general component structural deformation fuel damage mechanism are per Section 4 of the TR ANP-10337P-A (Reference 7.5). Specific to the GAIA spacer grid, additional criteria per Section 3.2 of the Deformable Grid Element TR (Reference 7.6) are applicable. Per Section 4 of Reference 7.5, OBE stress and load limits are set at the level A limits defined in the ASME Code, unless otherwise specified. SSE and LOCA stress and load limits are set at the Level D limits defined in the ASME Code, unless otherwise specified. Due to their special functions (i.e., forming a path for control rod insertion, ensuring coolable geometry is maintained), spacer grids and GTs are subject to more stringent service limits including:

OBE Spacer Grid Acceptance Criteria:

Spacer grid deformation experienced during an OBE event should not exceed the magnitude of the tolerance band to which the grid was designed. This acceptance criterion is established in the form of a grid impact load limit, which corresponds to a small amount of plastic deformation in the spacer grid that is within the envelope tolerance and does not exceed the deformation at the buckling point of the grid.

SSE/LOCA Spacer Grid Acceptance Criteria:

Spacer grid deformation experienced during an SSE/LOCA event ((

))

SSE/LOCA Guide Tube Acceptance Criteria:

Sudden and severe changes in the geometry of the GT (e.g., local collapse or plastic hinge) shall not occur. This acceptance criterion is further delineated by requiring that (1) stresses do not exceed a limit prohibiting local collapse of the GT, and (2) the structural stability of the GT must be maintained. The first criterion is met by limiting GT stresses to the Level C criteria in accordance with the ASME Code. The second criterion is satisfied by evaluating the critical buckling load margin.

Specific to the GAIA structural grid, per Section 3.2 of Reference 7.6, the spacer grid acceptance criteria above may be replaced as follows:

The limiting impact load is replaced with the limiting residual deformation. ((

))

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC-approved method associated with the general component structural deformation analysis is per Sections 7 and 8 of the Faulted topical report (Reference 7.5) and Sections 4 and 5 of the Deformable Grid Element topical report (Reference 7.6).

All component margins are positive, and the staff therefore concludes that the GAIA fuel assembly is structurally adequate for all postulated accident conditions and AOO's.

3.1.2 Thermal-Mechanical Design Design criteria relating to the fuel rod failure are applied in two ways. When they are applied to normal operation including AOOs, they are used as limits (i.e., specified acceptable fuel design limits (SAFDLs)), since fuel failure should not occur under these conditions. When they are applied to postulated accidents, limited fuel failures are permitted and must be accounted for in the fission product releases. Fuel rod failure is defined as the loss of fuel rod hermeticity. Each fuel rod failure mechanism listed in SRP Section 4.2 will be reviewed to confirm that the design criteria are not exceeded during normal operation and are properly accounted for during postulated accidents for the GAIA design.

Fuel rod thermal-mechanical evaluations are dependent on the rod power and core operating parameters. Because these parameters may vary for each operating cycle, verification on a cycle-specific basis is needed to ensure the actual cycle design will not result in SAFDL non-compliance. All fuel rod analyses are based on inputs which either represent or bound expected operating conditions at Braidwood and Byron.

3.1.2.1 Cladding Stress Normal Operation and AOO Licensing criteria for the normal operation and AOO cladding stress fuel damage mechanism are per the M5 TR (Reference 7.7, Section 10.1). The NRC-approved method associated with the fuel rod cladding stress normal operation and AOO analysis is per Section 8.1.1.2 of the GAIA Mechanical Design TR (Reference 7.1), with material properties per the M5 TR (Reference 7.7) used as inputs.

((

))

This LAR applied the NRC-approved criteria and methodology of cladding stress normal operation and AOO. NRC staff reviewed and concluded that the criteria for normal operation and AOO cladding stress have been met.

3.1.2.2 Cladding Buckling ((

))

Licensing criterion for the cladding buckling ((

)) fuel damage mechanism is per the M5 TR (Reference 7.7, Section 10.2).

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensed method for the fuel rod cladding buckling ((

)) analysis is based on the ASME Code and Reference 7.7.

((

))

This LAR applied the NRC-approved criteria and methodology of cladding buckling

((

)) NRC staff reviewed and concluded that the criterion for cladding buckling

((

)) has been met.

3.1.2.3 Cladding Structural Deformation (Faulted Stress and Buckling)

Licensing criteria for the faulted cladding buckling stress fuel damage mechanism are per the M5 TR (Reference 7.7, Section 10.1).

The licensed method for calculating the overall faulted stresses and margins is consistent with Section 8.1.1.2 of the GAIA Mechanical Design TR (Reference 7.1), with material properties per the M5 TR (Reference 7.7) used as input.

This LAR applied the NRC-approved criteria and methodology of faulted cladding stress and faulted cladding buckling. NRC staff reviewed and concluded the criteria for faulted cladding stress and faulted buckling have been met.

3.1.2.4 Cladding Fatigue The design criterion for M5 cladding fatigue is that the GAIA maximum fuel rod fatigue CUF shall not exceed 0.9. Licensing criterion for the cladding fatigue fuel damage mechanism is per Section 8.1.2.1 of the GAIA Mechanical Design TR (Reference 7.1). This design criterion is consistent with the acceptance criteria of SRP Section 4.2; therefore, this cladding fatigue criterion is acceptable for application to the GAIA fuel design.

The licensed method associated with the cladding fatigue analysis is per the M5 TR (Reference 7.7, Section 10.5).

The licensees procedures for the fatigue analysis follow those outlined in the ASME Code. The analysis uses all the Condition I and II events and one Condition III event to determine the total cladding fatigue usage factor. The maximum fatigue usage factor was determined to be well below the design criteria limit. Since the licensees methodology is consistent with the guidance in SRP Section 4.2 and the maximum fatigue is well below the design criteria limit, the staff concluded that the cladding fatigue acceptance criterion has been met.

3.1.2.5 Cladding Oxidation The criterion for the cladding oxidation fuel damage mechanism is per Section 8.1.4.1 of the GAIA Mechanical Design TR (Reference 7.1). The design criterion for cladding oxidation is that the GAIA fuel rod cladding best-estimate corrosion shall not exceed 100 microns. The M5 cladding material fuel rods are expected to have less than ((

)) Hydrogen pickup is a material-dependent property that is driven by cladding alloying elements and accelerated by thicker oxide layers. Initial hydrogen pickup is limited by vendor-controlled manufacturing processes that remain unchanged for GAIA. Additionally, the M5 cladding material has an

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION improved resistance to hydrogen pickup relative to previous generations of zirconium alloy claddings.

The licensed method associated with the cladding oxidation analysis is per the GALILEO TR (Reference 7.12, Section 3.4).

Criteria for cladding oxidation are intended to preclude potential fuel damage mechanisms. The SRP does not specify limits on cladding oxidation but does specify that its effects should be accounted for in the thermal and mechanical analyses performed for the fuel. The licensee accounts for the corrosion based on a database established for the M5 cladding material from its in-reactor performance over a number of years. The methodology and limits for cladding oxidation defined in Reference 7.1 are applicable and acceptable in the evaluation of the GAIA fuel assembly. This approach is acceptable because it uses realistic data that is representative of the material and burnup limits for the GAIA fuel assembly design.

Based on the data for M5 cladding material taken under prototypical irradiation conditions, staff concluded that there is reasonable assurance that oxidation will remain well below its acceptance criterion under the expected operating conditions at Braidwood and Byron.

3.1.2.6 Fuel Rod Internal Pressure Licensing criteria for the fuel rod internal pressure is per the GALILEO TR (Reference 7.12, Section 3.1). The rod internal pressure is limited to a maximum over pressure above reactor system pressure so the following criteria are met:

No clad lift-off during normal operation.

No reorientation of the hydrides in the radial direction in the cladding.

The licensee used the methodology in the GALILEO TR (Reference 7.12, Section 3.4) to ensure that the above criteria are met. For this reason, the staff concluded that the fuel rod internal pressure criteria have been met.

3.1.2.7 Internal Hydriding The design criterion for internal hydriding is that internal hydriding shall be precluded by appropriate manufacturing controls. The licensed method associated with fuel rod internal hydriding is per Section 8.2.1.2 of the GAIA Mechanical Design TR (Reference 7.1). For the GAIA assembly design, hydriding is prevented by keeping the level of moisture and hydrogenous impurities within the fuel to very low levels. GAIA UO2 and gadolinia fuel pellets have a total hydrogen content ((

))

Framatome maintains the low hydrogen levels in the fuel rod through manufacturing controls.

Because these controls will remain in place for the GAIA fuel assembly design and the limits are lower than the SRP Section 4.2 values, the staff concluded that the design criteria will continue to be met with the GAIA fuel assembly design.

3.1.2.8 Cladding Creep Collapse

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The design criterion for cladding collapse is that the predicted creep collapse life of the fuel rod must exceed the maximum expected in-core life. The SRP states that if axial gaps in the fuel pellet column occur due to densification, the cladding has the potential to collapse into such a gap. Because of the large local strains that accompany this process, any collapsed cladding is assumed to fail. Because the design criterion is consistent with the acceptance criteria of SRP Section 4.2, it is acceptable for application to the GAIA fuel assembly design.

Licensee uses their approved creep collapse methodology CROV TR (Reference 7.14) to determine the potential for creep collapse of the GAIA fuel assembly design with inputs provided per the GALILEO TR (Reference 7.12). This methodology uses conservative values to determine the creep collapse life of the fuel rod. When the ovality creep rate of the cladding exceeds ((

)) the cladding pressure differential exceeds the bifurcation buckling limit, or the generalized stress exceeds the generalized yield strength, the cladding is deemed failed. Based on these definitions of creep collapse, the creep collapse lifetime was shown to be greater than the allowable rod average burnup limit of 62 GWd/MTU. Therefore, NRC staff concludes that the GAIA fuel assembly design is adequately designed to prevent creep collapse for a service life up to 62 GWd/MTU, and the criteria for cladding creep collapse have been met.

3.1.3 Thermal-Hydraulic Design This section summarizes the thermal-hydraulic analyses associated with the GAIA fuel intended for implementation at both Braidwood and Byron.

3.1.3.1 Fuel Rod Bow The design criterion for fuel rod bow is that fuel rod bowing shall be evaluated with respect to the mechanical and thermal-hydraulic performance of the fuel assembly. There is not a specific limit for fuel rod bow specified in SRP Section 4.2; the SRP only calls for rod bow to be included in the design analysis to assure that it does not affect the satisfaction of other design criteria (e.g., DNB). DNB and linear heat generation rate (LHGR) burnup thresholds and penalties are calculated and considered on a cycle-by-cycle basis.

The methodology for fuel rod bow was approved in Exxon Nuclear Company TR XN-75-32(P)(A), Supplements 1, 2, 3, and 4, Computational Procedure for Evaluating Fuel Rod Bowing, dated October 1983 (Reference 7.15), with the exception of the fuel rod gap correlation which is updated according to Reference 7.7. The referenced methodologies incorporate DNBR and LHGR burnup thresholds and penalties that are verified on a cycle-specific basis, ensuring the criteria for fuel rod bow has been met.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 3.1.1 shows the limitations and conditions for the fuel rod bowing methodology, along with the licensee response.

Table 3.1.1 - Limitations and Conditions for TR XN-75-32 Limitation and Condition Licensee Response 1 The acceptance is not applicable to fuel designs which exhibit a greater propensity for bowing than that given in data from which the models reviewed were developed.

The fuel rod bow topical report is approved for use with the GAIA fuel assembly.

2 If the residual DNBR penalties due to fuel rod bowing are partially or totally offset by using generic or plant-specific DNBR margin, the margin used to offset these penalties must be documented in the bases to the TSs and any remnant penalties must be accommodated into the TSs.

Generic and/or plant-specific margins are not used to offset the application of residual DNBR rod bow penalties in the Byron/Braidwood AREA analysis. Since this method discussed in the condition was not performed, there is no requirement to document this in the TS bases.

3 If the inter-assembly gap distance increases by more than 50 mils, the NRC requires a more detailed analysis.

It was noted in the SER that assembly bow effects were not considered for FQ or DNBR analyses. The NRC stated that, due to a number of conservatisms, this was acceptable if the 95/95 inter-assembly gap increased by less than 50 mils. The ARITA topical report includes a methodology for calculating assembly bow penalties in Reference 11

[ANP-10339PA, Revision 0, ARITA -

ARTEMIS/RELAP Integrated Transient Analysis Methodology, 2023.], Section 9.1.3.5.

This methodology is conservative for the treatment of assembly bow.

4 For the Westinghouse fuel design, the Exxon procedure was found to be slightly non-conservative in determining DNBR values. However, ENC replied to the RAIs stating that a specific list of conservatisms exist to offset the deficiency of the model.

Framatomes current rod bow methodology is considered conservative and compliant.

The NRC staff has reviewed the licensee response to the four limitations and conditions for the rod bow methodology and finds each condition was met appropriately.

3.1.4 Neutronic Evaluation This section summarized the neutronic analyses associated with the GAIA fuel design for use at Braidwood and Byron. All neutronic design criteria are met up to the licensed peak UO2 fuel rod burnup of 62 GWd/MTU. The licensing criteria for reactivity coefficients are obtained from Section 8.4.7 of the GAIA Mechanical Design TR (Reference 7.1). From this TR, the Doppler

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION coefficient shall be negative at all operating conditions, and the power coefficient shall be negative at all operating power levels relative to hot zero power.

The nuclear design methodology is based on ARCADIA (References 7.9 and 7.10). The ARCADIA methodology was used to model both transition cores containing both VANTAGE+

and GAIA fuel and equilibrium cores of solely GAIA fuel. Therefore, the licensee used NRC-approved criteria and methodology. NRC staff concludes that these core designs show that sufficient margin exists between typical safety parameter values and the corresponding limits to ensure adequate protection of the fuel and allow flexibility in the development of reload cores.

3.2.

Core Source Term and Dose Consequences The NRC staff reviewed the regulatory and technical analyses, as related to the radiological consequences of the Maximum Hypothetical Accident (MHA) LOCA performed by CEG in support of proposed the LAR. Information regarding these analyses was provided in Attachments 1 and 4 of the LAR dated May 28, 2024, as supplemented on April 21, 2025. The NRC staff reviewed the assumptions, inputs, and methods used by CEG to assess the radiological consequences on the design basis MHA-LOCA from the proposed license amendment. All of the remaining design basis accident (DBA) dose consequence analyses, Main Steam Line Break (MSLB), Steam Generator Tube Rupture, Locked Rotor, Control Rod Ejection, and Fuel Handling Accident do not meet the criteria for requiring a LAR. The staffs findings are based on the descriptions of the licensees analyses and other supporting information submitted and docketed. CEG also calculated revised atmospheric dispersion factors used as input to the dose analyses. The majority of the initial conditions, inputs, and assumptions in the accident dose consequence analysis for the MHA-LOCA remain the same as the current licensing basis.

3.2.1 Onsite and Offsite Dispersion Modeling The accident-related radiological dose analyses accompanying the LAR required atmospheric dispersion parameters (i.e., relative concentrations or X/Qs) as direct inputs. These X/Qs are based on using appropriate dispersion models that rely, in part, on the input of representative meteorological (Met) data. The dispersion analyses for this LAR consider both offsite and onsite impacts. Offsite X/Qs are estimated at the Exclusion Area Boundary (EAB) and the outer boundary of the Low Population Zone (LPZ) to evaluate potential impacts to the public. Onsite X/Qs are estimated at the normal and/or emergency air intake locations to evaluate potential impacts to control room and Technical Support Center (TSC) habitability.

In conducting its review of the X/Qs and Met data input to those analyses, the NRC staff considered provisions in RG 1.183, as well as the guidance in RIS 2006-04.

Regulatory Position 5.3, Meteorology Assumptions, of RG 1.183 states that Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control room that were approved by the staff during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide. Similar language appears in Revision 1 of RG 1.183 with the following substantive changes to Revision 0:

o the addition of the phrase and, as applicable, the onsite emergency response facility (i.e., the TSC) following the phrase the control room, and

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION o appending a clarifying phrase to Sentence 1 that reads provided that such values remain relevant to the particular accident, release characteristics that affect plume rise, its release points, and receptor locations.

Section 4, Atmospheric Dispersion, of RIS 2006-04 called for greater resolution in the windspeed classes used to determine joint frequency distributions (JFDs) of windspeed, wind direction, and atmospheric stability class. A JFD represents the input of Met data to the dispersion modeling analyses for offsite receptors (i.e., at the EAB and LPZ) with the PAVAN model. RIS 2006-04 indicated this greater resolution, especially in the range of lower windspeeds, produced the best results.

The licensee reiterated a commitment to the NRC in Subsections 2.3.6.3.1, PAVAN Meteorological Database, of the Braidwood and Byron UFSARs (ML24103A226 and ML24103A227, respectively) to implement this finer windspeed resolution as reflected in RG 1.23, Revision 1. RIS 2006-04 was the basis for this portion of RG 1.23, Revision 1, over the previous windspeed class breakdown in Safety Guide 23, Onsite Meteorological Programs, but referred to as RG 1.23 in RG 1.183, Revision 0.

The dispersion modeling analyses for onsite receptors that impact the control room and TSC utilize a different model (i.e., ARCON96). The ARCON96 code uses an hourly Met database as one of its inputs. While the same period of record (POR) (i.e., 1994 to 1998) was used for both the offsite (PAVAN) and onsite (ARCON96) modeling analyses, only the PAVAN dispersion modeling analyses were affected by the change in resolution of the JFD windspeed classes.

3.2.1.1 Meteorological Data and Atmospheric Dispersion Estimates The LAR relies on the provision in RG 1.183 that allows the use of previously approved dispersion modeling results with some limitations. Consequently, the NRC staffs review of the offsite and onsite, accident-related atmospheric dispersion modeling results and related Met input data was limited in its scope. Therefore, no confirmatory checks of the dispersion modeling runs were made for either Braidwood or Byron.

Meteorological Data The Met data input to the previously approved offsite and onsite dispersion modeling analyses were from the Braidwood and Byron onsite Met monitoring programs. The NRC staffs evaluation and acceptance of the Met data for the 1994 to 1998 POR is documented in its February 7, 2014, issuance of license amendments regarding the measurement uncertainty recapture power uprate for Braidwood and Byron (ML13281A000). See the staffs discussions there under the headings Meteorological Data and Offsite Atmospheric Dispersion Factors.

The NRC staff notes that Regulatory Position 5.3 of RG 1.183, Revision 0, cites the Met monitoring guidance in Safety Guide 23. However, as mentioned above, the finer resolution of windspeed classes is associated with Revision 1 of RG 1.23; although neither Safety Guide 23 nor Revision 1 of RG 1.23 are specifically referenced in RG 1.183, Revision 1. The NRC staff also notes that Section B, Discussion, in RG 1.23, Revision 1, suggests that Met data used in dispersion analyses should not be older than 10 years. The 1994 to 1998 POR for the Met data used to generate dispersion factors in subsequently approved LARs for these facilities exceeds this criterion by more than 25 years.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Further, Regulatory Position 5.3 of RG 1.183, Revision 1, states that Licensees should ensure that any previously approved values remain accurate and do not include any misapplication of a methodology or calculational errors in the identified values. On that basis, the NRC staff evaluated the long-term representativeness of the 1994 to 1998 POR. The staff made cursory checks of more recent JFDs, two from each facilitys Met monitoring program, appearing in their annual radioactive effluent release reports (ARERRs). Specifically, the years 2018 and 2022 for Braidwood were reviewed (ML19121A463 and ML24250A119, respectively), and the years 2021 and 2022 were reviewed for Byron (ML22118A364 and ML23110A084, respectively).

The NRC staff recognized that year-to-year variations will be present in the Met data and resulting JFDs. These ARERRs list the JFDs on a quarterly rather than an annual basis. In addition, the breakdown of windspeed classes in these ARERRs is more similar to that in Safety Guide 23 compared to the finer resolution called for in RIS 2006-04 for dispersion modeling purposes. The staffs review focused only on stability classes E, F, and G because those stabilities result in the estimated X/Qs from which the controlling (highest) value(s) are identified. Further, F and G stability classes are associated with lower windspeeds as higher X/Q estimates are inversely proportional to the windspeed.

These ARERR summaries suggest that the 1994 to 1998 POR retained for this LAR is reasonably similar to the more recent JFDs for Byron. There is comparatively more variability in the more recent JFDs for Braidwood for E stability classes but better agreement in terms of the predominant wind direction sectors associated with the F and G stability classes. The occurrence of lower windspeeds is in general agreement for both sites.

Offsite and Onsite Atmospheric Dispersion Factors The NRC staff confirmed that the offsite and onsite X/Qs used in the May 28, 2024 LAR, represented the highest X/Q values based on previously approved LARs by verifying their traceability to earlier documents. The licensee used a conservative approach in selecting the overall highest X/Qs for input to the respective dose calculations. Specifically, the highest X/Q for a given offsite or onsite receptor and averaging time was identified regardless of whether it was associated with a potential accident release from Braidwood or Byron even though the distances and/or directions to the EAB, LPZ, control room, or TSC varied by site and unit, and the respective onsite Met datasets.

Section 4.1.2 of Attachment 4 of the LAR lists the highest offsite X/Qs at the EAB and LPZ.

Regarding traceability:

The overall highest X/Q at the EAB (i.e., for 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) is associated with a potential accident release from the Byron Unit 1 / Unit 2 outer containment wall. See Subsection 2.3.6.3.3 (Page 2.3-49e) in Revision 19 of the Byron UFSAR (ML24103A227) which is referenced to Revision 15 in December 2014 (ML14363A407).

The overall highest X/Qs at the LPZ (i.e., for 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (1 to 3 days), and 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (4 to 30 days)) are associated with a potential accident release modeled as the midpoint between Braidwood Units 1 and 2.

See Subsection 2.3.6.3.3 (Page 2.3-42d) in Revision 19 of the Braidwood UFSAR (ML24103A226) which is referenced to Revision 15 in December 2014 (ML14363A400).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Section 4.1.1 of Attachment 4 of the LAR lists the highest onsite X/Qs for the control room (CR).

Regarding traceability:

The overall highest X/Qs due to potential accident releases from containment leakage and modeled as a diffuse area source (i.e., for 0 to 0.3333 hours0.0386 days 0.926 hours 0.00551 weeks 0.00127 months at the CR fresh air intake, 0.3333 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at the turbine building emergency air intake for the CR, and 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (1 to 3 days), and 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (4 to 30 days) each at the turbine building emergency air intake for the CR) are associated with Braidwood Unit 1. See Table 2.3-53 (page 2.3-130) in Revision 19 of the Braidwood UFSAR (ML24103A226) which is referenced to Revision 12 in December 2008 (ML090270116).

The overall highest X/Qs due to emergency core cooling system leakage released through the plant vent are varied.

o For the 0 to 0.3333 hour0.0386 days 0.926 hours 0.00551 weeks 0.00127 months averaging period at the CR fresh air intake and the 0.3333 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months interval at the turbine building emergency air intake for the CR, Braidwood Unit 2 is the highest X/Q contributor. See Table 2.3-53 (page 2.3-130) in Revision 19 of the Braidwood UFSAR (ML24103A226) which is referenced to Revision 12 in December 2008 (ML090270116).

o For the 2 to 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months averaging period at the turbine building emergency air intake for the CR, Byron Unit 1 is the highest X/Q contributor. See Table 2.3-58 (page 2.3-140) in Revision 19 of the Byron UFSAR (ML24103A227) which is referenced to Revision 12 in December 2008 (ML090270120).

o For the 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (1 to 3 days), and 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (4 to 30 days) averaging periods each at the turbine building emergency air intake for the CR), Byron Unit 2 is the highest X/Q contributor. Again, see Table 2.3-58 (page 2.3-140) in Revision 19 of the Byron UFSAR (ML24103A227) which is referenced to Revision 12 in December 2008 (ML090270120).

Section 4.1.3 of Attachment 4 of the LAR lists the highest onsite X/Qs for the TSC. Regarding traceability:

The overall highest X/Qs due to containment leakage and modeled as a diffuse area source are varied.

o For the 0 to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 2 to 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months averaging periods at the TSC air intake, Braidwood Unit 1 is the highest X/Q contributor. See Table 4-3 (page 25 of 28) in an attachment to a letter from the Licensees to the NRC dated December 9, 2005 (ML060040081) titled Additional Information Related to Application of Alternative Radiological Source Term - Atmospheric Dispersion Coefficients for Braidwood Station Units 1 and 2 and Byron Station Units 1 and 2 (ML060040085).

o For the 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (1 to 3 days), and 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (4 to 30 days) averaging periods each at the TSC air intake, Byron Unit 1 is the highest X/Q contributor. See Table 4-1 (page 22 of 28) of the document indicated above.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The overall highest X/Qs due to emergency core cooling system leakage released through the plant vent are also varied.

o For the 0 to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months averaging period at the TSC air intake, Braidwood Unit 1 is the highest X/Q contributor. See Table 4-3 (page 25 of 28) of the document indicated in the first sub-bullet above.

o For the 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (1 to 3 days), and 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (4 to 30 days) averaging periods each at the TSC air intake, Byron Unit 1 is the highest X/Q contributor. See Table 4-1 (page 22 of 28) of the document indicated in the first sub-bullet above.

3.2.1.2 Dispersion Modeling Conclusion The NRC staff finds that the offsite and onsite X/Q values used as input to the respective dose calculations in the LAR are acceptable because they are consistent with a provision in Regulatory Provision 5.3 of RG 1.183 that allows the use of previously approved dispersion modeling results with some limitations. The staff notes that their previous acceptability was due in part to the conservative approach used for their selection as indicated earlier and a revision to the Met input data used for dispersion modeling at offsite receptors.

3.2.2 LOCA Radiological Consequences Analysis The revised design basis LOCA radiological analysis was performed using an NRC radiological consequence computer code, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, Version 3.03, described in NUREG/CR 6604, A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, dated April 1998 (ML15092A284). The RADTRAD code, developed by the Sandia National Laboratories for NRC, estimates transport and removal of radionuclides and radiological consequence doses at selected receptors. The NRC staff performed independent confirmatory dose evaluations, as needed, using the RADTRAD code version 3.03 (version used by the licensee in the application) and the latest version 5.04.

CEG performed an analysis of the radiological consequences of the design basis LOCA using the alternative radiological source term described in RG 1.183 Rev. 0 which is part of their current licensing basis (CLB). To show compliance with 10 CFR 50.67, the licensee calculated the TEDE at the EAB for any 2-hour period, the TEDE in the control room and at the boundary of the LPZ over the duration of the accident. The NRC staff reviewed the description of the analysis as submitted and output of the analytical codes, and finds that the licensee followed the guidance in RG 1.183 regarding dose consequence analysis calculation methodologies.

3.2.2.1 LOCA Source Term The radiological consequence analysis of the design basis LOCA assumes core melting, with release of the radioactive material to the reactor coolant system and then to the containment.

The source term was developed using the ORIGEN-ARP code which is consistent with RG 1.183, Revision 0. Release of radioactive material to the environment is assumed to occur through leakage from the containment, leakage of containment sump water from ECCS components outside containment after recirculation begins.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The maximum power level for both Braidwood and Byron is a core power level not in excess of 3,645 megawatts thermal (100 percent rated power). The thermal power uncertainty is

+/- 0.345 percent (UFSAR 15.0.3.2). This provides a core source term associated with a DBA power level of 3,657.6. The licensee uses a more conservative value of 3,658.3 megawatts thermal (MWth) as a basis for developing the core source term in this LAR.

The licensees assumptions, with regard to the source term release fractions, release timing, and radionuclide composition, follow the guidance in Regulatory Position 3 of RG 1.183.

Consistent with RG 1.183, the chemical form of radioiodine released to the containment should be assumed to be iodine chemical forms of 95 percent particulate as cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodine compounds if the containment sump pH can be shown to be at values of 7 or higher to prevent iodine re-evolution. The assumption of a pH value of greater than 7 is consistent with their CLB. The NRC staff finds that these assumptions are in accordance with the guidance in RG 1.183, Revision 0, and are acceptable.

3.2.2.2 LOCA Primary Containment (PC) Leakage In the LAR, the licensee assumed that the source term is released into the containment in accordance with RG 1.183, Table 2, PWR Core Inventory Fraction Release into Containment, and Table 4, LOCA Release Phases. The values for containment sprayed and unsprayed volumes and air exchange rate between the two volumes are unchanged from the CLB. The containment volume is 2.85x106 cubic feet with 82.5 percent of the containment volume sprayed and 17.5 percent in the unsprayed region. Transfer between these two volumes is assumed to be limited to that provided by the containment fan coolers, which are also known as deck fans.

Even without the fan coolers, the licensee considers that there would be significant mixing induced by the containment sprays and by the combination of steaming and heat transfer. Since the minimum number of fan coolers assured operable will be 2 out of 4, and the flow rate is 65,000 cubic feet per minute (cfm) per deck fan, the licensee assumed a total of 130,000 cfm flow between the unsprayed region to the sprayed region. These assumptions are also in the CLB as verified by the staff.

The licensees analysis took credit for removal of iodine in aerosol form in the containment through natural processes. The Powers Natural Deposition algorithm, based on NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, dated July 1996 (ML100130305) was used in CEG s analysis using RADTRAD.

No natural deposition of elemental or organic iodine was assumed. The lower bound, or 10th percentile, level of deposition values for aerosol deposition was used. Past 22.22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months , no further deposition is credited. These assumptions are in accordance with the guidance in RG 1.183, are part of the CLB, and are acceptable.

Credit was taken in the containment for removal of fission products other than organic iodine and noble gases by the containment spray system. The methodologies used by CEG to calculate the time constants for elemental iodine removal and particulate removal by sprays follows the guidance in RG 1.183 and more specific guidance in SRP 6.5.2, Revision 2. The Braidwood and Byron CLB includes spray removal using models from SRP 6.5.2, Revision 2.

Credit was taken for elemental iodine removal until a decontamination factor (DF) of 100 was reached. The licensee also assumed that the particulate spray removal was reduced after a DF of 50 was reached. The NRC staff verified that the spray model contains the same inputs and parameters in the previous DBA analysis contained in their alternate source term (AST)

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION (ML050560102). The licensees modeling of fission product removal in containment follows the guidance in RG 1.183, is consistent with their CLB, and is acceptable.

CEGs analysis assumed the containment leakage is at the TS value of La of 0.2 percent per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 0.1 percent per day thereafter. CEG used atmospheric dispersion factors for release from the containment in the LOCA radiological consequences analysis. The staffs review of the atmospheric dispersion factors is discussed above in Section 3.2.1, of this SE.

The NRC staff has reviewed the licensees calculations, including description of the analysis inputs, assumptions, and methodology, as well as their calculational code input and output. The staff finds that CEG s modeling of the LOCA containment leakage pathway follows the guidance of RG 1.183, and is, therefore, acceptable.

3.2.2.3 LOCA Emergency Core Cooling System Leakage The ECCS leakage is assumed to start when ECCS recirculation initiates and continues at a constant rate until the end of the accident period at 30 days. The ECCS is assumed to leak 276,000 cubic centimeters per hour (cc/hr), which is two times the administrative limit leakage value. The source term for the ECCS leakage is the radioactivity in the containment sump water, which is assumed to consist of only the iodine isotopes. The license provided an updated, more detailed description of the analysis. This update provides a filter efficiency of 100 percent aerosol, 90 percent elemental, and 90 percent organic iodine efficiency. With the iodine speciation provided in RG 1.183 of 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide, this results in an iodine speciation of 97 percent elemental and 3 percent organic available for release. This release fraction conforms with RG 1.183 source term assumptions. The volume of ECCS leakage and release fractions has not changed from the CLB with this additional detail in the analysis.

All of the above inputs to the RADTRAD code are part of the CLB. The NRC staff has reviewed the licensees calculations, including description of the analysis inputs, assumptions, and methodology, as well as their calculational code input and output associated with this LAR provided in Attachment B of the application. The staff finds that CEG s modeling of the LOCA ECCS leakage pathway follows the guidance of RG 1.183, and is, therefore, acceptable.

3.2.2.4 Control Room Modeling The air volumes of the control room at Braidwood and Byron are 230,830 ft3 and 232,872 ft3, respectively. For its analysis, CEG assumed a volume of 240,000 ft3 to model the control room at either Braidwood and Byron. The assumed volume, although larger than the actual air volume for either control room, is conservative for the calculation of dose in the control room. The licensee provided calculations in Attachment L of the application to demonstrate that the use of a slightly larger volume is conservative as opposed to the 200,000 ft3 in the CLB. The NRC staff performed confirmatory calculations using the CLB control room volume assumption and the proposed control room volume assumptions and found that the difference in control room dose is approximately 0.030 rem TEDE. This difference does not affect the ability to comply with control room dose limits. Further, the licensee performed several sensitivity studies to confirm the assumption of a slightly larger control room volume is conservative with respect to dose consequences. The NRC staff has reviewed this change to control room volume and finds the change acceptable.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION In the emergency mode of operation, Braidwood and Byron control room filtration systems have a filtered recirculation flow of 43,500 cfm and filtered outside air makeup airflow of 6,000 cfm

(+/-10 percent), totaling 49,500 cfm of total combined flow to the control room volume. A portion of this intake flow goes to the upper cable spreading room (CSR) volume and is un-recirculated.

The maximum value of this un-recirculated intake flow is 2,500 cfm. The un-recirculated intake flow for each plant is subtracted from the 49,500 cfm total combined flow, making the assumed adjusted combined flow to the control rooms to be 47,000 - 49,500 cfm. Due to the split in airflow to the upper CSR, there is a reduction of up to 303 cfm makeup flow, and a reduction of 2,917 cfm recirculation flow. CEG assumed makeup filtration efficiencies of 99 percent for particulates and 95 percent for elemental and organic iodine. These filter efficiency values and flow rates are the same values as the CLB.

The licensees analysis does not credit the control room filtration system emergency mode of operation for the first 20 minutes of the accident. This is a change from the CLB value of 30 minutes to align main control room ventilation in emergency mode. This new value is based upon a newly analyzed allowance for manual control room Mode 2 isolation. During this time, intake of unfiltered makeup air is assumed. The licensee addressed this in its response to request for additional information dated April 21, 2025. In the April 21, 2025, supplement, the licensee provided the Operator Response Time Validation Sheets which demonstrate the ability to realign the main control room ventilation system to emergency mode of operation which meets the 20 minutes used in the dose analysis calculations. These validation tests confirm that the use of 20 min critical action time is conservative. The validation tests demonstrate that the maximum amount of time required to shift ventilation is 14 minutes.

Therefore, the use of 20 minutes in the calculations meets the guidance provided in RG 1.194, which states that a conservative delay time should be assumed for the operator to complete the necessary actions to isolate the control room. The NRC staff reviewed the information provided to support the change to the operator action time and find the change acceptable.

In the last docketed calculation, the unfiltered inleakage was modeled as 1,000 cfm after control room emergency ventilation is initiated and remains at that value for the duration of the event.

This value was subsequently reduced to 500 cfm and then to 436 cfm via licensee-controlled processes. The current value of 436 cfm is supported and bounded by the maximum measured inleakage value of 318 standard cubic feet per minute (scfm) for Byron, and 155 scfm for Braidwood. The control room unfiltered inleakage was measured by means of tracer gas testing based on the ASTM standard ASTM E-741-00, Standard Test Method for Determining Air Change in a Single Zone by means of a Tracer Gas Dilution. The tests were completed on May 31, 2019, for Byron and November 16, 2024, for Braidwood. The NRC staff finds that the licensees assumption of 436 cfm control room unfiltered inleakage is acceptable.

The remainder of the initial conditions, inputs, and assumptions for the control room analysis are unchanged from the CLB. The NRC staff performed sensitivity studies using RADTRAD which showed that the changes were either conservative (higher calculated dose) or had negligible effects on control room dose. These sensitivity studies were performed on individual changes, as well as the changes as a collective. The NRC staff finds that CEG s modeling of the control room follows the guidance of RG 1.183, and is, therefore, acceptable.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.2.2.5 Control Room Gamma Shine Dose The calculation for gamma shine dose in the control room has not changed from their CLB and remains the same as the CLB. The licensee calculated total direct gamma shine dose in the control room is 0.051 rem TEDE.

3.2.2.6 Technical Support Center Dose The Braidwood and Byron UFSAR, Appendix E, Requirements Resulting from [Three Mile Island, Unit 2 (TMI-2)] Accident, addresses requirements resulting from the Three Mile Island, Unit 2 accident, discussed in NUREG-0737, Clarification of TMI Action Plan Requirements, dated November 1980 (ML051400209). CEG determined that the dose in the TSC is impacted by the implementation of the LAR. CEG recalculated the TSC dose using the same CLB AST LOCA release assumptions with new source term values. As in the control room analysis, the licensee did not take credit for operation of the emergency filtration system in the TSC for 30 minutes. The TSC nominal intake rate of 900 cfm was decreased by 10 percent during the initial 30-minute period. The licensees sensitivity calculations determined that this decrease is the most conservative assumption which resulted in the highest TSC doses. For the duration of the event, a TSC intake rate of 810 cfm (10 percent reduction of 900 cfm) was maintained.

Unfiltered inleakage was assumed to be 450 cfm (half of the nominal intake rate). The charcoal adsorber filter efficiency is reduced to 95% from the CLB 99%. The NRC staff finds that this is a conservative assumption and provides more operational margin and is acceptable. The TSC recirculation flow is assumed to be 990 cfm, which is the 1,100 cfm nominal flow minus 10 percent to allow for uncertainty and minimize activity removal. All of these assumptions are contained in the CLB. The licensee provided the RADTRAD output files for calculation of the TSC dose. The NRC staff performed confirmatory calculations, reviewed the licensees RADTRAD file and found initial conditions, inputs and assumptions for the TSC dose calculations to be acceptable. The NRC staff finds that the reanalyzed TSC dose results are within regulatory limits, and are, therefore, acceptable.

3.2.2.7 LOCA Radiological Consequences Conclusion The NRC staff reviewed the information provided in the licensees LAR, as supplemented, the Braidwood and Byron CLB, and the RADTRAD calculation output provided by the licensee. The staff verified that CEG s LOCA dose analysis used assumptions and inputs that follow the guidance in RG 1.183. The licensees most limiting calculated LOCA dose results are given in Table 3.2.1. Assumptions used by the licensee and evaluated by the NRC staff are listed in Tables 3.2.2 and 3.2.3 and show the changes from the CLB and the values used in the LAR. The licensees calculated doses are within the 10 CFR 50.67 and GDC 19 radiological dose acceptance criteria for a LOCA. These TEDE criteria are 25 rem at the EAB for the worst two hours, 25 rem at the LPZ for the duration of the accident and 5 rem in the control room for the duration of the accident.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 3.2.1 - MHA-LOCA Radiological Consequences (utilizing the limiting calculated dose results from GAIA Fuel)

Source Control Room (rem TEDE)

EAB (rem TEDE)

LPZ (rem TEDE)

PC Leakage 4.34 16.95 4.05 ECCS Leakage 0.35 0.29 0.29 Direct Dose from Containment, External Plume, & CR Filters 0.05 None None Total Dose 4.74 17.24 4.34 Regulatory Dose Limit 5

25 25 Table 3.2.2 - Control Room Parameters Input/Assumption CLB Value New LAR Value Reason for Change Control Room Volume 2.0E+05 ft3 (AST SE)

Byron 230,830 ft3 Braidwood 232,872 ft3 2.4E+05 ft3 Single volume used for both Braidwood and Byron control rooms.

Licensee provided calculations to demonstrate that using the larger bounding volume results in higher calculated control room dose and is therefore conservative Normal Operation Filtered Makeup Flow Rate 6000 cfm 6000 cfm No change. Values used in calculations are the same used in CLB Filtered Recirculation Flow Rate 43,500 +/- 10 percent cfm 43,500 +/- 10 percent cfm No change. Values used in calculations are the same used in CLB Emergency Operation Recirculation Mode Control Room Isolation (Mode 2) 30 min 20 min Licensee performed and provided sensitivity studies which support this change Filtered Makeup Flow Rate 6,000 +/- 10 percent (AST SE) 5,097 - 6,600 cfm 6000 cfm (used in calculations)

No change. Values used in calculations are the same used in CLB

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Filtered Recirculation Flow Rate 43,500 +/- 10 percent cfm 36,953 - 47,850 cfm 43,500 cfm (used in calculations)

No change. Values used in calculations are the same used in CLB Unfiltered Inleakage 1000 cfm (AST) 436 (CLB) 436 cfm Supported by current tracer gas tests Byron = 318 scfm (May 2019)

Braidwood = 155 scfm (Nov. 2024)

Filter Efficiencies CR Makeup Filters Elemental - 95%

Organic - 95%

Aerosol/Particulate -

99%

Elemental - 95%

Organic - 95%

Aerosol/Particulate -

99%

No change. Values used in calculations are the same used in CLB CR Recirculation Filters Elemental - 90%

Organic - 90%

Aerosol/Particulate -

80%

Elemental - 90%

Organic - 90%

Aerosol/Particulate - 0%

No credit is taken for Aerosol/Particulate filtration. This is a conservative assumption Occupancy Breathing Rate 3.50E-4 m3/sec (0-720 hr) 3.50E-4 m3/sec (0-720 hr)

No change. Values used in calculations are the same used in CLB 0-24 hours 1-4 days 4-30 days 100%

60%

40%

100%

60%

40%

No change. Values used in calculations are the same used in CLB Table 3.2.3 - LOCA Inputs and Assumptions Input/Assumption CLB Value for Offsite and Control Room New LAR Value for Offsite and Control Room Reason for Change Power Level (MWth) 3,658.3 3,658.3 No change. Values used in calculations are the same used in CLB Containment Containment Volume 2,850,000 ft3 2,850,000 ft3 No change. Values used in calculations are the same used in CLB Iodine Chemical Form 95% aerosol, 4.85%

elemental, and 0.15%

organic 95% aerosol, 4.85%

elemental, and 0.15%

organic No change. Values used in calculations are the same used in CLB Containment Sump pH

>7.0

>7.0 No change. Values used in calculations are the same used in CLB

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Containment Sprayed Volume 82.5% (2,351,000 ft3) 82.5% (2,351,000 ft3)

No change. Values used in calculations are the same used in CLB Containment unsprayed Volume 17.5% (500,000 ft3) 17.5% (500,000 ft3)

No change. Values used in calculations are the same used in CLB Containment Spray Start Time 0.025 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months (1.5 minutes) 0.025 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months (1.5 minutes)

No change. Values used in calculations are the same used in CLB Containment Spray Stop Time 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours No change. Values used in calculations are the same used in CLB Although sprays may run longer, credit for spray removal of aerosols is stopped at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Containment Spray Flow Rate 2950 gpm 2950 gpm No change. Values used in calculations are the same used in CLB Elemental Iodine Spray Removal Coefficient 20 hr-1 20 hr-1 No change. Values used in calculations are the same used in CLB Number of Deck Fans Operating, credited for mixing 2 of 4 2 of 4 No change. Values used in calculations are the same used in CLB Deck Fan Flow Rate (per fan) 65,000 scfm 65,000 scfm No change. Values used in calculations are the same used in CLB Aerosol Spray Removal Coefficient P1 P2 (after a decontamination factor of 50 is achieved) 6.0 hr-1 0.6 hr-1 6.0 hr-1 0.6 hr-1 No change. Values used in calculations are the same used in CLB Organic Iodine Spray Removal None None No change. Values used in calculations are the same used in CLB Natural Deposition Elemental, Organic-

None, Aerosol10%

Elemental, Organic-

None, Aerosol10%

No change. Values used in calculations are the same used in CLB

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Containment Leakage Rate 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours to 30 days 0.2%/day 0.1%/day 0.2%/day 0.1%/day No change. Values used in calculations are the same used in CLB Containment Leakage Filtration 0%

0%

No change. Values used in calculations are the same used in CLB Containment Sump Volume:

at switchover after balance of RWST injection 38,979 ft3 58,506 ft3 38,979 ft3 58,506 ft3 No change. Values used in calculations are the same used in CLB ECCS Leakage Flow Rate:

Administrative limit:

Assumed in analysis 138,000 cc/hr 276,000 cc/hr 138,000 cc/hr 276,000 cc/hr No change. Values used in calculations are the same used in CLB ECCS Flashing Fraction 10%

10%

No change. Values used in calculations are the same used in CLB Offsite Breathing Rates 0-8 Hours 8-24 Hours 1-30 Days 3.50E-4 m3/sec 1.80E-4 m3/sec 2.30E-4 m3/sec 3.50E-4 m3/sec 1.80E-4 m3/sec 2.30E-4 m3/sec No change. Values used in calculations are the same used in CLB 3.2.3 Core Source Term and Dose Consequences Conclusion As described above, the NRC staff reviewed the assumptions, inputs, and methods used by CEG to assess the radiological impacts of implementation of an MHA-LOCA at Braidwood and Byron. Consistent with SRP Section 15.0.1, the staff compared the results estimated by the licensee to results estimated by the staff in its confirmatory calculations based on docketed information submitted by the licensee. The staff finds that CEG used analysis methods and assumptions consistent with the regulatory requirements and guidance. The NRC staff compared the doses estimated by CEG to the applicable criteria identified in 10 CFR 50.67, as supplemented in Regulatory Position 4.4 of RG 1.183, and GDC 19, as supplemented by Section 6.4 of NUREG-0800. The NRC staff finds, with reasonable assurance, that the licensees estimates of the EAB, LPZ, and control room doses will continue to comply with these criteria. Therefore, the proposed changes are acceptable with regard to the radiological consequences of postulated DBAs.

3.3 Proposed TS Changes

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensee proposed revisions to several TSs to support loading Framatome GAIA fuel as described above in Section 2.1 of this SE. The NRC staffs review of the proposed changes is documented below.

3.3.1 TS 2.1.1, Reactor Core SLs The LAR describes the changes to the existing TS safety limits 2.1.1.1, 2.1.1.2, and 2.1.1.3 as follows, The safety limits are updated to reflect fuel-and method-specific limits. The proposed change will revise existing TS 2.1.1.1, TS 2.1.1.2, and TS 2.1.1.3 to clarify applicability to Westinghouse fuel; no changes to the values are proposed. The NRC staff finds that the proposed clarifications are appropriate to ensure compliance with safety limits during the two transition cycles when both GAIA and Westinghouse VANTAGE+ fuel will co-reside in the core.

The staff finds that these limits continue to provide adequate protection when utilized with the additional Framatome safety limits 2.1.1.4, 2.1.1.5, and 2.1.1.6 described further below.

After the two-cycle transition period, it is expected that the Westinghouse VANTAGE+ safety limits will no longer be needed, and the NRC staff finds that safety limits 2.1.1.4, 2.1.1.5, and 2.1.1.6 described further below will provide adequate protection when the core is composed entirely of Framatome GAIA fuel.

As part of this LAR, the licensee is adding additional safety limits for DNBR for the Framatome GAIA fuel as TS 2.1.1.4. Specifically, the language in the LAR states that 2.1.1.4 For Framatome fuel, the Departure from Nucleate Boiling Ratio (DNBR) shall be maintained 1.12 for the ORFEO-GAIA DNB correlation and 1.15 for the ORFEO-NMGRID DNB Correlation.

ORFEO-GAIA and ORFEO-NMGRID (ANP-10341P-A, Rev 0) have previously been approved for use with the Framatome ARITA-ARTEMIS/RELAP integrated transient analysis methodology (ANP-10339P-A, Rev 0) and is therefore acceptable to use in determining the safety limit as well. The Braidwood and Byron plant-specific technical reports for rod ejection (ANP-4086P) and non-LOCA transients (ANP-4087P) are also compatible with the ORFEO-GAIA and ORFEO-NMGRID correlations and the limitations and conditions have been found by the staff to have been correctly applied.

Given the ORFEO-GAIA and ORFEO-NMGRID methodologies are appropriate to be applied in either MODE 1 (Power Operation) or MODE 2 (Startup), the NRC staff finds that the proposed TS 2.1.1.4 will provide equivalent protection for the FRAMATOME GAIA fuel, as TS 2.1.1.1 and TS 2.1.1.2 provide for the Westinghouse VANTAGE+ fuel. The staff also finds that the proposed TS 2.1.1.4 meets the criteria for a safety limit for a nuclear reactor as defined in 10 CFR 50.36(c)(1)(A).

As part of this LAR, the licensee proposed adding additional safety limit equations in TS 2.1.1.5 and TS 2.1.1.6 to include the fuel centerline melt limits for both the uranium oxide fuel and the uranium oxide fuel with gadolinium, respectively. The fuel centerline melt limits in GALILEOTM are provided in ANP-10339-P-A, Revision 0, ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology, Section 4.2.4.7.1. These equations were approved as part of the GALILEO TR, ANP-10323P-A, Revision 1.

Specifically, the proposed TS 2.1.1.5 states For Framatome fuel, for UO2 fuel, the peak centerline temperature shall be maintained < 5090 °F, decreasing by 13.7 °F per 10,000 MWD/MTU burnup. The proposed TS 2.1.1.6 states For Framatome fuel, for

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION UO2-Gd2O3 fuel, the peak centerline temperature shall be maintained < 5090 °F - 9360*E2 -

399.6*E, decreasing by 13.7 °F per 10,000 MWD/MTU burnup where E is the gadolinia weight-fraction.

The NRC staff finds that the proposed safety limit 2.1.1.5 will provide adequate protection to centerline melting for fuel rods without homogeneous poisons, and that the limit is established appropriately using approved methodologies.

The NRC staff finds that the proposed safety limit 2.1.1.6 will provide adequate protection to centerline melting for fuel rods with gadolinia homogeneous poisons, and that the limit is established appropriately using approved methodologies.

Together the NRC staff finds that safety limits 2.1.1.5 and 2.1.1.6 provide equivalent protection for the Framatome GAIA fuel as safety limit 2.1.1.3 provides for the Westinghouse VANTAGE+

fuel and meets the criteria for a safety limit for a nuclear reactor as defined in 10 CFR 50.36(c)(1)(A).

Appendix C was revised for the Braidwood and Byron Facility Operating Licenses under Amendment No. 122 for Braidwood and Amendment No. 127 for Byron (ML020590491). The existing licensing bases reference Appendix C of the Braidwood and Byron Facility Operating Licenses, Units 1 and 2 contain a condition that states that if any fuel pellets incorporating homogeneous poisons are used, the TR documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in the license condition, as well as modification of TS 2.1.1.3 to include the fuel centerline melt temperature limit for the fuel with the homogeneous poison. This license condition is applicable only to VANTAGE+ fuel. The licensee stated that no homogeneous poisons will be used in the Westinghouse VANTAGE+

fuel going forward.

The NRC staff finds that it is appropriate to control safety limits 2.1.1.5 and 2.1.1.6 for Framatome GAIA fuel using the approved COLR methodologies being added to TS 5.6.5 as part of this amendment and therefore use of Appendix C is not required for Framatome GAIA fuel.

In the application, the licensee indicated that homogenous poisons are not currently being used in the installed Westinghouse VANTAGE+ fuel, therefore the NRC staff finds that use and applicability of Appendix C is precluded during the two-cycle transition period.

3.3.2 TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))

The licensees LAR submittal states that The proposed change to TS 3.2.1 includes replacing the Westinghouse FQW(Z) with the Framatome FQV(Z). This supports the re-baselining of specific accident analyses to Framatome methods. Accordingly, ACTIONS B is revised to remove REQUIRED ACTIONS B.1, B.3, and B.4.

The NRC staff reviewed the Framatome FQV(Z) methodology which is based on the Power Distribution Control A (PDC-A) methodology outlined in ANP-10339-P-A, Revision 0, ARITA -

ARTEMIS/RELAP Integrated Transient Analysis Methodology, Section 15.2. As stated previously this methodology has already been approved for use by the NRC staff.

Removal of REQUIRED ACTIONS B.1, B.3, and B.4 is appropriate because as stated in the licensees LAR submittal Reduction in the THERMAL POWER is not required in the event that

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION the FQV(Z) is not within the limit of TS 3.2.1.B because reduction of the axial flux distribution (AFD) limits accounts for reduction in the FQV(Z) that would have been realized with more restrictive AFD limits. This is consistent with the ARITA' methodology and allows for the removal of ACTIONS B.1 with no reduced thermal power, which allows for the removal of B.3 and B.4. The AFD reduction will be developed using approved methods and the result documented in the COLR.

The NRC staff finds that the basis for the changes to TS 3.2.1 described in the submittal are consistent with the approved methodology in ANP-10339-P-A, Rev. 0 and therefore should be implemented. Specifically, that the FQV(Z) methodology (ACTIONS B) allows for Axial Flux Difference (AFD) limits alone to provide equivalent protection in concert with the FQV(Z) methodology (ACTIONS A) controlling any hot channel power reductions as necessary.

The NRC staff also finds that proposed revisions to TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)), continue to provide equivalent protection for continuing plant operation and continue to maintain the requirements of 10 CFR 50.36(c)((2)(ii)(B) Criterion 2 for a limiting condition for operation that is an operating restriction which maintains initial conditions of a design basis accident which assumes failure or challenge to a fission product barrier.

3.3.3 TS 3.5.1, Accumulators The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a large break loss-of-coolant accident (LBLOCA), to provide inventory during the refill phase that follows thereafter, and to provide reactor coolant system (RCS) makeup for a SBLOCA. The minimum required boron concentration is necessary to assure reactor subcriticality in a post-LOCA environment, particularly in LBLOCA where no credit is taken for control rod assembly insertion. The maximum boron concentration is used in determining the cold leg to hot leg recirculation switchover time and minimum sump pH.

TS 3.5.1, Accumulators, provides operability requirements for the ECCS accumulators. To support the transition to GAIA fuel, the licensee proposed a change to the boron concentration limits defined in SR 3.5.1.4 and SR 3.5.1.5. The proposed change would increase the lower limit from 2,200 ppm to 2,600 ppm and the upper limit from 2,400 ppm to 2,900 ppm.

In Section 3.3 of Attachment 1 of the LAR, the licensee states that there are two significant changes between VANTAGE+ fuel and GAIA fuel:

The GAIA fuel rod diameter is 0.374 inches compared to the VANTAGE+ diameter of 0.360 inches.

GAIA fuel designs use gadolinia burnable poison, where VANTAGE+ use Integral Fuel Burnable Absorber (IFBA), a coating of Zirc Diboride onto selected fuel pellets).

The effects of the fuel properties above require higher boron concentrations to meet post-LOCA shutdown margin limits. The licensee evaluated representative GAIA cycles to forecast the boron concentration requirements. The licensee stated that the proposed values will be confirmed in cycle-specific core designs as part of the reload process.

In the Realistic Large Break Loss-of-coolant Accident (RLBLOCA) analysis, described in more detail below in Section 3.4.2 of this SE, the licensee used a conservative value for accumulator boron concentration of 2,150 ppm. The results from the analysis demonstrate that the

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION acceptance criteria for PCT, CWO and MLO from 10 CFR 50.46(b)(1), (2) and (3) have been met. Therefore, NRC staff finds the accumulator boron concentration lower limit of 2,600 ppm acceptable.

The licensee evaluated the post-LOCA containment recirculation sump pH using methodology consistent with the current licensing basis. The calculated minimum required containment recirculation sumps pH value decreased from 8.0 to 7.9, attributed to the increase in the maximum boron concentration of 2,900 ppm in the accumulators and RWST. The minimum pH value of 7.9 is greater than the minimum pH of 7.0 given in NUREG-0800, Branch Technical Position 6-1, pH for Emergency Coolant Water for Pressurized Water Reactors, to reduce the probability of stress-corrosion cracking of austenitic stainless-steel components, nonsensitized or sensitized, nonstressed or stressed.

The licensee stated that the solubility of boric acid is 4,563 ppm (2.61 weight percent) at 32°F with a higher value at higher temperatures. Given that the proposed upper limit on boron concentration of 2,900 ppm is lower than 4,563 ppm, the NRC staff finds that the proposed change has no effect on the boron solubility limits. Therefore, NRC staff finds the accumulator boron concentration upper limit of 2,900 ppm acceptable.

3.3.4 TS 3.5.4, Refueling Water Storage Tank (RWST)

The RWST supplies borated water to the chemical and volume control system during abnormal operating conditions, to the refueling pool during refueling, and to the ECCS and the containment spray system during accident conditions.

TS 3.5.4 provides operability requirements for the RWST. To support the transition to GAIA fuel, the licensee proposed a change to the boron concentration limits defined in SR 3.5.4.4. The proposed change would increase the lower limit from 2,300 ppm to 2,700 ppm and the upper limit from 2,500 ppm to 2,900 ppm. As discussed during the regulatory audit, the licensee conservatively does not model boron concentration in the RWST for RLBLOCA calculations. As discussed above in Section 3.2.3, the effects of the GAIA fuel properties require higher boron concentrations to meet post-LOCA shutdown margin limits. The licensee evaluated representative GAIA cycles to forecast the boron concentration requirements and stated that the proposed values will be confirmed in cycle-specific core designs as part of the reload process.

As described in Section 3.4.3, Post-LOCA Evaluations, of Attachment 1 to the LAR, the licensee plans to follow 10 CFR 50.59 for these evaluations which retain the same methods and acceptance criteria, except for pH requirements (discussed in the following paragraph). These evaluations include post-LOCA subcriticality, post-LOCA boron precipitation and hot leg switchover time, and post-LOCA long-term core cooling. Based on its review of the above information, NRC staff finds that the RWST boron concentration lower limit of 2,700 ppm is acceptable.

As discussed above in Section 3.2.4, the licensee evaluated the post-LOCA containment recirculation sump pH using methodology consistent with the current licensing basis. The licensee determined that the increase in the maximum boron concentration of 2,900 ppm in both the accumulators and RWST resulted in the minimum required containment recirculation sumps pH decreasing from 8.0 to 7.9. Since the value of 7.9 is greater than the minimum pH of 7.0 specified in NUREG-0800, Branch Technical Position 6-1, pH for Emergency Coolant Water for Pressurized Water Reactors, the NRC staff finds the RWST boron concentration upper limit of 2,900 ppm acceptable.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.3.5 TS 4.2.1, Fuel Assemblies TS 4.2.1 specifies the fuel rod cladding material approved for use. GAIA fuel uses M5Framatome cladding material. As described in Section 3.1 of this LAR, GAIA fuel with M5Framatome cladding is acceptable for use in Braidwood and Byron. The NRC staff finds that the addition of M5Framatome to TS 4.2.1 is acceptable.

3.3.6 TS 5.5.16, Containment Leakage Rate Testing Program TS 5.5.16 establishes reference values used in the licensees containment leakage rate testing program as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The containment leakage rate testing program acceptance criteria are based off peak calculated containment internal pressure for the design basis LOCA (Pa). As a result of changes in containment response evaluated in Section 3.6, the proposed change would increase Unit 2 Pa from its existing value of 38.4 psig to 38.7 psig. Based on the evaluation discussed in Section 3.6, the NRC staff finds the new Unit 2 Pa value acceptable.

3.3.7 TS 5.6.5, Core Operating Limits Report (COLR)

TS 5.6.5.a contains a list of safety limits and LCOs whose specific values are documented in the COLR. The proposed changes to TS 5.6.5.a are to include reference to LCO 3.1.4, Rod Group Alignment Limits, LCO 3.3.1, Reactor Trip System (RTS) Instrumentation, and LCO 3.3.9, Boron Dilution Protection System (BDPS) to better reflect information already contained in the COLR. NRC staff finds this change acceptable because it is an administrative action to reflect the actual contents of the existing COLR and is unrelated to the fuel transition.

TS 5.6.5.b contains a list of analytical methods used in-core operating limits development which have been previously approved for use by NRC staff. The proposed changes to the TS 5.6.5.b include the addition of 14 NRC-approved topical reports, as listed above in Section 2.1, for GAIA fuel used to develop core operating limits for this proposed fuel transition. The NRC staff finds this change acceptable based on the overall evaluation of the fuel transition as described in this safety evaluation.

3.4 Loss-of-Coolant Accident (LOCA) Analysis The NRC regulations require that licensees of operating LWRs analyze a spectrum of accidents including LOCA to assure adequate core cooling under the most limiting set of postulated design basis conditions. LOCAs are postulated accidents that would result in the loss of reactor coolant from piping breaks in the RCS primary boundary at a rate greater than the reactor coolant makeup system is able to replenish. Loss of significant quantities of reactor coolant would prevent heat removal from the reactor core unless the water is replenished. The small and large break LOCA analysis are discussed in Sections 3.4.1 and 3.4.2 respectively.

3.4.1 Small Break LOCA (SBLOCA) Analysis The licensee performed an SBLOCA analysis as documented in Framatome reports ANP-4074, Revision 0, Byron/Braidwood Unit 1 Small Break LOCA Analysis, dated February 2024 (Attachments 9 (non-proprietary) and 15 (proprietary) to the LAR), and ANP-4115, Revision 0, Byron/Braidwood Unit 2 Small Break LOCA Analysis, dated November 2024 (Attachments 2 (non-proprietary) and 8 (proprietary) from the supplement dated April 21, 2025) to support the planned fuel transition to GAIA fuel at Braidwood and Byron.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4.1.1 SBLOCA Description The postulated SBLOCA is defined as a break in the RCS pressure boundary with an area less than or equal to 10 percent of the cold leg pipe area. The reactor protection system (RPS) and ECCS are provided to mitigate these accidents. The most limiting break location for SBLOCA analysis performed is in the cold leg pipe on the discharge side of the reactor coolant pump (RCP). This break location results in the largest amount of RCS inventory loss, the largest fraction of ECCS fluid discharged out the break, and the largest pressure drop between the core exit and the top of the downcomer. The SBLOCA event progression develops in the following distinct phases: (1) subcooled depressurization (also known as blowdown), (2) natural circulation, (3) loop seal clearing, (4) core boiloff, and (5) core recovery and long-term cooling.

The duration of each of these phases is break size and system dependent. The licensee provided a detailed description of each of the phases in Framatome reports ANP-4074 and ANP-4115.

3.4.1.2 SBLOCA Methodology The licensee performed the SBLOCA analysis using the NRC-approved SBLOCA methodology documented in TR EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001, Supplement 1(P)(A) to TR EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, December 2016, and TR ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021.

The SBLOCA EM uses a deterministic approach based on the requirements of 10 CFR 50 Appendix K to determine the expected PCT, MLO and CWO response. The licensee used the EM for event response of the primary and secondary systems as well as the hot fuel rod and used the following two computer codes:

The GALILEO code to determine the burnup-dependent initial fuel rod conditions for the system calculations.

The S-RELAP5 code to predict the primary and secondary system T-H and hot rod transient response.

The S-RELAP5 code is used in the NRC-approved SBLOCA methodology as documented in TR EMF-2328(P)(A). The use of S-RELAP5 and the RODEX2A fuel performance code (FPC) is specified for SBLOCA analysis as described in Supplement 1 to TR EMF-2328(P)(A). However, in TR ANP-10349P-A, Framatome supplemented the approved EMs and implemented the GALILEO FPC in S-RELAP5. In the SE to TR ANP-10349P-A, the NRC staff concluded that the GALILEO FPC is an acceptable supplement for RODEX2A for SBLOCA evaluation models. The NRC staff therefore finds the GALILEO and S-RELAP5 codes appropriate for use with the applied methods.

3.4.1.3 SBLOCA Analysis The goal of the analysis is to demonstrate that the ECCS, while operating with GAIA fuel in the core, will continue to satisfy the ECCS acceptance criteria given in 10 CFR 50.46(b)(1) through (b)(4). A break spectrum analysis for SBLOCA was performed for breaks of varying diameters of up to 10 percent of the flow area for the cold leg pump discharge consistent with the EM. The spectrum analyzed included a break size range from 1.0 to 8.7 inches in diameter, with a break size interval sufficient to establish a PCT trend.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION In addition to the cold leg pump discharge break spectrum analysis, as required by the methodology, the licensee performed sensitivity studies for a delayed RCP trip, a break in an attached pipe, and a different ECCS temperature. For the delayed RCP trip, an operator action time of 5 minutes after the specified trip criteria is met was analyzed. The licensee performed an analysis of the ruptures in attached piping that compromise the ability to inject emergency coolant into the RCS. The piping study analyzed breaks in the accumulator line and high head safety injection line. The ECCS temperature sensitivity study analyzed the sensitivity to ECCS fluid temperatures different from those used in the break spectrum analysis.

The licensee performed SBLOCA analysis to support plant operation at a core power level of 3,658 megawatt thermal (MWt) (including measurement uncertainty), a maximum-allowed total peaking factor (FQ) of 2.6 (with uncertainties applied and an axial-dependent factor k(z) set to 1.0), a nuclear enthalpy rise factor/radial peaking factor of (FH) 1.70 (including measurement uncertainty), and up to 5 percent steam generator (SG) tube plugging per SG in Byron/Braidwood Unit 1 and up to 10 percent SG tube plugging/SG in Byron/Braidwood Unit 2.

3.4.1.4 SBLOCA Results The licensees SBLOCA break spectrum analysis for Braidwood and Byron Unit 1 resulted in a limiting PCT of 1,653°F for an 8.1-inch diameter cold leg pump discharge break. The same break produced the limiting CWO of 0.013 percent. The ECCS temperature study produced the limiting MLO of 3.69%. The total MLO value includes ((

))

For Braidwood and Byron Unit 2, the PCT is calculated to be 1,625°F for an 8.7-inch diameter cold leg pump discharge break. The same break size resulted in the limiting CWO of 0.010 percent and a limiting MLO of 3.35 percent. The total MLO value includes ((

))

Therefore, the NRC staff finds that the analysis results demonstrate the adequacy of the ECCS to satisfy the criteria given in 10 CFR 50.46(b)(1) to (b)(3). Further, maintaining compliance to 10 CFR 50.46(b)(1) to (b)(3) criteria also ensures that the 10 CFR 50.46(b)(4) criteria on maintaining the core amenable to cooling will be satisfied. The NRC staff evaluation on maintaining the coolable geometry of the fuel under the seismic and LOCA load combination to satisfy the 10 CFR 50.46(b)(4) criteria is provided in Section 3.1.1 of this SE.

3.4.1.5 Compliance with NRC Staff Imposed Limitations and Conditions on SBLOCA Methodologies Sections 3.5 of ANP-4074 and ANP-4115 discuss the limitations and conditions for the SBLOCA methodology. For each of the three methodologies used in the SBLOCA analysis, the limitations and conditions for each are discussed below.

For TR EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001, there was one condition imposed on the use of S-RELAP5 as follows:

That while it has been shown in Reference 53 [NUREG/CR-4945] that the thermal-hydraulic phenomena observed for breaks up to 10 percent of the cold leg flow area are the same, if the code is used for break sizes larger than 10 percent of the cold leg flow area additional assessments must be performed to ensure that the code is predicting the important phenomena which may occur.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensee performed analysis on a spectrum of cold leg break sizes from 1.0-inch diameter to 8.7-inch diameter (10 percent of cold leg pipe area). However, in the accumulator line pipe break sensitivity, the break area was equivalent to an 8.75-inch diameter (accumulator line inside diameter) which is larger than 10 percent of the cold leg flow area. This is addressed and found acceptable in Supplement 1 to EMF-2328 and therefore, NRC staff finds this limitation and condition has been met.

For Supplement 1(P)(A) to TR EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, December 2016, the SE stated:

The NRC staff mentions that it is necessary for all SBLOCA submittals utilizing the Reference 1 [EMF-2328(P)(A), Revision 0, Supplement 1, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2012]

methodology identify the critical break size, at and below which, only one loop seal clears of liquid. The NRC staff further requires that the largest small break that depressurizes to a pressure just above the SIT actuation pressure be included in the break spectrum evaluation. The ((

)) diameter break increment resolution is expected to capture this particular break size; however, it is mentioned and emphasized here since it is important to locate this break size as it could be the limiting small break.

Table 4-2 of both ANP-4074 and ANP-4115 provide results for the cold leg pump discharge SBLOCA break spectrum. Part of this table provides the loop seal clearing times. For Braidwood and Byron Unit 1, the results in ANP-4074 shows that the ((

)) For Braidwood and Byron Unit 2 results in ANP-4115, ((

))

As to the accumulator actuation, a footnote in Table 4-2 of ANP-4074 for Braidwood and Byron Unit 1 states ((

)) For Braidwood and Byron Unit 2, Table 4-2 of ANP-4115 states that ((

)) Based on its review of the above information, NRC staff finds the limitations and conditions met for Supplement 1(P)(A) to TR EMF 2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S RELAP5 Based, December 2016.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION For TR ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021, Section 4.0, Limitations and Conditions, of its SE states the following:

The demonstrated range of applicability of the methodology, specifically RLBLOCA and SBLOCA (EMF-2103 Rev 3, and EMF-2328 Rev 0 and Supplement 1) and applicable range (not related to the thermo-mechanical method) of applicability of GALILEO topical report (ANP-10323P) shall be implemented in the supplement EM (ANP-10349).

TR EMF-2103 is for the RLBLOCA analysis and is discussed below in Section 3.4.2.5, Compliance with NRC Staff Imposed Limitations and Conditions on RLBLOCA Methodologies, of this SE. The range of applicability for EMF-2328 and its supplement were discussed above.

The applicable range of ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, November 2020 are provided in Section 1.2, Limits of Code Applicability, of its SE. Table 3.4.1 below describes each item along with how the licensee meets the applicable conditions.

Table 3.4.1 - Range of Applicability for ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors Item Licensee response Pressurized water reactor designs using Low-Enriched Uranium (LEU) fuel loading This analysis was performed for the Byron/Braidwood Units 1 and 2 plants, which are PWRs, using LEU fuel.

Rod average burnups up to ((

] gigawatt days per metric ton of uranium (GWd/MTU) for Zircaloy-4 and up to ((

)) GWd/MTU for M5 cladding The fuel burnups applied in this analysis do not exceed the rod average burnup of

((

))

Zircaloy-4 and M5 cladding The analysis supports operation with M5Framatome cladding.

Rod diameter between ((

)) mm and

((

)) mm This analysis was performed using fuel with a rod outside diameter of 9.5 mm.

Uranium 235U enrichments up to 5 weight percent (wt%)

The 235U enrichments applied in this analysis do not exceed 5 weight percent.

Gadolinia concentrations up to 10 wt%

Gadolinia fuel is not analyzed as part of the SBLOCA methodology. Therefore, this parameter is not subject to the limitation for this LOCA analysis.

Nominal true pellet density ranging from

((

)) percent of the theoretical density of UO2 The initial pellet density is ((

)) percent of the theoretical density of UO2.

Fuel grain sizes ranging from ((

)) microns (mean linear intercept)

This analysis was performed using fuel pellets with a grain size of ((

))

Pellets manufactured by dry conversion and ammonium diuranate The fuel pellet manufacturing process for the fuel design considered in this analysis is dry conversion and ammonium diuranate.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Item Licensee response Fuel temperature up to the melting point to the approved burnup range This is related to thermo-mechanical methods and is not subject to the limitation for this LOCA analysis.

Cladding strain up to the approved transient clad strain limit This is related to thermo-mechanical methods and is not subject to the limitation for this LOCA analysis.

Internal rod pressure up to pressures that protect from clad lift-off and hydride reorientation This is related to thermo-mechanical methods and is not subject to the limitation for this LOCA analysis.

Fuel rod power not to exceed levels as limited by fuel melt, cladding strain, and rod pressure criteria This is related to thermo-mechanical methods and is not subject to the limitation for this LOCA analysis.

Based on the licensee responses to the limits of code applicability summarized in Table 3.4.1 above, the NRC staff finds the licensee meets all the limits of code applicability. It should be noted that the last four items in Table 3.4.1 are related to thermo-mechanical methods and are not subject to the limitation on appliable range for the LOCA analysis as stated in Section 4.0, Limitations and Conditions, of the SE approving ANP-10349-PA, Revision0, GALILEO Implementation in LOCA Methods. Therefore, based on the review of the licensee responses to the limits of code applicability summarized above, the NRC staff finds all limitations and conditions satisfied for the methodologies used in the SBLOCA evaluations.

3.4.1.6 SBLOCA Conclusions The NRC staff reviewed the information in the licensees submittal pertaining to the analysis of the SBLOCA event with Framatome GAIA fuel to support plant operation at a core power level of 3,658 MWt (includes measurement uncertainty), a maximum-allowed local peaking factor (FQ) of 2.6 (with uncertainties applied and an axial-dependent factor k(z) set to 1.0), a nuclear enthalpy rise factor/radial peaking factor (FH) of 1.7 (including measurement uncertainty), and up to 5 percent SG tube plugging per SG for Braidwood and Byron Unit 1 and up to 10 percent for Braidwood and Byron Unit 2.

The NRC staffs review verified that SBLOCA break spectrum analysis results meet the limiting PCT limits and the total MLO and CWO limits set by 10 CFR 50.46(b)(1) through (b)(3). The NRC staff finds the delayed RCP trip study performed by the licensee to be acceptable as it shows that there is at least 5 minutes for operators to trip all four RCPs after the trip criteria is met with margin to the 10 CFR 50.46(b)(1-4) criteria. The NRC staff finds the results from analysis of the ruptures in attached piping to be acceptable as they are less limiting than the limiting break spectrum case. The NRC staff finds that the licensees analysis showed it will continue to meet GDCs 4, 27, and 35 of Appendix A to 10 CFR 50 and the requirements of Appendix K to 10 CFR 50.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4.2 Realistic Large Break LOCA (RLBLOCA) Analysis The licensee performed an RLBLOCA analysis as documented in Framatome reports ANP-4075, Revision 0, Byron/Braidwood Unit 1 Large Break LOCA Analysis, dated February 2024 (Attachments 8 (non-proprietary) and 14 (proprietary) to the LAR), and ANP-4116, Revision 0, Byron/Braidwood Unit 2 Large Break LOCA Analysis, dated November 2024 (Attachments 3 (non-proprietary) and 9 (proprietary) to the supplement dated April 21, 2025), to support the planned fuel transition to GAIA fuel at Braidwood and Byron.

3.4.2.1 RLBLOCA Description During normal plant operation at full power, an LBLOCA is initiated by a postulated rupture of the RCS primary piping. The most limiting break is an instantaneously occurring break in the cold leg piping between the RCP and the reactor vessel. A worst-case single failure is also assumed to occur during the accident. The single failure for this analysis, as defined in the EM, is the loss of one ECCS injection train without the loss of containment spray.

The LBLOCA is described in three phases: the blowdown phase, the refill phase, and the reflood phase. The licensee described these phases in Section 3.2, Description of LBLOCA Event, of Framatome reports ANP-4075 and ANP-4116.

3.4.2.2 RLBLOCA Methodology The NRC-approved TR EMF-2103P-A, Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016, describes the Framatome methodology developed for the realistic evaluation of a LBLOCA for PWRs with recirculation (U-tube) SGs. It covers Westinghouse 3-loop and 4-loop plant designs and Combustion Engineering plants, all with fuel assembly lengths of 14 feet or less and ECCS injection to the cold legs. Since Braidwood and Byron are 4-loop Westinghouse-designed PWRs with recirculation SGs, have fuel assembly length of less than 14 feet, and ECCS injection into the cold legs, the NRC staff finds this methodology is applicable to Braidwood and Byron for the LBLOCA analysis. The EM in TR EMF-2103P-A for the LBLOCA response of the RCS, secondary system, and the fuel rods used in the analysis is based on the use of the following two computer codes:

GALILEO for computation of the initial fuel stored energy, fission gas release, and the transient fuel cladding gap conductance.

S-RELAP5 code for the thermal-hydraulic system calculations, which includes ICECON for containment response.

The use of S-RELAP5 and the COPERNIC FPC is specified for RLBLOCA analysis as described in TR EMF-2103P-A. However, in TR ANP-10349P, Framatome supplemented the approved EMs and implemented the GALILEO FPC in S-RELAP5. In the SE to TR ANP-10349P, the NRC staff concluded that the GALILEO FPC is an acceptable supplement for COPERNIC for RLBLOCA evaluation models. The NRC staff therefore finds the GALILEO and S-RELAP5 codes appropriate for use with the applied methods.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4.2.3 RLBLOCA Analysis The licensees LBLOCA analysis is based on a statistical realistic LOCA EM in accordance with the methodology in TR EMF-2103P-A instead of conservative EMs specified by 10 CFR 50, Appendix K. For performing the statistical analysis, the licensee created ((

)) The licensee sampled each key input parameter over a range established through code uncertainty assessment or expected operating limits provided either by TSs or plant data. The licensee considered the key LOCA parameters listed in Table A-6 of TR EMF-2103P-A, and the uncertainty range associated with each of these parameters given in Table A-7 of TR EMF-2103P-A.

Tables 4-1 of ANP-4075 and ANP-4116 show the plant parameters and ranges used for the analysis for Byron/Braidwood Units and Byron/Braidwood Unit 2 respectively. The analysis assumes full power operation at a core power level of 3,658 MWt (including measurement uncertainty), a local peaking factor (FQ) of 2.6, a nuclear enthalpy rise factor/radial peaking factor (FH) of 1.7, and up to 5 percent SG tube plugging per SG for Byron/Braidwood Unit 1 and up to 10 percent for Byron/Braidwood Unit 2. This analysis also addresses typical operational ranges or TS limits for items such as pressurizer pressure and level, accumulator pressure, temperature, and level, loop flow, and containment pressure and temperature. The analysis explicitly analyzes fresh and once-burned GAIA fuel assemblies with and without gadolinia.

The summary of the major parameters and event times for the demonstration case analysis are identified in Tables 4-5 and 4-6 of ANP-4075 and ANP-4116. The analysis uses the fuel swelling, rupture, and relocation (FSRR) model to determine if cladding rupture occurs and evaluate the consequences of FSRR on the transient response. Section 7.9.3.3, Clad Ballooning, Rupture and Area Adjustment Models, of TR EMF-2103P-A, provides a discussion and consequences of FSRR and is documented in the supporting analyses in the TR. ((

))

3.4.2.4 RLBLOCA Results Tables 4-4 of ANP-4075 and ANP-4116 provide the Unit 1 and Unit 2 results respectively of the licensees analysis for demonstrating compliance with 10 CFR 50.46(b)(1), (b)(2), and (b)(3).

Table 3.4-2 below (extracted from ANP-4075 and ANP-4116) shows the upper tolerance limit (UTL) for 95/95 simultaneous coverage/confidence results for PCT, MLO, and CWO for

((

)) cases.

Table 3.4.2 - RLBLOCA Upper Tolerance Limit for 95/95 Simultaneous Coverage/Confidence Results Criteria Unit 1 Unit 2 Acceptance Criteria

((

)) ((

))

PCT (°F) 1,680

((

))

1,767

((

))

2,200.0 MLO (%)

4.33

((

))

6.01

((

))

17.0 CWO (%)

0.04

((

))

0.05

((

))

1.0

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The results in Table 3.4.2 above shows the limiting ((

)) results for 95/95 simultaneous coverage/confidence meet the 10 CFR 50.46(b)(1-3) criteria. The analysis results for Byron/Braidwood Unit 1 have a PCT of 1,680°F, MLO of 4.33 percent, and a total CWO of 0.04 percent. The PCT of 1,680°F occurred in a fresh UO2 rod. Byron/Braidwood Unit 2 have a PCT of 1,767°F, MLO of 6.01 percent, and a total CWO of 0.05 percent. The PCT of 1,767°F occurred in a fresh UO2 rod. Therefore, the NRC staff finds that the results of the licensees RLBLOCA analysis demonstrate that acceptance criteria in 10 CFR 50.46(b)(1), (b)(2), and (b)(3) are met.

The NRC staff evaluation on maintaining the coolable geometry of the fuel under the seismic and LOCA load combination to satisfy the 10 CFR 50.46(b)(4) criteria is provided in Section 3.1.1 of this SE.

3.4.2.5 Compliance with NRC Staff Imposed Limitations and Conditions on RLBLOCA Methodologies For the application of the EMF-2103P-A methodology, there are 11 limitations and conditions listed in Section 4.0 of the NRC staffs SE for TR EMF-2103P-A. The licensees compliance statements for these are provided in Section 3.7 of ANP-4075 and ANP-4116.

Table 3.4.3 - Limitations and Conditions for EMF-2103P-A, Revision 3 Limitations and Conditions for EMF-2103P-A Licensee Response 1

This EM was specifically reviewed in accordance with statements in EMF-2103, Revision 3. The NRC staff determined that the EM is acceptable for determining whether plant-specific results comply with the acceptance criteria set forth in 10 CFR 50.46(b), paragraphs (1) through (3).

AREVA did not request, and the NRC staff did not consider, whether this EM would be considered applicable if used to determine whether the requirements of 10 CFR 50.46(b)(4),

regarding coolable geometry, or (b)(5), regarding long-term core cooling, are satisfied. Thus, this approval does not apply to the use of SRELAP5-based methods of evaluating the effects of grid deformation due to seismic or LOCA blowdown loads, or for evaluating the effects of reactor coolant system boric acid transport. Such evaluations would be considered separate methods.

The analysis applies only to the acceptance criteria set forth in 10 CFR 50.46(b), paragraphs (1) through (3).

2 EMF-2103, Revision 3, approval is limited to application for 3-loop and 4-loop Westinghouse-designed nuclear steam supply systems (NSSSs), and to Combustion Engineering-designed NSSSs with cold leg ECCS injection, only. The NRC staff did not consider model applicability to other NSSS designs in its review.

Byron/Braidwood Units 1 and 2 are 4-loop Westinghouse-designed NSSS PWRs with cold leg ECCS injection.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitations and Conditions for EMF-2103P-A Licensee Response 3

The EM is approved based on models that are specific to AREVA proprietary M5Framatome fuel cladding. The application of the model to other cladding types has not been reviewed.

The analysis was performed with M5Framatome cladding material.

4 Plant-specific applications will generally be considered acceptable if they follow the modeling guidelines contained in Appendix A to EMF-2103, Revision 3. Plant-specific licensing actions referencing EMF-2103, Revision 3, analyses should include a statement summarizing the extent to which the guidelines were followed, and justification for any departures.

Additional Discussion Should NRC staff review determine that absolute adherence to the modeling guidelines is inappropriate for a specific plant, additional information may be requested using the RAI process. For example, if a specific plant shows heightened PCT sensitivity to containment parameters, the NRC staff may request additional information seeking justification for the application of the containment modeling guidelines to that particular plant.

The analysis was performed using the modeling guidelines contained in Appendix A of EMF-2103P-A, Revision 3.

5 The response to RAI 15 indicates that the fuel pellet relocation packing factor is derived from data that extend to currently licensed fuel burnup limits (i.e., rod average burnup of

((

))). Thus, the approval of this method is limited to fuel burnup below this value.

Extension beyond rod average burnup of

((

)) would require a revision or supplement to EMF-2103, Revision 3, or plant-specific justification.

The burnup values applied in the analysis do not exceed the rod average burnup of

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitations and Conditions for EMF-2103P-A Licensee Response 6

The response to RAI 15 indicates that the fuel pellet relocation packing factor is derived from currently available data. Should new data become available to suggest that fuel pellet fragmentation behavior is other than that suggested by the currently available database, the NRC may request AREVA to update its model to reflect such new data.

Such a request would be tendered by a letter from the NRC to AREVA identifying the newly available data and requesting an update to the model, or an assessment to demonstrate that such an update is not needed.

The analysis uses the approved EMF-2103P-A, Revision 3 relocation packing factor application. ((

))

7 The regulatory limit contained in 10 CFR 50.46(b)(2), requiring cladding oxidation not to exceed 17 percent of the initial cladding thickness prior to oxidation, is based on the use of the Baker-Just oxidation correlation. To account for the use of the C-P [Cathcart-Pawel]

correlation, this limit shall be reduced to 13 percent, inclusive of pre-transient oxide layer thickness.

Additional Discussion Should the NRC staff position regarding the application of the 17 percent Baker-Just acceptance criterion to the C-P correlation change, the NRC will notify AREVA with a letter either revising this limitation or stating that it is removed.

For this analysis the MLO UTL is less than 13%.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitations and Conditions for EMF-2103P-A Licensee Response 8

In conjunction with Limitation 8 [7] above, C-P oxidation results will be considered acceptable, provided plant-specific ((

)) If second-cycle fuel is identified in a plant-specific analysis, whose

((

)),

the NRC staff reviewing the plant-specific analysis may request technical justification or quantitative assessment, demonstrating that

((

))

Additional Discussion This limitation ensures that the safety analysis retains sufficient margin to the Equivalent Cladding Reacted (ECR) analytic limit to

((

))

All second-cycle fuel rod ((

))

9 The response to RAI 13 states that all operating ranges used in a plant-specific analysis are supplied for review by the NRC in a table like Table 8-8 of EMF-2103, Revision 3. In plant-specific reviews, the uncertainty treatment for plant parameters will be considered acceptable if plant parameters are ((

)), as appropriate.

Alternative approaches may be used, provided they are supported with appropriate justification.

Additional Discussion This limitation ensures that the safety analysis adequately covers the range of permissible plant operation, as discussed in Section 3.4.2 of this SE. However, this limitation should not be construed to imply that exceeding limiting values by any amount is acceptable; sampling distributions for plant parameters should be realistic and justifiable.

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitations and Conditions for EMF-2103P-A Licensee Response 10

((

))

This analysis uses ((

))

11 Any plant submittal to the NRC using EMF-2103, Revision 3, which is not based on the first statistical calculation intended to be the analysis of record must state that a re-analysis has been performed and must identify the changes-that were made to the evaluation model and/or input in order to obtain the results in the submitted analysis.

Additional Discussion Adherence to this processensures that the fidelity of the chosen tolerance level is preserved in the analysis.

The present analysis is the first statistical application of EMF-2103P-A, Revision 3 for this plant.

The NRC staff reviewed the licensee response to all 11 limitation and conditions above in Table 3.4.3 and finds that the licensee has satisfactorily met each limitation and condition for use of TR EMF-2103P-A, Revision 3.

For TR ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021, Section 4.0, Limitations and Conditions, of its SE states the following:

The demonstrated range of applicability of the methodology, specifically RLBLOCA and SBLOCA (EMF-2103 Rev 3, and EMF-2328 Rev 0 and Supplement 1) and applicable range (not related to the thermo-mechanical method) of applicability of GALILEO topical report (ANP-10323P) shall be implemented in the supplement EM (ANP-10349).

With one exception, the licensee response to the limitations and conditions for the use of GALILEO for RLBLOCA are all identical to the responses for SBLOCA as described above in Section 3.4.1.5 of this SE. The one difference is the use of gadolinia fuel in the RLBLOCA evaluations. Given that the licensee states that the Gd2O3 concentration analyzed does not exceed 10 wt%, the NRC staff finds that the specific limitation related to gadolinia concentration is met.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4.2.6 RLBLOCA Conclusions The NRC staff reviewed the information in the licensees submittal pertaining to the analysis of the RLBLOCA event with Framatome GAIA fuel to support plant operation at a core power level of 3,658 MWt (includes measurement uncertainty), a maximum-allowed total peaking factor (FQ) of 2.6 (with uncertainties applied and an axial-dependent factor k(z) set to 1.0), a nuclear enthalpy rise factor/radial peaking factor (FH) of 1.7 (including measurement uncertainty), and up to 5 percent SG tube plugging per SG for Byron/Braidwood Unit 1 and up to 10 percent SG tube plugging for Byron/Braidwood Unit 2. The NRC staffs review verified that the RLBLOCA analysis used the approved methodology in TR EMF-2103P-A, Revision 3, and the results meet the limits as set by 10 CFR 50.46(b)(1) through (3) related to PCT, total MLO, and CWO. In addition, the licensee has met all the limitations and conditions appliable to the RLBLOCA methodologies.

3.4.3 Post-LOCA Evaluations The licensee stated several additional items affected by LOCA and the changes to the ECCS accumulators and RWST boron concentration change in Section 3.4.3 of the LAR which are not submitted for review or approval but will be implemented in accordance with the requirements of 10 CFR 50.59. The NRC staff reviewed these items for information and applicability to other sections of the application but is not approving or denying any of the listed items in this section.

The NRC inspection and enforcement programs continue to review site changes subject to 10 CFR 50.59 implementation.

3.4.4 Generic Safety Issue (GSI) - 191 As part of the proposed transition to GAIA fuel, the licensee re-examined the results from their assessment of Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (ML042360586), also commonly referred to as GSI-191, Assessment of Debris Accumulation on PWR Sump Performance. The objective is to ensure that post-accident debris blockage will not impede or prevent the operation of the ECCS and containment spray system in recirculation mode at PWRs during LOCAs or other high energy line break accidents for which sump recirculation is required. The Braidwood and Byron GL 2004-02 assessment for the current Westinghouse-designed fuel follows the methodology from WCAP-16793, Revision 2, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid. The NRC approved this assessment in a letter dated May 19, 2016 (ML15296A358). However, the methodology from WCAP-16793, Revision 2, is not directly applicable to GAIA fuel as the GAIA fuel was not tested or evaluated as part of that program.

Framatome has since performed testing in accordance with WCAP-17788, Volume 6, Comprehensive Analysis and Test Program for GSl-191 Closure (PA-SEE-1090) - Subscale Head Loss Test Program Report. The testing resulted in GAIA fuel specific limitations that have been incorporated in WCAP-17788, Revision 1.

WCAP-17788, Comprehensive Analysis and Test Program for GSI [Generic Safety Issue]-191 Closure, was intended to support the closure of GSI-191 and GL 2004-02 by creating a methodology to define an in-vessel fibrous debris limit to respond to in-vessel questions in GL 2004-02. The NRC staff reviewed WCAP-17788 but did not complete its review or determine that WCAP-17788 could be approved for use by licensees. However, during its review of WCAP-17788, the NRC staff identified a significant amount of evidence indicating that potential

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION in-vessel downstream effects have low safety significance as documented in a technical evaluation report (non-public). The NRC staff then developed review guidance (ML19228A011) and found that many of the methods developed in WCAP-17788 may be used by PWR licensees in demonstrating adequate long-term core cooling.

The licensee used methods and analytical results developed in WCAP-17788, Revision 1 to address in-vessel downstream debris effects for Braidwood and Byron. The licensee addressed the applicability to WCAP-17788 methods and results and found the results were slightly outside the WCAP-17788 criteria, and therefore, provided justification why these deviations are acceptable. A summary of the key parameters is presented below in Table 3.8.1 of this SE. The containment recirculation sump switchover time was found to be outside the assumptions in WCAP-17788. The analysis in WCAP-17788 assumes a sump switchover time of 20 minutes, while the Braidwood and Byron time was found to be 9.692 minutes, based on two trains of the ECCS and containment spray in operation. As stated in Section 2.3.1 of the NRC staff developed review guidance, the licensee can justify the earlier sump switchover time by crediting conservatism in the thermal power level assumed in the analysis relative to the plants actual rated thermal power level, as well as by evaluating the debris transport behavior assumed at sump switchover, relative to a more realistic consideration of the possible debris transport. The 20-minute time is based on the 10 CFR 50 Appendix K decay heat model (1.2 times the 1971 ANS Infinite Standard). The licensee is using a more realistic, yet still conservative, decay heat model based on the ANSI/ANS 51.1-1979 standard as described in the Realistic Large Break LOCA Methodology TR (EMF-2103). Using a core power level of 3,658 MWt, the WCAP-17788 core decay heat at 20 minutes is calculated as 87.4 MWt.

Using the same core power level, the Braidwood and Byron core decay heat at 9.692 minutes is calculated as ((

)) MWt, which is less than that assumed in WCAP-17788. In addition, additional margin is credited based on the debris transport time expected in the plant relative to that assumed in WCAP-17788. The generic analysis for Westinghouse-designed plants assumes that all debris begins to arrive at the time of sump switchover and increases to the maximum value over the next 60 seconds. For Braidwood and Byron, debris will not begin to arrive at the RCS for at least 45 seconds after the start of sump switchover assuming that debris starts at the containment recirculation sumps screen and travels at the containment recirculation sumps recirculation flow rate through the shortest path to the RCS. Maintaining the conservative assumption that the debris will subsequently build up over 60 seconds, the total amount of debris will not reach the core inlet until 45 + 60 = 105 seconds after sump switchover. The calculated core power at this time is ((

)) MWt, which is less than that assumed in WCAP-17788.

For the minimum ECCS recirculation flow, the analysis in WCAP-17788 models a hot leg break where the full ECCS flow reaches the reactor core. As stated in the supplement dated April 21, 2025, the total flow to the RCS for a single ECCS train is 3,208 gpm for the hot leg break. This results in an equivalent 16.6 gpm/FA for the single train of ECCS, therefore, the flow rate resulting from two ECCS trains will be significantly higher than the minimum flow rate of 18 gpm/FA from WCAP-17788.

Table 3.4.4 - Key Parameter Values for the In-Vessel Debris Effects

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Parameter WCAP-17788 Value Braidwood/Byron Value Licensee Evaluation Maximum Total In-Vessel Fiber Load (g/FA)

((

))

11.9 Maximum in-vessel fiber load is less than WCAP-17788 limit.

Maximum Core Inlet Fiber Load (g/FA)

((

))

11.9 Maximum core inlet fiber load is less than the limit for GAIA fuel.

Minimum Sump Switchover Time (min) 20 9.692 Credit for the RLBLOCA decay heat model results in lower decay heat at the time of sump switchover for the Braidwood and Byron Units compared to the analyzed value.

Minimum Chemical Precipitate Time (hr) 2.4 24 Potential for complete core inlet blockage due to chemical product generation would occur later than assumed.

Maximum Hot Leg Switchover Time (hr) 24 6.167 Latest hot leg switchover occurs well before the earliest potential chemical product generation.

Thermal Power (MWt) 3,658 3,658 Same thermal power yields same decay heat.

Maximum Alternative Flow Path (AFP)

Resistance Volume 4, Table 6-1 Volume 4 RAIs, Table RAI-4.2-24 AFP resistance is less than the analyzed value, which increases the effectiveness of the Alternate Flow Path.

Minimum ECCS Recirculation Flow (gpm/FA) 18*

16.6**

ECCS flow rate is greater than minimum analyzed value.

The minimum flow rate of 18 gpm/FA is the flow rate from two ECCS trains

- The value of 16.6 gpm/FA is for a single ECCS train as noted in the April 21, 2025 supplement.

Based on the above information, the key parameters for in-vessel debris effects meet the acceptance criteria as specified in WCAP-17788 and the NRC staff review guidance. Therefore, the NRC staff finds that the licensee meets the requirements of 10 CFR 50.46(b)(5) in regard to adequate long-term core cooling.

3.5 Containment Integrity Based on possible differences in the fuel decay heat and the stored sensible energy in the reactor internals (for example in fuel assemblies and other components), the fuel transition from a full Westinghouse core to a mixed Westinghouse and Framatome GAIA core may impact the LOCA containment AORs. For Braidwood and Byron, the licensee states in the submittal that the higher core stored energy will affect the calculated peak containment pressure, temperature, and the containment recirculation sumps water temperature following a LOCA. As a result, for

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Unit 2, the licensee proposes to change the Pa value in TS 5.5.16, Containment Leakage Rate Testing Program, from 38.4 psig to 38.7 psig. Unit 1 Pa remains unchanged.

3.5.1 Peak Containment Pressure and Pa Impact The submittal states the impact the change in fuel will have on the peak containment pressure using the GAIA fuel penalty as calculated for Braidwood and Byron Unit 1, as those values bound the Unit 2 results. UFSAR Table 6.2-27, Double-Ended Pump Suction Break Energy Balance Minimum Safeguards - Unit 1 lists the inputs used in containment analysis, including initial conditions of 23.64 MBTU of core stored energy with a total energy of 968.81 MBTU. The licensee calculated the GAIA fuel will result in an increase of 2.6 MBTU in-core stored energy at the start of an accident. The licensee then applied this increase in-core stored energy into the time period between the end of blowdown and the time of the reflood peak to determine the increased peak pressure, using linear extrapolation. This value is estimated to be 0.10 psig for a double-ended pump suction break which the licensee also applied to a double-ended hot leg break. For additional conservatism, the licensee applied a peak pressure increase of 0.3 psig to account for the change in fuel. This results in a proposed TS 5.5.16 Pa value of 42.8 psig for Braidwood and Byron Unit 1 and 38.7 psig for Braidwood and Byron Unit 2. The Unit 1 value remains the same as the existing TS 5.5.16 Pa while the proposed Unit 2 value is 0.3 psig higher. The NRC staff determined the peak containment pressure remains below the containment design pressure of 50 psig as described in UFSAR Table 6.2-1, Unit 1 Containment Peak Pressure and Temperature and Table 6.2-1a, Unit 2 Containment Peak Pressure and Temperature.

3.5.2. Peak Containment Temperature To calculate changes to peak containment temperature due to the fuel transition, the licensee correlated the change in temperature to the change in vapor saturation temperature related to the calculated difference in pressure. This correlates to a containment temperature increase of 0.3 °F for Braidwood and Byron. For Unit 1, peak containment temperature would be 263.8 °F and for Unit 2 it would be 257 °F. The NRC staff determined this peak temperature remains below the containment design temperature of 280 °F as described in UFSAR Section 6.2.1, Containment Functional Design.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.5.3 Containment Recirculation Sumps Water Temperature and Post-Accident Net Positive Suction Head (NPSH)

The licensee calculated the temperature increase due to the increased core stored energy from the GAIA fuel, applying the entire increase to the containment recirculation sumps. This action conservatively provides the highest potential increase in sump temperature, as it eliminates the possibility of the energy to be absorbed by the containment structure itself or removed by containment cooling systems. UFSAR Section 6.3, Emergency Core Cooling System and Section 6.5.2, Containment Spray System discuss the NPSH requirements of the residual heat removal pumps and the containment spray pumps and states that the NPSH calculations for the ECCS pumps recirculation mode assume that the vapor pressure of the liquid in the sump is equal to the containment ambient pressure. The NRC staff determined that the minimal temperature increase would result in a correspondingly small increase in vapor pressure which remains below the design limits of containment. In addition, the licensee states that the increased core energy from GAIA requires increased boron addition to the accumulators and RWST, which results in more mass to the sumps at a higher temperature, providing further margin to post-accident NPSH requirements.

3.5.4 Containment Analysis Supplement In its August 6, 2025, supplement, the licensee described a recently discovered issue related to the ECCS recirculation phase of a LOCA. The issue is related to a specific scenario where the component cooling flow to the Residual Heat Removal (RHR) heat exchanger is below the required value of 5,000 gpm that is credited in the containment integrity analysis. The licensee stated that Westinghouse performed sensitivity analysis and determined that there are small increases in the containment temperature, pressure and sump water temperature during the ECCS recirculation phase following a LOCA. However, as stated by the licensee, these small increases do not have any impact on the post-LOCA maximum calculated containment temperature and pressure as the peak values occur during the ECCS injection phase.

Therefore, NRC staff finds the identified issue in the August 6, 2025, supplement does not have an adverse impact on the limiting containment pressure and temperature results. In addition, NRC staff finds that the small increase in sump water temperature (< 0.25°F) over that reported in Table 3.5.3-1 of the LAR would not have an adverse impact on the pumps taking suction from the sump.

3.6 Control Rod Ejection Accident (REA)

The NRC staff reviewed the REA analysis provided in report ANP-4086P, Revision 0 (Reference 7.8) which was provided with the LAR (Attachment 10 for the non-proprietary version and Attachment 16 for the proprietary version).

3.6.1 Accident Description and Analysis Method The event is initiated by a postulated rupture of a control rod drive mechanism housing. Such a rupture would theoretically allow the full system pressure to act on the drive shaft, which would eject its control rod from the core. The consequences of the postulated failure would be a rapid positive reactivity insertion, a core power excursion, and an increase in radial power peaking, which potentially leads to localized fuel rod damage. The power excursion would be mitigated by the fuel temperature (Doppler) feedback, and, in some cases terminated by the RPS with a reactor trip in response to changes in neutron flux or system pressure. Although the initial increase in power would occur too rapidly for control rod scram to affect the power increase, the

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION negative reactivity inserted during scram would affect the fuel temperature and fuel rod cladding surface heat flux.

The analysis was performed based on a representative cycle design and is also applicable to transition cycles (containing co-resident VANTAGE+ fuel with GAIA fuel) and cycle designs containing a full core of GAIA, provided the conditions of the cycle design are bounded by the AREA analysis described in Reference 7.8. The analysis was performed using the ARCADIA code system (References 5.9 and 5.10) which demonstrates compliance with the criteria defined in RG 1.236. The criteria within the AREA methodology consists of the following:

((

))

The enthalpy rise limit is based on excess hydrogen as defined in RG 1.236. The enthalpy limit used for high-temperature cladding failure threshold in RG 1.236 is a function of internal pin pressure with a maximum of 170 calories per gram (cal/g) for internal pressures less than system pressure and a minimum limit of 100 cal/g for internal pressures higher than system pressure.

RG 1.236 has the following restrictions for coolability:

Peak radial average fuel enthalpy must remain below 230 cal/g.

A limited amount of fuel melting is acceptable provided it is restricted to the fuel centerline region and is less than 10% of pellet volume. The peak fuel temperature in the outer 90% of the pellets volume must remain below incipient fuel melting conditions.

Methodology Departures:

GALILEO (Reference 7.12) is used as the FPC in this analysis. ((

))

((

))

The update of the ((

)) the methodology as described in RG 1.236. The NRC staff has reviewed the methodology departures, as described above, and finds they are acceptable for the Braidwood and Byron analyses as both departures are conservative relative to the approved methodology.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Methodology Clarifications:

The version of GALILEO used in this analysis is provided in Reference 7.12. The version of GALILEO used in the approved AREA methodology (Reference 7.11) is a prior revision of GALILEO provided in Reference 7.13. The newer revision of GALILEO was benchmarked to the prior revision and showed comparable results, with some improvements in the results. These improvements were not factored into this analysis.

The NRC staff reviewed the above information and finds that not considering improvements as a result of code/methodology improvements is a conservative approach and acceptable for use in the Braidwood and Byron AREA evaluation.

For non-prompt scenarios with uncharacteristic pulse widths (the empirical database is compromised of narrow pulses), ((

))

The NRC staff reviewed the above information and finds that this approach is conservative and acceptable for use in the Braidwood and Byron AREA evaluation.

The enthalpy limits from Reference 7.11 are based on prompt critical testing and are to protect against clad overheating and PCMI failure. For non-prompt critical ejected rod events, there is no fast power pulse and the power deposition occurs over a longer time period so DNBR is a failure mechanism. This impacts selection of acceptance criteria and how the results are presented so ((

))

The NRC staff reviewed the above information and finds that the approach taken between prompt critical and non-prompt critical cases is consistent with the guidance in RG 1.236, Revision 0, Pressurized-Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents, and therefore, acceptable.

System pressure calculations were not performed as part of the Braidwood and Byron AREA analysis.

Based on its review, the NRC staff finds that not performing system pressure calculations is in accordance with the approved AREA methodology, and is acceptable.

3.6.2 Cycle Inputs The Byron/Braidwood REA analysis was performed for a full core of Framatome GAIA fuel with M5 cladding, and the impact of transition cycles is considered through the development of cycle-to-cycle biases.

The analysis is performed for ((

)) times in life (TIL) and at ((

)) power levels for each TIL. The TILs considered are ((

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

)) The selected power levels are ((

))

Table 3.6.1 lists the uncertainties applied in the AREA analysis with the ((

)) provided in Figure 2-1 of Reference 7.8.

Table 3.6.1 - Core Model Parameter Uncertainties

))

Table 2-3 of Reference 7.8 shows the core model parameter biases, which extend the REA to cover both transition and future cycles.

The NRC staff has reviewed the cycle inputs as provided in the LAR and finds the inputs used in the analysis are consistent with the NRC-approved AREA methodology (Reference 7.7) and are acceptable for the control element assembly ejection analysis.

3.6.3 REA Limits Generated by GALILEO

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The PCMI limits are specified in Section C.3.2 of RG 1.236. The excess hydrogen is calculated using GALILEO (Reference 7.12). ((

))

3.6.4 Fuel Integrity Summaries

((

)) The margins reported are based on the calculated value minus the limit, so that a negative number is favorable. A positive value indicates a violation of the limit.

Additional details are provided for the cases with the least margin to the limit for fuel melt, fuel rim melt, MDNBR, enthalpy, and enthalpy rise. Limiting cases for enthalpy, enthalpy rise, and fuel rim temperature are provided for those cases which are ((

))

((

))

The results reported in Tables 4-2 through 4-6 of Reference 7.8 are summarized below in Table 3.6.2, which provides limiting criteria for power level, cycle burnup, limiting value, and estimated level of conservatism (limiting value minus nominal value).

Table 3.6.2 - Measure of Conservatism for Limiting Results

((

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

))

The NRC staff reviewed the Braidwood and Byron REA analysis as described in Reference 7.8.

The NRC staff determined that the analysis was performed according to the NRC-approved AREA methodology and is consistent with RG 1.236, with the clarifications and departures as described above. The fuel-related acceptance criteria for this event are evaluated to support the fuel transition. The AREA methodology implementation for mixed core applications is addressed in the development of parameter biasing to account for cycle-to-cycle changes with co-resident fuel as described in Reference 7.8. The NRC staff determined that the REA analysis provides adequate margin to limits for fuel temperature, fuel rim temperature, MDNBR, and enthalpy rise which assures that any fuel rod failures are below the limits provided in RG 1.236.

3.7 Non-LOCA Transients/Accidents Attached to the LAR submittal, the Byron/Braidwood UFSAR Chapter 15 non-LOCA transient analyses supporting the introduction of GAIA fuel at Braidwood and Byron are provided in Attachments 11 (non-proprietary) ANP-4087NP and 17 (proprietary) ANP-4087P titled Byron Braidwood Non-LOCA Summary Report with ARITA Methodology.

3.7.1 Non-LOCA Transients/Accidents Analysis The analyses are performed in accordance with the NRC-approved method, ANP-10339P-A, Revision 0. The demonstration analyses are performed using a representative cycle design and illustrate how these approved methods will be applied to future Braidwood and Byron reloads.

Their future applicability to transition cycles and full core GAIA cycles will be evaluated on a cycle-specific basis. ARITA has been shown to adequately model the important phenomena and the integrated plant systems' response to the UFSAR anticipated operational occurrences and postulated accidents. Differences between the units (e.g., SGs) are also adequately represented. The staff finds that the approved ANP-10339P-A, Rev. 0, ARITA methodology is able to appropriately model non-LOCA transients at Braidwood and Byron.

The licensee LAR submittal states, Section 2 of Attachments 11 and 17 addresses the SE restrictions for each of the methodologies that are used to evaluate the non-LOCA transients.

Framatome also addresses departures from methodology and important clarifications (e.g., SG difference between units) for application of ARITA in Section 2. CEG concludes that the Braidwood and Byron non-LOCA transient demonstration analyses address all applicable SE limitations, adequately addresses mixed core conditions, and provides results representative of the Braidwood and Byron unit differences.

3.7.2 Non-LOCA Transients/Accidents Results The NRC staff conducted a thorough review of Methodology in Section 2 of ANP-4087P. The staff determined that the analysis of ANP-10339 (ARITA TR) detailed appropriate consideration and responses to all the Limitations and Conditions provided in the methodology approval in Section 2.1.1. Detailed discussion regarding the NRC staffs review is provided in Section 3.7.3, below. The report also properly analyzed Methodology Departures and Clarifications as well as Sensitivity Studies and Topical Updates. The NRC staff also noted that the methodology also ensured compatibility with supporting/interfacing analyses limitations and conditions and methodology clarifications as necessary for ANP-10297 (ARCADIA TR), ANP-10323 (GALILEO TR), ANP-10311 (COBRA-FLX TR), ANP-10341 (ORFEO-GAIA and ORFEO-NMGRID TR), BAW-10227 (M5 TR), and XN-75-32 (Rod Bow TR). The NRC staff

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION concluded that the ANP-4087P methodology is able to appropriately analyze non-LOCA transients and accidents.

The licensee LAR submittal states, Table 4-1 of Attachments 11 and 17 identifies UFSAR Chapter 15 events or sub-events that are not affected by the fuel transition and/or remain bounded by the existing UFSAR Analysis of Record (AOR). Long-term events driven by decay heat remain applicable since the decay heat associated with GAIA fuel is bounded by the decay heat used in these AORs. Several events are not applicable as they are relevant to boiling water reactors only. The remainder are driven by plant system responses or plant systems themselves and not dependent on fuel-related parameters or performance.

As part of the change to the ARITA methodology, Table 4-1 of Attachments 11 and 17 also captures that UFSAR 15.1.3 event, Excessive Increase in Secondary Steam Flow, now bounds the UFSAR 15.1.4 event, Inadvertent Opening of a Steam Generator Relief or Safety Valve. This is different from the current UFSAR 15.1.4 disposition under Westinghouse methodology, which is currently bounded by the UFSAR 15.1.5 event, Steam System Piping Failure at Zero Power, and UFSAR 15.1.6 event, Steam System Piping Failure at Full Power.

Additionally, CEG is aware of a typographical error in the UFSAR, which currently states in subsection 15.1.3 that steam flow increases greater than 10 percent are analyzed in subsections 15.1.4 and 15.1.5; this will be corrected during the transition implementation to point to subsections 15.1.5 and 15.1.6.

The NRC staff reviewed ANP-4087P, Table 4-1, Disposition of Events Summary. The staff found that the dispositions provided were appropriately considered and evaluated from a DNB analysis standpoint. The staff found that the dispositions provided were appropriately considered and evaluated from a fuel centerline melt analysis standpoint. The staff found that the dispositions provided were appropriately considered and evaluated from a Transient Cladding Strain Analysis standpoint. Other events were appropriately categorized as Bounded by the AOR or not applicable for the reasons specified in the LAR submittal. The NRC staff also reviewed the specific changes referenced to 15.1, Increase in Heat Removal by the Secondary System, events and found that the categorizations for those events were also correct and consistent with the information provided in the LAR submittal. The events were re-evaluated as necessary using the ARITA methodology.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.7.3 Compliance with NRC Staff Imposed Limitations and Conditions on Non-LOCA Transient/Accident Methodology Table 3.7.3 - Limitations and Conditions for ANP-10339P-A, Revision 0, ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology, October 2023 Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 1

For plants with a licensing basis that deviates from the SRP guidance current as of the issuance of this SE, the licensee shall assess the compatibility of the proposed ARITA methodology with unique or legacy aspects of the licensing basis in plant-specific implementation submittals. In cases where the plant-specific licensing basis deviates substantially and non-conservatively from current SRP guidance, the licensees implementation submittal shall propose modifications, as necessary, to the ARITA methodology or licensing basis to assure adequate conservatism. (Reference to Section 3.1.2 of Reference 1)

The ARITA analysis process includes a review of the existing plant licensing basis against the event descriptions included in the SRP as part of the assessment of applicability of the ARITA method for that event.

Examples of results of this review are provided in the demonstration analyses available for audit.

2 Plant-specific implementation submittals for the ARITA methodology ((

)) Alternatively, submittals may

((

)) (Reference to Section 3.1.3 of Reference 1)

((

))

3 For each analyzed event, plant-specific implementation submittals for the ARITA methodology must identify: (1) the FOMs considered, (2) the ((

)) discussed in Framatomes response to RAI 11, and (3) the method of determining ((

))

discussed in Framatomes response to RAI 11. (Sections 3.4.2.1 and 3.4.2.4)

The FOMs for each event analyzed as part of this submittal are provided.

Further, ((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 4

The ARITA methodology does not modify the scope of the postulated misloading events considered within the plants licensing basis. In addition, as applicable, licensees must justify credit for any equipment, surveillances, and associated acceptance criteria that are not included in TSs, but which are credited with the detection of misloading errors. (Section 3.4.2.5)

The ARITA methodology is not being applied to postulated misloading events as part of this submittal.

5 Proposed values for all ((

)) used in analyses performed with the ARITA methodology must be included in each plant-specific implementation submittal, including appropriate justification. The licensee must assure that all ((

)) used with the ARITA methodology are acceptable for each operating cycle. Once the ((

)) in the implementation submittal have been approved by the NRC staff, the licensee shall use those values in its analysis unless (1) the values need to be changed to maintain a conservative analysis or (2) the values are changed to reflect an updated plant design configuration with an appropriately conservative margin. (Sections 3.5.1.1 and 3.5.1.7)

Licensee-controlled parameters including ((

)) used in analyses will continue to be managed in accordance with existing procedures and 10 Code of Federal Regulations (CFR) 50.59. The

((

)) used in the ARITA analyses are available for audit.

6 In the absence of plant-specific data or other information ((

))

licensees implementing the ARITA methodology shall ((

)) (Section 3.5.1.2, PCS-7d and SEC-11c, and Section 3.5.1.3)

((

))

7 For events that consider overfill as a FOM, ((

))

licensees implementing the ARITA methodology shall either ((

)) (Section 3.5.1.2, PCS-20a)

No events considering pressurizer or steam generator overfill as a FOM are being analyzed with the ARITA methodology as part of this submittal.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 8

Absent justification for the insignificance of primary system and secondary system metal heat capacity on applicable FOMs, licensees implementing the ARITA methodology shall

((

)) irrespective of evaluation model variant, where the primary system significantly cools down relative to its initial condition, ((

)) (Section 3.5.1.2, PCS-27b)

((

))

9 In the absence of plant-specific data or other information conclusively demonstrating the ability to determine the reactor pressure vessel upper head fluid temperature to within ((

)) by the ARITA methodology, licensees adopting the ARITA methodology shall ((

)) (Section 3.5.1.2, PCS-28)

((

))

10 The parameter SEC-2b, ((

))

(Section 3.5.1.2, SEC-2b)

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 11 Licensees using the ARITA methodology must either (1) justify that the comparisons in ANP-10339P for the MB-2 MSLB testing or other relevant comparisons of the ARITA methodology against experimental data accurately or conservatively represent their plants with respect to the ((

)) or (2) otherwise demonstrate that the

((

)) is representative or conservative for each event where it is relevant to the prediction of applicable FOMs. (Section 3.5.1.2, SEC-6)

((

))

12 The parameter SEC-21a, ((

)) In the event that this prescribed approach prevents stable code execution, licensees shall describe and justify any alternative approach taken. (Section 3.5.1.2, SEC-21a)

An analysis was performed to

((

))

13 Licensees using the ARITA methodology shall provide justification for the applicability of parameter ((

)) (Section 3.5.1.2, SEC-21a)

An analysis has been performed to

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 14 Licensees using the ARITA methodology must assure that assumptions regarding operator actions remain conservative with respect to the FOMs calculated by the ARITA methodology.

(Section 3.5.1.2, HUM-1a)

Operator actions are not modeled for events terminated by an automatic reactor trip, unless justified. Operator actions that mitigate event progression are conservatively justified for events that initiate an automatic reactor trip but are not terminated by a trip.

15 For any event within the scope of the ARITA methodology that exhibits a prompt critical response: (1) the licensee shall perform any necessary analysis for the event using the coupled evaluation model variant and (2) justification must be provided at the time of application for the ((

))

(Section 3.5.1.2, GCN-2a and GCN-2b)

For any event within the scope of the ARITA methodology that exhibits a prompt critical response: (1) the coupled evaluation model variant will be used and (2) justification will be established for the ((

))

16 The parameter ((

))

(Section 3.5.1.2, GCN-3a)

The parameter ((

))

17 Unless otherwise justified, the parameter

((

))

consistent with the AREA methodology presented in ANP-10338P-A. (Section 3.5.1.2, LCN-15a3)

The parameter ((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 18 This SE does not approve the uncertainty equation specified in Section 9.1.3.2 of ANP-10339P for determining the ((

)) licensees may (1) provide adequate justification for the equation Framatome proposed in Section 9.1.3.2 of ANP-10339P, (2) conservatively estimate the magnitude of the

((

)) according to the equation FZ lFZl FQ FQ

or (3) justify other approaches for determining the

((

)) in their implementation submittals. (Section 3.5.1.2, LCN-15b1)

Justification is provided that the uncertainty equation specified in Section 9.1.3.2 of ANP-10339P for determining the ((

)) can be e of the

((

)) for the present application.

19 This SE does not approve the proposed uncertainty approach for parameter

((

licensees may (1) conservatively apply ((

)) to this parameter, or (2) justify alternative approaches (e.g., via citation of additional representative data) for determining ((

))

in their implementation submittals.

(Section 3.5.1.2, LCN-15b2)

The parameter ((

))

20

((

)) may not be representative of all contemporary or future fuel designs. ((

)) prior to the use of these fuel designs with the ARITA methodology or revise

((

)) to maintain a conservative analysis in accordance with Limitation and Condition Error!

Reference source not found.. (Section 3.5.1.2, TH-5, and TH-8)

The ((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 21 Unless otherwise justified, the ((

)) (Section 3.5.1.2, TH-9)

The ((

))

22 For uncertainty parameters ((

)) (Section 3.5.1.2, FRR-12 and Section 3.5.1.4)

For uncertainty parameters ((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 23 Unless otherwise justified, PORVs and PSVs will be ((

)) (Section 3.5.1.2,

((

))

Power-Operated Relief Valves (PORVs) and Pressurizer Safety Valves (PSVs) ((

))

24 Unless otherwise justified, licensees applying the ARITA methodology shall ((

)) in accordance with the treatment prescribed per GCN-7a. (Section 3.4.2.2)

((

))

25 In the absence of plant-specific data or other information ((

)) when addressing parameter GCN-7a, licensees shall

((

)) (Section 3.5.1.3)

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation on ANP-10339P-A, Revision 0 Summary of Licensee Response 26 The ((

)) proposed for the ARITA methodology have been reviewed and approved ((

)) The use of engineering analyses, on a case-by-case basis, ((

)) is not precluded by this limitation and condition.

(Section 3.5.1.6)

L&C 26 does not apply to Braidwood and Byron because Braidwood and Byron are ((

))

27 Licensees using the ARITA methodology

((

)) Licensees shall consider ((

)) on a cycle-specific basis per Limitation and Condition 5, above. (Section 4.3)

Braidwood and Byron will continue to collect measured data per existing plant programs/procedures. Where those measured values are relevant to the uncertainty and bias values used in the ARITA methodology, these will be assessed over time as an integral aspect of the reload engineering evaluation, which includes a review of plant parameters used in safety analysis against values applicable to the upcoming reload.

This scope is generally performed on a reload basis and addressed in the reload 50.59 process. Representative documents showing the scope of the reload plant parameter review will be available for audit during the NRC review period.

28 For each analyzed event, the applicability range of the ARITA methodology shall be limited to the range over which its constituent computational codes and models have been assessed and validated to provide acceptable predictions of relevant phenomena and processes. (Section 5.1)

All events evaluated with the ARITA method will include a review of the range of applicability of constituent codes and models vs. transient conditions. ((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff reviewed the licensee response to the 28 limitations and conditions above in Table 3.7.3 and finds that the licensee has satisfactorily met each limitation and condition for use of TR ANP-10339P-A, Revision 0. However, some of the limitations and conditions required confirmation of information or responses to RAIs before the staff could conclude the limitations and conditions were satisfactorily met. Several other limitations and conditions required additional considerations and examination of audit documentation for the staff to conclude the licensee satisfactorily met the limitations and condition. The discussion of these limitations and conditions and the licensee response is provided in the following subsection.

3.7.3.1 Discussion of Compliance with Limitations and Conditions for ANP-10339P-A, Revision 0 Limitation and Condition 2 Limitation and Condition 2 requires plant-specific submittals for the ARITA methodology

((

)) Section 2.1.1 of ANP-4087P, submitted as part of the licensees license amendment request, states ((

The NRC SRP 4.2 identifies a TCS damage threshold and a TCS failure threshold. In part, the damage threshold is to ensure that cladding is not so deformed during anticipated operational occurrences that the fuel rod cant be returned to service. The failure threshold is applicable to postulated accidents, wherein TCS is employed as a surrogate figure of merit for the PCMI failure mechanism. While SRP 4.2 specifies a 1 percent TCS criterion for both thresholds, applicants are free to propose an alternative with appropriate justification. In the present case, the licensee indicates that ((

))

As discussed in NUREG/KM-0019 (which documents the underlying technical basis for RG 1.236) and the NRC staffs SE for PWROG-21001-P-A, PCMI is cladding brittle fracture due to hydrogen embrittlement under reactivity insertion transient conditions. These events are characterized by short duration high mechanical loading. Rod ejection accidents, with their high enthalpy short-width pulses, are the most limiting reactivity insertion event; the high mechanical loading is a result, in part, of the rapid thermal expansion of the pellet and its subsequent pressing up against the cladding.

Failure threshold curves as a function of peak radial average fuel enthalpy rise versus excess cladding hydrogen content are defined in RG 1.236. These curves place limits on peak radial average fuel enthalpy rise for RIAs. Generally, for excess cladding hydrogen < 100 wppm, the peak radial average fuel enthalpy rise is limited to 150 cal/g for both PWR and BWR SRA and RXA cladding.

As mentioned above, the licensees license amendment request indicates ((

)) Several salient

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION cladding performance metrics, ((

)) were provided in support of this statement. In particular, ((

))

The NRC staff verified these points by auditing documentation demonstrating the performance of ((

)) the NRC staff noted ((

))

Based on the audited data, ((

)) Therefore, the NRC staff finds there is no need to specify a TCS limit ((

)) for SRP Chapter 15 non-LOCA postulated accidents because ((

)) As a result, the NRC staff finds the licensee meets the intent of Limitation and Condition 2 for the ARITA methodology.

Limitation and Condition 5 Limitation and Condition 5 requires that proposed values for all ((

)) used in analyses performed with the ARITA methodology must be included in each plant-specific implementation submittal, including appropriate justification. The licensee did not include a complete list of ((

)) in the submittal, but the licensee did have documentation readily available for audit that contained values for all

((

)) and the associated justifications. The NRC staff found this to be an acceptable approach because the limitation and condition serves to ensure NRC review of

((

)) is performed and to establish an acceptable baseline of

((

)) that can be referenced should values be updated in the future ba

, manifestation of non-conservatisms (e.g., drift in ((

))) or as a necessity to continue representing plant configuration with appropriate conservatism. To this end, the staff notes that the audited documentation that establishes the acceptable baseline of ((

)) is FS1-0064966, Revision 6.0, Byron and Braidwood Units 1 and 2 Analytical Input Summary of Generic Input Parameters for ARITA, dated October 9, 2024.

Appendix IV of the NRC staffs SE for ANP-10339P-A identifies a population of uncertainty parameters that ((

)) before they can be used in analyses performed with the ARITA methodology. In most instances, these ((

)) In a select few instances,

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

((

))

In auditing the documentation discussed above, NRC staff examined the source and derivation of the proposed values for ((

)) to ensure they are appropriately determined from representative plant data. There are ((

)) identified in the NRC staffs SE for ANP-10339P that require ((

)) license amendment implementing the ARITA methodology. Of these, the licensee indicated under Limitation and Condition 27 Method of Adherence in Section 2.1.1 of ANP-4087P, Revision 2 that the treatment of ((

)) will utilize an approach discussed in the response to RAI-9 of ANP-10339P-A. This treatment is further discussed in the response to RAI-11 of ANP-10339P-A. Succinctly, this treatment involves the use of ((

))

As part of the NRC staffs review of the present application, the treatment of ((

)) was examined to ensure it appropriately adhered to the approach described in the RAI response, and the staff concluded the approach is appropriately implemented. Because the ((

))

will be treated in a similar fashion, and because the NRC staff concluded the treatment approach itself, as discussed in the response to RAI-11, is being appropriately applied, the NRC staff finds the treatment for ((

)) is also acceptable.

Of the remaining ((

)) license amendment implementing the ARITA methodology, ((

)) The NRC staff examined the treatment of ((

)) and concludes the treatment is acceptable because either ((

)) is utilized in the treatment. Based on this and the discussions above, the NRC staff finds the licensee meets the intent of Limitation and Condition 5 for the ARITA methodology.

Limitation and Condition 6 Limitation and Condition 6 requires licensees implementing the ARITA methodology to ((

)) The licensee indicated in Section 2.1.1 of ANP-4087P, Revision 1 that ((

)) However, the NRC staff found that this response does not explicitly demonstrate adherence to the limitation and condition when analyses are performed with the ARITA methodology. Therefore, the NRC staff requested further clarification via RAI.

The licensees response indicated that ((

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

)) The results showed that ((

))

was not significant. In audit meetings that took place between February 24, 2025, through March 21, 2025, the NRC staff audited this analysis and verified the results. Because the licensees response discusses the approach taken and the results showed ((

)) the NRC staff finds the licensee meets the intent of Limitation and Condition 6 for the ARITA methodology.

Limitation and Condition 8 Limitation and Condition 8 requires licensees implementing the ARITA methodology ((

)) where the primary system significantly cools down relative to its initial condition, irrespective of the evaluation model variant. The licensee identified in Section 2.1.1 of ANP-4087P, Revision 1, the treatment for ((

)) but not all ARITA evaluation model variants were discussed. Therefore, the NRC staff requested further clarification via RAI regarding treatment for the evaluation model variants omitted from the discussion. The licensees response indicated the ((

)) The response further clarified that, ((

)) Because the response clearly discusses applicability of the omitted evaluation model variants and the associated treatment when using them, and the treatment is consistent with the staffs SE for ANP-10339P-A, the NRC staff finds that the licensee meets the intent of Limitation and Condition 8 for the ARITA methodology.

Limitation and Condition 9 Limitation and Condition 9 requires licensees implementing the ARITA methodology ((

)) Framatome indicated in Section 2.1.1 of ANP-4087P, Revision 1, that ((

)) However, the NRC staff found that this response does not explicitly demonstrate adherence to the limitation and condition when analyses are performed with the ARITA methodology. Therefore, the NRC staff requested further clarification via RAI.

The licensees response indicated that ((

)) In audit meetings that took place between February 24th, 2025,

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION through March 21st, 2025, the NRC staff audited this analysis and verified the results, noting

((

)) Because the response discusses the approach that was taken, that the

((

)) the NRC staff finds the licensee meets the intent of Limitation and Condition 9 for the ARITA methodology.

Limitation and Condition 11 Limitation and Condition 11 requires licensees adopting the ARITA methodology either 1) justify the comparisons in ANP-10339P for the Model Boiler-2 (MB-2) MSLB testing or other relevant comparisons of the ARITA methodology against experimental data accurately or conservatively represent their plants with respect to ((

)) or 2) otherwise demonstrate the ((

)) is representative or conservative for each event where it is relevant to the prediction of applicable FOMs. In Section 2.1.1 of ANP-4087P, Revision 1, Framatome discusses ((

)) This same comparison was provided in the ARITA topical report, ANP-10339P-A.

While the results provided in ANP-4087P and ANP-10399P-A demonstrate ((

)) which is conservative for a MSLB event, the NRC staff has concerns with regard to the scaling applicability of the test facility to full-size ((

] and the assurance that ((

)) will be conservatively predicted for future analyses for all relevant reactor designs. In particular, the scaling analysis report for the test facility indicates there are differences in the design of ((

)) Therefore, the NRC staff requested further justification via RAI regarding the scaling applicability of the ((

)) for the MSLB case.

Framatomes response indicated the ((

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

))

Framatomes response speaks to each of the underlying concerns associated with scaling applicability as identified in the RAI ((

)) indicating the testing results of the ((

)) for the 100 percent MSLB event. Therefore, the NRC staff finds the licensee meets the intent of Limitation and Condition 11 for the ARITA methodology.

Limitation and Condition 12 Limitation and Condition 12 requires parameter SEC-21a ((

)) The licensee indicated in Section 2.1.1 of ANP-4087P, Revision 1 that an analysis was performed to establish ((

)) for the treatment of parameter SEC-21a consistent with the requirements of the limitation and condition and this analysis is available for audit. In audit meetings conducted from February 24, 2025, through March 21, 2025, the NRC staff audited the applicable analysis and found the ((

)) which is consistent with the approach discussed within the staffs SE for ANP-10339P-A. Therefore, the NRC staff finds the licensee meets the intent of Limitation and Condition 12 for the ARITA methodology.

Limitation and Condition 13 Limitation and Condition 13 requires licensees implementing the ARITA methodology provide justification for the applicability of parameter ((

)) The licensee indicated in Section 2.1.1 of ANP-4087P, Revision 1 that an analysis was performed to establish

((

)) to be used in analysis as needed, and the analysis is available for audit. In audit meetings conducted from February 24, 2025, through March 21, 2025, Framatome indicated that ((

)) To further assess the adequacy of this approach, the NRC staff requested further information via RAI.

The licensees RAI response indicated ((

)) The NRC staff noted that the ((

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

)) Because of this, and because the justification is consistent with the approach presented and discussed for RAI-43 of ANP-10339P-A, the NRC staff finds the licensee meets the intent of Limitation and Condition 13 for the ARITA methodology.

Limitation and Condition 15 Limitation and Condition 15 requires licensees adopting the ARITA methodology perform any necessary analysis for events exhibiting a prompt critical response using the coupled evaluation model variant and provide justification for ((

)) The licensee indicated in Section 2.1.1 of ANP-4087P, Revision 1, that for any event that exhibits a prompt critical response, the coupled evaluation model variant will be used and justification provided for ((

)) However, the NRC staff found that this response does not explicitly demonstrate adherence the limitation and condition when analyses are performed with the ARITA methodology. Therefore, the NRC staff requested further clarification via RAI.

The licensees RAI response indicated ((

)) The NRC staff noted this is supported by ((

)) In audit meetings conducted from February 24, 2025 through March 21, 2025, NRC staff audited these analyses and verified the results, noting ((

)) Because justification was provided for ((

)) and because the coupled evaluation model variant will be used, the NRC staff finds the licensee meets the intent of Limitation and Condition 15 for the ARITA methodology.

Limitation and Condition 18

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation and Condition 18 requires licensees adopting the ARITA methodology conservatively estimate the magnitude of ((

)) according to the supplied equation or justify other approaches, which may include additional justification for the equation proposed in Section 9.1.3.2 of ANP-10339P-A (the ARITA formulation). In Section 2.2.1 of ANP-4087P, Revision 1, Framatome provided additional justification for the ARITA formulation. The justification provided focuses on demonstrating 1) the ((

)) when determined using the ARITA formulation and the formulation in Limitation and Condition 18, provided ((

)) and 2) the ARITA formulation ((

)) can be used to obtain a conservative estimate of ((

))

Table 2-2 of ANP-4087P, Revision 1, provides a comparison of the results from the different formulations and includes ((

)) as determined based on available measurement data. The results show the ARITA formulation and the Limitation and Condition 18 formulation ((

)) Additionally, Table 2-2 shows the ARITA formulation conservatively estimates ((

))

The NRC staff notes that the use of measured and predicted peaking factor data is instrumental to this approach. The measured and predicted peaking factor data used comes from topical report ANP-10297, Revision 0, Supplement 1P-A, Revision 1, which documents the NRC-approved ARCADIA core design and analysis methodology. The measured peaking factor data in this topical report ((

)) The use of measured peaking factor data suggests the ((

)) and thus appropriate for use as a common reference for comparison to the different formulations.

In audit meetings conducted from February 18, 2025, through March 21, 2025, NRC staff audited available analysis documents to verify the ((

)) The NRC staff noted ((

)) Based on this verification, the NRC staff finds there is reasonable assurance the ((

)) is representative of ((

)) and that, with respect to this, the ARITA formulation results in a conservative estimate ((

)) for the present application. Therefore, the NRC staff finds the licensee meets the intent of Limitation and Condition 18 for the ARITA methodology.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Limitation and Condition 19 Limitation and Condition 19 requires licensees adopting the ARITA methodology conservatively apply ((

)) or justify an alternative approach. In Section 2.1.1 of ANP-4087P, Revision 1, the licensee indicated ((

)) The NRC staff finds this is reasonable; if ((

)) then it follows ((

))

With regard to ((

)) Framatome indicated in an audit meeting on February 18, 2025, that, ((

))

Because this involves a separate calculation, Framatome made the calculation available for audit. NRC staff audited the calculation and found it is consistent with the approach discussed in ANP-10338P-A. Based on the justification provided and the discussion above, the NRC staff finds the licensee meets the intent of Limitation and Condition 19 for the ARITA methodology.

3.7.4 Methodology Departures Section 2.1.2 of ANP-4087, Revision 1, identifies eight departures from the NRC-approved ARITA methodology for the Braidwood and Byron application (the NRC-approved ARITA methodology is documented in ANP-10339P-A and the associated NRC safety evaluation). The NRC staffs review of each of these departures is discussed below.

Changes to the Power Distribution Control (PDC-A) Methodology Section 8.0 of ANP-4087P, Revision 1, discusses implementation of the PDC-A method with two departures from the approved approach discussed in Section 15 of ANP-10339P-A. The first departure implements a ((

)) to support the current plant Technical Specifications and operating procedures. The implementation is achieved by

((

)) The result is an analysis that bounds measured or actual plant operation. Because the analysis is bounding and consistent with the current plant Technical Specification, the NRC staff finds the methodology departure acceptable.

The second departure implements a change to the PDC-A method in response to Braidwood and Byron TS SRs. At present, the PDC methodology requires either an increase in of 2 percent or an increase in surveillance of if two successive measurements indicate an

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION increase in peak pin power,, with exposure. A similar requirement exists in the Braidwood and Byron TSs, however is monitored directly instead of. To remain consistent with the current Braidwood and Byron TSs, the PDC-A method was updated for the present application to implement surveillance. The modification to the method ((

)) Because this remains consistent with the existing PDC methods and is also consistent with the Braidwood and Byron TS SRs, the NRC staff finds the methodology departure acceptable.

Sampling Range for ((

))

Framatome indicates in Section 2.1.2 of ANP-4087P, Revision 1, that the sampling range ((

)) The NRC staff finds this acceptable because the use of ((

))

Number of ((

))

((

)) This is conservative for such cases, and the NRC staff therefore finds the methodology departure acceptable.

Additional ((

))

Section 2.1.2 of ANP-4087, Revision 1, indicates an additional ((

)) Because this will result in the application of either the current approved ((

)) or a more conservative one, the NRC staff finds the methodology departure acceptable.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Adjustment to The Sampling Range for ((

))

Section 2.1.2 of ANP-4087, Revision 1, indicates the sampling range for ((

))

The NRC staff finds this methodology departure acceptable because ((

)) is generally conservative for analyses and because ((

))

Inclusion of ((

))

Section 2.1.2 of ANP-4087, Revision 1, indicates ((

)) The NRC staff finds this methodology departure acceptable because ((

)) is just as conservative or more conservative than what is prescribed in the approved ARITA methodology.

Use of ((

)) Evaluation Model Variant ((

))

((

)) event analysis is performed using the ((

)) However, Section 2.1.2 of ANP-4087, Revision 1, indicates ((

)) The NRC staff notes that ((

)) Section 8.6.1 of ANP-10339P-A, Section D7 of ANP-10338P-A, and RAI response 1.D.1 of ANP-10338P-A demonstrate this; ((

)) The NRC staff therefore finds the methodology departure acceptable.

Adjusting ((

))

The S-RELAP5 plant input model in the ARITA methodology includes ((

)) In the NRC-approved input model, these ((

)) However, Section 2.1.2 of ANP-4087P, Revision 1, identifies that, ((

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

)) To address this, ((

)) The NRC staff notes discussion of these results were included in the response to RAI-27 of ANP-10339P-A. The NRC staff also audited associated analyses to verify the veracity ((

)) and conclude they are reasonable. Based on this, the NRC staff finds the methodology departure acceptable.

3.7.5 Non-LOCA Transients/Accidents Conclusions The NRC staff reviewed the information in the licensees submittal pertaining to the analysis of non-LOCA transients and accidents events with Framatome GAIA fuel to support plant operation for Braidwood and Byron. The NRC staffs review verified that the non-LOCA transients and accidents analyses appropriately applied the approved methodology in TR ANP-10339P-A, Revision 0, and the licensee has met all the applicable limitations and conditions. The NRC staff also reviewed the departures from the methodology approved in ANP-10339P-A, Revision 0, and concluded the departures are acceptably justified.

3.8 Co-Resident Fuel Considerations For the first two transition cycles, Braidwood and Byron will contain once-and twice-burned VANTAGE+ fuel from Westinghouse. This requires a technical consideration of the thermal-hydraulic and neutronic effects of these two co-resident fuel types in the core.

Co-resident fuel considerations are broken down into the following five areas.

3.8.1 Assembly Flow Penalties for Mixed Cores The two fuel types have different pressure drops at different elevations, which can affect assembly flow and the thermal hydraulics of the core. Westinghouse and Framatome are separately developing penalties for their fuel type for use in a mixed core.

Westinghouse has evaluated the hydraulic effects of co-resident GAIA fuel on VANTAGE+ fuel and assigned conservative FH peaking factor limit reduction. This evaluation involved obtaining and modeling GAIA fuel in the VIPRE-W code in representative mixed core configurations (representing the first and second transition cycle) alongside VANTAGE+ fuel. The DNBR results from these mixed core configurations were then ((

)) The DNBR effects were then converted into FH limit reductions and confirmed to be sufficient to offset the mixed core DNBR penalty for VANTAGE+ fuel.

As discussed in the April 21, 2025, supplement, the first transition cycle is set to have a minimum of 100 VANTAGE+ assemblies while the second transition cycle is set to have a minimum of 8 VANTAGE+ assemblies. This allows up to 93 GAIA assemblies in the first cycle and 185 GAIA assemblies in the second transition cycle. These assembly minimums were

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION intentionally chosen to not challenge the transition cycle designs. In its August 6, 2025, supplement, the licensee stated that recent calculation refinements have identified a revision to the second transition cycle FDH penalty previously provided and the corresponding COLR limits for both transition cycles were incorrect. For the first transition cycle, a ((

)) is to be applied to Westinghouse fuel, which will be documented in the COLR.

For the second transition cycle, the ((

)) The penalties correspond to Westinghouse ((

)) in the first transition cycles and ((

)) in the second transition cycle. The NRC staff reviewed the Westinghouse evaluation during the regulatory audit and finds that the penalties were appropriately calculated.

Framatome will develop cycle-specific penalties for the GAIA fuel due to the co-resident thermal-hydraulic considerations of the VANTAGE+ fuel. These penalties are developed following the methodology in Reference 7.16 (Section 12, RAI-77, and L&Cs 20 and 21).

Table 3.8.1 displays these two limitations and conditions and the licensee evaluation.

Table 3.8.1 - ANP-10339P-A Limitations and Conditions ANP-10339P-A L&C Licensee Evaluation

((

)) may not be representative of all contemporary or future fuel designs. ((

))

prior to the use of these fuel designs with the ARITA methodology or revise ((

)) to maintain a conservative analysis in accordance with Limitation and Condition 5.

(Section 3.5.1.2, TH-5, and TH-8)

The ((

))

Unless otherwise justified, the ((

))

((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff reviewed the licensee responses to the two limitations and conditions above.

Given that Framatome is considering the effects of the VANTAGE+ fuel to evaluate crossflow resistance and turbulent energy exchange and indicated that this limitation and condition will be adhered to when developing cycle-specific mixed core uncertainties, the NRC staff finds these two limitations and conditions met.

Additionally, Reference 7.16 applies the TS minimum flow rate in DNBR calculations, which is below the usual system flow rate, applying additional conservatism to DNBR calculations. This same assumption is used in the Westinghouse calculation of FH limit reduction for VANTAGE+

fuel.

The NRC staff finds that the above evaluations related to VANTAGE+ and GAIA fuel DNBR penalties adequately address assembly flow penalties for mixed cores for the Braidwood and Byron fuel transition.

3.8.2 Nuclear Design for Mixed Cores The ARCADIA methods (Reference 7.9) are used for neutronic design of the core, including GAIA and VANTAGE+ fuel. ARCADIA methods were validated for six different PWR types and six different fuel types, which included Westinghouse plants with 17x17 fuel. The NRC staff finds that the Framatome ARCADIA methods are acceptable for the neutronic design of the core, including GAIA and VANTAGE+ fuel types with fuel specific limits specified in the COLR as required.

3.8.3 DNB Affected Analysis for Mixed Cores For mixed cores, both Westinghouse and Framatome methods will be used. Each vendor will use their methods for analyzing the fuel performance of their fuel (i.e., VANTAGE+ fuel performance will be analyzed with Westinghouse methods, and GAIA fuel performance will be analyzed with Framatome methods). For this reason, the two vendor fuel types and methodologies will be looked at separately in this section. The ARCADIA methods (Reference 7.9) are used to develop the power, burnup, and power distributions for fuel performance evaluation.

Westinghouse - VANTAGE+ Fuel Under current VANTAGE+ core designs, Westinghouse does not re-run cycle-specific evaluations. Events are run with a representative core with limiting reactivity and power distribution characteristics to create an AOR, and the cycle-specific reactivity and power distribution characteristics are compared to the limiting characteristics from the AOR. For mixed cores, the process will generally be the same. ARTEMIS (Reference 7.16) will be used to determine the burnup and power distribution of the VANTAGE+ fuel with no additional uncertainties required due to use of the ARTEMIS methods. Additionally, the flow reduction penalty described in Section 3.8.1 of this SE will be applied to the VANTAGE+ fuel for these evaluations. Westinghouse AORs will predict limiting DNB rod census for the full core for each event, and if the AOR predicts fuel failure by DNB, the VANTAGE+ fuel failure census will be ratioed to the fraction of VANTAGE+ in the mixed core. Westinghouse AORs predict no FCM for all events other than control rod ejection (see Section 3.6 of this SE). The VANTAGE+ FCM confirmation during fuel transition will be evaluated as part of the transition reload safety analysis, and FCM rod census will be determined as described below in Section 3.8.4.

Framatome - GAIA Fuel

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Safety analyses for GAIA fuel will be identical between a full core of GAIA fuel and a mixed core configuration. The notable difference is mixed cores will follow the reduced assembly flow considerations describes in Section 3.8.1 of this SE.

The NRC staff concludes that the methodologies described above are adequate for addressing DNB affected analyses for mixed cores.

3.8.4 Other VANTAGE+ SAFDLs SAFDLs for GAIA Fuel are covered in Section 3.1 of this SE. SAFDLs related to VANTAGE+

are currently confirmed with the existing methodologies listed in TS 5.6.5.b. During transition cycles, Westinghouse will continue to confirm the SAFDLs related to VANTAGE+ fuel using these existing methodologies. As stated above in Section 3.8.2, ARCADIA will be used to develop inputs for the necessary for any SAFDL evaluations, including FCM rod census, rod internal pressure, cladding conditions, etc.

The NRC staff finds that Westinghouse using existing methodologies to confirm SAFDLs for VANTAGE+ fuel under the described process is acceptable.

3.8.5 Boundary/initial conditions As stated in Section 3.8 of the LAR, the design and safety analysis boundary conditions (plant initial conditions, power distribution limits, Reactor Protection Setpoints, etc.) for the current Westinghouse AORs will also apply to the mixed cores. The boundary conditions were discussed during the regulatory audit and exceptions to using the same boundary conditions are discussed in the April 21, 2025, supplement. There are five instances where the boundary/initial conditions are different between the existing Westinghouse AORs and the proposed mixed cores.

Setpoint for MSLB Pressure The MSLB outside containment (UFSAR 15.1.5 and 15.1.6) evaluation performed by Framatome uses a higher Safety Analysis Limit (SAL) (435.3 vs. 488.3 psig), while maintaining the existing TS value. This change was made to remove an unneeded conservatism. In the Westinghouse evaluation, radiological environmental effects from fuel damage was considered even though fuel damage was not predicted. However, since radiological environmental effects were not included in the Framatome evaluations, the conservatism is not needed.

Power Distribution Limit for FNDH The licensee stated that the Framatome fuel will be designed and verified acceptable up to the current COLR limit for FNDH of 1.7. For mixed cores, Westinghouse fuel will require a penalty to the limit to be used to compensate for the hydraulic mismatch with the GAIA fuel with respect to DNB. The Westinghouse penalty is discussed above in Section 3.8.1 of this SE.

Control System Response, Rod Control Speed The licensee stated that the Westinghouse and Framatome evaluations include the effects of the rod control system operating normally when that operation provides more limiting results than with control rods in manual. The Westinghouse evaluations consider rod insertion and

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION withdraw speeds up to 72 steps per minute. Based on an upcoming plant modification to limit the rod control system speed controls system with the introduction of GAIA fuel, the Framatome evaluations consider rod withdrawal speeds up to eight steps per minute and up to 72 steps per minute insertion.

Boron Concentrations in ECCS Subsystems The licensee stated that the boron concentrations in ECCS subsystems will be higher than the current Westinghouse limits. The change in boron concentrations is discussed above in Sections 3.3.3 and 3.3.4 of this SE.

Moderator Temperature Coefficient (MTC)

The licensee stated that the least negative MTC limits in TS 3.1.3 and the most negative MTC in the Westinghouse Analyses of Record are unchanged. The Framatome analyses will use the same MTC limits as Westinghouse had used for design basis events. However, the Westinghouse evaluations apply an upper limit to the MTC per current TS 5.6.5.b.5. This upper limit only applies to anticipated transients without scram events, and that limit was removed by another license amendment (ML25030A158). Both Westinghouse and Framatome methods evaluate the full range of MTC as specified in the COLR and TS 3.1.3 for all other events.

Based on its review of the above information, the NRC staff finds that the changes to boundary/initial conditions from the original Westinghouse evaluations are appropriate and, therefore, acceptable.

3.9 Additional Considerations The licensee stated several additional considerations in Section 3.9 of the application which are not submitted for review or approval but will be implemented in accordance with the requirements of 10 CFR 50.59. The NRC staff reviewed these considerations for information and applicability to other sections of the application but is not approving or denying any of the listed items in this section. The NRC inspection and enforcement programs continue to review site changes subject to 10 CFR 50.59 implementation.

The items included for additional consideration include: lower plenum flow anomaly (Byron Unit 1 only), misloaded fuel assembly, and environmental qualification.

3.10 Technical Conclusion The NRC staff has reviewed the LAR and its supplements to evaluate the acceptability of the Braidwood and Byron transition to Framatome GAIA fuel design with Framatome safety analyses and core design methodologies. Based on its review, the NRC staff has determined that the licensee provided an adequate technical basis to support the proposed LAR.

Specifically, the NRC staff has determined that the licensee has demonstrated that (1) the fuel assembly and fuel rod analyses satisfy the NRC-approved design criteria for normal and faulted conditions, (2) the licensee complies with the NRC staffs limitations and conditions imposed for application of TRs used in the execution of the LAR, (3) Framatome codes and methods are applicable for Braidwood and Byron, (4) the safety analysis results submitted to the NRC staff demonstrate compliance with applicable regulatory requirements, and (5) the proposed TS changes and license conditions are acceptable and satisfy the 10 CFR 50.36 requirements.

Therefore, the NRC staff finds there is reasonable assurance of public health and safety.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on May 22, 2025. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in Title 10 of the Code of Federal Regulations (10 CFR), Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, which was published in the Federal Register on November 5, 2024 (89 FR 87898), and there has been no public comment on such finding. Accordingly, the amendment meets 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 amendment.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

7.0 REFERENCES

7.1 ANP-10342P-A, GAIA Fuel Assembly Mechanical Design, Revision 0, dated September 2019.

7.2 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section Ill, Nuclear Power Plant Components, 1992 Edition 7.3 BAW-10240P-A, Incorporation of MS' Properties in Framatome ANP Approved Methods, Revision 0, dated May 2004 7.4 ANP-10334P-A, Q12 ' Structural Material, Revision 0, dated September 2017 7.5 ANP-10337P-A, PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations, Revision 0, dated April 2018.

7.6 ANP-10337P-A, Supplement 1P-A, Revision 0, Deformable Spacer Grid Element, Revision 0, dated September 2020.

7.7 BAW-10227P-A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, Revision 2, dated January 2023.

7.8 ANP-4086P, Byron and Braidwood Rod Ejection Accident Analysis, Revision 0, dated March 2024.

7.9 ANP-10297P-A, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, Revision 0, dated February 2013.

7.10 ANP-10297, Supplement 1P-A, Revision 1, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, Revision 0, dated December 2020.

7.11 ANP-10338P-A, AREA - ARCADIA Rod Ejection Accident, Revision 0, dated December 2017.

7.12 ANP-10323P-A, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, Revision 1, dated November 2020.

7.13 ANP-10323P, Fuel Rod Thermal-Mechanical Methodology for Boiling-Water Reactors and Pressurized Water Reactors, Revision 0, dated July 2013.

7.14 BAW-10084P-A, Program To Determine In-Reactor Performance of BWFC Fuel Cladding Creep Collapse, Revision 3, dated July 1995.

7.15 XN-75-32P-A, Supplements 1, 2, 3, & 4, Computational Procedure for Evaluating Fuel Rod Bowing, dated October 1983.

7.16 ANP-10339P-A, ARITA-ARTEMIS/RELAP Integrated Transient Analysis Methodology, Revision 0, dated October 2023.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 7.17 BAW-10183P-A, Fuel Rod Gas Pressure Criterion (FRGPC), Revision 0, dated July 1995.

7.18 Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss-of-Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits and Exemption Request for use of M5TM Cladding, dated May 2, 2023.

7.19 Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Proposed Amendment to Support Implementation of Framatome GAIA Fuel, dated October 30, 2023.

8.0 ACRONYMS/ABBREVIATIONS Acronym Definition ADAMS Agencywide Documents Access and Management System AFD Axial Flux Difference AC Alternating Current AFP Alternative Flow Path AFW Auxiliary Feedwater ANS American Nuclear Society ANSI American National Standards Institute AO Axial Offset AOO Anticipated Operational Occurrence AOR Analysis of Record AREA ARCADIA Rod Ejection Accident ARERR Annual radioactive effluent release reports ASME American Society of Mechanical Engineers AST Alternate source term BOC Beginning-of-Cycle BOL Beginning of Life BDMS Boron Dilution Mitigation System BWR Boiling-Water Reactor CAP Containment Accident Pressure CFR Code of Federal Regulations CHF Critical Heat Flux CLB Current Licensing Basis COLR Core Operating Limit Report CR Control Room CSR Cable Spreading Room CUF Cumulative usage factor CVCS Chemical and Volume Control System CWO Core Wide Oxidation DBA Design Basis Accident DF Decontamination Factor DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio DTC Doppler Temperature Coefficient

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Acronym Definition EAB Exclusion Area Boundary ECCS Emergency Core Cooling System ECR Equivalent Cladding Reacted EFPD Effective Full Power Days EM Evaluation Model EOC End-of-Cycle EOL End-of-Life ESFAS Engineered Safety Feature Actuation System oF Degrees Fahrenheit FH Nuclear Enthalpy Rise Factor/Radial Peaking Factor FQ Total Peaking Factor/Global Peaking Factor FCM Fuel Centerline Melt FPC Fuel Performance Code FR Federal Register FRGPC Fuel Rod Gas Pressure Criterion FSAR SP Final Safety Analysis Report (Standard Plant)

FSRR Fuel Swelling, Rupture, and Relocation GDC General Design Criteria GL Generic Letter gpm gallons per minute GSI Generic Safety Issue GWd Gigawatt days GT Guide Tubes HFP Hot Full Power HHSI High Head Safety Injection HMP High Mechanical Performance HZP Hot Zero Power IFBA Integral Fuel Burnable Absorber IFM Intermediate Flow Mixer IFWF Increase in Feedwater Flow IGM Intermediate GAIA Mixing IHSI Intermediate Head Safety Injection IN Information Notice ISG Intermediate Spacer Grid JFD Joint Frequency Distributions k(z)

Axial-Dependent Peaking Factor L&C Limitation and Condition LAR License Amendment Request LBLOCA Large Break Loss-of-Coolant Accident LCO Limiting Condition of Operation LEU Low-Enriched Uranium LHGR Linear Heat Generation Rate LHSI Low Head Safety Injection LOCA Loss-of-Coolant Accident LOOP Loss-of-Offsite Power LPZ Low Population Zone LWR Light-Water Reactor

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Acronym Definition M&E Mass and Energy MDNBR Minimum Departure from Nuclear Boiling Ratio MFIV Main Feedwater Isolation Valve MLO Maximum Local Oxidation MOC Middle of Cycle MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MTC Moderator Temperature Coefficient MTU Metric Ton Uranium MWt Megawatt thermal NPSH Net Positive Suction Head NQA Nuclear Quality Assurance NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake OTT Overtemperature Delta Temperature pcm percent mille (one-thousandth of a percent)

PC Primary Containment PCMI Pellet Clad Mechanical Interaction PCT Peak Cladding Temperature PDC Power Distribution Control PDMS Power Distribution Monitoring System PLC Pressure Loss Coefficient PLHGR Peak Linear Heat Generation Rate ppm parts per million psi pounds per square inch PSV Pressurizer Safety Valve PV Pressure-Velocity PWR Pressurized Water Reactor RAI Request for Additional Information RBHT Rod Bundle Heat Transfer RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System REA Rod Ejection Analysis RG Regulatory Guide RHR Residual Heat Removal RPS Reactor Protection System RTP Rated Thermal Power RTS Reactor Trip System RWST Refueling Water Storage Tank SAFDL Specified Acceptable Fuel Design Limit SBLOCA Small Break Loss-of-Coolant Accident SE Safety Evaluation SG Steam Generator SI Safety Injection SL Safety Limit SPC Siemens Power Corporation

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Acronym Definition SR Surveillance Requirement SRM Swelling and Rupture Model SRP Standard Review Plan SRSS Square Root of Sum of Squares SSE Safe Shutdown Earthquake TCS Transient Cladding Strain TEDE Total Effective Dose Equivalent T-H Thermal-Hydraulic TIL Time in Life TR Topical Report TS Technical Specification TSC Technical Support Center UET Unfavorable Exposure Time UFSAR Updated Final Safety Analysis Report UO2 Uranium Dioxide UTL Upper Tolerance Limit VQP Vendor Qualification Program Principal Contributors: D. Nold, NRR H. Wagage, NRR J. Vande Polder, NRR J. Ambrosini, NRR J. Lehning, NRR K. Heller, NRR M. Hamm, NRR M. Mazaika, NRR R. Fu, NRR R. Beaton. NRR S. Meighan, NRR M. Mazaika, NRR Date: August 21, 2025

v t e Letter Sequence Approval
EPID:L-2024-LLA-0072, (Open)
Initiation Acceptance Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Questions 26 March 2025 10 April 2025 Answers Results Approval, Approval Other: ML24299A255, ML25125A263
MONTHYEAR2025-04-24 [Table View]

Person / Time
ML25153A008
Site:	Byron, Braidwood (NPF-037, NPF-066, NPF-072, NPF-077)
Issue date:	08/21/2025
From:	Scott Wall Plant Licensing Branch III
To:	Rhoades D Constellation Energy Generation
Wall, S
References
EPID L-2024-LLA-0072
Download: ML25153A008 (1)
v • d • e

ML25153A008

Text

SUBJECT:

Dear Mr. Rhoades:

Enclosures:

Attachment:

Attachment:

Attachment:

Attachment:

1.0 PROPOSED CHANGE

2.0 REGULATORY EVALUATION

2.1 Proposed TS Changes

3.0 TECHNICAL EVALUATION

3.3 Proposed TS Changes

3.3 Proposed TS Changes

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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