Text

Response to Request for Additional Information Related to License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46
	ML25111A257
Person / Time
Site:	Braidwood, Byron
Issue date:	04/21/2025
From:	Steinman R; Constellation Energy Generation
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML25111A256	List: RS-25-088, Response to Request for Additional Information Related to License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46;
References
	RS-25-088
	Download: ML25111A257 (1)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office Proprietary Information - Withhold from Public Disclosure Under 10 CFR 2.390 Attachments 7 - 11 contain Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachments 7 - 11 this document is decontrolled.

April 21, 2025 10 CFR 50.90 RS-25-088 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Response to Request for Additional Information Related to License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, to Transition to Framatome GAIA Fuel and Exemptions to 10 CFR 50.46

Reference:

1. Letter from R. Steinman (Constellation Energy Generation, LLC) to U.S.

NRC, "License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, to transition to Framatome GAIA fuel and exemptions to 10 CFR 50.46 and 10 CFR 50 Appendix K", dated May 28, 2024 (ADAMS Accession No. ML24149A126)

2. Email from S. Wall (U.S. NRC) to L. Zurawski (Constellation Energy Generation, LLC), "RAI - Braidwood and Byron - License Amendment Request to Transition to Framatome GAIA Fuel (EPID No.

L-2024-LLA-0072)" dated March 17, 2025 (ADAMS Accession No. ML25085A297)

In Reference 1, Constellation Energy Generation, LLC (CEG) requested an amendment to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood) and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron). The proposed amendment is to change:

x TS 2.1.1, "Reactor Core SLs" to add limits for the GAIA fuel type.

x TS 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))" to replace the Westinghouse FQ W(Z) with the Framatome FQ V(Z). Additionally, ACTIONS B is revised to remove REQUIRED ACTIONS B.1, B.3, and B.4.

x TS 3.5.1, "Accumulators" to raise SR 3.5.1.4 and 3.5.1.5 ECCS boron limits.

U.S. Nuclear Regulatory Commission April 21, 2025 Page 2 x

TS 3.5.4, "Refueling Water Storage Tank (RWST)" to raise SR 3.5.4.4 ECCS boron limits.

x TS 4.2.1, "Fuel Assemblies" to add M5Framatome fuel rod cladding.

x TS 5.5.16, "Containment Leakage Rate Testing Program" to change Pa to 38.7 psig for Unit 2.

x TS 5.6.5, "Core Operating Limits Report (COLR)" to add Framatome analytical methods to TS 5.6.5.b. Additionally, add LCO 3.1.4, 3.3.1, and 3.3.9 to TS 5.6.5.a.

In Reference 2, the NRC requested additional information that is needed to complete review of the license amendment request. In response to this request, CEG is providing the attached information.

CEG has reviewed the information supporting the finding of No Significant Hazards Consideration, and the Environmental Consideration that were previously provided to the NRC.

The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

CEG is notifying the State of Illinois of this supplement to a previous application for a change to the operating license by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b).

There are no regulatory commitments included in this letter.

U.S. Nuclear Regulatory Commission April 21, 2025 Page 3 Should you have any questions concerning this letter, please contact Ms. Lisa Zurawski at 779-231-6196.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 21st day of April 2025.

Respectfully, Rebecca L Steinman Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:

1. Response to Request for Additional Information (Non-Proprietary)

2. 103-4115NP-000, Byron/Braidwood Unit 2 Small Break LOCA Analysis (Non-Proprietary)

3. 103-4116NP-000, Byron/Braidwood Unit 2 Large Break LOCA Analysis (Non-Proprietary)

4. Affidavit Constellation

5. Affidavit Framatome

6. Affidavit Westinghouse

7. Response to Request for Additional Information (Proprietary)

8. 103-4115P-000, Byron/Braidwood Unit 2 Small Break LOCA Analysis (Proprietary)

9. 103-4116P-000, Byron/Braidwood Unit 2 Large Break LOCA Analysis (Proprietary)

10. OP-AA-102-106 Operator Response Time Validation Sheet Attachment 1 for May 22, 2023 - TCA 25 (Proprietary)

11. OP-AA-102-106 Operator Response Time Validation Sheet Attachment 1 for May 21, 2024 - TCA 29 (Proprietary) cc:

Regional Administrator - NRC Region III NRC Senior Resident Inspector - Byron Nuclear Power Station NRC Senior Resident Inspector - Braidwood Nuclear Power Station Illinois Emergency Management Agency - Department of Nuclear Safety

ATTACHMENT 1 BRAIDWOOD STATION UNITS 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 Docket Nos. STN-50-456 and STN-50-457 BYRON STATION UNITS 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 Docket Nos. STN-50-454 and STN-50-455 Response to Request for Additional Information

ATTACHMENT 1 Page 1 of 9 By application dated May 28, 2024 (Agencywide Documents Access and Management System(ADAMS) Accession No.ML24149A125), Constellation Energy Generation, LLC (CEG; the licensee) submitted a license amendment request (LAR) for Braidwood Station, Units 1 and 2 (Braidwood), and Byron Station, Units 1 and 2 (Byron). The proposed amendments would revise technical specifications to allow the use of Framatome GAIA fuel at Braidwood and Byron.

The U.S. Nuclear Regulatory Commission (NRC) staff determined that the following information is needed to complete its review.

Nuclear Systems Performance Branch (SNSB) Questions SNSB-RAI-1 Regulatory Basis Appendix A to Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), General Design Criterion (GDC) 10, "Reactor Design," states: "The 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."

Issue Section 3.8.1 of the LAR states: "Westinghouse will evaluate the hydraulic effects of co-resident GAIA fuel on VANTAGE+ fuel and assign a conservative penalty to account for the potential VANTAGE+ assembly flow reduction. This Westinghouse evaluation considers, as input, the pressure drops of each component (nozzles, grids, fuel rods, etc.)

of both fuel types at each elevation. The result of this evaluation is a conservative peaking factor penalty on VANTAGE+ fuel. The peaking factor limit reduction on F h will compensate for the hydraulic mismatch with the GAIA fuel with respect to DNB acceptance criterion."

The LAR did not provide the results of this evaluation, specifically the value of the peaking factor penalty to be used on VANTAGE+ fuel during mixed core designs for the fuel transition to GAIA fuel.

Request x

Provide the numerical F h penalties to be used on VANTAGE+ fuel during the transition cycles to GAIA fuel at the Byron and Braidwood stations.

x Confirm the expected VANTAGE+ and GAIA fuel assembly counts during the transition cycles are consistent with the VANTAGE+ and GAIA fuel assembly counts used for the calculation of these penalties.

x Confirm how the penalties will be captured to ensure they will be used during transition cycles.

ATTACHMENT

Page 2 of 9 Constellation Response:

x The power peaking factor (FH) limit reduction is evaluated in NF-CB-24-084 Rev. 0, Byron and Braidwood DNB Mixed Core Analysis Report (provided during the audit) as a FH penalty to be included for Westinghouse VANTAGE+ fuel remaining in the core. For the first transition cycle, a [

]a,c is to be applied to Westinghouse fuel, which will be provided in the Core Operating Limits Report (COLR).

For the second transition cycle, the [

]a,c. The penalties correspond to Westinghouse [

]a,c in the first transition cycles and

[

]a,c in the second transition cycles.

x Per the evaluation (NF-CB-24-084 Rev. 0, Byron and Braidwood DNB Mixed Core Analysis Report) provided during the audit, the first transition cycle is set to have a minimum of 100 Vantage+ assemblies. 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 intentionally chosen to not challenge the transition cycle designs. The first BR1C26 transition loading pattern is already finalized and uses 88 GAIA assemblies. The BR2C26 and BY1C28 loading pattern development work is in progress and also uses an 88-feed loading pattern. In recent history Constellation has not needed to use more than 93 feed assemblies in any reload. Batch sizes are expected to decrease after the first transition cycles due to the larger fuel rod diameter of the GAIA fuel. The second transition cycles are expected to use 80 feed assemblies. However, even 88 feeds were conservatively assumed in the second fuel transition cycle this would still leave 17 (193 88) Westinghouse assemblies in the core. In addition, this minimum Westinghouse fuel assembly requirement is communicated to Framatome in cycle specific generation of the loading pattern requirements (referred to as the Enrichment Setting Groundrules).

x The peaking factor reduction FH is used to compensate for the hydraulic mismatch with the GAIA fuel with respect to DNB. This FH will be [

]a,c in the core design by Westinghouse in their cycle specific RSAC Evaluation to validate any Westinghouse AOR DNB limiting event will bound cycle-specific fuel design characteristics. The COLR will include a Westinghouse [

]a,c. This

[

]a,c will also be applied to the Westinghouse fluxmap peaking factor processing software.

SNSB-RAI-2 Regulatory Basis:

Requirements of GDC 10 Issue The fuel rod bowing methodology applied in the Rod Ejection Accident (XN-75-32(P)(A) and Supplements 1, 2, 3, & 4, Computational Procedure for Evaluating Fuel Rod Bowing, February 1983) has the following Limitation and Condition (L&C):

ATTACHMENT 1 Page 3 of 9 x

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 technical specifications and any remnant penalties must be accommodated into the technical specifications.

To address this L&C, the LAR states:

x Residual Departure from Nucleate Boiling Ratio (DNBR) penalties are not used to offset generic or plant-specific margins. Therefore, this L&C is met.

The resolution of the L&C does not address the concern that plant-specific or generic penalties are used to offset residual DNBR penalties due to fuel rod bowing. Instead, the resolution states plant-specific or generic margins are used without documenting their effects on the residual DNBR penalties. This L&C must be met so this fuel rod bowing methodology can be applied for the Rod Ejection Accident.

Request x

Clarify how the statement in the LAR addresses the limitation and condition related to generic or plant-specific DNBR margins offsetting residual DNBR penalties.

Constellation Response:

Upon further review of this question Framatome has determined that the L&C response offered in Attachment 16 Section 1.2, Page 1-16 of the original LAR submittal (ADAMS Accession No. ML24149A126) quoted below was incorrect.

Residual Departure from Nucleate Boiling Ratio (DNBR) penalties are not used to offset generic or plant-specific margins.

A correction to the response is offered as:

Generic and/or plant-specific margins are not used to offset the application of residual Departure from Nucleate Boiling Ratio (DNBR) rod bow penalties in the BYR/BRW AREA analysis. Since this method discussed in the condition was not performed, there is no requirement to document this in the technical specifications bases. Thus, this condition is met.

ATTACHMENT 1 Page 4 of 9 SNSB-RAI-3 Regulatory Basis x

10 CFR 50 Appendix A, GDC 27, "Combined Reactivity Control System Capability,"

states: "The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained."

x 10 CFR 50 Appendix A, GDC 28, "Reactivity Limits," states: "The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition."

Issue Section 3.8 of the LAR states: "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."

These boundary conditions are important to ensuring that Byron and Braidwood continue to meet GDCs 27 and 28, and any exceptions or changes are not detailed in the LAR.

Request x

Confirm if all the boundary conditions for the design and safety analyses from Westinghouse AORs will apply to mixed cores and describe any exceptions to these boundary conditions.

Constellation Response:

The Westinghouse and Framatome mixed core evaluations share the same plant specific inputs ("Boundary Conditions"):

x Licensed Thermal Power x

Technical Specifications:

o Setpoints for Reactor Protection and Engineering Safeguards o Reactivity and Power Distribution Limits o Reactor Coolant Flows o Emergency Core Cooling System (ECCS) Performance

ATTACHMENT 1 Page 5 of 9 x

Control System Setpoints x

Design of the Nuclear Steam Safety Systems including normal system flows (bypass flows, charging and letdown, etc.)

x Cycle Specific Core Loading Patterns Exceptions to this list are noted below:

x Setpoint for Main Steam Line Break Pressure - The Main Steam Line Break outside containment (UFSAR 15.1.5 and 15.1.6) evaluation performed by Framatome uses a slightly higher Safety Analysis Limit (SAL) (435.3 vs 488.3 psig). The Technical Specification Allowable Value is not changing, but the SAL is changing to remove unneeded conservatism. The value used in the Westinghouse evaluation considered radiological environmental effects from fuel damage even though fuel damage was not predicted (it is allowed in the licensing basis). This conservatism was included but not needed in the Westinghouse analysis; however, the radiological environmental effects were NOT included in the Framatome evaluations.

x Power Distribution Limit for FNDH - Framatome fuel will be designed and verified acceptable up to the current COLR limit for FNH,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.

x Control System Response, Rod Control Speed - 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 withdraw speeds up to 72 steps per minute. The Framatome evaluations consider rod withdrawal speeds up to 8 steps per minute (still, up to 72 steps per minute insertion). A plant modification limiting the rod control system speed controls system will be implemented with the introduction of GAIA fuel.

x Boron Concentrations in ECCS Subsystems -The Boron concentrations in ECCS subsystems will be higher than the current Westinghouse limits. The Framatome analyses which credit the reactivity effects of the boron in ECCS (Main Steam Line Break and Loss of Coolant Accident (LOCA)) will confirm that the reactivity effects meet design basis limits. The Framatome ARCADIA code system is qualified to confirm this for full cores of GAIA and for mixed cores with GAIA and VANTAGE+. The effects of higher boron concentrations on the Westinghouse fuel (clad corrosion, etc.) are being evaluated by Westinghouse with the currently approved methods.

x Moderator Temperature Coefficient (MTC) limits associated with Anticipated Transient Without SCRAM (ATWS) - The least negative MTC limits in Technical Specifications (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. The Westinghouse evaluations apply an upper limit to the MTC per current TS 5.6.5.b.5. This upper limit only applies to ATWS, and that limit is being removed by another license amendment request (ADAMS Accession No. ML25030A158). Both Westinghouse and Framatome methods evaluate the full range of MTC in the COLR and TS 3.1.3 for all other events.

ATTACHMENT 1 Page 6 of 9 SNSB-RAI-4 Regulatory Basis 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" x

(b)(1), "Peak cladding temperature," states: The calculated maximum fuel element cladding temperature shall not exceed 2200° F.

x (b)(2), "Maximum cladding oxidation," states: "The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation."

x (b)(3), "Maximum hydrogen generation," states: "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."

Issue The LAR does not provide any analysis or results for the Braidwood and Byron Unit 2 LOCA analyses to demonstrate compliance with 10 CFR 50.46(b)(1) through (3).

Sections 3.4.1 and 3.4.2 of Attachment 1 to the LAR states that the Braidwood and Byron Unit 2 LOCA analyses will be available for NRC Audit beginning in late 2024.

Request x

Provide the Braidwood and Byron Unit 2 LOCA analyses/results which demonstrate compliance with the requirements in 10 CFR 50.46(b)(1) through (3).

Constellation Response:

Please see attachments 2,3, 8 and 9 of this document.

- 103-4115NP-000, Byron/Braidwood Unit 2 Small Break LOCA Analysis - 103-4116NP-000, Byron/Braidwood Unit 2 Large Break LOCA Analysis - 103-4115P-000, Byron/Braidwood Unit 2 Small Break LOCA Analysis - 103-4116P-000, Byron/Braidwood Unit 2 Large Break LOCA Analysis SNSB-RAI-5 Regulatory Basis 10 CFR 50.46(b)(5), "Long-term cooling," states: "After any calculated successful initial operation of the ECCS [Emergency Core Cooling Systems], 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."

ATTACHMENT 1 Page 7 of 9 Issue Attachments 6 (non-proprietary) and 12 (proprietary) of the LAR address the Braidwood and Byron response to 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), also known as Generic Safety Issue (GSI)-191, "Assessment of Debris Accumulation on PWR Sump Performance." In these attachments, the licensee justifies application of WCAP-17788-P, Revision 1, "Comprehensive Analysis and Test Program for GSl-191 Closure (PA-SEE-1090)," to demonstrate adequate long term core cooling.

Table 6-1, "Summary of Key Inputs-Westinghouse Upflow Barrel/Baffle Plant Design," of Volume 4 of WCAP-17788-P, shows ECCS recirculation flow rate with values ranging from 8 to 40 gallons per minute (gpm)/flow assembly (FA). In Attachments 6 and 12 to the LAR, the "Key Parameter Values for the In-Vessel Debris Effects" table shows the WCAP-17788-P value of 8 gpm/FA along with the Byron/Braidwood plant specific value of 8.6 gpm/FA and concludes that the plant specific ECCS flow rate is greater than minimum analyzed value in WCAP-17788-P. However, Table 6-1 of WCAP-17788-P has a note that states "Only the 40 and 18 gpm/FA flows are used for the K max and t block cases." In this case, the Byron/Braidwood plant specific value of 8.6 gpm/FA would be lower than the minimum analyzed value in WCAP-17788-P of 18 gpm/FA."

Request x

Provide information that demonstrates the Byron/Braidwood plant specific minimum ECCS flow rate will be larger than the WCAP-17788-P value of 18 gpm/FA.

Constellation Response:

The ECCS flow rate reported in Attachment 12 of the LAR is 8.6 gpm/Fuel Assembly (FA).

This is the flow rate during ECCS Cold Leg Recirculation assuming a single failure of one train of ECCS and a cold leg break. However, as discussed here, this value should have been reported for a hot leg break and for two trains of ECCS.

Table 6-1 of WCAP-17788 Volume 4 lists the inputs for the thermal-hydraulic analysis documented in WCAP-17788. These analyses model a large hot leg break, which is limiting in terms of delivering debris to the core, because all ECCS injected to the RCS will reach the downcomer and core. Furthermore, as described in the response to RAI-4.14, "These values represent the maximum expected flow such as having all ECCS train in operation but with ECCS pump head loss degradation consistent with the degradation mechanisms assumed ECCS performance used in the short-term LOCA analyses."

(

Reference:

WCAP-17788, Volume 1 and 4 RAI Responses, Attachment 1 to LTR-SEE-17-102 Revision 0 (ADAMS Accession No. ML18143B745))

The flow rate of 8.6 gpm/FA reported in the LAR is the intact flow rate for one ECCS train of 1659 gpm divided by the number of assemblies (193). The ECCS hydraulic analysis

(

Reference:

BYR06-029/BRW-06-0016-M, Revision 6, "SI/RHR/CS/CV System Hydraulic

ATTACHMENT 1 Page 8 of 9 Analysis in Support of GSI-191") calculates flows during ECCS Cold Leg Recirculation assuming an RCS Cold Leg is broken. The ECCS flow to the broken loop is referred to as spill flow and was not credited as reaching the Reactor Core. Since the WCAP-17788 hydraulic analysis models a Hot Leg Break, no RCS Cold Leg is broken and the full ECCS flow reaches the Reactor Core. Therefore, the total flow to the RCS (intact plus spill) for one ECCS train is 3208 gpm for the RCS Hot leg break, which is equivalent to 16.6 gpm/FA. The flow rate resulting from two ECCS trains will be significantly higher than the minimum flow rate of 18 gpm/FA reported on Table 6.1 of WCAP-17788 Volume 4.

Radiation Protection and Consequence (ARCB) Questions ARCB-RAI-1 By letter dated September 8, 2006 (ML062340420), the NRC approved amendments that fully implemented an alternative source term (AST), pursuant to 10 CFR 50.67, at Braidwood and Byron. For that review, the licensee docketed the calculation of control room (CR) unfiltered in-leakage as 1,000 cubic feet per minute (cfm). By letter dated February 5, 2009, the NRC approved amendments that changed the licensing basis for Braidwood and Byron associated with the application of an AST methodology. For the February 5, 2009, the licensee reduced the calculation of CR unfiltered in-leakage to 500 cfm with the conclusion that "The new assumption of 500 cfm of unfiltered in-leakage bounds the maximum measured in-leakage value of 68 standard cubic feet per minute (scfm) for Byron and 29.3 scfm for Braidwood." In the May 28, 2024, application, CEG further reduces CR unfiltered in-leakage to 436 cfm.

x Provide the results of the most recent measured CR unfiltered in-leakage values.

Constellation Response:

The maximum results for the most recent Main Control Room unfiltered in-leakage testing are:

In-leakage (scfm)

Test Date Reference Work Order Braidwood 155 November 16, 2024 5285905 Byron 318 May 31, 2019 1958961 Note: The in-leakage testing is performed on a 6-year frequency.

ARCB-RAI-2 In support of the initial September 8, 2006 (ML062340420), AST approval, the licensee docketed calculation of CR isolation time critical action as 30 minutes. In the May 28, 2024, application, CEG reduces this value to 20 minutes. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," dated June 2003 (ML031530505), provides guidance that a conservative delay time should be assumed for the operator to complete the necessary actions and provides items that should be considered.

ATTACHMENT 1 Page 9 of 9 x

Provide information that demonstrates the reduced isolation time critical action meets the guidance.

Constellation Response:

With respect to the time critical action for Main Control Room (MCR) ventilation alignment:

There are six accident analysis that credit the operator action to align MCR ventilation:

Loss of Coolant Accident (LOCA), Fuel Handling Accident, Reactor Coolant Pump Locked Rotor, Steam Generator Tube Rupture, Control Rod Ejection, and Main Steam Line break.

As part of this LAR we provided a copy of the LOCA analysis for AST (dose). That analysis uses a 20 minute operator action for one of its assumptions. However, the Control Rod Ejection analysis is the most limiting case with an assumed operator action time of 14 minutes. Therefore, attachment 10 and 11 have 14 minutes as the acceptance criteria.

Please see attachments 10 and 11 to this document.

0 - OP-AA-102-106 Operator Response Time Validation Sheet Attachment 1 for May 22, 2023 - TCA 25 1 - OP-AA-102-106 Operator Response Time Validation Sheet Attachment 1 for May 21, 2024 - TCA 29

ATTACHMENT 2 BRAIDWOOD STATION UNITS 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 Docket Nos. STN-50-456 and STN-50-457 BYRON STATION UNITS 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 Docket Nos. STN-50-454 and STN-50-455 103-4115NP-000, Byron/Braidwood Unit 2 Small Break LOCA Analysis (Non-Proprietary)

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Controlled Document

RS-25-088, Response to Request for Additional Information Related to License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 to Transition to Framatome Gaia Fuel and Exemptions to 10 CFR 50.46

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