View source for 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

{{Adams
| number = ML25111A257
| issue date = 04/21/2025
| title = 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
| author name = Steinman R
| author affiliation = Constellation Energy Generation, LLC
| addressee name = 
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000456, 05000457, 05000454, 05000455
| license number = NPF-072, NPF-077, NPF-037, NPF-066
| contact person = 
| case reference number = RS-25-088
| package number = ML25111A256
| document type = Legal-Affidavit, Letter type:RS, Response to Request for Additional Information (RAI)
| page count = 1
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter: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&deg; 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|date=September 8, 2006|text=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|date=February 5, 2009|text=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) 

Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report ANP-4115NP Revision 0 November 2024 (c) 2024 Framatome Inc.
Document

ANP-4115NP Revision 0 Copyright &#xa9; 2024 Framatome Inc.
All Rights Reserved FRAMATOME TRADEMARKS GAIA, GALILEO, GRIP, HMP, M5, M5Framatome, MONOBLOC, and S-RELAP5 are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries.
Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page i Nature of Changes Item Section(s) or Page(s)
Description and Justification 1
All Initial Issue Controlled Document

Framatome Inc.
ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page ii Contents Page 1.0 INTRODUCTION............................................................................................... 1-1 2.0 

==SUMMARY==
OF RESULTS................................................................................. 2-1 3.0 DESCRIPTION OF ANALYSIS.......................................................................... 3-1 3.1 Acceptance Criteria................................................................................ 3-1 3.2 Description of SBLOCA Event................................................................ 3-1 3.3 Description of Analytical Methods........................................................... 3-4 3.4 Plant Description and Summary of Analysis Parameters........................ 3-5 3.5 Safety Evaluation Compliance................................................................ 3-7 4.0 SBLOCA ANALYSIS......................................................................................... 4-1 4.1 Cold Leg Pump Discharge Break Spectrum Results.............................. 4-1 4.2 Discussion of Transient for Limiting PCT Break...................................... 4-2 4.3 Attached Piping Break Results............................................................... 4-3 4.4 Delayed RCP Trip Study......................................................................... 4-4 4.5 ECCS Temperature Sensitivity Study..................................................... 4-5 5.0 REFERENCES.................................................................................................. 5-1 Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page iii List of Tables Table 2-1 10 CFR 50.46 (b) (1-4) Compliance........................................................ 2-2 Table 3-1 System Parameters and Initial Conditions.............................................. 3-9 Table 3-2 High Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum................................................................... 3-10 Table 3-3 Intermediate Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum................................................................... 3-11 Table 3-4 Low Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum................................................................... 3-12 Table 3-5 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications.......................................................................... 3-17 Table 4-1 Summary of Cold Leg Pump Discharge Break Spectrum Results.......... 4-6 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum..... 4-7 Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page iv List of Figures Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization................. 3-13 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization......................... 3-14 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization............................... 3-15 Figure 3-4 Axial Power Distribution Comparison.................................................... 3-16 Figure 4-1 Cold Leg Pump Discharge Break Spectrum Peak Cladding Temperature versus Break Size........................................................... 4-13 Figure 4-2 Reactor Power - 8.70-inch Break......................................................... 4-14 Figure 4-3 Primary and Secondary System Pressures - 8.70-inch Break............. 4-15 Figure 4-4 Break Mass Flow Rate - 8.70-inch Break............................................. 4-16 Figure 4-5 Break Vapor Void Fraction - 8.70-inch Break....................................... 4-17 Figure 4-6 Loop Seal Upside Collapsed Levels - 8.70-inch Break........................ 4-18 Figure 4-7 Downcomer Collapsed Liquid Level - 8.70-inch Break......................... 4-19 Figure 4-8 Primary System Masses - 8.70-inch Break.......................................... 4-20 Figure 4-9 RCS Loop Mass Flow Rates - 8.70-inch Break.................................... 4-21 Figure 4-10 Steam Generator Main Feedwater Flow Mass Rates - 8.70-inch Break.................................................................................................... 4-22 Figure 4-11 Steam Generator MSSV Mass Flow Rates - 8.70-inch Break.............. 4-23 Figure 4-12 Steam Generator Auxiliary Feedwater Flow Rate - 8.70-inch Break.... 4-24 Figure 4-13 Steam Generator Total Secondary Side Mass - 8.70-inch Break......... 4-25 Figure 4-14 Steam Generator Narrow Range Level - 8.70-inch Break.................... 4-26 Figure 4-15 High Head Safety Injection Mass Flow Rates - 8.70-inch Break.......... 4-27 Figure 4-16 Intermediate Head Safety Injection Mass Flow Rates - 8.70-inch Break.................................................................................................... 4-28 Figure 4-17 Low Head Safety Injection Mass Flow Rates - 8.70-inch Break........... 4-29 Figure 4-18 Accumulator Mass Flow Rates - 8.70-inch Break................................ 4-30 Figure 4-19 Total ECCS and Break Mass Flow Rates - 8.70-inch Break................ 4-31 Figure 4-20 Hot Assembly Collapsed Liquid Level - 8.70-inch Break...................... 4-32 Figure 4-21 Cladding Temperature at PCT Node - 8.70-inch Break........................ 4-33 Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page v Nomenclature Acronym Definition AFW Auxiliary Feedwater BOC Beginning-of-Cycle CFR Code of Federal Regulations CWO Core Wide Oxidation DC Downcomer ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EOC End-of-Cycle F'H Nuclear Enthalpy Rise Factor/Radial Peaking Factor FQ Total Peaking Factor Framatome Framatome Inc.
HEM Homogeneous Equilibrium Model HHSI High Head Safety Injection HMP High Mechanical Performance IGM Intermediate GAIA Mixing Grid IHSI Intermediate Head Safety Injection k(z)
Axial-Dependent Peaking Factor LEU Low-Enriched Uranium LHSI Low Head Safety Injection LOCA Loss-of-Coolant Accident LOOP Loss-of-Offsite Power LS Loop Seal LSC Loop Seal Clearing Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page vi Acronym Definition MFW Main Feedwater MLO Maximum Local Oxidation MSSV Main Steam Safety Valve NRC U.S. Nuclear Regulatory Commission RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RPS Reactor Protection System RT Reactor Trip RV Reactor Vessel PCT Peak Cladding Temperature PWR Pressurized Water Reactor PZR Pressurizer SBLOCA Small Break Loss-of-Coolant Accident SE Safety Evaluation SG Steam Generator SI Safety Injection SIAS Safety Injection Actuation Signal Tavg RCS Average Temperature TT Turbine Trip Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 1-1 

==1.0 INTRODUCTION

==
This report summarizes the small break loss-of-coolant accident (SBLOCA) analysis for Byron/Braidwood Unit 2. The purpose of the SBLOCA analysis is to support the fuel transition at Byron/Braidwood Unit 2 to the Framatome GAIA fuel design. This analysis was performed in accordance with the U.S. Nuclear Regulatory Commission (NRC)-approved S-RELAP5-based methodology described in Reference 1 as supplemented by Reference 2 and Reference 3.
Byron/Braidwood Unit 2 is a four-loop, Westinghouse-designed Pressurized Water Reactor (PWR). The Framatome GAIA fuel design with M5Framatome cladding for Byron/Braidwood Unit 2 consists of a 17x17 array with GAIA and intermediate GAIA Mixing (IGM) 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 nozzle.
The analysis supports plant operation at a core power level of 3658 MWt (includes measurement uncertainty), a maximum-allowed total peaking factor (FQ) of 2.6 (represents total peaking with uncertainties applied and an axial-dependent factor k(z) set to 1.0), a radial peaking factor of (F+) of 1.70 (includes measurement uncertainty),
and up to 10% steam generator (SG) tube plugging per SG.
A complete spectrum including cold leg pump discharge break sizes ranging from 1.00 inch diameter to 8.70 inch diameter and breaks in attached piping were considered. Other supporting analyses prescribed by the methodology to assess a delayed reactor coolant pump (RCP) trip and the sensitivity to Emergency Core Cooling System (ECCS) fluid temperature were performed.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 2-1 2.0

==SUMMARY==
OF RESULTS The SBLOCA analysis results demonstrate the adequacy of the ECCS to satisfy the 10 CFR 50.46(b) (1-4) criteria (Reference 5) for Byron/Braidwood Unit 2 operating with Framatome supplied GAIA fuel design with M5Framatome cladding. The limiting peak cladding temperature (PCT) is 1625&deg;F for an 8.70-inch diameter cold leg pump discharge break. The limiting PCT case also produced the limiting core wide oxidation (CWO) of 0.010% and a limiting maximum local oxidation (MLO) of 3.35%. The total MLO value includes [ 
 ] Compliance to the 10 CFR 50.46 (b) (1-4) criteria is summarized in Table 2-1.
In addition to the analysis of the cold leg pump discharge breaks, breaks in the attached piping were considered. The attached piping break analyses considered breaks in the accumulator line and HHSI line.
Two supporting analyses prescribed by the methodology were performed to investigate a delayed RCP trip and a different ECCS temperature. The results of the delayed RCP trip study demonstrated that there is a maximum of 5 minutes for operators to trip all four RCPs after the specified trip criteria are met. The ECCS temperature sensitivity study analyzed the sensitivity to ECCS fluid temperatures different from those used in the break spectrum analysis. The results of the ECCS temperature sensitivity study support the applicability of the ECCS temperature used in the SBLOCA analysis.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 2-2 Table 2-1 10 CFR 50.46 (b) (1-4) Compliance 10 CFR 50.46 Parameter Analysis Value 10 CFR 50.46 Criteria
: 1. PCT 1625&deg;F 
&deg;F
: 2. Maximum Local Oxidation 3.35% 

: 3. Core Wide Oxidation 0.010%

: 4. Coolable Geometry The application of the SBLOCA method and its models and results demonstrating compliance with the above criteria confirm a coolable geometry.
Maintain coolable geometry Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-1 3.0 DESCRIPTION OF ANALYSIS 3.1 Acceptance Criteria The purpose of the analysis is to verify the adequacy of the Byron/Braidwood Unit 2 ECCS by demonstrating compliance with the following 10 CFR 50.46(b) criteria (Reference 5):
: 1. The calculated maximum fuel element cladding temperature shall not exceed 2200&deg;F.
: 2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
: 3. 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. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
3.2 Description of SBLOCA Event The postulated SBLOCA is defined as a break in the Reactor Coolant System (RCS) pressure boundary with an area less than or equal to 10% of the cold leg pipe area.
The most limiting break location is typically in the cold leg pipe on the discharge side of the 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 difference between the core exit and the top of the downcomer (DC). This typically produces the greatest degree of core uncovery, the longest fuel rod heatup time and, consequently, the greatest challenge to the 10 CFR 50.46(b)(1-4) criteria (Reference 5).
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-2 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 boil-off, (5) core recovery and long-term cooling. The duration of each of these phases is break size and system dependent.
Following the break, the RCS rapidly depressurizes to the saturation pressure of the hot leg fluid. During the initial depressurization phase, a reactor trip signal is generated on low pressurizer pressure, and the turbine is tripped on the reactor trip. The assumption of a loss-of-offsite power (LOOP) concurrent with the reactor scram results in RCP trip.
In the second phase of the transient, the RCS transitions to a quasi-equilibrium condition in which the core decay heat, leak flow, SG heat removal, and system hydrostatic head balance combine to control the core inventory. During this period, the RCPs are coasting down and the system drains from the top of the RCS with the first voiding occurring at the top of the SG tubes, in the reactor vessel (RV) upper head, and at the top of the RV upper plenum region. Also, the loop seals remain plugged during this phase, trapping vapor generated by the core in the RCS, resulting in a low-quality flow at the break.
The third phase in the transient is characterized by loop seal clearing (LSC). During this phase the loop seal (i.e., liquid trapped in the RCP suction piping) can prevent steam from venting to the break. The maximum pressure difference between the RV upper head and DC is reached when the liquid level on the downside of the SG is depressed to the elevation of the horizontal loop seal piping. When this point is reached, the loop seal clears, and the trapped steam can be vented to the break. For some break sizes, the transient develops slowly, and the core can become temporarily uncovered before the loop seal clears. Following LSC, the break flow transitions to primarily steam and the core recovers to approximately the cold leg elevation as pressure imbalances throughout the RCS are relieved.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-3 The fourth phase is characterized as core boil-off. With the loop seal cleared, the venting of steam through the break causes a rapid RCS depressurization below the secondary pressure. As boiling increases in the core, the core mixture level decreases.
The core mixture level will reach a minimum, in some cases resulting in deep core uncovering. The transient boil-off period ends when the core liquid level reaches this minimum. At this time, the RCS has depressurized to the point where ECCS flow into the RV matches the rate of boil-off from the core.
The last phase of the transient is characterized as core recovery. The core recovery period extends from the time at which the core mixture level reaches a minimum in the core boil-off phase until all parts of the core are quenched and covered by a low-quality mixture. Core recovery is provided by pumped injection and passive accumulator injection when the RCS pressure decreases below the accumulator pressure.
Generally, PCT occurs at the beginning of the core recovery phase before the mixture level has increased enough to provide enhanced cooling to the PCT location on the hot rod.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-4 The SBLOCA transient progression is dependent on the size of the break and is typically broken into three different break size ranges. For break sizes towards the larger end of the break spectrum, significant RCS inventory loss results in more rapid RCS depressurization to the accumulator actuation pressure. Accumulator flow provides sufficient inventory early in the transient to limit the core uncovery and hot rod heatup.
For break sizes in the middle of the spectrum, the rate of inventory loss from the RCS is such that the HHSI pumps typically cannot preclude significant core uncovery. The RCS depressurization rate is slow, extending the time required to reach the accumulator injection pressure, if reached at all. Break sizes in this range, will either exhibit core recovery with the HHSI pumped injection alone while the RCS pressure remains barely above the accumulator injection setpoint, or exhibit core recovery from accumulator injection. For break sizes at the low end of the spectrum, the RCS pressure does not reach the accumulator injection pressure. However, RCS inventory loss is not significant and typically within the means of HHSI makeup capacity such that core uncovery is minimal if not precluded.
3.3 Description of Analytical Methods This analysis was performed in accordance with the NRC-approved S-RELAP5-based methodology, described in Reference 1 as supplemented by Reference 2 and Reference 3. The Framatome S-RELAP5 SBLOCA evaluation model for event response of the primary and secondary systems and the hot fuel rod is based on the use of two computer codes. The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50 (Reference 6), are incorporated.
The two Framatome computer codes used in this analysis are:
: 1. The GALILEO code was used to determine the burnup dependent initial fuel rod conditions for the system calculations.
: 2. The S-RELAP5 code was used to predict the primary and secondary system thermal-hydraulic and hot rod transient response.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-5 Representative system nodalization figures for a Westinghouse four-loop plant are shown in Figure 3-1 through Figure 3-3. See Section 3.4 for a description of the Byron/Braidwood Unit 2 analysis model.
3.4 Plant Description and Summary of Analysis Parameters The plant analyzed is the Byron/Braidwood Unit 2, Westinghouse-designed PWR, which has four loops, each with a hot leg, a U-tube steam generator, and a cold leg with an RCP. The RCS includes one pressurizer connected to a hot leg. Main feedwater (MFW) is injected into the downcomer of each SG. The auxiliary feedwater (AFW) system provides flow to the four SGs when normal feedwater is not available. The ECCS provides injection to each of the four loops via the centrifugal charging/HHSI system, SI/intermediate head safety injection (IHSI) system, residual heat removal (RHR)/low head safety injection (LHSI) system, and accumulators. For the purpose of this report, the centrifugal charging/HHSI, SI/IHSI, and RHR/LHSI systems are referred to as the HHSI, IHSI, and LHSI systems, respectively.
The RCS, SG, reactor vessel, pressurizer, and ECCS are explicitly modeled in the S-RELAP5 model to provide an accurate representation of the plant. The model includes four accumulators, a pressurizer, and four SGs with both primary and secondary sides modeled. For the secondary side, the model includes the main steam lines between their respective SGs and the turbine control valve, including the connected main steam safety valve (MSSV) inlet piping.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-6 For each RCS loop, the ECCS model includes an injection connection to the cold leg for the accumulator and another connection for HHSI. IHSI and LHSI are modeled with separate injection connections to each of the four accumulator lines. The accumulator and HHSI injection connections to the cold leg pipe are downstream of the RCP discharge. The ECCS pumped injection is modeled as a table of flow versus RCS backpressure. A model is included to account for the potential presence of nitrogen in the accumulator lines. This model works by delaying the injection of the accumulator line ECCS to the cold leg until an equivalent amount of liquid has been swept out of the accumulator line.
Important system parameters and initial conditions used in the analysis are given in Table 3-1. The heat generation rate in the S-RELAP5 reactor core model is determined from reactor kinetics equations with actinide and decay heating as prescribed by 10 CFR 50 Appendix K (Reference 6).
The break spectrum analysis assumes a LOOP concurrent with reactor scram, which is based on the reactor protection system (RPS) low pressurizer pressure reactor trip and includes delays as stated in Table 3-1. The assumption of LOOP concurrent with reactor scram results in an RCP trip.
The RCPs are tripped at the time of reactor scram, instead of the opening of the break (time zero). This is considered to be conservative, since continued RCP operation will delay LSC. This delay in LSC will result in additional RCS inventory loss since the break flow is mostly liquid until the time of LSC. After LSC, a path for steam venting is established and the break flow transitions from liquid to steam, lowering the break mass flow rate.
The single failure criterion required by 10 CFR 50 Appendix K (Reference 6) is satisfied by assuming the loss of one emergency diesel generator (EDG). The loss of an EDG disables one of the two ECCS trains. Thus, one AFW pump, one HHSI pump, one IHSI pump, and one LHSI pump are assumed unavailable.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-7 Following the safety injection actuation system (SIAS) activation on low pressurizer pressure, actuation of the HHSI, IHSI, and LHSI systems are delayed by 50 seconds.
Table 3-2, Table 3-3, and Table 3-4 show the minimum ECCS flow rates with one EDG failure for HHSI, IHSI, and LHSI, respectively, for a break in the RCS loop. The pumped safety injection flow to the intact loops is modeled to be distributed equally among the three intact loops.
With one of the two AFW pumps assumed unavailable for the single failure criterion, AFW flow is supplied via the remaining AFW pump to the four SGs. AFW injection is delayed by 75 seconds beyond the time of AFW system initiation on low-low SG level.
A SG tube plugging level of 10% is modeled in each SG. The MSSVs are set to open at their nominal setpoints plus 5% accumulation.
The axial power shapes for this analysis are shown in Figure 3-4. The figure shows the input axial power shape and the axial power shape after being adjusted so that it is consistent with the Technical Specification total and radial peaking factors.
3.5 Safety Evaluation Compliance The SBLOCA analysis for Byron/Braidwood Unit 2 presented herein is consistent with the submitted SBLOCA methodology documented in EMF-2328, Revision 0 (Reference 1) as modified by EMF-2328, Revision 0, Supplement 1, Revision 0 (Reference 2) and supplemented by ANP-10349P-A, Revision 0 (Reference 3). The limitations and conditions from the NRC Safety Evaluation (SE) for EMF-2328, Revision 0, are addressed below. There are no limitations and conditions from the NRC for EMF-2328, Revision 0, Supplement 1. Reference 3 contains a condition related to the range of applicability of GALILEO applications in Reference 4, which are addressed in Table 3-5. From the disposition provided below and the responses given in Table 3-5, the Byron/Braidwood Unit 2 SBLOCA analysis documented herein is compliant with all requirements of the applicable topical reports.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-8 There is one SE limitation and condition for the application of the SBLOCA evaluation model EMF-2328, Revision 0 (Reference 1), that S-RELAP5 is acceptable for modeling transients where the break flow area is less than or equal to 10% of the cold leg flow area. A spectrum of cold leg break sizes from 0.00545 ft2 (1.00-inch diameter) to 0.41282 ft2 (8.70-inch diameter, 10% of cold leg pipe area) are analyzed. This satisfies the limitation placed on EMF-2328, Revision 0 for the cold leg break spectrum.
The attached pipe break in the accumulator lines is greater than 10% of the cold leg flow area with a break area of 0.4176 ft2 (8.75-inch accumulator line inside diameter).
However, the flow area limitation on EMF-2328, Revision 0 (Reference 1) is addressed for this break location in Supplement 1 to EMF-2328 (Reference 2) which is applied for the Byron/Braidwood Unit 2 SBLOCA analyses. Therefore, no further actions are required to address the limitation for the accumulator line break area.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-9 Table 3-1 System Parameters and Initial Conditions Parameter Value Reactor Power (MWt) 3658(1)
Axial Power Shape Figure 3-4 Radial Peaking Factor (F'H) 1.70(1)
Maximum-Allowed Total Power Peaking Factor (FQ) 2.6(2)
Total RCS Flow Rate (gpm) 368,000 Pressurizer Pressure (psia) 2249.3 RCS Operating Temperature, Tavg (&deg;F) 588 SG Tube Plugging per SG (%)
10 SG Secondary Pressure (psia) 914 MFW Temperature (&deg;F) 447.5 RPS Low Pressurizer Pressure for Reactor Trip (psia) 1859.3 RPS Low Pressurizer Pressure Trip Delay(3) (sec) 2 SIAS Low Pressurizer Pressure Activation Setpoint (psia) 1714.3 Accumulator Pressure (psia) 602 Accumulator Fluid Temperature (&deg;F) 120 Accumulator Water Volume per Accumulator (ft3) 950 AFW Temperature (&deg;F) 125 Total AFW Flow Rate (gpm) 560(4)
AFW Initiation on Low-Low SG Narrow Range Level Setpoint (% Narrow Range Span) 18.6 AFW Injection Delay (sec) 75(5)
ECCS Pumped Injection Temperature (&deg;F) 120 HHSI, IHSI, and LHSI Injection Delay Time on SIAS (sec) 50 MSSV Lift Pressure and Accumulation Nominal + 5% Accumulation 1 Includes measurement uncertainty.
2 Includes uncertainties and k(z) set to 1.0.
3 Includes scram delay.
4 Total flow is split evenly among the four SGs.
5 Delay applied to all analyses regardless of offsite power availability.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-10 Table 3-2 High Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum RCS Pressure (psia)
Total Intact Loops Flow (gpm)
Broken Loop Flow (gpm) 15 278 114 20 278 114 40 277 114 50 276 114 60 276 113 80 276 113 100 275 113 120 273 112 156 272 111 157 271 111 180 270 111 182 270 111 300 270 111 400 263 108 500 257 105 600 249 102 700 242 99 800 235 96 900 226 93 1000 219 90 1100 210 86 1200 201 83 1300 192 79 1339 189 78 1400 163 67 1500 153 63 1600 142 58 1700 129 53 1800 116 48 1900 103 42 2000 87 36 2100 72 30 2200 43 18 2250 23 10 2282 0
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-11 Table 3-3 Intermediate Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum RCS Pressure (psia)
Total Intact Loops Flow (gpm)
Broken Loop Flow (gpm) 15 403 163 20 402 162 40 400 161 50 398 161 60 397 160 80 394 159 100 392 158 120 390 157 156 385 155 157 385 155 180 382 154 182 382 154 300 356 144 400 338 137 500 320 129 600 300 121 700 280 113 800 259 105 900 233 94 1000 204 83 1100 168 68 1200 123 50 1300 47 19 1339 0
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-12 Table 3-4 Low Head Safety Injection Flow Rates for Cold Leg Pump Discharge Break Spectrum RCS Pressure (psia)
Total Intact Loops Flow (gpm)
Broken Loop Flow (gpm) 15 2461 1053 20 2421 1036 40 2259 968 50 2175 932 60 2085 893 80 1890 810 100 1672 715 120 1417 607 156 862 365 157 605 254 180 33 10 182 0
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-13 Figure 3-1 S-RELAP5 SBLOCA Reactor Coolant System Nodalization Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-14 Figure 3-2 S-RELAP5 SBLOCA Secondary System Nodalization Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-15 Figure 3-3 S-RELAP5 SBLOCA Reactor Vessel Nodalization Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-16 Figure 3-4 Axial Power Distribution Comparison Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 3-17 Table 3-5 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications Ranges of Applicability (Section 1.2 of the SE in Reference 4) 

===Response===
Pressurized water reactor designs using Low-Enriched Uranium (LEU) fuel loading This analysis was performed for the Byron/Braidwood Unit 2 plant, which is a PWR, 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 loss-of-coolant accident (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.
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.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-1 4.0 SBLOCA ANALYSIS The analysis results demonstrate the adequacy of the ECCS to satisfy the criteria given in 10 CFR 50.46(b)(1-4) for Byron/Braidwood Unit 2 operating with Framatome supplied GAIA fuel design with M5Framatome cladding.
4.1 Cold Leg Pump Discharge Break Spectrum Results The Byron/Braidwood Unit 2 break spectrum analysis for SBLOCA includes breaks of varying diameter up to 10% of the flow area for the cold leg pump discharge. The spectrum includes a break size range from 1.00 to 8.70 inches in diameter, where the break size interval is sufficient to establish a PCT trend. Additional break sizes are analyzed with a smaller break interval once the potential limiting break size is determined to confirm the limiting break size, which is the case with the highest PCT.
[ 
 ] Figure 4-1 displays the PCT results as a function of break size. For the break spectrum analysis, RCP trip is assumed to occur on reactor scram.
The results of the cold leg pump discharge SBLOCA break spectrum analysis are presented in Table 4-1. The predicted event times for the break spectrum are provided in Table 4-2. The limiting PCT break size is determined to be 8.70 inches in diameter (0.41282 ft2), resulting in a PCT of 1625&deg;F. The limiting PCT case also produced the limiting core wide oxidation (CWO) of 0.010% and a transient MLO value of 0.69%. The fraction of total hydrogen generated was not directly calculated; however, it is conservatively bounded by the calculated total core wide percent oxidation, which is well below the 1 percent limit.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-2 4.2 Discussion of Transient for Limiting PCT Break The limiting PCT break from the cold leg pump discharge break spectrum is the 8.70-inch diameter break with a PCT of 1625&deg;F. The transient progression is shown in Figure 4-2 through Figure 4-21. The cladding temperature at the PCT location is shown in Figure 4-21. Key transient results are presented in Table 4-1. The sequence of events is provided in Table 4-2.
The break opens at t=0 seconds and initiates a subcooled depressurization of the RCS.
The RPS low pressurizer pressure trip setpoint is reached at 0.42 seconds and at 2.42 seconds the reactor is scrammed, coincident with the assumed RCP and turbine trips (Figure 4-2 and Figure 4-9). MFW is also isolated on reactor scram (Figure 4-10). The pressure in the secondary side does not reach the lowest MSSV lift setpoint pressure (Figure 4-11).
The SIAS is issued at 8.94 seconds. Following the EDG loading delay, the HHSI and IHSI begin to inject at 59 seconds (Figure 4-15 and Figure 4-16). Prior to this, the core began to uncover at 41 seconds with effective cooling lost to most of the hot assembly in a short period of time (Figure 4-20). HHSI and IHSI do not provide sufficient inventory to offset the large amount of coolant mass lost out the break at this time (Figure 4-19).
All four loop seals clear before the time of PCT, with the broken loop clearing first after 83 seconds, followed closely by the other three loops clearing between 83 and 95 seconds (Figure 4-6). The clearing of the loop seals produces a temporary increase in core level between 83 and 95 seconds (Figure 4-20). However, the mixture level remains near the bottom of the active core, resulting in continuation of the clad temperature excursion (Figure 4-21).
The accumulator injection is delivered to the cold leg between 145 and 147 seconds (Figure 4-18). The minimum RV mass occurs around 151 seconds (Figure 4-8). Clad rupture is predicted at 164 seconds.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-3 The cladding temperature excursion is terminated at 165 seconds with a PCT of 1625&deg;F (Figure 4-21). The core is quenched at approximately 210 seconds with the initial accumulator injection ending around the same time (Figure 4-18). LHSI initially begins injecting at 167 seconds, just after the time of PCT. However, the LHSI flow rate is significantly less than that of the accumulators and any impact on transient mitigation is considered minimal (Figure 4-17 and Figure 4-18). By the time of core quench, enough decay heat is being removed and an adequate mixture level is sustained primarily by pumped ECCS injection (Figure 4-15, Figure 4-16, and Figure 4-17).
4.3 Attached Piping Break Results The ECCS must cope with ruptures of the main RCS piping and breaks in attached piping. To demonstrate this, as prescribed by the NRC-approved supplement to EMF-2328 (Reference 2), an analysis of the ruptures in attached piping that compromise the ability to inject emergency coolant into the RCS is performed. The size of the rupture and the portion of ECCS lost directly to containment are dependent on the plant design.
The Byron/Braidwood Unit 2 plant design injects IHSI and LHSI to the accumulator injection line which is connected to each cold leg while HHSI is injected through a separate line that is connected to each cold leg. Therefore, two break locations are analyzed, one in the accumulator line and one in the HHSI line. The break areas analyzed represents a double-ended guillotine of the accumulator line and HHSI line.
The accumulator line break resulted in a PCT of 1525&deg;F, transient MLO of 0.26%, and CWO of 0.006%. The HHSI line break resulted in a PCT of 1203&deg;F, transient MLO of 0.12%, and CWO of 0.003%. The accumulator line and HHSI break analysis results are less limiting than those of the break spectrum analysis.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-4 4.4 Delayed RCP Trip Study The delayed RCP trip study is performed in accordance with the NRC-approved supplement to the EMF-2328 methodology (Reference 2). For plants such as Byron/Braidwood Unit 2 that do not have an automatic RCP trip, a delayed RCP trip can potentially result in a more limiting condition than tripping the RCPs at reactor scram.
Continued operation of the RCPs can result in more overall inventory loss out the break.
It has been postulated that tripping the pumps when the minimum RCS inventory occurs could cause a collapse of voids in the core, thus depressing the core level and provoking a deeper core uncovering, and a potentially higher PCT. Therefore, the methodology prescribes an RCP trip study for both the cold and hot leg breaks consistent with the plant licensing basis and Emergency Operating Procedures.
For Byron/Braidwood Unit 2, a delayed RCP trip time for operator action of 5 minutes following the RCP trip criteria is analyzed to evaluate the adequacy of the trip criteria and delay time by demonstrating compliance to 10 CFR 50.46(b)(1-4) criteria (Reference 5). The RCP trip criteria, which are based on plant-specific procedures, are modeled as the occurrence of the RCS pressure being below a specified value and at least one HHSI pump or one IHSI pump meeting certain flow criteria. The spectrum of cold and hot leg breaks in this study includes break sizes from 1.00 to 8.70 inches.
The results of the delayed RCP trip cases indicate that there is at least 5 minutes for operators to trip all four RCPs after the specified trip criteria is met with considerable margin to the 10 CFR 50.46(b)(1-4) criteria.
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-5 4.5 ECCS Temperature Sensitivity Study Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-6 Table 4-1 Summary of Cold Leg Pump Discharge Break Spectrum Results 6 No clad heat-up experienced, therefore, reported value is the initial clad temperature.
7 [ 
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8 [ 
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-7 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum 9 The time stated is when flow is delivered to the cold leg 10 No clad heat-up experienced, therefore, reported value is the initial clad temperature.
11 [ 
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12 [ 
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-8 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd) 13 The time stated is when flow is delivered to the cold leg Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-9 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd) 14 The time stated is when flow is delivered to the cold leg Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-10 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-11 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-12 Table 4-2 Sequence of Events for Cold Leg Pump Discharge Break Spectrum (contd)
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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-13 Figure 4-1 Cold Leg Pump Discharge Break Spectrum Peak Cladding Temperature versus Break Size Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-14 Figure 4-2 Reactor Power - 8.70-inch Break Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-15 Figure 4-3 Primary and Secondary System Pressures - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-16 Figure 4-4 Break Mass Flow Rate - 8.70-inch Break Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-17 Figure 4-5 Break Vapor Void Fraction - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-18 Figure 4-6 Loop Seal Upside Collapsed Levels - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-19 Figure 4-7 Downcomer Collapsed Liquid Level - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-20 Figure 4-8 Primary System Masses - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-21 Figure 4-9 RCS Loop Mass Flow Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-22 Figure 4-10 Steam Generator Main Feedwater Flow Mass Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-23 Figure 4-11 Steam Generator MSSV Mass Flow Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-24 Figure 4-12 Steam Generator Auxiliary Feedwater Flow Rate - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-25 Figure 4-13 Steam Generator Total Secondary Side Mass - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-26 Figure 4-14 Steam Generator Narrow Range Level - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-27 Figure 4-15 High Head Safety Injection Mass Flow Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-28 Figure 4-16 Intermediate Head Safety Injection Mass Flow Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-29 Figure 4-17 Low Head Safety Injection Mass Flow Rates - 8.70-inch Break Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-30 Figure 4-18 Accumulator Mass Flow Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-31 Figure 4-19 Total ECCS and Break Mass Flow Rates - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-32 Figure 4-20 Hot Assembly Collapsed Liquid Level - 8.70-inch Break Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 4-33 Figure 4-21 Cladding Temperature at PCT Node - 8.70-inch Break Controlled Document

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ANP-4115NP Revision 0 Byron/Braidwood Unit 2 Small Break LOCA Analysis Licensing Report Page 5-1 

==5.0 REFERENCES

==
: 1. EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.
: 2. EMF-2328(P)(A), Revision 0; Supplement 1(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, December 2016.
: 3. ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021.
: 4. ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, November 2020.
: 5. Code of Federal Regulations, Title 10, Part 50, Section 46, Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors, August 2007.
: 6. Code of Federal Regulations, Title 10, Part 50, Appendix K, ECCS Evaluation Models, June 2000.
Document

ATTACHMENT 3 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-50-455 103-4116NP-000, Byron/Braidwood Unit 2 Large Break LOCA Analysis (Non-Proprietary) 

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Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report ANP-4116NP Revision 0 November 2024 (c) 2024 Framatome Inc.
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ANP-4116NP Revision 0 Copyright &#xa9; 2024 Framatome Inc.
All Rights Reserved FRAMATOME TRADEMARKS GAIA, GALILEO, GRIP, HMP, M5, M5Framatome, MONOBLOC, and S-RELAP5 are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page i Nature of Changes Item Section(s) or Page(s)
Description and Justification 1
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page ii Contents Page 1.0 INTRODUCTION............................................................................................... 1-1 2.0 

==SUMMARY==
OF RESULTS................................................................................. 2-1 3.0 DESCRIPTION OF ANALYSIS.......................................................................... 3-1 3.1 Acceptance Criteria................................................................................ 3-1 3.2 Description of LBLOCA Event................................................................. 3-2 3.3 Description of Analytical Models............................................................. 3-3 3.4 GDC-35 Limiting Condition Determination.............................................. 3-6 3.5 Overall Statistical Compliance to Criteria................................................ 3-7 3.6 Plant Description..................................................................................... 3-7 3.7 Safety Evaluation Limitations.................................................................. 3-9 4.0 RLBLOCA ANALYSIS....................................................................................... 4-1 4.1 RLBLOCA Results.................................................................................. 4-1 4.2 Conclusions............................................................................................ 4-3 5.0 REFERENCES.................................................................................................. 5-1 For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page iii List of Tables Table 3-1 EMF-2103P-A, Revision 3, SE Limitations Evaluation............................. 3-10 Table 3-2 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications............................................................................................. 3-14 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges...................... 4-4 Table 4-2 Statistical Distribution Used for Process Parameters................................ 4-8 Table 4-3 Passive Heat Sinks and Material Properties in Containment Geometry................................................................................................... 4-9 Table 4-4 Compliance with 10 CFR 50.46(b)........................................................... 4-10 Table 4-5 Summary of Major Parameters for the Demonstration Case................... 4-11 Table 4-6 Calculated Event Times for the Demonstration Case.............................. 4-12 Table 4-7 Fuel Rod Rupture Ranges of Parameters............................................... 4-13 For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page iv List of Figures Figure 3-1 Representative Primary System Noding.................................................. 3-15 Figure 3-2 Representative Secondary System Noding............................................. 3-16 Figure 3-3 Representative Reactor Vessel Noding................................................... 3-17 Figure 4-1 Scatter Plot Key Parameters................................................................... 4-14 Figure 4-1 Scatter Plot Key Parameters (continued)................................................ 4-15 Figure 4-2 PCT versus PCT Time Scatter Plot......................................................... 4-16 Figure 4-3 PCT versus Break Size Scatter Plot........................................................ 4-17 Figure 4-4 MLO vs PCT Scatter Plot........................................................................ 4-18 Figure 4-5 CWO vs PCT Scatter Plot....................................................................... 4-19 Figure 4-6 Peak Cladding Temperature (Independent of Elevation)......................... 4-20 Figure 4-7 Break Flow.............................................................................................. 4-21 Figure 4-8 Core Inlet Mass Flux................................................................................ 4-22 Figure 4-9 Core Outlet Mass Flux............................................................................. 4-23 Figure 4-10 Void Fraction at RCS Pumps.................................................................. 4-24 Figure 4-11 ECCS Flows (Includes Accumulator, HHSI, IHSI and LHSI)................... 4-25 Figure 4-12 Upper Plenum Pressure.......................................................................... 4-26 Figure 4-13 Collapsed Liquid Level in the Downcomer.............................................. 4-27 Figure 4-14 Collapsed Liquid Level in the Lower Plenum.......................................... 4-28 Figure 4-15 Core Collapsed Liquid Level................................................................... 4-29 Figure 4-16 Containment and Loop Pressures........................................................... 4-30 Figure 4-17 Pressure Differences between Upper Plenum and Downcomer............. 4-31 Figure 4-18 [ 
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page v Nomenclature Acronym Definition CC Centrifugal Charging C-P Cathcart-Pawel CFR Code of Federal Regulations CHF Critical Heat Flux CSAU Code Scaling, Applicability and Uncertainty CWO Core-Wide Oxidation ECCS Emergency Core Cooling System ECR Equivalent Cladding Reacted EM Evaluation Model EMDAP Evaluation Model Development and Assessment Process F+
Nuclear Enthalpy Rise Factor/Radial Peaking Factor FQ Total Peaking Factor/Global Peaking Factor Framatome Framatome Inc.
FSRR Fuel Swell Rupture and Relocation Gd2O3 Gadolinia GDC General Design Criteria HHSI High Head Safety Injection HMP High Mechanical Performance IGM Intermediate GAIA Mixing Grid IHSI Intermediate Head Safety Injection k(z)
Axial-Dependent Peaking Factor LBLOCA Large Break Loss-of-Coolant Accident LCO Limiting Condition of Operation LEU Low-Enriched Uranium LHSI Low Head Safety Injection LOCA Loss-of-Coolant Accident For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page vi Acronym Definition LOOP Loss-of-Offsite Power MLO Maximum Local Oxidation No-LOOP No Loss-of-Offsite Power NRC U.S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PWR Pressurized Water Reactor RAI Request for Additional Information RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RLBLOCA Realistic Large Break Loss of Coolant Accident SE Safety Evaluation SG Steam Generator SI Safety Injection SIAS Safety Injection Actuation Signal TS Technical Specification UTL Upper Tolerance Limit For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 1-1 1.0 INTRODUCTION This report summarizes the realistic large break loss-of-coolant accident (RLBLOCA) analysis for Byron Unit 2 and Braidwood Unit 2 (referred to herein as Byron/Braidwood Unit 2). The purpose of the RLBLOCA analysis is to support the fuel transition at Byron/Braidwood Unit 2 to Framatome GAIA fuel. This analysis was performed in accordance with the U.S. Nuclear Regulatory Commission (NRC)-approved S-RELAP5-based methodology described in Reference 1 and supplemented by Reference 2.
Byron/Braidwood Unit 2 is a four-loop, Westinghouse-designed Pressurized Water Reactor (PWR). The Framatome GAIA fuel with M5Framatome cladding design for Byron/Braidwood Unit 2 consists of a 17x17 rod array with GAIA and intermediate GAIA Mixing (IGM) 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 nozzle. The fuel is standard UO2 fuel. The fuel assemblies may contain fuel rods with 2, 4, 6, or 8 weight-percent gadolinia (Gd2O3) as a burnable absorber. The analysis explicitly analyzes fresh and once-burned Framatome GAIA fuel with and without gadolinia.
The analysis assumes full-power operation at a core power level of 3658 MWt (includes measurement uncertainty), a maximum-allowed total peaking factor (FQ) of 2.6 (represents total peaking with uncertainties applied with an axial-dependent factor k(z) set to 1.0), a radial peaking factor (F+) of 1.70 (includes measurement uncertainty),
and up to 10% steam generator (SG) tube plugging per SG.
The analysis also addresses typical operational ranges or technical specification (TS) limits (whichever is applicable) with regard to [ 
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The plant parameter specification for this analysis is provided in Table 4-1.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 1-2 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.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 2-1 2.0 

==SUMMARY==
OF RESULTS The UTL results providing 95/95 simultaneous coverage from this evaluation meet the 10 CFR 50.46(b) criteria with a PCT of 1767&deg;F, a maximum local oxidation of 6.01 percent and a total core-wide oxidation of 0.050 percent. The PCT of 1767&deg;F occurred in a fresh UO2 rod. The results of the analysis demonstrate the adequacy of the ECCS to support the 10 CFR 50.46(b) Criteria 1-3 discussed in Section 3.0.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-1 3.0 DESCRIPTION OF ANALYSIS 3.1 Acceptance Criteria The purpose of the analysis is to verify the adequacy of the Byron/Braidwood Unit 2 ECCS by demonstrating compliance with the following 10 CFR 50.46(b) criteria (Reference 4):
: 1.
The calculated maximum fuel element cladding temperature shall not exceed 2200&deg;F.
: 2.
The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
: 3.
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.
The final two criteria, coolable geometry and long-term cooling, are treated in separate plant-specific evaluations.
Note: The original 17% value in the second acceptance criterion for MLO was based on the usage of the Baker-Just correlation. For present reviews on ECCS Evaluation Model (EM) applications, the NRC staff imposed a limitation specifying that the equivalent cladding reacted (ECR) results calculated using the Cathcart-Pawel (C-P) correlation are considered acceptable in conformance with 10 CFR 50.46(b)(2) if the ECR value is less than 13% (Section 3.3.3, NRC Final Safety Evaluation (SE) for EMF-2103P-A Revision 3, Reference 1). The limitation is addressed in Table 3-1.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-2 3.2 Description of LBLOCA Event A large break loss-of-coolant accident (LBLOCA) is initiated by a postulated rupture of the Reactor Coolant System (RCS) primary piping. The most challenging break location is in the cold leg piping between the reactor coolant pump and the reactor. The plant is assumed to be operating normally at full power prior to the accident and the break is assumed to open instantaneously. 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 pumped injection train without the loss of containment spray.
The LBLOCA event is typically described in three phases: blowdown, refill, and reflood.
Following the initiation of the break, the blowdown phase is characterized by a sudden depressurization from operating pressure down to the saturation pressure of the hot leg fluid. For larger cold leg breaks, an immediate flow reversal and stagnation occurs in the core due to flow out the break, which causes the fuel rods to pass through critical heat flux (CHF), usually within one second following the break. Following this initial rapid depressurization, the RCS depressurizes at a more gradual rate. Reactor trip and emergency injection signals occur when either the low pressure setpoint or the containment high-pressure setpoint are reached. However, for LBLOCA, reactor trip and scram are not modeled, and reactor shutdown is accomplished by moderator reactivity feedback. During blowdown, core cooling is supported by the natural evolution of the RCS flow pattern as driven by the break flow.
When the system pressure falls below the accumulator pressure, flow from the accumulator is injected into the cold legs ending the blowdown period and initiating the refill period. Once the system pressure falls below the respective shutoff heads of the safety injection systems and the system startup time delays are met, flow from the pumped safety injection systems is injected into the RCS. While some of the ECCS flow bypasses the core and goes directly out of the break, the downcomer and lower plenum gradually refill until the mixture in the lower head and lower plenum regions reaches the bottom of the active core and the reflood period begins. Core cooling is supported by the natural evolution of the RCS flow pattern as driven by the break flow and For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-3 condensation in the RCS promoted by safety injection. Towards the end of the refill period, heat transfer from the fuel rods is relatively low, steam cooling and rod-to-rod radiation being the primary mechanisms of core heat removal.
Once the lower plenum is refilled to the bottom of the fuel rod heated length, refill ends and the reflood phase begins. Substantial ECCS fluid is retained in the downcomer during refill. This provides the driving head to move coolant into the core. As the mixture level moves up the core, steam is generated and liquid is entrained, providing cooling in the upper core regions. The two-phase mixture extends into the upper plenum and some liquid may de-entrain and flow downward back into the cooler core regions. The remaining entrained liquid passes into the steam generators where it vaporizes, adding to the steam that must be discharged through the break and out of the system. The difficulty of venting steam is, in general, referred to as steam binding. It acts to impede core reflood rates. With the initiation of reflood, a quench front starts to progress up the core. With the advancement of the quench front, the cooling in the upper regions of the core increases, eventually arresting the rise in fuel rod surface temperatures. Later the core is quenched and a pool cooling process is established that can maintain the cladding temperature near saturation, so long as the ECCS makes up for the core boil off.
3.3 Description of Analytical Models The Framatome RLBLOCA methodology is documented in EMF-2103P-A (Reference 1). The methodology follows the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology (Reference 5) and the requirements of the Evaluation Model Development and Assessment Process (EMDAP) documented in Reference 6.
The CSAU method outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties in a Loss-of-Coolant Accident (LOCA) analysis.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-4 The methodology evaluation model for event response of the primary and secondary systems and the hot fuel rod used in this analysis is based on the use of two computer codes.
x GALILEO for computation of the initial fuel stored energy, fission gas release, and the transient fuel-cladding gap conductance.
x S-RELAP5 for the thermal-hydraulic system calculations (includes ICECON for containment response).
The methodology (Reference 1) has been reviewed and approved by the NRC to perform LBLOCA analyses. The governing two-fluid (plus non-condensable) model with conservation equations for mass, energy, and momentum transfer is used. The reactor core is modeled in S-RELAP5 with heat generation rates determined from reactor kinetics equations (point kinetics) with reactivity feedback, and with actinide and fission product decay heat.
The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. The effects of one phase on the other are accounted for by interfacial friction and heat and mass transfer interaction terms in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ.
The modeling of plant components is performed by following guidelines developed to ensure accurate accounting for physical dimensions and that the dominant phenomena expected during the LBLOCA event are captured. The basic building blocks for modeling are hydraulic volumes for fluid paths and heat structures for heat transfer. In addition, special purpose components exist to represent specific components such as the Reactor Coolant Pumps (RCPs) or the SG separators. All geometries are modeled at the resolution necessary to best resolve the flow field and the phenomena being modeled within practical computational limitations.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-5 The analysis considers blockage effects due to clad swelling and rupture as well as increased heat load due to fuel relocation in the ballooned region of the cladding in the prediction of the hot fuel rod PCT.
A typical calculation using S-RELAP5 begins with the establishment of a steady-state initial condition with all loops intact. The input parameters and initial conditions for this steady-state calculation are chosen to reflect plant technical specifications or to match measured data. Specific parameters are discussed in Section 3.6. Additionally, the GALILEO code provides initial conditions for the S-RELAP5 fuel models.
Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops. The evolution of the transient through blowdown, refill, and reflood is computed continuously using S-RELAP5. Containment pressure is calculated by the ICECON module within S-RELAP5.
A detailed assessment of the S-RELAP5 computer code was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena in a PWR LBLOCA. The final step of the methodology is to combine all the uncertainties related to the code and plant parameters and estimate values for the first three criteria of 10 CFR 50.46(b) with a probability of at least 95 percent with 95 percent confidence.
The steps taken to derive the uncertainty estimate are summarized below:
: 1.
Base Plant Input File Development First, base GALILEO and S-RELAP5 input files for the plant (including the containment input file) are developed. The code input development guidelines documented in Appendix A of Reference 1 are applied to ensure that model nodalization is consistent with the model nodalization used in the code validation.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-6
: 2.
Sampled Case Development The statistical approach requires that many sampled cases be created and processed. For every set of input created each key LOCA parameter is randomly sampled over a range established through code uncertainty assessment or expected operating limits (provided by plant technical specifications or data). Those parameters considered "key LOCA parameters" are listed in Table A-6 of Reference 1. This list includes both parameters related to LOCA phenomena, based on the PIRT provided in Reference 1, and to plant operating parameters. The uncertainty ranges associated with each of the model parameters are provided in Table A-7 of Reference 1.
: 3.
Determination of Adequacy of ECCS The methodology uses a non-parametric statistical approach to determine that the first three criteria of 10 CFR 50.46(b) are met with a probability higher than 95 percent with 95 percent confidence.
3.4 GDC-35 Limiting Condition Determination General Design Criteria (GDC)-35 (Reference 7) requires that a system be designed to provide abundant core cooling with suitable redundancy such that the capability is maintained in either the LOOP or No-LOOP condition. [ 
 ]
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-7 3.5 Overall Statistical Compliance to Criteria 3.6 Plant Description The plant analyzed is the Byron/Braidwood Unit 2, Westinghouse-designed PWR, which has four loops, each with a hot leg, a U-tube steam generator, and a cold leg with an RCP. The RCS includes one pressurizer connected to a hot leg. The ECCS provides injection to each of the four loops via the centrifugal charging (CC)/high head safety injection (HHSI) system, SI/intermediate head safety injection (IHSI) system, residual heat removal (RHR)/low head safety injection (LHSI) system, and accumulators. For the purpose of this report, the CC/HHSI, SI/IHSI, and RHR/LHSI systems are referred to as the HHSI, IHSI, and LHSI systems, respectively. The RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water for ECCS pumped injection does not need to be considered.
The RCS, reactor vessel, pressurizer, and ECCS are explicitly modeled in the S-RELAP5 model. For each RCS loop, the LOCA ECCS model includes an injection connection to the cold leg for the accumulator and another connection for HHSI. IHSI and LHSI are modeled with separate injection connections to each of the four accumulator lines. The accumulator and HHSI injection connections to the cold leg pipe are downstream of the RCP discharge. Also modeled is the secondary-side steam generator that is instantaneously isolated (closed main steam isolation valve and feedwater trip) at the time of the break. The ECCS pumped injection is modeled as a table of flow versus backpressure. A model is included to account for the potential For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-8 presence of nitrogen in the accumulator lines. This model works by delaying the injection of the accumulator line ECCS to the cold leg until an equivalent amount of liquid has been swept out of the accumulator line. The primary and secondary coolant systems for Byron/Braidwood Unit 2 were nodalized to be consistent with code input guidelines in Appendix A of Reference 1. Representative system nodalization details are shown in Figure 3-1 through Figure 3-3.
The results used to demonstrate compliance with the 10 CFR 50.46(b) criteria are only applicable to the Framatome GAIA fuel product. No results or comparisons to criteria apply to Westinghouse fuel. However, because Byron/Braidwood Unit 2 will transition from the Westinghouse resident fuel to Framatome fuel, the thermal-hydraulic effects of the Westinghouse resident fuel (mixed core configuration) were considered in the analysis. The resident fuel assemblies have different form loss coefficients for the grid spacers and upper and lower tie plates than the Framatome fuel. [ 
 ] Therefore, the results are also applicable for the transition from Westinghouse to Framatome fuel, and continued operation with Framatome fuel.
As described in Section 3.3, many parameters associated with LBLOCA phenomenological uncertainties and plant operation ranges are sampled. Values for process or operational parameters, including ranges of sampled process parameters, and fuel design parameters used in this analysis are given in Table 4-1. Table 4-2 presents a summary of the uncertainties used in the analysis. [ 
 ] The passive heat sinks and material properties used in the containment input model are provided in Table 4-3.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-9 3.7 Safety Evaluation Limitations The RLBLOCA analysis for Byron/Braidwood Unit 2 presented herein is consistent with the approved methodology documented in EMF-2103P-A, Revision 3 (Reference 1) and supplemented by ANP-10349P-A, Revision 0 (Reference 2). The limitations and conditions from the NRC SE for EMF-2103P-A, Revision 3, are addressed in Table 3-1.
The limitations and conditions from Reference 3 on GALILEO applications for ANP-10349P-A, Revision 0 (Reference 2) are addressed in Table 3-2. From the responses given in Table 3-1 and Table 3-2, the Byron/Braidwood Unit 2 RLBLOCA analysis documented herein is compliant with all requirements of the applicable topical reports.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-10 Table 3-1 EMF-2103P-A, Revision 3, SE Limitations Evaluation Limitations (Section 4.0 of the SE in Reference 1) 

===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 Unit 2 are 4-loop Westinghouse-designed NSSS PWRs with cold leg ECCS injection.
3 The EM is approved based on models that are specific to AREVA proprietary M5 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.
The analysis was performed using the modeling guidelines contained in Appendix A of EMF-2103P-A, Revision 3.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-11 Limitations (Section 4.0 of the SE in Reference 1) 

===Response===
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
[ 
 ]
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 correlation, this limit shall be reduced to 13 percent, inclusive of pre-transient oxide layer thickness.
For this analysis the MLO UTL is less than 13%.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-12 Limitations (Section 4.0 of the SE in Reference 1) 

===Response===
8 In conjunction with Limitation [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 [ 
]
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 B-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.
[ 
 ]
10 [
]
This analysis uses [ 
 ]
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-13 Limitations (Section 4.0 of the SE in Reference 1) 

===Response===
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.
The present analysis is the first statistical application of EMF-2103P-A, Revision 3 for this plant.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-14 Table 3-2 ANP-10349P-A, Revision 0, Limitations and Conditions on GALILEO Applications Ranges of Applicability (Section 1.2 of the SE in Reference 3) 

===Response===
Pressurized water reactor designs using Low-Enriched Uranium (LEU) fuel loading This analysis was performed for the Byron/Braidwood Unit 2 plant, which is a PWR, 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 burnup values applied in the analysis do not exceed the rod average burnup of [ 
 ]
Zircaloy-4 and M5 cladding The analysis was performed for fuel rods 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%
The Gd2O3 concentration analyzed does not exceed 10 wt%.
Nominal true pellet density ranging from [
 ] percent of the theoretical density of UO2 This analysis was performed using fuel pellets with a density of [ 
 ]
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 pellets considered in this analysis were manufactured using dry conversion and ammonium diuranate.
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.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-15 Figure 3-1 Representative Primary System Noding For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-16 Figure 3-2 Representative Secondary System Noding For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 3-17 Figure 3-3 Representative Reactor Vessel Noding For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-1 4.0 RLBLOCA ANALYSIS 4.1 RLBLOCA Results For a simultaneous coverage/confidence level of 95/95, the UTL values, [ 
 ] are a PCT of 1767&deg;F, an MLO of 6.01 percent, and a CWO of 0.050 percent. The fraction of total hydrogen generated was not directly calculated; however, it is conservatively bounded by the calculated total core wide percent oxidation, which is well below the 1 percent limit.
A summary of the major input parameters for the demonstration case is provided in Table 4-5. The sequence of event times for the demonstration case is provided in Table 4-6. Table 4-7 [ 
 ]
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-2 The analysis scatter plots for the case set are shown in Figure 4-1 through Figure 4-5.
Figure 4-1 shows linear scatter plots of the key parameters sampled for all cases.
These figures illustrate the parameter ranges used in the analysis. Visual examination of the linear scatter plots demonstrates that the spread and coverage of all the values used is appropriate and within the uncertainty ranges listed in Table 4-2. Key results such as the PCT and event timings are also listed for the case set.
Figure 4-2 and Figure 4-3 show PCT scatter plots versus the time of PCT and versus break size, respectively. The scatter plots for the maximum local oxidation and total core-wide oxidation are shown in Figure 4-4 and Figure 4-5, respectively.
Figure 4-2 shows about 10% of cases have PCT occurring during the blowdown phase (PCT time less than approximately 30 seconds). Blowdown PCT cases are dominated by rapid RCS depressurization and high stored energy in the core. The next large grouping of PCTs occurs during the reflood period. The ability to remove decay heat is key for reflood PCT cases. In general, plants with high accumulator pressure tend to inject early in the transient when the break flow is still high. The high pressure and high break flow can drive some of this fluid to bypass the core, delaying the progression of the core reflood. This can result in cases with higher PCTs in the reflood phase of the transient.
The high PCT cases in Figure 4-2 are influenced by the size of the break. This can be seen in Figure 4-3 which shows a general increasing trend in PCT with break size. From this, it also follows that smaller break sizes result in lower PCTs. From all sampled parameters, the break size is a dominant effect on PCT. This is expected since break size heavily influences the residual coolant inventory and the rate of primary system depressurization.
Figure 4-4 shows a general trend in increasing oxidation results with increasing PCT.
Since the MLO includes the pre-transient oxidation, the MLO is not only a function of cladding temperature but also of time in cycle (burnup). The CWO also shows a strong For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-3 correlation to PCT as demonstrated in Figure 4-5, as higher PCT cases would have higher oxidation throughout the core.
The demonstration case is a reflood peak case with a PCT timing of 167.2 seconds.
Figure 4-6 through Figure 4-17 show key parameters from the S-RELAP5 calculations for the demonstration case. The transient progression for the demonstration case follows that described in Section 3.2.
4.2 Conclusions This report describes and provides results from the RLBLOCA analysis for the Byron/Braidwood Unit 2 with the Framatome GAIA fuel design. The application of the methodology involves developing input decks, executing the simulations that comprise the uncertainty analysis, retrieving PCT, MLO, and CWO information, and determining the simultaneous UTL results for the criteria. [ 
 ] The UTL results providing a 95/95 simultaneous coverage/confidence level from this evaluation meet the 10 CFR 50.46(b) criteria with a PCT of 1767&deg;F, an MLO of 6.01 percent, and a CWO of 0.050 percent.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-4 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges Plant Parameter Parameter Value 1.0 Plant Physical Description 1.1 Fuel a) Cladding outside diameter 0.374 in.
b) Cladding inside diameter 0.329 in.
c) Pellet outside diameter 0.3225 in.
d) Initial Pellet density
[ 
]
e) Active fuel length 144 in.
f) Gd2O3 concentrations 2, 4, 6, 8 weight-percent 1.2 RCS a) Flow resistance Analysis b) Pressurizer location
[ 
]
c) Hot assembly location Anywhere in core d) Hot assembly type 17x17 e) SG tube plugging 10 percent 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Analyzed reactor power 3658(1) MWt b) FQ 2.6(1) c) F+
1.70(1) 2.2 Fluid Conditions a) Total Loop flow 137.3 Mlbm/hr d M d 165.3 Mlbm/hr b) RCS average temperature 565qF d T d 598qF c) Upper head temperature 
~RCS Cold Leg Temperature(2) d) Pressurizer pressure 2190 psia d P d 2310 psia e) Pressurizer liquid level 50 percent d L d 70 percent f) Accumulator pressure 602 psia d P d 677 psia g) Accumulator liquid volume 920 ft3 d V d 980 ft3 h) Accumulator temperature 60qF d T d 120qF(3) i) Accumulator resistance fL/D As-built piping configuration j) Minimum accumulator boron 2150 ppm 1 Includes measurement uncertainty.
2 Upper head temperature will change based on sampling of RCS temperature.
3 Coupled with containment temperature.
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-5 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges (Continued)
Plant Parameter Parameter Value 3.0 Accident Boundary Conditions a) Break location Cold leg pump discharge b) Break type Double-ended guillotine or split c) Break size (each side, relative to cold leg pipe area) d) ECCS pumped injection temperature 120&deg;F e) HHSI pump delay 37 s (No-LOOP) and 50 s (LOOP) f) IHSI pump delay 37 s (No-LOOP) and 50 s (LOOP) g) LHSI pump delay 37 s (No-LOOP) and 50 s (LOOP) h) Initial containment pressure 14.22 psia i) Initial containment temperature 60qF d T d 120qF j) Containment sprays delay 0 s k) Containment spray water temperature 32qF l) LHSI Flow RCS Cold Leg Pressure (psia)
Broken Loop Flow (gpm)
Total Intact Loops Flow (gpm) 15.0 1053.0 2460.9 20.0 1036.0 2421.0 40.0 968.0 2259.0 50.0 932.0 2175.0 60.0 893.0 2085.0 80.0 810.0 1890.0 100.0 715.0 1671.9 120.0 607.0 1416.9 156.0 365.0 861.9 157.0 254.0 605.1 180.0 10.0 33.0 182.0 0.0 0.0 For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-6 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges (Continued)
Plant Parameter Parameter Value m) IHSI Flow RCS Cold Leg Pressure (psia)
Broken Loop Flow (gpm)
Total Intact Loops Flow (gpm) 15.0 163.0 402.9 20.0 162.0 402.0 40.0 161.0 399.9 50.0 161.0 398.1 60.0 160.0 396.9 80.0 159.0 393.9 100.0 158.0 392.1 120.0 157.0 390.0 156.0 155.0 384.9 157.0 155.0 384.9 180.0 154.0 381.9 182.0 154.0 381.9 300.0 144.0 356.1 400.0 137.0 338.1 500.0 129.0 320.1 600.0 121.0 300.0 700.0 113.0 279.9 800.0 105.0 258.9 900.0 94.0 233.1 1000.0 83.0 204.0 1100.0 68.0 168.0 1200.0 50.0 123.0 1300.0 19.0 47.1 1339.0 0.0 0.0 For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-7 Table 4-1 RLBLOCA Analysis - Plant Parameter Values and Ranges (Continued)
Plant Parameter Parameter Value n) HHSI Flow RCS Cold Leg Pressure (psia)
Broken Loop Flow (gpm)
Total Intact Loops Flow (gpm) 15.0 114.0 279.0 20.0 114.0 279.0 40.0 114.0 276.0 50.0 114.0 276.0 60.0 113.0 276.0 80.0 113.0 276.0 100.0 113.0 276.0 120.0 112.0 273.0 156.0 111.0 273.0 157.0 111.0 270.0 180.0 111.0 270.0 182.0 111.0 270.0 300.0 111.0 270.0 400.0 108.0 264.0 500.0 105.0 258.0 600.0 102.0 249.0 700.0 99.0 243.0 800.0 96.0 234.0 900.0 93.0 225.0 1000.0 90.0 219.0 1100.0 86.0 210.0 1200.0 83.0 201.0 1300.0 79.0 192.0 1339.0 78.0 189.0 1400.0 67.0 162.0 1500.0 63.0 153.0 1600.0 58.0 141.0 1700.0 53.0 129.0 1800.0 48.0 117.0 1900.0 42.0 102.0 2000.0 36.0 87.0 2100.0 30.0 72.0 2200.0 18.0 42.0 2250.0 10.0 24.0 2282.0 0.0 0.0 For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-8 Table 4-2 Statistical Distribution Used for Process Parameters For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-9 Table 4-3 Passive Heat Sinks and Material Properties in Containment Geometry Heat Sink Surface Area, ft2 Thickness, ft Material 1 (Carbon Steel-0.025ft) 283617.00 0.025 Carbon Steel 2 (Carbon Steel-0.1325ft) 1154.70 0.1325 Carbon Steel 3 (Carbon Steel-0.00714ft) 29719.00 0.00714 Carbon Steel 4 (Carbon Steel-0.0039ft) 20411.00 0.0039 Carbon Steel 5 (Carbon Steel-0.2229ft) 1674.75 0.2229 Carbon Steel 6 (Carbon Steel-0.1156ft) 2013.84 0.1156 Carbon Steel 7 (Carbon Steel-0.04167ft) 42144.90 0.04167 Carbon Steel 8 (Carbon Steel-0.0104ft) 40000.00 0.0104 Carbon Steel 9 (Carbon Steel-0.1708ft) 663.49 0.1708 Carbon Steel 10 (Concrete-1.0ft) 116157.70 0.5 Concrete 0.5 Concrete 11 (Concrete/Steel-1.03362ft) 828.13 0.03362 Carbon Steel 1.0 Concrete 12 (Concrete/Steel-1.02292ft) 1850.20 0.02292 Carbon Steel 1.0 Concrete 13 (Concrete/Steel-1.01563ft) 10134.80 0.01563 Carbon Steel 1.0 Concrete 14 (Concrete/Steel-1.04899ft) 23489.55 0.04899 Carbon Steel 1.0 Concrete 15 (Concrete/Steel-1.15276ft) 3022.63 0.15276 Carbon Steel 1.0 Concrete 16 (Concrete/Steel-4.52083ft) 115872.75 0.02083 Carbon Steel 4.5 Concrete 17 (Steel-0.0039ft) 36114.00 0.0039 Carbon Steel 18 (Steel-0.01042ft) 30000.00 0.01042 Carbon Steel Heat Sink Material Thermal Conductivity Btu/hr-ft-&deg;F Volumetric Heat Capacity Btu/ft3-&deg;F Concrete 0.92 22.6 Carbon Steel 27.0 58.8 Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-10 Table 4-4 Compliance with 10 CFR 50.46(b)
UTL for 95/95 Simultaneous Coverage/Confidence Parameter Value Case Number PCT (&deg;F) 1767
[ ]
MLO (%)
6.01
[ 
 ]
CWO (%)
0.050
[ 
 ]
Characteristics of Case Setting the PCT UTL - the Demonstration Case PCT (&deg;F) 1767 PCT Rod Type Fresh UO2 Rod Time of PCT (s) 167.20 Elevation within Core (ft) 10.45 Local Maximum Oxidation (%)
6.48 Total Core-Wide Oxidation (%)
0.047 PCT Rod Rupture Time (s) 30.24 Rod Rupture Elevation within Core (ft) 9.60 For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-11 Table 4-5 Summary of Major Parameters for the Demonstration Case Parameter Value Core Power (MWt) 3658 Fresh Fuel Time in Cycle (hrs) 12458 Burned Fuel Time in Cycle (hrs) 25181 Fresh Fuel Assembly Avg. Burnup (GWd/mtU) 27.4 Burned Fuel Assembly Avg. Burnup (GWd/mtU) 45.8 Core Peaking Factor, FQ 2.45 Radial Peaking Factor, F+
1.70 Fresh Fuel Axial Offset 0.234 Burned Fuel Axial Offset 0.234 Break Type Split Break Size (ft2/side) 3.068
[ 
]
[ 
]
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-12 Table 4-6 Calculated Event Times for the Demonstration Case Event Time (sec)
Break Opens 0.0 RCP Trip 0.0 SIAS Issued 1.5 Start of Broken Loop Accumulator Injection 13.1 Start of Intact Loop Accumulator Injection
[ 
 ]
15.1, 15.1 and 15.1 Beginning of Core Recovery (Beginning of Reflood) 31.5 Charging/IHSI/LHSI Available 51.5 Broken Loop Charging Delivery Began 51.5 Intact Loop Charging Delivery Began
[
 ]
51.5, 51.5 and 51.5 Broken Loop IHSI Delivery Began 51.5 Intact Loop IHSI Delivery Began
[ 
 ]
51.5, 51.5 and 51.5 Broken Loop LHSI Delivery Began 51.5 Intact Loop LHSI Delivery Began
[
]
51.5, 51.5 and 51.5 Intact Loop Accumulator Emptied
[ 
]
53.5, 52.2 and 52.4 Broken Loop Accumulator Emptied 54.3 PCT Occurred 167.2 Transient Calculation Terminated 1165.8 Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-13 Table 4-7 Fuel Rod Rupture Ranges of Parameters For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-14 Figure 4-1 Scatter Plot Key Parameters For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-15 Figure 4-1 Scatter Plot Key Parameters (continued)
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-16 Figure 4-2 PCT versus PCT Time Scatter Plot For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-17 Figure 4-3 PCT versus Break Size Scatter Plot For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-18 Figure 4-4 MLO vs PCT Scatter Plot For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-19 Figure 4-5 CWO vs PCT Scatter Plot For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-20 Figure 4-6 Peak Cladding Temperature (Independent of Elevation)
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-21 Figure 4-7 Break Flow For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-22 Figure 4-8 Core Inlet Mass Flux For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-23 Figure 4-9 Core Outlet Mass Flux For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-24 Figure 4-10 Void Fraction at RCS Pumps For Information Only

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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-25 Figure 4-11 ECCS Flows (Includes Accumulator, HHSI, IHSI and LHSI)
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ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-26 Figure 4-12 Upper Plenum Pressure For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-27 Figure 4-13 Collapsed Liquid Level in the Downcomer For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-28 Figure 4-14 Collapsed Liquid Level in the Lower Plenum For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-29 Figure 4-15 Core Collapsed Liquid Level For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-30 Figure 4-16 Containment and Loop Pressures For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-31 Figure 4-17 Pressure Differences between Upper Plenum and Downcomer For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 4-32 Figure 4-18
[ 
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For Information Only

Framatome Inc.
ANP-4116NP Revision 0 Byron/Braidwood Unit 2 Large Break LOCA Analysis Licensing Report Page 5-1 5.0 REFERENCES
: 1. EMF-2103P-A, Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016.
: 2. ANP-10349P-A, Revision 0, GALILEO Implementation in LOCA Methods, November 2021.
: 3. ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, November 2020.
: 4. Code of Federal Regulations, Title 10, Part 50, Section 46, Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors, August 2007.
: 5. NUREG/CR-5249, Quantifying Reactor Safety Margins, Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large Break, Loss-of-Coolant Accident, U.S. NRC, December 1989.
: 6. Regulatory Guide 1.203, Transient and Accident Analysis Methods, U.S. NRC, December 2005.
: 7. Code of Federal Regulations, Title 10, Appendix A to Part 50, General Design Criteria for Nuclear Power Plants, August 2007.
For Information Only

ATTACHMENT 4 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 Affidavit Constellation 

AFFIDAVIT
: 1. My name is Rebecca L. Steinman. I am Senior Manager, Licensing for Constellation Energy Generation, LLC (CEG) and as such I am authorized to execute this Affidavit.
: 2. I am familiar with the criteria applied by CEG to determine whether certain CEG information is proprietary. I am familiar with the policies established by CEG to ensure the proper application of these criteria.
: 3. I am familiar with the CEG information contained in Attachment 10 and 11 to CEG letter RS-25-088 dated April 21, 2025, with 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" and referred to herein as "Document." Information contained in this Document has been classified by CEG as proprietary in accordance with the policies established by CEG for the control and protection of proprietary and confidential information.
: 4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by CEG and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
: 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information. "
: 6. The following criteria are customarily applied by CEG to determine whether information should be classified as proprietary:
(a) The information reveals details of CEGs research and development plans and programs or their results.
(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for CEG.
(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for CEG in product optimization or marketability.
(e) The information is vital to a competitive advantage held by CEG, would be helpful to competitors to CEG, and would likely cause substantial harm to the competitive position of CEG. The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d), and 6(e) above.
: 7. In accordance with CEGs policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside CEG only as required and under suitable agreement providing for nondisclosure and limited use of the information.
: 8. CEG policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
: 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on: (4/21/25)
Rebecca.Steinman@Constellation.com 4300 Winfield Road Warrenville, IL 60555 

ATTACHMENT 5 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 Affidavit Framatome 

A F F I D A V I T
: 1.
My name is Morris Byram. I am Manager, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.
: 2.
I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.
: 3.
I am familiar with the Framatome information contained in Attachments 8 and 9 to Constellation letter RS-25-088, with 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, and referred to herein as Documents. Information contained in these Documents has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.
: 4.
These Documents contain information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in these Documents as proprietary and confidential.
: 5.
These Documents have been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is 

requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.
: 6.
The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:
(a)
The information reveals details of Framatomes research and development plans and programs or their results.
(b)
Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.
(c)
The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.
(d)
The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.
(e)
The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.
The information in these Documents is considered proprietary for the reasons set forth in paragraph 6(c), 6(d), and 6(e) above.
: 7.
In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.
: 8.
Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
: 9.
The foregoing statements are true and correct to the best of my knowledge, information, and belief.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on: (4/18/2025)
(NAME) morris.byram@framatome.com 2101 Horn Rapids Road Richland, WA 99354

ATTACHMENT 6 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 Affidavit Westinghouse 

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-25-019 Page 1 of 3 Commonwealth of Pennsylvania:
County of Butler:
(1)
I, Jerrod Ewing, Manager, Operating Plants Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).
(2)
I am requesting the proprietary portions of RS-24-088, Revision 0, be withheld from public disclosure under 10 CFR 2.390.
(3)
I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.
(4)
Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.
(ii)
The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.
(iii)
Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. 

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-25-019 Page 2 of 3 (5)
Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).
(c)
Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
(6)
The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit. 

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-25-019 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.
Executed on: 4/18/2025 Signed electronically by Jerrod Ewing}}