ML25085A297
| ML25085A297 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 03/26/2025 |
| From: | Scott Wall NRC/NRR/DORL/LPL3 |
| To: | Zurawski L Constellation Energy Generation, Constellation Nuclear |
| References | |
| EPID L-2024-LLA-0072 | |
| Download: ML25085A297 (1) | |
Text
From:
Scott Wall To:
Zurawski, Lisa M: (Constellation Nuclear)
Cc:
Gantt, Danii M:(Constellation Nuclear); Seawright, Brian: (Constellation Nuclear); Steinman, Rebecca L:
(Constellation Nuclear)
Subject:
Final RAI - Braidwood and Byron - License Amendment Request to Transition to Framatome GAIA Fuel (EPID No.
Date:
Wednesday, March 26, 2025 11:46:40 AM
Dear Lisa Zurawski,
By letter 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 and2 (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 NRC staff has reviewed the submittal and determined that additional information is needed to complete its review. The specific question is found in the enclosed request for additional information (RAI). On March 25, 2025, the CEG staff indicated that a response to the RAIs would be provided by April 21, 2025.
If you have questions, please contact me at 301-415-2855 or via e mail at Scott.Wall@nrc.gov.
Scott P. Wall Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 301.415.2855 Scott.Wall@nrc.gov
Docket Nos.: STN 50-456, STN 50-457, STN 50-454, and STN 50-455
Enclosure:
Request for Additional Information
cc: Listserv
RAI (Transition to Framatome GAIA Fuel)
REQUEST FOR ADDITIONAL INFORMATION
LICENSE AMENDMENT REQUEST TO TRANSITION TO FRAMATOME GAIA FUEL
CONSTELLATION ENERGY GENERATION, LLC
BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2
DOCKETNOS.STN50456, STN50457, STN50454, AND STN50455
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 and2 (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
AppendixA 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 Fh 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
Provide the numerical Fh penalties to be used on VANTAGE+ fuel during the transition cycles to GAIA fuel at the Byron and Braidwood stations.
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.
Confirm how the penalties will be captured to ensure they will be used during transition cycles.
SNSB-RAI-2
Regulatory Basis: Requirements of GDC10
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, February1983) has the following Limitation and Condition (L&C):
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:
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
Clarify how the statement in the LAR addresses the limitation and condition related to generic or plant-specific DNBR margins offsetting residual DNBR penalties.
SNSB-RAI-3
Regulatory Basis
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.
10 CFR 50 Appendix A, GDC28, 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
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.
SNSB-RAI-4
Regulatory Basis
10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors (b)(1), Peak cladding temperature, states: The calculated maximum fuel element cladding temperature shall not exceed 2200° F.
(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.
(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 0.01times 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 and3.4.2 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
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).
SNSB-RAI-5
Regulatory Basis
10CFR50.46(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.
Issue
Attachments 6 (non-proprietary) and 12 (proprietary) of the LAR address the Braidwood and Byron response to Generic Letter (GL) 200402, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, dated September13,2004 (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 WCAP17788P, Revision1, 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 40gallons 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 WCAP17788P value of 8gpm/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 WCAP17788P. However, Table6-1 of WCAP17788P has a note that states Only the 40 and 18 gpm/FA flows are used for
the Kmax and tblock cases. In this case, the Byron/Braidwood plant specific value of 8.6gpm/FA would be lower than the minimum analyzed value in WCAP17788P of 18gpm/FA.
Request
Provide information that demonstrates the Byron/Braidwood plant specific minimum ECCS flow rate will be larger than the WCAP17788P value of 18gpm/FA.
Radiation Protection and Consequence (ARCB) Questions
ARCB-RAI-1
By letter dated September 8, 2006 (ML062340420), the NRC approved amendments that fully implemented an alternative source term (AST), pursuant to 10CFR50.67, at Braidwood and Byron. For that review, the licensee docketed the calculation of control room (CR) unfiltered in-leakage as 1,000cubic feet per minute (cfm). By letter dated February5,2009, the NRC approved amendments that changed the licensing basis for Braidwood and Byron associated with the application of an AST methodology. For the February5,2009, the licensee reduced the calculation of CR unfiltered in-leakage to 500cfm with the conclusion that The new assumption of 500cfm of unfiltered in-leakage bounds the maximum measured in-leakage value of 68standard cubic feet per minute (scfm) for Byron and 29.3scfm for Braidwood. In the May28,2024, application, CEG further reduces CR unfiltered in-leakage to 436cfm.
Provide the results of the most recent measured CR unfiltered in-leakage values.
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 30minutes. In the May28,2024, application, CEG reduces this value to 20minutes. Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, dated June2003 (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.
Provide information that demonstrates the reduced isolation time critical action meets the guidance.