RS-22-040, Response to Request for Additional Information Related to the License Amendment Request to Transition to GNF3 Fuel

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Response to Request for Additional Information Related to the License Amendment Request to Transition to GNF3 Fuel
ML22101A146
Person / Time
Site: Quad Cities  Constellation icon.png
Issue date: 04/11/2022
From: Simpson P
Constellation Energy Generation
To:
Document Control Desk, Office of Nuclear Reactor Regulation
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ML22101A145 List:
References
RS-22-040
Download: ML22101A146 (34)
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4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 RS-22-040 10 CFR 50.90 April 11, 2022 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Response to Request for Additional Information Related to the License Amendment Request to Transition to GNF3 Fuel

References:

1. Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Request for Licensing Amendment Regarding Transition to GNF3 Fuel," dated September 14, 2021 (ML21257A419)
2. Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Supplemental Request for Information Related to Request for Licensing Amendment Regarding Transition to GNF3 Fuel," dated November 3, 2021 (ML21307A444)
3. Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information for the GNF3 Fuel Transition License Amendment Request", dated January 11, 2022 (ML22011A319)
4. Email from R. Kuntz (U.S. Nuclear Regulatory Commission) to R. Steinman (Constellation Energy Generation, LLC),

Subject:

Request for Additional Information Related to the License Amendment Request to Transition to GNF3 Fuel, dated March 15, 2022 (ML22076A112)

In the Reference 1 letter, Exelon Generation Company, LLC, (EGC) requested an amendment to Renewed Facility Operating License Nos. DPR-29 for Quad Cities Nuclear Power Station (QCNPS), Unit 1 and DPR-30 for QCNPS, Unit 2. Subsequently, the QCNPS facility operating licenses were transferred to Constellation Energy Generation, LLC (CEG). The proposed Attachments 6 and 7 contain Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachments 6 and 7, this document is decontrolled.

April 11, 2022 U.S. Nuclear Regulatory Commission Page 2 changes support the transition from Framatome (formerly AREVA) ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF-A) GNF3 fuel at QCNPS. Additional information supporting the license amendment request were submitted by References 2 and 3.

The following attachments are included in support of CEG's response to the NRC request for additional information (RAI) in Reference 4:

1. Response to NRC Request for Additional Information (Non-Proprietary Version)
2. Marked-up QCNPS Units 1 and 2, Technical Specification Pages
3. Retyped QCNPS Units 1 and 2 Technical Specifications Pages
4. GNF-A 10 CFR 2.390 Affidavit for Withholding
5. Framatome 10 CFR 2.390 Affidavit for Withholding
6. Response to NRC Request for Additional Information (Proprietary Version)
7. Tabular Summary of GEXL98 Data Supporting SFNB-RAI-2 (Attachment 1 to GNF Report 007N0608P) (Proprietary Version)

Attachments 6 and 7 contain information proprietary to GNF and/or Framatome. As a result, these documents are supported by signed affidavits from the owners of the information, which are included as Attachments 4 and 5, respectively. Each affidavit sets forth the basis on which the corporation's information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information, which is proprietary to GNF and Framatome be withheld from public disclosure.

A redacted non-proprietary version of the RAI responses provided in Attachment 6 is provided as Attachment 1. The data table in Attachment 7, which supports one of the RAI responses, is proprietary in its entirety. A redacted version of this table would not be meaningful; therefore, no redacted version is provided.

CEG has reviewed the information supporting the finding of no significant hazards consideration, and the environmental consideration that were previously provided to the NRC in Reference 1.

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 contained within this letter.

Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at (630) 657-2831.

April 11, 2022 U.S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 11th day of April 2022.

Respectfully, Patrick R. Simpson Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:

1. Response to NRC Request for Additional Information (Non-Proprietary Version)
2. Marked-up QCNPS Units 1 and 2 Technical Specification Pages
3. Retyped QCNPS Units 1 and 2 Technical Specifications Pages
4. GNF-A 10 CFR 2.390 Affidavit for Withholding
5. Framatome 10 CFR 2.390 Affidavit for Withholding
6. Response to NRC Request for Additional Information (Proprietary Version)
7. Tabular Summary of GEXL98 Data Supporting SFNB-RAI-2 (Attachment 1 to GNF Report 007N0608P) (138 pages) (Proprietary Version) cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector, Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Response to NRC Request for Additional Information Non-Proprietary Version In this attachment, Framatome information is indicated by the use of single brackets (e.g., [ ]) as described in and GNF-A information is indicated using double brackets (e.g., (( ))) as described in Attachment 4.

ATTACHMENT 1 Response to NRC Request for Additional Information INTRODUCTION By letter dated September 14, 2021, Exelon Generation, licensee at the time, submitted a request to amend the Quad Cities Nuclear Power Station, Units 1 and 2 (Quad Cities) Facility Operating Licenses as necessary to utilize the Global Nuclear Fuels (GNF) GNF-3 fuel design (Agencywide Document Access and Management System (ADAMS) Accession No. ML21257A419). On February 1, 2022 (ADAMS Accession No. ML22032A333), Exelon Generation Company, LLC was renamed Constellation Energy Generation, LLC.

SFNB GEXL RAIs BACKGROUND Part of the amendment included is the review and approval of the report NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel." The report summarizes the development of the ATRIUM 10XM GEXL98 correlation. The ATRIUM 10XM GEXL98 correlation will be used to determine the critical power performance of the Framatome (formerly AREVA) ATRIUM 10XM fuel in a mixed core of ATRIUM 10XM and GNF fuel. This report describes the process used in the development of the GEXL98 correlation for prediction of critical power for the ATRIUM 10XM fuel and the determination of the overall uncertainty of that correlation in prediction of the ATRIUM 10XM critical power performance.

APPLICABLE REGULATORY REQUIREMENTS GDC 10 in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, is the principal regulation associated with this report. This criterion introduces the concept of specified acceptable fuel design limits (SAFDLs). In essence, SAFDLs are those limits placed on certain variables to ensure that the fuel does not fail. One such SAFDL is associated with critical power performance. Because the decrease in heat transfer following critical power could result in fuel failure, a SAFDL is used to demonstrate that critical power does not occur during normal operation and anticipated operational occurrences (AOOs). Therefore, fuel failure is precluded during normal operation and AOOs.

SRP Section 4.4 includes the SAFDLs used in accounting for the uncertainties involved in developing and predicating critical power performance model and ensuring that fuel failure is precluded: "At least 99.9 percent of the fuel rods in the core will not experience a critical power during normal operation or AOOs."

In order to determine that the GEXL98 correlation for ATRIUM 10XM can satisfy the associated SAFDL, the NRC staff requests the following information.

SFNB-RAI-2 For the data submitted, identify which of the data points were used for training and which of the data points were used for validation.

Page 1 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information Constellation Response to SFNB-RAI-2 The requested data has been organized into a table by GNF-A, the owner of the proprietary information. Attachment 7 contains the following information:

Case: unique identifier number (#), and for each:

Database identifier for data usage:

1 - Data point used for training Data point used for validation The unique case number used in Attachment 7 corresponds with the similarly named column of data previously provided in Attachment 4 of Constellation letter RS-22-005 dated January 11, 2022 (ML22011A320).

SFNB-RAI-3 Provide a summary of the random subspace analysis performed which demonstrated that the GEXL98 correlation was relativity insensitive to the training data chosen.

Constellation Response to SFNB-RAI-3 The following table describes the distribution and statistics for multiple sets of random subspaces of the total statistical database (all full-length rods (FLRs)) generated by the final GEXL98 coefficients. The statistics are relatively insensitive to the random subspace, including the smaller sized, complementary sets.

Table SFNB-RAI-3.1: GEXL98 Experimental Critical Power Ratio (ECPR) Statistics by Random Subspace and its Complement (RS and RSC)

Set Cases* Distribution Mean Std. Dev. Maximum Minimum RS1 ((

RS2 RS3 DDB** ))

Set Cases* Distribution Mean Std. Dev. Maximum Minimum RS1C ((

RS2 C RS3 C VDB*** ))

All FLRs (( ))

(( ))

    • DDB (Development Database used to develop the final GEXL98 correlation coefficients
      • VDB (Verification Database, DDBC) of final GEXL98 correlation coefficients Page 2 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information SFNB-RAI-4 Demonstrate that the application domain of GEXL98 is bounded by the application domain of ACE/ATRIUM-10 XM. Specifically address the (( )) limit.

Constellation Response to SFNB-RAI-4 For development of the GEXL98 critical power correlation, GNF requested critical power data for ATRIUM 10XM fuel using the Framatome ACE/ATRIUM 10XM correlation at numerous specific [ ] conditions. The critical power data provided by Framatome were calculated consistent with an approved methodology, as documented in ANP-10298P, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 1 (Reference 1). The data for various conditions, [ ], are explicitly bounded by the upper and lower values of the domain range in Table 2-1 of Reference 1.

Specifically, for the [

]

Framatome followed its approved methodology to calculate the critical power data for GNFs use in developing the GEXL98 correlation. The Framatome data for the GEXL98 development was provided to GNF through transmittal of design information. The critical power(s) at dryout determined by the Framatome ACE/ATRIUM 10XM critical power correlation were based on specified boundary conditions. GEXL98 is not based on any information extrapolated or otherwise inferred from the requested ACE/ATRIUM 10XM critical power data set. As such, the GEXL98 correlation is consistent with the application domain of the approved ACE/ATRIUM 10XM correlation methodology.

SFNB-RAI-5

((

)) Justify the use of GEXL98 correlation's uncertainty in these subregions as it seems that uncertainty in these subregions should be higher.

Page 3 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information Constellation Response to SFNB-RAI-5 The statistics of experimental CPR (ECPR), the ratio of GEXL calculated to ACE calculated critical power, for subregions of interest are presented in Tables SFNB-RAI-5.1 and SFNB-RAI-5.2.

((

))

Both Tables SFNB-RAI-5.1 and SNFB-RAI-5.2 present statistics for individual power shapes as well as all power shapes:

B: Bottom-Peaked C: Cosine-Peaked T: Top-Peaked D: Double-Humped The assessment of the ECPR statistics for each subregion is provided below:

((

Page 4 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information

))

Bold font in Tables SNFB-RAI-5.1 and SNFB-RAI-5.2 indicates greater than the GEXL98 overall mean ECPR + one standard deviation.

Page 5 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information Table SNFB-RAI-5.1: ECPR Statistics at (( ))

Flow, All Axial Power Shape kg/s Shapes B C T D Mean StDev ((

Count Mean StDev Count Mean StDev Count ))

Table SNFB-RAI-5.2: ECPR Statistics at ((

))

Flow, All Axial Power Shape kg/s Shapes B C T D Mean StDev ((

Count Mean StDev Count Mean StDev Count Mean StDev Count ))

Page 6 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information

((

))

Figure SFNB-RAI-5.1. Typical Axial Power Shapes at EOC SFNB-RAI-6 Will the GEXL98 correlation be implemented in computer codes other than the one used to perform the validation analysis? If so, provide the criteria which will be used to ensure an appropriate implementation.

Constellation Response to SFNB-RAI-6 A computer program module, GEXLM02, was used in the validation analysis of the GEXL98 correlation. The same module is called in steady-state three-dimensional (3-D) core simulator and core monitoring code.

Three computer codes, ISCOR, TASC, and TRACG, are identified as using their own respective routine to calculate critical power (CP) or critical power ratio (CPR) of BWR fuels rather than calling the GEXLM02.

First, it should be noted that there is no separate implementation of the GEXL98 correlation in these computer codes. The code specific routine using the approved form per GESTAR II (Reference 2) was already implemented in each code. Information specific to GEXL98 such as correlation coefficients and geometric data are provided as input to the GEXLM02 or other code specific routines. There is no practical difference in the implementation of the GEXL critical power calculation between the GEXLM02 and other routines. As discussed in Section 4 of Page 7 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information NEDC-33930P (Reference 6), the critical quality versus boiling length correlation uses bundle average quantities to calculate the axial quality distributions, both critical quality and local equilibrium quality. The bundle average quantities are uniquely determined using mass and energy balance, GEXL input, and material properties.

((

)) ISCOR is a steady-state code to provide thermal hydraulic information of the core. Models and methods in ISCOR are consistent with those described in GESTAR II (Reference 2). As discussed earlier, the steady-state critical power calculation in ISCOR is the same as that in the GEXLM02.

TASC is used to predict the change in CPR (CPR) over initial CPR (ICPR) or CPR/ICPR for QCNPS Units 1 and 2 anticipated operational occurrence (AOO) analysis. The use of TASC for AOO analysis was reviewed and accepted by the NRC, as documented in Reference 3.

TRACG is used to perform stability analysis of QCNPS Units 1 and 2. SLMCPR protection calculations for the long-term stability solutions Option 1-D, II and III rely on the DIVOM (Delta CPR Over Initial MCPR Versus Oscillation Magnitude) curve as established in Reference 4.

QCNPS Units 1 and 2 currently use the Option III solution. The use of the latest TRACG code was reviewed and accepted by the NRC, as documented in Reference 5.

In summary, the implementations of the generic GEXL critical power calculation within steady-state analysis codes are identical because of the use of bundle average quantities and not using subchannel thermal hydraulic information in the GEXL correlation. The generic GEXL application to CPR/ICPR for plant AOO analysis and DIVOM for plant stability analysis were reviewed and accepted by the NRC.

SFNB-RAI-7 BACKGROUND Section 3.3, "Core Inventory Update and Resulting Dose Consequences," of Attachment 1 to the September 14, 2021, submittal, states that the core inventory at reactor shutdown with no isotopic decay assumed, was calculated for use in the post-loss-of-coolant accident (post LOCA) and control rod drop accident (CRDA) dose analyses. For the post-CRDA dose consequence analysis, consistent with section S.2.2.3.1.4 of General Electric Licensing Topical Report NEDE-24011P-A, "General Electric Standard Application for Reactor Fuel" (GESTAR-II),

US Supplement, the number of fuel rods that would reach 170 calories per gram (cal/gm) is used.

The 170 cal/gm acceptance criterion is understood to refer to a threshold, above which fuel rods are assumed to fail, as predicted by a transient simulation of the postulated CRDA. This estimated number of failed fuel rods is used to determine the fuel inventory released to the coolant.

Page 8 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information APPLICABLE REGULATORY REQUIREMENTS The regulatory requirements applicable to the CRDA include General Design Criteria (GDCs) 13 and 28 contained in Appendix A, "General Design Criteria for Nuclear Power Plants," to part 50, "Domestic Licensing of Production and Utilization Facilities," of Title 10, "Energy" of the Code of Federal Regulations (10 CFR 50 Appendix A). However, Quad Cities Units were licensed prior to the promulgation of these criteria and hence meet the intent of the draft Principal Design Criteria published in 1967 by the Atomic Energy Commission, as discussed in Chapter 3 of the Quad Cities Updated Final Safety Analysis Report.

The analogous criteria to GDC 13, "Instrumentation and Control," within the Quad Cities Licensing Basis are Criterion 12, "Instrumentation and Control Systems," Criterion 13, "Fission Process Monitors and Controls," and Criterion 14, "Core Protection Systems." These criteria require the availability of instrumentation to monitor variables and systems over their anticipated ranges to assure safety, and of appropriate controls to maintain these variables and systems within prescribed operating ranges. These criteria apply because the sequence of events associated with the CRDA includes automatic actuations of protection systems, and potentially manual actions, and the sequence of events must be justified based on the expected values of the relevant monitored parameters and instrument indications.

The analogous criterion to GDC 28, "Reactivity Limits" within the Quad Cities Licensing Basis is Criterion 32, "Maximum Reactivity Worth of Control Rods." This criterion requires that the potential effects of a sudden or large change of reactivity, such as a dropped control rod, cannot (a) rupture the reactor coolant boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling. This criterion applies because (1) the transient fuel enthalpy must be assessed for whether cladding failure or fuel rupture is predicted and to what extent; (2) the coolability of the core following the event must be established; and (3) the maximum reactor coolant pressure must be predicted to demonstrate stress limits for the reactor pressure vessel are not exceeded. These requirements provide assurance that fuel damage and reactor vessel pressure will not be excessive in the CRDA.

Finally, 10 CFR 50.67, "Accident Source Term," also applies, insofar as it establishes radiation dose limits for individuals at the boundary of the exclusion area and at the outer boundary of the low population zone. The fission product inventory released from all failed fuel rods is an input to the radiological evaluation; hence, an underprediction of the number of failed fuel rods following a postulated CRDA could lead to a nonconservative estimation of the post-CRDA dose, leading to inadequate assurance that the requirements of 10 CFR 50.67 are met.

ISSUE The amendment proposes to transition from cold-worked, stress-relieved (CWSR) Zircaloy-2 cladding associated with the ATRIUM10 fuel design, to recrystallized, annealed (RXA)

Zircaloy 2 cladding associated with the GNF-3 fuel design. The RXA cladding is more susceptible to fuel failure due to pellet-cladding mechanical interaction (PCMI) than the resident CWSR fuel cladding. In addition, given the new basis for the post-CRDA consequences, it is not clear how this basis addresses high-temperature cladding failure, molten fuel cladding failure, the magnitude of the predicted pressure surge, or core coolability. Such information is Page 9 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information required to determine whether the CRDA analysis used to determine input to the post-CRDA radiological consequence analysis is consistent with the Quad Cities licensing basis design criteria identified above.

REQUEST Demonstrate that the number of post-CRDA failed fuel rods discussed in the September 14, 2021, submittal accounts for the following three fuel damage mechanisms: (1) PCMI; (2) high temperature cladding failure; and (3) molten fuel cladding failure. Additionally, demonstrate that core coolability is maintained, and the reactor coolant pressure boundary remains intact, following this postulated event. In demonstrating the above, provide justification for any acceptance criteria, e.g., deposited fuel enthalpy, selected.

Constellation Response to SFNB-RAI-7 The method for determining the number of damaged rods during a CRDA at QCNPS is in accordance with §S.2.2.3.1.1 to S.2.2.3.1.4 of Reference 2 which is based on a 170 cal/gm cladding failure threshold established in Reference 8 from 1960s-era data. Additionally, Reference 2 addresses the impact of PCMI in Sections 4.2.1.2.5 and 4.2.1.2.6, molten fuel in Section 4.2.1.2.4, and high temperature cladding in Section 4.2.1.2.3. However, these descriptions are not based on the more modern data and recommendations from the 2015 NRC reviews of this threshold. This Reference 2 methodology is the current licensing basis at QCNPS.

The enthalpy threshold currently used a QCNPS does not have the research data background documentation to rigorously and explicitly address each of the proposed failure mechanisms identified in the 2015 NRC recommendations; however, the overall rod damage method combined with the conservatisms included in the dose calculation ensure the resultant CDRA dose consequences adequately compensate for uncertainties in the number of failed rods.

Conservatisms in the fuel damage assessment are discussed below; there are many other distinct conservatisms within the CRDA calculation that affect final tabulated doses, however they are not directly related to fuel damage assessment.

In transitioning to the new GNF3 fuel type, a fuel damage comparison was performed to legacy fuel types such as the Optima2 and ATRIUM 10XM fuel. The GNF3 analysis showed bounding results with (( )) rods damaged in a CRDA compared to zero or 546 rods for Optima2 and ATRIUM 10XM, respectively. Note that the Optima2 damaged fuel rod number is provided for comparison purposes, but there will be no Optima2 remaining in the core at the time GNF3 fuel is introduced. The dose consequence analysis assumes the entire core is composed of GNF3 fuel in order to maximize the number of failed rods since explicit modeling of a mixed core would reduce the core damage fraction and hence overall dose consequences of a CRDA event.

The CRDA failed rods calculation is only used as an input to the CRDA Alternative Source Term dose analysis. Sensitivity studies were performed using this calculation to determine susceptibility of the resultant dose consequences to changes in rod failure assumptions. The results of those sensitivity studies indicate that an additional 955 rods (above the (( ))

Page 10 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information failed rods currently assumed) could fail before the dose results reach the Regulatory Guide 1.183 / Standard Review Plan (SRP) 15.0.1 acceptance criteria. Similarly, an additional 4470 rods could fail before the dose results approach the limits in 10 CFR 50.67.

The bundles with the highest burnup and fission product inventory are by necessity third-cycle bundles, and will be located at peripheral low-power locations. Although these outer bundles would actually operate at a lower power, for the purpose of the CRDA dose analysis full power (with additional power measurement uncertainty) is assumed at the time of the accident and a 1.70 radial peaking factor multiplier is also applied. This assumption conservatively models the lowest power region of the core as its limiting highest power area.

Newer methodologies that explicitly address the postulated failure mechanisms have shown a significant reduction in the number of failed rods, typically demonstrating zero failed rods for similar startup sequences (References 2 and 10). Based on these experiences at other sites it would be expected that QCNPS would also see a comparable reduction in the number of failed fuel rods under a methodology that includes more explicit analysis of PCMI, high PCT, and fuel melting criteria. Therefore, it can be reasonably concluded that the legacy failed rods calculational method and thresholds are conservative relative to more modern methods when it comes to the calculation of the CRDA dose consequences.

Regarding the reactor coolant pressure boundary integrity, it has been shown that CRDA is not the pressurization limiting event (Reference 10 Sections 15.4.10.3.2 & 5.2 and Reference 2).

More limiting pressurization events (e.g. MSIV fast closure) continue to be analyzed on a cycle-specific basis to ensure adherence to American Society of Mechanical Engineers (ASME) pressure vessel criteria. CRDA overpressurization discussion in References 2 and 12 shows approximately 15 psi increase, significantly below the psi required to challenge vessel integrity.

Therefore, margin exists between the CRDA peak pressure and ASME pressure limitations, ensuring integrity of the overall QCNPS pressure boundary.

Core coolability is maintained so long as there are not gross geometry changes of fuel. A conservative measure of gross geometry change is documented in GESTAR II (Reference 2),

where maintaining fuel enthalpy below 280 cal/gm ensures no violent expulsion of fuel into an uncoolable geometry. This 280 cal/gm value is based on the design limit for BWR fuel and is verified experimentally in Reference 8, again based on 1960s-era techniques. However there exists significant margin to this limit based on the Reference 2 statement that in "all cases the peak fuel enthalpy from a CRDA would be much less than the 280 cal/gm limit".

In summary, while the specific cladding failure threshold of 170 cal/gm cannot be readily shown bounding against more modern criteria in Reference 9, the QCNPS CRDA dose consequence analysis in aggregate is highly conservative and sufficient to offset the uncertainty in the fuel failure threshold. As a result, it is concluded that there is no credible means by which the uncertainty in the fuel failure criteria could result in a dose consequence that increases risk to the health and safety of the public or station personnel greater than that associated with the current dose limits.

Page 11 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information STSB-RAI-1 BACKGROUND TS 5.6.5, "Core Operating Limits Report (COLR)," provides the list of approved methods to be used in determining the core operating limits. The transition to GNF3 fuel requires addition of a new topical report, NEDC-33930P, "GEXL98 Correlation for ATRIUM 10XM Fuel", to TS 5.6.5.b to able to document the GNF correlation applicable to the ATRIUM 10XM fuel for use in the previously approved GESTAR II methodologies for performing licensing analysis.

Quad Cities TS 5.6.5, "Core Operating Limits Report (COLR)," currently requires, "The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents."

APPLICABLE REGULATORY REQUIREMENTS 10 CFR 50.36(c)(5), "Administrative controls," requires that provisions relating to organization and management, procedures, recording keeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner must be included in a licensee's TS.

ISSUE Standard Technical Specifications - General Electric BWR/4 Plants NUREG-1433, Vol. 1, Rev. 5, "Standard Technical Specifications General Electric BWR/4Plants: Specifications",

Section 5.6.3, "Core Operating Limits Report" Reviewer's Note states, "Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date."

REQUEST As described above, the amendment request proposed a change to the Quad Cities, TS 5.6.5.b to add a COLR reference to Report 006N8642-P. The guidance provided in NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications,"

recommends that the staff Safety Evaluation (SE) for a plant-specific methodology by NRC letter and date be included in the citation of plant-specific methodology. This formatting would also be consistent with TS 5.6.5 and the guidance provided in NUREG-1433, Revision 5, "Standard Technical Specifications General Electric BWR/4 plants." Therefore, describe how the proposed TS change would ensure that the NRC staff SE is considered in COLR revisions or revise the citation similar to the following (as underlined):

006N8642-P, Revision 1, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels," January 2022, as approved by NRC Staff SE dated XXX XX, 20XX.

Page 12 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information Constellation Response to STSB-RAI-1 The proposed change to Technical Specification (TS) 5.6.5.b associated with the GNF3 transition LAR (Reference 7) is to add a COLR reference to GNF Report NEDC-33930-P, not report 006N8642-P as noted in the NRC request for information. A revised mark-up of the proposed Technical Specification 5.6.5.b citation for GNF report NEDC-33930-P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel, dated February 2021 is provided in Attachment 2.

A corresponding clean version is provided in Attachment 3. These versions of the TS mark-up and clean pages supersede the previous versions provided in ML21307A444 (Constellation letter RS-21-113 dated November 3, 2021) in their entirety.

References

1. Framatome Licensing Topical Report ANP-10298P-A, "ACE/ATRIUM 10XM Critical Power Correlation," Revision 1, dated March 2014 (ADAMS Accession Number ML14183A734 for the non-proprietary version)
2. GE Licensing Topical Report NEDE-24011-P-A-31, "General Electric Standard Application for Reactor Fuel (GESTAR II, Main)," Revision 31, dated November 2020 (ADAMS Accession Number ML20330A199 for the non-proprietary version)
3. GE Licensing Topical Report NEDC-32084P-A, "TASC-03A, A Computer Program for Transient Analysis of a Single Channel," Revision 2, dated July 2002 (ADAMS Accession Number ML100220484 for the non-proprietary version)
4. GE Licensing Topical Report NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," dated August 1996 (ADAMS Accession Number ML14093A210 for the non-proprietary version)
5. GE Licensing Topical Report Supplement NEDE-32465 Supplement 1P-A, "Migration to TRACG04/PANAC11 from TRACG02/PANAC10 for Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," Revision 1, dated October 2014 (ADAMS Accession Number ML14247A244 for the non-proprietary version)
6. GNF Report NEDO-33930P, "GEXL98 Correlation for ATRIUM 10XM Fuel," Revision 0, dated February 2021 (ADAMS Accession Number ML21336A420 for the non-proprietary version)
7. Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Request for Licensing Amendment Regarding Transition to GNF3 Fuel,"

dated September 14, 2021 (ADAMS Accession Number ML21257A419)

8. GNF Report NEDO-10527, "Rod Drop Accident Analysis for Large Boiling Water Reactors," Revision 0, dated March 1972 (ADAMS Accession Number ML10870249)
9. NRC Memo from P. Clifford to T.J. McGinty,

Subject:

Technical and Regulatory Basis For The Reactivity-Initiated Accident Acceptance Criteria and Guidance, Revision 1, dated March 16, 2015 (ADAMS Accession Number ML14188C423)

Page 13 of 14

ATTACHMENT 1 Response to NRC Request for Additional Information

10. Letter from D. Murray (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Application to Incorporate Licensing Topical Report NEDE-33885P-A, Revision 1, 'GNF CRDA Application Methodology'," dated February 10, 2021 (ADAMS Accession Number ML21041A490)
11. Quad Cities Nuclear Power Station Updated Final Safety Analysis Report (UFSAR),

Revision 16 (Enclosed with ADAMS Accession Number ML21298A208)

12. Topical Report Evaluation of GNF Report NEDO-21778-A, "Transient Pressure Rises Affecting Fracture Toughness Requirements for Boiling Water Reactors," dated January 1978 (ADAMS Accession Number ML20197D187)

Page 14 of 14

ATTACHMENT 2 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 MARKED-UP QCNPS UNITS 1 AND 2 TECHNICAL SPECIFICATIONS PAGES

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

3. The LHGR for Specification 3.2.3.
4. Control Rod Block Instrumentation Setpoint for the Rod Block MonitorUpscale Function Allowable Value for Specification 3.3.2.1.
5. The OPRM setpoints for the trip function for SR 3.3.1.3.3.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

delete

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."
2. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
3. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."
4. WCAP-16081-P-A, "10x10 SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2."
5. WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 2 to Code Description, Qualification and Application."

6. WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."

7. WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - Supplement 1."
8. WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENPD-287."
9. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

(continued)

Quad Cities 1 and 2 5.6-3 Amendment No. 264/259

Reporting Requirements 5.6 5.6 Reporting Requirements delete 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

10. WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application," Revision 1, October 2011.

11.

3 XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.

4

12. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs,"

Advanced Nuclear Fuels Corporation, May 1995.

5

13. EMF-85-74(P) Revision 0 Supplement 1 (P)(A) and Supplement 2 (P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.

6

14. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors,"

AREVA NP, February 2008.

7

15. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors

- Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.

8

16. XN-NF-80-19(P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.

9

17. XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX:

Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.

18.

10 EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation, October 1999.

(continued)

Quad Cities 1 and 2 5.6-4 Amendment No. 264/259

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) delete 19.

11 EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000.

12

20. EMF-2209(P)(A) Revision 3, "SPCB Critical Power Correlation," AREVA NP, September 2009.

21.

13 ANP-10298P-A Revision 1, "ACE/ATRIUM 10XM Critical Power Correlation," AREVA, March 2014.

22.

14 ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.

23.

15 XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient ThermalHydraulic Core Analysis," Exxon Nuclear Company, February 1987.

24.

16 ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses,"

Advanced Nuclear Fuels Corporation, August 1990.

25.

17 EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001.

26.

18 EMF-2292 (P)(A) Revision 0, "ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.

27.

19 ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors,"

Framatome ANP, September 2005.

20

28. EMF-CC-074(P)(A) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, August 2000.

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

(continued)

21. NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," February 2021, as approved by NRC Staff SE dated XXX XX, 20XX.

Quad Cities 1 and 2 5.6-5 Amendment No. 264/259

ATTACHMENT 3 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 RETYPED QCNPS UNITS 1 AND 2 TECHNICAL SPECIFICATIONS PAGES

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

3. The LHGR for Specification 3.2.3.
4. Control Rod Block Instrumentation Setpoint for the Rod Block MonitorUpscale Function Allowable Value for Specification 3.3.2.1.
5. The OPRM setpoints for the trip function for SR 3.3.1.3.3.
b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."
2. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
3. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.
4. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs,"

Advanced Nuclear Fuels Corporation, May 1995.

5. EMF-85-74(P) Revision 0 Supplement 1 (P)(A) and Supplement 2 (P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.
6. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.
7. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors -

Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.

(continued)

Quad Cities 1 and 2 5.6-3 Amendment No. 264/259

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. XN-NF-80-19(P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.
9. XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.
10. EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation, October 1999.
11. EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000.
12. EMF-2209(P)(A) Revision 3, "SPCB Critical Power Correlation," AREVA NP, September 2009.
13. ANP-10298P-A Revision 1, "ACE/ATRIUM 10XM Critical Power Correlation," AREVA, March 2014.
14. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.
15. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient ThermalHydraulic Core Analysis," Exxon Nuclear Company, February 1987.
16. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.
17. EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001.

(continued)

Quad Cities 1 and 2 5.6-4 Amendment No. 264/259

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

18. EMF-2292 (P)(A) Revision 0, "ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.
19. ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.
20. EMF-CC-074(P)(A) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, August 2000.
21. NEDC-33930P Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," February 2021, as approved by NRC Staff SE dated XXX XX, 20XX.

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

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Quad Cities 1 and 2 5.6-5 Amendment No. 264/259

ATTACHMENT 4 GNF-A 10 CFR 2.390 Affidavit for Withholding

Global Nuclear Fuel - Americas, LLC AFFIDAVIT I, Kent Halac, state as follows:

(1) I am the Senior Engineer, Global Nuclear Fuel - Americas, LLC (GNF-A), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GNF-A proprietary report 007N0608P, Revision 0, GNF Responses to NRC RAI 2 through RAI 6 Quad Cities Nuclear Power Station, Units 1 and 2 License Amendment Request to Transition to GNF3 Fuel, dated April 2022. GNF-A proprietary information in 007N0608P Revision 0 is identified by a dotted underline inside double square brackets. ((This sentence is an example {3})). GNF-A proprietary information in figures and large objects is identified by double square brackets before and after the object. In each case, the superscript notation

{3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 U.S.C. §552(b)(4), and the Trade Secrets Act, 18 U.S.C. §1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without a license from GNF-A constitutes a competitive economic advantage over other companies;
b. Information that, if used by a competitor, would reduce its expenditure of resources or improve its competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
c. Information that reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A; 007N0608P Revision 0 Affidavit Page 1 of 3

Global Nuclear Fuel - Americas, LLC

d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions for proprietary or confidentiality agreements or both that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GNF-A.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary and/or confidentiality agreements.

(8) The information identified in paragraph (2) is classified as proprietary because it contains the detailed GNF-A methodology for fuel analyses for the GNF-A Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the fuel analyses were achieved at a significant cost to GNF-A.

The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GNF-A asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and 007N0608P Revision 0 Affidavit Page 2 of 3

Global Nuclear Fuel - Americas, LLC analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GNF-A. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without there having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

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

Executed on this 4th day of April 2022.

Kent Halac Senior Engineer Global Nuclear Fuels Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 kent.halac@ge.com 007N0608P Revision 0 Affidavit Page 3 of 3

ATTACHMENT 5 Framatome 10 CFR 2.390 Affidavit for Withholding

AFFIDAVIT

1. My name is Morris Byram. I am Product 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 Attachment 4 to Constellation letter RS-22-040 dated April 11, 2022, with subject Response to Request for Additional Information Related to the License Amendment Request to Transition to GNF3 Fuel and referred to herein as Document. Information contained in this Document 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. The Framatome proprietary information is identified by single bolded square brackets.
4. This Document contains 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 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 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 this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 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: (4/4/2022)

BYRAM Morris Digitally signed by BYRAM Morris Date: 2022.04.04 08:10:23 -07'00' (NAME) morris.byram@framatome.com 2101 Horn Rapids Road Richland, WA 99354

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