Text

Request to Expand Applicability of Prime Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel
	ML22020A399
Person / Time
Site:	Quad Cities
Issue date:	01/20/2022
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML22020A398	List: RS-22-006, Request to Expand Applicability of Prime Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel;
References
	RS-22-006
	Download: ML22020A399 (52)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 10 CFR 50.90 RS-22-006 January 20, 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:

Request to Expand Applicability of PRIME Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel at Quad Cities

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. Global Nuclear Fuel, "The PRIME Model for Analysis of Fuel Rod Thermal -

Mechanical Performance," NEDC-33840P-A Revision 1, August 2017 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2.

The proposed change supports the transition from Framatome (formerly AREVA)

ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF-A) GNF3 fuel at QCNPS.

Specifically, EGC proposes to revise Technical Specifications (TS) 5.6.5, "Core Operating Limits Report (COLR)," paragraph b, to add a report that supplements the analysis methodologies included in General Electric Standard Application for Reactor Fuel (GESTAR) to the list of approved methods to be used in determining the core operating limits in the COLR. Other operational aspects of the fuel transition, including other TS 5.6.5.b changes, are addressed by the on-going NRC review of Reference 1.

Attachment 6 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390.

When separated from Attachment 6, this document is decontrolled.

January 20, 2022 U.S. Nuclear Regulatory Commission Page 2 The PRIME computer code is used to calculate fuel rod thermal and mechanical performance.

The NRC previously approved application of the PRIME methodology for licensing analyses of GNF-A boiling water reactor (BWR) fuel rod designs (Reference 2) with certain limitations on its application to non-GNF co-resident fuel in the core. EGC requests NRC approval to expand the use of PRIME and its associated methodologies to demonstrate compliance with the fuel melt and cladding strain criteria for the co-resident Framatome ATRIUM 10XM fuel in the QCNPS transition cores. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked-up to show the proposed TS changes. Attachment 3 provides TS Bases pages marked up to show the associated TS Bases changes and is provided for information only. Attachments 4 and 6 are the public and proprietary versions of a GNF-A report with data used to demonstrate the applicability of PRIME to the Framatome ATRIUM 10XM co-resident fuel. Attachment 6 contains information proprietary to GNF-A. As a result, this document is supported by a signed affidavit from the owner of the information, which is included as Attachment 5. The affidavit sets forth the basis on which GNF-A'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-A be withheld from public disclosure. A redacted non-proprietary version of the GNF-A report is provided in Attachment 4.

The proposed change has been reviewed by the QCNPS Plant Operations Review Committees, in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed license amendment by January 20, 2023. There are no regulatory commitments contained in this submittal. Once approved, the amendment will be implemented within 60 days.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), a copy of this application, with attachments, is being provided to the designated State Officials.

Should you have any questions concerning this submittal, please contact Ms. Rebecca L.

Steinman at (630) 657-2831.

January 20, 2022 U.S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. This statement was executed on the 20th day of January 2022.

Respectfully, Patrick R. Simpson Sr. Manager Licensing Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Changes

2. Mark-up of QCNPS, Units 1 and 2 Technical Specifications Pages

3. Mark-up of QCNPS, Units 1 and 2 Technical Specifications Bases Pages - For Information Only

4. 006N8642-NP, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels," Revision 1, dated January 2022 (Non-Proprietary Version)

5. Global Nuclear Fuel - Americas, LLC Affidavit for Withholding

6. 006N8642-P, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels," Revision 1, dated January 2022 (Proprietary Version) cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector, Quad Cities Nuclear Power Station NRC Project Manager, Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

Request to Expand Applicability of PRIME Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel at Quad Cities 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Applicability of the PRIME Safety Evaluation 2.2 Detailed Description of Change

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration 4.3 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION Exelon Generation Company, LLC (EGC) requests an amendment to the Technical Specifications (TS) for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed change supports the transition from Framatome (formerly AREVA) ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF-A) GNF3 fuel at QCNPS. Specifically, EGC proposes to revise Technical Specifications (TS) 5.6.5, "Core Operating Limits Report (COLR),"

paragraph b, to add a report that supplements the analysis methodologies included in General Electric Standard Application for Reactor Fuel (GESTAR) to the list of approved methods to be used in determining the core operating limits in the COLR.

PRIME calculates fuel rod thermal and mechanical performance. The NRC previously approved application of the PRIME methodology for licensing analyses of GNF-A boiling water reactor (BWR) fuel rod designs as described with certain limitations on its application to non-GNF co-resident fuel in the core (Reference 6.1). EGC requests NRC approval to expand the use of PRIME and its associated methodologies to demonstrate compliance with the fuel melt and cladding strain criteria for the transition cores containing Framatome ATRIUM 10XM fuel. This is to be used as an alternative to or in conjunction with the currently approved approach of using overpower limits as provided by the non-GNF fuel manufacturing vendor.

2.0 DETAILED DESCRIPTION 2.1 Applicability of the PRIME Safety Evaluation EGC plans to begin loading GNF3 fuel into the QCNPS Unit 1 core beginning with the refueling outage scheduled for the spring of 2023 and Unit 2 during the refueling outage scheduled for the spring of 2024. After the spring 2022 refuel outage, the QCNPS Units 1 and 2 reactors are fueled with Framatome ATRIUM 10XM fuel bundles, and these bundles will remain in the cores for a transition period following the initial loading of GNF3 fuel. For these transition cycles, compliance with fuel acceptance criteria must be demonstrated for both the GNF3 and the co-resident ATRIUM 10XM fuel. The transition cores will be evaluated using GESTAR methods, which includes the use of the PRIME computer code for the evaluation of fuel rod thermal and mechanical performance.

The GNF-A PRIME computer code and associated methodologies have been reviewed and approved by the NRC. References 6.1 through 6.4 are the current PRIME licensing topical reports. GNF-A has concluded that the Limitations and Conditions imposed in Section 4.0 of Reference 6.1 currently prevent the application of the PRIME thermal overpower (TOP) /

mechanical overpower (MOP) methodologies to demonstrate that non-GNF fuel complies with the acceptance criteria for fuel melt and cladding strain without additional NRC review and approval. The TOP methodology is proposed to be applied to the Framatome ATRIUM 10XM fuel with no changes, while the MOP methodology will be applied with the changes described in Section 5.2.3 of Reference 6.5 to ensure fuel damage due to fuel melting or excessive cladding strain does not occur as a result of anticipated operational occurrences (AOOs). The proposed methodology allows the indicated criteria of fuel melt temperature and cladding strain to be evaluated directly instead of through a simplification which requires excess conservatism due to the incompatibilities between GNF-A and non-GNF fuel manufacturing vendor methods. The

ATTACHMENT 1 Evaluation of Proposed Changes proposed method is to be used as an alternative to or in conjunction with the currently approved approach of using overpower limits provided by the non-GNF fuel manufacturing vendor.

2.2 Detailed Description of Change GNF-A prepared a report (Reference 6.5) that demonstrates PRIME and its application methodologies are qualified and appropriate for use in directly assessing compliance to fuel centerline melt and cladding strain criteria for the non-GNF fuel type that will be in the QCNPS transition cores. EGC is requesting that this GNF-A report, 006N8642-P, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels", be added to TS 5.6.5.b to document the expansion of the applicability of PRIME and its associated methodologies to demonstrate that the co-resident non-GNF fuel complies with the acceptance criteria for fuel melt and cladding strain. Attachment 2 contains a marked-up version of the TS that shows the combined effect of this proposed addition to the reference list and the changes previously submitted to the NRC in Attachment 2 of Reference 6.6.

QCNPS will make supporting changes to the TS Bases in accordance with TS 5.5.10, "Technical Specifications (TS) Bases Control Program." Attachment 3 provides the marked-up TS Bases pages associated with the TS 5.6.5.b change discussed above. The TS Bases marked-up pages are being submitted for information only and do not cover aspects of the fuel transition that may also impact these Bases pages.

3.0 TECHNICAL EVALUATION

QCNPS plans to load GNF3 fuel into Unit 1 during the spring 2023 refueling outage and into Unit 2 during the spring 2024 refueling outage. The application of PRIME and its approved methodologies is proposed for assessing the fuel melt and cladding strain criteria during fast and slow transients for ATRIUM 10XM fuel for the transition cycles. The key parameters for assessing PRIME's capabilities with respect to these criteria are the ability to reasonably predict fuel temperature and transient cladding deformation.

Justification for the application of PRIME to the co-resident QCNPS Framatome ATRIUM 10XM fuel is provided in Reference 6.5 as follows:

Section 2 addresses the applicability of PRIME for assessing the fuel melt criterion.

Section 3 addresses the applicability of PRIME for assessing the cladding strain criterion.

Section 4 addresses the cladding material differences between the GNF3 and ATRIUM 10XM fuel designs in relation to internal PRIME models for irradiation growth and creep. This section also justifies the application of PRIME with non-GNF vendor supplied Thermal Mechanical Operating Limits (TMOLs) which may have attributes outside the previously approved applicability of PRIME.

Page 3

ATTACHMENT 1 Evaluation of Proposed Changes

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 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. 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 the addition of a new report, 006N8642-P, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels", to TS 5.6.5.b to document the expansion of the applicability of using PRIME and its associated methodologies to demonstrate that non-GNF fuel complies with the acceptance criteria for fuel melt and cladding strain.

4.2 No Significant Hazards Consideration Overview Exelon Generation Company, LLC (EGC) proposes to revise Technical Specifications (TS) 5.6.5, "Core Operating Limits Report (COLR)," paragraph b, to add a new report that supplements the analysis methodologies included in General Electric Standard Application for Reactor Fuel (GESTAR) to the list of approved methods to be used in determining the core operating limits in the COLR.

EGC has evaluated the proposed change against the criteria of 10 CFR 50.92(c) to determine if the proposed changes result in any significant hazards. The following is the evaluation of each of the 10 CFR 50.92(c) criteria:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No TS 5.6.5, "Core Operating Limits Report (COLR)," lists NRC approved analytical methods used at Quad Cities Nuclear Power Station (QCNPS) to determine core operating limits. The proposed changes will add one new reference to the list of administratively controlled analytical methods in TS 5.6.5.b.

QCNPS Unit 1 is scheduled to load Global Nuclear Fuel - Americas, LLC (GNF-A) fuel during its upcoming outage in the spring of 2023. The proposed change to TS 5.6.5 will support the initial insertion of GNF3 fuel by allowing the use of PRIME and its application methodologies to assess Framatome ATRIUM 10XM co-resident fuel compliance with fuel centerline melt and cladding strain acceptance criteria. The application of this methodology has no effect on any accident initiator or precursor previously evaluated and does not change the manner in which the core is operated, as it is the method of assessment for safety analysis that is changing and not the criteria themselves. PRIME is a previously approved General Electric (GE) method whose approved application is being expanded to allow two additional assessments for a non-GNF fuel design in use at QCNPS. This expanded application has no effect on the type or Page 4

ATTACHMENT 1 Evaluation of Proposed Changes amount of radiation released and has no effect on predicted offsite doses in the event of an accident.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed methodology changes to evaluate the transition cores containing two fuel types do not affect the performance of any QCNPS structure, system, or component credited with mitigating any accident previously evaluated. The proposed change expands the applicability of PRIME to evaluate fuel melt and cladding strain compliance for the QCNPS ATRIUM 10XM fuel.

The proposed analysis methodology change does not introduce any new modes of system operation or failure mechanisms.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change adds one new reference to the list of administratively controlled analytical methods in TS 5.6.5 that can be used to evaluate the thermal-mechanical behavior of the transition cores. The proposed analysis methodology change does not modify the safety limits or setpoints at which protective actions are initiated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The modified methodology implicitly includes conservatism, aligns with the standard review plan, and adds additional conservatism to compensate for uncertainties associated with evaluating fuel manufactured by another vendor. The proposed methodology change allows the criteria of fuel melt temperature and cladding strain to be evaluated directly instead of through a simplification which requires excess conservatism due to the incompatibilities between GNF-A and other fuel manufacturing vendor methods. Therefore, QCNPS has determined that the proposed change provides an equivalent level of protection as that currently provided.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

4.3 Conclusion Based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Page 5

ATTACHMENT 1 Evaluation of Proposed Changes

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 Global Nuclear Fuel NEDC-33256-P-A, Revision 2, The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance Part 1 - Technical Bases, dated October 2021 (ADAMS Accession No. ML21279A283 (non-proprietary version))

6.2 Global Nuclear Fuel NEDC-33257-P-A, Revision 2, The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance Part 2 - Qualification, dated October 2021 (ADAMS Accession No. ML21279A283 (non-proprietary version))

6.3 Global Nuclear Fuel NEDC-33258-P-A, Revision 2, The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance Part 3 - Application Methodology, dated October 2021 (ADAMS Accession No. ML21279A283 (non-proprietary version))

6.4 Global Nuclear Fuel NEDC-33840-P-A, Revision 1, The PRIME Model for Transient Analysis of Fuel Rod Thermal-Mechanical Performance, dated August 2017 (ADAMS Accession No. ML17230A012 for NEDO-33840, which is the non-proprietary version) 6.5 Global Nuclear Fuel 006N8642, Revision 1, Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels,, dated January 2022 (RS-22-006 Attachments 4 (non-proprietary) and 6 (proprietary))

6.6 Letter from P. 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 (ADAMS Accession No. ML21307A444)

Page 6

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 MARK-UP OF 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)

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.

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.

22. 006N8642-P, Revision 1, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels,"

January 2022.

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 MARK-UP OF QCNPS, UNITS 1 AND 2 TECHNICAL SPECIFICATIONS BASES PAGES -

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7. 006N8642, Revision 1, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels,"

January 2022.

ATTACHMENT 4 006N8642-NP, Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels, Revision 1, dated January 2022 Non-Proprietary Version

Global Nuclear Fuel 006N8642-NP Revision 1 January 2022 Non- Proprietary Information Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels Copyright 2022 Global Nuclear Fuels All Rights Reserved

006N8642-NP Revision 1 Non - Proprietary Information INFORMATION NOTICE This is a non-proprietary version of the document 006N8642-P, Revision 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by open and closed brackets as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document are furnished in accordance with the contract between Exelon and GNF, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GNF makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

ii

006N8642-NP Revision 1 Non - Proprietary Information TABLE OF CONTENTS

1.0 INTRODUCTION

............................................................................................................... 1 2.0 FUEL MELTING COMPLIANCE ................................................................................... 2 2.1 Phenomenon and Qualification.................................................................................... 2 2.2 Temperature Limit ....................................................................................................... 4 2.3 Methodology ................................................................................................................ 6 3.0 CLADDING STRAIN (PCMI) COMPLIANCE .............................................................. 7 3.1 Phenomenon and Qualification.................................................................................... 7 3.2 Strain Limit ................................................................................................................ 10 3.3 Methodology .............................................................................................................. 13 4.0 CWSR ZIRC-2................................................................................................................... 16 4.1 Irradiation Growth ..................................................................................................... 17 4.2 Irradiation Creep ........................................................................................................ 19 4.3 Compliance Sensitivity .............................................................................................. 21 5.0 ACRONYMS ..................................................................................................................... 25

6.0 REFERENCES

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006N8642-NP Revision 1 Non - Proprietary Information LIST OF TABLES Table 4-1: PRIME Cladding Models ............................................................................................16 Table 4-2: Creep Coefficients CCOEFF .......................................................................................21 Table 4-3: PRIME Cases Analyzed ..............................................................................................22 LIST OF FIGURES Figure 2-1: PRIME Nominal Predicted over Measured (P/M) of Fuel Centerline Temperature ...3 Figure 2-2: PRIME 2-Sigma Predicted over Measured (P/M) of Fuel Centerline Temperature ....4 Figure 2-3: UO2 Melt Data .............................................................................................................5 Figure 3-1: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain ...8 Figure 3-2: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain ...9 Figure 3-3: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain Including PRIME Transient Qualification ....................................................................................10 Figure 3-4: GNF Hydrogen Burst Test Results ............................................................................12 Figure 3-5: NRC Approved BWR Zircaloy-2 Hydrogen Uptake Model .....................................13 Figure 3-6: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain using Increased Power Multiplier .................................................................................................14 Figure 4-1: Comparison of PRIME Models in NEDC-33256P-A and FAST (Reference 6)

Models for Irradiation Growth ......................................................................................................18 Figure 4-2: PRIME Axial Growth Models ...................................................................................19 Figure 4-3: Impact of CWSR Properties on Melt Margin and Strain Compliance for a Limiting BWR/3 HPCI AOO.......................................................................................................................23 iv

006N8642-NP Revision 1 Non - Proprietary Information ABSTRACT This report provides justification for the use of PRIME and its associated licensed methodologies to demonstrate compliance directly to the no fuel melting and cladding transient strain criteria for non-GNF fuel designs. Currently, the no fuel melting criterion is evaluated using the Thermal Overpower (TOP) methodology and the cladding strain criterion is evaluated using the Mechanical Overpower (MOP) methodology described in References 1 and 2. In addition, this report describes the models to be used in PRIME for cold work stress relieved (CWSR) Zircaloy-2 for irradiation growth and creep. Lastly, a demonstration calculation is provided showing the relative sensitivity these models have on a limiting Anticipated Operational Occurrence (AOO) event. Specifically, this justification is being written to support license amendment requests (LARs) for the Exelon plants Quad Cities Units 1 and 2 and Dresden Units 2 and 3 to support the transition from Framatome ATRIUM 10XM fuel to GNF3 fuel.

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1.0 INTRODUCTION

GNF is supporting Exelon in the transition of the Quad Cities Units 1 and 2 and Dresden Units 2 and 3 reactors from Framatomes ATRIUM 10XM fuel design to GNFs GNF3 design. As part of this effort, GNF is tasked with demonstrating thermal-mechanical compliance to the no fuel centerline melting and transient cladding strain criteria. For these transition cycles, compliance must be shown for both the newer GNF3 and the co-resident ATRIUM 10XM fuel bundles.

The purpose of this document is to establish a technical basis to justify using PRIME to show compliance to no fuel melting and cladding strain criteria as an alternative to, or in conjunction with, the currently approved approach of using overpower limits for non-GNF fuels as provided by the manufacturing vendor. The basis applies to showing compliance for both slow (i.e., using the steady-state methodology defined in Reference 1) and fast (i.e., using the transient methodology defined in Reference 2) transients. The approval would eliminate the restriction related to overpower limits in the second-to-last sentence in limitation and condition (L&C) 1.a of NEDC-33256P-A (Reference 3), which is also referenced in NEDC-33840P-A, in support of these core transitions.

The transient methodologies are generally incompatible between GNF and the manufacturing vendor, necessitating simplification of overpower limits provided by the manufacturing vendor.

This simplification requires the addition of conservatism to ensure compliance with the aforementioned criteria. GNF may then be unable to meet the conservative overpower limits when applied using GNF methodology, resulting in significant operational penalties to the as-provided Thermal Mechanical Operating Limit (TMOL) for the non-GNF fuel. The use of PRIME to show compliance directly to the no fuel centerline melt and cladding strain criteria allows GNF to remove this excess conservatism and support operation using the as-provided TMOL.

Furthermore, this document establishes a basis for the properties to use in PRIME to model Zirc-2 CWSR cladding for material properties ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` )), per the Technical Evaluation of the Safety Evaluation Report (SER) in Reference 1. The use of these models is restricted to showing compliance to no fuel melting and cladding strain criteria. In addition, these models may be used to support the input generation for downstream safety analyses and demonstrating downstream overpower compliance for non-GNF fuels.

Lastly, this document establishes a procedure by which GNF will confirm transient compliance and downstream input generation for non-GNF fuels provided with TMOLs that exceed the PRIME limit of ((` ` ` ` ` ` ` ` ` ` ` ` ` ` )).

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006N8642-NP Revision 1 Non - Proprietary Information 2.0 FUEL MELTING COMPLIANCE Compliance to the GESTAR II criterion that loss of fuel rod mechanical integrity will not occur due to fuel melting (Section 1.2.2.B.ix of Reference 4) is demonstrated by utilizing PRIME and its licensed methodologies References 1 and 2 under AOOs. The thermal overpower screening parameter used by GNF to show compliance to the no fuel centerline melting criterion is referred to as TOP. This section provides a basis, as well as updates to the methodology, for showing no fuel melting compliance for non-GNF fuels.

2.1 Phenomenon and Qualification The maximum temperature in the fuel during an AOO is largely controlled by the thermal conductivity of the fuel and thermal conductance of the fuel/cladding gap. For fast transients, the radial power distribution and specific heat capacity of the fuel are also important characteristics affecting the ability of the heat to transfer from the pellet to the coolant.

Of the properties identified above, only the gap conductance is highly influenced by fuel vendor manufacturing. For steady-state and AOO conditions, the gap conductance is comprised of conduction through solid points of contact between the fuel and cladding and conduction through the gap-gas, as described in Section 3.2 of NEDC-33256P-A. The conduction through solid contact is a function of thermal properties (e.g., fuel and cladding conductivities), geometry (e.g., fuel and cladding diameters), power (via temperatures and fuel relocation) and other fabrication parameters (e.g., surface roughness). The conduction through the gas is controlled by the gap-gas composition, which is controlled largely by the fuel temperatures (namely power and as-fabricated geometry).

Nominal geometrical parameters (along with the TMOL) are provided by the non-GNF fuel vendor, while manufacturing uncertainties are described in more detail in Section 2.3.

Thermal properties (such as thermal conductivity and specific heat) are generally based on the material composition and, to a lesser extent, the fabrication specifics such as porosity. The commonly used correlations for fuel thermal properties are based on legacy measurements and have been adjusted over the years to account for material property changes, such as different as-fabricated densities and additives, and improved understanding of the phenomenon (e.g., thermal conductivity degradation (TCD) with exposure). PRIMEs models also account for these variations (e.g., as-fabricated density) and phenomenon (e.g., TCD) as described in Reference 3.

Furthermore, in a consistent approach, the US Nuclear Regulatory Commissions (NRCs) thermal-mechanical code FAST (Reference 5) uses a single set of fuel material properties for all varieties of UO2 fuel independent of reactor type (BWR/PWR) or fuel vendor (GNF, Westinghouse, Framatome) as detailed in Section 2.1 of Reference 6. Given that the non-GNF fuel to be analyzed falls within the range of applicability listed in Table 2.1 of Reference 3, the fuel thermal models used in PRIME are applicable to non-GNF fuel.

The PRIME qualification database for fuel centerline temperature is reproduced in Figure 2-1 and Figure 2-2 for the nominal and model uncertainty cases, respectively. The database for fuel centerline temperature is comprised of over ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

experiments. The model uncertainty was established in Reference 1 to ensure that the qualification 2

006N8642-NP Revision 1 Non - Proprietary Information data is statistically bounded for a one-sided confidence interval (i.e., such that PRIME predicts a fuel temperature above the measured temperature). It is important to note that the PRIME qualification on fuel centerline temperature predictions is an integral qualification, accounting for the interplay of numerous fuel rod models and manufacturing variability, including gap conductance, which may be influenced by vendor specific parameters. The model uncertainty is established to encompass all of these uncertainties, and in PRIMEs statistical application methodology, the model uncertainty ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

((

` ` ` ` ))

Figure 2-1: PRIME Nominal Predicted over Measured (P/M) of Fuel Centerline Temperature 3

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((

` ` ` ` ))

Figure 2-2: PRIME 2-Sigma Predicted over Measured (P/M) of Fuel Centerline Temperature The fuel temperature qualification database is comprised ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) Therefore, the use of this database and its derivative uncertainties are appropriate for assessing PRIMEs ability to model the fuel temperature of non-GNF fuels consistent within the range of applicability listed in Table 2.1 of NEDC-33256P-A.

2.2 Temperature Limit The currently approved PRIME temperature requirement is that the maximum fuel centerline temperature shall remain below the fuel melting point at a 95% confidence interval (Reference 1).

This requirement applies to both steady-state operation and slow and fast AOOs. The figure of merit used to comply with the criterion above in PRIME licensing evaluations is referred to as the melt margin. The melt margin is calculated as the difference between the fuel melting temperature and the upper 95% confidence interval prediction of fuel temperatures. A positive melt margin indicates that the criterion to preclude fuel melting has been satisfied.

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006N8642-NP Revision 1 Non - Proprietary Information The fuel melting correlation used in PRIME, reproduced in Figure 2-3, is a function of ((` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) (Equation 3-61 of Reference 3). The correlation is derived from ((` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` )). As the current PRIME correlation and updated database below is comprised of a mixture of fuel manufacturers, ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) the PRIME correlation is equally applicable to both GNF and non-GNF fuels when the PRIME uncertainty is considered.

((

` ` ` ` ))

Figure 2-3: UO2 Melt Data As stated in NUREG-0800 Standard Review Plan (SRP) Chapter 4.2, the assumption that centerline melting results in fuel failure is conservative. One reason for this statement is due to 5

006N8642-NP Revision 1 Non - Proprietary Information the criterion of no fuel melting being applied to the hottest part of the fuel pellet, the centerline, for AOOs. Given the poor thermal conductivity of UO2, significant thermal gradients (on the order of > 1000°F) exist in the fuel pellet making it extremely unlikely that the centerline melt can propagate radially to reach the cladding given the cooler fuel temperatures as it approaches the heat transfer surface. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )). Furthermore, axial power gradients would eliminate significant molten fuel axial migration, which has historically been demonstrated to be the dominant mechanism for cladding swelling (and thus failure) due to fuel volume expansion. Although fuel volumetric expansion due to phase change may result in some additional cladding strain, minimal strain is expected to occur with limited centerline melting.

Given that GNF uses the no fuel centerline melting criterion and TOP statistical application methodology to demonstrate no centerline melting, the application of this criterion to non-GNF fuel is conservative.

2.3 Methodology The PRIME methodology for demonstrating compliance to the no fuel melting criterion is described in References 1 and 2. At a high level, the methodology uses a statistical framework to determine the lower 95% melt margin (as described in Section 2.2) when ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) Of the ((` ` ` ` ` ` ` ` ))

perturbations analyzed, the ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )). For any uncertainties not provided by the non-GNF fuel vendor (due to the proprietary nature of those values or the uncertainties used by PRIME methodologies being used in a different manner than other vendors), current-production GNF parameters will be used due to the minimal effect of these parameters on fuel temperature uncertainty.

The modeling uncertainty used in the PRIME methodology was developed using a fuel temperature database comprised of both GNF and non-GNF fuels (including BWRs and PWRs with as-fabricated densities ranging from ((` ` ` ` ` ` ` ` ` ` ` ` )) theoretical density), as shown in Figure 2-2, and is therefore applicable for use on non-GNF BWR fuel designs. Furthermore, the conservatism in the PRIME melting correlation, as well as the use of fuel melting as the failure criterion, provides significant margin between the analyzed failure limits and those required to cause actual fuel failure due to overheating of the pellet.

Therefore, the methodology used for demonstrating compliance to the no fuel melting criterion for GNF fuel is applicable unchanged to non-GNF fuel.

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006N8642-NP Revision 1 Non - Proprietary Information 3.0 CLADDING STRAIN (PCMI) COMPLIANCE Compliance to the GESTAR II criterion that loss of fuel rod mechanical integrity will not occur due to pellet-cladding mechanical interaction (PCMI, Section 1.2.2.B.x of Reference 4) is demonstrated by utilizing PRIME and its licensed methodologies in References 1 and 2 under normal operation and AOOs. The screening parameter used by GNF to show compliance to cladding strain limits is referred to as MOP. This section provides a basis, as well as updates to the methodology, for showing strain compliance for non-GNF fuels.

3.1 Phenomenon and Qualification The strain imparted by the fuel onto the cladding during an AOO is caused primarily by thermal expansion of the fuel pellet. As the power increases, the relatively poor thermal conductivity of the fuel causes the fuel temperatures to increase therefore resulting in fuel expansion into the cladding. Given the significant difference in strength of Zircaloy vs UO2, the cladding yields resulting in both an instantaneous plastic strain as well as the potential for creep strain. For slow AOOs, the PRIME steady-state methodology applies a ((` ` ` ` ` ` ` ` ` ` ` ` ` )) hold to allow for operator action, which increases the permanent strain of the cladding due to creepout. During the AOO event, the overpower puts the cladding under high tensile stress as well as elevated temperatures, thereby increasing the creep rate relative to normal operating conditions. The hold time also further increases fuel temperatures, resulting in increased thermal expansion and additional cladding strain.

In addition to thermal expansion, an additional swelling mechanism known as gaseous swelling has been observed at higher exposures (References 10 and 11). This mechanism is believed to be due to a combination of 1) gases trapped in the grains (intra-granular swelling) and 2) gas bubbles trapped on the grain faces. The result of this swelling is an increase in the strain imparted onto the cladding during a power ramp. It has been demonstrated that the duration of the power increase has a significant effect on the total gaseous swelling, with a majority of the swelling occurring during pellet annealing (on the order of several minutes to hours (Reference 10)). Furthermore, this phenomenon is highly dependent on fuel temperatures, with both a lower and upper bound for which the phenomenon is observed. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

The PRIME steady-state (i.e., instantaneous) transient strain qualifications during a power ramp are re-produced in Figure 3-1 and Figure 3-2. These cases are consistent with the original PRIME Licensing Topical Report (LTR) qualification (Reference 12) as supplemented by the 2015 PRIME update in Reference 13. The PRIME Transient methodology strain qualification is shown in Figure 3-3, reproduced from NEDC-33840P-A.

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((

` ` ` ` ))

Figure 3-1: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain 8

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((

` ` ` ` ))

Figure 3-2: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain 9

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((

` ` ` ` ))

Figure 3-3: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain Including PRIME Transient Qualification 3.2 Strain Limit The transient strain limit for GNF fuel designs, approved for use as defined in Reference 1, is reproduced below, where exposure is defined in terms of peak pellet exposure (PPE):

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

The limit is applied as a transient diametral strain limit, taken as the maximum strain during the event minus the strain at the onset of the event. For slow transients, the maximum strain is always at the end of the event, but for fast transients the timing of peak strain may vary as the power and temperatures change. The purpose of these limits is to ensure adequate ductility remains in the cladding. The ductility is reduced from the as-fabricated state due to both irradiation hardening and hydrogen pickup / hydride formation.

As discussed in the Economic Simplified Boiling-Water Reactor (ESBWR) Design Certification Application RAI-4.2-2 S03 and 4.2-4 S02 (Reference 14) response justifying the aforementioned allowable strains, PCMI loading is more consistent with strain controlled testing as opposed to load controlled testing. Therefore, total elongation results are more relevant to PCMI failure due to the effect of the test specimen geometry on the uniform elongation caused by dislocation channeling. Unirradiated Zircaloy cladding has significant ductility, with uniform elongation of 10

006N8642-NP Revision 1 Non - Proprietary Information greater than 7% (up to ~20%, and total elongations of > 25%) (References 15 and 16). However, under irradiation the ductility is reduced due to fast neutrons damaging the lattice. The predominant defect affecting ductility is the <a> dislocation loop, which forms early in the cladding lifetime and saturates at relatively low fluence levels. In PRIME, ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) (Reference 3).

Furthermore, the water chemistry may result in further ductility reduction due to hydrogen pickup.

This ductility reduction is accounted for in the PRIME methodology ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

``````````````````````````````````````````````````````````` `````````````````````````````````````

`````````````````````````````````````````````````````````` ``````````````````````````````````````

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )).

The exposure of ((` ` ` ` ` ` ` ` ` ` ` ` ` ` )) is a surrogate for a quantity of hydrogen in the cladding.

This exposure was chosen to ensure that the cladding hydrogen ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` )) criterion was chosen based on testing conducted by GE on irradiated cladding specimens containing varying levels of hydrogen as discussed in Reference 14. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

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((

` ` ` ))

Figure 3-4: GNF Hydrogen Burst Test Results The NRC approved hydrogen pickup model for BWR Zircaloy-2 as defined in RG-1.236 is reproduced in Figure 3-5. It can be demonstrated from this model that the hydrogen concentration for NRC-approved non-GNF Zircaloy-2 claddings is sufficiently low to ensure adequate ductility up to the current PRIME limit of ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )). Therefore, the basis of the transient strain limit described above is also appropriate for non-GNF Zirc-2 claddings and the same limits may be applied.

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006N8642-NP Revision 1 Non - Proprietary Information Figure 3-5: NRC Approved BWR Zircaloy-2 Hydrogen Uptake Model 3.3 Methodology The PRIME methodology for demonstrating cladding strain compliance is described in References 1 and 2. At a high level, the methodology uses a worst-tolerance (WSTOL) framework to calculate the transient ((` ` ` ` ` ` ` ` ` ` ` ` ` )) cladding strain when ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )). The WSTOL methodology puts selected manufacturing tolerances, along with ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )), at their maximum (or minimum) values to ensure the largest strain is calculated in the cladding. Of the perturbed manufacturing tolerances, the most significant contributors are ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )), the others having only minor sensitivity or some variability in the direction which increases calculated cladding strains. These two parameters are both ((` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) For manufacturing tolerance inputs not available for non-GNF fuel, GNF fabrication parameters will be used. This is acceptable because GNF fabrication tolerances are applied to the non-GNF fuel in the demonstration of WSTOL predictions for cladding diametral strain in the multiple fabricator qualification plots shown in Figure 3-1 and Figure 3-2.

In addition to manufacturing tolerances, the WSTOL methodology biases ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) to reduce cladding strength and increase creep rates (via increased temperatures) and increased metal thinning due to oxidation. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

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` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) Oxide values will be provided or confirmed to be appropriate for non-GNF claddings.

An additional conservatism applied in WSTOL analyses is a power penalty, which is used to account for power spiking due to densification (a low-burnup phenomenon associated with highly densifying fuel materials) as well as additional uncertainties ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )). To account for additional uncertainties of non-GNF fuels, including additional gaseous swelling beyond what exists in the current GNF qualification database, the power penalty factor applied for WSTOL

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` )) Given the uncertainty is applied as a direct power multiplier, this results in a larger strain increase driven by an increase in fuel thermal expansion. The PRIME qualification using this increased uncertainty multiplier is shown in Figure 3-6, as compared to the PRIME qualification shown in Figure 3-2 which does not include any power penalty.

((

` ` ` ` ))

Figure 3-6: PRIME Worst-Tolerance Methodology Predictions for Cladding Diametral Strain using Increased Power Multiplier Lastly, to be compliant with the requirements of Limitations & Conditions (L&C) 4.f of NEDC-33256P-A (Reference 3), which states ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

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` ` )), this will be confirmed with the values used for PRIME. Manufacturing tolerances provided by the non-GNF fuel vendor that differ from standard GNF tolerances will be used to demonstrate that the ramp strain qualification database is bounded to the required extent. If the provided 14

006N8642-NP Revision 1 Non - Proprietary Information manufacturing tolerances do not sufficiently bound relative to the SER requirements of ((` ` ` ` ` ` ` `

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` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

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006N8642-NP Revision 1 Non - Proprietary Information 4.0 CWSR ZIRC-2 After reviewing documentation associated with the ATRIUM 10XM fuel transitions for Quad Cities and Dresden, it was determined that the cladding material is CWSR Zircaloy-2 (page 3-1 of Reference 17). ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` )) The PRIME Technical Bases LTR (Reference 3) provides a list of cladding models used by PRIME, along with their applicability basis, reproduced in Table 4-1. These models are also used by the PRIME Transient LTR (Reference 2) with the same limitations on application.

Table 4-1: PRIME Cladding Models Item Applicability Range per SER

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((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ````````````````````````````````````

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````````````````````

````````````````` ````````````````````````````````````

```````````````````````````

````````````````

````````````````````

```````````````````````````

````````````````````````````````````````````````

``````````````````

``````````

````` ``````````````

````````` ````````````````````````````````````````````````

```` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

Of the models listed above, ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )), as stated in the SER. These models are consistent with the models identified as being influenced by the heat treatment in the RODEX4 Supplement 1 Safety Evaluation (SE) in Reference 18 which updates RODEX4 applicability from being only applicable to CWSR Zircaloy-2 to both RXA and CWSR Zircaloy-2. In addition, it is also expected that the final heat treatment may affect the oxidation and hydrogen pickup due to an alteration to the grain morphology and second phase particle (SPP) size and distribution.

16

006N8642-NP Revision 1 Non - Proprietary Information Of these four items identified, only irradiation growth and creep are models internal to PRIME.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) and will be provided by the non-GNF fuel vendor. The remainder of this section describes the models to be used in PRIME for CWSR Zircaloy-2 for irradiation growth and creep. Lastly, a demonstration calculation is provided showing the relative sensitivity these models have on a limiting AOO event.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ))

4.1 Irradiation Growth The irradiation growth of cold worked Zircaloy materials has been demonstrated to be ~2x higher than that of fully recrystallized materials, as shown in Figure 4-6 of Reference 19 for various degrees of cold work and in Figure 7 of Reference 20 for Zirc-2 RXA vs CWSR. The degree of cold work affects the axial growth, with increasing amounts of cold work increasing the initial growth rate relative to fully recrystallized materials. This is due to the increased dislocation density at beginning of life. However, at high fluences when the dislocation densities are similar, the growth rates also become similar.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

17

006N8642-NP Revision 1 Non - Proprietary Information

((

` ` ` ` ))

Figure 4-1: Comparison of PRIME Models in NEDC-33256P-A and FAST (Reference 6)

Models for Irradiation Growth

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

`````````````````````````````````````````````````

(4-1)

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

18

006N8642-NP Revision 1 Non - Proprietary Information

((

` ` ` ` ))

Figure 4-2: PRIME Axial Growth Models 4.2 Irradiation Creep The creep rate of CWSR Zircaloy-2 is known to be faster (~2x) than that of RXA Zircaloy-2 under normal operating conditions (Reference 20). This is consistent with the change in cold work for other Zirconium-based alloys as well, such as Zirc-4, shown in Figure 4-33 of Reference 19. The dominant creep mechanism under low stress conditions is climb and glide of dislocations (Reference 22). Therefore, the increased creep rate at low stress for cold-worked materials is due to the higher dislocation density as compared to RXA (Reference 19).

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````

(4-2)

(4-3) 19

006N8642-NP Revision 1 Non - Proprietary Information

`````````````````````````````````````````````````````````````````````````````````````````````

(4-4)

(4-5)

`````

`` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

`` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

```````````````````````````

`````````````````````

```````````````````````

````````````````````````````````````````````````````````````````````````````````````````````````

``````````````````````````````````````````````````````````````` ))

20

006N8642-NP Revision 1 Non - Proprietary Information Table 4-2: Creep Coefficients CCOEFF Coefficient Value Coefficient Value

`````````` `````````

(( ````

`````````` `````````

```` ````````` ```` `````````

```` ``````` ```` ```````

``````````

```` ```` ```````

``````````

```` `````````` ```` ````````

`````````

```` ```` ```````

`````````

```` `````````` ```` ````````

````````

```` `````````` ````

`````````

````` `````````` ` ` ` ` ))

4.3 Compliance Sensitivity This section demonstrates the effect of using standard Zirc-2 RXA properties in PRIME to the updated Zirc-2 CWSR properties proposed for use in the preceding sections. The transient being analyzed is a typical BWR/3 limiting fast transient involving the inadvertent start-up of the High-Pressure Coolant Injection (HPCI) system. The methodology being used to perform this study is consistent with the Method #1 approach described in Section 5.1 of NEDC-33840P-A, which directly evaluates compliance to fuel centerline melting and cladding transient strain using statistical and worst tolerance methodologies, respectively. The four cases analyzed are listed in Table 4-3.

21

006N8642-NP Revision 1 Non - Proprietary Information Table 4-3: PRIME Cases Analyzed Case Description

((` ` ` ` ` ` ` ` ` ` ````````````````````````````````````````````````````````````````````

````````````````` `````````````````````````````````````````````````````````````````````````

`````` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` 4.1` ` ` `

````````````````` `````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` 4.2` ` ` `

`````````````````````````````````````````````````````````````````````````

`````````````````

`````````````````````````````````````````````````````````````````````````

````````````

` ` ` ` ` ` ` ` ` ` 4.1` ` ` ` ` ` ` ` 4.2` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

The sensitivity study was performed for the range of exposures of (( ` ` ` ` ` ` ` ` ` ` ` ` ` )) PPE. The lower bound of ((` ` ` ` ` ` ` ` ` ` ` ` ` ` )) was chosen due to the fuel to be analyzed during this transition not being fresh fuel and was conservatively taken as the lower bound (in reality, the lower bound exposure of limiting bundles is expected to be closer to ~15-20 GWd/MTU PPE for second cycle fuel at the beginning of the transition cycles). ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

The results of this study are shown in Figure 4-3. The figure shows the relative effect of the different studies relative to RXA analyses (i.e., case X minus the results of the Zirc-2 RXA case).

A negative value in melt margin indicates that the modeled case would predict higher fuel temperatures, and a negative value in transient strain margin indicates the modeled case would predict more transient cladding strain (i.e., in both cases this would indicate being closer to the licensing criterion). In this study, the limiting melt margin and strain values occur around ((` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

````````````````````````````````````````````````````````````````

````````````````````````````````````````````````````````````````````````````````````````````````

22

006N8642-NP Revision 1 Non - Proprietary Information

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

((

```

` ))

Figure 4-3: Impact of CWSR Properties on Melt Margin and Strain Compliance for a Limiting BWR/3 HPCI AOO The results of this study demonstrate ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) The use of the proposed models ensures GNF continues to use, at a minimum, the same amount of conservatism derived in the methodologies for ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )).

It is worth noting that the relative sensitivity shown above is not directly applicable to other regulatory criteria, such as no cladding liftoff due to rod internal pressure and creep collapse. These 23

006N8642-NP Revision 1 Non - Proprietary Information criteria are much more sensitive to the low stress creep rate of the cladding. As such, compliance to these criteria will continue to be demonstrated by the manufacturing fuel vendor.

24

006N8642-NP Revision 1 Non - Proprietary Information 5.0 ACRONYMS Acronym Explanation AOO Anticipated Operational Occurrence BWR Boiling Water Reactor CWSR Cold Worked Stress Relieved ESBWR Economic Simplified Boiling-Water Reactor GE General Electric GNF Global Nuclear Fuels HPCI High Pressure Coolant Injection IFE Institute for Energy Technology (parent company of Halden)

INEL Idaho National Engineering Laboratory INL Idaho National Laboratory L&C Limitations and Conditions LAR License Amendment Request LTR Licensing Topical Report MOP Mechanical Overpower NFD / NNFD Nippon Nuclear Fuel Development (known today as NFD)

NRC Nuclear Regulatory Commission P/M Predicted Over Measured PCI Pellet/Cladding Interaction PCMI Pellet/Cladding Mechanical Interaction PNL Pacific Northwest Laboratory (today known as PNNL)

PNNL Pacific Northwest National Laboratory PPE Peak Pellet Exposure PWR Pressurized Water Reactor 25

006N8642-NP Revision 1 Non - Proprietary Information Acronym Explanation RXA Recrystallized Annealed SE Safety Evaluation SER Safety Evaluation Report SPP Second Phase Particle SRP Standard Review Plan TCD Thermal Conductivity Degradation TER Technical Evaluation Report TMOL Thermal Mechanical Operating Limits TOP Thermal Overpower UO2 Uranium Dioxide WEC Westinghouse Electric Company WSTOL Worst-Tolerance 26

006N8642-NP Revision 1 Non - Proprietary Information

6.0 REFERENCES

1. Global Nuclear Fuel. The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance Part 3 - Application Methodology. NEDC-33258P-A Revision 2, October 2021.

2. Global Nuclear Fuel. The PRIME Model for Transient Analysis of Fuel Rod Thermal-Mechanical Performance. NEDC-33840P-A Revision 1, August 2017.

3. Global Nuclear Fuel. The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance Part 1 - Technical Bases. NEDC-33256P-A Revision 2, October 2021.

4. Global Nuclear Fuel. General Electric Standard Application for Reactor Fuel (GESTAR II). NEDE-24011-P-A Revision 31, November 2020.

5. IE Porter et. al. FAST-1.0: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients. PNNL-29720, March 2020.

6. KJ Geelhood et. al. MatLib-1.0: Nuclear Material Properties Library. PNNL-29728, March 2020.

7. INEL. MATPRO - Version 11 A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior. NUREG/CR-0497, February 1979.

8. HC Brassfield et. al. RECOMMENDED PROPERTY AND REACTION KINETICS DATA FOR USE IN EVALUATING A LIGHT-WATER-COOLED REACTOR LOSS-OF-COOLANT INCIDENT INVOLVING ZIRCALOY-4 OR 304-SS-CLAD UO2. GEMP-482, April 1968.

9. J. A. Christensen et. al. Melting Point of Irradiated Uranium Dioxide. WCAP-6065, 1965.

10. S. Kashibe, K. Une and K. Nogita. Formation and growth of intragranular fission gas bubbles in UO2 fuels with burnup of 6-83 GWd/t. Journal of Nuclear Materials, May 1993.

11. White, R.J. The development of grain-face porosity in irradiated oxide fuel. Journal of Nuclear Materials, October 2003.

12. Global Nuclear Fuel. The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance Part 2 - Qualification. NEDC-33257P-A Revision 2, October 2021.

13. Global Nuclear Fuel. The PRIME Model for Analysis of Fuel Rod Thermal-Mechanical Performance 2015 5-year Update. NEDC-33257P Supplement 1 Revision 0, August 2015.

14. GE Hitachi Nuclear Energy. Response to Portion of NRC Request for Additional Information Letter No. 110 - Related to ESBWR Design Certification Application - RAI Numbers 4.2-2 S03, 4.2-4 S02 and 4.8-6 S01. MFN 08-347, May 2008.

15. A. Hermann, M. Martin, P. Prschke and S. Yagnik. Ductility degradation of irradiated fuel cladding. Scientific Report 2000 Volume IV Nuclear Energy and Safety, ISSN 1423-7334. March 2001. pp.95-103.

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006N8642-NP Revision 1 Non - Proprietary Information

16. G.F. Rieger and D. Lee. Strength and Ductility of Neutron Irradiated and Textured Zircaloy-2. Zirconium in Nuclear Applications, ASTM STP 551. American Society Testing Materials, 1974. pp. 355-369.

17. Framatome. ATRIUM10XM Fuel Rod Thermal-Mechanical Evaluation with RODEX2A for Quad Cities and Dresden. ANP-3918NP Revision 0, ML21257A420, April 2021.

18. US NRC, FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT BAW-10247PA, REVISION 0, SUPPLEMENT 1P, REVISION 0, REALISTIC THERMAL-MECHANICAL FUEL ROD METHODOLOGY FOR BOILING WATER REACTORS SUPPLEMENT 1: QUALIFICATION OF RODEX4 FOR RECRYSTALLIZED ZIRCALOY-2 CLADDING. ML17129A009, May 2017.

19. R. Adamson and B. Cox. Impact of Irradiation on Material Performance. A.N.T.

International, January 2006.

20. F. Garzarolli, R. Adamson and P. Rudling. Optimization of BWR fuel rod cladding condition for high burnups. Proceedings of 2010 LWR Fuel Performance/TopFuel/WRFPM, September 2010.

21. L.W. Newman. The Hot Cell Examination of Oconee 1 Fuel Rods After Five Cycles of Irradiation. DOE/ET/34212-50, BAW-1874, October 1986.

22. R. Adamson, F. Garzarolli and C. Patterson. In-Reactor Creep of Zirconium Alloys.

A.N.T. International, September 2009.

28

ATTACHMENT 5 Global Nuclear Fuel -Americas, LLC Affidavit for Withholding

Global Nuclear Fuel - Americas, LLC AFFIDAVIT I, Brian R. Moore, state as follows:

(1) I am General Manager, Core & Fuel Engineering, 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 proprietary report 006N8642-P, Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels Revision 1, dated January 2022. GNF proprietary information in 006N8642-P Revision 1 is identified by a dotted underline inside double square brackets. ((This sentence is an example.{3})). GNF 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 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's competitors without a license from GNF 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 customer-funded development plans and programs, resulting in potential products to GNF;

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

006N8642-P Revision 1 Affidavit Page 1 of 3

Global Nuclear Fuel - Americas, LLC (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, 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, 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.

(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 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 justification for the use of PRIME and its associated licensed methodologies to demonstrate compliance to the no fuel melting and cladding transient strain criteria for non-GNF fuel designs for the GNF Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification and analyses requirements were achieved at a significant cost to GNF.

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 asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF'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 analytical methodology and includes development of the expertise to determine and apply 006N8642-P Revision 1 Affidavit Page 2 of 3

Global Nuclear Fuel - Americas, LLC 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. 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's competitive advantage will be lost if its competitors are able to use the results of the GNF 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 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 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 14th day of January 2022.

Brian R. Moore General Manager, Core & Fuel Engineering Global Nuclear Fuel - Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 Brian.Moore@ge.com 006N8642-P Revision 1 Affidavit Page 3 of 3

v t e Letter Sequence Request
EPID:L-2022-LLA-0121, Updated Final Safety Analysis Report (Ufsar), Revision 16 and Fire Protection Report (Fpr), Revision 25 (Open)
Initiation Request, Request, Request, Request Acceptance Supplement, Supplement Administration Questions 12 January 2023 Answers 11 January 2022 11 April 2022 23 January 2023 Results Approval, Approval Other: ML22335A565, ML23088A224
MONTHYEAR2022-09-13 [Table View]

RS-22-006, Request to Expand Applicability of Prime Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel

Text

Subject:

References:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

1.0 INTRODUCTION

6.0 REFERENCES

1.0 INTRODUCTION

6.0 REFERENCES

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