Sequoyah Nuclear Plant, Units 1 and 2, Application to Modify the Technical Specification to Allow for Transition to Westinghouse

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402

CNL-20-014

September 23, 2020

10 CFR 50.90

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject: Application to Modify the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09)

References: 1. Summary of November 14, 2018, Meeting with Tennessee Valley Authority Regarding Presubmittal Licensing Actions - Sequoyah Nuclear Plant, Units 1 and 2 (ML18344A567)

2. Summary of January 31, 2019, Meeting with Tennessee Valley Authority Regarding Presubmittal Licensing Actions - Sequoyah Nuclear Plant, Units 1 and 2 (ML19038A469)

3. Summary of February 20, 2020, Public Meeting with Tennessee Valley Authority Regarding the Future Submittal of License Amendment Requests RE: Fuel Transition and Operation with One Control Rod Removed (ML20055C231)

4. NRC Letter to TVA, “Sequoyah Nuclear Plant, Unit 1 - Issuance of Exigent Amendment No. 348 to Operate One Cycle With One Control Rod Removed (EPID L-2019-LLA-0239),” dated November 21, 2019 (ML19319C831)

5. NRC Letter to TVA, “Sequoyah Nuclear Plant, Unit 2 - Issuance of Exigent Amendment No. 342 to Operate One Cycle with One Control Rod Removed (EPID L 2020 LLA 0078),” dated April 23, 2020 (ML20108F049)

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2

U.S. Nuclear Regulatory Commission CNL-20-014 Page 2 September 23, 2020

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, “Application for amendment of license, construction permit, or early site permit,” Tennessee Valley Authority (TVA) is submitting a request for an amendment to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN) Units 1 and 2. This amendment is needed to facilitate transition to Westinghouse RFA-2 fuel with Optimized ZIRLOTM 1 cladding.

The proposed change revises these Technical Specifications (TSs):  TS 2.0 SAFETY LIMITS (SLs) TS 2.1.1, “Reactor Core SLs” Figure 2.1.1-1, “Reactor Core Safety Limit – Four Loops in Operation” is revised and relocated to the Core Operating Limits Report (COLR), Reactor Core Safety Limit 2.1.1.1 is revised to implement the methodology in the Westinghouse WRB-2M departure from nucleate boiling (DNB) correlation, and Reactor Core Safety Limit 2.1.1.2 is revised to implement the Westinghouse Performance Analysis and Design Model (PAD5) methodology for fuel melt temperature.  TS 3.1.4, “Rod Group Alignment Limits” is revised to be consistent with revised TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(Z))” Surveillance Requirement numbering.  TS 3.1.7, “Rod Position Indication” is revised to implement BEACONTM 2 core power distribution measurement.  TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(X,Y,Z))” is revised to reflect Standard Technical Specification (STS) 3.2.1 in NUREG-1431 and the methodology in WCAP-17661-P-A Revision 1, “Improved RAOC and CAOC FQ Surveillance Technical Specifications” with deviations to the Condition B Required Action Completion Times.  TS 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor F∆H(X,Y),” and 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” are revised to implement Westinghouse STS format and content consistent with NUREG-1431, and TS 3.2.4 is revised to implement BEACON core power distribution measurement.  TS 3.3.1, “RTS Instrumentation” is revised to implement BEACON core power distribution measurement and relocation of OT∆T and OP∆T setpoint parameter values to the COLR consistent with TSTF-339-A, Rev. 2, “Relocate TS Parameters to COLR,” and WCAP-14483-A, “Generic Methodology for Expanded Core Operating Limits Report.”  TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits” is revised to reflect a lower Thermal Design Flow rate and the relocation of parameter values to the COLR consistent with TSTF-339-A and WCAP-14483-A. ______

1Optimized ZIRLO is a trademark or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners. 2BEACON, FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

U.S. Nuclear Regulatory Commission CNL-20-014 Page 3 September 23, 2020

 TS 4.2.1, “Fuel Assemblies” is revised to add Optimized ZIRLO as a fuel assembly cladding material in accordance with WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO.”  TS 4.2.1, “Fuel Assemblies,” is revised to delete the reference to Framatome lead test assemblies and the corresponding BAW-2328 topical report.  TS 4.2.2, “Control Rod Assemblies” is revised to reflect 52 rod cluster control assembly (RCCAs) for Units 1 and 2.  TS 5.6.3, “Core Operating Limits Report” is revised to reflect Westinghouse core safety analysis methodologies and to replace the loss-of-coolant accident (LOCA) analysis evaluation model references with the FULL SPECTRUM™3 Loss-of-Coolant Accident (FSLOCA™3) Evaluation Model analysis applicable to SQN Units 1 and 2.

In addition, the following Operating License (OL) License Conditions are proposed for replacement with new restrictions associated with the mixed or transition cores.  Unit 1 OL, License Condition 2(C)25  Unit 2 OL, License Condition 2(C)18

In Reference 1, TVA proposed an alternative approach for a fuel transition license amendment. NRC provided feedback on this approach via public meeting in Reference 2, noting specific concerns related to the fuel transition to be included for review for the staff to independently verify the fuel transition could be implemented in conformance with NRC requirements. Therefore, in addition to the TS and OL changes discussed above, the proposed change presents the following to address the staff’s feedback:  The core configurations for the transition based on the expected transition and equilibrium core designs being presented.  The operating limits for the RFA-2 fuel and the Framatome HTP fuel in the transition cores.  The use of the Westinghouse methods for core design and for determining core operating limits in the transition cores and subsequent equilibrium cores containing only RFA-2 fuel.  The safety analyses, specifically, that these analyses satisfy the applicable conditions and limitations associated with the employed methods and that these analyses meet applicable safety analysis and regulatory acceptance criteria.

Enclosure 1 provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Attachments 1 and 2 to the enclosure provide the existing TS pages marked-up to show the proposed changes for

______3 FULL SPECTRUM and FSLOCA are trademarks of Westinghouse Electric Company LLC

U.S. Nuclear Regulatory Commission CNL-20-014 Page 4 September 23, 2020

SQN Unit 1 and Unit 2, respectively. Attachments 3 and 4 to the enclosure provide the SQN Unit 1 and Unit 2 TS pages retyped to show the proposed changes. Attachment 5 to the enclosure provides the existing SQN Unit 1 TS Bases pages marked-up to show the proposed changes. Only the Unit 1 TS Bases pages have been provided, as the Unit 2 changes will be nearly identical except for some editorial differences. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.

As discussed in References 4 and 5, both units have approved temporary Technical Specification changes to allow the RCCA in core location H-08 to be removed. As a result of the timing of the fuel transition, the request for permanent removal of the RCCA has been integrated into the attached amendment request.

Enclosure 1 also contains:  Attachment 6 – revised Core Operating Limits Report (COLR) Template for SQN Units 1 and 2 (For Information Only)  Attachment 7 – revised Technical Requirements Manual (TRM) (Mark-Ups) for SQN Unit 1 (For Information Only)  Attachment 8 – Compliance with Limitations and Conditions from NRC-Approved Newly Applied Topical Reports  Attachment 9 – Justification for Permanent Removal of SQN Units 1 and 2 RCCA H-08  Attachment 10 – Sequoyah Safety Analysis UFSAR Impact Summary for the WEC RFA-2 Fuel Transition

Enclosure 2 to this letter provides a summary of the SQN Units 1 and 2 LOCA Analysis with the FSLOCA Evaluation Methodology. This document contains information that Westinghouse Electric Company LLC considers to be proprietary in nature and subsequently, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 2.390, “Public inspections, exemptions, requests for withholding,” paragraph (a)(4), TVA requests that this proprietary information be withheld from public disclosure. Enclosure 3 contains a non-proprietary version of the summary of the SQN Units 1 and 2 LOCA Analysis with the FSLOCA Evaluation Methodology.

Enclosure 4 provides the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-20-5063 affidavit supporting this proprietary withholding request, signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission (“Commission”) and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Accordingly, TVA requests that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-20-5063 and should be addressed to Camille T. Zozula, Manager,

U.S. Nuclear Regulatory Commission CNL-20-014 Page 5 September 23, 2020

Regulatory Compliance & Corporate Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 1, Cranberry Township, Pennsylvania 16066.

In addition, as the Westinghouse RFA-2 fuel will have Optimized ZIRLO cladding, an exemption request has been included for NRC approval in Enclosure 5.

TVA has determined that there are no significant hazards considerations associated with the proposed amendment and TS changes. The proposed amendment and TS changes qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and attachments to the Division of Radiological Health – Tennessee State Department of Environment and Conservation.

As discussed in Reference 3, the technical evaluations, calculations and safety analyses for the transition to Westinghouse RFA-2 fuel have been completed and will be made available for regulatory audits to leverage the review efficiency as suggested by NRC in Reference 2.

TVA requests approval of the proposed license amendment within one year from the date of this submittal. The TS changes are to be implemented when the Westinghouse RFA-2 fuel is loaded for Cycle 26 in both SQN units. Approximate dates for implementation are November 1, 2022 for SQN Unit 1 and March 15, 2023 for SQN Unit 2.

There are no new regulatory commitments associated with this submittal. If you have any questions about this proposed change, please contact Gordon Williams, Senior Manager, Fleet Licensing (Acting) at (423) 751-2687.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 23rd day of September 2020.

Respectfully,

James Barstow Vice President, Nuclear Regulatory Affairs & Support Services

Enclosures cc: see Page 6

U.S. Nuclear Regulatory Commission CNL-20-014 Page 6 September 23, 2020

Enclosure 1: Evaluation of the Transition to Westinghouse RFA-2 Fuel Enclosure 2: Application of Westinghouse FULL SPECTRUM LOCA Evaluation Model to the Sequoyah Nuclear Plant (Proprietary) FSLOCA Enclosure 3: Application of Westinghouse FULL SPECTRUM LOCA Evaluation Model to the Sequoyah Nuclear Plant (Non-Proprietary) FSLOCA Enclosure 4: Affidavit Enclosure 5: Exemption Request cc (Enclosures)

NRC Regional Administrator – Region II NRC Resident Inspector – Sequoyah Nuclear Plant NRC Project Manager – Sequoyah Nuclear Plant Director, Division of Radiological Health – Tennessee State Department of Environment and Conservation

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel

Subject: Application to Modify the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09)

CONTENTS

1.0 Summary Description ...... 1 2.0 Detailed Description ...... 2 2.1 Background ...... 2 2.2 Need for Proposed Changes ...... 3 2.3 Proposed Changes ...... 3 2.4 Condition Intended to Resolve ...... 6 3.0 TECHNICAL EVALUATION ...... 7 3.1 System Description ...... 7 3.1.1 Seismic/LOCA Impact on Fuel Assemblies ...... 8 3.1.2 Safety Analysis Impact Summary ...... 9 3.1.3 Source Terms ...... 9 3.1.4 Nuclear Core Design ...... 9 3.2 Technical Analysis ...... 27 3.2.1 TS 2.0 Safety Limits Changes ...... 27 3.2.2 TS 3.1 Reactivity Control Systems Changes ...... 30 3.2.3 TS 3.2 Power Distribution Limits Changes ...... 32 3.2.4 TS 3.3.1 Reactor Trip System (RTS) Instrumentation Changes ...... 39 3.2.5 TS 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Changes ...... 41 3.2.6 TS 4.2.1 Fuel Assemblies Changes ...... 43 3.2.7 TS 4.2.2 Control Rod Assemblies Changes ...... 44 3.2.8 TS 5.6.3 Core Operating Limits Report Changes ...... 44 3.2.9 Operating License Conditions 2.C (25) and 2.C (18) Changes ...... 49 3.3 Conclusion ...... 49 4.0 REGULATORY EVALUATION ...... 50 4.1 Applicable Regulatory Requirements/Criteria ...... 50 4.1.1 Regulations ...... 50 4.1.2 General Design Criteria ...... 51 4.1.3 Regulatory Guidance and Miscellaneous References ...... 53 4.2 Precedent ...... 54 4.3 No Significant Hazards Consideration ...... 55

Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel

4.4 Conclusions ...... 60 5.0 ENVIRONMENTAL CONSIDERATION ...... 61 6.0 REFERENCES ...... 61

ATTACHMENTS 1. Proposed TS Changes Mark-Ups for SQN Unit 1 2. Proposed TS Changes Mark-Ups for SQN Unit 2 3. Proposed TS Changes (Final Typed) for SQN Unit 1 4. Proposed TS Changes (Final Typed) for SQN Unit 2 5. Revised TS Bases Page Changes (Mark-Ups) for SQN Unit 1 (For Information Only) 6. Revised Core Operating Limits Report (COLR) Template for Units 1 and 2 (For Information Only) 7. Revised Technical Requirements Manual (TRM) (Mark-Ups) for Units 1 and 2 (For Information Only) 8. Compliance with Limitations and Conditions from NRC-Approved Newly Applied Topical Reports 9. Justification for Permanent Removal of Units 1 and 2 RCCA H08 10. Sequoyah Safety Analysis UFSAR Impact Summary for the WEC RFA-2 Fuel Transition

Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel

1.0 SUMMARY DESCRIPTION

This evaluation supports a request to amend Renewed Facility Operating License No. DPR-77 for the Sequoyah Nuclear Plant (SQN) Unit 1, and Renewed Facility Operating License No. DPR-79 for the SQN Unit 2.

The proposed amendment revises the following SQN Units 1 and 2 Technical Specifications (TSs):

 TS 2.0 SAFETY LIMITS (SLs) TS 2.1.1, “Reactor Core SLs” Figure 2.1.1-1, “Reactor Core Safety Limit – Four Loops in Operation,” is revised and relocated to the Core Operating Limits Report (COLR), Reactor Core Safety Limit 2.1.1.1 is revised to implement the methodology in the Westinghouse WRB-2M departure from nucleate boiling (DNB) correlation, and Reactor Core Safety Limit 2.1.1.2 is revised to implement the Westinghouse Performance Analysis and Design Model (PAD5) methodology for fuel melt temperature.  TS 3.1.4, “Rod Group Alignment Limits,” is revised to be consistent with revised TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(Z))” Surveillance Requirement numbering.  TS 3.1.7, “Rod Position Indication,” is revised to implement BEACONTM 4 core power distribution measurement.  TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(XY,Z)),” is revised to reflect STS 3.2.1 in NUREG-1431 and the methodology in WCAP-17661-P-A Revision 1, “Improved RAOC and CAOC FQ Surveillance Technical Specifications” with deviations to the Condition B Required Action Completion Times.  TS 3.2.2, “F∆H(X,Y),” and 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” are revised to implement Westinghouse Standard Technical Specification (STS) format and content consistent with NUREG-1431, and TS 3.2.4 is revised to implement BEACON core power distribution measurement.  TS 3.3.1, “RTS Instrumentation” is revised to implement BEACON core power distribution measurement and relocation of OT∆T and OP∆T setpoint parameter values to the COLR consistent with TSTF-339-A, Rev. 2, “Relocate TS Parameters to COLR,” and WCAP-14483-A, “Generic Methodology for Expanded Core Operating Limits Report.”  TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” is revised to reflect a lower Thermal Design Flow rate and the relocation of parameter values to the COLR consistent with TSTF-339-A and WCAP-14483-A.  TS 4.2.1, “Fuel Assemblies” is revised to add Optimized ZIRLOTM 5 as a fuel assembly cladding material in accordance with WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO.”  TS 4.2.1, “Fuel Assemblies,” is revised to delete the reference to Framatome lead test assemblies and the corresponding BAW-2328 topical report.  TS 4.2.2, “Control Rod Assemblies” is revised to reflect 52 RCCAs for Units 1 and 2. ______

4BEACON, FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners. 5Optimized ZIRLO is a trademark or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners. CNL-20-014 E1 1 of 63 Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel x TS 4.2.2, “Control Rod Assemblies” is revised to reflect 52 RCCAs for Units 1 and 2. x TS 5.6.3, “Core Operating Limits Report” is revised to reflect Westinghouse core safety analysis methodologies and to replace the loss-of-coolant accident (LOCA) analysis evaluation model references with the FULL SPECTRUM™6 Loss-of-Coolant Accident (FSLOCA™6) Evaluation Model analysis applicable to SQN Units 1 and 2.

The proposede also chan revisesg the S QN Units 1 and 2 Operating License (OL) to replace OL condition 2.C(25) and 2.C(18), respectively. These Conditions will no longer be applicable with Westinghouse fuel in the core. However, during the transition or mixed cores, new penalties will be established, as discussed in Section 3.2.8 below.

The proposed changes to the SQN Units 1 and 2 TSs are consistent with NUREG-1431, “Standard Technical Specifications Westinghouse Plants,” Revision 4.

Attachments 1 and 2 to this Enclosure provide the TS pages marked-up to show the proposed changes for SQN Unit 1 and Unit 2, respectively. Attachments 3 and 4 to the enclosure provide the SQN Unit 1 and Unit 2 TS pages retyped to show the proposed changes.

Attachment 5 to the enclosure provides the existing SQN Unit 1 TS Basese pag s marked-up to show the proposed changes. Only the Unit 1 TS Bases pages have been provided, as the Unit 2 changes will be nearly identical except for some editorial differences. Changes to the existing TS Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.

The enclosure also includes: x Attachment 6 – revised Core Operating Limits Report (COLR) Template for SQN Units 1 and 2 (For Information Only) x Attachment 7 – revised Technical Requirements Manual (TRM) (Mark-Ups) for SQN Unit 1 (For Information Only) x Attachment 8 – Compliance with Limitations and Conditions from NRC-Approved Newly Applied Topical Reports x Attachment 9 – Justification for Permanent Removal of SQN Units 1 and 2 RCCA H08 x Attachment 10 – Sequoyah Safety Analysis UFSAR Impact Summary for the WEC RFA-2 Fuel Transition

2.0 DETAILED DESCRIPTION

2.1 Background

Tennessee Valley Authority (TVA) plans to transition Sequoyah Units 1 and 2 from Framatome-supplied high thermal performance (HTP) fuel to Westinghouse 17x17 Robust Fuel Assembly-2 (RFA-2), commencing with Unit 1 Cycle 26 and Unit 2 Cycle 26. The 17x17 RFA-2 fuel for SQN Units 1 and 2 includes the following features: x Integral fuel burnable absorbers (IFBAs) x Robust protective grid (RPG) ______6 FULL SPECTRUM and FSLOCA are trademarks of Westinghouse Electric Company LLC CNL-20-014 E1 2 of 63 Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel x Standardized debris filter bottom nozzle (SDFBN) x High-burnup bottom grid x Debris mitigating long fuel rod bottom end plugs x Wet annular burnable absorbers (WABA) x Removable top nozzle (RTN) x Three ZIRLO intermediate flow mixer (IFM) grids x Six ZIRLO RFA-2 structural mid-grids x Reduced rod bow (RRB) INCONEL® top grid x Optimized ZIRLO high-performance fuel cladding with a coated cladding feature x Thicker-walled guide thimble and instrumentation tubes to improve fuel assembly stiffness and to address incomplete rod insertion (IRI) considerations.

The 17x17 RFA-2 fuel is similar to the fuel in use at North Anna Units 1 and 2. Except for Optimized ZIRLO cladding, the 17x17 RFA-2 fuel is similar to the fuel in several other plants, including Millstone 3, Seabrook, Beaver Valley, and Watts Bar. An exemption request for the use of Optimized ZIRLO cladding at Sequoyah is included with this amendment request as Enclosure 5.

The structural mid-grid design used in the RFA-2 fuel assembly is a modification of the low-pressure-drop mid-grid design that was accepted by the NRC for use in the VANTAGE-5-Hybrid (V5H) fuel assembly design (Reference 1). The RFA-2 mid-grid design was the culmination of several changes evaluated by means of the NRC-approved Fuel Criteria Evaluation Process (FCEP) (Reference 2). By complying with the requirements of the FCEP, it has been demonstrated that the mid-grid design meets all design criteria of the tested mid-grids that form the basis of the WRB-2M correlation database and that the WRB-2M correlation applies to the new RFA-2 mid-grid as well as the IFM grid (References 3, 4, and 5).

The proposed transition to Westinghouse 17x17 RFA-2 fuel for SQN Units 1 and 2 will require changes to the TS and OL as discussed below.

2.2 Need for Proposed Changes

The proposed TS and OL changes and the exemption request are needed for the SQN Unit 1 and 2 transition to Westinghouse RFA-2 fuel with Optimized ZIRLO cladding, which includes the Westinghouse transition core methodology, Westinghouse core safety analysis methodology, the FSLOCA Evaluation Model, the BEACON core monitoring capability, and the PAD5 fuel performance analysis and design model. This license amendment request (LAR) also includes a request for permanent removal of a rod cluster control assembly (RCCA) at core location H-08 for both units.

2.3 Proposed Changes

The proposed change revises the SQN Units 1 and 2 TS and OL as follows.

TS 2.1.1 Reactor Core Safety Limits (SL) Changes

The proposed change would revise TS 2.1.1, “Reactor Core SLs” Figure 2.1.1-1, “Reactor Core Safety Limit – Four Loops in Operation” which specifies the acceptable operation

CNL-20-014 E1 3 of 63 Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel

domain with consideration given to the fraction of Rated Thermal Power, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure. The Reactor Core Safety Limit Figure would be revised based upon the application of Nuclear Regulatory Commission (NRC) approved WCAP-11397-P-A, “Revised Thermal Design Procedure” (Reference 6) evaluation methodology. The proposed change to TS 2.1.1 would also relocate the revised figure to the COLR consistent with NRC-approved TSTF-339-A, Rev. 2, “Relocate TS Parameters to COLR” (Reference 7), and NRC-approved WCAP-14483-A, “Generic Methodology for Expanded Core Operating Limits Report” (Reference 8). Relocating the Reactor Core Safety Limit Figure from TS 2.1.1 to the COLR will also be consistent with TS 2.1.1 in NUREG-1431, “Standard Technical Specifications Westinghouse Plants” (Reference 9).

The proposed change would revise Reactor Core Safety Limit 2.1.1.1 based upon the application of NRC approved WCAP-15025-P-A, “Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids” (Reference 10) evaluation methodology. The change would replace the three Framatome departure from nucleate boiling (DNB) correlation values with a single WRB-2M DNB correlation value.

The proposed change would revise Reactor Core Safety Limit 2.1.1.2 based upon the application of NRC approved WCAP-17642-P-A, Revision 1, “Westinghouse Performance Analysis and Design Model (PAD5)” (Reference 11) evaluation methodology. The change would replace the peak fuel centerline temperature derived from Framatome methodology with the burnup-dependent Westinghouse PAD5 peak fuel centerline temperature limit.

TS 3.1 REACTIVITY CONTROL SYSTEMS Changes

The proposed change would revise TS 3.1.4, “Rod Group Alignment Limits,” Required Action B.2.4 to add SR 3.2.1.2, consistent with changes to TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(X,Y,Z)).” Surveillance Requirements (SRs) 3.2.1.1 and 3.2.1.2 would be revised consistent with TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(Z)),” from NRC-approved WCAP-17661-P-A Revision 1, “Improved RAOC and CAOC FQ Surveillance Technical Specifications” (Reference 12) which incorporated improvements to the relaxed axial offset control (RAOC) version of TS 3.2.1B from NUREG-1431, Revision 4 (Reference 9). Steady state and transient FQ surveillances are also consistent with the RAOC version of TS 3.2.1 in Reference 9.

The proposed change would revise TS 3.1.7, “Rod Position Indication,” Required Actions A.1, A.2.1, B.3, and C.1 to implement BEACON core power distribution measurement consistent with NRC-approved WCAP-12472-P-A, “BEACON Core Monitoring and Operations Support System” (Reference 13), and WCAP-12472-P-A, Addendum 4-A, Revision 0, “BEACON Core Monitoring and Operation Support System, Addendum 4” (Reference 14), and replace the discussion of flux maps obtained solely using the movable incore detectors with more generic terminology (i.e., “core power distribution measurement information”).

TS 3.2 POWER DISTRIBUTION LIMITS Changes

The proposed change would revise TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(X,Y,Z)),” based upon STS 3.2.1, “Heat Flux Hot Channel Factor (FQ(Z)),” from Reference 9 as revised by NRC-approved WCAP-17661-P-A, Appendix A for RAOC plants, with deviations based on

CNL-20-014 E1 4 of 63 Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel

References 30 and 31. The change would also implement BEACON core power distribution measurement consistent with WCAP-12472-P-A,” and WCAP-12472-P-A, Addendum 4-A.

The proposed change would revise TS 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor Fǻ+(X,Y),” to be consistent with STS 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor N (F 'H)” in Reference 9.

The proposed change would revise TS 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” Required Actions A.3 and A.6 to be consistent with the Surveillance Requirement numbering in revised TS 3.2.1 and revised TS 3.2.2. SR 3.2.4.2 would be changed to implement BEACON core power distribution measurement consistent with WCAP-12472-P-A and WCAP-12472-P-A, Addendum 4-A, and replace the discussion of flux maps obtained solely using the movable incore detectors with more generic terminology (i.e., “core power distribution measurement information”).

TS 3.3.1 RTS Instrumentation Changes

The proposed change would revise TS 3.3.1, “RTS Instrumentation,” SRs 3.3.1.3 and 3.3.1.6 to implement BEACON core power distribution measurement consistent with NRC-approved WCAP-12472-P-A and WCAP-12472-P-A, Addendum 4-A. Table 3.3.1-1 Note 1, ³2YHUWHPSHUDWXUHǻ7´ZRXOGEHUHYLVHGWRLQFUHDVHWKH.2 DQG.3 FRHIILFLHQWVLQWKH27ǻ7 reactor trip function .2 ZLOOEHLQFUHDVHGIURPq)WRq).3 will be increased from 0.00055/psig to 0.0008/psig) using NRC-approved WCAP-8745-P-A, “Design Bases for the 7KHUPDO2YHUSRZHU¨7DQG7KHUPDO2YHUWHPSHUDWXUH¨77ULS)XQFWLRQV´(Reference 15) evaluation methodology.

Table 3.3.1-1RWHVDQGZRXOGEHUHYLVHGWRUHORFDWH27ǻ7DQG23¨7SDUDPHWHUYDOXHV (nominal RCS average temperature and operating pressure at full power, LaPlace transform FRQVWDQWV.1 WKURXJK.6DQG3URFHVV3URWHFWLRQ6\VWHPFDUGGHOD\WLPHVĲ1 WKURXJKĲ5) to the COLR consistent with NRC approved TSTF-339-A, Rev. 2, and WCAP-14483-A.

TS 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Changes

The proposed change would revise TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to remove the specified Minimum Measured Flow rate of 378,400 gpm and insert the revised Thermal Design Flow rate of 360,000 gpm. The proposed change would relocate the current parameter values from LCO 3.4.1 and SRs 3.4.1.1 – 3.4.1.4 to the COLR. The LCO 3.4.1, SR 3.4.1.3, and SR 3.4.1.4 RCS total flow rate is revised from ³JSP´WRWKH7KHUPDO'HVLJQ)ORZ 7') YDOXHRI³JSP” in conjunction with “and greater than or equal to the limit specified in the COLR.” This change is consistent with TSTF-339-A, Rev. 2, WCAP-14483-A, and NUREG-1431.

TS 4.2.1 Fuel Assemblies Changes

The proposed change would revise TS 4.2.1, “Fuel Assemblies” to add Optimized ZIRLO as a fuel assembly cladding material in accordance with NRC-approved WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO” (Reference 16). The Optimized ZIRLO clad material contains a lower tin content than ZIRLO cladding, thereby reducing corrosion. An

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Evaluation of the Transition to Westinghouse RFA-2 Fuel exemption to 10 CFR 50.46 is also requested in Enclosure 5 consistent with WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A. The proposed change would also delete the discussion of Framatome lead test assemblies and report BAW-2328 consistent with the proposed implementation of Westinghouse core safety analysis methodology.

TS 4.2.2 Control Rod Assemblies

The proposed change would revise TS 4.2.2, “Control Rod Assemblies” to require 52 rod cluster control assemblies (RCCAs) with no full-length control rod assembly in core location H-08 for Units 1 and 2. This change would also remove any cycle-specific restraints associated with this configuration. This proposed change is discussed in Attachment 9 to this Enclosure.

TS 5.6.3 Core Operating Limits Report Changes

The proposed change would revise TS 5.6.3.a, “Core Operating Limits Report,” to add additional core operating limits for Limiting Conditions for Operation (LCOs) 2.1.1, 3.1.4, 3.1.8, 3.3.1, and 3.4.1. In addition, the existing LCO titles will be revised consistent with the proposed TS changes discussed above. TS 5.6.3.b is revised to change analytical methods used to determine the core operating limits from Framatome methods to Westinghouse core safety analysis methodology references. This includes the application of the FSLOCA Evaluation Methodology (Reference 17) to evaluate the peak cladding temperatures for SQN Units 1 and 2 large-break and small-break LOCAs (LBLOCA and SBLOCA).

Renewed License No. DPR-77 and DPR-79

The proposed change also revises the SQN Units 1 and 2 Operating License (OL) to replace OL condition 2.C (25) and 2.C (18) respectively, as a result of the implementation of Westinghouse core safety analysis methodology. Specifically, the following license condition is proposed. When the Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies the: N x HTP fuel assemblies F ǻ+ shall be maintained less than 1.61. x RFA-2 fuel assemblies the DNBR limit shall be reduced by: — 0.25% for the WRB-2M critical heat flux correlation — 0.50% for the ABB-NV critical heat flux correlation

2.4 Condition Intended to Resolve

The proposed change will allow TVA to use Westinghouse core safety analysis methodologies as part of a transition from Framatome-supplied HTP fuel to Westinghouse 17x17 RFA-2 fuel with Optimized ZIRLO cladding. This includes the use the FSLOCA Evaluation Model to evaluate the peak cladding temperatures for LBLOCAs and SBLOCAs for SQN Units 1 and 2. The proposed change will allow TVA to operate with 52 RCCAs and no full-length control rod assembly in core location H-08 for Units 1 and 2 for the fuel transition and subsequent cycles.

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

3.0 TECHNICAL EVALUATION

3.1 System Description

Beginning in Cycle 9, both SQN units were reloaded with Mark-BW fuel supplied by Framatome. Beginning with Cycle 19 for Unit 2 and Cycle 20 for Unit 1, each SQN unit was loaded with Advanced W17 high thermal performance (HTP) fuel supplied by AREVA NP, Inc. (Framatome). The Advanced W17 HTP fuel assembly consists of 264 fuel rods, 24 guide tubes and one instrument tube in a 17x17 square array. The guide tubes are annealed Zircaloy-4 and provide guidance for control rod insertion. The fuel assembly contains 7 Zircaloy-4 spacer grid assemblies including the uppermost grid and 1 nickel alloy 718 lowermost spacer grid assembly. The Advanced W17 HTP fuel assembly also includes 3 Zircaloy-4 intermediate flow mixing spacer grid assemblies. The bottom nozzle is a proven debris resistant design.

The basic design parameters of the Framatome Mark-BW and Advanced W17 HTP fuel assemblies are comparable to those of the Westinghouse Standard and Vantage 5H 17x17 fuel assemblies. The Mark-BW and Advanced W17 HTP fuel assemblies incorporate proven design features while maintaining compatibility with the Westinghouse reactor internals and fuel assemblies.

TVA plans to transition SQN Units 1 and 2 from Framatome-supplied HTP fuel to Westinghouse 17x17 Robust Fuel Assembly-2 (RFA-2), commencing with Unit 1 Cycle 26 and Unit 2 Cycle 26. The results of evaluations and analyses performed to confirm the acceptable transition to Westinghouse 17x17 RFA-2 fuel for SQN Units 1 and 2 are provided in this enclosure and supporting attachments.

Based on a comparison between the 17x17 V5H and RFA-2 designs, it was determined that the RFA-2 design is compatible with the HTP design. This was determined based on the compatibility of the RFA-2 fuel with the V5H fuel, explicit analysis, and the previous experience of mixed cores of V5H and Mark-BW fuel, and the ability for Sequoyah to adequately transition from Framatome Mark-BW fuel to HTP fuel. The changes between the V5H and RFA-2 fuel do not impact the compatibility of the RFA-2 fuel with the HTP fuel. The 17x17 RFA-2 fuel assembly is concluded to be mechanically and hydraulically compatible with the Framatome 17x17 HTP fuel assembly in full or transition cores since the Framatome 17x17 Mark-BW fuel assembly was once mixed with the 17x17 V5H design.

Westinghouse has established and obtained NRC approval for the design bases for several PWR fuel system designs (e.g., WCAP-9500-A, “Reference Core Report – 17 x 17 Optimized Fuel Assembly,” May 1982, WCAP 10444-P-A, “Westinghouse Reference Core Report – VANTAGE 5 Fuel Assembly,” September 1985, and WCAP-10125-P-A, “Extended Burnup Evaluation of Westinghouse Fuel,” December 1985). These design bases, or Specified Acceptable Fuel Design Limits (SAFDLs), and their respective evaluations comply with the “Acceptance Criteria” of SRP 4.2, “Fuel System Design,” and as such provide assurance that the fuel system is mechanically designed to perform safely. The plant specific requirements outlined in the NRC’s Safety Evaluation Reports (SER) for WCAP-9500-A and WCAP-10444-P-A will be adhered to in applying the Fuel Criteria Evaluation Process (FCEP – as described in WCAP-12488-A) to plant specific applications.

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

The NRC-approved generic design criteria used to assess the performance of the RFA-2 fuel assemblies were developed to satisfy certain objectives as described in Reference 2 (WCAP-12488-A, “Westinghouse Fuel Criteria Evaluation Process,” October 1994). These objectives are used for designing fuel assemblies to provide the following assurances. x The fuel assembly (system) shall not fail as a result of normal operation and anticipated operational occurrences. The fuel assembly (system) dimensions shall be designed to remain within operational tolerances and the functional capabilities of the fuels shall be established to either meet or exceed those assumed in the safety analysis. x Fuel assembly (system) damage shall never prevent control rod insertion when it is required. x The number of fuel rod failures shall be conservatively estimated for postulated accidents. x Fuel coolability shall always be maintained. x The mechanical design of fuel assemblies shall be compatible with co-resident fuel and the reactor core internals.

The generic criteria are applied to the fuel rod and fuel assembly designs.

3.1.1 Seismic/LOCA Impact on Fuel Assemblies

The evaluation of the 17x17 RFA-2 fuel and Framatome high thermal performance (HTP) fuel assembly structural integrity in both mixed transition cores and full 17x17 RFA-2 core was performed with the established method focusing on the Beginning-of-Life (BOL) condition, as presented in WCAP-12610-P-A. Additionally, a fuel assembly analysis to address the RFA-2 fuel at the End-of-Life (EOL) condition has also been performed according to PWROG 16043-P-A, Rev. 2. The seismic analysis considered the safe shutdown earthquake (SSE) and operating basis earthquake (OBE) conditions. The LOCA analysis considered two different pipe break situations: accumulator (ACC) line break and pressurizer (PZR) surge line break. The seismic/LOCA analysis results were obtained using the time history numerical integration technique. The maximum grid impact forces obtained from both seismic and LOCA transient cores were combined using the square root sum of the squares (SRSS) method. The maximum grid impact forces were compared to the allowable grid crush strength.

The detailed site-specific seismic and LOCA fuel assembly analyses have been performed in accordance with NRC-approved methodologies: WCAP 12610-P-A (Reference 32), WCAP 12488-A (Reference 2), and WCAP-9401 (Reference 18). The standard BOL analysis was supplemented by fuel assembly seismic/LOCA analysis to address RFA-2 fuel at the EOL condition in both full core (RFA-2 fuel only) and co-resident (with Framatome fuel) cores using methodology approved by the NRC in PWROG-16043-P-A (Reference 19).

The analysis predicted no permanent grid deformation (grid crush) to occur in both homogeneous core and mixed cores under combined seismic and LOCA loadings. Because there was no thimble tube damage observed during the grid crush testing and the stress analysis shows that no fragmentation occurs for the fuel rods under the combined seismic and LOCA loads, it is concluded that the RCCA insertability is maintained and coolable geometry is maintained.

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

An additional seismic and LOCA analysis was performed for a mixed core of HTP and RFA-2 fuel per the Framatome method of evaluation described in Reference 29. This analysis determines the impact loads on the HTP fuel in the mixed core and demonstrates that coolable geometry, control rod insertability, and fuel rod integrity are maintained for the HTP fuel.

3.1.2 Safety Analysis Impact Summary

As expected, the majority of the Updated Final Safety Analysis Report (UFSAR) safety analyses are revised to reflect the inputs and results from applying the Westinghouse methods and codes used to support the RFA-2 fuel, in addition to applying the FSLOCA and PAD5 methods. Attachment 10 (Sequoyah Safety Analysis UFSAR Impact Summary for the WEC RFA-2 Fuel Transition) summarizes the impact to the UFSAR safety analysis.

3.1.3 Source Terms

To address the transition to Westinghouse fuel, an assessment of the design basis source term (core activity inventory) for SQN Units 1 and 2 has been performed. The assessment consisted of comparing the current analysis of record (AOR) source key term input parameters to corresponding key input parameters for the representative transition and equilibrium fuel cycles associated with the reload transition.

The core loading parameters (core thermal power level, Uranium mass, cycle average burnup, Uranium-235 fuel enrichment) of the representative reload transition cycles were compared to the AOR core loading parameters. The AOR core loading parameters are bounding for all inputs based on conservatisms incorporated into the AOR, such as core thermal power which is directly proportional to the nuclide-specific activities.

It is concluded that the AOR core activity inventory remains applicable for the transition to Westinghouse fuel. This is due mainly to the similar and/or conservative input parameters applied to the AOR calculations.

3.1.4 Nuclear Core Design

Representative core designs were developed for the first two transition cycles that include Framatome fuel. If future designs desire to utilize HTP fuel, the requirements will be similar (if not identical) to the requirements during the first and second transition, which can be dealt with on a case by case basis. An equilibrium cycle design with only Westinghouse fuel was used to approximate subsequent reloads. Note that the core designs utilized are not intended to represent limiting designs, but were instead developed with the intent to determine if sufficient margin exists between typical parameter values and the corresponding safety analysis limits to allow flexibility in designing future cores. The margins for the Sequoyah core designs were compared to the values for recent Watts Bar reload cycles to evaluate the adequacy of margins between typical safety parameter values and the corresponding limits. As mentioned previously, cycle-specific calculations will confirm that the actual values are within the safety analysis limits.

Additional core designs were developed, apart from those already mentioned, for use in the FSLOCA methodology limit setting. These cores exhibit unrealistically high peaking factors which are intended to bound future reload peaking factors. These core designs are fully

CNL-20-014 E1 9 of 63 Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel representative of the range of core characteristics expected during and after the fuel transition from Framatome HTP fuel to Westinghouse RFA-2 fuel.

Further, two transition core designs in addition to two equilibrium cycles were developed as representative core designs to move Sequoyah from Framatome HTP fuel to Westinghouse RFA-2 fuel. The purpose of the loading patterns is to determine acceptability of core safety parameters with representative designs. The loading patterns were developed based on typical cycle energy requirements with current operating parameters for power, flow, and temperatures. The loading patterns are representative of both Sequoyah Units 1 and 2.

The transition cycles contain both Framatome HTP and Westinghouse RFA-2 fuel. The first transition cycle feeds 85 assemblies of Westinghouse RFA-2 fuel with both once and twice burned Framatome HTP fuel. A higher than usual feed batch is utilized to maximize the population of Westinghouse RFA-2 fuel in the first transition core. The second transition cycle includes feed and once burned Westinghouse RFA-2 fuel and twice burned Framatome HTP fuel. The two equilibrium cycles are representative loading patterns once the entire core contains Westinghouse RFA-2 fuel.

Figures 1 through 4 depict the representative loading patterns that were utilized for the reload transition safety analysis. Figures 5 through 8 show cycle characteristics of critical boron, axial RIIVHW)4DQG)ǻ+UHVSHFWLYHO\. Figures 9 through 16 show the beginning of cycle and end of cycle power and burnup distributions for each loading pattern.

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

Figure 1: Representative First Sequoyah Transition Loading Pattern

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

Figure 2: Representative Second Sequoyah Transition Loading Pattern

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

Figure 3: Representative 77 Feed Sequoyah Equilibrium Loading Pattern

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

Figure 4: Representative 76 Feed Sequoyah Equilibrium Loading Pattern

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

1400

1200

1000

800 Equilibrium 2 Equilibrium 1 600 Transition 2

Critical Boron (ppm) Boron Critical Transition 1 400

200

0 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 Burnup (MWD/MTU)

Figure 5: Critical Boron Concentrations for Transition and Equilibrium Cycles

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

6 Equilibrium 2 4 Equilibrium 1 Transition 2 2 Transition 1 HFP Axial OFfset (%)

0 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000

-2

-4 Burnup (MWD/MTU)

Figure 6: HFP Axial Offset Comparisons for Transition and Equilibrium Cycles

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

1.8

1.75

1.7 Equilibrium 2

FQ Equilibrium 1 1.65 Transition 2 Transition 1

1.6

1.55 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 Burnup (MWD/MTU)

Figure 7: FQ Comparisons for Transition and Equilibrium Cycles

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

1.48

1.46

1.44

1.42 Equilibrium 2 H ǻ

F Equilibrium 1 1.4 Transition 2 Transition 1 1.38

1.36

1.34 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 Burnup (MWD/MTU)

)LJXUH)ǻ+&RPSDULVRQVIRU7UDQVLWLRQDQG(TXLOLEULXP&\FOHV

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

H G F E D C B A

TR1 P1 TR1 P1 TR1 P1 TR1 P2 1.297 1.172 1.271 1.234 1.344 1.122 1.067 0.472 8 1.385 1.239 1.395 1.309 1.453 1.214 1.264 0.736 0 21464 0 21404 0 22499 0 31839 P1 TR1 P1 P1 P1 TR1 TR1 P2 1.171 1.346 1.181 1.229 1.228 1.138 1.06 0.44 9 1.239 1.415 1.257 1.302 1.311 1.257 1.242 0.72 21486 0 22845 21451 19559 0 0 33987 TR1 P1 TR1 P1 TR1 P1 TR1 P2 1.269 1.179 1.313 1.262 1.233 1.088 1.09 0.407 10 1.393 1.256 1.402 1.341 1.397 1.16 1.305 0.712 0 22835 0 19292 0 22761 0 35861 P1 P1 P1 TR1 P1 TR1 TR1 P2 1.231 1.223 1.257 1.275 1.11 1.143 1.013 0.313 11 1.306 1.293 1.336 1.398 1.197 1.266 1.265 0.641 21400 21615 19427 0 20886 0 0 41560 TR1 P1 TR1 P1 P1 TR1 P1 1.339 1.221 1.228 1.107 1.119 1.16 0.607 12 1.449 1.303 1.391 1.195 1.188 1.332 0.899 0 19704 0 20958 21516 0 22544 P1 TR1 P1 TR1 TR1 TR1 P3 1.12 1.132 1.082 1.143 1.166 1.048 0.323 13 1.211 1.252 1.155 1.266 1.334 1.342 0.666 22448 0 22845 0 0 0 36726 TR1 TR1 TR1 TR1 P1 P2 1.065 1.055 1.083 1.012 0.618 0.371 14 1.261 1.238 1.298 1.262 0.898 0.771 0 0 0 0 22225 34098 P2 P2 P2 P2 0.471 0.437 0.395 0.31 15 0.734 0.722 0.676 0.635 31889 33649 37102 41846

Region Assembly Power Max Rod Power Assembly Burnup

Figure 9: BOC Power and Burnup Distributions for Transition 1

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

H G F E D C B A

TR1 P1 TR1 P1 TR1 P1 TR1 P2 1.304 1.019 1.291 0.975 1.303 1.059 1.216 0.549 8 1.341 1.059 1.348 1.028 1.362 1.101 1.33 0.809 25914 42184 25184 41514 26587 44688 24088 42063 P1 TR1 P1 P1 P1 TR1 TR1 P2 1.019 1.286 0.978 0.955 1.049 1.29 1.223 0.516 9 1.059 1.328 1.039 1.003 1.131 1.342 1.352 0.789 42201 25665 42814 41247 41479 26060 24182 43564 TR1 P1 TR1 P1 TR1 P1 TR1 P2 1.291 0.978 1.257 1.049 1.343 1.063 1.227 0.48 10 1.348 1.039 1.312 1.127 1.39 1.108 1.361 0.786 25177 42795 25173 41110 26896 44732 24129 44673 P1 P1 P1 TR1 P1 TR1 TR1 P2 0.975 0.954 1.047 1.297 1.025 1.313 1.129 0.373 11 1.028 1.001 1.126 1.339 1.068 1.369 1.334 0.707 41493 41352 41195 26202 41865 26096 21892 48302 TR1 P1 TR1 P1 P1 TR1 P1 1.304 1.048 1.342 1.024 1.011 1.267 0.655 12 1.363 1.13 1.39 1.067 1.063 1.369 0.935 26559 41560 26849 41913 42368 24949 35012 P1 TR1 P1 TR1 TR1 TR1 P3 1.06 1.29 1.063 1.314 1.27 1.055 0.36 13 1.102 1.342 1.107 1.37 1.371 1.277 0.692 44636 26015 44763 26096 25042 20467 43201 TR1 TR1 TR1 TR1 P1 P2 1.217 1.223 1.225 1.13 0.662 0.401 14 1.331 1.352 1.36 1.335 0.932 0.777 24084 24125 24043 21886 34871 41482 P2 P2 P2 P2 0.549 0.515 0.471 0.371 15 0.807 0.792 0.755 0.704 42103 43207 45708 48543

Region Assembly Power Max Rod Power Assembly Burnup

Figure 10: EOC Power and Burnup Distributions for Transition 1

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

H G F E D C B A

TR1 TR1 TR1 TR1 TR2 TR1 TR2 TR1 1.342 1.373 1.322 1.196 1.134 1.122 1.043 0.485 8 1.394 1.427 1.375 1.28 1.235 1.177 1.243 0.749 25912 25181 24086 25669 0 25184 0 26583 TR1 TR2 TR1 TR2 TR1 TR1 TR2 P1 1.373 1.298 1.334 1.208 1.182 1.145 1.019 0.366 9 1.427 1.403 1.407 1.329 1.275 1.21 1.221 0.566 25174 0 24123 0 24125 26899 0 35018 TR1 TR1 TR2 TR1 TR1 TR2 TR2 P2 1.322 1.336 1.26 1.307 1.186 1.102 0.989 0.329 10 1.375 1.409 1.358 1.371 1.267 1.222 1.208 0.549 24082 24044 0 25061 26114 0 0 43564 TR1 TR2 TR1 TR1 TR2 TR1 TR2 P1 1.197 1.21 1.308 1.364 1.186 1.213 1.057 0.274 11 1.281 1.33 1.372 1.436 1.328 1.283 1.33 0.531 25663 0 24959 20473 0 21897 0 41192 TR2 TR1 TR1 TR2 TR1 TR2 TR1 1.134 1.182 1.186 1.187 1.174 1.205 0.63 12 1.235 1.275 1.268 1.328 1.224 1.354 0.959 0 24181 26099 0 26198 0 26057 TR1 TR1 TR2 TR1 TR2 TR2 P2 1.122 1.145 1.102 1.214 1.206 1.05 0.315 13 1.177 1.21 1.222 1.284 1.354 1.373 0.655 25170 26881 0 21891 0 0 48488 TR2 TR2 TR2 TR2 TR1 P2 1.043 1.02 0.991 1.058 0.631 0.315 14 1.243 1.22 1.209 1.332 0.959 0.655 0 0 0 0 26013 48517 TR1 P1 P2 P1 0.485 0.366 0.331 0.274 15 0.749 0.566 0.555 0.532 26555 34878 43207 41107

Region Assembly Power Max Rod Power Assembly Burnup

Figure 11: BOC Power and Burnup Distributions for Transition 2

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

H G F E D C B A

TR1 TR1 TR1 TR1 TR2 TR1 TR2 TR1 0.968 1.033 1.011 1.052 1.283 1.055 1.257 0.611 8 0.987 1.062 1.049 1.091 1.336 1.088 1.363 0.89 46711 47453 45583 47395 25216 46269 23433 37258 TR1 TR2 TR1 TR2 TR1 TR1 TR2 P1 1.033 1.248 1.1 1.282 1.058 1.087 1.252 0.48 9 1.063 1.322 1.167 1.337 1.085 1.143 1.377 0.701 47448 25786 47437 25954 45733 48836 23316 43266 TR1 TR1 TR2 TR1 TR1 TR2 TR2 P2 1.011 1.1 1.3 1.09 1.054 1.33 1.201 0.424 10 1.049 1.168 1.354 1.15 1.105 1.396 1.364 0.68 45581 47381 26878 48190 47957 25953 22565 50884 TR1 TR2 TR1 TR1 TR2 TR1 TR2 P1 1.052 1.282 1.091 1.136 1.318 1.141 1.096 0.327 11 1.091 1.337 1.152 1.235 1.37 1.185 1.306 0.609 47391 25970 48114 44502 26618 45095 21132 46910 TR2 TR1 TR1 TR2 TR1 TR2 TR1 1.283 1.058 1.054 1.318 1.089 1.243 0.633 12 1.336 1.085 1.105 1.37 1.123 1.358 0.892 25216 45783 47950 26622 48684 24704 38055 TR1 TR1 TR2 TR1 TR2 TR2 P2 1.055 1.087 1.33 1.141 1.243 1.023 0.335 13 1.089 1.143 1.396 1.186 1.358 1.262 0.65 46257 48818 25960 45097 24708 19891 54545 TR2 TR2 TR2 TR2 TR1 P2 1.257 1.252 1.202 1.096 0.633 0.335 14 1.363 1.377 1.364 1.306 0.892 0.65 23434 23320 22581 21146 38021 54574 TR1 P1 P2 P1 0.611 0.481 0.425 0.328 15 0.89 0.701 0.681 0.609 37233 43145 50564 46839

Region Assembly Power Max Rod Power Assembly Burnup

Figure 12: EOC Power and Burnup Distributions for Transition 2

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

H G F E D C B A

EQ1 N1 EQ1 N1 EQ1 N1 EQ1 N1 1.303 1.364 1.217 1.215 1.176 1.282 1.15 0.523 8 1.392 1.421 1.353 1.283 1.289 1.363 1.364 0.803 0 24462 0 24144 0 24142 0 25892 N1 N1 N1 EQ1 N1 N1 EQ1 N2 1.364 1.311 1.24 1.153 1.182 1.248 1.102 0.427 9 1.421 1.379 1.355 1.254 1.247 1.348 1.332 0.704 24462 25148 23088 0 25575 24112 0 42938 EQ1 N1 EQ1 N1 N1 EQ1 EQ1 N2 1.217 1.242 1.181 1.18 1.167 1.129 1.035 0.379 10 1.353 1.357 1.303 1.251 1.224 1.237 1.232 0.651 0 22993 0 26430 26222 0 0 46381 N1 EQ1 N1 N1 EQ1 N1 EQ1 N2 1.215 1.156 1.181 1.324 1.166 1.219 1.051 0.299 11 1.283 1.256 1.251 1.413 1.289 1.292 1.341 0.607 24144 0 26430 19341 0 20908 0 47156 EQ1 N1 N1 EQ1 N1 EQ1 N2 1.176 1.185 1.168 1.167 1.16 1.162 0.543 12 1.289 1.253 1.221 1.289 1.206 1.323 0.951 0 25634 26260 0 25740 0 44046 N1 N1 EQ1 N1 EQ1 EQ1 N2 1.282 1.254 1.133 1.221 1.163 0.997 0.293 13 1.363 1.353 1.241 1.294 1.324 1.323 0.61 24142 24305 0 20946 0 0 47708 EQ1 EQ1 EQ1 EQ1 N2 N2 1.15 1.114 1.042 1.054 0.544 0.293 14 1.364 1.342 1.236 1.345 0.952 0.61 0 0 0 0 44083 47779 N1 N2 N2 N2 0.523 0.446 0.383 0.302 15 0.803 0.747 0.654 0.612 25892 36950 46949 46930

Region Assembly Power Max Rod Power Assembly Burnup

Figure 13: BOC Power and Burnup Distributions for Equilibrium 1

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

H G F E D C B A

EQ1 N1 EQ1 N1 EQ1 N1 EQ1 N1 1.221 1.053 1.287 1.152 1.3 1.105 1.249 0.592 8 1.276 1.116 1.353 1.194 1.359 1.15 1.359 0.857 24890 46728 26003 47914 26103 47093 24331 36720 N1 N1 N1 EQ1 N1 N1 EQ1 N2 1.053 1.007 1.081 1.3 1.054 1.111 1.24 0.504 9 1.116 1.031 1.106 1.36 1.102 1.148 1.368 0.769 46728 46576 45462 25958 47475 46999 23941 51852 EQ1 N1 EQ1 N1 N1 EQ1 EQ1 N2 1.287 1.082 1.28 1.04 1.05 1.324 1.192 0.45 10 1.353 1.108 1.331 1.082 1.102 1.388 1.354 0.728 26003 45400 25632 48061 47939 26083 22834 54357 N1 EQ1 N1 N1 EQ1 N1 EQ1 N2 1.152 1.301 1.04 1.142 1.316 1.142 1.076 0.343 11 1.194 1.361 1.082 1.247 1.372 1.186 1.295 0.657 47914 25999 48067 43040 26261 44009 20759 53213 EQ1 N1 N1 EQ1 N1 EQ1 N2 1.3 1.054 1.049 1.316 1.093 1.224 0.562 12 1.359 1.102 1.101 1.372 1.131 1.352 0.888 26103 47563 47981 26262 48122 23912 54399 N1 N1 EQ1 N1 EQ1 EQ1 N2 1.105 1.111 1.323 1.141 1.223 1.001 0.322 13 1.15 1.147 1.388 1.185 1.352 1.25 0.625 47093 47242 26123 44053 23914 19070 53423 EQ1 EQ1 EQ1 EQ1 N2 N2 1.249 1.246 1.193 1.076 0.562 0.321 14 1.359 1.371 1.355 1.295 0.887 0.623 24331 24126 22916 20790 54438 53485 N1 N2 N2 N2 0.592 0.518 0.451 0.344 15 0.857 0.808 0.724 0.659 36720 46299 54964 53019

Region Assembly Power Max Rod Power Assembly Burnup

Figure 14: EOC Power and Burnup Distributions for Equilibrium 1

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H G F E D C B A

EQ1 EQ1 EQ1 EQ1 EQ2 EQ1 EQ2 EQ1 1.334 1.345 1.258 1.258 1.184 1.284 1.147 0.518 8 1.383 1.425 1.313 1.314 1.291 1.364 1.362 0.797 24890 24331 26003 23914 0 23912 0 26103 EQ1 EQ2 EQ1 EQ2 EQ1 EQ1 EQ2 N1 1.345 1.237 1.233 1.171 1.182 1.248 1.1 0.428 9 1.425 1.362 1.321 1.29 1.249 1.35 1.33 0.7 24331 0 22916 0 25958 23941 0 43040 EQ1 EQ1 EQ2 EQ1 EQ1 EQ2 EQ2 N1 1.258 1.234 1.167 1.182 1.167 1.127 1.031 0.377 10 1.313 1.324 1.263 1.25 1.22 1.235 1.227 0.647 26003 22834 0 26261 26083 0 0 46728 EQ1 EQ2 EQ1 EQ1 EQ2 EQ1 EQ2 N1 1.258 1.175 1.183 1.325 1.164 1.217 1.046 0.297 11 1.314 1.294 1.247 1.414 1.287 1.289 1.335 0.603 23914 0 26262 19070 0 20759 0 47242 EQ2 EQ1 EQ1 EQ2 EQ1 EQ2 N1 1.184 1.185 1.168 1.165 1.158 1.161 0.544 12 1.291 1.244 1.22 1.288 1.203 1.321 0.949 0 25999 26123 0 25632 0 44009 EQ1 EQ1 EQ2 EQ1 EQ2 EQ2 N1 1.284 1.253 1.13 1.219 1.162 1.001 0.302 13 1.364 1.354 1.239 1.291 1.321 1.323 0.643 23912 24126 0 20790 0 0 45462 EQ2 EQ2 EQ2 EQ2 N1 N1 1.147 1.114 1.039 1.051 0.545 0.302 14 1.362 1.34 1.232 1.34 0.95 0.644 0 0 0 0 44053 45400 EQ1 N1 N1 N1 0.518 0.449 0.383 0.3 15 0.797 0.743 0.653 0.609 26103 36720 47093 46999

Region Assembly Power Max Rod Power Assembly Burnup

Figure 15: BOC Power and Burnup Distributions for Equilibrium 2

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H G F E D C B A

EQ1 EQ1 EQ1 EQ1 EQ2 EQ1 EQ2 EQ1 1 1.059 1.011 1.114 1.294 1.108 1.252 0.591 8 1.02 1.111 1.036 1.177 1.35 1.154 1.362 0.856 45965 46381 47151 46903 25892 46949 24461 36951 EQ1 EQ2 EQ1 EQ2 EQ1 EQ1 EQ2 N1 1.059 1.261 1.082 1.286 1.049 1.114 1.244 0.508 9 1.111 1.328 1.14 1.34 1.093 1.151 1.371 0.771 46381 25148 45057 25575 47708 46930 24112 52078 EQ1 EQ1 EQ2 EQ1 EQ1 EQ2 EQ2 N1 1.011 1.083 1.297 1.046 1.053 1.326 1.195 0.452 10 1.036 1.142 1.35 1.087 1.108 1.39 1.357 0.73 47151 45003 25740 47948 47881 26222 22993 54762 EQ1 EQ2 EQ1 EQ1 EQ2 EQ1 EQ2 N1 1.114 1.287 1.046 1.148 1.32 1.146 1.078 0.343 11 1.177 1.342 1.09 1.252 1.376 1.189 1.297 0.658 46903 25634 47963 42938 26429 44046 20908 53340 EQ2 EQ1 EQ1 EQ2 EQ1 EQ2 N1 1.294 1.05 1.052 1.319 1.096 1.228 0.566 12 1.35 1.095 1.107 1.375 1.134 1.355 0.891 25892 47779 47927 26430 48203 24142 54496 EQ1 EQ1 EQ2 EQ1 EQ2 EQ2 N1 1.108 1.114 1.326 1.145 1.227 1.008 0.333 13 1.154 1.151 1.39 1.188 1.355 1.254 0.656 46949 47156 26260 44083 24143 19341 51364 EQ2 EQ2 EQ2 EQ2 N1 N1 1.252 1.25 1.197 1.078 0.566 0.333 14 1.362 1.374 1.358 1.297 0.89 0.656 24461 24305 23088 20946 54546 51307 EQ1 N1 N1 N1 0.591 0.524 0.454 0.345 15 0.856 0.809 0.728 0.661 36951 46230 55194 53137

Region Assembly Power Max Rod Power Assembly Burnup

Figure 16: EOC Power and Burnup Distributions for Equilibrium 2

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3.2 Technical Analysis

3.2.1 TS 2.0 Safety Limits Changes

Background

Based on design comparisons and hydraulic testing of the fuel assemblies, the Framatome HTP fuel and the Westinghouse 17x17 RFA-2 fuel assembly designs are considered to be hydraulically compatible. The Westinghouse 17x17 RFA-2 fuel design is compatible with the core components and the SQN resident Framatome HTP fuel. The dimensions of the two fuel types are essentially equivalent, i.e., the grid envelope, fuel rod, thimble and instrument tube diameters are unchanged between the Westinghouse 17x17 RFA-2 design and the Framatome HTP fuel. The thermal-hydraulic (T/H) evaluation of the 17x17 RFA-2 fuel has shown that the Westinghouse 17x17 RFA-2 and Framatome HTP fuel types are hydraulically compatible.

Methodology

The T/H analysis for the Westinghouse 17x17 RFA-2 fuel for SQN Units 1 and 2 is based on the following. 1. The Westinghouse Revised Thermal Design Procedure (RTDP) Departure from Nucleate Boiling Ratio (DNBR) Evaluation Methodology (WCAP-11397-P-A) 2. The VIPRE-W code (WCAP-14565-P-A [Reference 20]) 3. The WRB-2M DNB correlation (WCAP-15025-P-A and Letter from D. S. Collins (USNRC) to J. A. Gresham (Westinghouse), “Modified WRB-2 Correlation WRB-2M for Predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,” February 2006 [Reference 21].

WCAP-11397-P-A is referenced as the design method (RTDP) used to meet the DNB design basisin the Sequoyah current licensing basis (Reference 95 in UFSAR Section 4.4.6). In the RTDP method, uncertainties in plant operating parameters nuclear thermal parameters, fuel fabrication parameter, computer codes, and DNB correlation predictions are considered statistically to obtain DNB uncertainty factors. Based on the DNB uncertainty factors, RTDP design limit DNBR values are determined such that there is at least a 95% probability at a 95% confidence level that DNB will not occur on the most limiting fuel rod during normal operation and operational transients and during transient conditions arising from faults of moderate frequency (Condition I and II events as defined in ANSI N18.2). Because the uncertainties in these parameters are considered in determining the design DNBR value, the plant safety analyses are performed using input values without uncertainties for these parameters.

The applicability of the WRB-2M DNB correlation to analyses for Westinghouse 17x17 RFA-2 fuel is documented in WCAP-14565-P-A. The VIPRE-W models employ the WRB-2M DNB correlation for at-power events and for analyses applicable to the region above the first mixing vane grid. When any of the conditions are outside the range of the WRB-2M DNB correlation, the W-3 and W-3 Alternative (ABB-NV and WLOP) DNB correlations and a deterministic treatment of key DNBR analysis uncertainties is used. These correlations have been approved by the NRC and are described in WCAP-14565-P-A, Addendum 1-A (Reference 22), and WCAP-14565-P-A, Addendum 2-P-A (Reference 23).

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The Westinghouse PAD5 fuel performance code from WCAP-17642-P-A was used to evaluate fuel temperatures and rod internal pressure for Safety Analysis considering the Sequoyah Units 1 and 2 fuel transition conditions. Analyses were performed to predict best estimate and upper bound fuel centerline temperatures as a function of rod local burnup and power based on a limiting steady-state power history. These fuel centerline temperature analyses are the same as those performed to provide maximum fuel temperature input to LOCA and Non-LOCA safety analysis. Using these upper bound fuel centerline temperatures in conjunction with the burnup dependent fuel melting temperature limit, the local power level at which fuel centerline melt will occur is determined for each burnup step. The safety analyses confirmed that the peak linear heat rate (expressed in kW/ft) does not provide a value that could result in fuel centerline melting.

The PAD5 fuel pellet overheating (power-to-melt) analysis confirmed that fuel rods will not fail due to fuel centerline melting for Condition I and Condition II events for the Sequoyah Units 1 and 2 fuel transition conditions and will be confirmed as part of the cycle-specific Reload Safety Analysis Checklist (RSAC) along with the other PAD5 generated data for Safety Analysis.

Refer to Attachment 8 for how the Limitations and Conditions were met for each of the NRC approved methodologies applied to the proposed Sequoyah Units 1 and 2 fuel transition.

Results

TS 2.1.1, “Reactor Core SLs” Figure 2.1.1-1, “Reactor Core Safety Limit – Four Loops in Operation” resulting from the combination of Thermal Power, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure is revised from the application of WCAP-11397-P-A evaluation methodology. The revised reactor core safety limit figure would also be relocated to the COLR consistent with NRC-approved TSTF-339-A and WCAP-14483-A. As discussed below from the NRC Safety Evaluation for WCAP- 14483-A, the justification is provided for why this figure may be relocated to the COLR and replaced with the DNB design basis limit and the fuel centerline melt limit.

“The current TS figure (2.1.1-1) presents core limits on RCS temperature conditions (T-avg) as a function of pressurizer pressure and fractional rated thermal power. This figure was originally included in the Westinghouse TS to satisfy the requirements of 10CFR50.36 which states that ‘safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.’ However, the figure is not a complete representation of reactor core safety limits but is intended to provide the relationship between the process variables that are available to the operator (i.e., T-avg, pressurizer pressure, and thermal power) and the DNB design basis safety limit.

To ensure that the requirements of 10CFR50.36 are met, i.e., limits upon important process variables, the WOG has proposed to retain the requirement for a Reactor Core Limits figure in the Safety Limits TS, but relocate the actual figure to the COLR and replace it with the DNB design basis limit and the fuel centerline melt limit (Ref. 2). Both of these limits are criteria that must be satisfied for normal operation and for AOOs to prevent overheating of the fuel cladding and possible cladding perforation which would

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result in the release of fission products to the RCS and are, therefore, the true safety limits. The reactor protection system (RPS) and the Reactor Core Limits figure would then be used to determine whether the actual DNB and fuel centerline melt safety limits were violated should an event occur that could potentially challenge them. Appropriate functioning of the RPS and the steam generator safety valves ensures that for variations in the thermal power, RCS pressure, RCS average temperature, RCS flow rate, and ǻI (percent power in top half of core minus percent power in bottom half of core), the reactor core safety limits will be satisfied during steady-state operation, normal operational transients, and AOOs. Therefore, in the event of an AOO, verification that the RPS and the main steam system safety valves are functioning as designed will ensure that all safety limits are met. In the event that the RPS is not functioning as designed, an evaluation of any transient condition would be required to determine whether or not the safety limits have been violated.

In addition, since the Reactor Core Limits figure is based on the nuclear enthalpy rise N hot channel factor limit, F 废宋. and the RCS total flow rate, both of which may be in the COLR, relocation of the figure to the COLR would eliminate the need for a license amendment if cycle-dependent changes to these parameters were to exist.

The staff concludes that the Reactor Core Limits figure may be relocated to the COLR and replaced with the DNB design basis limit and the fuel centerline melt limit. The NRC-approved methodology used to derive the parameters in the figure is contained in WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," dated July 1985, (or other applicable approved reload methodology), and will be referenced in the Reporting Requirements section of the TS.”

Relocating the reactor core safety limit figure from TS 2.1.1 to the COLR will also be consistent with TS 2.1.1 in NUREG-1431. The DNB design basis limit and the fuel centerline melt limit are contained in Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2.

Reactor Core Safety Limit 2.1.1.1 is revised from the application of WCAP-15025-P-A evaluation methodology. The change would replace the three Framatome correlation departure from nucleate boiling ratio (DNBR) values with the WRB-2M correlation DNBR value. With the significant improvement in the accuracy of the critical heat flux (CHF) prediction by using the WRB-2M correlation instead of previous DNB correlations, a DNBR limit of 1.14 is applicable for the 17x17 RFA-2 Standard fuel assembly.

Reactor Core Safety Limit 2.1.1.2 is revised from the application of WCAP-17642-P-A evaluation methodology. The change would replace the peak fuel centerline temperature derived from Framatome methodology with the Westinghouse PAD5 peak fuel centerline temperature value. With PAD5, an improved melting temperature model is used in which burnup has less of an impact on the fuel melt temperature limit, as indicated by the following equation that is conservatively based on the UO2 fuel melting point equation used in PAD5.

Tmelt = 5080°F – __9__ (BU) 10000 where, Tmelt = UO2 melting temperature, °F BU = UO2 burnup, MWD/MTU

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Summary and Conclusions

The T/H evaluation of the 17x17 RFA-2 fuel has shown that the Westinghouse 17x17 RFA-2 and Framatome HTP fuel types are hydraulically compatible. The existing DNB methodology is applicable with the addition of the transition core DNB effect. The design criteria associated with rod bow, fuel temperatures, bypass flow, core component analysis, void fraction, and hydrodynamic stability have been satisfied.

The PAD5 fuel performance code was used to assess the fuel rod design criteria and generate the fuel performance data (i.e., fuel temperatures, rod internal pressure, and fuel centerline melt for the respective downstream analyses). Fuel performance evaluations were completed for each of the representative fuel regions to demonstrate that the design criteria can be satisfied for all fuel rod types in the core under the planned operating conditions for the Sequoyah Units 1 and 2 proposed fuel transition. Cycle-specific core designs and fuel performance analyses are performed for each reload cycle to guarantee that the fuel rod design criteria will remain satisfied for cycle-specific operating conditions.

3.2.2 TS 3.1 Reactivity Control Systems Changes

Background

TS 3.1.4, “Rod Group Alignment Limits” Required Action B.2.4 is revised to be consistent with FQ surveillance requirement numbering changes to TS 3.2.1. The proposed changes to TS 3.2.1 (discussed in subsection 3.2.3 of this enclosure) will address the issues identified in Westinghouse Nuclear Safety Advisory Letter (NSAL) NSAL-09-5, Revision 1, “Relaxed Axial Offset Control FQ Technical Specification Actions” (Reference 24), and also will address issues identified in NSAL-15-1, “Heat Flux Hot Channel Factor Technical Specification Surveillance” (Reference 25).

TS 3.1.7, “Rod Position Indication” Required Actions A.1, A.2.1, B.3, and C.1 are revised to implement BEACON core power distribution measurement consistent with WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A, and replace using “movable incore detectors” with “core power distribution measurement information”. The proposed TS 3.1.7 changes will allow the use of a dedicated on-line core power distribution monitoring system (PDMS) to enhance surveillance of core thermal limits. The PDMS to be used at SQN Units 1 and 2 is the NRC-approved Westinghouse proprietary core analysis system called Best Estimate Analyzer for Core Operations – Nuclear (BEACON). Currently, for conditions involving inoperable rod position indicators, Required Actions A.1, A.2.1, B.3, and C.1 require plant operators to verify the position of the rod(s) with inoperable position indicators by using movable incore detectors. The generic phrase “core power distribution measurement information” is substituted for “movable incore detectors,” as this would allow the use of either a PDMS or the movable incore detectors for verifying the position of the rod(s) with an inoperable rod position indicator.

Methodology

As discussed in WCAP-17661-P-A, the Heat Flux Hot Channel Factor [FQ(Z)] TS provides assurance that the heat flux hot channel factor, FQ, will remain within the limits assumed in the plant safety analyses when the core is operated within its allowed operating space. Performing SR 3.2.1.1 and SR 3.2.1.2 verifies that that FQ(Z), as approximated by C equilibrium FQ(Z) [FQ (Z)] and transient FQ(Z) for future non-equilibrium operation within the

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W allowed operating space [FQ (Z)], are within the limits specified in the COLR. The application of WCAP-17661-P-A is discussed in more detail in subsection 3.2.3 of this enclosure.

As described in WCAP-12472-P-A, the Westinghouse BEACON PDMS was developed to improve the operational support for pressurized water reactors (PWRs). BEACON is an advanced core monitoring and support package that utilizes existing plant instrumentation such as core exit thermocouple temperatures, reactor coolant system (RCS) cold leg temperatures, control bank positions, power range detector output, and reactor power level. These data are sent by the plant computer in the form of a file that BEACON can interpret to perform nodal power distribution prediction calculations. The PDMS does not provide any protection or control system function.

The BEACON-TSM (Technical Specification Monitor) system level was developed to provide licensees with the functionality needed to integrate BEACON into the plant TS for monitoring of current TS thermal power limits such as peak linear power density (FQ – TS 3.2.1, Heat N Flux Hot Channel Factor (FQ(Z))) and peak enthalpy rise (F ǻ+ – TS 3.2.2, Nuclear N Enthalpy Rise Hot Channel Factor (F ǻ+)). The application of WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A is discussed in detail in subsection 3.2.3 of this enclosure.

Refer to Attachment 8 of this Enclosure for how the Limitations and Conditions were met for each of the NRC approved methodologies that were applied for the proposed Sequoyah Units 1 and 2 fuel transition.

Results

TS 3.1.4 Required Action B.2.4 for continued operation with a misaligned rod verifies that hot channel factors FQ(Z) are within limits. Required Action B.2.4 is revised to add “and SR 3.2.1.2” consistent with changes to TS 3.2.1 Surveillance Requirements (SRs) from Appendix A of WCAP-17661-P-A. The revised TS 3.1.4 Required Action B.2.4 would C require performance of both SR 3.2.1.1 to “Verify FQ (Z) is within limit”, and SR 3.2.1.2 to W “Verify FQ (Z) is within limit”. Performing SR 3.2.1.1 and SR 3.2.1.2 verifies that FQ(Z), as C W approximated by FQ (Z) and FQ (Z), is within the required limits and ensures that operation DW573ZLWKDURGPLVDOLJQHGZLOOQRWUHVXOWLQSRZHUGLVWULEXWLRQVWKDWPD\LQYDOLGDWH safety analysis assumptions at full power.

Rod position may be determined indirectly by use of core power distribution measurement information. Core power distribution measurement information can be obtained from flux maps using the movable incore detectors, or from a functional PDMS. TS 3.1.7, “Rod Position Indication” Required Actions A.1, A.2.1, B.3, and C.1 are revised to implement BEACON core power distribution measurement consistent with WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A, and replace using “movable incore detectors” with “core power distribution measurement information.”

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Summary and Conclusions

The Reactivity Control Systems TS 3.1.4 and TS 3.1.7 Required Action changes are consistent with the changes from the application of WCAP-17661-P-A, WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A. Application of these NRC approved topical reports is discussed in more detail in subsection 3.2.3 of this enclosure.

3.2.3 TS 3.2 Power Distribution Limits Changes

TS 3.2.1 Change Discussion

TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(X,Y,Z)),” is revised to implement the methodology and Technical Specification markups consistent with TS 3.2.1, “Heat Flux Hot Channel Factor (FQ(Z)),” from Appendix A of WCAP-17661-P-A (Reference 12), with deviations to the Condition B Required Action Completion Times based on References 30 and 31, for RAOC plants using the format and much of the content contained within NUREG-1431 Revision 4 (Reference 9). Additionally, TS 5.6.3.b would be revised to list WCAP-17661-P-A Revision 1 as a COLR methodology reference.

NSAL-09-5, Revision 1, “Relaxed Axial Offset Control FQ Technical Specification Actions,” dated September 23, 2009, (Reference 24), notified Westinghouse customers of an issue associated with the Required Actions for Condition B of NUREG-1431 (Reference 9) TS 3.2.1B, “Heat Flux Hot Channel Factor (FQ(Z) (RAOC-W(Z) Methodology),” for plants that have implemented the W RAOC methodology. In certain situations where transient FQ, FQ (Z), is not within its limit, the W existing Required Actions may be insufficient to restore FQ (Z) to within the limit.

NSAL-15-1, “Heat Flux Hot Channel Factor Technical Specification Surveillance,” dated February 3, 2015, (Reference 25) notified Westinghouse customers of an issue associated with Surveillance Requirement (SR) 3.2.1.2 of NUREG-1431 (Reference 9). Specifically, one aspect of the SR may not be sufficient to assure that the peaking factor assumed in the licensing basis analysis remains valid under all conditions between the instances of performance of SR 3.2.1.2. Implementing the TS changes of WCAP-17661-P-A, Revision 1, will resolve the non-conservatism identified in NSAL-15-1 by inclusion of the penalty factor from the NUREG-1431 SR 3.2.1.2 NOTE in the surveillance formulation and therefore make it applicable at all times as part of the SQN Unit 1 and 2 transition to Westinghouse RFA-2 fuel.

These non-conservatisms identified in NSAL-09-5 Revision 1 and NSAL-15-1 are not applicable to the current SQN TS 3.2.1, but are applicable to the NUREG-1431 TS 3.2.1B. The proposed changes to TS 3.2.1 implement a modified version of NUREG-1431 TS 3.2.1B from WCAP-17661-P-A that will resolve the non-conservatisms identified in NSAL-09-5 Revision 1 and NSAL-15-1 as part of the SQN Unit 1 and 2 transition to Westinghouse RFA-2 fuel.

The improved FQ surveillance methodology in WCAP-17661-P-A resolves the issues identified in NSAL-09-5 Revision 1 and NSAL-15-1. The new surveillance methodology requires the measurement of FXY(Z), which is then multiplied by factors that characterize the maximum transient T(Z) values postulated to occur during non-equilibrium operation. This formulation essentially eliminates the sensitivity of the surveillance to the surveillance axial power shape. Additionally, the improved FQ surveillance methodology incorporates various RAOC operating spaces, consisting of combinations of control bank rod insertion, AFD, and thermal power limits

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that provides sufficient FQ margin for future operation. These improvements to TS 3.2.1 are discussed below in the "Methodology for 3.2.1 Changes” section of this enclosure.

The Completion Times for Required Actions B.2.1, B.2.2, and B.2.3 have been revised from the Technical Specification markups of WCAP-17661-P-A Revision 1 Appendix A, based on information contained in References 30 and 31

TSTF-241-A Revision 4 (Reference 30) approved the addition of “after each FQ(Z) determination” to the Completion Times of STS 3.2.1A Required Actions A.1 through A.4 and C “after each FQ (Z) determination” to the Completion Times of STS 3.2.1B Required Actions A.1 through A.3. These changes were made to assure that the Required Actions were continued, with the associated Completion Time resets, after subsequent peaking factor determinations were made while continuing to operate under the applicable Condition with the peaking factor parameter not within its limit. The NRC approved Reference 30 on January 13, 1999.

TSTF-290-A Revision 0 (Reference 31) made the following changes of note for this license amendment request.

x Deleted Required Action A.2 of STS 3.2.1A (renumbering the subsequent Required Actions of that LCO) and renamed STS 3.2.1A to “FQ(Z) (CAOC-Fxy Methodology)”

x Added new Required Actions B.2 and B.3 to STS 3.2.1B and renamed STS 3.2.1B to “FQ(Z) (RAOC-W(Z) Methodology)”

x Added new STS 3.2.1C (“FQ(Z) (CAOC-W(Z) Methodology)”

The NRC approved Reference 31 on June 30, 1999.

Both References 30 and 31 were included in STS NUREG-1431 Revision 2 (published by the NRC in June 2001). However, the related effects of References 30 and 31 were not reconciled in the various versions of the FQ STS 3.2.1 prior to publication of STS NUREG-1431 Revision 2. It appears that all Required Actions in Reference 31 which required either a power reduction or reactor trip setpoint change should have had a Completion Time that was reset after each FQ determination, the intent of Reference 30. The logic for this is as follows.

TSTF-290 introduced a new Condition B with Required Actions and Completion Times similar to those in Condition A. However, the parallel work on TSTF-241 appears to have not been reconciled to TSTF-290 to recognize the new Condition B and did not include the appropriate Completion Time changes for Condition B.

TS 3.2.1 Bases in WCAP-17661-P-A Revision 1 Appendix B makes the following statements for Required Action B.2.1:

“The Completion Time of 4 hours provides an acceptable time to reduce the THERMAL POWER and AFD limits in an orderly manner to preclude entering an unacceptable condition during future non-equilibrium operation. The limit on THERMAL POWER initially determined by Required Action B.2.1 may be affected by subsequent W determinations of FQ (Z) that are not within limit and would require power reductions W within 4 hours of the FQ (Z) determination if necessary to comply with the decreased

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W THERMAL POWER limit. Decreases in FQ (Z) would allow increasing the THERMAL POWER limit and increasing THERMAL POWER up to this revised limit.”

C The steady state FQ (Z) peaking factor Completion Times for new Actions A.1, A.2, and A.3 in C TS 3.2.1 allow for subsequent Completion Time resets after each steady state FQ (Z) peaking factor determination. However, the Completion Times for new Actions B.2.1 through B.2.3 in WCAP-17661-P-A Revision 1 Appendix A do not allow for subsequent Completion Time resets W after each transient FQ (Z) peaking factor determination.

W This license amendment request proposes that subsequent transient FQ (Z) peaking factor determinations while the plant continues to operate in Condition B, either pursuant to Required Action B.2.4 or from the nominal performance of SR 3.2.1.2 if it should come due while the plant remains in Condition B, should reset the Completion Times for Required Actions B.2.1, B.2.2, W and B.2.3. Without the required Completion Time resets, subsequent transient FQ (Z) peaking factor determinations that occur after 4 hours (Required Action B.2.1) or 72 hours (Required Actions B.2.2 and B.2.3) would be met with an expired Completion Time resulting in a required shutdown to MODE 2.

TS 3.2.2 Change Discussion

TS 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor Fǻ+(X,Y),” is revised to reflect N TS 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor (F 'H),” in Reference 9. Adopting the latest version of the NRC-approved Standard Technical Specification 3.2.2 essentially returns the Sequoyah Units 1 and 2 Technical Specifications to their licensing basis for the

F¨H LCO prior to Reference 26 which addressed their conversion from Westinghouse fuel to Framatome Cogema Fuel, except for two Completion Times (CTs).

Prior to Reference 26, with F¨H not within limit, Sequoyah Units 1 and 2 had an Action to reduce Thermal Power to less than 50% of Rated Thermal Power (RTP) within 2 hours. During the industry efforts following the NRC’s Interim Policy Statement on Technical Specification Improvements (52 FR 3788) which led up to NUREG-1431 Revision 0 (September 1992), WCAP-12159, “MERITS Program, Phase II, Technical Specifications and Bases,” March 1989, used the Vogtle Unit 1 Technical Specifications of that vintage as the latest licensed version of TS 3.2.2. The CT for the 50% power reduction was 4 hours in the Vogtle licensing basis at that time. The 4-hour CT was also used in WCAP-13029 Volume 1, “MERITS Program, Phase III, Comments on Draft NUREG-1431, Standard Technical Specifications – Westinghouse Plants,” July 1991, and has been used ever since with the NRC’s approval in NUREG-1431 Revisions 0 through 4. The 4-hour CT for this power reduction is acceptable because it provides an appropriate time to reach the required power level from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time.

Prior to Reference 26, with F¨H not within limit, Sequoyah Units 1 and 2 also had an Action to reset the power range neutron flux WULSVHWSRLQWVWR55% RTP within 6 hours. Similar to the discussion above, this CT was 8 hours in the Vogtle licensing basis at that time, which was also used in WCAP-12159 and WCAP-13029 Volume 1 and used with the NRC’s approval in NUREG-1431 Revisions 0 and 1. Prior to the issuance of NUREG-1431 Revision 2, the NRC approved TSTF-95-A to increase this CT to 72 hours by letter from Christopher I. Grimes (NRC OTSB) to James Davis (NEI) dated 9-27-96. TSTF-95-A was cited in the NRC’s approval of the 72-hour CT in LA274/155 for Beaver Valley Units 1 and 2 dated

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2-27-06 (ADAMS Accession Number ML060330636). The 72-hour CT for resetting the trip setpoints is acceptable, recognizing that after the power level is reduced the safety analysis assumptions are satisfied. This CT balances the non-urgent need to reset the trip setpoints against the sensitivity of an operation that could inadvertently trip the reactor.

TS 3.2.4 Change Discussion

TS 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” Required Actions A.3 and A.6 are changed to be consistent with the Surveillance Requirement numbering in revised TS 3.2.1 and revised TS 3.2.2. SR 3.2.4.2 would also be changed to implement BEACON core power distribution measurement consistent with WCAP-12472-P-A (Reference 13) and WCAP-12472-P-A, Addendum 4-A (Reference 14), and replace the discussion of flux maps obtained solely using the movable incore detectors with more generic terminology (i.e., “core power distribution measurement information”). The basis for implementing the BEACON methodology was previously discussed in subsection 3.2.2.

Methodology for TS 3.2.1 Changes

Heat Flux Hot Channel Factor (FQ(Z))

FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density and is a measure of the peak fuel pellet power within the reactor core. The values of FQ vary along the axial height (Z) of the core. FQ(Z) also varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution. The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density.

FQ(Z) is measured periodically using either the movable incore detector system (MIDS) or the power distribution monitoring system (PDMS). Because these measurements are generally taken with the core at or near equilibrium conditions, the measured FQ(Z) does not include the variations which would be present during non-equilibrium situations, such as load following or power ascension.

To account for these possible variations, the equilibrium values of FQ(Z) are adjusted by elevation-dependent factors that account for the expected maximum values postulated to occur during RAOC operation.

The proposed changes to TS 3.2.1 involve a re-formulation of these elevation-dependent factors, designated as [T(Z)]COLR. The proposed TS 3.2.1 incorporates various RAOC Operation Spaces (ROS) that define the corresponding elevation-dependent factors, [T(Z)]COLR. Each ROS is composed of corresponding COLR limits associated with TS 3.2.3, “Axial Flux Difference (AFD),” and TS 3.1.6, “Control Bank Insertion Limits,” assumed in the calculation of each particular [T(Z)]COLR function.

Axial Flux Difference (AFD)

The purpose of TS 3.2.3 is to establish limits on the values of AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumption used in the safety analyses. Limiting power distribution skewing over time

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also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control.

AFD is the difference in normalized flux signals between the top and bottom halves of a two-section excore neutron detector. AFD is a measure of the axial power distribution skewing to either the top or bottom half of the core. AFD is sensitive to many core-related parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration.

The allowed range of AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements. The limits on AFD ensure that FQ(Z) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition II, III, or IV events as described in Chapter 15 of the SQN Updated Final Safety Analysis Report (reference Attachment 10 for a summary of the safety analysis changes to the UFSAR).

RAOC, as described in WCAP-10216-P-A, Revision 1A, “Relaxation of Constant Axial Offset Control/FQ Surveillance Technical Specification,” (Reference 27), is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of AFD. Subsequently, power peaking factors and power distributions are examined to ensure that the loss of coolant accident (LOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD limits invalidates the conclusions of the accident and transient analyses with regard to fuel cladding integrity.

The RAOC methodology establishes a xenon distribution library with tentatively wide AFD limits. One-dimensional axial power distribution calculations are then performed to demonstrate that normal operation power shapes are acceptable for LOCA and loss of flow accident and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysis requirements.

Control Bank Insertion Limits

The insertion of the control banks directly affects core power and fuel burnup distributions and assumptions of available ejected rod worth, shutdown margin (SDM), and initial reactivity insertion rate. Rod insertion limits (RILs) are established and rod positions are monitored against the RILs and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control (shutdown banks C and D have only one group each). A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. Except for shutdown banks C and D, a bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one-step of each other. There are four control banks and four shutdown banks.

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TS 3.1.5 requires each shutdown bank to be within the insertion limits as specified in the COLR. TS 3.1.6 requires the control banks to be within the insertion, sequence, and overlap limits as specified in the COLR. The control banks are operated in sequence by withdrawal of Bank A, Bank B, Bank C, and then Bank D. The control banks are sequenced in reverse order upon insertion.

Overlap is the distance travelled together by two control banks. Upon initiation of control bank withdrawal, control bank A is withdrawn by itself. At a predetermined position, control bank B begins withdrawing, resulting in both banks withdrawing simultaneously until control bank A is fully withdrawn. Control bank B will continue withdrawing until, at a subsequent predetermined position, control bank C begins withdrawing. This process continues until control bank D is withdrawn to the full power position. As such, each bank’s overlap is the number of steps that each bank travelled from the following bank’s predetermined position to the fully withdrawn position.

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, TS 3.1.4, “Rod Group Alignment Limits,” TS 3.1.5, TS 3.1.6, TS 3.2.3, and TS 3.2.4, “Quadrant Power Tilt Ratio (QPTR),” provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria. The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core.

Refer to Attachment 8 for how Limitations and Conditions were met.

Results

As described in detail in WCAP-17661-P-A, the proposed change implements an improved RAOC FQ Surveillance formulation and TS. The new formulation also improves accuracy of part-power surveillances. Finally, the improved RAOC FQ Surveillance TS incorporates the concept of RAOC operating spaces that are defined in the COLR. If the FQ limit is exceeded during a surveillance, a more restrictive RAOC operating space can be implemented that provides the required additional FQ margin for future operation.

Sequoyah Technical Specification 3.2.1 is entirely replaced with a new TS which is a composite of References 9, 12, 30, and 31. New Sequoyah TS 3.2.1 is changing as compared to Westinghouse Standard TS (STS) 3.2.1B, “Heat Flux Hot Channel Factor (FQ(Z) (RAOC-W(Z) Methodology)” as follows.

Condition A

C x Revises setpoint reductions required when FQ (Z) limit is exceeded. Required Actions C A.2 and A.3 are being revised replacing “1 percent for each 1 percent FQ (Z) exceeds limits” with “1% for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action A.1.” x The Note is being revised to clarify when SR 3.2.1.2 is required. x These changes are evaluated in Section 4.2 of the NRC Final Safety Evaluation included in WCAP-17661-P-A.

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Condition B

x A new Required Action B.1.1 is included, which allows the plant to “Implement a RAOC W operating space specified in the COLR that restores FQ (Z) to within limits” whenever W FQ (Z) is determined to be not within the limits. The RAOC operating space is a unique combination of axial flux difference (AFD) limits and control bank insertion limits. The operating spaces are pre-analyzed using the approved WCAP-17661-P-A methodology and included in the COLR. x A new Required Action B.1.2 assures that for situations involving control rod movement, C W SRs 3.2.1.1 and 3.2.1.2 will be performed to ensure that FQ (Z) and FQ (Z) remain within limits. x Completion Times for Required Actions B.2.1, B.2.2, and B.2.3 with the resets discussed W above (i.e., after each FQ (Z) determination). x Other than the last bulleted item, these changes are evaluated in Section 4.3 of the NRC Final Safety Evaluation included in WCAP-17661-P-A Revision 1.

Removal of Notes for FQ Surveillance

x Two Notes are deleted in the revised SRs. The first removed Note applied to both SR 3.2.1.1 and SR 3.2.1.2 and required obtaining the power distribution map for C W measuring FQ (Z) and FQ (Z) at equilibrium conditions during power escalation at the W beginning of each cycle. The effect of the change is that FQ (Z) will not be determined until 24 hours after exceeding 75 percent of RTP, instead of within 12 hours of achieving equilibrium conditions after exceeding 75 percent RTP following refueling outages as currently specified. W x The second removed Note applies to SR 3.2.1.2 and required multiplication of FQ (Z) by a factor and increased surveillance under certain conditions. In the improved methodology, the penalty factor is embedded in the methodology and a separate penalty factor is not applicable. x The deletion of these notes is evaluated in Section 4.4 of the NRC Final Safety Evaluation included in WCAP-17661-P-A.

Revision of Second Surveillance Frequency for SRs 3.2.1.1 and 3.2.1.2

x The time interval for completing the SRs is increased from 12 to 24 hours. x These changes are evaluated in Section 4.5 of the NRC Final Safety Evaluation included in WCAP-17661-P-A.

Deletion of Note in SR 3.2.1.2

x The deleted Note required increasing the Frequency to once per 7 effective full power days (EFPD) for certain conditions until the conditions are satisfied. In the new W methodology, the required penalty factor is part of the FQ (Z) formulation. x This change is evaluated in Section 4.6 of the NRC Final Safety Evaluation included in WCAP-17661-P-A.

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Change in Frequency of SR 3.2.1.2 during Power Escalations

W x The Frequency is being changed to require FQ (Z) to be verified within the limits following each refueling within 24 hours after THERMAL POWER exceeds 75 percent RTP. x This change is evaluated in Section 4.7 of the NRC Final Safety Evaluation included in WCAP-17661-P-A.

Summary and Conclusions for TS 3.2.1 Changes

This proposed amendment resolves non-conservative NUREG-1431 TS 3.2.1 Required Actions identified via Westinghouse NSAL-09-5, Revision 1 and also resolves non-conservative NUREG-1431 TS 3.2.1 Surveillance Requirements identified via Westinghouse NSAL-15-1. The proposed amendment also revises TS 5.6.3.b to include WCAP-17661-P-A, Revision 1, as a COLR methodology reference.

The new FQ formulation will remove the surveillance sensitivity to the predicted axial power shapes and remove the potential non-conservatism in TS 3.2.1. Additional actions for the new Condition B Required Action Completion Times have also been proposed. The detailed proposed changes to SQN TS are provided in mark-up and retyped forms in Attachments 1 through 4 to this enclosure. The detailed changes to the SQN (Unit 1) TS Bases are provided for information in markup form in Attachment 5 to this enclosure.

3.2.4 TS 3.3.1 Reactor Trip System (RTS) Instrumentation Changes

Background

TS 3.3.1, “Reactor Trip System (RTS) Instrumentation” SR 3.3.1.3 and 3.3.1.6 are revised to implement BEACON core power distribution measurement consistent with WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A, and replace “incore detector” with “core power distribution” measurement. The proposed TS 3.3.1 changes will allow the use of a dedicated on-line core PDMS to enhance surveillance of core thermal limits. The PDMS to be used at SQN Units 1 and 2 is the NRC-approved Westinghouse proprietary core analysis system called BEACON.

Currently, SR 3.3.1.3 requires plant operators to compare the Nuclear Instrumentation System (NIS) excore neutron detectors channel output to the incore system AFD measurement by using movable incore detectors. SR 3.3.1.6 requires plant operators to calibrate excore channels to agree with the incore measurement system. The generic phrase “core power distribution” is substituted for “incore detector,” as this would allow the use of either a PDMS or the movable incore detectors to compare the results of the incore and NIS excore measurements.

Table 3.3.1-1 Note 1, “2YHUWHPSHUDWXUHǻ7´LVUHYLVHGWRLQFUHDVHWKH.DQG.FRHIILFLHQW YDOXHVLQWKH27ǻ7UHDFWor trip function .2 ZLOOEHLQFUHDVHGIURPq)WRqF; .3 will be increased from 0.00055/psig to 0.0008/psig) to protect the core thermal limits associated with this program. In addition, the Tƍ (nominal Tavg at RTP), Pƍ (nominal RCS RSHUDWLQJSUHVVXUH .-.FRHIILFLHQWVȫȫȫand ȫparameter values are relocated to the COLR consistent with TSTF-339-A, Rev. 2, and WCAP-14483-A. Table 3.3.1-1

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Note ³2YHUSRZHUǻ7´ is revised to relocate the TƎ (nominal Tavg at RTP), .-. coefficients, and ȫ3parameter values to the COLR consistent with TSTF-339-A, Rev. 2, and WCAP-14483-A.

Methodology

The NRC-approved setpoint methodology, WCAP-8745-P-A, “Design Bases for the Thermal Overpower ǻ7 and Thermal Overtemperature ǻ7 Trip Functions,” is included in the references to be added to TS 5.6.3 as discussed in subsection 3.2.8 of this enclosure. The .DQG.FRHIILFLHQWVLQWKH27ǻ7UHDFWRUWULSIXQFWLRQZHUHERWKLQFUHDVHGWRSURWHFWWKH core thermal limits associated with this program to account for uncertainty. The adequacy of WKH27ǻ7UHDFWRUWULSVHWSRLQWVZDVFRQILUPHGE\VKRZLQJWKDWWKH'1%GHVLJQEDVLVLVPHW in the analyses of those events that credit these functions for accident mitigation.

Refer to Attachment 8 for how the Limitations and Conditions were met for each of the NRC approved methodologies that were applied for the Sequoyah Units 1 and 2 fuel transition.

As discussed in the NRC Safety Evaluation for WCAP-14483-A, the justification is provided for why these parameter values may be relocated to the COLR. Cycle-specific TS parameters of Westinghouse plants have been approved by the NRC for inclusion in COLRs. The NRC has also extended this philosophy to the cycle-dependent OTǻ7 and OPǻ7 setpoint parameters and function modifiers. This allows these setpoints to be based on cycle-specific core design parameters, which are verified on a cycle-specific basis, thereby avoiding the necessity of overly conservative TS limits.

Results

NIS excore neutron detectors channel output can be compared to the incore measurement either through the PDMS or by using the movable incore detectors. SR 3.3.1.3 and 3.3.1.6 are revised to implement BEACON core power distribution measurement and replace using “incore detectors” with “core power distribution” measurement, consistent with WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A.

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7KH.DQG.FRHIILFLHQWvalues LQWKH27ǻ7UHDFWRUWULSIXQFWLRQDUHUHYLVHGDQGUHORFDWHG to the COLR along with other cycle-specific and cycle-FRQILUPHG27ǻ7DQG23¨7 parameter values consistent with TSTF-339-A, Rev. 2, and WCAP-14483-A.

Summary and Conclusions

The SR 3.3.1.3 and 3.3.1.6 changes are consistent with the changes from the application of WCAP-12472-P-A, and WCAP-12472-P-A, Addendum 4-A. Application of these NRC approved topical reports is discussed in detail in subsection 3.2.3 of this enclosure. Table 3.3.1-1RWH27ǻ7DQG1RWH23¨7 parameter values are relocated to the COLR.

3.2.5 TS 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Changes

Background

The proposed change would revise TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” to revise the specified RCS total flow rate, and relocate parameter values from LCO 3.4.1 and SRs 3.4.1.1 through 3.4.1.4 to the COLR, consistent with NRC approved TSTF-339-A, Rev. 2, and WCAP-14483-A.

The LCO 3.4.1, SR 3.4.1.3, and SR 3.4.1.4 RCS total flow rate is revised from ³JSP´ WRWKH5&67KHUPDO'HVLJQ)ORZ 7') YDOXHRI³JSPDQGJUHDWHUWKDQRUHTXDOWR the limit specified in the COLR.” The LCO 3.4.1 and SR 3.4.1.1 pressurizer pressure value is relocated to the COLR. The LCO 3.4.1 and SR 3.4.1.2 and RCS average temperature value is also relocated to the COLR.

Methodology

As discussed in the NRC SER for WCAP-14483-A, the justification is provided below for why these parameter values may be relocated to the COLR.

The TS limits on the DNB parameters assure that pressurizer pressure, RCS flow, and the RCS T-avg will be maintained within the limits of steady-state operation assumed in the accident analyses. These limits must be consistent with the initial full power conditions considered in the FSAR safety analysis for normal operation and anticipated operational occurrences (AOOs) in which precluding DNB is the primary criterion. The DNB parameter limits are also based on initial conditions assumed for accidents in which precluding DNB is not a criterion.

The minimum RCS total flow rate value currently specified in LCO 3.4.1 will be relocated to the COLR. In accordance with NRC approved WCAP-14483-A, the minimum limit for RCS flow (TDF) is proposed for inclusion in the LCO.

The justification for the minimum limit for RCS flow is contained within WCAP-14483-A and the associated NRC SER. The reason why there is a flow value in LCO 3.4.1 as well as a

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Evaluation of the Transition to Westinghouse RFA-2 Fuel reference to the COLR is explained below. The SER for WCAP-14483-A agreed with the Westinghouse Owners Group request to:

1. Revise TS 3.4.1 of NUREG-1431, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, to relocate the pressurizer pressure, RCS average temperature (T-avg), and RCS total flow rate values to the COLR. The minimum limit for total flow based on that used in the reference safety analysis will be retained in the TS.

The reasoning behind the minimum limit remaining in the TS was discussed in the WCAP-14483-A SER:

Although some plants operate with lower steam generator tube plugging levels and thus higher RCS flow rates than those assumed in the safety analyses, a change in RCS flow is an indication of a physical change to the plant which should be reviewed by the NRC staff. Because of this, the staff recommended that if RCS flow rate were to be relocated to the COLR, the minimum limit for RCS total flow based on a staff approved analysis (e.g., maximum tube plugging) should be retained in the TS to assure that a lower flow rate than reviewed by the staff would not be used.

Because the lowest RCS flow rate used in any of the safety analyses is the Thermal Design Flow (TDF), the value for the ‘minimum limit for total flow’ which complies with the WCAP-14483-A SER requirement is the TDF value of 360,000 gpm.

Cycle-specific TS parameters of Westinghouse plants have been approved by the NRC for inclusion in COLRs. This allows the cycle-specific core design parameters to be verified on a cycle-specific basis, thereby avoiding the necessity of overly conservative TS limits.

Results

The pressurizer pressure value of “2220 psia” in LCO 3.4.1 and SR 3.4.1.1 is relocated to the COLR, and replaced with “greater than or equal to the limit specified in the COLR.”

The RCS average temperature value “583°F” in LCO 3.4.1 and SR 3.4.1.2 is relocated to the COLR, and replaced with “less than or equal to the limit specified in the COLR.”

The RCS total flow rate is revised from “378,400 JSP´WR³360,000 gpm and greater than or equal to the limit specified in the COLR” in LCO 3.4.1, SR 3.4.1.3, and SR 3.4.1.4.

These changes are not required to support the SQN transition to Westinghouse fuel; rather, it is an opportunity to make TS 3.4.1 more consistent with NUREG-1431 and provide for more operational flexibility in the future.

Summary and Conclusions

The RCS total flow rate is revised, and the pressurizer pressure and RCS average temperature values are relocated to the COLR with other cycle-confirmed parameter values.

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3.2.6 TS 4.2.1 Fuel Assemblies Changes

Background

This proposed license amendment (Enclosure 1) would modify the SQN Units 1 and 2 Technical Specifications (TS) 4.2.1, “Fuel Assemblies,” and 5.6.3, “Core Operating Limits Report,” to allow the use of Optimized ZIRLO as an approved fuel rod cladding material. The change would revise TS 4.2.1 to add Optimized ZIRLO as a fuel assembly cladding material in accordance with WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A (Proprietary) “Optimized ZIRLO” (Reference 16). The proposed change to TS 4.2.1 would also delete the discussion of Framatome report BAW-2328 consistent with the proposed implementation of Westinghouse core safety analysis methodologies, which are discussed in subsection 3.2.8 of this enclosure. Also, as mentioned earlier, Attachment 10 has been added to provide a summary of the revised safety analysis impact to the UFSAR.

In support of this license amendment request, Enclosure 5 contains an exemption request in accordance with 10 CFR 50.12, “Specific exemptions,” from certain requirements of 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,” DQG&)5$SSHQGL[.³ECCS Evaluation Models.” This exemption request relates solely to the specific type of cladding material specified in these regulations for use in light water reactors. As written, the regulations presume use of either Zircaloy or ZIRLO fuel rod cladding. The exemption is required because Optimized ZIRLO has a slightly different composition than Zircaloy or ZIRLO.

Optimized ZIRLO was developed to meet the needs of longer operating cycles with increased fuel discharge burnup and fuel duty. Fuel rod internal pressure (resulting from the increased fuel duty, use of integral fuel burnable absorbers, and corrosion and temperature feedback effects) has become more limiting with respect to fuel rod design criteria. Reducing the associated corrosion buildup and thus minimizing temperature feedback effects provides additional margin to the fuel rod internal pressure design criterion. Compared to ZIRLO, the lower tin content and microstructure difference of Optimized ZIRLO provides a reduced corrosion rate while maintaining the benefits of mechanical strength and resistance to accelerated corrosion from abnormal chemistry conditions.

TVA currently plans to install Optimized ZIRLO clad fuel rods as part of the fuel transition.

Methodology

Reference 16 provides the details and test results of Optimized ZIRLO compared to ZIRLO fuel rod cladding material. Reference 16 also contains the material properties to be used in various models and methodologies when analyzing Optimized ZIRLO fuel rod cladding. The NRC safety evaluation for Reference 16 contains ten limitations and conditions that are addressed in Attachment 8 to this enclosure.

Subsection 3.2.8 of this enclosure discusses the proposed changes to SQN Units 1 and 2 TS 5.6.3, “Core Operating Limits Report” which includes adding Westinghouse Electric Company LLC (Westinghouse) topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLO,” (Reference 16) to the list of analytical methods used to determine the core operating limits approved by the NRC.

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Lead test assemblies (LTAs) with fuel pellets made from highly enriched reprocessed uranium blended down with natural uranium were evaluated by BAW-2328, “Blended Uranium Lead Test Assembly Design Report,” July 1998. These LTAs will not be used in reload cores after the SQN Unit 1 and 2 transition to Westinghouse RFA-2 fuel.

Results

TS 4.2.1, “Fuel Assemblies,” is revised to add “Optimized ZIRLO”.

The following sentence is deleted from TS 4.2.1: “Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 1 Operating Cycle 12.”

Summary and Conclusions

TS 4.2.1 is revised to add Optimized ZIRLO as an approved fuel rod cladding material and the reference to Framatome report BAW-2328 is removed from TS 4.2.1.

3.2.7 TS 4.2.2 Control Rod Assemblies Changes

Background

TS 4.2.2 is revised to require 52 RCCAs with no full-length control rod assembly in core location H-08 for both SQN Units 1 and 2. This proposed change is discussed in Attachment 9 to this enclosure.

Summary and Conclusions

Based on the assessments summarized in Attachment 9, TVA has determined that the SQN reactor cores can be safely designed and operated for the life of the plant with 52 control rods.

3.2.8 TS 5.6.3 Core Operating Limits Report Changes

Background

TS 5.6.3.a, “Core Operating Limits Report,” is revised to add additional core operating limits for Limiting Conditions for Operation (LCOs) 2.1.1, 3.1.4, 3.1.8, 3.3.1, and 3.4.1. In addition, the existing LCO titles will be revised consistent with the proposed TS changes discussed above. TS 5.6.3.b is revised to change analytical methods used to determine the core operating limits from Framatome methods to Westinghouse core safety analysis methodology references. This includes the application of the FSLOCA EM to evaluate the peak cladding temperatures for SQN Units 1 and 2 large-break and small-break LOCAs (LBLOCA and SBLOCA).

The thermal limits on the Framatome HTP fuel ensure that DNB and fuel centerline melt (FCM) will not occur if DNB and FCM do not occur in the Westinghouse RFA-2 fuel. Therefore, SLs 2.1 do not contain explicit limits for the Framatome HTP fuel. In addition, keeping the Framatome HTP fuel within operating limits of previous HTP cores, and applying additional thermal limit restrictions on this fuel, ensures that the Westinghouse

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RFA-2 fuel will always be limiting and establishes core operating limits for the transition cores. While the Framatome HTP fuel remains within compliance with the Framatome methodologies listed in the current TS 5.6.3, these Framatome methods no longer establish core operating limits or PCT during a LOCA and have been removed.

Methodology

Eighteen Westinghouse topical reports represent the methodologies used to determine the values presented in the revised SQN Units 1 and 2 COLR template (Attachment 6).

WCAP-8745-P-$ PHWKRGRORJ\IRU27¨7DQG23¨75HDFWRU7ULS6\VWHPVHWSRLQWVLQ76

WCAP-9272-P-A (methodology for Shutdown Margin, Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, Axial Flux Difference, Heat Flux Hot Channel Factor, Nuclear Enthalpy Rise Hot Channel Factor, DNB Limits, Refueling Pool Boron Concentration)

WCAP-10216-P-A Revision 1A (methodology for Axial Flux Difference limits with Relaxed Axial Offset Control and Heat Flux Hot Channel Factor (W(z)) Surveillance Requirements for FQ)

WCAP-10444-P-A (methodology for Axial Flux Difference and Heat Flux Hot Channel Factor Limits)

WCAP-10444-P-A Addendum 2-A (methodology for Axial Flux Difference and Heat Flux Hot Channel Factor Limits)

WCAP-10965-P-A (methodology for Shutdown Margin, Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, Axial Flux Difference, Heat Flux Hot Channel Factor, Nuclear Enthalpy Rise Hot Channel Factor, Refueling Pool Boron Concentration)

WCAP-10965-P-A, Addendum 2-A, Revision 0 (methodology for Shutdown Margin, Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, Heat Flux Hot Channel Factor, Nuclear Enthalpy Rise Hot Channel Factor, Axial Flux Difference, TS 3.4.1 DNB Limits, Refueling Pool Boron Concentration)

WCAP-11397-P-A (methodology for Reactor Core Safety Limits, Nuclear Enthalpy Rise Hot Channel Factor, TS 3.4.1 DNB Limits)

WCAP-12610-P-A (methodology for Axial Flux Difference and Heat Flux Hot Channel Factor Limits)

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A (methodology for Axial Flux Difference and Heat Flux Hot Channel Factor Limits)

WCAP-14565-P-A (methodology for DNB Safety Limit, Nuclear Enthalpy Rise Hot Channel Factor, and TS 3.4.1 DNB Limits)

WCAP-14565-P-A, Addendum 1-A, Revision 0 (methodology for DNB Safety Limit)

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WCAP-14565-P-A, Addendum 2-P-A, Revision 0 (methodology for DNB Safety Limit)

WCAP-15025-P-A (methodology for DNB Safety Limit and Nuclear Enthalpy Rise Hot Channel Factor)

WCAP-16045-P-A (methodology for Shutdown Margin, Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, Heat Flux Hot Channel Factor, Nuclear Enthalpy Rise Hot Channel Factor, Axial Flux Difference, TS 3.4.1 DNB Limits, Refueling Pool Boron Concentration)

WCAP-16045-P-A, Addendum 1-A (methodology for Moderator Temperature Coefficient)

N WCAP-16996-P-A, Revision 1 (methodology for FQ and F ¨H limits)

WCAP-17661-P-A, Revision 1 (methodology for control bank insertion limits, FQ limits, and AFD limits)

Many of the methodology topical reports being added to TS 5.6.3.b are discussed elsewhere in this amendment request and listed below.

WCAP-8745-P-A (see subsection 3.2.4) WCAP-9272-P-A (see subsection 3.2.1) WCAP-11397-P-A (see subsection 3.2.1) WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A (see subsection 3.2.6) WCAP-14565-P-A and Addendum 1-A and Addendum 2-A (see subsection 3.2.1) WCAP-15025-P-A (see subsection 3.2.1) WCAP-16996-P-A Revision 1 (see Enclosures 2 and 3) WCAP-17661-P-A, Revision 1 (see subsection 3.2.3)

Refer to Attachment 8 for how Limitations and Conditions were met for each of these core safety analysis methodology references.

Redistribution of flow in pressurized water reactor cores is a well-documented and modeled phenomenon that occurs generally because of differences in hydraulic resistance, i.e., the differences in mixing vane grid loss coefficients between the two fuel types. In a mixed core, with assemblies having different overall hydraulic resistance, the local hydraulic resistance differences are also a mechanism for flow redistribution. This redistribution results in the fluid velocity vector having a lateral component as well as the dominant axial component. The lateral component is commonly referred to as crossflow. The crossflow induced by local hydraulic resistance differences will typically impact the mechanical design of the fuel assemblies, as well as the safety analysis of the core.

In safety analyses, crossflow affects DNB because the flow redistribution affects both mass velocity and enthalpy distributions. With the current DNB correlations that will be licensed for use at Sequoyah, WRB-2M and ABB-NV, the flow redistribution affects the prediction of minimum DNBR. As such, the design procedure for the DNB analysis of a mixed core is based on the principle that once the transition core DNBR penalty is determined, all further plant-specific analyses may proceed as if it were a full core of one fuel type.

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Transition cores are analyzed as if they were full cores of one assembly type by applying the applicable transition core DNBR penalties. Transition core DNBR penalties are calculated according to the methodology documented in WCAP-11837-P-A (Reference 28). It is noted that with the exception of the inlet group, the fuel assembly loss coefficient and the section/grid loss coefficients for the Westinghouse 17x17 RFA-2 fuel (w/IFMs) are lower than that of the Framatome HTP fuel. It is therefore expected that there will be a transition core penalty levied against the Westinghouse fuel for bottom-skewed shapes, but not for double-humped or top-skewed shapes. The penalties on the 17x17 RFA-2 fuel are calculated with the VIPRE-W code (Reference 20).

Typically, transition core penalties are reported as a function of the percentage of each fuel type in the core, as approved in Reference 28. However, for the Sequoyah reload transition safety analysis, it was determined that a bounding transition core penalty would be reported for the 17x17 RFA-2 fuel, for separate use with the WRB-2M and ABB-NV correlations.

The transition core DNBR penalty for the Sequoyah reload transition safety analysis is based on a comparison of the DNBR results obtained by modeling two different configurations of 3x3 fuel assembly arrays. One model represents a uniform array of Westinghouse 17x17 RFA-2 fuel assemblies (w/IFMs) and the other model represents a mixed array of the Westinghouse and Framatome fuel types, i.e., one Westinghouse assembly surrounded by eight Framatome assemblies.

The VIPRE-W models employ the WRB-2M DNB correlation for at-power events and for analyses applicable to the region above the first mixing vane grid. The WRB-2M correlation is applicable for pressures ranging from 1495 psia to 2425 psia, and for a local quality ranging from -0.10 to 0.29. For the rod withdrawal from subcritical (RWFS) (below the first mixing vane grid) and hot zero power (HZP) steam line break (SLB) events and any other analyses/conditions where the WRB-2M correlation is not applicable, the W-3 or the W-3 alternative correlations are employed. The ABB-NV correlation is applicable for pressures UDQJLQJIURPWRSVLDDQGIRUTXDOLW\+=36/%FDVHVZLOl be evaluated using the WLOP CHF correlation. The WLOP correlation is applicable for pressures ranging from 185 SVLWRSVLDQGIRUDTXDOLW\

The transition core DNBR penalties that are reported in subsection 3.2.9 are only applicable to Westinghouse 17x17 RFA-2 fuel; they are expected to bound the safety analyses for reload evaluations and for anticipated Chapter 15 events. For conservatism, the design limit transition core penalty has been rounded up by >10.0 percent from the transition core penalty that was obtained from the mixed core analysis. The DNBR results were computed within the respective correlation limits for quality and pressure. Sufficient available DNBR margin to cover the cycle-specific transition core penalties for each fuel type will be demonstrated on a cycle-specific basis.

A separate calculation was performed with the VIPRE-W code to analyze the impact of the transition core effect on the Framatome HTP fuel. The same methodology and process that was used on the RFA-2 assemblies described above was applied to the HTP fuel. The WRB-1 correlation was used in the HTP transition core evaluation, although it is not directly licensed for use with the HTP fuel. The WRB-1 correlation was used to show the general impact of a transition core scenario on the HTP fuel, as it is an older more conservative correlation compared to the WRB-2M and newer correlations, and is applicable over the full range of plant

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conditions. The main focus on the transition core evaluation for the HTP fuel was quantifying the flow loss present during a transition core scenario with the VIPRE-W code.

The transition core DNBR penalty for the Sequoyah reload transition safety analysis is based on a comparison of the DNBR results obtained by modeling two different configurations of 3x3 fuel assembly arrays. One model represents a uniform array of Framatome 17x17 HTP fuel assemblies and the other model represents a mixed array of the Westinghouse and Framatome fuel types, i.e., one HTP assembly surrounded by eight RFA-2 assemblies.

The maximum calculated transition core DNB penalty on the Framatome HTP fuel was 2.14 percent in a mixed core scenario with the WRB-1 correlation. The total flow reduction in the limiting HTP assembly through the top of the heated length was approximately 1.64 percent. Another 7.30 percent flow left the limiting HTP assembly above the heated length, for a total of 8.94 percent flow reduction. The 7.30 percent flow loss after the heated length of the fuel will have little impact on the CHF calculations. Therefore, this flow reduction’s impact on CHF and '1%ZLOOEHRIIVHWE\WKHSHUFHQW)ǻ+UHGXFWLRQDSSOLHGWRWKH+73IXHOGXULQJWKHWUDQVLWLRQ core cycles. $FDVHZDVUXQZLWKWKH+73IXHODWWKHUHGXFHG)ǻ+YDOXHLQD[matrix surrounded by RFA-2 fuel and resulted in a DNB increase from the full core case due to the UHGXFHG)ǻ+YDOXHVKRZLQJWKDWWKHSHUFHQW)ǻ+UHGXFWLRQRIIVHWVWKHWUDQVLWLRQFRUH'1% effect.

As an additional note, Westinghouse methods address the loss of the Leading Edge Flow Meter (LEFM) through administrative procedures. The uncertainty associated with the LEFM is 0.7% of 3455 MWt and the uncertainty associated with the feedwater venturis is 2% of 3411 MW. Upon loss of the LEFM, operation may continue at the uprated thermal power of 3455 MWt while continuing to use the power indications from the Nuclear Instrumentation System (NIS) power range channels. If the LEFM has not been returned to service prior to the required 12-hour NIS output adjustment (SR 3.3.1.2), then the reactor power must be administratively restricted to 98.7 percent of 3455 MWt (3411 MWt) and operated at a reduced power level until the LEFM is brought back into service to allow for the 2% venturi-based power measurement uncertainty. Consistent with the approach used at plants with Westinghouse fuel, this measure precludes the need for adjustment of the COLR limits to maintain the validity of the safety analyses).

Results

Attachment 6 presents the revised SQN Units 1 and 2 COLR template. Eighteen Westinghouse topical report references represent the methodologies used to determine the values presented in Attachment 6. The previous B&W, Areva, or Framatome methods are not needed to justify the transition to Westinghouse RFA-2 fuel. Where necessary, transition core penalties or N conservative operational limits are set for the RFA-2 vs. HTP fuel (e.g., a 5 percent F ǻ+ reduction (from 1.70 to 1.61) will be applied to the Framatome HTP fuel during the transition core cycles, a minimal (<1.0%) transition core DNBR penalty will be applied to the RFA-2 fuel while co-resident HTP fuel is loaded, the surveillance FQ limit for all fuel will be set at 2.62 in order to maintain a single FQ surveillance limit consistent with the current limit for Framatome HTP fuel, and a transition core PCT penalty of 23°F for Framatome HTP fuel resulting from a hydraulic mismatch during transition cycles to Westinghouse RFA-2 fuel).

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Summary and Conclusions

The analytical methods used to determine the core operating limits have been previously reviewed and approved by the NRC. This satisfies the stipulation in NRC Generic Letter 88-16 for relocating cycle-specific variables to the COLR. Attachment 8 goes into more detail on the application of these methodologies as well as others, such as PAD5, that aren’t specifically cited in the COLR.

3.2.9 Operating License Conditions 2.C (25) and 2.C (18) Changes

Background

When the Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies the: N x HTP fuel assemblies F ǻ+ shall be maintained less than 1.61. x RFA-2 fuel assemblies the DNBR limit shall be reduced by: — 0.25% for the WRB-2M critical heat flux correlation — 0.50% for the ABB-NV critical heat flux correlation

Summary and Conclusions

Removal of the current license conditions is appropriate as the conditions were associated with a mixed core DNBR penalty resulting from the Framatome fuel conversion in 1997. The proposed license conditions are appropriate for the mixed cores until a homogeneous core of Westinghouse RFA-2 fuel exists. Additional information is provided above in subsection 3.2.8.

3.3 Conclusion

The proposed changes are needed to support the transition to Westinghouse RFA-2 fuel. The changes are based upon NRC approved methods and methodologies and with the exception of the application of WCAP-17661-P-A Revision 1, have received NRC approval for identical requested changes from other licensees.

Additionally, due to the application of a large number of Topical Reports for the fuel transition, Attachment 8 has been created to address SQN Units 1 and 2 compliance with the Limitations and Conditions stipulated in the NRC safety evaluation approving each of the Topical Reports, with one exception. The FSLOCA summary report (Enclosure 2 of this LAR) contains proprietary information, including the proprietary disposition to the Limitations and Conditions contained in the NRC safety evaluation approving WCAP-16996-P-A, Revision 1, “Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),” November 2016.

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4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

4.1.1 Regulations

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications (TSs) as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public. The U.S. Nuclear Regulatory Commission’s (NRC’s) requirements related to the content of the TSs are contained in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) which requires that the TSs include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements per 10 CFR 50.36(c)(3); (4) design features; and (5) administrative controls.

This amendment request involves each of the five categories in 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5).

10 CFR 50.46 requires, in part, that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in 10 CFR 50.46(b). $SSHQGL[.WR 10 CFR Part 50 establishes the regulations for conservative ECCS evaluation models. Enclosure 5 contains an exemption UHTXHVWIURP&)5DQG$SSHQGL[.WR&)53DUW, in accordance with 10 CFR 50.12. In Reference 16, the NRC approved Addendum 1-A to WCAP-12610-P-A and CENPD-404-P-A for the use of Optimized ZIRLO as an acceptable fuel rod cladding material for Westinghouse fuel designs.

TS 4.2.2, “Control Rod Assemblies,” describes a Design Feature required per 10 CFR 50.36(c)(4). The proposed change does not eliminate the design feature requiring control rod assemblies. Rather, it allows for a revised number of control rod assemblies.

TS 3.1.4, “Rod Group Alignment Limits,” requires all shutdown and control rods to be operable. Because the control rod in location H-08 would be permanently removed under the proposed change, this TS requirement would not be applicable to that control rod position. As such, no changes to TS 3.1.4 are required.

The requirements of 10 CFR 50.62(c) concerning an alternate rod injection system were also assessed.

Removal of Control Rod H-08 does not impact either the reactor protection system or ATWS Mitigation System Actuation Circuitry because the trip reactivity remains bounding. The changes to other parameters described in the license amendment request do not impact the ATWS analysis. The all-rods-out moderator temperature coefficient (MTC) is not impacted by removal of Control Rod H-08. Therefore, the requirements of 10 CFR 50.62(c)(1) continue to be met and there is no impact to the ATWS analysis.

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The requirements of 10 CFR 50.62(c)(2) are not applicable because SQN is a Westinghouse pressurized water reactor. Also, 10 CFR 50.62(c)(3), (4), and (5) are applicable to boiling water reactors and therefore not applicable to SQN.

4.1.2 General Design Criteria

As noted in the SQN UFSAR Section 3.1.2, the Sequoyah Nuclear Plant was designed to meet the intent of the Proposed General Design Criteria for Nuclear Power Plant Construction Permits published in July, 1967. The Sequoyah construction permit was issued in May, 1970. This UFSAR, however, addresses the NRC General Design Criteria (GDC) published as Appendix A to 10 CFR 50 in July 1971.

The SQN UFSAR contains the following relevant GDCs. Compliance with these GDCs is described in Section 3.1.2 of the SQN UFSAR.

Criterion 4 – Environmental and Missile Design Bases Structures, systems, and components important to safety are designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCA. These structures, systems and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

Criterion 10 – Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Criterion 11 – Reactor Inherent Protection The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

Criterion 12 – Suppression of Reactor Power Oscillations The reactor core and associated coolant, or, control and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

Criterion 20 -- Protection System Functions The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

Criterion 23 – Protection System Failure Modes The Protection System shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the

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Criterion 25 – Protection System Requirements for Reactivity Control Malfunctions The Protection System shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the Reactivity Control Systems, such as accidental withdrawal (not ejection or dropout) of control rods.

Criterion 26 – Reactivity Control System Redundance and Capability Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

Criterion 27 – Combined Reactivity Control Systems Capability The Reactivity Control Systems shall be designed to have a combined capability, in conjunction with poison addition by the Emergency Core Cooling System (ECCS), of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

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

Criterion 29 – Protection against Anticipated Operational Occurrences The Protection and Reactivity Control Systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

Criterion 35 – Emergency Core Cooling A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite

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electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

With the implementation of the proposed changes, SQN Units 1 and 2 continue to meet the applicable regulations and requirements, subject to the previously approved exceptions.

4.1.3 Regulatory Guidance and Miscellaneous References

NRC Generic Letter (GL) 88-16

GL 88-16, “Removal of Cycle-Specific, Parameter Limits from Technical Specifications,” dated October 4, 1988, provides that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle-specific COLR that is required to be submitted to the NRC every operating cycle or each time it is revised. Consistent with the guidance in NRC GL 88-16, SQN Units 1 and 2 TS 5.6.3, “Core Operating Limits Report (COLR)” requires the following.

x The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. x The COLR, including any mid-cycle revisions or supplements, is to be provided upon issuance for each reload cycle to the NRC. x The TS include a list of references of the NRC-approved methodologies that are used to determine the cycle-specific core operating limits. TS 5.6.3 identifies the NRC-approved analytical methodologies that are used to determine the core operating limits for SQN Units 1 and 2.

Upon approval of the proposed LAR, the guidance in GL 88-16 continues to be met because the proposed change will continue to specify the NRC-approved methodologies used to determine the core operating limits open and therefore provide input to the associated RTS channel.

Standard Review Plan NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition” (SRP), Section 4.2, “Fuel System Design,” provides regulatory guidance to the NRC staff for the review of fuel rod cladding materials and fuel system. According to SRP Section 4.2, the fuel system safety review provides assurance as follows. x The fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs). x Fuel system damage is never so severe as to prevent control rod insertion when it is required. x The number of fuel rod failures is not underestimated for postulated accidents. x Coolability is always maintained.

Upon approval of the proposed LAR, the guidance in SRP 4.2 continues to be met because the proposed change will continue to comply with the NRC-approved design standards discussed in Section 3.1 of this enclosure.

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Regulatory Guide (RG) 1.236 RG 1.236 Revision 0, “Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents,” was reviewed for impact on this LAR. The results of this review are discussed in footnote 1 to the table in Attachment 10 to this enclosure. No unacceptable impact was found during this review.

Conclusion

There are no changes being proposed in this amendment application such that commitments to the regulations and regulatory guidance documents above would come into question. The evaluations documented above confirm that Sequoyah Units 1 and 2 will continue to comply with all applicable regulatory requirements.

4.2 Precedent

The following precedents are related to the proposed TS change in this submittal.

FSLOCA Relative to FSLOCA, this LAR is similar to one approved by the NRC for the Diablo Canyon Nuclear Power Plant (ML19316A109), which revised Diablo Canyon TS 5.6.5b, “Core Operating Limits Report (COLR),” to replace the existing LOCA methodologies with the NRC-approved LOCA methodology contained in WCAP-16996-P-A, Revision 1.

Use of Optimized ZIRLO Relative to the use of Optimized ZIRLO, this LAR is similar in nature to the following LARs and exemption requests approved by the NRC that authorized the use of Optimized ZIRLO fuel rod cladding: x Beaver Valley Power Station, Units 1 and 2 (ML18022B116 and ML17313A550) x Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (ML17319A214 and ML17319A107) x Wolf Creek Generating Station (ML16179A293 and ML16179A440)

TVA also reviewed the related RAI responses associated with the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 LAR (ML17153A373) and the Wolf Creek Generating Station LAR (ML16161A509) and incorporated the requested information as appropriate.

Removal of RCCA at location H-08 The changes proposed in this LAR for permanent operation of SQN-1 and SQN-2 with no RCCA in core location H-08 is similar to one approved by NRC for South Texas Project, Unit 1, in LA No. 211 issued on December 21, 2016 (ML16319A010).

COLR Parameter Relocation The changes proposed in this LAR for relocating RCS-related cycle specific parameters are consistent with similar changes approved by the NRC for other nuclear power plants. These include changes approved for the following. x Braidwood Station, Units 1. and 2, and Byron Station, Units 1 and 2, by License Amendment (LA) Nos. 106 and 113, respectively, issued on May 15, 2000 (ML003717646) x Wolf Creek Generating Station by LA No. 144, issued on March 28, 2002 (ML020180190)

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x Catawba Nuclear Station, Units 1 and 2 by LA Nos. 210 and 204, respectively, issued on December 19, 2003 (ML033570127) x McGuire Nuclear Station, Units 1 and 2 by LA Nos. 219 and 201, respectively, issued on January 14, 2004 (ML040140090) x Millstone Power Station, Unit No. 3 by LA No. 218 issued on March 9, 2004 (ML033450435) x Indian Point Nuclear Generating Unit No. 3 by LA No. 225, issued on March 24, 2005 (ML050600380).

BEACON Relative to the use of BEACON, this LAR is similar to one approved by NRC for Watts Bar Unit 1 in LA No. 82, issued on October 27, 2009 (ML092710381)

4.3 No Significant Hazards Consideration

The Tennessee Valley Authority (TVA) is proposing an amendment to revise the Sequoyah Nuclear Plant (SQN) Units 1 and 2 Technical Specifications (TSs) as follows.

The proposed change would revise SQN Units 1 and 2 Technical Specification (TS) 5.6.3, “Core Operating Limits Report,” to replace the loss-of-coolant accident (LOCA) analysis evaluation model references with reference to the FULL SPECTRUM Loss-of-Coolant Accident (FSLOCA) Evaluation Model analysis applicable to both SQN Unit 1 and Unit 2 with replacement steam generators.

TVA is applying the FSLOCA Evaluation Model to determine the peak cladding temperatures for LBLOCA and SBLOCA. The use of the FSLOCA Evaluation Model results in a reduction in the peak cladding temperature in analyses of LBLOCA and SBLOCA. The safety analysis process for each reload design will continue to demonstrate that all regulatory criteria are met.

TVA is applying topical report WCAP-17661-P-A, Revision 1, which corrects a non-conservatism in the Standard Technical Specifications for Westinghouse PWRs (NUREG-1431). Specifically, the proposed change revises TS 3.2.1 Conditions and Surveillance Requirements to eliminate non-conservatisms described in NSAL-09-5 Revision 1 and NSAL-15-1. The proposed change also revises TS 5.6.3 to include WCAP-17661-P-A, Revision 1, as a reference. STS 3.2.1 from NUREG-1431 is adopted in areas not specifically revised by WCAP-17661-P-A Revision 1. Deviations to the TS 3.2.1 Condition B Completion Times in WCAP-17661-P-A Revision 1 are also proposed.

TS 3.2.2, “Nuclear Enthalpy Rise Hot Channel Factor Fǻ+(X,Y),” is revised to reflect TS 3.2.2, N “Nuclear Enthalpy Rise Hot Channel Factor (F 'H),” in NUREG-1431. Adopting the latest version of the NRC-approved Standard Technical Specification 3.2.2 essentially returns the

Sequoyah Units 1 and 2 Technical Specifications to their licensing basis for the F¨H LCO prior to the conversion from Westinghouse fuel to Framatome Cogema Fuel, except for two Completion Times (CTs).

This proposed license amendment would modify the SQN Units 1 and 2 Technical Specifications (TS) 4.2.1, “Fuel Assemblies,” and 5.6.3, “Core Operating Limits Report,” to allow the use of Optimized ZIRLO as an approved fuel rod cladding material. The change would revise TS 4.2.1 to add Optimized ZIRLO as a fuel assembly cladding material in accordance with WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A (Proprietary), “Optimized ZIRLO”

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(Reference 16). The proposed change to TS 4.2.1 would also delete the discussion of Framatome report BAW-2328 consistent with the proposed implementation of Westinghouse core safety analysis methodologies.

TS 4.2.2 is revised to require 52 RCCAs with no full-length control rod assembly in core location H-08 for both Units 1 and 2.

The proposed changes will relocate cycle-specific parameter limits from TSs 2.1.1, 3.3.1, and 3.4.1 to the CORE OPERATING LIMITS REPORT (COLR). The proposed changes to TS 2.1.1, “Reactor Core Safety Limits,” will relocate the revised Reactor Core Safety Limits figure to the COLR and add new Safety Limits for departure from nucleate boiling ratio (DNBR) and peak fuel centerline temperature. The proposed changes to TS Table 3.3.1-1, “Reactor Trip System Instrumentation,” will relocate the Overtemperature ¨T and Overpower ¨T setpoint parameters (nominal RCS average temperature, nominal RCS operating pressure, .YDOXHVand dynamic compensation time constants (Ĳ values) to the COLR. The proposed changes to TS 3.4.1, “RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,” will relocate the pressurizer pressure and RCS average temperature values to the COLR and revise the thermal design flow rate. TS 5.6.3, “CORE OPERATING LIMITS REPORT (COLR),” will be revised to add several topical reports and allow all cited topical reports to be identified by title and number only.

The proposed changes to the TSs also involve allowing the use of the Best Estimate Analyzer for Core Operations Nuclear (BEACON) Power Distribution Monitoring System (PDMS) to perform core power distribution surveillances. The proposed changes allow for the power distribution surveillances to be performed by PDMS rather than using the Movable Incore Detector (MID) System. In addition, a RAOC-type axial flux distribution methodology is planned to be implemented along with the proposed implementation of PDMS. The proposed TS changes for PDMS and RAOC are supported by the NRC approved methodologies. All reload specific input will be confirmed via an approved reload methodology employed by TVA and Westinghouse.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, “Issuance of amendment,” as discussed below:

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

Response: No

The proposed change to SQN Units 1 and 2 TS 5.6.3, “Core Operating Limits Report,” to replace the LOCA analysis evaluation model references with the FSLOCA Evaluation Model analysis applicable to both SQN Unit 1 and Unit 2 with replacement steam generators. These changes implement a Nuclear Regulatory Commission (NRC) approved LOCA evaluation model. The analysis results for SQN Units 1 and 2, based on using the new evaluation model, meet the regulatory requirements of 10 CFR 50.46. The use of a new NRC-approved LOCA evaluation model will not increase the potential for an accident. Therefore, the possibility of an accident is not increased by the proposed changes. Because the reactor core meets the regulatory requirements of 10 CFR 50.46 after a

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

postulated LOCA, the consequences of an accident are not increased by the proposed changes.

The proposed amendment will allow the use of Optimized ZIRLO clad nuclear fuel at SQN Units 1 and 2. The NRC approved topical report WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, which addresses Optimized ZIRLO fuel rod cladding and demonstrates that Optimized ZIRLO fuel rod cladding has essentially the same properties as ZIRLO fuel rod cladding. The use of Optimized ZIRLO fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO meets the fuel design acceptance criteria and hence does not significantly affect the consequences of an accident.

The proposed changes described in WCAP-17661, Revision 1, resolve non-conservative TS Required Actions identified via Westinghouse NSAL-09-5, Revision 1. The proposed changes also resolve non-conservative TS Surveillance Requirements identified via Westinghouse NSAL-15-1. Operation in accordance with the revised TS ensures that the assumptions for initial conditions of key parameter values in the safety analyses remain valid and does not result in actions that would increase the probability or consequences of any accident previously evaluated.

TVA has performed a multi-cycle assessment on SQN reactor cores. Operation of SQN with the H-08 control rod removed will not involve a significant increase in the probability or consequences of an accident previously evaluated. Shutdown Margin (SDM) is reduced by the absence of the H-08 control rod, but remains bounded by the limits specified by the COLR. Because the impacts on the cycle-specific nuclear design parameters are bounded by the conservative input values used in the fuel transition accident analyses.

Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no design changes. The design of the reactor trip system (RTS) instrumentation and engineered safety feature actuation system (ESFAS) instrumentation will be unaffected and these protection systems will continue to function in a manner consistent with the plant design basis. All design, material, and construction standards that were applicable prior to this amendment request will be maintained.

The proposed changes will not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. Accidents/events were analyzed/evaluated using approved methods and codes. Results were within established acceptance criteria, and the plant’s assumed source term remains bounding; therefore, there’s no increase in consequences.

These changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions. Additional Safety Limits on the DNB design basis and peak fuel centerline temperature are being imposed in TS 2.1.1, “Reactor Core Safety Limits,” and the Reactor Core Safety Limits figure is being relocated to the COLR. The additional Safety Limits are consistent with the values stated in the UFSAR. The proposed changes do not, by themselves, alter any of the relocated parameter limits. The

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removal of the cycle-specific parameter limits from the TS does not eliminate existing requirements to comply with the parameter limits. TS 5.6.3 ensures that the analytical methods used to determine the core operating limits meet NRC reviewed and approved methodologies. TS 5.6.3 ensures that applicable limits of the safety analyses are met.

Power Distribution Monitoring System (PDMS) performs continuous core power distribution monitoring. It in no way provides any protection or control system functionality. Fission product barriers are not impacted by these proposed changes. The proposed changes occurring with PDMS will not result in any additional challenges to plant equipment that could increase the probability of any previously evaluated accident. The changes associated with the PDMS do not affect plant systems such that their function in the control of radiological consequences is adversely affected. These proposed changes will therefore not affect the mitigation of the radiological consequences of any accident described in the Updated Final Safety Analysis Report (UFSAR).

Continuous on-line monitoring through the use of PDMS provides significantly more information about the power distributions present in the core than is currently available. This results in more time (i.e., earlier determination of an adverse condition developing) for operator action prior to having any adverse condition develop that could lead to an accident condition or to unfavorable initial conditions for an accident.

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

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

Response: No

The proposed change to SQN Units 1 and 2 TS 5.6.3 to replace the LOCA analysis evaluation model references with the FSLOCA Evaluation Model. The use of this new analytical methodology will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The use of Optimized ZIRLO fuel rod cladding material will not result in adverse changes to the operation or configuration of the facility. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A demonstrated that the material properties of Optimized ZIRLO fuel rod cladding are similar to those of ZIRLO fuel rod cladding. Therefore, Optimized ZIRLO fuel rod cladding will perform similarly to ZIRLO fuel rod cladding, thus precluding the possibility of the fuel rod cladding becoming an accident initiator and causing a new or different kind of accident.

The proposed change to the adopted Westinghouse Standard TS (NUREG-1431) 3.2.1 and 3.2.2 does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). Operation in accordance with the revised TS and its limits precludes new challenges to systems, structures, or components that might introduce a new type of accident. Applicable design and performance criteria will continue to be met and no new single failure mechanisms will be created. The proposed change for resolution of Westinghouse NSAL-09-5, Revision 1 and NSAL-15-1 does not involve the alteration of plant equipment or introduce unique operational modes or accident precursors.

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Operation of SQN with the H-08 control rod removed will not create the possibility of a new or different kind of accident from any accident previously evaluated and the safety evaluations performed with the H-08 control rod removed validated that the impacts to the nuclear design parameters were within the bounds of those assumed in the fuel transition accident analyses. Additionally, by installing a flow restrictor in the H-08 upper internals control rod guide tube, the hydraulic characteristics of the reactor vessel upper internals are unchanged and all plant equipment will continue to meet applicable design and safety requirements.

Relocation of cycle-specific parameter limits has no influence on, nor does it contribute to, the possibility of a new or different kind of accident. The relocated cycle specific parameter limits will continue to be calculated using the NRC reviewed and approved methodologies. The proposed changes do not alter assumptions made in the safety analyses. Operation within the core operating limits will continue to be observed.

As stated previously, the implementation of the PDMS system has no influence or impact on plant operations or safety, nor does it contribute in any way to the probability or consequences of an accident. No safety-related equipment, safety function, or plant operation will be altered as a result of this proposed change. The possibility for a new or different type of accident from any accident previously evaluated is not created because the changes associated with PDMS do not result in a change to the design basis of any plant component or system. The evaluation of the effects of the PDMS changes shows that all design standards and applicable safety criteria limits are met. These changes, therefore, do not cause the initiation of any accident nor create any new failure mechanisms. All equipment important to safety will operate as designed. Component integrity is not challenged. The proposed changes do not result in any event previously deemed incredible being made credible. The PDMS changes will not result in more adverse conditions and will not result in any increase in the challenges to safety systems. The cycle specific variables required by the PDMS are calculated using NRC approved methods. The Technical Specifications (TS) will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.

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

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

Response: No

The proposed change to SQN Units 1 and 2 TS 5.6.3 to replace the LOCA analysis evaluation model references with the FSLOCA Evaluation Model. The analysis results for SQN Units 1 and 2, based on using the new evaluation model, meet the regulatory requirements of 10 CFR 50.46 with increased margin after a postulated LOCA.

WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, demonstrated that the material properties of the Optimized ZIRLO fuel rod cladding are similar to those of ZIRLO fuel rod cladding. Optimized ZIRLO fuel rod cladding is expected to perform similarly to ZIRLO fuel rod cladding for normal operating and accident scenarios, including both loss-of-coolant

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accident (LOCA) and non-LOCA scenarios. The use of Optimized ZIRLO fuel rod cladding will not result in adverse changes to the operation or configuration of the facility.

Operation in accordance with the revised TS and its limits does not impact assumptions made in the safety analyses. This ensures that applicable design and performance criteria associated with the safety analysis will continue to be met and that the margin of safety is not affected. In addition, to address mixed core safety analysis impacts, a combination of PCT and DNB penalties has been imposed.

Operation of SQN with the H-08 control rod removed will not involve a significant reduction in a margin of safety. The margin of safety is established by setting safety limits and operating within those limits. The proposed change does not alter any design basis or safety limit and does not change any setpoint at which automatic actuations are initiated. The proposed change has been evaluated for effects on available shutdown margin, boron worth, trip reactivity as a function of time, and moderator temperature coefficient. The results of these evaluations show that the proposed change does not exceed or alter a design basis or safety limit.

For the relocated values, the proposed changes do not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications. The nominal RTS and ESFAS trip setpoints will remain unchanged. None of the acceptance criteria for any accident analysis will be changed as a result of the relocated values. The development of cycle-specific parameter limits for future reload designs will continue to conform to NRC reviewed and approved methodologies, and will be performed pursuant to 10 CFR 50.59 to assure that plant operation remains within cycle-specific parameter limits.

The margin of safety is not affected by the implementation of PDMS. The margin of safety presently provided by current TS remains unchanged. Appropriate measures exist to control the values of these cycle-specific limits. The proposed changes continue to require operation within the core limits that are based on NRC approved reload design methodologies. The proposed changes continue to ensure that appropriate actions will be taken if limits are violated. These actions remain unchanged. The development of the reload specific limits, including RAOC bands, for future reloads will continue to conform to the methods described in WCAP-9272-P-A which includes a review to assure that operation of the units, within the cycle specific limits, will not involve a reduction in the margin of safety as defined in the basis for any Technical Specification.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92 (c), and, accordingly, a finding of “no significant hazards consideration” is justified.

4.4 Conclusions

In 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,

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that 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 increases 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 criteria 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 needs to be prepared in connection with the issuance of the amendment.

6.0 REFERENCES

1. Westinghouse Report WCAP-10444-P-A, Addendum 2-A (Proprietary), “VANTAGE 5H FUEL ASSEMBLY,” February 1989.

2. Westinghouse Report WCAP-12488-A, Revision 0, “Westinghouse Fuel Criteria Evaluation Process,” October 1994.

3. Westinghouse Letter NSD-NRC-98-5796, “Fuel Criteria Evaluation Process Notification for the 17x17 Robust Fuel Assembly with IFM Grid Design,” October 13, 1998. (ML20154M323)

4. Letter from Henry A. Sepp (Westinghouse) to J. S. Wermiel (U.S. NRC), LTR-NRC-01-44, “Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design (Proprietary),” December 19, 2001.

5. Letter from Henry A. Sepp (Westinghouse) to J. S. Wermiel (U.S. NRC), LTR-NRC-02-55, “Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design, Revision 1 (Proprietary),” November 13, 2002. (ML023190181)

6. WCAP-11397-P-A (Proprietary), “Revised Thermal Design Procedure,” April 1989.

7. TSTF-339-A, Revision 2, “Relocate TS Parameters to COLR,” June 2000. (ML003723269)

8. WCAP-14483-A, “Generic Methodology for Expanded Core Operating Limits Report,” January 1999.

9. NUREG-1431, Revision 4, “Standard Technical Specifications Westinghouse Plants,” November 2011.

10. WCAP-15025-P-A (Proprietary), “Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,” April 1999.

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

11. WCAP-17642-P-A, Revision 1 (Proprietary) “Westinghouse Performance Analysis and Design Model (PAD5),” November 2017.

12. WCAP-17661-P-A, Revision 1 (Proprietary), “Improved RAOC and CAOC FQ Surveillance Technical Specifications,” February 2019. (ML19225C081)

13. WCAP-12472-P-A (Proprietary) “BEACON Core Monitoring and Operations Support System,” August 1994. (ML12270A386)

14. WCAP-12472-P-A, Addendum 4, Revision 0 (Proprietary) “BEACON Core Monitoring and Operation Support System, Addendum 4,” September 2012.

15. WCAP-8745-P-A (Proprietary) ³'HVLJQ%DVHVIRUWKH7KHUPDO2YHUSRZHU¨7DQG Thermal 2YHUWHPSHUDWXUH¨77ULS)XQFWLRQV´6HSWHPEHU

16. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A (Proprietary) “Optimized ZIRLO,” July 2006. (ML062080563)

17. WCAP-16996-P-A, Revision 1 (Proprietary) “Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),” November 2016. (ML17277A131)

18. WCAP-9401-P-A, “Verification Testing and Analysis of the 17 x 17 Optimized Fuel Assembly,” August 1981.

19. PWROG-16043-P-A, Revision 2, “PWROG Program to Address NRC Information Notice 2012-09: ‘Irradiation Effects on Fuel Assembly Spacer Grid Cush Strength’ for Westinghouse and CE PWR Fuel Designs,” November 2019.

20. WCAP-14565-P-A, “VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,” October 1999.

21. Letter from D. S. Collins (USNRC) to J. A. Gresham (Westinghouse), “Modified WRB-2 Correlation WRB-2M for Predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,” February 2006. (ML060320290)

22. WCAP-14565-P-A, Addendum 1-A, Revision 0, “Addendum 1 to WCAP 14565-P-A Qualification of ABB-NV Critical Heat Flux Correlations with VIPRE-01 Code,” August 2004.

23. WCAP-14565-P-A, Addendum 2-P-A, Revision 0, “Addendum 2 to WCAP-14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications,” April 2008.

24. NSAL-09-5, Revision 1, “Relaxed Axial Offset Control FQ Technical Specification Actions,” September 23, 2009.

25. NSAL-15-1, “Heat Flux Hot Channel Factor Technical Specification Surveillance,” February 3, 2015.

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Evaluation of the Transition to Westinghouse RFA-2 Fuel

26. Sequoyah Units 1 and 2 License Amendments 223/214 dated 4-21-97 (ML013320456).

27. WCAP-10216-P-A, Revision 1A (Proprietary) “Relaxation of Constant Axial Offset Control / FQ Surveillance Technical Specification,” February 1994.

28. WCAP-11837-P-A, Revision 0, “Extension of Methodology for Calculating Transition Core DNBR Penalties,” January 1990.

29. BAW-10133NP-A, Rev. 1, Addendum 1 and Addendum 2, “Mark-C Fuel Assembly LOCA-Seismic Analysis,” October 2000. (ML003767624)

30. TSTF-241-A, Revision 4, “Allow Time for Stabilization after Reducing Power due to QPTR out of Limit” (NRC approved on January 13, 1999). (ML040611034)

31. TSTF-290-A, Revision 0, “Revision to Hot Channel Factor Specifications, (NRC approved on June 30, 1999). (ML040630063)

32. Westinghouse Report WCAP-12610-P-A, Revision 0, “VANTAGE+ Fuel Assembly Reference Core Report,” April 1995.

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ATTACHMENT 1

Proposed TS Changes (Mark-Ups) for SQN Unit 1

CNL-20-014 - 11-

F. Post Accident Sampling (Section 22.3, II.B.3)

This condition has been deleted.

H. Instruments for Inadequate Core Cooling (Section 22.3, II.F.2)

(1) By January 1, 1982, TVA shall install a backup indication for incore thermocouples. This display shall be in the control room and cover the temperature range of 200 F - 2000 F.

(2) At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall install reactor vessel water level instrumentation which meets NRC requirements.

I. Upgrade Emergency Support Facilities (Section 22.3, II.A.1.2)

(1) At the first outage of sufficient duration, but no later than startup following the second refueling outage, TVA shall update the Technical Support Facilities to meet NRC requirements.

(2) TVA shall maintain interim emergency support facilities (Technical Support Center, Operations Support Center and the Emergency Operations Facility) until the final facilities are complete.

J. Relief and Safety Valve Test Requirements (Section 22.2, II.D.1)

TVA shall conform to the results of the EPRI test program. TVA shall provide documentation for qualifying (a) reactor coolant system relief and safety valves, (b) piping and supports, and (c) block valves in accordance with the review schedule given in SECY 81-491 as approved by the Commission.

(24) Compliance with Regulatory Guide 1.97

TVA shall implement modifications necessary to comply with Revision 2 of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following An Accident," dated December 1980 by startup from the Unit 2 Cycle 4 refueling outage. Transition Core Peaking Penalties (25) Mixed Core DNBR Penalty

When Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies: TVA will obtain NRC approval prior to startup for any cycle's core that involves a (a) The HTP fuel assemblies FN shall be maintained less than 1.61; reduction in the ǻdepartureH from nucleate boiling ratio initial transition core penalty (b) The RFA-2 fuel assemblies DNBR limit shall be reduced by: below that value stated in TVA's submittal on Framatome fuel conversion dated 1. 0.25% for the WRB-2M critical heat flux correlation 2. 0.50% forApril the 6,ABB-NV 1997. critical heat flux correlation

Renewed License No. DPR 77 September 28, 2015 For information only

This note is for NRC information

TABLE OF CONTENTS Page

1.0 USE AND APPLICATION 1.1 Definitions ...... 1.1-1 1.2 Logical Connectors ...... 1.2-1 1.3 Completion Times ...... 1.3-1 1.4 Frequency ...... 1.4-1

2.0 SAFETY LIMITS (SLs) ...... 2.0-1 2.1 SLs 2.2 SL Violations

3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ...... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ...... 3.0-4

3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(X,Y,Z)) (RAOC-T(Z) Methodology) ...... 3.2.1-1 N 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F ¨H(X,Y)) ...... 3.2.2-1 3.2.3 AXIAL FLUX DIFFERENCE (AFD) ...... 3.2.3-1 3.2.4 QUADRANT POWER TILT RATIO (QPTR) ...... 3.2.4-1

3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation ...... 3.3.1-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation ...... 3.3.2-1 3.3.3 Post Accident Monitoring (PAM) Instrumentation ...... 3.3.3-1 3.3.4 Remote Shutdown Monitoring Instrumentation ...... 3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ...... 3.3.5-1 3.3.6 Containment Ventilation Isolation Instrumentation ...... 3.3.6-1 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation ...... 3.3.7-1 3.3.8 Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation ...... 3.3.8-1 3.3.9 Boron Dilution Monitoring Instrumentation (BDMI) ...... 3.3.9-1

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ...... 3.4.1-1 3.4.2 RCS Minimum Temperature for Criticality ...... 3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits ...... 3.4.3-1 3.4.4 RCS Loops - MODES 1 and 2 ...... 3.4.4-1

SEQUOYAH - UNIT 1 i Amendment 334, SLs 2.0

2.0 SAFETY LIMITS (SLs)

2.1 SLs the COLR 2.1.1 Reactor Core SLs 9

specified In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits shown in Figure 2.1.1-1; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.132 for the BHTP correlation, 1.21 for the BWU-N correlation, and 1.21 for the BWCMV correlation. 1.14 for the WRB-2M

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained < 5080 )GHFUHDVLQJE\)SHU0:'078RIEXUQXSIRU COPER1,&DSSOLFDWLRQVDQG)GHFUHDVLQJE\)SHU 10,000 MWD/MTU of burnup for TACO3 applications.

2.1.2 Reactor Coolant System Pressure SL

In MODES 1, 2, 3, 4, and 5, the RCS SUHVVXUHVKDOOEHPDLQWDLQHG 2735 psig.

2.2 SAFETY LIMIT VIOLATIONS

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

SEQUOYAH – UNIT 1 2.0-1 Amendment 334 6/V Relocate to COLR

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ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B.2.1.2 Initiate boration to restore 1 hour SDM to within limit.

AND

B.2.2 Reduce THERMAL 2 hours 32:(5WR RTP.

AND

B.2.3 Verify SDM is within the Once per limits specified in the 12 hours COLR.

AND

B.2.4 Perform SR 3.2.1.1 and SR KRXUV 3.2.1.2.

AND

% Perform SR 3.2.2.1. KRXUV

AND

B.2.6 Re-evaluate safety GD\V analyses and confirm results remain valid for duration of operation under these conditions.

C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition B not met.

D. More than one rod not D.1.1 Verify SDM is within the 1 hour within alignment limit. limits specified in the COLR.

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6(482<$+±81,7 $PHQGPHQW FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

3.2 POWER DISTRIBUTION LIMITS

3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) (RAOC-T(Z) Methodology)

C W LCO 3.2.1 FQ(Z), as approximated by FQ(Z) and FQ (Z), shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. ------NOTE------A.1 Reduce THERMAL 15 minutes after each Required Action A.4 C POWER 1% RTP for FQ(Z) determination shall be completed C each 1% FQ(Z) exceeds whenever this Condition limit. is entered prior to increasing THERMAL AND POWER above the limit of Required Action A.1. A.2 Reduce Power Range 72 hours after each SR 3.2.1.2 is not C Neutron Flux – High trip F (Z) determination required to be Q setpoints 1% for performed if this each 1% that THERMAL Condition is entered POWER is limited below prior to THERMAL RTP by Required POWER exceeding Action A.1. 75% RTP after a

refueling. AND ------

A.3 Reduce Overpower ¨T trip FC(Z) not within limit. 72 hours after each Q setpoints 1% for each 1% C FQ(Z) determination that THERMAL POWER is limited below RTP by Required Action A.1.

AND

A.4 Perform SR 3.2.1.1 and Prior to increasing SR 3.2.1.2. THERMAL POWER above the limit of Required Action A.1

SEQUOYAH – UNIT 1 3.2.1-1 Amendment 334, FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

W B. FQ (Z) not within limits B.1.1 Implement a RAOC 4 hours operating space specified in the COLR that restores W FQ (Z) to within its limits.

AND

B.1.2 Perform SR 3.2.1.1 and 72 hours SR 3.2.1.2 if control rod motion is required to comply with the new operating space.

B.2.1 ------NOTE------Required Action B.2.4 shall be completed whenever Required Action B.2.1 is performed prior to increasing THERMAL POWER above the limit of Required Action B.2.1. ------

Limit allowable THERMAL 4 hours after each W POWER and AFD limits as FQ (Z) determination specified in the COLR.

AND

B.2.2 Limit Power Range 72 hours after each W Neutron Flux - High trip FQ (Z) determination setpoints 1% for each 1% that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND

SEQUOYAH – UNIT 1 3.2.1-2 Amendment 334, FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

CONDITION REQUIRED ACTION COMPLETION TIME

B.2.3 Limit Overpower ¨T trip 72 hours after each W setpoints 1% for each 1% FQ (Z) determination that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND

B.2.4 Perform SR 3.2.1.1 and SR 3.2.1.2. Prior to increasing THERMAL POWER above the limit of Required Action B.2.1

C. Required Action and C.1 Be in MODE 2. 6 hours associated Completion Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

C SR 3.2.1.1 Verify FQ(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP

AND

Once within 24 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL

SEQUOYAH – UNIT 1 3.2.1-3 Amendment 334, FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

POWER at which C FQ(Z) was last verified

AND

In accordance with the Surveillance Frequency Control Program

W SR 3.2.1.2 Verify FQ (Z) is within limit. Once after each refueling within 24 hours after THERMAL POWER exceeds 75% RTP

AND

Once within 24 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which W FQ (Z) was last verified

AND

In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 1 3.2.1-4 Amendment 334, Replace with new TS 3.2.2 on the following pages Fǻ+(X,Y) 3.2.2

3.2 POWER DISTRIBUTION LIMITS

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor Fǻ+(X,Y)

LCO 3.2.2 Fǻ+(X,Y) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. ------NOTE------A.1 Reduce allowable 2 hours Required Actions A.3 THERMAL POWER from and A.5 must be 573E\55+PXOWLSOLHG completed whenever times the Fǻ+ min margin. Condition A is entered. ------AND

Fǻ+ min margin < 0. A.2 Reduce Power Range 72 hours Neutron Flux – High trip VHWSRLQWVE\55+ multiplied times the Fǻ+ min margin.

AND

A.3 Perform SR 3.2.2.1. 24 hours

AND

A.4 Reduce Overtemperature 48 hours ¨T trip setpoint by TRH multiplied times the Fǻ+ min margin.

AND

SEQUOYAH – UNIT 1 3.2.2-1 Amendment 334 N F¨H 3.2.2

3.2 POWER DISTRIBUTION LIMITS

N 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F¨H)

N LCO 3.2.2 F¨H shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

N A. ------NOTE------A.1.1 Restore F¨H to within limit. 4 hours Required Actions A.2 and A.3 must be OR completed whenever Condition A is entered. A.1.2.1 Reduce THERMAL 4 hours ------POWER to < 50% RTP.

N F¨H not within limit. AND

A.1.2.2 Reduce Power Range 72 hours Neutron Flux – High trip setpoints to 55% RTP.

AND

A.2 Perform SR 3.2.2.1. 24 hours

AND

A.3 ------NOTE------THERMAL POWER does not have to be reduced to comply with this Required Action. ------

Perform SR 3.2.2.1. Prior to THERMAL POWER exceeding 50% RTP

AND

SEQUOYAH – UNIT 1 3.2.2-1 Amendment 334, N F¨H 3.2.2

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

Prior to THERMAL POWER exceeding 75% RTP

AND

24 hours after THERMAL POWER reaching 95% RTP

B. Required Action and B.1 Be in MODE 2. 6 hours associated Completion Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

N SR 3.2.2.1 Verify F¨H is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75% RTP

AND

In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 1 3.2.2-2 Amendment 334, QPTR 3.2.4

3.2 POWER DISTRIBUTION LIMITS

3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 7KH4375VKDOOEH

APPLICABILITY: 02'(ZLWK7+(50$/32:(5!573

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. QPTR not within limit. A.1 Reduce THERMAL 2 hours after each 32:(5IURP573IRU QPTR determination HDFKRI4375!

AND

A.2 Determine QPTR. Once per 12 hours

AND

A.3 Perform SR 3.2.1.1, 24 hours after SR 3.2.1.2, SR achieving equilibrium 3.2.1.3,and conditions from a SR 3.2.2.1., and SR 3.2.2.2. THERMAL POWER reduction per Required Action A.1

AND

2QFHSHUGD\V thereafter

AND

SEQUOYAH – UNIT 1 3.2.4-1 Amendment 334 QPTR 3.2.4

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

A.4 Reevaluate safety analyses Prior to increasing and confirm results remain THERMAL POWER valid for duration of operation above the limit of under this condition. Required Action A.1

AND

$ ------NOTES------1. Perform Required ActLRQ$RQO\DIWHU Required Action A.4 is completed.

2.Required Action A.6 shall be completed whenever 5HTXLUHG$FWLRQ$LV performed. ------

Normalize excore detectors to Prior to increasing restore QPTR to within limit. THERMAL POWER above the limit of Required Action A.1 AND

A.6 ------NOTE------Perform Required Action A.6 RQO\DIWHU5HTXLUHG$FWLRQ$ is completed. ------

Perform SR 3.2.1.1, Within 24 hours after SR 3.2.1.2, SR 3.2.1.3,and achieving equilibrium SR 3.2.2.1., and SR 3.2.2.2. conditions at RTP QRWWRH[FHHG hours after increasing THERMAL POWER above the limit of Required Action A.1

SEQUOYAH – UNIT 1 3.2.4-2 Amendment 334 QPTR 3.2.4

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. Required Action and %1 Reduce THERMAL 4 hours associated Completion POWER to RTP. Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.2.4.1 ------NOTES------ With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER 573WKHUHPDLQLQJWKUHHSRZHUUDQJH channels can be used for calculating QPTR.

SR 3.2.4.2 may be performed in lieu of this Surveillance. ------

Verify QPTR is within limit by calculation. In accordance with the Surveillance Frequency Control Program

SR 3.2.4.2 ------NOTE------Only required to be performed if input to QPTR from one or more Power RangeNeutron Flux channels are inoperable with7+(50$/ 32:(5! RTP. ______

SEQUOYAH - UNIT 1 3.2.4-3 Amendment 3 4 3I Verify QPTR is within limit using core power Once within 12 distribution measurement information.the movable hours incore detectors. AND

In accordance with the Surveillance Frequency Control Program

SEQUOYAH - UNIT 1 3.2.4-3 Amendment 3 4 3I RTS Instrumentation 3.3.1

SURVEILLANCE REQUIREMENTS ------NOTE------Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function. ------

SURVEILLANCE FREQUENCY

SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program

SR 3.3.1.2 ------NOTE------Not required to be performed until 12 hours after THERMAL POWER is t 573 ------

Compare results of calorimetric heat balance In accordance calculation to power range channel output. Adjust with the power range channel output if absolute difference is Surveillance ! Frequency Control Program

SR 3.3.1.3 ------NOTE------1RWUHTXLUHGWREHSHUIRUPHGXQWLOKRXUV after THERMAL POWER is t 573 ------

Compare results of the incore core power In accordance distribution detector measurements to Nuclear with the Instrumentation System (NIS) AFD. Adjust NIS Surveillance channel if absolute difference is t Frequency Control Program

SR 3.3.1.4 ------NOTE------This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service. ------

Perform TADOT. In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 1 3.3.1- Amendment 334 RTS Instrumentation 3.3.1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

SR Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program

SR 3.3.1.6 ------NOTE------Not required to be performed until 24 hours after 7+(50$/32:(5LV573 ------

Calibrate excore channels to agree with incore In accordance core power distribution detector with the measurements. Surveillance Frequency Control Program

SR ------NOTE------Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours after entry into MODE 3. ------

Perform COT. In accordance with the Surveillance Frequency Control Program

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6(482<$+±81,7 $PHQGPHQW Design Features 4.0

4.0 DESIGN FEATURES

4.1 Site Location

The Sequoyah Nuclear Plant is located on a site near the geographical center of Hamilton County, Tennessee, on a peninsula on the western shore of Chickamauga Lake at Tennessee River mile (TRM) 484.5. The Sequoyah site is approximately 7.5 miles northeast of the nearest city limit of Chattanooga, Tennessee, 14 miles west-northwest of Cleveland, Tennessee, and approximately 31 miles south-southwest of TVA's Watts Bar Nuclear Plant.

4.2 Reactor Core Optimized ZirloTM

4.2.1 Fuel Assemblies

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or M5 clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Sequoyah is authorized to place a limited number of lead test assemblies into the reactor as described in the Framatome-Cogema Fuels report BAW-2328, beginning with the Unit 1 Operating Cycle 12.

4.2.2 Control Rod Assemblies

------NOTE------Operation with 52 full length control rod assemblies (with no control rod assembly installed in core location H-08) is permitted during Cycle 24. ------

The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium, and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. 52

4.3 Fuel Storage (with no full length control rod assembly in core location H-08) 4.3.1 Criticality

4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

SEQUOYAH – UNIT 1 4.0-1 Amendment 334 348 Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: 1. LCO 2.1.1, "Reactor Core Safety Limits"; 2. 1. LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

3. 2. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)"; 4. LCO 3.1.4, "Rod Group Alignment Limits"; 5. 3. LCO 3.1.5, "Shutdown Bank Insertion Limits";

6. 4. LCO 3.1.6, "Control Bank Insertion Limits"; 7. LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2"; 8. 5. LCO 3.2.1, "Heat Flux Hot Channel Factor (F (X, Y, Z))"; Q N

9. 6. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (F¨+(X,Y))"; (RAOC-T(Z) 10. 7. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)"; Methodology) ;

11. 8. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," f1(¨I) limits for 2YHUWHPSHUDWXUH¨7DQGf2(¨I) limits for 2YHUSRZHU¨71RPLQDO Trip Setpoints; and 12. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from 13. 9. LCO 3.9.1, "Boron Concentration." Nucleate Boiling (DNB) Limits"; and

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:

Add 1. BAW-10180-A, Revision 1, "NEMO - Nodal Expansion Method Insert A Optimized," March 1993;

2. BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989;

3. BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for Westinghouse-Designed PWRs," June 1989;

SEQUOYAH – UNIT 1 5.6-2 Amendment 334 Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT (continued)

4. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 2001;

5. BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003;

6. BAW-10186P-A, Revision 2, "Extended Burnup Evaluation," June 2003;

7. EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003;

8. BAW-10241P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005;

9. BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996;

10. BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design," January 1996;

11. BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990; and

12. BAW-10231P-A, Revision 1, "COPERNIC Fuel Rod Design Computer Code" January 2004.

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 within 30 days of issuance for each reload cycle to the NRC.

SEQUOYAH – UNIT 1 5.6-3 Amendment 334 INSERT A

1. WCAP-8745-P-$³'HVLJQ%DVHVIRUWKH7KHUPDO2YHUSRZHU¨7DQG7KHUPDO 2YHUWHPSHUDWXUH¨77ULS)XQFWLRQV´6HSWHPEHU;

2. WCAP-9272-P-A, “Westinghouse Reload Safety Evaluation Methodology,” July 1985;

3. WCAP-10216-P-A, Revision 1A, “Relaxation of Constant Axial Offset Control – FQ Surveillance Technical Specification,” February 1994;

4. WCAP-10444-P-A, “Reference Core Report VANTAGE 5 Fuel Assembly,” September 1985;

5. WCAP-10444-P-A Addendum 2-A, “VANTAGE 5H Fuel Assembly,” February 1989;

6. WCAP-10965-P-A, “ANC: A Westinghouse Advanced Nodal Computer Code,” September 1986;

7. WCAP-10965-P-A, Addendum 2-A, Revision 0, “Qualification of the New Pin Power Recovery Methodology,” September 2010;

8. WCAP-11397-P-A, “Revised Thermal Design Procedure,” April 1989;

9. WCAP-12610-P-A, “VANTAGE+ Fuel Assembly Reference Core Report,” April 1995;

10. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, “Optimized ZIRLOTM,” July 2006;

11. WCAP-14565-P-A, “VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,” October 1999;

12. WCAP-14565-P-A, Addendum 1-A, Revision 0, “Addendum 1 to WCAP 14565-P-A Qualification of ABB-NV Critical Heat Flux Correlations with VIPRE-01 Code,” August 2004;

13. WCAP-14565-P-A, Addendum 2-P-A, Revision 0, “Addendum 2 to WCAP-14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications,” April 2008;

14. WCAP-15025-P-A, “Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,” April 1999;

15. WCAP-16045-P-A, “Qualification of the Two-Dimensional Transport Code PARAGON,” August 2004;

16. WCAP-16045-P-A, Addendum 1-A, “Qualification of the NEXUS Nuclear Data Methodology,” August 2007;

17. WCAP-16996-P-A, Revision 1, “Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),” November 2016; and

18. WCAP-17661-P-A, Revision 1, "Improved RAOC and CAOC FQ Surveillance Technical Specifications," February 2019. Enclosure 1

Evaluation of the Transition to Westinghouse RFA-2 Fuel

ATTACHMENT 2

Proposed TS Changes (Mark-Ups) for SQN Unit 2

CNL-20-014 - 11 -

s. Primary Coolant Outside Containment (Section 22.2, III.D.1.1)

Prior to exceeding 5 percent power level, TVA is required to complete the leak tests on Unit 2, and results are to be submitted within 30 days from the completion of the testing.

(17) Surveillance Interval Extension

The performance interval for the 36-month surveillance requirements in TS 4.3.2.1.3 shall be extended to May 18, 1996, to coincide with the Cycle 7 refueling outage. The extended interval shall not exceed a total of 50 months for the 36-month surveillances. Transition Core Peaking Penalties (18) Mixed Core DNBR Penalty When Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies: N (a) The HTP fuel assembliesTVA will obtain F ǻH shallNRC be approval maintained prior less to thanstartup 1.61; for any cycle's core that involves a (b) The RFA-2 fuelreduction assemblies in theDNBR departure limit shall from be reduced nucleate by: boiling ratio initial transition core penalty 1. 0.25% for thebelow WRB-2M that value critical stated heat flux in TVA'scorrelation submittal on Framatome fuel conversion dated 2. 0.50% for theApril ABB-NV 6, 1997. critical heat flux correlation

(19) Steam Generator Replacement Project

During the Unit 1 Cycle 12 refueling and steam generator replacement outage, lifts of heavy loads will be performed in accordance with Table 3.1 of NRC Safety Evaluation dated March 26, 2003.

(20) Control Room Air Conditioning System Maintenance

TVA commits to the use of a portable chiller package and air-handling unit to provide alternate cooling if both trains of the control room air condition system become inoperable during the maintenance activities to upgrade the compressors and controls or immediately enter Technical Specification 3.0.3.

(21) Mitigation Strategy License Condition

Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy and with the following elements: 1. Pre-defined coordinated fire response strategy and guidance 2. Assessment of mutual aid fire fighting assets 3. Designated staging areas for equipment and materials 4. Command and control 5. Training of response personnel

Renewed License No. DPR 79 September 28, 2015 For Information Only

This note is for NRC information TABLE OF CONTENTS Page

1.0 USE AND APPLICATION 1.1 Definitions ...... 1.1-1 1.2 Logical Connectors ...... 1.2-1 1.3 Completion Times ...... 1.3-1 1.4 Frequency ...... 1.4-1

2.0 SAFETY LIMITS (SLs) ...... 2.0-1 2.1 SLs 2.2 SL Violations

3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ...... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ...... 3.0-4

3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) ...... 3.1.1-1 3.1.2 Core Reactivity ...... 3.1.2-1 3.1.3 Moderator Temperature Coefficient (MTC) ...... 3.1.3-1 3.1.4 Rod Group Alignment Limits ...... 3.1.4-1 3.1.5 Shutdown Bank Insertion Limits ...... 3.1.5-1 3.1.6 Control Bank Insertion Limits ...... 3.1.6-1 3.1.7 Rod Position Indication ...... 3.1.7-1 3.1.8 PHYSICS TESTS Exceptions - MODE 2 ...... 3.1.8-1 (RAOC-T(Z) Methodology) 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (F Q(X,Y,Z)) ...... 3.2.1-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F¨+(X,Y)) ...... 3.2.2-1 3.2.3 AXIAL FLUX DIFFERENCE (AFD) ...... 3.2.3-1 3.2.4 QUADRANT POWER TILT RATIO (QPTR) ...... 3.2.4-1

N 3.3 INSTRUMENTATION F ȜH 3.3.1 Reactor Trip System (RTS) Instrumentation ...... 3.3.1-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation ...... 3.3.2-1 3.3.3 Post Accident Monitoring (PAM) Instrumentation ...... 3.3.3-1 3.3.4 Remote Shutdown Monitoring Instrumentation ...... 3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ...... 3.3.5-1 3.3.6 Containment Ventilation Isolation Instrumentation ...... 3.3.6-1 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation ...... 3.3.7-1 3.3.8 Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation ...... 3.3.8-1 3.3.9 Boron Dilution Monitoring Instrumentation (BDMI) ...... 3.3.9-1

SEQUOYAH - UNIT 2 i Amendment 327 SLs 2.0

2.0 SAFETY LIMITS (SLs)

2.1 SLs the COLR 2.1.1 Reactor Core SLs 9 specified In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits shown in Figure 2.1.1-1; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.132 for the BHTP correlation, 1.21 for the BWU-N correlation, and 1.21 for the BWCMV correlation. 1.14 for the WRB-2M 2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained )GHFUHDVLQJE\)SHU0:'078RIEXUQXSIRU < 5080 COPER1,&DSSOLFDWLRQVDQG)GHFUHDVLQJE\)SHU 10,000 MWD/MTU of burnup for TACO3 applications.

2.1.2 Reactor Coolant System Pressure SL

In MODES 1, 2, 3, 4, and 5, the RCS SUHVVXUHVKDOOEHPDLQWDLQHG 2735 psig.

2.2 SAFETY LIMIT VIOLATIONS

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

SEQUOYAH – UNIT 2 2.0-1 Amendment 327 SLs 2.0

Relocate to COLR

0.0 0.2 0.4 0.6 0.8 1.0 1.2

FRACTION OF RATED THERMAL POWER

Figure 2.1.1-1 (page 1 of 1) Reactor Core Safety Limit - Four Loops in Operation

SEQUOYAH – UNIT 2 2.0-2 Amendment 327 Rod Group Alignment Limits 3.1.4

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B.2.1.2 Initiate boration to restore 1 hour SDM to within limit.

AND

B.2.2 Reduce THERMAL 2 hours POWER to 75% RTP.

AND

B.2.3 Verify SDM is within the Once per limits specified in the 12 hours COLR.

AND and SR 3.2.1.2

B.2.4 Perform SR 3.2.1.1. 72 hours

AND

B.2.5 Perform SR 3.2.2.1. 72 hours AND

B.2.6 Re-evaluate safety analyses and confirm 5 days results remain valid for duration of operation under these conditions.

C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition B not met.

D. More than one rod not D.1.1 Verify SDM is within the 1 hour within alignment limit. limits specified in the COLR.

SEQUOYAH – UNIT 2 3.1.4-2 Amendment 327 Rod Position Indication 3.1.7

3.1 REACTIVITY CONTROL SYSTEMS

3.1.7 Rod Position Indication

LCO 3.1.7 The Rod Position Indication System and the Demand Position Indication System shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS ------NOTES------1. Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator.

2. LCO 3.0.4.a and b are not applicable for Required Actions A.2.1 and A.2.2 following a startup from a refueling outage, or following entry into MODE 5 of sufficient duration to safely repair an inoperable rod position indication. ------

CONDITION REQUIRED ACTION COMPLETION TIME

A. One rod position A.1 Verify the position of the Once per 12 hours indicator per bank rods with inoperable inoperable. position indicators indirectly by using movable incore detectors.

------NOTE------Required Actions A.2.1 and A.2.2 may only be applied to one inoperable rod position indicator. ------

A.2.1 Verify position of the 8 hours rod with inoperable position indicator indirectly AND by using movable incore detectors. core power distribution measurement information

SEQUOYAH – UNIT 2 3.1.7-1 Amendment 327 Rod Position Indication 3.1.7

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. More than one rod B.1 Place the control rods Immediately position indicator per under manual control. bank inoperable. AND

B.2 Monitor and record Reactor Once per 1 hour Coolant System Tavg.

AND

B.3 Verify the position of the Once per 12 hours rods with inoperable position indicators indirectly by using the movable incore detectors.

AND

B.4 Restore inoperable position 24 hours indicators to OPERABLE status such that a maximum of one rod position indicator per bank is inoperable.

C. One or more rods with C.1 Verify the position of the Immediately inoperable position rods with inoperable indicators have been position indicators indirectly moved in excess of 24 by using movable incore steps in one direction detectors. since the last determination of the OR rod’s position. C.2 Reduce THERMAL 8 hours POWER to < 50% RTP.

core power distribution measurement information

SEQUOYAH – UNIT 2 3.1.7-3 Amendment 327 Replace with new TS 3.2.1 on the FQ(X,Y,Z) 3.2.1 following pages

3.2 POWER DISTRIBUTION LIMITS

3.2.1 Heat Flux Hot Channel Factor (FQ(X,Y,Z))

LCO 3.2.1 FQ(X,Y,Z) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. ------NOTE------A.1 Reduce THERMAL 15 minutes after each C Required Action A.5 32:(5 1% RTP for FQ (X,Y,Z) C shall be completed each 1% FQ (X,Y,Z) determination whenever this Condition exceeds limit. is entered. ------AND

C FQ (X,Y,Z) not within the A.2 Reduce, by administrative 2 hours after each C steady state limit. means, $)'OLPLWVIRU FQ (X,Y,Z) C each 1% FQ (X,Y,Z) exceeds determination limit.

AND

A.3 5HGXFH2YHUSRZHU¨7WULS 48 hours after each C VHWSRLQWVIRUHDFK1% FQ (X,Y,Z) C FQ (X,Y,Z) exceeds limit. determination

AND

A.4 Reduce Power Range 72 hours after each C Neutron Flux – High trip FQ (X,Y,Z) VHWSRLQWVIRUHDFK determination C FQ (X,Y,Z) exceeds limit.

AND

A.5 Perform SR 3.2.1.1, Prior to increasing SR 3.2.1.2, and SR 3.2.1.3. THERMAL POWER above the limit of Required Action A.1

SEQUOYAH – UNIT 2 3.2.1-1 Amendment 327 FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

3.2 POWER DISTRIBUTION LIMITS

3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) (RAOC-T(Z) Methodology)

C W LCO 3.2.1 FQ(Z), as approximated by FQ(Z) and FQ (Z), shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

refueling. AND ------

A.3 Reduce Overpower ¨T trip FC(Z) not within limit. 72 hours after each Q setpoints 1% for each 1% C FQ(Z) determination that THERMAL POWER is limited below RTP by Required Action A.1.

AND

A.4 Perform SR 3.2.1.1 and Prior to increasing SR 3.2.1.2. THERMAL POWER above the limit of Required Action A.1

SEQUOYAH – UNIT 2 3.2.1-1 Amendment 327, FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

W B. FQ (Z) not within limits B.1.1 Implement a RAOC 4 hours operating space specified in the COLR that restores W FQ (Z) to within its limits.

AND

B.1.2 Perform SR 3.2.1.1 and 72 hours SR 3.2.1.2 if control rod motion is required to comply with the new operating space.

B.2.1 ------NOTE------Required Action B.2.4 shall be completed whenever Required Action B.2.1 is performed prior to increasing THERMAL POWER above the limit of Required Action B.2.1. ------

Limit allowable THERMAL 4 hours after each W POWER and AFD limits as FQ (Z) determination specified in the COLR.

AND

B.2.2 Limit Power Range 72 hours after each W Neutron Flux - High trip FQ (Z) determination setpoints 1% for each 1% that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND

SEQUOYAH – UNIT 3.2.1-2 Amendment 334, FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

CONDITION REQUIRED ACTION COMPLETION TIME

B.2.3 Limit Overpower ¨T trip 72 hours after each W setpoints 1% for each 1% FQ (Z) determination that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND

B.2.4 Perform SR 3.2.1.1 and SR 3.2.1.2. Prior to increasing THERMAL POWER above the limit of Required Action B.2.1

C. Required Action and C.1 Be in MODE 2. 6 hours associated Completion Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

C SR 3.2.1.1 Verify FQ(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP

AND

Once within 24 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL

SEQUOYAH – UNIT 3.2.1-3 Amendment 334, FQ(Z) (RAOC-T(Z) Methodology) 3.2.1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

POWER at which C FQ(Z) was last verified

AND

In accordance with the Surveillance Frequency Control Program

W SR 3.2.1.2 Verify FQ (Z) is within limit. Once after each refueling within 24 hours after THERMAL POWER exceeds 75% RTP

AND

Once within 24 hours after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which W FQ (Z) was last verified

AND

In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 3.2.1-4 Amendment 334, Replace with new TS 3.2.2 on Fǻ+(X,Y) the following pages 3.2.2

3.2 POWER DISTRIBUTION LIMITS

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor Fǻ+(X,Y)

LCO 3.2.2 Fǻ+(X,Y) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

Fǻ+ min margin < 0. A.2 Reduce Power Range 72 hours Neutron Flux – High trip VHWSRLQWVE\55+ multiplied times the Fǻ+ min margin.

AND

A.3 Perform SR 3.2.2.1. 24 hours

AND

A.4 Reduce Overtemperature 48 hours ¨T trip setpoint by TRH multiplied times the Fǻ+ min margin.

AND

SEQUOYAH – UNIT 2 3.2.2-1 Amendment 327 N F¨H 3.2.2

3.2 POWER DISTRIBUTION LIMITS

N 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F¨H)

N LCO 3.2.2 F¨H shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

N F¨H not within limit. AND

A.1.2.2 Reduce Power Range 72 hours Neutron Flux – High trip setpoints to 55% RTP.

AND

A.2 Perform SR 3.2.2.1. 24 hours

AND

A.3 ------NOTE------THERMAL POWER does not have to be reduced to comply with this Required Action. ------

Perform SR 3.2.2.1. Prior to THERMAL POWER exceeding 50% RTP

AND

SEQUOYAH – UNIT 2 3.2.2-1 Amendment 327, N F¨H 3.2.2

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

Prior to THERMAL POWER exceeding 75% RTP

AND

24 hours after THERMAL POWER reaching 95% RTP

B. Required Action and B.1 Be in MODE 2. 6 hours associated Completion Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

N SR 3.2.2.1 Verify F¨H is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75% RTP

AND

In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 2 3.2.2-2 Amendment 327, QPTR 3.2.4

3.2 POWER DISTRIBUTION LIMITS

3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 7KH4375VKDOOEH 1.02.

APPLICABILITY: MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. QPTR not within limit. A.1 Reduce THERMAL 2 hours after each 32:(5 3% from RTP for QPTR determination each 1% of QPTR > 1.02.

AND

A.2 Determine QPTR. Once per 12 hours

AND

A.3 Perform SR 3.2.1.1, 24 hours after SR 3.2.1.2, SR 3.2.1.3, achieving equilibrium SR 3.2.2.1, and SR 3.2.2.2. conditions from a THERMAL POWER reduction per and Required Action A.1

AND

Once per 7 days thereafter

AND

SEQUOYAH – UNIT 2 3.2.4-1 Amendment 327 QPTR 3.2.4

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

A.4 Reevaluate safety analyses Prior to increasing and confirm results remain THERMAL POWER valid for duration of operation above the limit of under this condition. Required Action A.1

AND

A.5 ------NOTES------1. Perform Required Action A.5 only after Required Action A.4 is completed.

2. Required Action A.6 shall be completed whenever Required Action A.5 is performed. ------

Normalize excore detectors to Prior to increasing restore QPTR to within limit. THERMAL POWER above the limit of Required Action A.1 AND

A.6 ------NOTE------Perform Required Action A.6 only after Required Action A.5 is completed. ------

Perform SR 3.2.1.1, Within 24 hours after SR 3.2.1.2, SR 3.2.1.3, achieving equilibrium SR 3.2.2.1, and SR 3.2.2.2. conditions at RTP not to exceed 48 hours after and increasing THERMAL POWER above the limit of Required Action A.1

SEQUOYAH – UNIT 2 3.2.4-2 Amendment 327 QPTR 3.2.4

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to 50% RTP. Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.2.4.1 ------NOTES------1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance. ------

Verify QPTR is within limit by calculation. In accordance with the Surveillance Frequency Control Program

SR 3.2.4.2 ------NOTE------Only required to be performed if input to QPTR from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP. ------

Verify QPTR is within limit using the movable incore Once within 12 detectors. hours

AND core power distribution measurement In accordance information with the Surveillance Frequency Control Program SEQUOYAH – UNIT 2 3.2.4-3 Amendment 327 336 RTS Instrumentation 3.3.1

SURVEILLANCE REQUIREMENTS ------NOTE------Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function. ------

SURVEILLANCE FREQUENCY

SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program

SR 3.3.1.2 ------NOTE------Not required to be performed until 12 hours after THERMAL POWER is t 15% RTP. ------

SR 3.3.1.3 ------NOTE------Not required to be performed until 96 hours after THERMAL POWER is t 15% RTP. core power distribution ------

Compare results of the incore detector In accordance measurements to Nuclear Instrumentation System with the (NIS) AFD. Adjust NIS channel if absolute Surveillance difference is t 3%. Frequency Control Program

SR 3.3.1.4 ------NOTE------This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service. ------

Perform TADOT. In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 2 3.3.1-9 Amendment 327 RTS Instrumentation 3.3.1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

SR 3.3.1.5 Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program

SR 3.3.1.6 ------NOTE------Not required to be performed until 24 hours after 7+(50$/32:(5LV 50% RTP. ------core power distribution Calibrate excore channels to agree with incore In accordance detector measurements. with the Surveillance Frequency Control Program

SR 3.3.1.7 ------NOTE------Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours after entry into MODE 3. ------

Perform COT. In accordance with the Surveillance Frequency Control Program

SEQUOYAH – UNIT 2 3.3.1-10 Amendment 327 RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 7 of 9) Reactor Trip System Instrumentation

Note 2YHUWHPSHUDWXUH¨7

7KH2YHUWHPSHUDWXUH¨7)XQction Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 1.9% RI¨7VSDQ

§1W S · § 1W S · ½ 4 1 c c 'T ¨ ¸ 0 ® 'd KKT 21 ¨ ¸>@ 3 1 ' IfPPKTT )()( ¾ ©1W 5 S ¹ ¯ ©1W 2 S ¹ ¿

Where: ¨T is measured RCS ¨T,°F. ¨T0 LVWKHLQGLFDWHG¨7DW573) S is the Laplace transform operator, sec-1. T is the measured RCS average temperature,°F. ' T is the nominal Tavg DW573 578.2°F.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, = 2235 psig **

K1 1.15 K2 0.011/°F K3 = 0.00055/psig W 1 33 sec ** W 2 4 sec W 4 5 sec W 5 3 sec

and f1 ('I) is a function such that:

(i) for qt - qb between QTNL* and QTPL* f1 ('I) = 0

* (ii) for each percent that the magnitude of (qt - qb) exceeds QTNL , the 'T nominal trip setpoint shall be automatically reduced by QTNS* of its value at RATED THERMAL POWER.

* (iii) for each percent that the magnitude of (qt - qb) exceeds QTPL , the 'T nominal trip setpoint shall be automatically reduced by QTPS* of its value at RATED THERMAL POWER.

Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

*QTNL, QTPL, QTNS, and QTPS are specified in the COLR.

The values denoted with ** are specified in the COLR.

SEQUOYAH – UNIT 2 3.3.1-20 Amendment 327 RTS Instrumentation 3.3.1

Table 3.3.1-1 (page 8 of 9) Reactor Trip System Instrumentation

Note 2YHUSRZHU¨7

7KH2YHUSRZHU¨7)XQFWLRQ$OORZDEOH9DOXHVKDOOQRWH[FHHGWKHIROORZLQJNominal Trip Setpoint by more than 1.7% RI¨7VSDQ

§1W S · § W S · ½ 'T 4 'd KKT ¨ 3 ¸ ' IfTTKT )(" ¨ ¸ 0 ® 54 ¨ ¸ 6 2 ¾ ©1W 5 S ¹ ¯ ©1W 3S ¹ ¿

Where: ¨T is measured RCS ¨T,°F. ¨T0 LVWKHLQGLFDWHG¨7DW573) S is the Laplace transform operator, sec-1. T is the measured RCS average temperature,°F. " T is the nominal Tavg DW573 578.2°F.

** " K4 1.087 K5 0.02/°F for increasing Tavg K6 0.0011/°F when T > T " 0/°F for decreasing Tavg 0)ZKHQ7 T ** W 3 10 sec W 4 5 sec W 5 3 sec

and f2 ('I) is a function such that:

(i) for qt - qb between QPNL* and QPPL* f2 ('I) = 0

* (ii) for each percent that the magnitude of (qt - qb) exceeds QPNL , the 'T nominal trip setpoint shall be automatically reduced by QPNS* of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (qt - qb) exceeds QPPL*, the ǻT nominal trip setpoint shall be automatically reduced by QPPS* of its value at RATED THERMAL POWER.

Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

*QPNL, QPPL, QPNS, and QPPS are specified in the COLR.

The values denoted with ** are specified in the COLR.

SEQUOYAH – UNIT 2 3.3.1-21 Amendment 327 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits greater than or equal to the limit specified

in the COLR LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is 2220 psia; less than or equal to the limit b. RCS average temperature is 583°F; and specified in the

COLR c. 5&6WRWDOIORZUDWH 378,400 gpm. 360,000

APPLICABILITY: MODE 1. and greater than or equal to the limit specified in the COLR ------NOTE------Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute;

b. THERMAL POWER step > 10% RTP;

c. PHYSICS TESTS; or

d. Performance of SR 3.1.3.2. ------

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more RCS DNB A.1 Restore RCS DNB 2 hours parameters not within parameter(s) to within limit. limits.

B. Required Action and B.1 Be in MODE 2. 6 hours associated Completion Time not met.

SEQUOYAH – UNIT 2 3.4.1-1 Amendment 327 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.1.1 Verify pressurizer pressure is SVLD. In accordance with the Surveillance greater than or equal to the limit Frequency specified in the COLR Control Program

SR 3.4.1.2 Verify RCS average temperature is ). In accordance with the less than or equal to the limit Surveillance specified in the COLR 360,000 Frequency Control Program

SR 3.4.1.3 9HULI\5&6WRWDOIORZUDWHLV 378,400 gpm. In accordance with the Surveillance and greater than or equal to the limit Frequency specified in the COLR Control Program

SR 3.4.1.4 Verify by measurement that RCS total flow rate is In accordance 378,400 gpm. with the 360,000 Surveillance Frequency and greater than or equal to the limit specified in the COLR Control Program

SEQUOYAH – UNIT 2 3.4.1-2 Amendment 327 Design Features 4.0

4.0 DESIGN FEATURES

4.1 Site Location

4.2 Reactor Core Optimized ZirloTM 4.2.1 Fuel Assemblies

4.2.2 Control Rod Assemblies

------NOTE------Operation with 52 full length control rod assemblies (with no control rod assembly installed in core location H-08) is permitted during Cycles 24. ------52 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium, and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.

4.3 Fuel Storage (with no full length control rod assembly 4.3.1 Criticality in core location H-08)

4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

SEQUOYAH – UNIT 2 4.0-1 Amendment 327, 342 Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT

3. 2. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)"; 4. LCO 3.1.4, "Rod Group Alignment Limits"; 5. 3. LCO 3.1.5, "Shutdown Bank Insertion Limits";

6. 4. LCO 3.1.6, "Control Bank Insertion Limits"; 7. LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2"; 8. 5. LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(X, Y, Z))"; N (RAOC T(Z) Methodology) 9. 6. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (F¨+(X,Y))";

10. 7. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)"; ; 11. 8. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," f1(¨I) limits for 2YHUWHPSHUDWXUH¨7DQGf2(¨I) limits for 2YHUSRZHU¨71RPLQDO Trip Setpoints; and 12. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from 13. 9. LCO 3.9.1, "Boron Concentration." Nucleate Boiling (DNB) Limits"; and

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:

Add 1. BAW-10180-A, Revision 1, "NEMO - Nodal Expansion Method Insert A Optimized," March 1993;

2. BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989;

3. BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for Westinghouse-Designed PWRs," June 1989;

SEQUOYAH – UNIT 2 5.6-2 Amendment 327 Reporting Requirements 5.6