Text

NON-PROPRIETARY - Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 356 and 349 Regarding the Transition to Westinghouse Robust Fuel Assembly-2 (RFA-2) Fuel
	ML21245A267
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Site:	Sequoyah
Issue date:	10/27/2021
From:	Perry Buckberg; Plant Licensing Branch II
To:	Jim Barstow; Tennessee Valley Authority
	Buckberg P
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OFFICIAL USE ONLY PROPRIETARY INFORMATION October 26, 2021 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 356 AND 349 REGARDING THE TRANSITION TO WESTINGHOUSE ROBUST FUEL ASSEMBLY-2 (RFA-2) FUEL (EPID L-2020-LLA-0216)

Dear Mr. Barstow:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 356 to Renewed Facility Operating License (RFOL) No. DPR-77, and Amendment No. 349 to Renewed Facility Operating License No. DPR-79, for the Sequoyah Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated September 23, 2020, as supplemented by letters dated May 5, 2021, and August 13, 2021.

The amendments revise the RFOLs and Technical Specifications (TSs) to allow the use of Westinghouse Robust Fuel Assembly-2 (RFA-2) fuel with Optimized ZIRLOTM cladding. The amendments also revise the TS 5.6.3, Core Operating Limits Report, to replace the loss-of-coolant accident analysis evaluation model references with the FULL SPECTRUMTM Loss-of-Coolant Accident Evaluation Model. Finally, the amendments revise the TSs to permit the use of 52 full-length control rods with no full-length control rod assembly in core location H-08.

Enclosure 3 to this letter contains proprietary information. When separated from Enclosure 3, this document is DECONTROLLED.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION J. Barstow A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Perry H. Buckberg, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.: 50-327 and 50-328

Enclosures:

1. Amendment No. 356 to DPR-77

2. Amendment No. 349 to DPR-79

3. Safety Evaluation (Proprietary)

4. Safety Evaluation (Non-Proprietary) cc w/o Enclosure 3: Listserv OFFICIAL USE ONLY PROPRIETARY INFORMATION

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 356 Renewed License No. DPR-77

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated September 23, 2020, as supplemented by letters dated May 5, 2021, and August 13, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)

Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 356 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented when the Westinghouse RFA-2 fuel is loaded into the core during the Cycle 26 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 26, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 356 SEQUOYAH NUCLEAR PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace pages 3 and 11 of the Renewed Facility Operating License with the attached pages 3 and 11.

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 2.0-1 2.0-1 2.0-2 3.1.4-2 3.1.4-2 3.1.7-1 3.1.7-1 3.1.7-3 3.1.7-3 3.2.1-1 3.2.1-1 3.2.1-2 3.2.1-2 3.2.1-3 3.2.1-3 3.2.1-4 3.2.1-4 3.2.1-5 3.2.2-1 3.2.2-1 3.2.2-2 3.2.2-2 3.2.2-3 3.2.2-4 3.2.4-1 3.2.4-1 3.2.4-2 3.2.4-2 3.2.4-3 3.2.4-3 3.3.1-9 3.3.1-9 3.3.1-10 3.3.1-10 3.3.1-20 3.3.1-20 3.3.1-21 3.3.1-21 3.4.1-1 3.4.1-1 3.4.1-2 3.4.1-2 4.0-1 4.0-1 5.6-2 5.6-2 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5 Enclosure 2

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.

(25) Transition Core Peaking Penalties When Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies:

(a) The HTP fuel assemblies FNH shall be maintained 5% less than the RFA-2 fuel FNH value.

(b) The RFA-2 fuel assemblies margin to the DNBR limit shall be adjusted by subtracting the following:

1. 0.25% for the WRB-2M critical heat flux correlation

2. 0.50% for the ABB-NV critical heat flux correlation Amendment 356 Renewed License No. DPR 77

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)

LCO 3.2.1 FQ (Z), as approximated by FCQ (Z) and FW Q (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 POWER 1% RTP for FCQ (Z) determination shall be completed each 1% FCQ (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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each SR 3.2.1.2 is not Neutron Flux - High trip FCQ (Z) determination required to be performed setpoints 1% for if this Condition is each 1% that THERMAL entered prior to POWER is limited below THERMAL POWER RTP by Required exceeding 75% RTP Action A.1.

after a refueling.

AND FCQ (Z) not within limit. A.3 Reduce Overpower T trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each setpoints 1% for each 1% FCQ (Z) determination that THERMAL POWER is limited below RTP by Required Action A.1.

AND Prior to increasing A.4 Perform SR 3.2.1.1 and THERMAL POWER SR 3.2.1.2. above the limit of Required Action A.1 SEQUOYAH - UNIT 1 3.2.1-1 Amendment 334, 356

FQ(Z) (RAOC-T(Z) Methodology) 3.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. FW Q (Z) not within limits B.1.1 Implement a RAOC 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months operating space specified in the COLR that restores FWQ (Z) to within its limits.

AND B.1.2 Perform SR 3.2.1.1 and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.2.1.2 if control rod motion is required to comply with the new operating space.

OR 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after each POWER and AFD limits as FW Q (Z) determination specified in the COLR.

AND B.2.2 Limit Power Range Neutron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each Flux - High trip setpoints FW Q (Z) determination 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, 356

FQ(Z) (RAOC-T(Z) Methodology) 3.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.2.3 Limit Overpower T trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each setpoints 1% for each 1% FW Q (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 Prior to increasing SR 3.2.1.2. THERMAL POWER above the limit of Required Action B.2.1 C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FCQ (Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL SEQUOYAH - UNIT 1 3.2.1-3 Amendment 334, 356

FQ(Z) (RAOC-T(Z) Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY POWER at which FCQ (Z) was last verified AND In accordance with the Surveillance Frequency Control Program SR 3.2.1.2 Verify FW Q (Z) is within limit.

Once after each refueling within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER exceeds 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FWQ (Z) was last verified AND In accordance with the Surveillance Frequency Control Program SEQUOYAH - UNIT 1 3.2.1-4 Amendment 334, 356

FNH 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN H )

LCO 3.2.2 FNH shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------ NOTE ----------- A.1.1 Restore FNH to within limit. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Actions A.2 and A.3 must be OR completed whenever Condition A is entered. A.1.2.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

POWER to < 50% RTP.

FNH not within limit. AND A.1.2.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Neutron Flux - High trip setpoints to 55% RTP.

AND A.2 Perform SR 3.2.2.1. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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, 356

FNH 3.2.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Prior to THERMAL POWER exceeding 75% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER reaching 95% RTP B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FNH 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, 356

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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 15% RTP.

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

> 2%. Frequency Control Program SR 3.3.1.3 -----------------------------NOTE-------------------------------

Not required to be performed until 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months after THERMAL POWER is 15% RTP.

Compare results of the core power distribution In accordance measurements to Nuclear Instrumentation System with the (NIS) AFD. Adjust NIS channel if absolute Surveillance difference is 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 1 3.3.1-9 Amendment 334, 356

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 9)

Reactor Trip System Instrumentation Note 1: Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 1.9% of T span.

1+ 1+

[ ] + ( ) ()

1+ 1+

Where: T is measured RCS T,°F.

T0 is the indicated T at RTP,°F.

S is the Laplace transform operator, sec-1.

T is the measured RCS average temperature,°F.

T is the nominal Tavg at RTP, **°F.

P is the measured pressurizer pressure, psig P is the nominal RCS operating pressure, = ** psig K ** K **/°F K = **/psig

- sec ** sec

- sec ** sec and f1 (I) is a function such that:

(i) for qt - qb between QTNL* and QTPL* f 1 (I) = 0 (ii) for each percent that the magnitude of (q t - 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 (q t - q b ) 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 1 3.3.1-20 Amendment 334, 356

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 9)

Reactor Trip System Instrumentation Note 2: Overpower T The Overpower T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 1.7% of T span.

1+

" ()

1+ 1+

Where: T is measured RCS T,°F.

T0 is the indicated T at RTP,°F.

S is the Laplace transform operator, sec-1.

T is the measured RCS average temperature,°F.

T" is the nominal Tavg at RTP, **°F.

K ** K **/°F for increasing Tavg K **/°F when T > T"

- /°F for decreasing Tavg **/°F when T T"

- sec ** sec ** sec and f2 (I) is a function such that:

(i) for qt - qb between QPNL* and QPPL* f 2 (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 T 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 1 3.3.1-21 Amendment 334, 356

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 349 Renewed License No. DPR-79

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated September 23, 2020, as supplemented by letters dated May 5, 2021, and August 13, 2021, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-79 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 349 are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented when the Westinghouse RFA-2 fuel is loaded into the core during the Cycle 26 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 26, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 349 SEQUOYAH NUCLEAR PLANT, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace pages 3 and 11 of the Renewed Facility Operating License with the attached pages 3, 11, and 12.

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 2.0-1 2.0-1 2.0-2 3.1.4-2 3.1.4-2 3.1.7-1 3.1.7-1 3.1.7-3 3.1.7-3 3.2.1-1 3.2.1-1 3.2.1-2 3.2.1-2 3.2.1-3 3.2.1-3 3.2.1-4 3.2.1-4 3.2.1-5 3.2.2-1 3.2.2-1 3.2.2-2 3.2.2-2 3.2.2-3 3.2.2-4 3.2.4-1 3.2.4-1 3.2.4-2 3.2.4-2 3.2.4-3 3.2.4-3 3.3.1-9 3.3.1-9 3.3.1-10 3.3.1-10 3.3.1-20 3.3.1-20 3.3.1-21 3.3.1-21 3.4.1-1 3.4.1-1 3.4.1-2 3.4.1-2 4.0-1 4.0-1 5.6-2 5.6-2 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5

s. Primary Coolant Outside Containment (Section 22.2, I11.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.

(18) Transition Core Peaking Penalties When Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies:

(a) The HTP fuel assemblies Ff8 shall be maintained 5% less than the RFA-2 fuel FH value.

(b) The RFA-2 fuel assemblies margin to the DNBR limit shall be adjusted by subtracting the following:

1. 0.25% for the WRB-2M critical heat flux correlation

2. 0.50% for the ABB-NV 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:

Amendment 349 Renewed License No. DPR 79

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)

LCO 3.2.1 FQ(Z), as approximated by FCQ (Z) and FW Q (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 POWER 1% RTP for FCQ (Z) determination shall be completed each 1% FCQ (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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each SR 3.2.1.2 is not Neutron Flux - High trip FCQ (Z) determination required to be performed setpoints 1% for if this Condition is each 1% that THERMAL entered prior to POWER is limited below THERMAL POWER RTP by Required exceeding 75% RTP Action A.1.

after a refueling.

AND FCQ (Z) not within limit. A.3 Reduce Overpower T trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each setpoints 1% for each 1% FCQ (Z) determination that THERMAL POWER is limited below RTP by Required Action A.1.

AND Prior to increasing A.4 Perform SR 3.2.1.1 and THERMAL POWER SR 3.2.1.2. above the limit of Required Action A.1 SEQUOYAH - UNIT 2 3.2.1-1 Amendment 327, 349

FQ(Z) (RAOC-T(Z) Methodology) 3.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. FW Q (Z) not within limits B.1.1 Implement a RAOC 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months operating space specified in the COLR that restores FWQ (Z) to within its limits.

AND B.1.2 Perform SR 3.2.1.1 and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.2.1.2 if control rod motion is required to comply with the new operating space.

OR 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after each POWER and AFD limits as FW Q (Z) determination specified in the COLR.

AND B.2.2 Limit Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each Neutron Flux - High trip FW Q (Z) determination setpoints 1% for each 1%

that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND SEQUOYAH - UNIT 2 3.2.1-2 Amendment 327, 349

FQ(Z) (RAOC-T(Z) Methodology) 3.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.2.3 Limit Overpower T trip 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each setpoints 1% for each 1% FW Q (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 Prior to increasing SR 3.2.1.2. THERMAL POWER above the limit of Required Action B.2.1 C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FCQ (Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL SEQUOYAH - UNIT 2 3.2.1-3 Amendment 327, 349

FQ(Z) (RAOC-T(Z) Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY POWER at which FCQ (Z) was last verified AND In accordance with the Surveillance Frequency Control Program SR 3.2.1.2 Verify FW Q (Z) is within limit.

Once after each refueling within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER exceeds 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FWQ (Z) was last verified AND In accordance with the Surveillance Frequency Control Program SEQUOYAH - UNIT 2 3.2.1-4 Amendment 327, 349

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 9)

Reactor Trip System Instrumentation Note 1: Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 1.9% of T span.

1+ 1+

[ ] + ( ) ()

1+ 1+

Where: T is measured RCS T,°F.

T0 is the indicated T at RTP,°F.

S is the Laplace transform operator, sec-1.

T is the measured RCS average temperature,°F.

T' is the nominal Tavg at RTP, **°F.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, = ** psig K ** K **/°F K = **/psig

- sec ** sec

- sec ** sec and f1 (I) is a function such that:

(i) for qt - qb between QTNL* and QTPL* f 1 (I) = 0 (ii) for each percent that the magnitude of (q t - 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 (q t - 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, 349

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 9)

Reactor Trip System Instrumentation Note 2: Overpower T The Overpower T Function Allowable Value shall not exceed the following Nominal Trip Setpoint by more than 1.7% of T span.

1+

" ()

1+ 1+

Where: T is measured RCS T,°F.

T0 is the indicated T at RTP,°F.

S is the Laplace transform operator, sec-1.

T is the measured RCS average temperature,°F.

T" is the nominal Tavg at RTP, **°F.

K ** K **/°F for increasing Tavg K **/°F when T > T"

- /°F for decreasing Tavg **/°F when T T"

- sec ** sec ** sec and f2 (I) is a function such that:

(i) for qt - qb between QPNL* and QPPL* f 2 (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 T 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, 349

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (continued)

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

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.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits;

2. LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

System"; and

3. LCO 3.5.2, ECCS - Operating.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves";

2. Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation; and

3. Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

c. The PTLR shall be provided to the NRC within 30 days of issuance for each reactor vessel fluence period and for any revision or supplement thereto.

SEQUOYAH - UNIT 2 5.6-4 Amendment 327, 349

OFFICIAL USE ONLY PROPRIETARY INFORMATION

- 104 -

ENCLOSURE 4 (NON-PROPRIETARY)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 356 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 349 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 Proprietary information has been redacted from this document pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations.

Redacted information is identified by blank space enclosed within ((double brackets)).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 356 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 349 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

By application dated September 23, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20267A617 and proprietary attachment ML20267A618), as supplemented by letters dated May 5, 2021, and August 13, 2021 (ADAMS Accession Nos.

ML21125A347 and ML21225A249, respectively), the Tennessee Valley Authority (the licensee) submitted a license amendment request (LAR) for Sequoyah Nuclear Plant (SQN), Units 1 and 2. The requested changes would modify the Technical Specifications (TSs) and modify license conditions to allow for the transition to Westinghouse Robust Fuel Assembly-2 (RFA-2) fuel with Optimized ZIRLOTM cladding. Further, the proposed amendments would revise TS 5.6.3, Core Operating Limits Report [COLR], to replace the loss-of-coolant accident analysis evaluation model references with the FULL SPECTRUMTM Loss-of-Coolant Accident (FSLOCA) evaluation model. Finally, the proposed amendments would revise the TSs to permit the use of 52 full-length control rods with no full-length control rod assembly in core location H-08.

From February 4, 2021, through March 12, 2021, the U.S. Nuclear Regulatory Commission (NRC, the Commission) staff conducted a regulatory audit to support its review of the amendment request, as discussed in the staffs audit plan dated February 4, 2021 (ADAMS Accession No. ML21036A109), and audit summary dated April 8, 2021 (ADAMS Accession No. ML21090A228).

The supplements dated May 5, 2021, and August 13, 2021, provided addition information that clarified and corrected the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs initial proposed no significant hazards consideration determination as published in the Federal Register on December 1, 2020, (85 FR 77265).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

2.0 REGULATORY EVALUATION

2.1 System Description 2.1.1 Reactor Fuel The SQN Units 1 and 2 reactor cores consists of fuel rods that are constructed of cylindrical Zircaloy tubes containing uranium oxide (UO2) fuel pellets. The fuel rods are grouped together in bundles, known as fuel assemblies, that are arranged in a pattern approximating a right circular cylinder. Specifically, SQN Units 1 and 2 reactor cores are loaded with the Advanced W17 High Thermal Performance (HTP) fuel supplied by AREVA NP, Inc. The Advanced W17 HTP fuel assembly is a 17 x 17, standard lattice fuel assembly that is designed for use in Westinghouse-designed reactors. The Advanced W17 HTP fuel assemblies consist of 264 fuel rods, 24 guide tubes, and one instrument tube in a 17 x 17 square array. The guide tubes are annealed Zircaloy-4 and provide guidance for control rod insertion. The fuel assemblies contain 7 Zircaloy-4 spacer grid assemblies. The use of Zircaloy-4 in the Advanced W17 HTP fuel assemblies is intended to provide the following benefits: 1) low neutron absorption cross section, 2) high strength to resist deformation due to differential pressures and mechanical interaction between fuel and clad, 3) high corrosion resistance, and 4) high reliability.

The Westinghouse 17x17 Robust Fuel Assembly-2 (RFA-2) fuel design has been in operation since 2003. The LAR describes the following as features of the RFA-2 design:

Integral fuel burnable absorbers (IFBAs)

Robust protective grid (RPG)

Standardized debris filter bottom nozzle (SDFBN)

High-burnup bottom grid Debris mitigating long fuel rod bottom end plugs Wet annular burnable absorbers (WABA)

Removable top nozzle (RTN)

Three ZIRLO intermediate flow mixer (IFM) grids Six ZIRLO RFA-2 structural mid-grids Reduced rod bow (RRB) INCONEL top grid Optimized ZIRLO high-performance fuel cladding with a coated cladding feature Thicker-walled guide thimble and instrumentation tubes to improve fuel assembly Stiffness and to address incomplete rod insertion (IRI) considerations.

The Optimized ZIRLO' fuel cladding is different from standard ZIRLO in two respects: (1) the tin content is lower and (2) the microstructure is different. Optimized ZIRLO' is intended to provide improved corrosion resistance when compared to prior fuel cladding materials.

2.1.2 Reactivity Control System The SQN Units 1 and 2 reactor cores normally contain 53 full-length control rod assemblies divided into four control banks (Control Banks A, B, C, and D) and four shutdown banks (Shutdown Banks A, B, C, and D). Control Bank D is used for reactivity control during normal, at-power operation, while the control banks are normally used for reactor startup and shutdown.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The shutdown banks are intended to provide additional negative reactivity to meet shutdown margin (SDM) requirements.

Each control rod is moved by a full-length control rod drive mechanism (CRDM) consisting of a stationary gripper, movable gripper, and a lift pole. Should both sets of grippers be de-energized simultaneously, the corresponding control rod would drop into the core. The primary function of the CRDMs is to insert, withdraw, or hold control rods within the core to control average core temperature and to shut down the reactor. Control rod H-08 is part of Control Bank D and is located in the center of the core.

2.1.3 Reactor Trip System Instrumentation For SQN Units 1 and 2, the Overpower Delta Temperature (OPT) and Overtemperature Delta Temperature (OTT) trip functions are used to derive the limiting safety system settings for power-related reactor trips. There are four OPT channels and four OTT channels. These trips are activated on a two-out-of-four logic. The LCO requires all four channels of the OPT and four channels of the OTT trip functions to be OPERABLE.

The OTT trip function is provided to ensure that the design limit nucleate boiling ratio (DNBR) is met. This trip function also limits the range over which the OPT trip function must provide protection.

The OPT trip function ensures that protection is provided to ensure the integrity of the fuel (i.e.,

no fuel pellet melting and less than 1 percent cladding strain) under all possible overpower conditions. This trip function also limits the required range of the OTT trip function and provides a backup to the Power Range Neutron Flux - High Setpoint trip. The OPT trip function ensures that the allowable heat generation rate of the fuel, in kilowatts per foot (kW/ft),

is not exceeded.

2.1.4 FSLOCA Methodology As described in 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 (ADAMS Accession No. ML17277A131), the purpose of the FSLOCA evaluation model is to build on the Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology by extending the applicability of the WCOBRA/TRAC Code to include the treatment of small break LOCA (SBLOCA) and intermediate break LOCA (IBLOCA) scenarios. The term Full Spectrum specifies that the new model is intended to resolve the full spectrum of LOCA scenarios that result from a postulated break in the cold leg of a pressurized-water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA methodology include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine rupture with a break flow area equal to 2 times the pipe area.

2.2 Licensees Proposed Changes In the LAR, the licensee requested the following changes to the SQN, Units 1 and 2, operating licenses and respective TSs (additions in bold, deletions in strikeout):

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Unit 1 Operating License (25) Mixed Core DNBR Penalty TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6, 1997.

Transition Core Peaking Penalties When Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies:

(a) The HTP fuel assemblies FNH shall be maintained 5% less than the RFA-2 fuel FNH value.

(b) The RFA-2 fuel assemblies margin to the DNBR limit shall be adjusted by subtracting the following:

1. 0.25% for the WRB-2M critical heat flux correlation

2. 0.50% for the ABB-NV critical heat flux correlation Unit 2 Operating License (18) Mixed Core DNBR Penalty TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6, 1997.

Transition Core Peaking Penalties When Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies:

(a) The HTP fuel assemblies FNH shall be maintained 5% less than the RFA-2 fuel FNH value.

(b) The RFA-2 fuel assemblies margin to the DNBR limit shall be adjusted by subtracting the following:

1. 0.25% for the WRB-2M critical heat flux correlation

2. 0.50% for the ABB-NV critical heat flux correlation TS 2.1.1, Reactor Core SLs [Safety Limits]

2.1.1 Reactor Core SLs 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 specified in Figure 2.1.1 1 the COLR; and the following SLs shall not be exceeded:

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 1.14 for the WRB-2M correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained 4901<5080°F, decreasing by 13.79°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications.

In addition, the licensee proposed to delete Figure 2.1.1-1, Reactor Core Safety Limit - Four Loops in Operation, from the TSs and relocate it to the COLR.

TS 3.1.4, Rod Group Alignment Limits Proposed changes to the Required Action for Condition B (One rod not within alignment limits).

CONDITION REQUIRED ACTION COMPLETION TIME B.2.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limit.

AND B.2.2 Reduce THERMAL to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 75% RTP.

AND B.2.3 Verify SDM is within the Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limits specified in the COLR.

AND B.2.4 Perform SR 3.2.1.1 and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.2.1.2.

AND B.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.6 Re-evaluate safety 5 days analyses and confirm results remain valid for OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION duration of operation under these conditions.

TS 3.1.7, Rod Position Indication Changes to Condition A CONDITION REQUIRED ACTION COMPLETION TIME A. One rod position A.1. Verify the position of the Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months indicator per bank rods with inoperable inoperable. position indicators indirectly by using movable incore detectorscore power distribution measurement information.

OR

NOTE---------------

Required Actions A.2.1 and A.2.2 may only be applied to one inoperable rod position indicator. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

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

Changes to Conditions B and C CONDITION REQUIRED ACTION COMPLETION TIME B. More than one rod B.1 Place the control rods Immediately position indicator under manual control.

per bank inoperable. AND OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION B.2 Monitor and record Once per 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Reactor Coolant System Tavg.

AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.3 Verify the position of the rods with inoperable position indicators indirectly by using the movable incore detectorscore power distribution measurement 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months information.

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

C. One or more rods C.1 Verify the position of the Immediately with inoperable rods with inoperable position indicators position indicators have been moved indirectly by using in excess of 24 movable incore steps in one detectorscore power direction since the distribution last determination measurement of the rods information.

position.

OR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C.2 Reduce THERMAL POWER to < 50% RTP.

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

The licensee proposed deleting TS 3.2.1 and replacing it with the following:

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

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION LCO 3.2.1 FQ(Z), as approximated by FQC(Z) and FQW(Z), shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Reduce THERMAL 15 minutes after each Required Action POWER 1% RTP for FQC(Z) determination A.4 shall be each 1% FQC(Z) exceeds completed limit.

whenever this Condition is AND entered prior to increasing A.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each THERMAL Neutron Flux - High FQC(Z) determination POWER above the trip setpoints 1% for limit of Required each 1% that THERMAL Action A.1. POWER is limited SR 3.2.1.2 is not below RTP by Required required to be Action A.1.

performed if this Condition is AND entered prior to THERMAL A.3 Reduce Overpower T 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each POWER trip setpoints 1% for FQC(Z) determination exceeding 75% each 1% that THERMAL RTP after a POWER is limited refueling. below RTP by Required

Action A.1.

FQC(Z) not within limit. 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 B. FQW(Z) not within B.1.1 Implement a RAOC 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits operating space specified in the COLR that restores FQW(Z) to within its limits.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.1.2 Perform SR 3.2.1.1 and SR 3.2.1.2 if control rod motion is required to comply with the new operating space.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION OR 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.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after each Limit allowable FQW(Z) determination THERMAL POWER and AFD limits as specified in the COLR.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each B.2.2 Limit Power Range FQW(Z) determination Neutron Flux - High trip setpoints 1% for each 1% that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each B.2.3 Limit Overpower T FQW(Z) determination trip setpoints 1% for each 1% that THERMAL POWER is limited below RTP by Required Action B.2.1.

AND Prior to increasing B.2.4 Perform SR 3.2.1.1 and THERMAL POWER SR 3.2.1.2. above the limit of Required Action B.2.1 C. Required Action C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and associated OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQC(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQC(Z) was last verified AND In accordance with the Surveillance Frequency Control Program SR 3.2.1.1 Verify FQW(Z) is within limit Once after each refueling within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER exceeds 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQW(Z) was last verified AND OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION In accordance with the Surveillance Frequency Control Program TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor FH(X,Y)

The licensee proposed deleting TS 3.2.2 and replacing it with the following:

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH)

LCO 3.2.2 FNH shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1.1 Restore FNH to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Actions limit.

A.2 and A.3 must be completed OR whenever Condition A is A.1.2.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months entered. POWER to < 50% RTP.

FNH not within AND limit.

A.1.2.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Neutron Flux - High trip setpoints to 55%

RTP.

AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months A.2 Perform SR 3.2.2.1.

AND A.3 ------------NOTE-------------

THERMAL POWER does not have to be reduced to comply with this Required Action.

Prior to THERMAL Perform SR 3.2.2.1. POWER exceeding 50%

RTP OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION AND Prior to THERMAL POWER exceeding 75%

RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER reaching 95% RTP B. Required Action B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FNH is within limits specified in Once after each the COLR. refueling prior to THERMAL POWER exceeding 75% RTP AND In accordance with the Surveillance Frequency Control Program TS 3.2.4, Quadrant Power Tilt Ratio (QPTR)

Proposed changes to Required Action A.3 for Condition A (QPTR not within limit).

A.3 Perform SR 3.2.1.1, SR 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving 3.2.1.2, SR 3.2.1.3, and equilibrium conditions SR 3.2.2.1, and SR from a THERMAL 3.2.2.2. POWER reduction per Required Action A.1 AND OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Once per 7 days thereafter Proposed changes to Required Action A.6 A.6 -------------NOTE------------ Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after Perform Required Action achieving equilibrium A.6 only after Required conditions at RTP not to Action A.5 is completed. exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after

increasing THERMAL POWER above the limit Perform SR 3.2.1.1, of Required Action A.1 SR 3.2.1.2, and SR 3.2.1.3, SR 3.2.2.1, and SR 3.2.2.2.

Proposed changes to SR 3.2.4.2 SR 3.2.4.2 ------------------------NOTE------------------------- Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Only required to be performed if input to QPTR from one or more Power Range AND Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP. In accordance with the

Surveillance Frequency Control Program Verify QPTR is within limit using core power distribution measurement information.the movable incore detectors.

TS 3.3.1, Reactor Trip System (RTS) Instrumentation Proposed changes to Surveillance Requirement (SR) 3.3.1.3 SR 3.3.1.3 ------------------------NOTE-------------------------

Not required to be performed until 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months after THERMAL POWER is 15% RTP.

Compare results of the incore detectorcore In accordance with the power distribution measurements to Surveillance Frequency Nuclear Instrumentation System (NIS) Control Program OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION AFD. Adjust NIS channel if absolute difference is 3%.

Proposed changes to SR 3.3.1.6 SR 3.3.1.6 ------------------------NOTE-------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 50% RTP.

Calibrate excore channels to agree with In accordance with the incore detectorcore power distribution Surveillance Frequency measurements. Control Program Proposed changes to Table 3.3.1-1, Note 1:

Where: T is measured RCS T,°F.

T0 is the indicated T at RTP,°F.

S is the Laplace transform operator, sec-1.

T is the measured RCS average temperature,°F.

T' is the nominal Tavg at RTP, 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**/oF K3 = 0.00055**/psig 1 33** sec 2 4** sec 4 5** sec 5 3** sec The values denoted with ** are specified in the COLR.

Proposed changes to Table 3.3.1-1, Note 2:

Where: T is measured RCS T,°F.

T0 is the indicated T at RTP,°F.

S is the Laplace transform operator, sec-1.

T is the measured RCS average temperature,°F.

T is the nominal Tavg at RTP, 578.2**°F.

K4 1.087** K5 0.02**/oF for increasing Tavg K6 0.0011**/oF when T > T 0**/oF for decreasing Tavg 0**/oF when T T 3 10** sec 4 5** sec 5 3** sec The values denoted with ** are specified in the COLR.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION TS 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits Proposed changes to the Limiting Condition for Operation (LCO)

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 psiagreater than or equal to the limit specified in the COLR;

b. RCS average temperature is 583oFless than or equal to the limit specified in the COLR; and

c. RCS total flow rate 378,400360,000 gpm and greater than or equal to the limit specified in the COLR.

Proposed changes to Surveillance Requirements (SRs) 3.4.1.1 through 3.4.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is 2220 psia In accordance with the greater than or equal to the limit Surveillance Frequency specified in the COLR. Control Program SR 3.4.1.2 Verify RCS average temperature is In accordance with the 583oFless than or equal to the limit Surveillance Frequency specified in the COLR. Control Program SR 3.4.1.3 Verify RCS total flow rate is In accordance with the 378,400360,000 gpm and greater than or Surveillance Frequency equal to the limit specified in the COLR. Control Program SR 3.4.1.4 Verify by measurement that RCS total flow In accordance with the rate is 378,400360,000 gpm and greater Surveillance Frequency than or equal to the limit specified in Control Program the COLR.

TS 4.2, Reactor Core OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 4.2.1 Fuel Assemblies [Unit 1]

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Optimized ZIRLOTM, 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.

Fuel Assemblies [Unit 2]

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Optimized ZIRLOTM, Zircaloy or M5 . 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 2 Operating Cycle 10 core.]

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 5352 full length and no part length control rod assemblies (with no full length control rod assembly in core location H-08). 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.

TS 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; OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 1.2. LCO 3.1.1, SHUTDOWN MARGIN (SDM);

2.3. LCO 3.1.3, Moderator Temperature Coefficient (MTC);

4. LCO 3.1.4, Rod Group Alignment Limits; 3.5. LCO 3.1.5, Shutdown Bank Insertion Limits; 4.6. LCO 3.1.6, Control Bank Insertion Limits;

7. LCO 3.1.8, PHYSICS TESTS Exceptions - MODE 2; 5.8. LCO 3.2.1, Heat Flux Hot Channel Factor (FQ(X, Y, Z))(RAOC-T(Z)

Methodology);

6.9. LCO 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNH(X,Y));

7.10. LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFD);

8.11. LCO 3.3.1, Reactor Trip System (RTS) Instrumentation,; f1(I) limits for Overtemperature T and f2(I) limits for Overpower T Nominal Trip Setpoints; and

12. LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and 9.13. LCO 3.9.1, Boron Concentration.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. BAW 10180 A, Revision 1, NEMO Nodal Expansion Method 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;

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; OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION

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.

1. WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986;

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; OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION

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.

2.3 Regulatory Review The NRC staff reviewed the licensees application to determine whether (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that the activities proposed will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or the health and safety of the public. The NRC staff considered the following regulatory requirements, guidance, and licensing and design-basis information during its review of the proposed changes.

Section 50.36, Technical specifications, of Title 10 of the Code of Federal Regulations (10 CFR) establishes the regulatory requirements related to the content of TSs.

Section 50.36(a)(1) requires an application for an operating license to include proposed TSs. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Pursuant to 10 CFR 50.36(b), each license must include TSs, which are derived from the safety analysis report, as amended. The Commission may include such additional TSs as it finds appropriate.

Pursuant to 10 CFR 50.36(c), TSs for operating reactors are required, in part, to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

Section 50.36(c)(1)(i)(A) of 10 CFR states, in part, 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 of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down.

10 CFR 50.36(c)(1)(ii)(A) states in part, that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.

Section 50.36(c)(2) of 10 CFR states, in part, that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility, and when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

Section 50.36(c)(2)(ii) of 10 CFR states, in part, that an LCO must be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Section 50.36(c)(3) of 10 CFR states, in part, that SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Section 50.36(c)(4) of 10 CFR states, in part, that design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3).

Section 50.36(c)(5) of 10 CFR states, in part, that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

The regulation, 10 CFR 50.46(b), states, in part, the following:

OFFICIAL USE ONLY PROPRIETARY INFORMATION

1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200 °F [degrees Fahrenheit].

2) Maximum cladding oxidation. The calculated total oxidation of the cladding, nowhere exceed 0.17 times the total cladding thickness before oxidation.

3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

SQN, Units 1 and 2 were designed to meet the intent of the proposed General Design Criteria for Nuclear Power Plant Construction Permits, dated July 1967 (ADAMS Accession No. ML043310029). The SQN, Units 1 and 2 construction permits were issued in May 1970. The Updated Final Safety Analysis Report (UFSAR) addresses the General Design Criteria (GDCs) published as Appendix A to 10 CFR Part 50 in July 1971. In the UFSAR, each criterion is followed by a discussion of the design features and procedures that meet the intent of the criteria. Any exception to the 1971 GDCs resulting from the earlier commitments is identified in the discussion of the corresponding criterion.

Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 establishes the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The following GDCs are applicable to this review:

GDC 2 - Design Bases for Protection Against Natural Phenomena states: Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

GDC 4 - Environmental and Missile Design Bases states: Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

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. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION GDC 10 - Reactor Design states: The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Guidance documents NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 4.2, Revision 3, Fuel System Design, dated March 2007 (ADAMS Accession No. ML070740002).

NUREG-0800, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, dated March 2007 (ADAMS Accession No. ML070550016).

The NRC staffs guidance for review of TSs are Chapter 16, Technical Specifications, of NUREG-0800, Revision 3, dated March 2010 (ADAMS Accession No. ML100351425) and NUREG-1431, Volume 1, Revision 4.0, Standard Technical Specifications [STSs]

Westinghouse Plants, dated April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228).

NRC Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, dated May 1989 (ADAMS Accession No. ML003739584).

NRC Regulatory Guide 1.203, Transient and Accident Analysis Methods, dated December 2005 (ADAMS Accession No. ML053500170).

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees application to determine whether the proposed changes are consistent with the guidance, regulations, and plant-specific design and licensing basis discussed in the Section 2.3 of this safety evaluation. The NRC staff reviewed the licensees statements in the LAR, the referenced NRC-approved reload methodology, and the relevant sections of the SQN, Units 1 and 2 UFSAR. The staff reviewed all limitations and conditions for all of the proposed topical reports to assure they are met.

The staff also reviewed the proposed changes to verify that the new LOCA methodology is an approved NRC code and that all limitations and conditions are met, that the licensee appropriately applied the LOCA Evaluation Model (EM) to SQN, Units 1 and 2, and that the results meet the acceptance criteria of 10 CFR 50.46(b)(1) through (4).

As part of the LAR, a significant number of changes are proposed to the TS as shown in Section 2.2 of this SE. Each change is described in detail in the following sections.

3.1 Safety Limits Changes (TS 2.0)

As shown in Section 2.2 of this SE, the three Framatome correlation departure from nucleate boiling ratio (DNBR) values in the Reactor Core Safety Limit 2.1.1.1 as specified for the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Framatome HTP fuel are proposed to be replaced with the WRB-2M correlation DNBRs value of 1.14 due to the transition from Framatome HTP fuel to the Westinghouse RFA-2 fuel. This replacement is based on the application of WCAP-15025-P-A evaluation methodology.

To assure the applicability of WCAP-15025-P-A to SQNs use of Westinghouse RFA-2 fuel, the licensee performed an applicability evaluation and concluded that the methodology is applicable to SQN. The NRC staff reviewed the licensees evaluation and found that the licensees evaluation and conclusion for the applicability is acceptable because the licensee addressed the limitations and conditions stated in the NRC safety evaluation on the application of WCAP-15025-P-A. The specific limitations and conditions are described in detail in Section 3.8.14 of this SE. However, the licensee only evaluated the applicability of WCAP-15025-P-A for the use of Westinghouse RFA-2 fuel in SQN. In the forthcoming transition cycles, the transition core will still contain the legacy non-Westinghouse RFA-2 fuel, i.e., the Framatome HTP fuel. The licensee provided justification that only one DNBR limit and correlation is applicable to a transition core that contains two different types of fuel, as discussed below.

The licensee stated that the DNBR limit provided by Westinghouse as an update to TS Safety Limit 2.1.1.1 will only be applied to Westinghouse fuel and that the DNBR limit for Framatome fuel will continue to be applied as listed in UFSAR Section 4.5.4.1.1. The licensee also stated that the Framatome analysis of record (AOR) is established with a nuclear enthalpy rise hot channel factor (FNH) value of 1.70 as listed in UFSAR Section 4.5.4.3.2.1. However, the licensee proposed a 5 percent reduction in the surveilled FNH from the Westinghouse FNH limit to ensure the impacts due to the mixed core are adequately addressed for Framatome fuel.

The 5 percent FNH reduction is described in Section 3.9 of this SE and was determined to be an acceptable penalty factor to allow the Framatome HTP fuel to be non-limiting in mixed transition cores. Therefore, the staff finds the use of the single DNBR limit of 1.14 in TS Safety Limit 2.1.1.1 acceptable.

As shown in Section 2.2 of this SE, the current maximum local fuel pin centerline temperature values in Reactor Core Safety Limit 2.1.1.2 are proposed to be replaced with a new value and new decreasing rate. This replacement values are determined based on the application of WCAP-17642-P-A evaluation methodology.

To assure the applicability of WCAP-17642-P-A to SQNs use of Westinghouse RFA-2 fuel, the licensee performed an applicability evaluation and concluded that the application is applicable to SQN. The NRC staff reviewed the licensees evaluation and found that the licensees evaluation and conclusion for the applicability is acceptable because the licensee addressed the limitations and conditions stated in the NRC safety evaluation on the application of WCAP-17642-P-A. The specific limitations and conditions are described in detail in Section 3.11.5.5 of this SE. However, the licensee only evaluated the applicability of WCAP-17642-P-A for the use of Westinghouse RFA-2 fuel in SQN. In the forthcoming transition cycles the transition reactor core will still contain the legacy non-Westinghouse RFA-2 fuel, i.e. the Framatome HTP fuel. The licensee provided justification that only one maximum local fuel pin centerline temperature limit and associated limit degraded rate is specified for a transition core that contains two different types of fuel, as discussed below.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated that the local fuel pin centerline temperature limit provided by Westinghouse as an update to TS Safety Limit 2.1.1.2 will only be applied to Westinghouse fuel, and is not applicable to the Framatome fuel. As to prevention of Framatome fuel centerline melt, the licensee stated that the maximum linear heat rate for Framatome fuel, based on the Framatome centerline temperature limits, will continue to be applied as listed in UFSAR Section 4.5.4.2.2.6 as an analytical limit. Specifically, the kW/ft limits of 20.05 for UO2 rods and 18.95 kW/ft for UO2-Gd2O3 will be confirmed analytically to ensure the prevention of fuel centerline melt. The heat flux hot channel factor (FQ) limit for COLR item 2.6.1 (corresponding with TS 3.2.1) is based on the most recent COLR supporting Framatome HTP fuel and will be utilized with all fuel surveilled to a limit of 2.62. Therefore, the licensee concluded that these surveilled limits in combination with the analytical confirmation of acceptable kW/ft based on the Framatome centerline melt analysis ensure the centerline temperature for Framatome will be met, while the fuel pin centerline TS limit ensures the Westinghouse fuel analysis acceptance criteria will be met. Given the licensees statements, staff finds that the licensee is indirectly confirming, through analysis, that the fuel centerline melt limit is met by virtue of meeting the linear power and FQ limits. The Framatome fuel will be at its maximum power generation rate at the beginning of the first transition cycle, which is when analytical uncertainty would be at a minimum. As the once-burned Framatome fuel burns out, its power level will drop so that there would be little concern that the fuel centerline melt limit would be violated, even if the analytical uncertainty becomes large due to depletion-related uncertainty. Therefore, NRC staff finds the use of a single maximum local fuel pin centerline temperature, based on WCAP-17642-P-A, of <

5080°F, decreasing by 9°F per 10,000 MWD/MTU of burnup, acceptable.

The licensee proposed to 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 domain with consideration given to the current operating conditions (power, temperature, pressure, etc.). In addition, the licensee proposed moving the figure to the COLR. The revised figure will be based on the NRC-approved methodology in WCAP-11397-P-A. The staff finds that the licensee satisfied all the limitations and conditions on WCAP-11397-P-A, as discussed in Section 3.8.8 of this SE. Therefore, the staff finds that relocation of the figure to the COLR is acceptable based on NRC-approved Technical Specification Task Force (TSTF)-339-A, Revision 2, (ADAMS Accession No. ML003723269), Relocate TS Parameters to COLR, and the TS 5.6.3 requirement to document the limit in the COLR.

Based on the above, the NRC staff finds the proposed changes to Reactor Core Safety Limits 2.1.1.1 and 2.1.1.2 are consistent with the approved changes documented in the topical reports, provide the limits on the DNB and maximum local fuel pin centerline temperature to prevent the onset of the DNB and fuel pin melt, and are consistent with the GDC 10 requirement to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. Therefore, the proposed TSs meet the 10 CFR 50.36(c)(1) requirement to limit important process variables that are necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity and are acceptable.

3.2 Reactivity Control Systems Changes (TS 3.1)

The licensee proposed changes to TS 3.1.4, Rod Group Alignment Limits to be consistent with the revised TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)) (RAOC-T(Z) Methodology)

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Surveillance Requirement numbering, and TS 3.1.7, Rod Position Indication to implement BEACONTM core power distribution measurement.

For TS 3.1.4, Rod Group Alignment Limits, the proposed change would add and SR 3.2.1.2 to REQUIRED ACTION B.2.4 for Condition B, One rod not within alignment limits. The current required action B.2.4 specifies performance of SR 3.2.1.1, which verifies that the FQC(X, Y, Z) is within the steady state limit. However, this is based on the current heat flux hot channel factor methodology. As described below in Section 3.3, the licensee is proposing a change to use the relaxed axial offset control (RAOC) methodology. Under the RAOC methodology, new SRs 3.2.1.1 and 3.2.1.2 verify that FQ(Z) is within the allowed operating limits as specified in the COLR. The NRC staff finds that the addition of and SR 3.2.1.2 to REQUIRED ACTION B.2.4 of TS 3.1.4 is consistent with the changes to TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))

(RAOC-T(Z) Methodology), which are described in Section 3.3 of this SE, and meets the 10 CFR 50.36(c)(2) requirement that an LCO specify the lowest functional capability or performance levels of equipment required for safe operation of the facility and remedial actions that are to be taken until the condition can be met.

For TS 3.1.7, Rod Position Indication, the proposed change would replace movable incore detectors with core power distribution measurement information in REQUIRED ACTIONS A.1, A.2.1, B.3, and C.1. The staff finds this change is consistent with the changes to TS 3.3.1, Reactor Trip System (RTS) Instrumentation, which are described in Section 3.4 of this SE, meets the 10 CFR 50.36(c)(2) requirement that an LCO specify the remedial actions that are to be taken until the condition can be met, and is therefore acceptable.

3.3 Power Distribution Limits Changes (TS 3.2)

The licensee proposed to revise TS 3.2.1, Heat Flux Hot Channel Factor (FQ(X,Y,Z)) in order to address non-conservatisms identified in Westinghouse NSAL-09-5, Revision 1, Relaxed Axial Offset Control FQ Technical Specification Actions, and NSAL-15-1, Heat Flux Hot Channel Factor Technical Specification Surveillance. The non-conservatisms are related to instances where (1) the Heat Flux Hot Channel Factor is not within its limits and the Required Actions may be insufficient to restore the Heat Flux Hot Channel Factor to within the limit, and (2) SR 3.2.1.2 of NUREG-1431 may not be sufficient to assure that the peaking factor assumed in the licensing basis analysis remains valid. The licensee proposed to apply the NRC-approved methodology in WCAP-17661-P-A to resolve the issues identified in NSAL-09-5, Revision 1 and NSAL-15-1.

The licensee further proposed to change TS 3.2.1 by employing the approved changes made in WCAP-17661-P-A, Appendix A for relaxed axial offset control (RAOC) plants. The approved changes are based on NUREG-1431, STS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)). On these changes, the licensee applied the deviations from the STS based on TSTF-241-A, Revision 4, Allow Time for Stabilization after Reducing Power due to QPTR out of Limit, which the NRC approved on January 13, 1999 (ADAMS Accession No. ML040611034) and TSTF-290-A, Revision 0, Revision to Hot Channel Factor Specifications, which the NRC approved on June 30, 1999 (ADAMS Accession No. ML040630063).

The licensee also proposed amendments that would also revise TS 5.6.3.b to include WCAP-17661-P-A, Revision 1, in the list of analytical methods that must be used to determine the core operating limits documented in the COLR.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION To be consistent with changes to TS 3.2.1, the licensee proposed to revise TS 3.1.4, Rod Group Alignment Limits, Required Action B.2.4.

Summary of Proposed Changes Related to TS 3.2.1 The licensee proposed to replace SQN TS 3.2.1 in its entirety with a new TS, including a change to the title, which is based on the following references:

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

TSTF-241-A, Revision 4

TSTF-290-A, Revision 0 Section 3.2.3 of Enclosure 1 to the LAR (pages E1 37 of 63 to E1 39 of 63) provided the change details for the new SQN TS 3.2.1.

In Section 3.2.8 of Enclosure 1 to the LAR, the licensee proposed to add the following methodology to the COLR TS 5.6.3 to support TS 3.2.1 changes:

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

TS 3.1.4 Required Action B.2.4 is revised to add and SR 3.2.1.2.

The NRC staff finds that the proposed changes to SQN TS 3.2.1, including the revised title, are consistent with NUREG-1431 and updates provided in TSTF-241-A, Revision 4, and TSTF-290-A, Revision 0.

Regarding the applicability of WCAP-17661-P-A to SQN TS 3.2.1 related changes, the licensee performed an applicability evaluation in LAR Enclosure 1 Attachment 8 and concluded that the methodology is applicable to SQN. The NRC staff reviewed the licensees evaluation and finds that the evaluation and conclusion as to applicability is acceptable because the licensee satisfied the Limitations and Conditions stated in the safety evaluation for the application of WCAP-17661-P-A. The specific limitations and conditions are described and evaluated in Section 3.8.18 of this SE.

The licensee provided the detailed changes to SQN TS 3.2.1 in markup and retyped clean formats in Attachments to LAR Enclosure 1. The NRC staff reviewed and found all the proposed changes to be consistent with Appendix A of WCAP-17661-P-A except the following two deviations:

1. The new Required Action (RA) B.2.1 proposed for SQN, Limit allowable THERMAL POWER and AFD limits as specified in the COLR, differs from the B.2.1 approved in WCAP-17661-P-A B.2.1, Limit THERMAL POWER to less than RATED THERMAL POWER and reduce AFD limits as specified in the COLR.

2. The licensee proposed to add the phrase after each FQW(Z) determination to the Completion Time for the new B.2.1, B.2.2, and B.2.3.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION In its letter dated May 5, 2021, the licensee stated that deviation 1 is intended to clarify that the limits for both the THERMAL POWER and AFD (Axial Flux Difference) are specified in the COLR (example COLR is Attachment 6 to LAR Enclosure 1; Table 6 of Attachment 6 provides the limits). The NRC staff finds this justification to be acceptable and finds the proposed RA B.2.1 acceptable.

As to deviation 2, the licensee stated that the proposed addition is not based solely on TSTF-241-A Revision 4, but on a combination of the related impacts associated with TSTF-241-A and TSTF-290-A as discussed in the LAR Enclosure 1, Section 3.2.3 (page 33 of 63). As stated by the licensee, TSTF-241-A did not include the B.2.x Action Completion Time resets because the B.x.x actions for transient FQ from TSTF-290-A were not present in TS 3.2.1 at the time of TSTF-241-A issuance. TSTF-290-As B.2.x completion times were not reconciled with the TSTF-241-A changes when incorporated into NUREG-1431, Revision 2.

TSTF-241-A Revision 4 approved the addition of after each FQ(Z) determination to the completion times of TS 3.2.1A Required Actions A.1 through A.4 and after each FQC(Z) determination to the Completion Times of TS 3.2.1B Required Actions A.1 through A.3. As stated in TSTF-241-A Revision 4, this is to ensure that Actions are continued until the parameter is within its limits.

As stated in the TS 3.2.1 Bases of Appendix B to WCAP-17661-P-A, Revision 1, The limit on THERMAL POWER initially determined by Required Action B.2.1 may be affected by subsequent determinations of FQW(Z) that are not within limit and would require power reductions within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the FQW(Z) determination if necessary to comply with the decreased THERMAL POWER limit. The licensee, therefore, concluded that the Completion Time resets, though absent from the TS markups of WCAP-17661-P-A, Revision 1, are necessary to allow additional THERMAL POWER and AFD limit adjustments based on the subsequent FQW(Z) determinations. The NRC staff finds the licensee justification to be acceptable and finds that the addition of after each FQW(Z) determination to the Completion Time for the new B.2.1, B.2.2, and B.2.3 is acceptable.

Based on the above, the NRC staff finds the proposed changes to TS 3.2.1 acceptable because they are consistent with the approved changes documented in the topical reports, consistent with the limits placed on the DNB and maximum local fuel pin centerline temperature, consistent with the GDC 10 requirement to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation or anticipated operational occurrences, and meet the 10 CFR 50.36(c)(2) requirement that an LCO require that a licensee shut down the reactor or follow any remedial action permitted by the TS until the LCO can be met. The staff also finds that proposed SR change meets 10 CFR 50.36(c)(3) requirements because the SRs, as revised, will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

The licensee proposed to revise TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor FH(X,Y) to reflect Westinghouse STS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNH), in NUREG-1431. Adopting the latest version of the NRC-approved STS 3.2.2 essentially returns the SQN, Units 1 and 2 Technical Specifications to its licensing basis for the FH LCO prior to SQN, Units 1 and 2 License Amendments 223/214 dated April 21, 1997 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION (ADAMS Accession No. ML013320456), which addressed the conversion from Westinghouse fuel to Framatome-Cogema Fuel, except for two Completion Times (CTs).

Summary of Proposed Changes Related to TS 3.2.2 The licensee proposed to replace SQN TS 3.2.2, including a change to the title, with a new TS that incorporates information from the following references:

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WCAP-12472-P-A (Proprietary) BEACON Core Monitoring and Operations Support System, August 1994. (ML12270A386) and WCAP-12472-P-A, Addendum 4, Revision 0 (Proprietary) BEACON Core Monitoring and Operation Support System, Addendum 4, September 2012.

The change details for the new SQN TS 3.2.2 are provided in Attachment 1 to LAR Enclosure 1.

TS 3.2.4, Quadrant Power Tilt Ratio (QPTR), SRs 3.2.1.3 and 3.2.2.2 would be deleted from Required Actions A.3 and A.6 and the movable incore detectors in SR 3.2.4.2 would be replaced with core power distribution measurement information.

The staff finds that the NUREG-1431, Revision 4, WCAP-12472-P-A and Addendum 4 related changes to SQN TS 3.2.2 are acceptable because SQN is a Westinghouse PWR plant and any WCAP limitations and conditions are met.

For the WCAP-12472-P-A and Addendum 4 related changes to SQN TS 3.2.2, the licensee performed an applicability evaluation in Attachment 8 to LAR Enclosure 1 and concluded that the changes are applicable to SQN. The NRC staff reviewed the licensees evaluation and finds that the licensees evaluation and conclusion is acceptable because the licensee satisfied the conditions stated in the NRC staff safety evaluation on application of WCAP-12472-P-A and Addendum 4. The details of the staffs review of the specific limitations and conditions are in Section 3.11.5.2 and 3.11.5.3 of this SE.

The licensee provided the detailed changes to SQN TS in markup and retyped clean formats in Attachments to LAR Enclosure 1. The staff reviewed and found all proposed changes consistent with NUREG-1431, Revision 4, and WCAP-12472-P-A and Addendum 4.

Based on the above, the NRC staff finds the proposed changes to TS 3.2.2, including the revised title, acceptable because they are consistent with the approved changes documented in the STS and topical reports and the limits placed on the hot channel factor act as acceptable fuel design limits to prevent exceeding the acceptance criteria for fuel; and consistent with GDC 10 requirements. The staff finds that 10 CFR 50.36(c)(2) requirements will continue to be met because the TS, as revised by the proposed changes, will continue to require the licensee shut down the reactor or follow any remedial action permitted by the TS until the LCO can be met.

The staff also finds that 10 CFR 50.36(c)(3) requirements will continue to be met because the SRs, as revised by the proposed changes, will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

3.4 Reactor Trip System (RTS) Instrumentation Changes (TS 3.3.1)

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee proposed changes to TS 3.3.1 to 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 BEACON. As part of this change, the licensee proposed to revise the existing SR 3.3.1.3 and 3.3.1.6 by replacing incore detector with core power distribution measurement. As stated in the LAR, the current SR 3.3.1.3 requires plant operators to compare the NIS excore neutron detectors channel output to the incore system AFD measurement by using movable incore detectors and SR 3.3.1.6 requires plant operators to calibrate excore channels to agree with the incore measurement system. The proposed change would allow the use of either a PDMS or the movable incore detectors to compare the results of the incore and NIS excore measurements.

Westinghouse developed the BEACON system to improve the monitoring support for Westinghouse-designed PWRs. The BEACON PDMS is a core monitoring and support package that uses Westinghouse standard instrumentation in conjunction with an analytical methodology for on-line generation of three-dimensional power distributions to provide core monitoring, core measurement reduction, core analysis, and core predictions. The BEACON system is described in Westinghouse Topical Report WCAP-12472-P-A, which was approved by the NRC staff for Westinghouse reactors in its letter dated February 16, 1994 (ADAMS Accession No. ML19346E687). In the SE on WCAP-12472-P-A, the staff found that the BEACON system is acceptable for performing core monitoring and operations support functions for Westinghouse PWRs. Note that limitations and conditions on the use of WCAP-12472-P-A are described in Section 3.11.5.2 of this SE. Given that BEACON is an acceptable core monitoring system, the staff finds it acceptable to allow the use of either a PDMS or the movable incore detectors in SRs 3.3.1.3 and 3.3.1.6 by replacing incore detector with core power distribution measurement. The staff also finds that 10 CFR 50.36(c)(3) requirements will continue to be met because the SRs, as revised by the proposed changes, will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

The purpose of the overtemperature T (OTT) and overpower T (OPT) trips is to ensure that the design limit DNBR is met and that protection is provided to ensure the integrity of the fuel. The licensee proposed relocating the OTT and OPT trip setpoint parameter values from Table 3.3.1-1 of the TSs to the COLR. As noted in NRC-approved Topical Report WCAP-14483-A, Generic Methodology for Expanded Core Operating Limits Report (ADAMS Accession No. ML020430092), the justification for moving the OTT and OPT setpoint parameter values to the COLR is based primarily on these parameters being important to the reload design. Relocating these parameters to the COLR would allow the licensee to use cycle-specific values that are then verified on a cycle-specific basis, thereby avoiding the necessity of overly conservative TS limits. This would also minimize the need for a TS change to alter a reload related parameter. The licensee proposed to use the NRC-approved setpoint methodology, WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Overtemperature T Trip Function, dated April 17, 1986 (ADAMS Accession No. ML073521507), to determine the overtemperature T and overpower T setpoints. The limitations and conditions for this methodology are described below in Section 3.8.1 of this SE.

As required, the licensee also proposed adding WCAP-8745-P-A as an analytical method that is required to be used to determine cycle-specific parameters in the TS 5.6.3, CORE OPERATING OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION LIMITS REPORT. Technical Specification Task Force Traveler TSTF-339-A, Revision 2, Relocate TS Parameters to COLR, was developed to relocate TS parameters to the COLR consistent with WCAP-14483-A. Given that the licensee proposes to use an approved setpoint methodology, which is listed in TS 5.6.3, relocating the setpoint parameters is consistent with both TSTF-339-A, Revision 2, and WCAP-14483-A, and the TS requires the cycle-specific parameters be established consistent with the safety analyses, the staff finds the relocation of the OTT and OPT trip setpoint parameter values from Table 3.3.1-1 of the TSs to the COLR acceptable. In addition, the staff finds that 10 CFR 50.36(c)(2) requirements will continue to be met because the TS, as revised by the proposed changes, will continue to specify the lowest functional capability or performance level of equipment required for the safe operation of the facility.

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

The licensee proposed to revise TS 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits to reflect Westinghouse STSs in NUREG-1431, Revision 4, to revise the specified RCS total flow rate, and to 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 staff finds that the application of NUREG-1431, Revision 4, and WCAP-14483-A to SQN TS 3.4.1 changes are acceptable because SQN is a Westinghouse PWR plant.

For the WCAP-14483-A to SQN TS 3.4.1 changes, the licensee performed an applicability evaluation in LAR Enclosure 1 Attachment 8 and concluded that the application is applicable to SQN. The NRC staff reviewed the licensees evaluation and finds that the licensees evaluation and applicability conclusion is acceptable because the licensee satisfied the conditions stated in the safety evaluation on the application of WCAP-14483-A. The staff review of the limitations and conditions on the use of WCAP-114483-A is discussed in Section 3.11.5.4 of this SE.

The licensee provided the detailed changes to SQN TS in markup and retyped clean formats in LAR Attachments to LAR Enclosure 1. The staff review found all proposed changes to be consistent with the approved information in NUREG-1431, Revision 4, and WCAP-14483-A.

Based on the above, the NRC staff finds the proposed changes acceptable because they are consistent with the approved changes documented in the STSs and topical report, and the limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed and act as acceptable fuel design limits to prevent exceeding the acceptance criteria for fuel; and therefore is consistent with the requirements in GDC 10. The staff finds that the 10 CFR 50.36(c)(2) requirements will continue to be met because the TS, as revised by the proposed changes, will continue to require the licensee shut down the reactor or follow any remedial action permitted by the TS until the LCO can be met. The staff also finds that 10 CFR 50.36(c)(3) requirements will continue to be met because the SRs, as revised by the proposed changes, will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

3.6 Fuel Assemblies Changes (TS 4.2.1)

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OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee proposed modifying TSs 4.2.1, Fuel Assemblies, and 5.6.3, Core Operating Limits Report, to allow the use of Optimized ZIRLOTM as an approved fuel rod cladding material. The proposed change to TS 4.2.1 would add Optimized ZIRLOTM and delete the reference to Framatome-Cogema Fuels report BAW-2328. As part of the LAR, the licensee requested an exemption 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, and 10 CFR 50, Appendix K, ECCS Evaluation Models.

The exemption request is necessary because the regulations either specify or presume use of either Zircaloy or ZIRLO fuel rod cladding. The staff separately processed the exemption request and granted exemptions from 10 CFR 50.46 and Part 50, Appendix K.I.5, on October 26, 2021 (ADAMS Accession No. ML21166A166). Therefore, the staff finds it acceptable to add Optimized ZIRLOTM to TS 4.2.1. In addition, since the licensee does not want to use Framatome methods for lead test assemblies, staff finds it acceptable to remove the sentences 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 for Unit 1 and 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 2 Operating Cycle 10 core for Unit 2.

3.7 Removal of Control Rod Assembly H-08 (TS 4.2.2)

During testing and inspections performed during SQN Unit 1 refueling outage Cycle 23, wear of the CRDM stationary gripper latch mechanism was discovered that resulted in the inability to maintain the H-08 control rod in the fully withdrawn or nearly fully withdrawn position. Similar conditions were discovered later with SQN Unit 2 during the Unit 2 refueling outage Cycle 23.

For both SQN, Units 1 and 2, the licensee submitted exigent LARs to permit temporary operation with 52 out of 53 full length control assemblies with no control rod installed in core location H-08. For each unit, the staff found the licensees proposed use of 52 control rod assemblies acceptable as being a design change that is consistent with the current design basis and that does not challenge the safety analyses detailed in Chapter 15 of the UFSAR. The staff reviewed and approved the LAR for Unit 1 in Amendment No. 348 dated November 21, 2019 (ADAMS Accession No. ML19319C831) and Unit 2 in Amendment No. 342 dated April 23, 2020 (ADAMS Accession No. ML20108F049). Based on approval of these amendments, the RCCA at core location H-08 was removed for Unit 1 Cycle 24 which began in the fall of 2019 and the RCCA at core location H-08 was removed for Unit 2 Cycle 24 which began in the spring of 2020.

The licensee stated that the following work activities have been performed in support of removing control rod H-08.

Unlatched the control rod drive shaft from the RCCA and CRDM and completely removed the drive shaft from the reactor vessel

Removed RCCA located in core location H-08

Installed a flow restrictor in the H-08 control rod guide tube (CRGT)

Removed H-08 control rod inputs to the Rod Position Indication (RPI) system

Modified plant computer position indication and alarm points for the H-08 control rod OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION

Removed visual indications of rod position and rod bottom light for the H-08 control rod on the Main Control Room M-4 panel

Removed rod control system fuses for control power to the H-08 CRDM and lift cables for both control and position indication

Reprogrammed the Integrated Computer System computer to account for the H-08 control rod being removed The licensee submitted a third LAR on June 12, 2020 (ADAMS Accession No. ML20164A270) to allow operation without the H-08 control rod for an additional operating cycle for each unit (Cycle 25). The staff review and approved this LAR for Units 1 and 2 in Amendment Nos. 354 and 347, respectively, dated March 3, 2021 (ADAMS Accession No. ML21021A349).

In its June 12, 2020, LAR, the licensee evaluated options for repairing or replacing control rod H-08 as an alternative to operating with control rod H-08 permanently removed and concluded that the consequences and uncertainties associated with a CRDM repair or replacement are significant. Repairing or replacing the H-08 CRDM would require the use of specialized remote tooling and processes that do not currently exist and would necessitate cutting and welding on the reactor coolant system pressure boundary. In its September 23, 2020 LAR, the licensee proposed to revise TS 4.2.2, Control Rod Assemblies to require that each core contain 52 rod cluster control assemblies (RCCAs) with no full-length control rod assembly in core location H-08 for Units 1 and 2. This proposed change would remove any cycle-specific restraints associated with this configuration. Unless otherwise stated, the licensee provided information below is from the Cycle 24 and Cycle 25 LARs.

3.7.1 Design Change Process - September 2020 LAR As part of its design change process, the licensee reviews plant design changes to determine any impacts to the UFSAR, including the Chapter 15 accidents. If any change affects core reload design of UFSAR Chapter 6 and 15 safety analysis, the impacts of the change are evaluated in the core reload design process to ensure UFSAR Chapter 15 safety analyses remain bounding.

The Westinghouse reload safety evaluation methodology, WCAP-9272-P-A, which is requested in the LAR dated September 23, 2020, is used as part of the overall core reload design process.

In this process, plant design changes are reviewed for potential impact to core parameters. The licensee stated that it performs the core reload design process for each new fuel cycle regardless of whether there are any plant design changes that could impact the core design for that cycle.

The licensee stated that the reload safety analysis methods are not invalidated by the removal of the H-08 control rod from the core design because these methods are not dependent on a particular RCCA configuration. The reload safety analysis methods and supporting computer codes remain applicable to model and evaluate the as-designed/operated configuration of the plant as the reload methodology is not dependent upon control bank configuration. The licensee performs cycle-specific reload evaluations of TS limits, safety analysis limits, and operating limits to ensure core protective and operating limits remain satisfied and safety analysis limits remain bounded.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.7.2 Flow Restrictor 3.7.2.1 Flow Restrictor Thermal-Hydraulic Impacts Upon removal of control rod H-08 in each unit, the licensee installed a flow restrictor in the H-08 control rod guide tube. The flow restrictor is intended to maintain RCS flow characteristics through the upper internals guide tube. The licensee stated that the flow restrictor assembly is a readily available part designed so that flow entering or exiting the upper plenum will be essentially unchanged when compared to the original guide tube housing plate design with the control rod drive shaft in place.

The licensee performed an evaluation to show that the control rod H-08 removal has a very small impact in core thimble bypass flow and remains bounded by the flow assumed in the safety analysis supporting the fuel transition. The NRC staff reviewed the analysis during the regulatory audit and finds that the installation of the flow restrictor on a permanent basis is acceptable because the flow restrictor maintains adequate thermal-hydraulic configuration of the reactor vessel upper internals.

3.7.2.2 Use of Flow Restrictor - Structural Evaluation 3.7.2.2.1 Dynamic Analysis The licensee stated that the changes in reactor coolant system water volume and metal mass is negligible due to removal of the control rod assembly and the installation of the flow restrictor.

The licensee concluded that the effect on the dynamic analysis is negligible. The NRC staff also considers the change in system masses due to the removal of the control rod assembly H-08 to be negligible because the mass of one control rod assembly is small compared to the mass of the reactor coolant system. Therefore, the staff finds that the impact on the dynamic analyses that predicts the stresses in the CRDM, reactor vessel, vessel supports, and reactor internals when subjected to seismic or loss of coolant accident (LOCA) excitations is negligible.

The licensee stated that there is no impact on the functionality or structural integrity of the reactor vessel upper internals with the removal of the control rod drive shaft and RCCA at core location H-08 when a flow restrictor is installed in its place. A licensee evaluation shows that control rod assembly H-08 removal has a very small impact in core thimble bypass flow.

Therefore, the licensee concluded that there is insignificant impact on the current reactor vessel internals analyses. The NRC staff concludes that there should not be any significant impact on the current reactor vessel internals analysis because the hydraulic loads on reactor internals will be similar, as the flow restrictor provides flow and pressure loss at core location H-08 for Units 1 and 2 that is similar to the removed control rod assembly.

Based on the above, the NRC staff finds that the current CRDM dynamic stress evaluations due to seismic and LOCA excitations in the UFSAR remain valid because the impact of the mass change or reduction is not significant; and, therefore, continue to meet the requirements established in GDC 2.

3.7.2.2.2 Flow Restrictor Design OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated that the installed flow restrictor is a standard component used to hydraulically simulate the CRDM drive shaft clearance with the guide tube housing opening.

The flow restrictor will establish hydraulically equivalent flow conditions in the upper internals when the drive shaft is removed. The licensee performed a generic structural analysis of the restrictor plate/orifice assembly using a bounding pressure differential load for the faulted service condition (i.e., LOCA). The analysis conservatively assumed no orifice holes in the assembly to maximize the differential pressure load. The analysis demonstrated that all membrane and bending, bearing, and shear stress intensities satisfy the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code),Section III. The licensees analysis also demonstrated that bolting was adequate to resist assembly separation for maximum LOCA pressure loads. The NRC staff finds that the licensees generic structural analysis bounds the SQN, Units 1 and 2 plant-specific service conditions because the generic structural analysis maximizes the differential load and is conservative.

Further, the licensee stated in the LAR for Cycle 24 (ADAMS Accession No. ML20108F672) that the materials used for the flow restrictor assembly conform to the ASME BPV Code,Section II, Part A. The material is Type 304 stainless steel for restrictor assembly, as well as the guide tube, and is compatible with fluid conditions in the reactor vessel upper internals. The NRC staff concludes that there will be no differential thermal expansion effects because the restrictor assembly and the guide tube are the same material.

The licensee also discussed the design to prevent the possibility of loose parts in the reactor coolant system. Installation of the restrictor is controlled to ensure that the required hex bolt preload is obtained, securely locking the flow restrictor in place at the top of the guide tube. A locking cup, which is tack welded to the flow restrictor, is crimped onto the hex bolt to prevent hex bolt rotation. The NRC staff concludes that the capture features of the flow restrictor (i.e.,

locking fingers, hex bolt cup, and hex bolt) provide assurance that the flow restrictor is securely installed and will not result in the generation of loose parts and, therefore, continue to meet GDC 4 requirements regarding compatibility with the environmental conditions associated with normal operation and postulated accidents.

The licensee stated that the reactor internals are designed and analyzed to the requirements of UFSAR Section 3.9.3, NSSS Components Not Covered by the ASME Code. The basis for the design stress and deflection criteria is summarized in Section 4.2.2.5 of the UFSAR. While the restrictor assembly does not perform a core support or safety function, all of the calculated stresses meet the ASME BPV Code allowable stress limits. The NRC staff determined that the flow restrictor assembly materials, fabrication, and design analysis meet the intent of ASME BPV Code, Subsection NG, consistent with the SQN, Units 1 and 2 design basis described in the UFSARs as supplemented by the LAR safety analysis.

3.7.2.2.3 Component Wear and Flow-Induced Vibration The licensee stated that removal of the control rod drive shaft tends to increase the thermal sleeve outside diameter wear due to the absence of damping effect from the drive rod on the thermal sleeve motion. The licensee performed a wear projection evaluation based on measurements of the thermal sleeve wall wear taken during refueling outages and estimated the wear life to be greater than 60 effective full-power years for both SQN, Units 1 and 2. The OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION thermal sleeve inside diameter wear will no longer progress when the drive shaft is removed as there is no longer anything for the thermal sleeve inside wall to contact. The licensee also stated that removal of the drive shaft would increase the thermal sleeve flange wear due to the absence of damping effect from the drive shaft. The flange wear measurements from fall 2019 for SQN, Unit 1 and fall 2018 for SQN, Unit 2 show moderate wear of the thermal sleeve flange at core location H-08. The licensee concluded that it is unlikely that thermal sleeve wear would progress to the point of separation over the next operating cycle. Even if separation should occur as a result of thermal sleeve outside diameter wear or flange wear, the separated condition will not result in a condition adverse to safety. The NRC staff finds that it is unlikely that the thermal sleeve wear and flange wear would progress to the point of separation over the next operating cycle because past inspection indicated that the wears are moderate. Thermal sleeve wear will be inspected in accordance with the Sequoyah in-service inspection program required by 10 CFR 50.55a during future outages. In addition, even if separation should occur, it is unlikely that the separated parts would adversely affect the integrity of the fuel or core support structures and result in a condition adverse to safety because the thermal sleeve and flange will be captured by the upper core plate and will not become loose parts in the reactor coolant system.

The licensee stated that the removal of the control rod drive shaft and installation of the flow restrictor without fuel assembly thimble plugging will not affect the upper head plenum flow and hydraulic characteristics; therefore, the flow-induced vibration of the drive rods and control rods in the remaining guides tubes, core exit thermal couples, mechanical instrumentation remnants, or their associated conduits is not affected. The NRC staff finds that the removal of the drive shaft and installation of the flow restrictor will unlikely affect the flow-induced vibration of the upper reactor internals components because the changes to the upper head plenum flow and hydraulic characteristics are insignificant.

3.7.3 Evaluation of Core Reload Design Impacts As part of its LAR, the licensee proposed using the NRC-approved methodology described in WCAP-9272-P-A to ensure each core reload design is acceptable. The licensee stated that the analyses used in the WCAP-9272-P-A methodology is not dependent on a minimum number of RCCAs in a core, a particular RCCA configuration, or the existence of a symmetric RCCA pattern but instead uses bounded reference safety analyses. The impact of the exact configuration of the core loading pattern, including RCCAs, is explicitly captured in the calculations performed each cycle to assess cycle-to-cycle loading patterns and minor fuel changes. Since RCCAs are not explicitly modeled in the bounding reference safety analysis, the licensee provided a general overview of how the removal of RCCA at location H-08 impacts the different key safety parameters as defined by the methodology.

As stated in the SE for WCAP-9272-P-A:

The key safety parameters form the basis for determining whether the reference safety analysis applies for a reload cycle following changes in fuel assemblies and configuration, and performance or setpoint changes in the reactor plant systems. The basis of selection of a parameter as a key safety parameter for a given accident is that (i) the parameter could change as a result of core rearrangement, and (ii) if it changed, it could affect the accident consequence. For each reload cycle, values of the key safety OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION parameters are determined for the reload core during the nuclear, thermal and hydraulic, and fuel rod design processes.

As part of the methodology in WCAP-9272-P-A, when a reload key safety parameter is not bounded, each accident, which includes the parameter, is separately evaluated to determine the impact of the deviation on the accident. If the magnitude of the effect is not easily quantifiable by the bounding evaluation, then a reanalysis is performed to ensure the required margin of safety is maintained for each affected accident.

3.7.3.1 Impacts on WCAP-10216-P-A, Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification, February 1994 and WCAP-17661-P-A As described above in Section 3.2 of this SE, the licensee is proposing use of WCAP-10216-P-A, Revision 1A, which provides methodology for RAOC. The use of this methodology does not depend upon the control rod pattern or number of control rods. The specific neutronics codes explicitly account for the removal of the RCCA at core location H-08.

Therefore, the staff finds that the removal of the RCCA at location H-08 does not impact the applicability of the methodology in WCAP-10216-P-A, Revision 1A.

As described above in Section 3.3 of this SE, the licensee is proposing use of WCAP-17661-P-A, Revision 1, which provides an improved method for FQ surveillance. Since the calculations supporting this methodology are from WCAP-10216-P-A, Revision 1A, the staff finds the removal of control rod H-08 does not impact the applicability of WCAP-17661-P-A methodology.

3.7.3.2 Neutronic Code Capability The nuclear design analytical methods and codes used in the application of the WCAP-9272-P-A methodology include the Advanced Nodal Code (ANC) (WCAP-10965-P-A) and PARAGON/NEXUS (WCAP-16045-P-A, WCAP-16045-P-A, Addendum 1-A). The licensee stated that these codes/methodologies are rigorously benchmarked and qualified for a variety of reactor types (Westinghouse 2, 3, 4 loop, Combustion Engineering), fuel types (various lattice types and fuel rod diameters) and burnable poison types (IFBA, WABA, Pyrex, Gadolinia).

ANC is a nodal neutronics code for multidimensional reactor core calculations, including the prediction of such design parameters as reactivity, assembly average power, rod power and flux, Doppler coefficients, moderator coefficients, boron worth, control rod worth, burnable absorber worth, depletion, and other safety-related parameters.

The lattice code the licensee used to provide multi-group data to ANC has been updated to PARAGON/NEXUS. The licensee stated that the qualification of ANC included a broad spectrum of reactor, fuel and burnable absorber designs. The approved topical reports WCAP-16045-P-A, Addendum 1-A and WCAP-10965-P-A, demonstrate that ANC is an accurate analytical tool for multidimensional nuclear calculations performed in the design, safety analyses, and operational follow of pressurized water reactor cores. The licensee stated that a change in the number of RCCAs is sufficiently represented by the broad spectrum of reactor, fuel, and burnable absorber designs as well as the off normal condition analyses evaluated in WCAP-16045-P-A, Addendum 1-A and WCAP-10965-P-A and does not impact the capabilities of the codes/methodology or the calculational uncertainties used in the methodology.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Based on the above, staff finds that the neutronics codes used do not depend on the number of RCCAs and the removal of control rod H-08 does not impact the capabilities of the codes and cited methodologies.

3.7.3.3 Key Safety Parameter Impact Per the methodology in WCAP-9272-P-A, the key safety parameters are evaluated for each cycle-specific core loading pattern. The licensee provided a summary of these parameters as follows:

Moderator Temperature Coefficient (MTC)/Moderator Density Coefficient (MDC): Slight impact to the most positive MDC because this parameter is conservatively calculated assuming all RCCAs are inserted into the core. A cycle-specific evaluation of the MTC/MDC values with the removal of Control Rod H-08 will be performed for each core reload design to confirm the most positive MDC remains bounding.

Fuel Temperature (Doppler) Coefficient: Potential increase to the fission rate primarily in a single assembly depending on the control bank D position. This increase is typically in a low power area during hot full power (HFP) operation primarily affecting a single assembly which has negligible impacts impact on global average fuel temperatures.

Boron Worth: Minor impact as the primary factor in determining boron worth is the total boron concentration.

Effective Delayed Neutron Fraction: Potential increase to the fission rate in the top portion of a single assembly depending on the control bank D position. This increase is primarily in a low power area (top end of the fuel assembly) which has negligible impacts on power sharing and the delayed neutron fraction.

Prompt Neutron Lifetime: Potential increase to the fission rate in the top portion of a single assembly depending on the control bank D position. This increase is primarily in a low power area (top end of the fuel assembly) which has negligible impacts on power sharing and the core average prompt neutron lifetime.

Control rod H-08 is part of control bank D. This control bank can be inserted during at-power operating in accordance with the rod insertion limits (TS 3.1.6, Control Bank Insertion Limits).

The licensee provided a summary of the impacts of the removal of RCCA H-08 on the following control rod worth key parameters from WCAP-9272-P-A:

Insertion Limits: Potential impact since the RCCA H-08 can be inserted during at power operation. The insertion limits are confirmed to be acceptable each cycle through analysis of accident scenarios starting from both all rods out (ARO) and rods at the insertion limits. This will be evaluated on a cycle specific basis.

Total Rod Worth: Total rod worth will be reduced with the removal of RCCA H-08. This key safety parameter is evaluated on a cycle specific basis to ensure shutdown margin and trip reactivity limits are met.

Trip Reactivity: Reduction of trip reactivity as a function of rod insertion position, which reduces the trip reactivity as a function of time after the RCCAs begin to fall -- a cycle specific evaluation is performed to confirm the trip reactivity remains bounded by the key safety parameter value.

Differential Rod Worth: Reduction in differential rod worth of the control bank D.

Differential rod worth differences are implicitly confirmed via analysis of accident OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION scenarios starting from both ARO and rods at the insertion limits on a cycle specific basis.

The licensee provided the impacts of the removal of RCCA H-08 on the following nuclear design key parameters from WCAP-9272-P-A on the analysis of certain Chapter 15 events:

LOCA: Potential impact to the core hot channel factor parameter (FQ) and axial power distribution since RCCA H-08 can be inserted during at-power operation. The FQ determination for the LOCA event is performed on a cycle-specific basis.

Uncontrolled Boron Dilution Event: Increases the N-1 control rods inserted critical boron concentration. The removal of RCCA H-08 effectively creates an N-2 control rods inserted situation. This key safety parameter is confirmed on a cycle-specific basis.

Note that N refers to the total number of control rods, so N-X implies that X control rods will not be inserted during the event.

Single RCCA Events: Reduced impact since there are fewer rods inserted at HFP conditions. The reduction of control rods present minimizes the impact on the core power distribution and can decrease the FNH. These key safety parameters are evaluated on a cycle-specific basis.

Control Rod Ejection Accident: Ejected rod worth and FQ are impacted by the absence of RCCA H-08. This impacts the power distribution prior to event initiation and the potential rods investigated for rod ejection. This accident is evaluated for acceptability on a cycle specific basis.

Steam Line Break Accident: Potential increase in post-trip reactivity due to the removal of RCCA H-08. The power distribution at hot zero power (HZP) with all rods inserted is also impacted. These key safety parameters are confirmed for each reload on a cycle specific basis.

The licensee provided its assessment of the impacts of the removal of RCCA H-08 on the following thermal and hydraulic analyses key parameters:

Engineering Hot Channel Factors: No impact to heat flux engineering hot channel factor FQE and the enthalpy rise engineering hot channel factor FHE1 because RCCA H-08 removal does not impact fuel pellet physical characteristics and because RCCA H-08s removal has a negligible impact on at-power core power distributions.

Axial Fuel Stack Shrinkage: No impact because removal of RCCA H-08 does not impact the physical design of the fuel.

Fuel Temperatures: No impact on the physical fuel design parameters and negligible impact on the at-power core power distributions assumed in the calculation of fuel temperatures.

Rod Internal Pressure: No impact because the removal of RCCA H-08 does not affect the physical design of the fuel and its removal has a negligible impact on the at-power core power distributions assumed in the calculation of rod internal pressure.

Core Limit Lines: No impact because the removal of RCCA H-08 has no impact on the parameters presented in Table 4-2 of WCAP-9272-P-A.

The licensee performed a sub-channel thermal hydraulic analysis using the VIPRE thermal hydraulic computer code to ensure the DNBR stays above the acceptable limits for the following events in Chapter 15 of the UFSAR:

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Zero and full power steam line break

Loss of forced RCS flow

Uncontrolled RCCA bank withdrawal from a subcritical or low-power startup condition

Control rod out of position

Dropped RCCA 3.7.3.4 UFSAR Chapter 15 Safety Analysis The licensee assessed the non-LOCA safety analysis for the fuel transition program in order to determine the impact of removal of the H-08 control rod. The LOCA analysis is discussed in Section 3.10 of this SE and is not affected by removal of the H-08 control rod. While these analyses do not directly model individual control rods, the analyses are indirectly affected by changes to core neutronics safety parameters.

The core neutronics safety parameters used in the non-LOCA safety analyses were selected by the licensee to conservatively bound the values expected in subsequent reload cycles (i.e., they are intended to be cycle-independent). The licensee compares the cycle specific values to the bounding values and if the bounding values are found to remain conservative, then the analyses remain valid. The licensee found that while removal of the H-08 control rod impacts the cycle-specific values for parameters such as shutdown margin, trip reactivity, boron concentration, and moderator temperature coefficient, the values modeled in the non-LOCA safety analyses remain bounding with the exception of the HZP stuck rod coefficients.

The licensee noted that HZP conditions were not explicitly analyzed for the SQN, Units 1 and 2 in support of the fuel transition, however, there are two events that were considered. In the feedwater malfunction from HZP, Westinghouse performed a generic study to demonstrate that the consequences of an HZP feedwater malfunction with an increased feedwater flow rate of less than 150 percent of the nominal full power flow rate are non-limiting. For the steam line break from hot zero power which models a maximum break size of 1.40 ft2 and offsite power available, the analysis was rerun with the revised stuck rod coefficients. The licensee confirmed that the DNBR and fuel centerline melting safety analysis limit values continued to be met.

The licensee performed analysis for the non-LOCA events from Chapter 15 of the SQN, Units 1 and 2 UFSAR. For each event, listed below, the licensee provided a summary of the event and stated that the results support permanent removal of the H-08 control rod.

Uncontrolled Rod Cluster Control Assembly Withdrawal from Subcritical (UFSAR Section 15.2.1)

Uncontrolled RCCA Withdrawal at Power (UFSAR Section 15.2.2)

RCCA Misalignment (UFSAR Section 15.2.3)

Uncontrolled Boron Dilution (UFSAR Section 15.2.4)

Partial Loss of Flow (UFSAR Section 15.2.5), Complete Loss of Flow (UFSAR Section 15.3.4), and Locked Rotor (UFSAR Section 15.4.4)

Startup of an Inactive Reactor Coolant Loop (UFSAR Section 15.2.6)

Loss of Load / Turbine Trip (UFSAR Section 15.2.7)

Loss of Normal Feedwater / Loss of Offsite Power (UFSAR Sections 15.2.8 and 15.2.9)

Feedwater Malfunction (UFSAR Section 15.2.10)

Excessive Load Increase (UFSAR Section 15.2.11)

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Accidental Depressurization of the RCS (UFSAR Section 15.2.12)

Accidental Depressurization of the Main Steam System (UFSAR Section 15.2.13)

Inadvertent Operation of the Emergency Core Cooling System at Power (UFSAR Section 15.2.14)

Minor Secondary System Pipe Breaks (UFSAR Section 15.3.2)

Steamline Break with Coincident Rod Withdrawal at Power (UFSAR Section 15.3.7)

Hot Zero Power Steamline Break (UFSAR Section 15.4.2.1)

Feedline Break (UFSAR Section 15.4.2.2)

RCCA Ejection (UFSAR Section 15.4.6)

Anticipated Transients without Scram (Not a UFSAR Chapter 15 event)

The licensee concluded that the SQN, Units 1 and 2 non-LOCA safety analyses results support operation with the permanent removal of the H-08 control rod. The staff finds the methodology used by the licensee to make these determinations acceptable and is not dependent on the number of RCCAs. The NRC staff also finds that the analyses of the reactor core modification demonstrate that the design is consistent with the GDC 10 requirement that appropriate margin be maintained to assure that specified acceptable fuel design limits are not exceeded. Given that the new core design meets GDC 10 and that the licensee will perform cycle-specific evaluations for UFSAR Chapter 15 safety analysis parameters and confirm that the values assumed in the safety analysis remain bounding per the methodologies, the staff finds the permanent removal of control rod H-08 acceptable.

3.7.4 Nuclear Design Multi-Cycle Margin Assessment The licensee performed a multi-cycle margin assessment to determine the effect that removal of RCCA H-08 would have on the equilibrium and transition core designs generated for the SQN, Units 1 and 2 Reload Transition Safety Report (RTSR). The core designs include two potential transition cycle designs (containing both Framatome HTP and Westinghouse RFA-2 fuel) and two potential equilibrium cycles with only Westinghouse RFA-2 fuel). The licensee notes that while these core designs have not been utilized, they are representative of the upcoming transition from Framatome HTP fuel to Westinghouse RFA-2 fuel and the subsequent operation with only Westinghouse RFA-2 fuel.

The licensee discussed parameters that could potentially be impacted by the removal of control rod H-08. The licensee presented results with and without control rod H-08 and note that the removal of control rod H-08 generally reduces the margin for some key safety parameters including shutdown margin while the impacts on the analysis of some postulated accidents or transients are negligible. The values presented in the LAR are representative and the actual values will be determined during each cycles reload design confirmation. The licensee stated that the equilibrium and transition cores developed for this LAR were not designed with removal of control rod H-08 in mind and that actual core designs could potentially gain back some of the margin lost due to removal of control rod H-08.

3.7.5 Conclusion The licensee proposed modification of TS 4.2.2 to reflect use of 52 rod cluster control assemblies with no control rod assembly installed in core location H-08. The staff reviewed the information provided by the licensee regarding the design change process, thermal-hydraulic impacts and core reload design impacts that occur when control rod H-08 is removed. The staff OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION concludes that the licensees proposed use of 52 control rod assemblies, including the use of a flow restrictor, is acceptable because the design change is consistent with the current design basis and does not challenge the safety analyses detailed in Chapter 15 of the UFSAR. The staff further concludes that the licensee meets GDC 4 requirements regarding compatibility with the environmental conditions associated with normal operation and postulated accidents and GDC 10 requirements related to 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.

3.8 Core Operating Limits Report Changes (TS 5.6.3)

The licensee proposed addition of new items to the list in TS 5.6.3.a, CORE OPERATING LIMITS REPORT, including reactor core safety limits (LCO 2.1.1), shutdown margin (LCOs 3.1.4 and 3.1.8) and pressurizer pressure, RCS average temperature, and RCS flow (LCO 3.4.1). Each of the listed LCOs will reference the COLR for the limits on the specific parameters. Details on these LCOs are described above in the following sections of this SE:

Section 3.1 - LCO 2.1.1, Reactor Core Safety Limits Section 3.2 - LCO 3.1.4, Rod Group Alignment Limits Section 3.2 - LCO 3.1.8, PHYSICS TESTS Exceptions - Mode 2 Section 3.5 - LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits As noted in Section 3.1 above, the staff found the proposed relocation of LCO 2.1.1, Figure 2.1.1-1, Reactor Core Safety Limit - Four Loops in Operation to the COLR acceptable.

Because the listing of the 2.1.1 SL would require that the core operating parameter be documented in the COLR and provides controls necessary to assure operation of the facility in a safe manner as required by 10 CFR 50.36(c)(5), the NRC staff finds it acceptable to add LCO 2.1.1 to the list of core operating limits in TS 5.6.3.a.

The existing LCOs 3.1.4 and 3.1.8 include Required Actions that verify shutdown margin is within limits in the COLR. The proposed change would add these two LCOs to the list of core operating limits that TS 5.6.3.a requires be established for each reload cycle and documented in the COLR. Given that the listing of LCO 3.1.4 and 3.1.8 would require that this core operating parameter be documented in the COLR and provides controls necessary to assure operation of the facility in a safe manner as required by 10 CFR 50.36(c)(5), the staff finds it acceptable to add LCOs 3.1.4 and 3.1.8 to the list of core operating limits in TS 5.6.3.a.

As noted in Section 3.5 above, the NRC staff found the proposed relocation of specific values for limits for pressurizer pressure, RCS average temperature and RCS total flow rate in LCO 3.4.1 and SRs 3.4.1.1, 3.4.1.2 and 3.4.1.3 to the COLR acceptable. Because the listing of the LCO 3.4.1 would require that these core operating parameters be established for each reload cycle and documented in the COLR, providing controls necessary to assure operation of the facility in a safe manner as required by 10 CFR 50.36(c)(5), the staff finds the addition of LCO 3.4.1 to the list of core operating limits in TS 5.6.3.a. acceptable.

The license proposed two other changes to TS 5.6.3.a. The first would remove the parameters listed with the title for LCO 3.3.1 in TS 5.6.3.a by changing from Reactor Trip System (RTS)

Instrumentation, f1(l) limits for Overtemperature T and f2(l) limits for Overpower T Nominal OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Trip Setpoints; and to Reactor Trip System (RTS) Instrumentation. The OTT and OPT trip setpoint values from Table 3.3.1-1 are discussed above in Section 3.4 of this SE. Table 3.3.1-1 already requires that the parameters being removed be specified in the COLR. All other entries are the LCO number and title and do not identify specific parameters after the LCO title. The removal of f1(l) limits for Overtemperature T and f2(l) limits for Overpower T Nominal Trip Setpoints; and from the TS 5.6.3.a list does not alter the requirement that these parameters be documented in the COLR. Therefore, staff finds it acceptable to remove these trip setpoint values and the word and because the revised text continues to provide controls necessary to assure operation of the facility in a safe manner as required by 10 CFR 50.36(c)(5) and the deletion of and reflects the change of position in the list.

The second proposed change to TS 5.6.3.a would be consistent with the changes to the heat flux hot channel factor and nuclear enthalpy rise hot channel factor LCO title changes as discussed above in Section 3.3 of this SE. The proposed changes to TS 5.6.3.a include changing the title of LCO 3.2.1 from Heat Flux Hot Channel Factor (FQ(X, Y, Z)) to Heat Flux Hot Channel Factor (FQ(Z)) (RAOC-T(Z) Methodology) and the title of LCO 3.2.2 from Nuclear Enthalpy Rise Hot Channel Factor (FH(X,Y)) to Nuclear Enthalpy Rise Hot Channel Factor (FNH). Staff finds the two changes to the LCO titles in TS 5.6.3.a acceptable as it appropriately references the LCOs by their updated titles.

The NRC staff finds the two additional title changes to TS 5.6.3.a list of parameters to be documented in the COLR are consistent with the other changes previously evaluated in sections 3.1 through 3.5 of this SE. Given that the revisions require that these parameters be established for reload cycles and documented in the COLR, the staff finds the proposed changes to TS 5.6.3.a provide controls necessary to assure operation of the facility in a safe manner and are acceptable as required by 10 CFR 50.36(c)(5). In addition, the staff finds that (1) the proposed renumbering of the list of LCOs in TS 5.6.3.a accurately reflects the inclusion of the additional items and is therefore acceptable, and (2) the changes to TS 5.6.3.a are consistent with the STS.

The licensee proposed removing the 12 Framatome methods in TS 5.6.3.b and replacing them with 18 Westinghouse core safety analysis methodologies. The following 18 Westinghouse topical reports represent the methodologies used to determine the values located, and required to be documented, in the COLR for each reload cycle. The staff evaluation of each topical report contains the following three items: 1) the limitation and condition(s) stated in the NRC safety evaluation on the topical report; 2) the licensees September 23, 2020 LAR statements as to compliance with the limitation and condition(s); and 3) the respective NRC staff findings.

There are numerous limitations and conditions that refer to some variant of the 17x17 VANTAGE, VANTAGE 5, VANTAGE 5H or VANTAGE+ fuel. Because 17x17 RFA-2 fuel is an updated version of the base 17x17 VANTAGE design and is similar to previous designs, the NRC staff finds that all limitations and conditions on 17x17 VANTAGE designs apply to 17x17 RFA-2.

3.8.1 WCAP-8745-P-A Limitation and condition OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION We have reviewed the Westinghouse design bases for the thermal overpower and overtemperature T Trip functions described in WCAP-8745 and find them acceptable for referencing by Westinghouse in licensing documents for plants that operate under constant axial offset control [CAOC].

Although Section 1 of the topical report specifies its applicability to Westinghouse plants that reference RESAR-3S and operate under CAOC, Westinghouse has indicated that they consider WCAP-8745 applicable to all Westinghouse plants that employ overpower and overtemperature T trip for core protection. Westinghouse has stated that new methods and technology developed after the submittal of WCAP-8745 are described in separate topical reports, and do not invalidate the conclusions of WCAP-8745. As examples of such new methods, Westinghouse has cited changes in DNB analysis methodology (Improved Thermal Design Procedure and WRB-1 and WRB-2 correlations), fuel design (Optimized Fuel Assembly), and plant operating procedure (Relaxed Axial Offset Control), and referenced topical reports describing these changes.

While we agree that the basic design philosophy described in WCAP-8745 is not invalidated by changes in DNB analysis methodology, fuel design, and plant operating procedure, the application of this methodology must account for changes in system design and operation. The adequacy of the standard power shapes in establishing the core DNB protection system must be evaluated whenever changes are introduced that could potentially affect the core power distribution.

Licensee Compliance The licensee stated that Sequoyah will operate under Relaxed Axial Offset Control (RAOC) as approved in WCAP-10216-P-A. The NRC SE for WCAP-14483-A acknowledges the acceptability of using WCAP-8745-P-A as a setpoint methodology to be referenced in Sequoyah Technical Specification 5.6.3.b. Overpower T and Overtemperature T setpoint adequacy for limiting power distributions will be verified on a cycle-specific basis using the Westinghouse reload methodology described in WCAP-9272-P-A.

The NRC staff finds the licensee meets this limitation and condition because it will use an approved axial offset control method, along with assessing the adequacy of the standard power shapes on a cycle-specific basis using an NRC-approved methodology. In addition, the NRC SE on WCAP-14483-A acknowledges the acceptability of using WCAP-8745-P-A as a setpoint methodology, and shows these methodologies work together.

3.8.2 WCAP-9272-P-A Limitation and Condition Since quantitative criteria are not available for determining when an accident re-evaluation rather than a reanalysis can be performed, justification for any reevaluation should be presented in individual Reload Safety Evaluation reports.

Licensee Compliance The licensee stated that the Reload Safety Evaluation process will be implemented for Sequoyah Units 1 and 2 beginning with the first partial core loading of RFA-2 fuel.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee states that the Reload Safety Evaluation process will be implemented. The process in WCAP-9272-P-A specifies that the licensee justify reevaluations as described in Request for Additional Information (RAI) 19 to WCAP-9272-P-A (ADAMS Accession No. ML19274C484). Therefore, the NRC staff finds this limitation and condition is met.

3.8.3 WCAP-10216-P-A, Revision 1A Limitation and Condition The proposed revisions to the FQ Surveillance Technical Specification in those reactors using CAOC or RAOC for power distribution control are acceptable. These revisions would allow the incorporation of a larger penalty to account for FQ(z) increases greater than 2 percent between measurements. These penalties may be incorporated in either the plant PFLR [Peaking Factor Limit Report] or COLR, as described above, and will be calculated with NRC-approved methods. The approved version of WCAP-10216-P, Rev.

1 must be included in the Administrative Reporting Requirements Section of the TS for those plants incorporating the penalty factor in the COLR. Also, TS Surveillance 4.2.2.2.e.1 must be modified to reflect inclusion of this parameter in the PFLR or COLR.

Licensee Compliance The licensee stated that the LCO in Sequoyah TS 3.2.1 requires compliance with the FQ limit specified in the COLR TS 5.6.3.b would be revised to list WCAP-10216-P-A as a COLR methodology reference.

The NRC staff finds that the licensee meets this limitation and condition. The proposed FQ LCO requires compliance with the value referenced in the COLR and the cited methodology is being added to the COLR reference list in TS 5.6.3.b. WCAP-10216-P, Revision 1A, approved in 1994, addressed TSs with specific values related to FQ(z), however, the specific values have since been relocated to the COLR. Given that the proposed TS 3.2.1 would require that the reactor meet the limits specified in the COLR and is consistent with the STS, the NRC staff finds the condition that TS Surveillance 4.2.2.2.e.1 be modified to reflect inclusion of this parameter in the PFLR or COLR is met.

3.8.4 WCAP-10444-P-A Limitation and Condition #1 The statistical convolution method described in WCAP-10125 for the evaluation of initial fuel rod to nozzle growth gap has not been approved. This method should not be used in VANTAGE 5.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The Rod Axial Growth model in Section 5.9 of the PAD5 topical report (WCAP-17642-P-A) was used for the Sequoyah reload transition safety report (RTSR) analysis.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff finds this limitation and condition is not applicable because the PAD5 model is not the WCAP-10125 statistical convolution method.

Limitation and Condition #2 For each plant application, it must be demonstrated that the LOCA/seismic loads considered in WCAP-9401 bound the plant in question; otherwise additional analysis will be required to demonstrate the fuel assembly structural integrity.

Licensee Compliance The licensee stated that detailed site-specific and LOCA fuel assembly analyses for Sequoyah Units 1 and 2 have been performed in accordance with approved methodologies. These methodologies were approved by the NRC in WCAP-12610-P-A, WCAP-12488-A, WCAP-9401, and PWROG-16043-P-A.

The analysis predicted no permanent grid deformation (grid crush) to occur in both homogeneous core and mixed cores for Sequoyah Units 1 and 2 under combined seismic and LOCA loadings.

The fuel assembly stress evaluation was performed for the limiting seismic and LOCA (accumulator injection line and pressurizer surge line breaks) loads assuming both a full homogeneous core of RFA-2 fuel and a mixed core of RFA-2 and Framatome HTP fuel.

The results show that all grid impact forces on the 17x17 RFA-2 fuel remain below the grid impact strength for both the homogeneous and mixed core configurations. The stresses in the guide thimbles remain below the allowable stress limits for the ZIRLO thimble tubes and the fuel rod stresses remain below the allowable stress limits for the Optimized ZIRLOTM cladding. For the Sequoyah units, it is concluded that the 17x17 RFA-2 seismic and LOCA analysis for Conditions I, II, III, and IV demonstrates that core coolability is maintained, full control rod insertion within the allowed rod drop time is maintained, and fuel rod fragmentation will not occur. This discussion will be added to Sequoyah UFSAR Section 4.2.1.3.2.

The NRC staff finds that this limitation and condition is met because the licensee performed site-specific seismic and LOCA fuel assembly analyses using NRC-approved methodologies and demonstrated acceptable results. See Section 3.11.2 of this SE for more information on the seismic/LOCA loads.

Limitation and Condition #3 An irradiation demonstration program should be performed to provide early confirmation performance data for the VANTAGE 5 design.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. Since the time of this SE, the VANTAGE 5, VANTAGE 5H, RFA, and RFA-2 fuel designs have collective operating experience in the hundreds of reactor-years.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION While the NRC staff finds this limitation and condition to be applicable due to the similarities in the VANTAGE and RFA-2 designs, the staff finds that the VANTAGE 5, VANTAGE 5H, RFA, and RFA-2 fuel designs have significant collective operating experience from the 1980s to the present. As part of the regulatory audit, NRC staff examined a fuel examination report that concluded RFA-2 fuel assemblies performed well and that the fuel assembly and single rod/cell examinations data are within the bounds of the existing databases. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #4 For those plants using the ITDP [Improved Thermal Design Procedure], the restrictions enumerated in Section 4.1 of this report must be addressed and information regarding measurement uncertainties must be provided.

Licensee Compliance The licensee stated that the Revised Thermal Design Procedure (RTDP) was used for the Sequoyah RTSR analyses. The discussion of WCAP-11837-P-A [Extension of Methodology for Calculating Transition Core DNBR Penalties, January 1990] also contains relevant information on transition core DNBR penalties.

As discussed in Section 3.2.1 of Enclosure 1 to the LAR, the WCAP-11397-P-A methodology 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. These uncertainty factors are used to determine the RTDP design limit DNBR values such that there is at least a 95 percent probability at a 95 percent 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.

Because the licensee will determine the DNBR at the 95 percent probability at a 95 percent confidence level, the NRC staff finds that the restriction enumerated in Section 4.1 from the limitation and condition is met. Note that the transition core DNBR penalties are calculated using the methodology in WCAP-11837-P-A. The staff describes and finds these penalties acceptable in Section 3.9 of this SE.

Limitation and Condition #5 The WRB-2 correlation with a DNBR limit of 1.17 is acceptable for application to 17x17 VANTAGE 5 fuel. Additional data and analysis are required when applied to 14x14 or 15x15 fuel with an appropriate DNBR limit. The applicability range of WRB-2 is specified in Section 4.2.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The WRB-2M DNBR correlation with a DNBR limit of 1.14 is used for RFA-2 fuel to be loaded at the Sequoyah units as discussed under WCAP-15025-P-A.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff finds that this limitation and condition is not applicable in that the licensee uses a different DNBR correlation. The DNBR limit and WCAP-15025-P-A is discussed and found acceptable in Section 3.1 of this SE.

Limitation and Condition #6 For 14x14 and 15x15 VANTAGE fuel designs, separate analyses will be required to determine a transitional mixed core penalty. The mixed core penalty and plant specific safety margin to compensate for the penalty should be addressed in the plant Technical Specifications Bases.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. 17x17 RFA-2 fuel is being loaded at the Sequoyah units.

While the NRC staff finds that this limitation and condition is not applicable as the licensee is using 17x17 RFA-2 fuel, the staff notes that the licensee addressed the intent of the limitation and condition on mixed core penalties as described and found acceptable in Section 3.9 of this SE.

Limitation and Condition #7 Plant specific analysis should be performed to show that the DNBR limit will not be violated with the higher value of FH.

Licensee Compliance The licensee stated that there will be no DNBR limit violations, as discussed under WCAP-11837-P-A, WCAP-11397-P-A, and WCAP-14545-P-A. A 5 percent FNH reduction (from 1.70 to 1.61) will be applied to the Framatome HTP fuel during the transition core cycles.

As described in Section 3.9 of this SE, the licensee will apply a 5 percent FNH reduction, as specified in the proposed license condition, which the NRC staff determined to be an acceptable penalty factor to allow the Framatome HTP fuel to be non-limiting. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #8 The plant-specific safety analysis for the steam system piping failure event should be performed with the assumption of loss of offsite power if that is the most conservative case.

Licensee Compliance The licensee stated that the main steam line rupture was analyzed under both cases, with and without offsite power available.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION As noted in Section 15.4.2.1.1 of the UFSAR, the main steam line rupture is performed with and without offsite power. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #9 With regard to the RCS pump shaft seizure accident, the fuel failure criterion should be the 95/95 DNBR limit. The mechanistic method mentioned in WCAP-10444 is not acceptable.

Licensee Compliance The licensee stated that in the locked rotor analysis performed for the Sequoyah RTSR, rods experiencing DNB are assumed to fail with respect to the radiological consequence analysis. The rods-in-DNB were calculated using the VIPRE-W code (WCAP-14565-P-A). The maximum percentage of fuel rods calculated to experience DNB was confirmed to be less than the 10 percent limit used in the radiological dose analysis.

The NRC staff finds the licensee meets this limitation and condition as the pump shaft seizure (locked rotor) accident analysis assumes that any rods experiencing DNB fail. While not part of the limitation and condition, the licensee also states that the existing radiological dose analysis currently uses a conservative number of failed rods.

Limitation and Condition #10 If a positive MTC is intended for VANTAGE 5, the same positive MTC consistent with the plant Technical Specifications should be used in the plant specific safety analysis.

Licensee Compliance The licensee stated that the moderator temperature coefficient (MTC) upper limit in the Technical Specifications, as specified in the COLR, is 0 pcm/°F. However, certain events are analyzed with an MTC of +5 pcm/°F as noted in the markups to UFSAR Chapter 15.

The licensee states that the upper limit in TSs, as specified in the COLR, is 0 pcm/°F. While this meets the limitation and condition, the NRC staff notes that the value in the COLR could change from cycle to cycle. As the licensee noted, certain events are already analyzed with a conservative MTC (+5 pcm/°F). However, since the TS requirement currently requires a non-positive MTC, the staff finds this limitation and condition is met.

Limitation and Condition #11 The LOCA analysis performed for the reference plant with higher FQ of 2.55 has shown that the PCT [peak cladding temperature] limit of 2200°F is violated during transitional mixed core configuration. Plant specific LOCA analysis must be done to show that with the appropriate value of FQ, the 2200°F criterion can be met during use of transitional mixed core.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Licensee Compliance The licensee stated that the FSLOCA evaluation model (EM) (WCAP-16996-P-A Revision 1) analysis performed for the Sequoyah RTSR used an FQ of 2.65 (a surveillance limit of 2.62 will be used until a full core of RFA-2 fuel is loaded) and calculated a small break (Region I) peak cladding temperature of 1213°F and a large break (Region II) PCT of 1878°F. During transitional mixed core configurations with resident HTP fuel, the Westinghouse RFA-2 fuel will not be penalized due to its overall lower loss coefficients. While the FSLOCA EM analysis does not explicitly address the co-resident HTP fuel, and this L&C [limitation and condition] does not therefore directly apply, it is noted that a transition core assessment has determined that the 2200°F criterion continues to be met by the HTP fuel during transition cycles.

Given that the licensee has performed plant specific LOCA analysis and demonstrated that the PCT remains below 2,200°F, the NRC staff finds that the limitation and condition is met. While details are discussed in Section 3.10 of this SE, it should be noted that during the transition cores, the PCT with Framatome methods and HTP fuel is larger than the PCT of the Westinghouse methods and RFA-2 fuel, however, it is still below the 10 CFR 50.46(b)(1) limit of 2,200°F.

Limitation and Condition #12 Our SER on Westinghouses extended burnup topical report WCAP-10125 is not yet complete; the approval of the VANTAGE 5 design for operation to extended burnup levels is contingent on NRC approval of WCAP-10125. However, VANTAGE 5 fuel may be used to those burnups to which Westinghouse fuel is presently operating. Our review of the Westinghouse extended burnup topical report has not identified any safety issues with operation to the burnup value given in the extended burnup report.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. This limit was increased to a lead rod average burnup of 62 GWd/MTU for Westinghouse fuel (WCAP-10444-P-A and WCAP-12610-P-A) provided that PAD 4.0 (WCAP-15063-P-A) or later fuel performance codes were used (NRC letter from J. D. Peralta to Westinghouse dated May 25 2006, Approval for Increase in Licensing Burnup Limit to 62,000 MWD/MTU (TAC NO. MD1486), ADAMS Accession Number ML061420458).

PAD5 and the FSLOCA EM have been approved for use up to a lead rod average burnup of 62 GWd/MTU for all approved types of cladding.

The NRC staff finds this limitation and condition is not applicable as the licensee is not using WCAP-10125. However, the NRC staff reviewed the letter cited in the licensees response and confirmed that the burnup limits were previously approved as noted by the licensee, and therefore, the staff finds the licensee meets the intent of this limitation and condition.

Limitation and Condition #13 Recently, a vibration problem has been reported in a French reactor having 14-foot fuel assemblies; vibration below the fuel assemblies in the lower portion of the reactor vessel OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION is damaging the movable incore instrumentation probe thimbles. The staff is currently evaluating the implications of this problem to other cores having 14 foot long fuel bundle assemblies. Any limitations to the 14-foot core design resulting from the staff evaluation must be addressed in plant-specific evaluations Licensee Compliance The licensee stated that this limitation and condition is not applicable to this license amendment request. The RFA-2 fuel assemblies to be loaded at Sequoyah are of the standard 12-foot length.

The NRC staff finds that this limitation and condition is not applicable as SQN, Units 1 and 2 do not use 14-foot fuel assemblies.

3.8.5 WCAP-10444-P-A Addendum 2-A Limitation and Condition The WCAP-10444-P-A Addendum 2 provides an acceptable method for the application of the WCAP-10444-P-A information in the use of the VANTAGE 5H fuel assemblies in complete and transition core configurations.

For transition cores, the transition core configuration penalty specified in WCAP-10444-P-A will apply for the estimation of the peak clad temperature in large LOCA analyses.

Licensee Compliance The licensee stated that the Sequoyah Units 1 and 2 analysis with the FSLOCA EM (WCAP-16996-P-A Revision 1) was performed assuming a full core of Westinghouse RFA-2 fuel. For the initial cycles in which Westinghouse RFA-2 fuel is used, however, Framatome high thermal performance (HTP) fuel will also be present. While the two fuel designs have generally similar mechanical designs, a transition core evaluation was performed to address the mixed core effects for both fuel types.

The loss coefficient of the Westinghouse RFA-2 fuel is slightly lower than the Framatome HTP fuel, and the analysis with the FSLOCA EM assuming a homogeneous core of RFA-2 fuel is bounding of the mixed core effect for Westinghouse fuel, because the RFA-2 fuel would receive a flow benefit in the presence of the relatively starved HTP fuel.

An evaluation was therefore performed to assess the PCT effect of the RFA-2 on the Framatome HTP fuel within the context of the existing analysis of record (AOR) supporting operation with that fuel type.

For SBLOCA transients, core-wide collapsed liquid levels correspond closely to a 1-dimensional (1-D) flow pattern, and the effects of grid loss coefficient differences among the assemblies are not significant in determining the PCT. As such, the existing OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION analysis of record supporting operation with Framatome HTP fuel is applicable for the Framatome HTP fuel during the transition cycle(s) to Westinghouse RFA-2 fuel.

For LBLOCA transients, conditions during blowdown and reflood can be affected by mixed core conditions arising from a hydraulic mismatch. The existing AOR performed with the best estimate (BE) AREVA/Framatome RLBLOCA methodology resulted in a PCT occurring around 265 seconds after the postulated break during the reflood period.

The PCT increase for Framatome HTP fuel resulting from the hydraulic mismatch was estimated to be 23°F, based on the expected effects on a transient with the reflood time and cladding heatup rate consistent with the Sequoyah Units 1 and 2 RLBLOCA analysis. The PCT estimate of effect of 23°F is applicable for the Framatome HTP fuel during transition cycles to Westinghouse RFA-2 fuel.

The transition core PCT penalty discussed in Sections 1.2.d, 5.2.1, and 5.2.3 of WCAP-10444-P-A is not applicable to the HTP fuel or the use of the FSLOCA EM in this license amendment request. Even if that obsolete transition core PCT penalty were to be applied, it would be insignificant in comparison to the 322°F margin to the 2200°F regulatory limit. See also the discussion of Limitation and Condition #11 under WCAP-10444-P-A.

The licensees application of the FSLOCA methodology is discussed and found acceptable in Section 3.10 of this SE. The licensee clarified that the 23°F penalty is only applicable to the HTP fuel as a change to the existing Framatome AOR, which continues to support the HTP fuel during the transition cores. In addition, the 23°F PCT penalty is not applicable to the FSLOCA EM analysis results, as the overall lower loss coefficients of the Westinghouse RFA-2 fuel would not lead to a PCT penalty for the RFA-2 fuel. The NRC staff finds that this limitation and condition is met as an appropriate transition core configuration penalty is applied for the estimation of the peak clad temperature in the LBLOCA analyses.

3.8.6 WCAP-10965-P-A Limitation and Condition The ANC code provides an accurate calculation of core reactivity, reactivity coefficients, critical boron, rod worths and core power distribution for use in design and safety analyses. The qualification presented in WCAP-10965 demonstrates that the accuracy of the ANC prediction of these quantities is generally comparable to that of TORTISE.

Licensee Compliance The licensee stated that the ANC code is used in accordance with the approved topical report for the Sequoyah RTSR. This topical report is discussed in Sequoyah UFSAR Sections 4.3.2.3.1 and 4.3.3.3 (cited as Reference 31 - SQN UFSAR will be updated to the approved September 1986 reference). See also the PARAGON and NEXUS/ANC discussions under WCAP-16045-P-A and WCAP-16045-P-A Addendum 1-A.

The NRC staff finds that the LAR incorrectly concludes that the technical position in the NRC SE that approved WCAP-10965-P-A is a limitation and condition. Because the staff finds that the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION licensee response is consistent with the technical position, the use of the topical report is acceptable.

3.8.7 WCAP-10965-P-A, Addendum 2-A, Revision 0 Limitation and Condition Westinghouse has provided a series of simulations to demonstrate the performance of the ANC code with the new pin power recovery methodology for control rod histories that exacerbate heterogeneous environments. These calculations confirm that the new pin power methodology is accurate for both unrodded and rodded pin power predictions.

Pin power histories computed with the ANC code, as presented in this safety evaluation, are dependent on the results of the NRC approved codes PARAGON (Reference 4; see discussion of WCAP-16045-P-A) and NEXUS (Reference 5; see discussion of WCAP-16045-P-A Addendum 1-A). Thus, the use of ANC with the new pin power recovery methodology described in the WCAP-10965-P-A, Addendum 2 /

WCAP-10966-A, Addendum 2, TR in licensing applications requires the concomitant application of the NRC approved lattice code PARAGON and the NRC approved cross section parameterization and reconstruction methodology of the NEXUS code system.

The NRC staff has reviewed the TR submitted by Westinghouse and determined that WCAP-10965-P, Addendum 2/WCAP-10966-A, Addendum 2, is an adequate enhancement to replace the pin power recovery methodologies of NRC-approved methodologies WCAP-10965-P-A, and where appropriate of WCAP-10965-P-A, Addendum 1.

Licensee Compliance The licensee stated that the pin power recovery method follows WCAP-10965-P-A Addendum 2-A in concert with NEXUS/ANC which uses the PARAGON lattice code.

Because the licensee uses the pin power recovery methodology in WCAP-10965-P-A Addendum 2-A in concert with NEXUS/ANC, which uses the PARAGON lattice code, the NRC staff finds the licensee meets this limitation and condition.

3.8.8 WCAP-11397-P-A Limitation and Condition #1 Sensitivity factors used for a particular plant and their ranges of applicability should be included in the Safety Analysis Report or reload submittal.

Licensee Compliance The licensee stated that sensitivity factors were calculated using the WRB-2M, ABB-NV, and WLOP DNB correlations and the VIPRE-W code for parameter values applicable to the RFA-2 fuel transition at Sequoyah. These sensitivity factors were used to determine the RTDP design limit DNBR values for the DNB correlations. The safety analysis DNBR limit is 1.58 for the Sequoyah RTSR DNB analyses. The DNBR design limit value OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION is 1.23 for both typical and thimble cells with the WRB-2M correlation for RFA-2 fuel.

The DNBR design limit values for the ABB-NV correlation are 1.18 for typical cells and 1.19 for thimble cells below the first mixing vane region of the RFA-2 fuel.

The NRC staff finds the licensee meets this limitation and condition because the licensee calculated the appropriate sensitivity factors for the DNB correlations. The sensitivity factors are required to be included in the Safety Analysis Report when updated per 10 CFR 50.71(e).

Limitation and Condition #2 Any changes in DNB correlation, THINC-IV correlations, or parameter values listed in Table 3-1 of WCAP-11397 outside of previously demonstrated acceptable ranges require re-evaluation of the sensitivity factors and of the use of Equation (2-3) of the topical report.

Licensee Compliance The licensee stated that because the VIPRE-W code is used to replace the THINC-IV code for the Sequoyah RTSR, sensitivity factors for the RTDP methodology were calculated using the VIPRE-W code for parameter values applicable to the RFA-2 fuel transition at Sequoyah, as discussed in the response to Limitation and Condition #1 above. See the Response to Limitation and Condition #3 for a discussion of the use of Equation (2-3) of the topical report.

The NRC staff finds that the licensee meets this limitation and condition because it appropriately calculated the sensitivity factors for the DNB correlations.

Limitation and Condition #3 If the sensitivity factors are changed as a result of correlation changes or changes in the application or use of the THINC code, then the use of an uncertainty allowance for application of Equation (2-3) must be re-evaluated and the linearity assumption made to obtain Equation (2-17) of the topical report must be validated.

Licensee Compliance The licensee stated that, as described in WCAP-14565-P-A, the VIPRE-W code has been demonstrated to be equivalent to the THINC code. Equation (2-3) of WCAP-11397-P-A and the linearity approximation made to obtain Equation (2-17) were confirmed to be valid for the Sequoyah RTSR for the combination of the WRB-2M correlation and the VIPRE-W code as well as for the combination of the ABB-NV and WLOP correlations and the VIPRE-W code.

Given that the VIPRE-W code has previously been approved as a replacement for the THINC code and the licensees statement that Equation (2-3) and the linearity approximation made to obtain Equation (2-17) of WCAP-11397-P-A were confirmed to be valid, the NRC staff finds that the licensee meets this limitation and condition.

Limitation and Condition #4 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Variances and distributions for input parameters must be justified on a plant-by-plant basis until generic approval is obtained.

Licensee Compliance The licensee stated that there were no changes to the operating parameter uncertainties for the Sequoyah RTSR.

Given that WCAP-11397-P-A is referenced as the design method (RTDP) used to meet the DNB design basis in the current SQN, Units 1 and 2 licensing basis (Reference 95 in UFSAR Section 4.4.6) and there were no changes to the parameters, the NRC staff finds that this license condition is met.

Limitation and Condition #5 Nominal initial condition assumptions apply only to DNBR analyses using RTDP. Other analyses, such as overpressure calculations, require the appropriate conservative initial condition assumptions.

Licensee Compliance The licensee stated that for the Sequoyah RTSR, nominal initial conditions were only applied to DNBR calculations that used the RTDP.

A review of UFSAR Chapter 15 and Attachment 10, Sequoyah Safety Analysis UFSAR Impact Summary for the WEC RFA-2 Fuel Transition, to Enclosure 1 of the LAR shows that conservative initial conditions are used. Given that the licensee only used nominal initial condition for the DNBR calculations that used RTDP, the NRC staff finds the licensee meets this limitation and condition.

Limitation and Condition #6 Nominal conditions chosen for use in analyses should bound all permitted methods of plant operation.

Licensee Compliance The licensee stated that the nominal conditions used in the Sequoyah RTSR are discussed under Limitation and Condition #2 of WCAP-14565-P-A. The continued applicability of the bounding input assumptions will be verified on a cycle-specific basis using the Westinghouse reload methodology described in WCAP-9272-P-A.

The NRC staff finds the licensee meets this limitation and condition as the bounding input assumptions are verified on a cycle-specific basis.

Limitation and Condition #7 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The code uncertainties specified in Table 3-1 (+/- 4 percent for THINC-IV and +/- 1 percent for transients) must be included in the DNBR analyses using RTDP.

Licensee Compliance The licensee stated that the code uncertainties specified in Table 3-1 of WCAP-11397-P-A remain unchanged and were included in the DNBR analyses using RTDP. The THINC-IV uncertainty was applied to VIPRE-W, based on the equivalence of the VIPRE-W model approved in WCAP-14565-P-A to THINC-IV.

The SQN DNBR analysis used the code uncertainties specified in Table 3-1 of WCAP-11397-P-A and WCAP-14565-P-A showed that the validation results, including uncertainty quantification, for VIPRE-W yielded equivalent results as THINC-IV. Therefore, the uncertainties remain applicable, and the NRC staff finds this limitation and condition is met.

3.8.9 WCAP-12610-P-A Note that the -A version of this topical report includes three separate safety evaluations.

Limitations and conditions for each are described below.

Limitation and Condition #1 (from July 1, 1991, NRC SE)

The staff has reviewed the Westinghouse VANTAGE+ fuel design and mechanical analyses report described in Reference 1 and find it acceptable for licensing applications up to a rod-average burnup level of 60 MWd/KgM. Our findings on applications beyond a rod-average burnup of 60 MWd/kgM which is addressed in Appendix B of Reference 1, will be reported separately.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. This limit was increased to a lead rod average burnup of 62 GWd/MTU for Westinghouse fuel (WCAP-10444-P-A and WCAP-12610-P-A) provided that PAD 4.0 (WCAP-15063-P-A) or later fuel performance codes were used (NRC letter from J. D. Peralta to Westinghouse dated May 25 2006, Approval for Increase in Licensing Burnup Limit to 62,000 MWD/MTU (TAC NO. MD1486), ADAMS Accession Number ML061420458).

PAD5 and the FSLOCA EM have been approved for use up to a lead rod average burnup of 62 GWd/MTU for all approved types of cladding.

The NRC staff finds that this limitation and condition no longer applies because the staff approved a burnup level increase to 62 GWd/MTU after the approval of WCAP-12610-P-A. The staff finds that the licensees response meets the intent of the limitation and condition by providing information that shows the licensee will use methods that are approved for the accepted burnup value.

Limitation and Condition #2 (from July 1, 1991, NRC SE)

This approval is subject to confirmation of expected fuel performance in the fuel surveillance program described in Section 6.0 [PNL TER] of the attached evaluation. It OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION is expected that any adverse surveillance information with respect to ZIRLO corrosion or hydrogen pickup or other properties considered in the design evaluation will be reported to the NRC.

Licensee Compliance The licensee stated to see the discussion of Limitation and Condition #3 under WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A and Sections 7.4.5 and 7.4.6 of the PAD5 topical report (WCAP-17642-P-A).

The licensees discussion on Limitation and Condition #3 of WCAP-12610-P-A &

CENPD-404-P-A, Addendum 1-A states that the maximum fuel rod waterside corrosion and hydrogen pickup will be limited to the appropriate criteria and that the methodologies used for the normal reload process under WCAP-9272-P-A will ensure the limits are maintained. While Sections 7.4.5 and 7.4.6 of WCAP-17642-P-A discuss clad oxidation and clad hydrogen pickup, Section 3.3 of WCAP-17642-P-A discusses the cladding corrosion models and presents comparisons to data to demonstrate confirmation of expected fuel performance. Given these models have been approved by the NRC, the NRC staff finds this limitation and condition is met.

Limitation and Condition #3 (from July 1, 1991, NRC SE)

This approval does not include the LOCA evaluation methods described in Appendices F and G of WCAP-12610.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The LOCA analysis for Sequoyah Units 1 and 2 was performed in accordance with the FSLOCA EM topical report WCAP-16996-P-A Revision 1).

Given that the licensee is not using the LOCA evaluation methods described in Appendices F and G of WCAP-12610, the NRC staff finds that this limitation and condition is not applicable.

Discussion of the licensees use of FSLOCA is in Section 3.10 of this SE.

Limitation and Condition #1 (from October 9, 1991, NRC SE, Enclosure 1)

We find the modifications to the use of the Westinghouse NOTRUMP/LOCTA-IV small break evaluation model to analyze VANTAGE+ fuel with ZIRLO cladding and thimble tubes in conformance with the requirements of 10 CFR Part 50, Appendix K and therefore acceptable. The limitations and conditions identified in previous SERs for the NOTRUMP/LOCTA-IV small break LOCA model continue to apply to this usage of the model.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The small break LOCA analysis for Sequoyah Units 1 and 2 was performed in accordance with the FSLOCA EM topical report WCAP-16996-P-A Revision 1).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Given that the licensee is not using NOTRUMP/LOCTA-IV for the small break LOCA analysis, the NRC staff finds that this limitation and condition is not applicable. Discussion of the licensees use of FSLOCA is in Section 3.10 of this SE.

Limitation and Condition #2 (from October 9, 1991, NRC SE, Enclosure 1)

Although ZIRLO is similar to ZIRCALOY, the criteria of acceptance (10 CFR 50.44, 10 CFR 50.46 and 10 CFR 50, Appendix K) cited in the evaluation are specifically identified as appropriate for ZIRCALOY clad fuel. Thus, exemptions must be obtained to allow application of those criteria to ZlRLO-clad fuel.

Licensee Compliance The licensee stated that an exemption request is included with the Sequoyah license amendment request (see Enclosure 5).

The staff finds that the licensee meets this limitation and condition to the extent applicable. On September 16, 2003, the NRC revised 10 CFR 50.44 (68 FR 54141), eliminating text that specified various types of fuel cladding, therefore, an exemption to 10 CFR 50.44 is no longer required. The LAR requested an exemption from 10 CFR 50.46 and Part 50, Appendix K related to the use of Optimized ZIRLO, which the NRC staff processed separately and granted on October 26, 2021 (ADAMS Accession No. ML21166A166).

Limitation and Condition #1 (from October 9, 1991, NRC SE, Enclosure 2)

The staff finds that WCAP-12610, Appendix G, describes LOCA analyses performed with acceptable methods and that results of these analyses demonstrate conformance with the criteria given in 10 CFR 50.46 and 10 CFR Part 50 Appendix K, for a reference plant. We find the criteria applicable and the analyses acceptable. This SER addresses only the reference plant and fuel designs discussed herein. Other applications must be justified.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The LOCA analysis for Sequoyah Units 1 and 2 was performed in accordance with the FSLOCA EM topical report (WCAP-16996-P-A Revision 1).

Given that the licensee is not using the LOCA methodology described in Appendix G to WCAP-12610, the NRC staff finds that this limitation and condition is not applicable. Discussion of the licensees use of FSLOCA is in Section 3.10 of this SE.

Limitation and Condition #2 (from October 9, 1991, NRC SE, Enclosure 2)

Although ZIRLO is similar to Zircaloy, the criteria of acceptance (10 CFR 50.46 and 10 CFR 50, Appendix K) cited in the evaluation are specifically identified as appropriate for Zircaloy-clad fuel. Thus, exemptions must be obtained to allow application of those criteria to ZIRLO-clad fuel. Similarly, exemptions must be obtained to allow application OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION of 10 CFR 50.44 dealing with hydrogen generation and combustible gas control to plants with ZIRLO-clad fuel.

Licensee Compliance The licensee stated that an exemption request is included with the Sequoyah license amendment request (see Enclosure 5).

The licensee meets this limitation and condition to the extent applicable. As noted above, an exemption from 10 CFR 50.44 is no longer required due to a 2003 rulemaking (68 FR 54141).

The NRC granted exemptions from 10 CFR 50.46 and Part 50, Appendix K.I.5, on October 26, 2021 (ADAMS Accession No. ML21166A166).

Limitation and Condition (from September 15, 1994, NRC SE)

The staff has reviewed the W fuel rod growth model described in WCAP-12610, Appendix B, Addendum 1 and finds it acceptable for licensing applications up to 60,000 MWD/MTU rod average.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. This limit was increased to a lead rod average burnup of 62 GWd/MTU for Westinghouse fuel (WCAP-10444-P-A and WCAP-12610-P-A) provided that PAD 4.0 (WCAP-15063-P-A) or later fuel performance codes were used (NRC letter from J. D. Peralta to Westinghouse dated May 25 2006, Approval for Increase in Licensing Burnup Limit to 62,000 MWD/MTU (TAC NO. MD1486), ADAMS Accession Number ML061420458).

PAD5 and the FSLOCA EM have been approved for use up to a lead rod average burnup of 62 GWd/MTU for all approved types of cladding.

The NRC staff finds that this limitation and condition no longer applies because the staff approved a burnup level increase to 62 GWd/MTU after the approval of WCAP-12610-P-A. The staff finds that the licensees response meets the intent of the limitation and condition by providing information that shows the licensee will use methods that are approved for the accepted burnup value.

3.8.10 WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A Limitation and Condition #1 Until rulemaking to 10 CFR Part 50 addressing Optimized ZIRLOTM has been completed, implementation of Optimized ZIRLOTM fuel clad requires an exemption from 10 CFR 50.46 and 10 CFR Part 50 Appendix K.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Licensee Compliance The licensee stated that an exemption request is included in the Sequoyah RTSR license amendment submittal (see Enclosure 5).

The licensee included an exemption request in the LAR and the staff processed the exemption separately. Exemptions from 10 CFR 50.46 and Part 50, Appendix K.I.5, were granted on October 26, 2021 (ADAMS Accession No. ML21166A166). Thus, the NRC staff finds that this limitation and condition satisfied.

Limitation and Condition #2 Revised by NRC letter from D. C. Morey to Westinghouse dated May 8, 2019, Modification of the Condition and Limitation 2 of the U.S. Nuclear Regulatory Commission Safety Evaluation Report for Westinghouse Electric Company Topical Report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (EPID L-2019-TOP-0004), ADAMS Accession Number ML19119A127, to read:

For Westinghouse fuel designs, the fuel rod average burnup limit for this approval remains at the currently established limit of 62 GWd/MTU. For the CE16NGF assembly design, the fuel rod average burnup limit for this approval is 62 GWd/MTU when analyzed with NRC approved analytical methods which properly accounts for TCD. This includes licensees that use penalties and allowances to account for TCD that have been explicitly approved by the NRC. For CE16NGF without NRC approved analytical methods which properly accounts for TCD and all remaining CE fuel designs, the fuel rod average burnup limit for this approval remains at the currently established limit of 60 GWd/MTU.

Licensee Compliance The licensee stated that the lead rod average burnup limit for this license amendment request is 62 GWd/MTU. Sequoyah will transition to Westinghouse RFA-2 fuel.

Because the licensee is using an approved burnup limit of 62 GWd/MTU with Westinghouse RFA-2 fuel and does not propose to use CE16NGF fuel, the NRC staff finds that the limitation and condition is not applicable. The NRC staff also notes that the licensee is using the PAD5 code, which properly accounts for TCD.

Limitation and Condition #3 The maximum fuel rod waterside corrosion, as predicted by the best-estimate model, will

[satisfy proprietary limits (included in the topical report and proprietary version of the NRC Safety Evaluation)] of hydrides for all locations of the fuel rod.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Licensee Compliance The licensee stated that the maximum fuel rod waterside corrosion and hydrogen pickup will be limited to the criteria specified in Limitation and Condition #3. The best estimate maximum predicted oxide thickness will remain below 100 microns. Hydrogen pickup will be limited to the proprietary limit approved in the topical report. The methodologies used for the normal reload process under WCAP-9272-P-A will ensure that the maximum fuel rod waterside corrosion limits will be confirmed to be less than the specified limits for all fuel rod locations as a normal part of the reload design process.

The corrosion and hydrogen pickup are calculated using the methodology in WCAP-17642-P-A.

Given the use of an approved methodology and following the reload process under WCAP-9272-P-A will ensure the results satisfy the acceptance criteria. Therefore, the NRC staff finds that the licensee meets this limitation and condition.

Limitation and Condition #4 All the conditions listed in previous NRC SE approvals for methodologies used for standard ZIRLO and Zircaloy-4 fuel analysis will continue to be met, except that the use of Optimized ZIRLOTM cladding in addition to standard ZIRLO and Zircaloy-4 cladding is now approved.

Licensee Compliance The licensee stated that all conditions for previously approved cladding-related methodologies, listed in the proposed changes to Sequoyah Technical Specification 5.6.3.b, will continue to be met and confirmed as part of the normal reload design process discussed in WCAP-9272-P-A. See also the discussion of the Limitations and Conditions for WCAP-12610-P-A.

The NRC staff finds that the licensee meets all limitations and conditions for the previously approved cladding-related methodologies and these will be confirmed as part of the normal reload design process. Therefore, this limitation and condition is met.

Limitation and Condition #5 All methodologies will be used only within the range for which ZIRLOTM and Optimized ZIRLOTM data were acceptable and for which the verifications discussed in Addendum 1 and responses to RAls were performed.

Licensee Compliance The licensee stated that all conditions for previously approved cladding-related methodologies, listed in the proposed changes to Sequoyah Technical Specification 5.6.3.b, will continue to be met within the range of acceptable data and confirmed as part of the normal reload design process discussed in WCAP-9272-P-A. See also the discussion of the Limitations and Conditions for WCAP-12610-P-A.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff finds that following the normal reload design process, required by revised TS 5.6.3.b and in WCAP-9272-P-A, will confirm that the conditions are met for all previously approved cladding-related methodologies, therefore, this limitation and condition is met.

Limitation and Condition #6 The licensee is required to ensure that Westinghouse has fulfilled the following commitment:

Westinghouse shall provide the NRC staff with a letter(s) containing the following information (Based on the schedule described in response to RAI #3 [Reference 3]

[Letter from J. A. Gresham (Westinghouse) to U.S. Nuclear Regulatory Commission, Westinghouse Responses to NRC Request for Additional Information (RAls) on Optimized ZIRLOTM Topical - Addendum 1 to WCAP-12610-P-A, LTR-NRC-04-44, August 4, 2004 (ADAMS Accession No. ML042240408)].

Confirmation of the approved models applicability up through the projected end of cycle burnup for the Optimized ZIRLOTM fuel rods must be completed prior to their initial batch loading and prior to the startup of subsequent cycles. For example, prior to the first batch application of Optimized ZIRLOTM, sufficient LTA data may only be available to confirm the models applicability up through 45 GWd/MTU. In this example, the licensee would need to confirm the models up through the end of the initial cycle. Subsequently, the licensee would need to confirm the models, based upon the latest LTA data, prior to re-inserting the Optimized ZIRLOTM fuel rods in future cycles. Based upon the LTA schedule, it is expected that this issue may only be applicable to the first few batch implementations since sufficient LTA data up through the burnup limit should be available within a few years.

Licensee Compliance The licensee stated that limitation and Condition 6 has been satisfied by letter from NRC (Kevin Hsueh) to Westinghouse (J. A. Gresham), Satisfaction of Conditions 6 and 7 of the U.S. Nuclear Regulatory Commission Safety Evaluation for Westinghouse Electric Company Addendum 1 to WCAP-12610-P-A & CENPD-404-P-A, OPTIMIZED ZIRLOTM, Topical Report, dated August 3, 2016, ADAMS Accession Number ML16173A354.

The NRC staff finds that the licensee meets this limitation and condition because the letter cited by the licensee indicates that Westinghouse previously provided the requested information to the NRC.

Limitation and Condition #7 The licensee is required to ensure that Westinghouse has fulfilled the following commitment.

Westinghouse shall provide the NRC staff with a letter containing the following information (Based on the schedule described in response to RAI #11 [Reference 3]):

a. Vogtle growth and creep data summary reports.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

b. Using the standard ZIRLOTM and Optimized ZIRLOTM database including the most recent Vogtle data, confirm applicability with currently approved fuel performance models (e.g., level of conservatism in W rod pressure analysis, measured vs. predicted, predicted minus measured vs. tensile and compressive stress).

Confirmation of the approved models applicability up through the projected end of cycle burnup for the Optimized ZIRLOTM fuel rods must be completed prior to their initial batch loading and prior to the startup of subsequent cycles. For example, prior to the first batch application of Optimized ZIRLOTM, sufficient LTA data may only be available to confirm the models applicability up through 45 GWd/MTU. In this example, the licensee would need to confirm the models up through the end of the initial cycle. Subsequently, the licensee would need to confirm the models, based upon the latest LTA data, prior to re-inserting the Optimized ZIRLOTM fuel rods in future cycles. Based upon the LTA schedule, it is expected that this issue may only be applicable to the first few batch implementations since sufficient LTA data up through the burnup limit should be available within a few years.

Licensee Compliance The licensee stated that limitation and Condition 7 has been satisfied by letter from NRC (Kevin Hsueh) to Westinghouse (J. A. Gresham), Satisfaction of Conditions 6 and 7 of the U.S. Nuclear Regulatory Commission Safety Evaluation for Westinghouse Electric Company Addendum 1 to WCAP-12610-P-A & CENPD-404-P-A, OPTIMIZED ZIRLOTM, Topical Report, dated August 3, 2016, ADAMS Accession Number ML16173A354.

The NRC staff finds that the licensee meets this limitation and condition because the letter cited by the licensee indicates that Westinghouse previously provided the requested information to the NRC.

Limitation and Condition #8 The licensee shall account for the relative differences in unirradiated strength (YS and UTS) between Optimized ZIRLOTM and standard ZIRLOTM in cladding and structural analyses until irradiated data for Optimized ZIRLOTM have been collected and provided to the NRC staff.

a. For the Westinghouse fuel design analyses:

i. The measured, unirradiated Optimized ZIRLOTM strengths shall be used for

[beginning of cycle life] BOL analyses.

ii. Between BOL up to a radiation fluence of 3.0 x 1021 n/cm2 (E> 1 MeV), pseudo-irradiated Optimized ZIRLOTM strength set equal to linear interpolation between the following two strength level points: At zero fluence, strength of Optimized ZIRLOTM equal to measured strength of Optimized ZIRLOTM and at a fluence of 3.0 x 1021 n/cm2 (E>1 MeV), irradiated strength of standard ZIRLOTM at the fluence of 3.0 x 1021 n/cm2 (E>1 MeV) minus 3 ksi.

iii. During subsequent irradiation from 3.0 x 1021 n/cm2 up to 12 x 1021 n/cm2, the differences in strength (the difference at a fluence of 3 x 1021 n/cm2 due to tin content) shall be decreased linearly such that the pseudo-irradiated Optimized OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION ZIRLOTM strengths will saturate at the same properties as standard ZIRLOTM at 12 x 1021 n/cm2.

b. For the CE fuel design analyses, the measured, unirradiated Optimized ZIRLOTM strengths shall be used for all fluence levels (consistent with previously approved methods).

Licensee Compliance The licensee stated that limitation and Condition 8.a was addressed in Westinghouse letter LTR-NRC-15-84, Submittal of Response to Condition 8.a of the Safety Evaluation Report (SER) on WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A, Optimized ZIRLOTM, (Proprietary/Non-Proprietary), dated September 29, 2015, ADAMS Accession Number ML15279A113. Based on that Westinghouse letter, the following conclusions were drawn:

Therefore, the current Condition 8.a requirement is not consistent with the data.

Instead, based on the data, the following model is proposed for Westinghouse fuel design analyses:

i. The measured, un-irradiated Optimized ZIRLOTM cladding strength shall be used for BOL analyses.

ii. Between the beginning of life up to the proprietary fluence value listed in Section 6 of the attachment to LTR-NRC-15-84, cladding strength will be modeled as described in that letter for Westinghouse fuel design analysis.

iii. During subsequent irradiation beyond that level, the irradiated Optimized ZIRLOTM material strengths are the same as the designed irradiated standard ZIRLO cladding strengths.

The fuel analysis of Optimized ZIRLOTM clad rods will be confirmed to meet the requirements of LTR-NRC-15-84 as part of the normal reload design process discussed in WCAP-9272-P-A.

Limitation and Condition 8.b is not applicable to the Sequoyah units given the proposed loading of transition cores with Westinghouse RFA-2 fuel.

The NRC staff reviewed LTR-NRC-15-84 and finds that Westinghouse previously addressed this limitation and condition by providing generic information that satisfies 8.a, above. The licensee stated that it will confirm the appropriate information as part of the normal reload design process. The staff also finds that because SQN does not use a Combustion Engineering fuel design, that 8.b does not apply. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #9 As discussed in response to RAI #21 (Reference 3), for plants introducing Optimized ZIRLOTM that are licensed with LOCBART or STRIKIN-II and have a limiting PCT that occurs during blowdown or early reflood, the limiting LOCBART or STRIKIN-II calculation will be rerun using the specified Optimized ZIRLOTM material properties.

Although not a condition of approval, the NRC staff strongly recommends that, for future OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION evaluations, Westinghouse update all computer models with Optimized ZIRLOTM specific material properties.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The current licensing basis for the Sequoyah units does not include LOCBART or STRIKIN-II. The LOCA analyses for the Sequoyah RTSR were performed using the FSLOCA EM (see the discussion of WCAP-16996-P-A Revision 1).

Because the licensee is not using the LOCBART of STRIKIN-II codes for LOCA analysis, the NRC staff finds this limitation and condition inapplicable.

Limitation and Condition #10 Due to the absence of high temperature oxidation data for Optimized ZIRLOTM, the Westinghouse coolability limit on PCT during the locked rotor event shall be [less than the proprietary limit included in the proprietary versions of the topical report and NRC Safety Evaluation].

Licensee Compliance The licensee stated that for the implementation of Optimized ZIRLOTM fuel cladding, the PCT calculated for the locked rotor event is 1852°F, which is considerably less than the limit specified in Limitation and Condition #10. For subsequent core reload designs, the calculated PCT will be confirmed as part of the normal reload design process conducted per WCAP-9272-P-A.

Because the calculated PCT of 1,852°F for the locked rotor event is considerably less than the proprietary limit given in the limitation and condition, and the licensee will confirm the calculated PCT is less than the proprietary limit as part of the reload design process, the NRC staff finds that this limitation is met.

3.8.11 WCAP-14565-P-A Limitation and Condition #1 Selection of the appropriate CHF correlation, DNBR limit, engineered hot channel factors for enthalpy rise and other fuel-dependent parameters for a specific plant application should be justified with each submittal.

Licensee Compliance The licensee stated that the WRB-2M correlation with a 95/95 correlation limit of 1.14 approved in WCAP-15025-P-A was used in the VIPRE-W DNBR analyses for at-power events and for analyses applicable to the region above the first mixing vane grid.

VIPRE-W is the Westinghouse version of EPRIs VIPER-01 code with NRC-approved input options, heat transfer models, and proprietary critical heat flux correlations added.

The ABB-NV and WLOP DNB correlations (approved in WCAP-14565-P-A Addenda 1-A OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION and 2-P-A) were used when the WRB-2M DNB correlation is not applicable. The ABB-NV and WLOP DNB correlation limits used in the DNBR calculations are 1.13 and 1.18, respectively. The Sequoyah RTSR utilizes the Westinghouse 17x17 RFA-2 fuel design, starting with a mixed core of Framatome 17x17 HTP fuel and Westinghouse RFA-2 fuel.

The effect of mixed cores on departure from nucleate boiling (DNB) analyses is discussed above under WCAP-11837-P-A. The HTP fuel utilizes a lower FH to preclude the need to penalize the DNB which maintains the current safety analysis. The Westinghouse fuel utilizes a direct DNB penalty applied by the thermal-hydraulic (T/H) design group. As a result, the FH technical specification will be modified and the Framatome HTP fuel will be monitored to a more restrictive value of FH than the Westinghouse RFA-2 fuel.

The NRC staff finds that the licensee selected appropriate DNB correlations, limits, etc., that have been approved for the Westinghouse RFA-2 fuel and therefore, that this limitation and condition is met. Additional discussion of the DNB correlation is in Section 3.1 of this SE.

Limitation and Condition #2 Reactor core boundary conditions determined using other computer codes are generally input into VIPRE for reactor transient analyses. These inputs include core inlet coolant flow and enthalpy, core average power, power shape and nuclear peaking factors.

These inputs should be justified as conservative for each use of VIPRE.

Licensee Compliance The licensee stated that the core boundary conditions used in VIPRE-W DNBR calculations for the RFA-2 fuel transition are generated from NRC-approved codes and analysis methodologies. Table 7.1-1 [of attachment 8 of the LAR] summarizes the thermal-hydraulic (T/H) design parameters for Sequoyah Units 1 and 2 that were used in this analysis. The T/H design parameters remain the same as those presented in the Sequoyah Units 1 and 2 Updated Final Safety Analysis Report, with the following exceptions:

Thermal Design Flow is increased from 87,000 gpm/loop to 90,000 gpm/loop.

A 5 percent FNH reduction (from 1.70 to 1.61) will be applied to the Framatome HTP fuel during the transition core cycles.

The FQ surveillance limit will be reduced from 2.65 to 2.62 until the resident Framatome HTP fuel has been fully replaced with Westinghouse RFA-2 fuel.

All current UFSAR T/H design criteria are satisfied. Continued applicability of the core boundary conditions as VIPRE-W input will be verified on a cycle-by-cycle basis using the reload methodology described in WCAP-9272-P-A.

As stated by the licensee, the core boundary conditions used in VIPRE-W DNBR calculations are generated from NRC-approved codes and analysis methodologies and the TH design parameters, with the noted exceptions, are consistent with the current licensing basis. The three exceptions are described elsewhere in this SE as follows: thermal design flow in OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Section 3.5, FNH in Section 3.9, and FQ in Section 3.8.4. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #3 The NRC staffs generic SER for VIPRE (Ref. 2) [Safety Evaluation Report on EPRI NP-2511-CCM VIPRE-01] set requirements for use of new CHF correlations with VIPRE.

Westinghouse has met these requirements for using the WRB-1, WRB-2 and WRB-2M correlations. The DNBR limit for WRB-1 and WRB-2 is 1.17. The WRB-2M correlation has a DNBR limit of 1.14. Use of other CHF correlations not currently included in VIPRE will require additional justification.

Licensee Compliance The licensee stated that as discussed in the response to Limitation and Condition #1, the WRB-2M correlation with a 95/95 correlation limit of 1.14 (approved in WCAP-15025-P-A) was used in the VIPRE-W DNBR calculations for the RFA-2 fuel transition at Sequoyah Units 1 and 2. The ABB-NV DNBR limit and the WLOP DNBR limit were previously approved in WCAP-14565-P-A Addenda 1-A and 2-P-A for use with the VIPRE-W code.

Given that the licensee is using the approved correlation and limits, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #4 Westinghouse proposes to use the VIPRE code to evaluate fuel performance following postulated design-basis accidents, including beyond-CHF heat transfer conditions.

These evaluations are necessary to evaluate the extent of core damage and to ensure that the core maintains a coolable geometry in the evaluation of certain accident scenarios. The NRC staffs generic review of VIPRE (Ref. 2) did not extend to post CHF calculations. VIPRE does not model the time-dependent physical changes that may occur within the fuel rods at elevated temperatures. Westinghouse proposes to use conservative input in order to account for these effects. The NRC staff requires that appropriate justification be submitted with each usage of VIPRE in the post-CHF region to ensure that conservative results are obtained.

Licensee Compliance The licensee stated that application of the VIPRE-W code does not model the time-dependent physical changes that may occur within the fuel rods at elevated temperatures in the post-CHF (critical heat flux) region.

Given that the licensee is not proposing to use VIPRE-W to model the time-dependent physical changes that may occur in the post-CHF (critical heat flux) region, the NRC staff finds that this limitation and condition not applicable.

3.8.12 WCAP-14565-P-A, Addendum 1-A, Revision 0 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Limitation and Condition #1 Addendum 1 to the WCAP-14565-P-A VIPRE-01 model must remain consistent with that for the DNB data analysis described in WCAP-14565-P-A VIPRE-01.

Licensee Compliance The licensee stated the VIPRE-W model used for the RTSR analyses was utilized as described in WCAP-14565-P-A. The required consistency between the VIPRE-W model and DNB data analysis has been maintained.

Given that the licensee is using the approved model as described in WCAP-14565-P-A, the NRC staff finds that the required consistency will be maintained and therefore, that this limitation and condition is met.

Limitation and Condition #2 The current 95/95 DNBR limit of 1.13 remains unchanged.

Licensee Compliance The ABB-NV DNBR Correlation is used for the rod withdrawal from subcritical analysis below the first mixing vane grid. The 1.13 correlation limit is used.

Given that the licensee is using the approved DNBR limit, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #3 DNBR calculations for CE-PWR fuels are within the current applicable range defined in Table 2-1 of the TR.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request.

Sequoyah is transitioning from Framatome HTP fuel to Westinghouse RFA-2 fuel.

Since the licensee is not using CE fuels, the NRC staff finds that this limitation and condition is not applicable.

3.8.13 WCAP-14565-P-A, Addendum 2-P-A, Revision 0 Limitation and Condition #1 The applicable range of the ABB-NV and WLOP correlations are presented in Table 1 and Table 2, respectively, of this SE.

Licensee Compliance OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated that the ABB-NV DNBR Correlation is used for the rod withdrawal from subcritical analysis below the first mixing vane grid. The applicability range in Table 1 of the SE is met and the 1.13 correlation limit is used. The WLOP DNBR Correlation is used for the hot zero power steam line break analysis. The applicability range in Table 2 of the SE is met and the 1.18 correlation limit is used.

Because the licensee has stated that the correlations are used within the range of applicability as presented in the SE to WCAP-14565-P-A, Addendum 2-P-A, Revision 0, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #2 The ABB-NV correlation and the WLOP correlation must use the same FC factor for power shape correction as used in the primary DNB correlation for a specific fuel design.

Licensee Compliance The licensee stated that the FC Tong factor (also referred to as the non-uniform power shape factor) from WCAP-8762-P-A, New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids, July 1984, is used with the ABB-NV and WLOP correlations as required by Limitation and Condition #2 of the NRC Safety Evaluation.

Given that the licensee is using a consistent FC factor, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #3 Selection of the appropriate DNB correlation, DNBR limit, engineering hot channel factors for enthalpy rise, and other fuel-dependent parameters will be justified for each application of each correlation on a plant specific basis.

Licensee Compliance See the discussion of Limitation and Condition #1 under WCAP-14565-P-A.

As discussed above for limitation and condition #1 for WCAP-14565-P-A, the NRC staff finds that the licensee selected appropriate DNB correlations, limits, etc., that have been approved for the Westinghouse RFA 2 fuel. Therefore, the staff finds that this limitation and condition is met.

Limitation and Condition #4 The ABB-NV correlation for Westinghouse PWR application and the WLOP correlation must be used in conjunction with the Westinghouse version of the VIPRE-01 (VIPRE) code since the correlations were justified and developed based on VIPRE and the associated VIPRE modeling specifications.

Licensee Compliance OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated that the ABB-NV and WLOP correlations are used with VIPRE-W in the Sequoyah RTSR analyses.

Given that the licensees RTSR analysis used the specified correlations in conjunction with VIPRE-W, the NRC staff finds that this limitation and condition has been met.

3.8.14 WCAP-15025-P-A Limitation and Condition #1 Since WRB-2M was developed from test assemblies designed to simulate Modified 17x17 Vantage 5H fuel, the correlation may only be used to perform evaluations for fuel of that type without further justification. Modified Vantage 5H fuel with or without modified intermediate flow mixer grids may be evaluated with WRB-2M.

Licensee Compliance The licensee stated that Sequoyah plans to transition 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 Sequoyah Units 1 and 2 includes the following features:

Integral fuel burnable absorbers (IFBAs)

Robust protective grid (RPG)

Standardized debris filter bottom nozzle (SDFBN)

High-burnup bottom grid Debris mitigating long fuel rod bottom end plugs Wet annular burnable absorbers (WABA)

Removable top nozzle (RTN)

Three ZIRLO intermediate flow mixer (IFM) grids Six ZIRLO RFA-2 structural mid-grids Reduced rod bow (RRB) INCONEL top grid ZIRLO and Optimized ZIRLO' high-performance fuel cladding with a coated cladding feature Thicker-walled guide thimble and instrumentation tubes to improve fuel assembly stiffness and to address incomplete rod insertion (IRI) considerations.

The licensee provided general information on the Westinghouse RFA-2 fuel and described differences between the 17x17 Vantage 5H and 17x17 RFA-2 fuel. The licensee stated that 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 (WCAP-10444-P-A Addendum 2-A). The RFA-2 mid-grid design was evaluated by the licensee with the NRC approved Fuel Criteria Evaluation Process (FCEP)

(WCAP-12488-A, Revision 0). The licensee stated that by complying with the requirements of FCEP, it demonstrated that the new mid-grid design meets all design criteria of existing 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 intermediate flow mixer grid.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Because the licensee used the NRC-approved FCEP and met applicable design criteria, the NRC staff finds that the WRB-2M DNBR correlation is acceptable for RFA-2 fuel.

Limitation and Condition #2 Since WRB-2M is dependent on calculated local fluid properties these should be calculated by a computer code that has been reviewed and approved by the NRC staff for that purpose. Currently WRB-2M with a DNBR limit of 1.14 may be used with the THINC-IV computer code. The use of VIPRE-01 by Westinghouse with WRB-2M is currently under separate review.

Licensee Compliance The licensee stated that a DNBR limit of 1.14 was used for WRB-2M. The NRC Safety Evaluation for WCAP-14565-P-A Addendum 1-A (ADAMS Accession Number ML041060018) notes that the WRB-2M correlation was licensed and incorporated into VIPRE-W for Westinghouse analyses.

After this limitation and condition was specified, the NRC staff approved the use of the WRB-2M correlation, which is incorporated into the VIPRE-W code, for use in Westinghouse analyses. In addition, inputs to VIPRE-W are generated with approved codes as discussed in limitation and condition #2 to WCAP-14565-P-A described above in Section 3.8.11. Therefore, staff finds this limitation and condition is met.

Limitation and Condition #3 WRB-2M may be used for PWR plant analyses of steady state and reactor transients other than loss of coolant accidents. Use of WRB-2M for loss of coolant accident analysis will require additional justification that the applicable NRC regulations are met and the computer code used to calculate local fuel element thermal/hydraulic properties has been approved for that purpose.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The FSLOCA EM (WCAP-16996-P-A Revision 1) is used for the Sequoyah RTSR LOCA analysis.

Because the LOCA analysis is performed with the NRC-approved FSLOCA methodology, the NRC staff finds that this limitation and condition is not applicable. Section 3.10 of this SE describes the licensees application of FSLOCA.

Limitation and Condition #4 The correlation should not be used outside its range of applicability defined by the range of the test data from which it was developed. This range is listed in Table 1.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Licensee Compliance The licensee stated that WRB-2M is used within the SE Table 1 range of applicability.

The licensee states that the correlation will be used within its range of applicability; therefore, the NRC staff finds this limitation and condition is met.

3.8.15 WCAP-16045-P-A Limitation and Condition #1 The PARAGON code can be used as a replacement for the PHOENIX-P lattice code, wherever the PHOENIX-P code is used in NRC-approved methodologies.

Licensee Compliance The licensee stated that PARAGON is used for the Sequoyah RTSR as discussed in WCAP-16045-P-A Addendum 1-A below (NEXUS/ANC).

The NRC staff finds that this limitation and condition is met as the licensee is using the PARAGON code in place of the PHOENIX-P code.

Limitation and Condition #2 The data base is insufficient to enable the staff to reach a conclusion regarding PARAGONs ability to predict depletion characteristics for a MOX fueled core at this time.

Licensee Compliance The licensee stated that this is not applicable to this license amendment request. The Sequoyah transition to RFA-2 fuel does not use MOX fuel.

Given that the proposed amendment would not authorize the use of MOX fuel, the NRC staff finds that this limitation and condition is not applicable.

3.8.16 WCAP-16045-P-A, Addendum 1-A Limitation and Condition Westinghouse has provided a series of assessments for the NEXUS/ANC code system in order to qualify the system as an NRC-approved code. While the re-parameterization within the NEXUS approach captures the dominant physical phenomena, the NEXUS/ANC code system is only approved to perform calculations on uranium-fueled, PWRs.

The NEXUS/ANC code system is limited to uranium-fueled, PWR applications as the only plant data assessments presented were for uranium-fueled PWRs. While Westinghouse has provided comparisons of the relative performance of PARAGON/ANC OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION and NEXUS/ANC for calculations with MOX fueled, PWR fuel assemblies, the PARAGON/ANC code system was not approved for this purpose. In the absence of actual plant data, NEXUS/ANC has not been approved for MOX applications.

Licensee Compliance The licensee stated that NEXUS/ANC is used for the Sequoyah transition to RFA-2 fuel.

It is not used for MOX applications.

Given that the licensee does not seek permission to use MOX fuel, the NRC staff finds that this limitation and condition is not applicable.

3.8.17 WCAP-16996-P-A, Revision 1 Limitation and Condition 1 The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.

Licensee Compliance The licensee stated that the analysis for Sequoyah Units 1 and 2 with the FSLOCA EM is only being used to demonstrate compliance with 10 CFR 50.46 (b)(1) through (b)(4).

Given that the licensee does not use the FSLOCA EM to demonstrate compliance with 10 CFR 50.46(b)(5), the NRC staff finds that the licensee meets this limitation and condition.

Limitation and Condition 2 The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.

Licensee Compliance The licensee stated that Sequoyah Units 1 and 2 are Westinghouse-designed 4-loop PWRs with cold-side injection, so it is within the NRC-approved methodology. The analysis for Sequoyah Units 1 and 2 utilized the NRC-approved FSLOCA methodology, except for the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in LTR-NRC-18-30 [ADAMS Accession No. ML19288A174].

The NRC staff reviewed the attachment to the letter. Westinghouse categorized these changes or errors into two separate groups, non-discretionary changes and discretionary changes. Four of those changes affected FSLOCA EM. The estimated effect of three of the changes lead to no impact on calculated PCT. The fourth change is an improvement in the pump component momentum equation at low pump speed which will either have no impact or a small penalizing impact on Region I analyses.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Given that SQN, Units 1 and 2, are Westinghouse-designed 4-loop PWRs with cold-side injection, the NRC staff finds that the FSLOCA EM is applicable to them. In addition, the NRC staff finds that the licensee has appropriately applied the FSLOCA EM with the changes or errors, as described above. Therefore, the NRC staff finds that the licensee meets limitation and condition 2.

Limitation and Condition 3 For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are modeled). This includes utilizing a plant-specific initial containment temperature, and only taking credit for containment coatings which are qualified and outside of the break zone-of-influence.

Licensee Compliance The licensee stated that the containment pressure calculation for the Sequoyah Units 1 and 2 analysis was performed consistent with the NRC-approved methodology.

Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A plant-specific initial temperature associated with normal full-power operating conditions was modeled, and no coatings were credited on any of the containment structures.

The NRC staff requested that the licensee explain how the plant-specific initial containment temperature that was modeled is expected to reduce the containment pressure. In response, the licensee stated that the initial temperature for the upper compartment (110°F), the lower compartment (130°F), the dead-ended compartment (130°F), and the ice bed compartment (32°F) were all initialized to the maximum temperature at full power operation. For ice condenser containment designs, the use of a maximum modeled temperature minimizes the predicted containment backpressure, as the effect of the reduced initial air mass more than offsets other effects such as reduced heat removal from the containment passive heat sinks.

The NRC staff finds that the licensee used the NRC-approved methodology for Region II, utilizing a plant-specific initial temperature associated with normal full-power operating conditions, and taking no credit for the coatings on any of the containment structures that should reduce the containment pressure, as required by the limitation and condition. Therefore, the NRC staff finds that the licensee meets limitation and condition 3.

Limitation and Condition 4 The decay heat uncertainty multiplier will be ((

)) The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Licensee Compliance The licensee stated that, consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was ((

)) for the Sequoyah Units 1 and 2 analysis. The analysis simulations were all executed for no longer than 10,000 seconds following reactor trip. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results have been provided in units of sigma and approximate absolute units in Table 7 [of enclosure 2 to the LAR].

The NRC staff finds that the licensee appropriately modeled decay heat per the limitation and condition and reported the resulting sampled values in units of sigma and absolute units for the limiting cases. Therefore, the NRC staff finds that the licensee meets limitation and condition 4.

Limitation and Condition 5 The maximum assembly and rod length-average burnup is limited to

(( )) respectively.

Licensee Compliance The licensee stated that the maximum analyzed assembly and rod length-average burnup were less than or equal to (( ))

respectively, for Sequoyah Units 1 and 2.

Based on the above, the NRC staff finds that the licensee meets limitation and condition 5.

Limitation and Condition 6 The fuel performance data for analyses with the FSLOCA EM should be based on the PAD5 code (at present), which includes the effect of thermal conductivity degradation.

The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PAD5 methodology.

Licensee Compliance The licensee stated that PAD5 fuel performance data were utilized in the Sequoyah Units 1 and 2 analysis with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of

[WCAP-17642-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5), November 2017 (ADAMS Accession No. ML17338A396)], and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of

[WCAP-17642-A, Revision 1].

Given that the licensee used the latest NRC-approved fuel performance code (i.e., PAD5) and used appropriate conservative inputs, the NRC staff finds that the licensee meets limitation and condition 6.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Limitation and Condition 7 The YDRAG uncertainty parameter should be ((

))

Licensee Compliance The licensee stated that, consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was ((

)) for the Sequoyah Units 1 and 2 Region I analysis.

The NRC staff finds that the licensee appropriately used the specified interfacial drag uncertainty parameter and, therefore, meets limitation and condition 7.

Limitation and Condition 8 The ((

))

Licensee Compliance The licensee stated that, consistent with the NRC-approved methodology, the ((

)) for the Sequoyah Units 1 and 2 Region I analysis.

The NRC staff finds that the licensee appropriately used the specified biased uncertainty parameters and, therefore, meets limitation and condition 8.

Limitation and Condition 9 For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm that the ((

)) for the plant design being analyzed. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Licensee Compliance The licensee stated that Sequoyah Units 1 and 2 are Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 11 [Westinghouse letter No. LTR-NRC-18-50, dated July 13, 2018].

The NRC staff reviewed the attachment to Westinghouse letter No. LTR-NRC-18-50, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs),

July 2018 (ADAMS Accession No. ML18198A041). This document describes the sensitivity studies done on the selected parameters and demonstrates that ((

] Therefore, the NRC staff finds that the licensee meets limitation and condition 9.

Limitation and Condition 10 For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to: 1) demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, and 2) ensure that the ((

)) must cover the equivalent 2 to 4-inch break range using RCS-volume scaling relative to the demonstration plant. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Additionally, the minimum sampled break area for the analysis of Region II should be 1 ft2.

Licensee Compliance The licensee stated that Sequoyah Units 1 and 2 are Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 11 [Westinghouse letter No.

LTR-NRC-18-50, dated July 13, 2018].

The minimum sampled break area for the Sequoyah Units 1 and 2 Region II analysis was 1 ft2.

The NRC staff reviewed the attachment to Westinghouse letter No. LTR-NRC-18-50, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs),

July 2018 (ADAMS Accession No. ML18198A041). This document describes the sensitivities performed to demonstrate that the boundary between Region I and Region II breaks is appropriate for a 4-loop Westinghouse-designed plant.

The limitation and condition specifies that plants with larger RCS fluid volumes than the 3-loop plant test example in WCAP-16996-P-A, Revision 1, should cover the same 2-to-4 inch range using break area to RCS volume scaling to ensure that the break range is preserved and not artificially truncated. The licensee applied ((

))

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION In order to demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, the analysis examined breaks ((

)) were not considered, the NRC staff finds that these would be bounded by the LBLOCA (Region II) similar to the IBLOCAs.

In addition, the Region II analysis for SQN, Units 1 and 2 considered a minimum break area of 1.0 ft2 consistent with the requirement in the limitation and condition.

Therefore, the NRC staff finds that limitation and condition 10 is met as the licensee performed the necessary sensitivity study to determine the appropriate break size range for Region I and boundary between Region I and Region II.

Limitation and Condition 11 There are various aspects of this Limitation and Condition, which are summarized below:

1. The (( )) the Region I and Region II analysis seeds, and the analysis inputs will be declared and documented prior to performing the Region I and Region II uncertainty analyses. The (( ))

and the Region I and Region II analysis seeds will not be changed throughout the remainder of the analysis once they have been declared and documented.

2. If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for PCT, MLO [maximum local oxidation], and CWO [core wide oxidation]

which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.

3. Plant operating ranges which are sampled within the uncertainty analysis will be provided in the analysis submittal for both regions.

Licensee Compliance The licensee stated that this Limitation and Condition was met for the Sequoyah Units 1 and 2 analysis as follows:

1. The (( )) the Region I and Region II analysis seeds, and the analysis inputs were declared and documented prior to performing the Region I and Region II uncertainty OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION analyses. The (( ))

and the Region I and Region II analysis seeds were not changed once they were declared and documented.

2. The analysis inputs were not changed once they were declared and documented.

3. The plant operating ranges which were sampled within the uncertainty analyses are provided for Sequoyah Units 1 and 2 in Table 1 [of Enclosure 2 to the LAR].

The licensee declared and documented the appropriate inputs and did not change these values once declared and documented; therefore, the NRC staff finds that the licensee meets limitation and condition 11.

Limitation and Condition 12 The plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves (MSSVs) must be adequately accounted for in analysis with the FSLOCA EM.

Licensee Compliance The licensee stated that a bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Sequoyah Units 1 and 2 analysis.

Because the licensee used a bounding dynamic pressure loss from the steam generator secondary side to the main steam safety valves, the NRC staff finds that the licensee meets limitation and condition 12.

Limitation and Condition 13 In plant-specific models for analysis with the FSLOCA EM: 1) the ((

)) and 2) the ((

))

Licensee Compliance The licensee stated that the ((

)) in the analysis for Sequoyah Units 1 and 2. The ((

)) in the analysis.

Based on the above, the NRC staff finds that the licensee meets limitation and condition 13.

Limitation and Condition 14 For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.

Licensee Compliance The licensee stated that, for the Sequoyah Units 1 and 2 analysis, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then summed with the pre-existing corrosion for comparison against the 10 CFR 50.46 local oxidation acceptance criterion of 17%.

The NRC staff finds that by using the Baker-Just correlation, converting to an ECR, and accounting for pre-existing corrosion, the licensee meets limitation and condition 14.

Limitation and Condition 15 The Region II analysis will be executed twice; once assuming loss-of-offsite power (LOOP) and once assuming offsite power available (OPA). The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.

The (( ))

Licensee Compliance The licensee stated that the Region II uncertainty analysis for Sequoyah Units 1 and 2 was performed twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the 10 CFR 50.46 acceptance criteria (see Section 5.0).

The ((

))

The LOOP configuration and the OPA configuration produced the same PCT (1,878°F) and are the limiting configurations for SQN, Units 1 and 2 (See Table 4 of Enclosure 2 to the LAR).

Given that the licensee performed the Region II analysis for both LOOP and OPA and that the results from both satisfy the acceptance criteria in 10 CFR 50.46(b)(1) through (b)(4), the NRC staff finds that the licensee meets limitation and condition 15.

3.8.18 WCAP-17661-P-A, Revision 1 Limitation and Condition #1, Use of AXY and AQ As discussed in Section 4.1.1 of this SE, the use of Methods 1 and 2 are acceptable for calculating AXY and AQ when performing RAOC and CAOC W(Z) surveillances, subject to the following limitations:

OFFICIAL USE ONLY PROPRIETARY INFORMATION

1. The NRC-approved methods provided in the response to RAI 15.b must be used to perform the surveillance-specific AXY or AQ calculations. Newer methods with similar capabilities may be considered acceptable provided the NRC staff specifically approves them for calculating AXY and AQ factors.

2. The depletion calculation used to determine the numerator and denominator of the AXY or AQ factor must be performed similarly to the original design calculation, as described in the response to RAI 15.c.

3. The use of Method 1 for calculating AQ is only acceptable subject to the constraints discussed in the response to RAI 15.a. The surveillance Axial Offset must be within 1.5-percent of the target AO, and there must be assurance that the limiting FQW(Z) location does not lie within a rodded elevation at the time of surveillance. Note that the use of Method 1 remains acceptable when surveillance-specific W(Z) functions are used.

Licensee Compliance The licensee stated:

1. TVA will use NRC-approved methods including those listed in the response to RAI 15.b (OG-18-35 Enclosure 1 dated February 15, 2018 in Appendix G of WCAP-17761-P-A Revision 1). Newer methods with similar capabilities may be used if the NRC specifically approves them for the AXY calculation. The AQ calculation is only applicable to constant axial offset control (CAOC) plants; Sequoyah will use relaxed axial offset control (RAOC) (AXY).

2. TVA will perform depletion calculations to determine the numerator and denominator of the AXY factor similarly to the original design calculations, that is, either with the BEACONTM core monitoring system without using nodal calibration factors, or with Advanced Nodal Code using the same nuclear model and depletion basis used to generate the original T(Z) function. The AQ calculation is only applicable to CAOC plants; Sequoyah will use RAOC (AXY).

3. This limitation applies to CAOC TS only and, thus does not apply to Sequoyah.

As noted by the licensee, the AQ calculation is only applicable to CAOC plants while SQN, Units 1 and 2 uses RAOC (as discussed in Section 3.3 of this SE), therefore, the staff finds that the portions of this limitation and condition that are related to AQ are not applicable. Given that the licensee will use NRC-approved methods for AXY calculations, the NRC staff finds that the licensee meets this limitation and condition.

Limitation and Condition #2, Power Level Reduction to 50 Percent RTP As noted in Section 4.3.2 of the SE on WCAP-17661-P-A, Revision 1, the use of 50 percent as the final power level reduction in the event of failed FQ surveillance is not included in the TS, but rather in the BASES and in the COLR. As such, this final power level, 50 percent, must be implemented on a plant-specific basis and included in COLR input generated using this methodology, in order to use this TR.

Licensee Compliance OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated that TVA will implement a final power level of 50 percent in the event of a failed FQ surveillance. This will be on a plant-specific basis and included in COLR input generated using this methodology upon implementing the License Amendment that allows adoption of the topical report.

Given that the licensee will implement, via its COLR, a final power level of 50 percent in the event of a failed FQ surveillance, the NRC staff finds that the licensee meets this limitation and condition.

3.8.19 Conclusion The NRC staff reviewed all of the limitations and conditions related to the 18 topical reports discussed above in this section. The NRC staff finds that the licensee meets all applicable limitations and conditions and that the use of these topicals is acceptable.

3.9 Operating License Conditions 2.C.(25) and 2.C.(18) Changes As stated by the licensee, TVA plans to transition SQN Units 1 and 2 from Framatome-supplied HTP fuel to Westinghouse RFA-2 fuel, commencing with Unit 1 Cycle 26 and Unit 2 Cycle 26.

As part of the transition, the licensee developed two transition cores which include both fuel types. The licensee anticipates that there will be two cycles which contain both fuel types before all the Framatome HTP fuel is removed leading to a final equilibrium core of Westinghouse RFA-2 fuel.

Due to differences in the design of each fuel type, the overall hydraulic resistances, including mixing vane grid loss coefficients, are different for each fuel type. When different types of fuel are co-resident, these differences result in a flow redistribution which results in changes to the flow in both the normal axial direction as well as the lateral, or crossflow, direction versus a core containing a single fuel type. The changes in flow through the fuel assemblies result in changes to the DNBR calculated with the VIPRE-W code using the WRB-2M and ABB-NV correlations.

These correlations are applicable over certain pressure and local quality ranges. In addition, the VIPRE-W models use the WRB-2M correlation for at-power events and for analyses applicable to the region above the first mixing vane grid while the ABB-NV correlation is used below the first mixing vane grid.

The licensee proposed to analyze the transition cores as if they were full cores of one assembly type. In order to do this, changes to the DNBR due to the flow redistribution must be taken into account. This is done by applying a DNBR penalty to the results of the full core analysis. The licensee computed transition core DNBR penalties according to the methodology documented in WCAP-11837-P-A. The transition core penalties described in WCAP-11837-P-A are determined as a function of the percentage of each fuel type in the core, however, the licensee chose to use the maximum penalty as a bounding value regardless of the percentage of each type of fuel in the transition cores. The NRC staff finds the use of a bounding penalty value acceptable as the licensee used the largest DNBR penalty over the range of percentage of each fuel type in the core.

For the Westinghouse RFA-2 fuel, the penalties are based on comparing the DNBR results from two different configurations as described in WCAP-11837-P-A. The first configuration uses a uniform array of Westinghouse RFA-2 fuel assemblies in a 3x3 grid while the second OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION configuration assumes one Westinghouse assembly surrounded by eight Framatome assemblies.

The licensee stated that the fuel assembly loss coefficient and the section/grid loss coefficients for the Westinghouse RFA-2 fuel are lower than that of the Framatome HTP fuel and, therefore, it is expected that there will be a transition core penalty levied against the Westinghouse fuel for bottom-skewed axial power shapes, but not for double-humped or top-skewed shapes. Given that the proposed license condition does not address power shape, the licensee stated that the calculated transition core penalty is applied generically for each critical heat flux (CHF) correlation region along the bundle axial length and that the penalties are conservatively applied to all analyses.

For conservatism, the licensee rounded the penalty up by more than10 percent from the penalty obtained from the mixed core analysis. The resulting penalties were determined to be 0.25 percent for the WRB-2M correlation and 0.5 percent for the ABB-NV correlation. As stated on page 27 of 63 of Enclosure 1 to the LAR, 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. To justify the lack of a penalty on the WLOP correlation, the licensee stated that the WLOP CHF correlation is only used for the hot zero power steam line break accident analysis and that a review of the analysis determined that none of the cases considered were limited in the lower portions of the fuel. Therefore, the licensee concluded that there is no requirement for a mixed core penalty on the WLOP CHF correlation. In addition, the licensee stated that this conclusion is supported by the WRB-2M DNBR penalty calculation in which no penalty was calculated for any accident other than rod withdrawal from subcritical which uses an extremely bottom limited axial power shape. The NRC staff finds this justification acceptable in that any locations in the fuel where the WLOP CHF correlation would be applied are not limiting in the present analyses and do not require a penalty.

While these penalties are expected to bound the safety analysis for reload evaluations and for Chapter 15 events, the licensee stated that sufficient available DNBR margin to cover the cycle-specific transition core penalties for each fuel type will be demonstrated on a cycle-specific basis. The NRC staff finds these penalties are acceptable as they were calculated with the approved methodology in WCAP-11837-P-A. The NRC staff finds that all limitations and conditions are met for WCAP-11837-P-A as described in Section 3.11.5.1 of this SE.

For the transition core effects on the Framatome HTP fuel, the licensee chose a different approach. Rather than apply a DNBR penalty to the Framatome HTP fuel, the licensee chose to reduce the nuclear enthalpy rise hot channel factor (FNH) by 5 percent over the value used for Westinghouse RFA-2 fuel (i.e., 1.7 to 1.61). The licensee performed a similar analysis as described above for the Westinghouse RFA-2 fuel to determine the DNBR penalty for the Framatome HTP fuel. In this case, two core configurations were examined. As before, the first configuration uses a uniform 3x3 array of Framatome HTP fuel assemblies while the second configuration assumes one Framatome HTP assembly surrounded by eight Westinghouse RFA-2 assemblies. In this case, the licensee used the WRB-1 correlation rather than the WRB-2M correlation. The licensee noted that while the WRB-1 correlation is not directly licensed for use with the HTP fuel, the correlation was intended to show the general impact of a transition core scenario on the HTP fuel. The WRB-2M correlation is not licensed for the Framatome HTP fuel either, but the WRB-1 correlation is an older and more conservative OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION correlation compared to the WRB-2M and newer correlations and is applicable over the full range of plant conditions.

For this analysis, the licensee computed a maximum transition core DNB penalty of 2.14 percent on the Framatome HTP fuel and determined the flow losses. The licensee then ran a case with the HTP fuel with the FNH reduced by 5 percent to demonstrate that the FNH reduction offsets the transition core DNB effect. The licensee also stated that to assure the 5 percent FNH reduction was sufficient to offset the mixed core impacts, a variety of core conditions (the same conditions considered for the transition core penalty calculation) were considered using a bounding core configuration and the most limiting power shape from the transition core penalty calculation. In addition, for each case considered, the minimum DNBR for the mixed core case (with FNH reduction) was greater than that of the homogenous core case, demonstrating that the 5 percent FNH reduction is sufficient to offset the mixed core impacts. The NRC staff finds that the licensee appropriately considered various core conditions along with the limiting axial power shape and finds that the 5 percent reduction in FNH is acceptable for the HTP fuel when co-resident with the RFA-2 fuel.

The existing license condition regarding a mixed core DNBR penalty states, TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6, 1997. In its LAR, the licensee proposed a revision to the SQN Units 1 and 2, Operating License (OL) to replace OL Condition 2.C.(25) and 2.C.(18),

respectively, to implement the Westinghouse core safety analysis methodology. Specifically, the licensee proposed that the following license condition replace the existing mixed core condition in each License.

When the Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies the:

HTP fuel assemblies FNH shall be maintained less than 1.61.

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 The existing mixed core license conditions were intended for the 1997 Framatome fuel transition. The NRC staff finds that removal of each of these conditions is appropriate and acceptable given that the proposed new conditions address the licensees application to operate with mixed core configurations where Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies.

For the proposed reduction in FNH, the licensee proposed to maintain the Framatome HTP fuel assemblies at less than 1.61. In its safety analysis, this was done by reducing the value to 5 percent of the Westinghouse RFA-2 proposed value of 1.70. The licensee explained that, while the 1.70 value for Westinghouse RFA-2 fuel will be documented in the COLR and is subject to change from cycle to cycle, the intent of the 1.61 value in the originally proposed license condition was based on the Westinghouse fuel using a value of 1.70 for FNH. The licensee further noted that if a different value were used for Westinghouse fuel (and documented in the COLR), the Framatome HTP fuel limit should be set by reducing the Westinghouse fuel limit by 5 percent. NRC staff asked the licensee to clarify if a change in the value for Westinghouse fuel OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION could result in a value for Framatome HTP fuel being inconsistent with the specific value of 1.61 given in the originally proposed license condition (ADAMS Accession No. ML21095A048).

Therefore, in its supplement dated May 5, 2021, the licensee revised the proposed license condition, in part, to read as follows:

When the Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies the:

HTP fuel assemblies FNH shall be maintained 5% less than the RFA-2 fuel FNH value The NRC staff finds that the above proposed license condition acceptable as it does not rely on a fixed value of 1.70 for the Westinghouse RFA-2 fuel and will remain applicable if the value for the RFA-2 fuel changes.

For the proposed DNBR penalties on the Westinghouse RFA-2 fuel in the LAR, the proposed license condition states the DNBR limit shall be reduced by: . To clarify how the DNBR penalties would be applied, the licensee stated that the condition is intended to refer to the calculated margin to the DNBR limit and that the DNBR penalties of 0.25 percent and 0.5 percent for the WRB-2M and ABB-NV CHF correlations, respectively, will be subtracted from the calculated margin to the DNBR limit. In addition, the licensee provided an example of how the penalties will be applied. In the supplement dated May 5, 2021, the licensee revised the proposed license condition in part to read as follows:

When the Framatome HTP fuel assemblies are co-resident with the Westinghouse RFA-2 fuel assemblies the:

RFA-2 fuel assemblies margin to the DNBR limit shall be adjusted by subtracting the following:

- 0.25% for the WRB-2M critical heat flux correlation

- 0.50% for the ABB-NV critical heat flux correlation The NRC staff finds that the revised license condition is acceptable as the penalties were conservatively determined and it correctly states that the margin is adjusted rather than the limit itself.

3.10 Full Spectrum LOCA 3.10.1 Description of FULL SPECTRUM LOCA Methodology As described in WCAP-16996-P-A, Revision 1, the purpose of the FSLOCA Methodology EM is to build on the ASTRUM EM, by extending the applicability of the WCOBRA/TRAC Code to include the treatment of SBLOCA and IBLOCA scenarios. The term Full Spectrum specifies that the new EM is intended to resolve the full spectrum of LOCA scenarios that result from a postulated break in the cold leg of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA methodology include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine rupture with a break flow area equal to two times the pipe area.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as SBLOCAs. Region II includes break sizes that are typically defined as LBLOCAs.

The FSLOCA EM explicitly considers the effects of fuel pellet thermal conductivity degradation (TCD) and other burnup-related effects by calibrating to fuel rod performance data input generated by the PAD5 code, which explicitly models TCD and is benchmarked to high burnup data. The fuel pellet thermal conductivity model in the WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet thermal conductivity degradation.

The safety evaluation for WCAP-16996-P-A, Revision 1 (ADAMS Package Accession No. ML17207A124) contains 15 limitations and conditions that must be met in order to implement the NRC-approved FSLOCA EM. A summary of each limitation and condition, the September 23, 2020, LAR statement of how the limitation and condition is met or inapplicable, and the NRC staff findings regarding each item are provided above in Section 3.8.17 of this SE.

3.10.2 Results and Compliance with 10 CFR 50.46 The licensee presented the results for PCT, MLO, and CWO in Table 4 of Enclosure 2 to the LAR for SQN, Units 1 and 2. To demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(4), the following criteria must be met:

1. PCT;

2. Maximum cladding oxidation;

3. Maximum hydrogen generation; and

4. Coolable geometry.

Each of the above four 10 CFR 50.46 criteria is discussed below.

Note that the FSLOCA EM does not address 10 CFR 50.46(b)(5), Long-term cooling.

Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The licensee stated that the actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM.

PCT The regulation, 10 CFR 50.46(b)(1), requires that [t]he calculated maximum fuel element cladding temperature shall not exceed 2200°F. The licensee stated that the analysis for PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level, and the resulting limiting PCTs are 1,878°F for the Westinghouse fuel and 2,024°F (which includes a penalty due to mixed-core affects, as discussed in Section 3.10.3 below) for the Framatome fuel. Given that the resulting PCTs are less than 2,200°F, the analyses with the FSLOCA EM confirm that the 10 CFR 50.46(b)(1) acceptance criterion is satisfied. The licensee presented the results for the Westinghouse fuel in Table 4 of Enclosure 2 to the LAR, and for Framatome fuel in the supplementary information provided on May 5, 2021.

Given that the maximum calculated PCT is below the 2,200°F PCT limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(1) is met.

Maximum Cladding Oxidation OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The regulation,10 CFR 50.46(b)(2), states, in part, that [t]he calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

The licensee stated that the analysis for MLO corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO is less than 17 percent when converting the time-at-temperature to an equivalent cladding reacted using the Baker-Just correlation and adding the pre-transient corrosion, the analysis confirms that the 10 CFR 50.46(b)(2) acceptance criterion is satisfied. The licensee presented the results in Table 4 of Enclosure 2 to the LAR for SQN, Units 1 and 2.

Given that the resulting MLO is below the 17 percent limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(2) is met.

Maximum Hydrogen Generation The regulation, 10 CFR 50.46(b)(3), states, The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

The licensee stated that the analysis for CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the resulting CWO is less than 1 percent, the analysis confirms that the 10 CFR 50.46(b)(3) acceptance criterion is satisfied.

The licensee presented the results in Table 4 of Enclosure 2 to the LAR for SQN, Units 1 and 2.

Given that the resulting CWO is below the 1 percent limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(3) is met.

Coolable Geometry The regulation, 10 CFR 50.46(b)(4), states, Calculated changes in core geometry shall be such that the core remains amenable to cooling. The licensee stated that this criterion is met by demonstrating compliance with criteria 10 CFR 50.46(b)(1), (b)(2), and (b)(3), and by ensuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. The NRC staff finds this acceptable because meeting the criteria for PCT, maximum cladding oxidation, maximum hydrogen generation, and fuel assembly grid deformation due to combined LOCA and seismic loads, provides reasonable assurance that the core geometry does not degrade to a condition that prevents the core from adequate cooling.

Section 32.1 of the NRC-approved FSLOCA EM documents that the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to the combined LOCA and seismic loads was calculated to not occur for SQN, Units 1 and 2. In Section 4.2.1.1.3.2 of the SQN, Units 1 and 2 UFSAR, the licensee stated that a calculation of the maximum LOCA and seismic grid impact forces, combined using the square root sum of the squares method (in accordance with SRP 4.2, Appendix A), which demonstrates that the maximum value is less than the allowable grid strength. The FSLOCA EM analyses did not invalidate the existing OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION seismic/LOCA analysis. Additional discussion of LOCA and seismic loads is in Section 3.11.2 of this SE.

Given that the criteria in 10 CFR 50.46(b)(1), (b)(2), and (b)(3) are met and that the analysis of seismic loads showed that inboard grid deformation due to the combined LOCA and seismic loads did not occur for SQN, Units 1 and 2, the NRC staff finds that the 10 CFR 50.46(b)(4) acceptance criterion is met.

3.10.3 Assessment of Mixed-Core Effects The SQN Units 1 and 2 analysis with the FSLOCA EM was performed assuming a full core of Westinghouse RFA-2 fuel. However, for the initial cycles in which Westinghouse RFA-2 fuel is used, Framatome HTP legacy fuel will also be present. While the two fuel designs have generally similar mechanical designs, a transition core evaluation was performed to address the mixed-core effects for both fuel types.

The loss coefficient of the Westinghouse RFA-2 fuel is slightly lower than the Framatome HTP fuel, and thus the RFA-2 fuel would receive a flow benefit in the presence of the relatively flow starved HTP fuel. For LBLOCA transients, conditions during blowdown and reflood can be affected by mixed-core conditions arising from a hydraulic mismatch, and the PCT increase for HTP fuel resulting from the hydraulic mismatch was estimated to be 23°F, based on the expected effects on a transient with the reflood time and cladding heatup rate. The licensee clarified that the Framatome Realistic LBLOCA (RLBLOCA) analysis predicts a PCT of 2,001°F for the HTP fuel, which becomes 2,024°F after the application of a mixed-core PCT increase of 23°F. The licensee further clarified that TVA will retain both vendors LOCA analyses and will report both of the PCT values in the annual/special 10 CFR 50.46 reporting. The staff understands that the 2,024°F PCT for HTP fuel will continue to be bounding in the second transition core because the HTP legacy fuel will be further depleted after the first transition cycle.

For SBLOCA transients, core-wide collapsed liquid levels correspond closely to a one-dimensional flow pattern, and the effects of grid loss coefficient differences among the assemblies are not significant in determining the PCT. As such, the SBLOCA PCTs for the RFA-2 fuel and HTP fuel are unchanged when adjacent in a mixed core, i.e., neither fuel type is flow starved in a SBLOCA scenario and the PCTs calculated based on the respective methodologies remain valid for mixed-core conditions, and no penalty due to mixed-core effect is necessary.

Given that the calculated PCTs for Westinghouse RFA-2 and Framatome HTP legacy fuels have adequate margin even after applying the PCT penalty for mixed-core effects, the NRC staff concludes that no unacceptable adverse effect of mixed-core operation is expected during the transition cycles.

3.10.4 FSLOCA Technical Evaluation Summary The licensee proposed to replace the existing NRC-approved LOCA methodologies in TS 5.6.3.b with the NRC-approved LOCA methodology contained in WCAP-16996-P-A, Revision 1.

The NRC staff concludes that the proposed TS change is acceptable because all of the limitations and conditions are satisfied for this plant and the licensee appropriately applied the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION FSLOCA EM to SQN Units 1 and 2 and finds that the resulting analysis meets the criteria in 10 CFR 50.46(b)(1) through (4).

3.11 Other Considerations 3.11.1 Nuclear Core Design As part of the fuel transition from Framatome HTP to Westinghouse RFA-2 fuel, the licensee plans to have two transition cycles that contain both fuel types. The licensee states in the LAR that the first transition cycle feeds 85 assemblies of Westinghouse RFA-2 fuel with both once and twice burned Framatome HTP fuel and that this higher than usual feed batch is utilized to maximize the population of RFA-2 fuel in the first transition core. The second transition cycle will contain feed and once burned RFA-2 fuel and twice burned HTP fuel. The licensee developed representative core designs for the two transition cores as well equilibrium cores containing only RFA-2 fuel. The licensee stated that the core designs utilized are not intended to represent limiting designs, rather they are 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. As stated in the LAR, the licensee performs cycle-specific core designs and fuel performance analyses for each reload cycle to guarantee that the fuel rod design criteria will remain satisfied for cycle-specific operating conditions.

The licensee developed additional core designs for use specifically with the FSLOCA methodology. These core designs are intended to be fully representative of the range of core characteristics expected during and after the fuel transition. These cores also use unrealistically high peaking factors that are intended to bound future reload peaking factors. Discussion of the application of FSLOCA is in Section 3.10 of this SE.

3.11.2 Seismic/LOCA Impact on Fuel Assemblies In order to demonstrate the fuel assembly structural integrity, the licensee performed detailed site-specific seismic and LOCA fuel assembly analyses for SQN, Units 1 and 2 in accordance with the following approved methodologies: WCAP-12610-P-A, WCAP-12488-A, WCAP-9401, and PWROG-16043-P-A. These analyses were also done to address limitation and condition

2 to WCAP-10444-P-A (see Section 3.8.4 in this SE). The licensees analyses considered both transition cores containing both Framatome HTP and Westinghouse RFA-2 fuel, and a full RFA-2 core. The analyses were performed with the established method focusing on the BOL condition, as presented in WCAP-12610-P-A. Additionally, analyses were performed to address the RFA-2 fuel at the EOL condition according to PWROG-16043-P-A.

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 licensee stated that the seismic analysis considered the safe shutdown earthquake and operating basis earthquake conditions. The LOCA analysis considered two different pipe break situations: accumulator line break and pressurizer surge line break. To address the fact that WCAP-9401-P-A considered other break locations, the licensee stated in the May 5, 2021, supplement that WCAP-9401-P-A was written and approved in 1981, which was before the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Leak-Before-Break (LBB) concept was codified with a change to GDC 4. The licensee referenced a letter dated July 19, 1989, from Suzanne Black (NRC) to Oliver Kingsley (TVA) titled Elimination of Primary Loop Pipe Breaks, General Design Criterion 4 (TAC Nos.

72829/72830) - Sequoyah Nuclear Plant Units 1 and 2, (ADAMS Accession No. ML20247B450 (letter), and ML20247B489 (associated safety evaluation)), which approved LBB for SQN, Units 1 and 2. The conclusion from the SE that approved the LBB for SQN states that the probability or likelihood of large pipe breaks occurring in the primary coolant system loops of Sequoyah is sufficiently low such that dynamic effects associated with postulated pipe breaks need not be a design basis. Therefore, the NRC staff finds the licensee justification acceptable and that the larger breaks considered in WCAP-9401-P-A do not need to be considered in the SQN, Units 1 and 2 design bases.

The licensee stated that 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, the licensee concluded that control rod insertability is maintained and coolable geometry is maintained.

In addition, the licensee performed an additional analysis for a mixed core of HTP and RFA-2 using the current Framatome methodology, BAW-10133NP-A, Revision 1, Addendum 1 and Addendum 2, Mark-C Fuel Assembly LOCA-Seismic Analysis, October 2000. The results of the analysis determined the impact loads on the HTP fuel in the mixed core and demonstrated that coolable geometry, control rod insertability, and fuel rod integrity will be maintained for the HTP fuel.

Given that the licensee analysis, performed in accordance with approved methodologies, predicted no permanent grid deformation to occur in both homogeneous core and mixed cores and found that control rod insertability is maintained under combined seismic and LOCA loadings, the NRC staff finds that fuel structural integrity is maintained and the requirements for core coolability in 50.46 are met.

3.11.3 UFSAR Changes Including Safety Analysis Impact As part of the fuel transition, a significant number of changes will need to be made to the UFSAR to account for both the RFA-2 fuel as well as the new methods. The majority of changes will be to Chapters 4, Reactor, and 15, Accident Analysis. While the UFSAR updates were not provided in the LAR, the licensee is required to update the SAR as specified in 10 CFR 50.71(e).

In Attachment 10 to the LAR, the licensee provided a summary of impact to the UFSAR safety analysis. For each applicable UFSAR section, the licensee included a description of the event/analysis, list of computer codes used, initial conditions, and an analysis conclusion. In all cases, the licensee concluded that all acceptance criteria are still met.

3.11.4 Source Terms The licensee performed an assessment of the design basis source term (core activity inventory).

As stated by the licensee, this assessment consisted of comparing the current analysis of OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 licensee compared the core loading parameters (core thermal power level, uranium mass, cycle average burnup, and uranium-235 fuel enrichment) of the representative cycles to the AOR core loading parameters and found that the AOR core loading parameters are bounding for all inputs based on conservatisms incorporated into the AOR. Therefore, the NRC staff finds that the AOR core activity inventory remains applicable for the transition to Westinghouse fuel.

3.11.5 Use of Other Topical Reports In Attachment 8 to the LAR, the licensee presented its position regarding compliance with topical report limitations and conditions. The review of the 18 topicals that are referenced in TS 5.6.3.b are presented in Section 3.8 of this SE. In addition to the 18 that will be in the COLR reference list, the licensee addressed the limitations and conditions of five additional topical reports. For each topical report listed below, three items are presented: 1) the limitation and condition(s) stated in the NRC SE for each specific topical report, 2) the explanation in the September, 23, 2020, LAR as to how the licensee complied with the applicable limitation and condition(s), and 3) the NRC staff findings regarding each item.

3.11.5.1 WCAP-11837-P-A, Revision 0 Limitation and Condition #1 Because the basic DNBR computational methodology is the same as has been previously used and approved, the restrictions and recommendations with respect to the DNBR methodology cited in the previous SERs will also be applicable in this extension.

Licensee Compliance The limitations and conditions for DNBR methodologies are discussed under the WRB-2M, ABB-NV, and WLOP topical reports.

As discussed in Section 3.8 of this SE, the NRC staff found that the licensee met all the limitations and conditions related to the DNBR methodologies. These include WCAP-15025-P-A and WCAP-14565-P-A for WRB-2M, WCAP-14565-P-A, Addendum 1-A for ABB-NV and WCAP-14565-P-A, Addendum 2-P-A for WLOP. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #2 Because the resulting curve fit used in the extension is a statistical fit through a set of data points corresponding to the DNBR penalties at different fractions of VANTAGE 5 fuels in the transition core, there is an error bound associated with it (Ref. 8)

[Westinghouse letter NS-NRC-89-3450 dated August 1, 1989]. Therefore, whenever data is read from the curve, its associated error bound should be simultaneously quoted and the data used should incorporate the uncertainty bound in a conservative manner.

Licensee Compliance OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION For the Sequoyah RTSR, 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 RTSR 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 (with intermediate flow mixers (IFM)) 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.

For the sake of 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.

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 was applied to the HTP fuel.

The transition core DNBR penalty for the Sequoyah RTSR 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 staff notes that the methodology described above is from WCAP-11837-P-A. The licensee chose to use a bounding DNBR penalty and not use a curve fit based on the different percentages of each fuel type in the transition cores as described in Section 3.9 of this SE.

While the licensee calculated a DNBR penalty for the Framatome fuel, it did not apply the DNBR penalty; rather, the licensee decreased the nuclear enthalpy rise hot channel factor (FH),

which was shown to offset the transition core DNBR penalty. By selecting the maximum penalty (for the RFA-2 fuel) and increasing it by >10 percent for conservatism, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #3 Since the case which represents the smallest fraction of VANTAGE 5 in the core in this submittal resulted in the bounding DNBR penalty, should a transition core involve a smaller fraction than 0.11 of VANTAGE 5 fuel assemblies, its DNBR penalty must be explicitly computed rather than using an extrapolated value from this curve.

Licensee Compliance Not applicable to this license amendment request. The first transition core will contain RFA-2 fuel assemblies in more than 11% of the core. Subsequent transition cores will contain a higher percentage of RFA-2 fuel.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated in the LAR that the first transition core will contain 85 assemblies of Westinghouse RFA-2 fuel and the second transition core will contain additional RFA-2 assemblies. Given there are 193 total fuel assemblies, the fraction of RFA-2 in the transition cores will be much greater than 11 percent, therefore, the NRC staff finds that this limitation and condition is not applicable.

Limitation and Condition #4 Westinghouses conclusion with respect to the regions of applicable penalties as described in the concluding paragraph of the subject topical report requires further qualification. Since (a) the study described in Reference 3 [Westinghouse letter NS-EPR-2643 dated August 17, 1982] only considered transition from STD [17x17 standard fuel assemblies] to OFA [17x17 optimized fuel assemblies] using a limited set of conditions, and (b) the limiting set of conditions identified in References 1 and 4

[WCAP-11837-P-A and WCAP-10444-P-A] was not used, and (c) all transition core effect studies presented in these reports analyzed generic cores, before any actual penalty from this reference is applied in a plant-specific transient analysis, Westinghouse should demonstrate that (1) the conditions for the analysis then under consideration are within the range of applicability (ranges of power, pressure, mass flux, and inlet temperature, and one of the three axial power shapes) considered in this study, and (2) that the case then being analyzed was otherwise similar to the cases used to develop the curves in this study (i.e., DNB occurred in the same general region of the core so that fluid element history was similar). If Westinghouse is unable to so demonstrate at such time, further justification will be necessary.

Licensee Compliance The conditions for the analysis under consideration are within the range of applicability (ranges of power, pressure, mass flux, and inlet temperature, and one of the three axial power shapes) considered in WCAP-11837-P-A, and the cases analyzed were otherwise similar to the cases used to develop the curves in WCAP-11837-P-A.

Given that the conditions for analysis are within the range of applicability, the NRC staff finds that this limitation and condition is met.

3.11.5.2 WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994 Limitation and Condition The staff has reviewed the BEACON system as described in References 2 [Topical Report WCAP-12472-P, BEACON Core Monitoring and Operations support Systems, April 1990, C. Beard and T. Morita] and 3 [Letter, with enclosed report, from N. Liparulo, Westinghouse, to R. Jones, NRC, November 4, 1992, Responses to NRC Request for Additional Information for WCAP-12472-P], the proposed technical specifications for operating with and without that system operable as described in Reference 3, and the attached Brookhaven TER.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION BEACON provides a greatly improved continuous on-line power distribution measurement and display, limit surveillance, and operation prediction information system for Westinghouse reactors. No new instrumentation or calculation system (other than interface systems and integration analysis) is introduced. The system is potentially suitable for other reactors, but has been examined in the topical report and by the staff only for current standard Westinghouse systems. Acceptance for others would require further review and approval.

As discussed in the attached TER (section 4.0, Technical Position), the system review has concluded that BEACON is acceptable for performing core monitoring and operations support, subject to conditions stated in TER sections 3.3 and 3.4 on uncertainties. As discussed in this staff evaluation, the revised proposed TSs provided in Reference 3 (Chapter 7 revision) are acceptable, with the exception of the placement of the thermocouple and incore detector operating limits in the COLR. They should be returned to TS 3/4.3.3.12. The condition stated in TER section 4.0.(3) with references to the TSs has been satisfied in the Reference 3 revision.

Section 4.0 of the Brookhaven Technical Evaluation Report (TER) states:

The BEACON Core Monitoring and Operations Support System Topical Report WCAP-12472-P and supporting documentation provided in Reference 9

[Responses to NRC Request for Additional Information for WCAP-12472-P (Proprietary), Letter, N. J. Liparulo (W) to R. C. Jones (NRC), dated November 4, 1992] have been reviewed in detail. Based on this review, it is concluded that the BEACON system is acceptable for performing core monitoring and operations support functions for Westinghouse PWRs subject to the conditions stated in Section-3 of this evaluation and summarized in the following.

1) In the cycle-specific applications of BEACON, the power peaking uncertainties Uh and UQ must provide 95% probability upper tolerance limits at the 95% confidence level (Section 3.3).

2) In order to insure that the assumptions made in the BEACON uncertainty analysis remain valid, the generic uncertainty components may require reevaluation when BEACON is applied to plant or core designs that differ sufficiently to have a significant impact on the WCAP-12472-P data-base (Section 3.4).

3) The BEACON Technical Specifications should be revised to include the changes described in Section 3 concerning Specifications 3.1.3.1 and 3.1.3.2 and the Core Operating Limits Report (Section 3.6).

Licensee Compliance Implementation of BEACON at the Sequoyah units uses the current standard Westinghouse systems. There are no changes to BEACON core monitoring methodology for the Sequoyah units.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The following discussions address the findings in the Brookhaven TER.

1) The 95/95 requirement in TER Section 3.3 is met.

The uncertainties to be applied to the BEACON power distribution measurements are calculated differently than those applied to the traditional flux map systems because BEACON uses a more comprehensive set of instrumentation.

The uncertainty in the BEACON power peaking resulting from errors in the model calibration and core exit thermocouple (CETC) calibration is determined using a Monte Carlo error propagation technique. In this analysis, the BEACON three-step calibration, model update, and power distribution update procedure are simulated. The model and CETC calibration factors are subjected to random variations, based on their uncertainties, and the resulting variations in the BEACON power distribution are used to determine the 95% probability upper tolerance limit on the assembly power for the approximately twenty highest powered assemblies.

The analysis is performed for a range of operating conditions including off-normal power distributions and extended calibration intervals. A typical set of CETC uncertainties is used together with a relatively large tolerance factor, which results in substantial smoothing of the CETC measurements. The upper tolerance limit on the assembly power peaking factor is calculated and found to increase as the square root of the CETC uncertainty.

The FNH and FQ(Z) uncertainties are determined by a statistical combination of the assembly peaking factor, axial peaking factor, model calibration interval, inoperable movable incore detectors, inoperable core exit thermocouples, grid factor, and local power peaking component uncertainties. The FNH and FQ(Z) uncertainties are continuously updated by the PDMS for actual operating conditions.

The accuracy of the BEACON analysis decreases as the calibration intervals increase and the power distribution diverges from the reference power shape. In order to minimize BEACON uncertainty, the reference power distribution is updated every 15 minutes, or when significant changes occur in the AFD or reactor power.

If the PDMS is functional, the core power distribution measurement uncertainty (UFH) to be applied to the FHN using the PDMS shall be the greater of:

UFH = 1.04 or UFH = 1.0 + (UH / 100)

Where:

UH = Uncertainty for power peaking factor as defined in Equation 5-19 from WCAP-12472-P-A.

This uncertainty is calculated and applied automatically by the PDMS.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION If the PDMS is not functional, the core power distribution measurement uncertainty (UFH) to be applied to the FHN shall be 1.04.

If the PDMS is functional, the core power distribution measurement uncertainty (UFQ) to be applied to FQ(Z) using the PDMS shall be calculated by:

UFQ = (1.0 + (UQ / 100))

UE where:

UQ = Uncertainty for power peaking factor as defined in Equation 5-19 from WCAP-12472-P-A.

UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the PDMS.

If the PDMS is not functional, the core power distribution measurement uncertainty (UFQ) to be applied to FQ(Z) shall be calculated by:

UFQ = UQU

UE where:

UQU = Base FQ measurement uncertainty of 1.05.

UE = Engineering uncertainty factor of 1.03.

The uncertainty values with the PDMS not functional will be specified in the Sequoyah COLR.

2) For the Sequoyah units 1 and 2, unit-specific thermocouple uncertainty parameters will be determined using data collected during the initial power ascension of each fuel cycle. The core design does not differ sufficiently from those used in the uncertainty evaluation of WCAP-12472-P-A to have a significant impact on the WCAP-12472-P-A database.

PDMS requires information on current plant and core conditions to determine the core power distribution using the core peaking factor measurement and measurement uncertainty methodology described in WCAP-12472-P-A. The core and plant condition information is used as input to the continuous core power distribution measurement software that continuously and automatically determines the current peaking factor values. The core power distribution calculation software provides the measured peaking factor values at nominal one-minute intervals.

For PDMS to accurately determine the core peaking factor value, the continuous core power distribution measurement software requires accurate information measured by the plant instrumentation (e.g., current reactor power level, average reactor vessel inlet temperature, control bank positions, power range detector calibrated voltage values, measured temperatures from a minimum number and distribution of operable CETCs).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The individual uncertainty components in the BEACON monitored power peaking are discussed in detail in WCAP-12472-P-A, Section 5.7. These components are grouped into three categories, (i.e., a) generic components, b) plant/cycle specific components, and c) input related to the plant operating conditions). The core instrumentation, in particular, can have different characteristics from plant to plant and cycle to cycle.

Therefore, the uncertainties are generated on a plant-specific basis for each cycle.

Additionally, PDMS continuously updates the uncertainty depending on the reactor operating conditions and the time since the last calibration constant update. The equations and constants to be used to determine the applicable measurement uncertainties to be applied to the core peaking factors determined by PDMS in the event that PDMS is inoperable are defined above.

For each plant/cycle specific application of PDMS, the reload design and plant/cycle specific information (e.g., COLR information, instrumentation data, RCCA data) is updated. This process also generates the cycle-specific PDMS constants (i.e.,

reference model), which includes the initial calculated calibration information.

3) The BEACON topical report Technical Specifications are not applicable to this amendment request. The only changes to the Sequoyah Technical Specifications are to replace various references to the movable incore detectors with core power distribution measurement information in Technical Specifications 3.1.7, 3.2.4, and 3.3.1. The PDMS will be used to perform the required functions in those Technical Specifications when the PDMS is functional. When not functional, the movable incore detectors will be used.

Consistent with previous licensing precedents, the functionality requirements for the PDMS will be located in the Sequoyah Technical Requirements Manual (TRM).

As described in Section 3.3 of the Brookhaven TER to WCAP-12472-P-A, the Monte Carlo analyses presented to determine the power peaking 95 percent probability upper tolerance limits do not include the tolerance factor to insure a 95 percent confidence level. Per part 1 of this limitation and condition, the licensee discussed the methodology as to how the uncertainties are calculated and stated that the uncertainty values with the PDMS not functional will be specified in the SQN, Units 1 and 2 COLR. For part 2, the licensee stated that unit-specific values will be determined during startup of each cycle and that the core design does not differ sufficiently from those used in WCAP-12472-P-A. For part 3, this limitation and condition is not applicable as BEACON will be located in the SQN, Units 1 and 2 TRM and not the TS as BEACON does not meet the requirements for inclusion in TS per the requirements of 10 CFR 50.36. Therefore, the NRC staff finds that this limitation and condition, as applicable, is met.

3.11.5.3 WCAP-12472-P-A, Addendum 4-A, Rev. 0, BEACON Core Monitoring and Operation Support System, Addendum 4, September 2012 Limitation and Condition The NRC staff has reviewed the Westinghouse submittal TR WCAP-12472-P/

WCAP-12472-NP, Addendum 4, Revision 0, and found the updated thermocouple methodology, the use of approved Westinghouse design model methodology, and the use of higher order polynomial fits for FID [fixed in-core detector] uncertainties provided OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION in the TR acceptable. The basis for acceptance is due to the provided qualitative and quantitative technical material contained in the TR.

Licensee Compliance The use of BEACON at the Sequoyah units will meet Addendum 4, including the use of the NEXUS/ANC model.

The NRC staff finds that the limitation condition listed by the licensee is a staff conclusion and not a limitation and condition. However, the NRC staff finds the licensee statement that the use of BEACON will meet Addendum 4, including the use of the NEXUS/ANC model, is consistent with the staff conclusion in WCAP-12472-NP, Addendum 4-A, Revision 0.

3.11.5.4 WCAP-14483-A, Generic Methodology for Expanded Core Operating Limits Report, January 1999 Limitation and Condition #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.

Licensee Compliance The thermal design flow (revised to 360,000 gpm total for four loops as part of the RTSR project) will be added to Technical Specification 3.4.1. This is the minimum RCS total flow rate corresponding to the maximum approved steam generator tube plugging limit.

The other DNB limits will be relocated to the COLR per TSTF-339-A Revision 2.

Section 3.5 of this SE discusses the changes to TS 3.4.1 and the relocation of certain parameters to the COLR. The proposed revision to TS 5.6.3.a requires that the parameters be documented in the COLR. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #2 Revise TS Table 3.3.1-1 of NUREG-1431, Reactor Trip System Instrumentation, to relocate the overtemperature T and overpower T (K) constant values and dynamic compensation (t) values, and the breakpoint and slope values for the f(l) penalty function(s) to the COLR.

Licensee Compliance These values are relocated to the COLR per TSTF-339-A Revision 2.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Section 3.4 of this SE discusses the changes to TS 3.3.1 and relocation of the setpoints to the COLR. The proposed revision to TS 5.6.3.a requires that the parameters be documented in the COLR. Therefore, the NRC staff finds that this limitation and condition has been met.

Limitation and Condition #3 Revise TS 2.1 Safety Limits of NUREG-1431, and the associated bases to relocate Figure 2.1.1-1 to the COLR and replace it with more specific requirements regarding the safety limits (i.e., the fuel DNB design basis and the fuel centerline melt design basis).

The NRC-approved methodology used to derive the parameters in the figure will be referenced in the Reporting Requirements section of the TS.

Licensee Compliance The requirement for compliance with the Safety Limits Figure will be retained in Technical Specification 2.1.1; however, the actual Figure 2.1.1-1 itself will be relocated to the COLR. Technical Specification 5.6.3.b has been revised per the proposed Technical Specification markups to include the approved analytical methods used to determine the core operating limits.

Section 3.1 of this SE discusses the changes to TS 2.1.1 Safety Limits and the relocation of Figure 2.1.1-1 to the COLR. Section 3.8 of this SE discusses the changes to TS 5.6.3.b and the addition of the approved Westinghouse methodologies. Therefore, the NRC staff finds that this limitation and condition is met.

3.11.5.5 WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5), November 2017 Limitation and Condition #4.1a The NRC staff limits the applicability of the PAD5 code and methodology for cladding, fuel types and reactor for the ranges that are listed [in the SE].

Licensee Compliance No exceptions are taken to the cladding, fuel, and reactor limitations in the PAD5 analyses performed for the RFA-2 fuel assemblies to be loaded at Sequoyah Units 1 and

2. 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). PAD5 was also used to provide the Westinghouse Transient Analysis (TA) group with representative fuel temperatures to demonstrate that the expected fuel temperatures associated with the Framatome HTP fuel are similar to those calculated for the RFA-2 fuel. Similar consideration for fuel temperatures was completed for prior fuel transitions (between V5H and the Framatome fuel product (see Section 5.2.2.26.4 of AREVA (Framatome)

Report ANP-2986(NP), Revision 2, Sequoyah HTP Fuel Transition (Non-Proprietary),

June 2011 (ADAMS Accession No. ML11172A070).

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The PAD5 inputs will be validated on a cycle-specific basis under the reload process performed per WCAP-9272-P-A.

Given that the licensee stated that there are no exceptions taken to the limitations in the PAD5 analyses performed for the RFA-2 fuel assemblies, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #4.1b The application of PAD5 should at no time exceed the fuel melting temperature as calculated by PAD5 due to the lack of properties for molten fuel in PAD5 and other properties such as thermal conductivity and FGR [fission gas release].

Licensee Compliance The peak fuel centerline temperature Safety Limit added to the Sequoyah Units 1 and 2 Technical Specifications as Safety Limit 2.1.1.2 in this license amendment request will assure that fuel melt is precluded during conditions for normal operation and anticipated operational occurrences. To assure that there will be a low probability for fuel melt for Condition I/II operation, the fuel centerline temperature is used to calculate maximum allowable local powers that are checked as part of the Reload Safety Analysis Checklist process to mitigate fuel melting. The definition for the fuel melt has been updated based on the maximum local fuel pin centerline temperature with the approval of PAD5. The specific changes to TS 2.1.1.2 are taken from Equation 6-14 in Section 6.1.5 of WCAP-17642-P-A with consideration given to conversion factors associated with temperature and burnup units. See ADAMS Accession Number ML17338A396 for the non-proprietary version and ADAMS Accession Number ML17334A841 for the proprietary version withheld from public availability. Both versions of the approved topical report were transmitted by Westinghouse letter LTR-NRC-17-75 dated November 27, 2017 (see ADAMS Accession Number ML17334A826 for the cover letter and application for withholding).

As discussed above in Section 3.1 of this SE, the licensee proposed a change to the peak fuel centerline temperature Safety Limit. As stated by the licensee, this safety limit will preclude fuel melt during normal operation and anticipated operational occurrences. Therefore, the NRC staff finds that this limitation and condition is met.

Limitation and Condition #4.1c The NRC staff did not review the [proprietary approach] that was suggested as a response to RAI-22 since there is no mention of this in the revised TR.

Licensee Compliance Not applicable to this license amendment request.

Given that the licensee did not propose to use the proprietary approach, the NRC staff finds that this limitation and condition is not applicable.

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Limitation and Condition #4.1d The NRC staff has reviewed the revised TR Chapter 8 on the model and methods improvement process (MMIP) and the RAI response (Reference 4) from Westinghouse.

The NRC staff acknowledges Westinghouse response that it is no longer requesting approval of the MMIP process, and as such, the NRC staff does not approve the streamlined MMIP whereby the models, uncertainty bound or methods that would be changed from those described in the revised TR and the RAI responses unless Westinghouse seeks NRC review and approval.

Licensee Compliance Not applicable to this license amendment request.

Given that the licensee is not proposing to use the MMIP, the NRC staff finds that this limitation and condition is not applicable.

Limitation and Condition #4.1e Since there are many references to the requests for additional information (RAIs) and their responses throughout the SE, the NRC staff is requiring Westinghouse to include the SE, the revised TR and all the responses to the RAIs in the final accepted version

(-A) of this TR on PAD5. This will enable clarity and completeness.

Licensee Compliance Not applicable to this license amendment request. The NRC accepted the -A version of WCAP-17642-P-A in their verification letter dated January 8, 2018 (ADAMS Accession Number ML17338A453). Future updates to PAD5 will be submitted by Westinghouse as needed.

The NRC staff finds that this limitation and condition does not apply to the licensee and applies to Westinghouse. Staff notes that Westinghouse previously met this limitation and condition when it submitted the final -A approved version of the topical report, which included the revised topical report, the RAI responses and the NRC staff SE.

4.0 Technical Evaluation Conclusion The NRC staff reviewed the proposed changes to TS 2.1, 3.1.4, 3.1.7, 3.2.1, 3.2.2, 3.2.4, 3.3.1, 3.4.1, 4.2.1, and 5.6.3, to support a transition to Westinghouse RFA-2 fuel with Optimized ZIRLOTM cladding. The licensee proposed to change analytical methods used to determine the core operating limits from Framatome methods to Westinghouse core safety analysis methodologies. The NRC staff finds that the licensee meets all applicable limitations and conditions specified in the proposed approved methodologies and that the methodologies are applicable to, and acceptable for use at, SQN, Units 1 and 2. The LAR included an exemption request for the use of Optimized ZIRLOTM , which the staff processed separately. Exemptions from 10 CFR 50.46 and Part 50, Appendix K.I.5 were granted on October 26, 2021 (ADAMS Accession No. ML21166A166). The NRC staff reviewed all the above proposed TS changes OFFICIAL USE ONLY PROPRIETARY INFORMATION

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and finds them consistent with both NUREG-1431 and the Westinghouse core safety analysis methodologies, except for minor deviations that were also determined to be appropriate. The staff finds that (1) the NRC-approved Westinghouse methodologies provide an acceptable method to determine operating limits and performance of core surveillance in a way that demonstrates compliance with the requirements identified in 10 CFR 50.46 for LOCAs and GDC 10 related to maintaining appropriate margin to assure that specified acceptable fuel design limits are not exceeded, and (2) the proposed revised TSs meet the requirements of 10 CFR 50.36(c)(2) for LCOs, 10 CFR 50.36(c)(3) for SRs, and 10 CFR 50.36(c)(5) for Administrative Controls, the proposed TS changes are acceptable.

The NRC staff further reviewed the analyses performed by the licensee to support Westinghouse RFA-2 fuel implementation for transition cycles and full-core conditions. These analyses included seismic/LOCA impact on fuel assemblies, source term (core activity inventory) and UFSAR safety analyses including LOCA analyses with the FSLOCATM methodology. The NRC staff finds these analyses support the intended transition to RFA-2.

The NRC staffs conclusion that RFA-2 fuel is acceptable for use in both units is based, in part, upon the inclusion of a license condition in each license, as specified in Section 2.2 of this SE, related to transition core peaking penalties.

As discussed in Sections 2.2 and 3.7 of this SE, the licensee proposed to modify TS 4.2.2 to reflect operation with 52 full-length control rod assemblies (with no control rod assembly installed in core location H-08). The staff reviewed the information provided by the licensee, including current licensing basis, thermal-hydraulic impacts, core reload design impacts, key safety parameter impact, and the UFSAR Chapter 15 safety analyses. The staff concludes that the licensee used methods consistent with regulatory requirements and the guidance identified in Section 2.0 above, and that future reactor cores with 52 control rods can be designed within the criteria established in the current SQN, Units 1 and 2 UFSAR Chapter 15 safety analyses and associated safety margins to support the fuel transition to Westinghouse RFA-2 fuel. The staff found that operation with 52 full-length control rod assemblies (with no control rod assembly installed in core location H-08) continues to meet the requirements established in GDC 2 regarding withstanding the effects of natural phenomena and GDC 4 regarding compatibility with the environmental conditions associated with normal operation and postulated accidents.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on July 28, 2021. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public OFFICIAL USE ONLY PROPRIETARY INFORMATION

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comment on such finding published in the Federal Register on December 1, 2020 (85 FR 77265). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: R. Beaton, NRR M. Razzaque, NRR S. Peng, NRR J. Grasso, NRR H. Vu, NRR Y. Wong, NRR Dated: October 26, 2021 OFFICIAL USE ONLY PROPRIETARY INFORMATION

J. Barstow

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 356 AND 349 REGARDING THE TRANSITION TO WESTINGHOUSE RFA-2 FUEL (EPID L-2020-LLA-0216)

DATED OCTOBER 26, 2021 DISTRIBUTION:

PUBLIC PM File Copy RidsNrrDorlLpl2-2 RidsNrrPMSequoyah RidsNrrRButler RidsACRS_MailCTR RidsRgn2MailCenter RidsNrrDeEicb RidsNrrDeEmib RidsNrrDssSfnb RidsNrrDssSnsb RidsNrrDssStsb RBeaton, NRR KHsu, NRR JLehning, NRR MRazzaque, NRR RStattel, NRR CTilton, NRR HVu, NRR YWong, NRR ADAMS Accession Package No.: ML21210A171 Proprietary Amendment: ML21210A172 Nonproprietary Amendment: ML21245A267 OFFICE NRR/DORL/LPLII-2/PM NRR/DORL/LPLII-2/LA NRR/DSS/SNSB/BC NAME PBuckberg RButler SKrepel DATE 8/11/2021 8/11/2021 6/17/2021 OFFICE NRR/DEX/EICB/BC NRR/DEX/EMIB/BC (A) NRR/DSS/STSB/BC (A)

NAME MWaters ABuford NJordan (M. Hamm for)

DATE 6/3/2021 5/19/2021 8/13/2021 OFFICE OGC - NLO NRR/DORL/LPLII-2/BC NRR/DORL/LPLII-2/PM NAME MYoung DWrona PBuckberg DATE 10/22/2021 10/25/2021 10/26/2021 OFFICIAL RECORD COPY OFFICIAL USE ONLY PROPRIETARY INFORMATION

v t e Letter Sequence Approval
EPID:L-2020-LLA-0216, Application to Modify the Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09) (Approved, Closed)
Initiation Request, Request Acceptance Administration Withholding Request Acceptance Questions 5 April 2021 Answers 5 May 2021 Results Approval Other: ML21035A322, ML21036A109, ML21090A228
MONTHYEAR2021-02-04 [Table View]

ML21245A267

Text

SUBJECT:

Dear Mr. Barstow:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

5.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

SUBJECT:

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