Text

Response to Request for Additional Information Re Application to Modify Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09)
	ML21125A347
Person / Time
Site:	Sequoyah
Issue date:	05/05/2021
From:	Polickoski J; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CNL-21-039, EPID L-2020-LLA-0216, SQN-TS-20-09
	Download: ML21125A347 (22)
	v • d • e

1101 Market Street, Chattanooga, Tennessee 37402 CNL-21-039 May 5, 2021 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject:

Response to Request for Additional Information Regarding Application to Modify the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification to Allow for Transition to Westinghouse RFA-2 Fuel (SQN-TS-20-09)

(EPID L-2020-LLA-0216)

References:

1. TVA letter to NRC, CNL-20-014, Application to Modify the Sequoyah Nuclear Plant Units 1 and 2 Technical Specification to Allow for Transition to W estinghouse RFA-2 Fuel (SQN-TS-20-09), dated September 23, 2020 (ML20267A617)

2. NRC Electronic Mail to TVA, Request for Additional Information Regarding Request to Transition to W estinghouse Fuel (EPID L2020LLA0216),

dated April 5, 2021 (ML21095A048)

In Reference 1, Tennessee Valley Authority (TVA) submitted a request for amendments to Renewed Facility Operating License Nos. DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2, respectively. The proposed amendm ents would revise Technical Specifications (TS) to allow the use of W estinghouse RFA-2 fuel with Optimized ZIRLOTM 1 cladding. Further, the proposed amendm ents would revise TS 5.6.3, Core Operating Limits Report to replace the loss-of-coolant accident analysis evaluation model references with the FULL SPECTRUMTM 2 Loss-of-Coolant Accident (FSLOCA TM 2) Evaluation Model. Finally, the proposed amendm ents 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.

1 Optimized ZIRLO is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved.

Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

2 FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world.

U.S. Nuclear Regulatory Commission CNL-21-039 Page 2 May 5, 2021 In Reference 2, the NRC issued a request for additional information (RAI) and requested TVA respond by May 5, 2021. Enclosure 1 to this letter provides the TVA response to the RAI. to this letter provides revised pages to Enclosure 1 of the Reference 1 license amendment request (LAR). Enclosures 3 and 4 provided the corrected markups and final typed pages, respectively, of the license conditions from Attachments 1 and 2 in Enclosure 1 of the LAR.

This letter does not change the no significant hazard considerations nor the environmental considerations contained in Reference 1. Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the enclosures to the Tennessee Department of Environment and Conservation.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Kimberly D. Hulvey at (423) 751-3275.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 5th day of May 2021.

Respectfully, James T. Polickoski Director, Nuclear Regulatory Affairs

Enclosures:

1. Response to NRC Request for Additional Information

2. Corrected Sections of Enclosure 1 to CNL-20-014

3. Corrections (markups) to SQN Unit 1 and Unit 2 License Conditions

4. Corrections (final-typed version) to SQN Unit 1 and Unit 2 License Conditions cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee State Departm ent of Environm ent and Conservation

Enclosure 1 Response to NRC Request for Additional Information NRC INTRODUCTION By application dated September 23, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20267A617), the Tennessee Valley Authority (TVA or the licensee) submitted a license amendment request (LAR) to revise the Technical Specifications for Sequoyah Nuclear Plant, Units 1 and 2 (Sequoyah). The proposed amendments would allow transition to Westinghouse Robust Fuel Assembly-2 (RFA-2) fuel with Optimized ZIRLOTM cladding at Sequoyah. Further, the proposed amendments would revise 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 (FSLOCA) Evaluation Model (FSLOCA EM). 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.

The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing the LAR and has identified where additional information is needed to complete its review. The NRC staffs request for additional information (RAI) is below.

REGULATORY BASIS The regulatory basis for RAI-1 through RAI-8 is General Design Criterion (GDC) 10, Reactor design. GDC 10 states that, 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.

The regulatory basis for RAI-9 through RAI-11 is Section 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, of Title 10 of the Code of Federal Regulations. Section 50.46 of 10 CFR, in part, establishes standards for the calculation of emergency core cooling system (ECCS) accident performance and acceptance criteria for that calculated performance NRC STAFF REQUESTS

RAI-1

TVA plans to transition Sequoyah Units 1 and 2 from Framatome-supplied High Thermal Performance (HTP) fuel to Westinghouse 17x17 RFA-2, commencing with Unit 1 Cycle 26 and Unit 2 Cycle 26. The licensee developed two transition core designs in addition to two equilibrium cycles as representative core designs to move Sequoyah from Framatome HTP fuel to Westinghouse RFA-2 fuel. The transition cycles contain both Framatome HTP and Westinghouse RFA-2 fuel.

For the proposed changes to TS 2.1.1, Reactor Core SLs [Safety Limits], the licensee has proposed a single departure from nucleate boiling ratio (DNBR) and single maximum local fuel pin centerline temperature.

CNL-21-039 E1 1 of 10

Enclosure 1 a) The NRC staff requests the licensee provide justification that only one DNBR limit and correlation is applicable for a transition core that contains two different types of fuel.

b) The NRC staff requests the licensee provide 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. The justification should address the fact that the new limit and associated degraded rate is much higher (i.e. non-conservatively) than the current limits and associated limit degraded rates.

TVA Response TVA Response to RAI-1a The Departure from Nucleate Boiling Ratio (DNBR) limit provided by Westinghouse as an update to Technical Specification (TS) Safety Limit 2.1.1.1 will be applied only to Westinghouse fuel. The DNBR limit for Fram atome fuel will continue to be applied as listed in Updated Final Safety Analysis Report (UFSAR) Section 4.5.4.1.1. The Framatome analysis of record (AOR) is established with a F NH value of 1.70 as listed in UFSAR Section 4.5.4.3.2.1. The 5% reduction in surveilled F NH from the W estinghouse F NH limit for the Framatom e fuel ensures that the impacts due to the mixed core are adequately addressed for Framatom e fuel.

TVA Response to RAI-1b The local fuel pin centerline temperature limit provided by Westinghouse as an update to Technical Specification (TS) Safety Limit 2.1.1.2 will only be applied to Westinghouse fuel, and is not applicable to the Fram atome fuel. Instead, for prevention of Framatom e fuel centerline m elt, 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 kilowatts per foot (kW/ft) limits of 20.05 for UO2 rods and 18.95 kW/ft for UO2-Gd2O3 will be confirm ed analytically to ensure the prevention of fuel centerline melt. The FQ limit for Core Operating Limits Report (COLR) item 2.6.1 (corresponding with TS 3.2.1) is based on the m ost recent COLR supporting Framatom e HTP fuel and will be utilized with all fuel surveilled to a limit of 2.62. Therefore, 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 Framatom e will be met, while the fuel pin centerline TS limit ensures the W estinghouse fuel analysis will be m et.

CNL-21-039 E1 2 of 10

Enclosure 1

RAI-2

The NRC staff identified the following two deviations in the proposed changes to TS 3.2.1, Heat Flux Hot Channel Factor (FQ(X,Y,Z)):

Deviation 1: The new Required Action (RA) B.2.1 proposed in SQN as Limit allowable THERMAL POWER and AFD limits as specified in the [Core Operating Limits Report] COLR which is different from the B.2.1 approved in WCAP-17661-P-A B.2.1 as Limit THERMAL POWER to less than RATED THERMAL POWER and reduce AFD limits as specified in the COLR.

Deviation 2: The licensee proposed to add after each FQw(z) determination to the Completion Time for the new B.2.1, B.2.2 and B.2.3.

(a) The NRC staff requests the licensee address and provide a justification for Deviation 1, as identified above.

(b) The licensee referred to TSTF-241 in its justification for Deviation 2; however, the basis referred to from TSTF-241 for this justification is only applicable to the repeating power reduction process required by RAs A.1, A.2 and A.3, which is not required for RAs B.2.1, B.2.2 and B.2.3. Therefore, the NRC staff requests the licensee provide additional justification for the proposed deviation.

TVA Response TVA Response to RAI-2a Deviation 1 is a semantics issue intended to make the action clearer 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). There is no technical difference in the terminology used.

TVA Response to RAI-2b Deviation 2 is not based solely on TSTF-241-A Rev. 4, but instead on a combination of the related impacts associated with TSTF-241-A and TSTF-290-A as discussed in the LAR , Section 3.2.3 (page 33 of 63). 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 tim e 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 the Standard Technical Specifications (STS) NUREG-1431, Revision 2. Based on the TS 3.2.1 Bases discussion of WCAP-17661-P-A, Revision 1, 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 FQ W(Z) determinations. This deviation is in discussion within the Pressurized W ater Reactor Owners Group (PWROG) as a potential revision to WCAP-17661-P-A, Revision 1.

CNL-21-039 E1 3 of 10

Enclosure 1

RAI-3

Page 47 of 63 of Enclosure 1 to the LAR states It is therefore expected that there will be a transition core penalty levied against the Westinghouse fuel for bottom-skewed shapes, but not for double-humped or top-skewed shapes. Given that the proposed license condition does not address power shape, the NRC staff requests the licensee to confirm that the transition core penalty will be applied to all Westinghouse RFA-2 fuel, when mixed with Framatome HTP fuel, regardless of the power shape and applied to all analyses. Further, clarify whether the penalty is applied over the entire axial length of the fuel rod.

TVA Response to RAI-3 The calculated transition core penalty is applied generically for each critical heat flux (CHF) correlation region along the bundle axial length. The ABB-NV correlation is used below the first mixing vane grid, for bottom-skewed power shapes and incurs a maximum DNBR penalty of 0.5%; whereas, the WRB-2M correlation is used above the first mixing vane grid, for all shapes (i.e., bottom-skewed, double-humped, or top-skewed) and will incur a conservative 0.25% DNBR penalty, regardless of the mixed core composition. The W RB-2M transition core penalty is only necessary for the Rod Withdrawal from Subcritical (bottom-skewed power shape) analysis but will be conservatively applied to all analyses using the WRB-2M correlation. The ABB-NV correlation is only used below the first mixing vane grid; thus, the ABB-NV DNBR penalty will only be applied in that region.

RAI-4

Page 48 of 63 of Enclosure 1 to the LAR states Where necessary, transition core penalties or conservative operational limits are set for the RFA-2 vs. HTP fuel (e.g., 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 NRC staff requests the licensee clarify whether the single case, as discussed on page 48 of 63, was the only analysis done to assure the 5 perfect reduction is an appropriate value. In addition, given that the 1.7 value will be located in the COLR and is subject to change from cycle to cycle, justify why a fixed value of 1.61 is acceptable for use in the proposed license condition.

TVA Response to RAI-4 The 5% reduction in F NH is credited in the thermal-hydraulic transition core analysis to offset the transition core effects on the Framatom e HTP fuel. To assure that the 5% F NH reduction was sufficient to offset the mixed core impacts, a variety of core conditions (the sam e 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. For each case considered, the minimum DNBR for the mixed core case (with F NH reduction) was greater than that of the hom ogenous core case, demonstrating that the 5% F NH reduction is sufficient to offset the mixed core impacts.

F NH limits for each fuel type will be surveilled in accordance with the limits as listed in the COLR. The intent of the 1.61 value in the license condition was based on the Westinghouse fuel using a 1.70 F NH value. If a lesser value were published in the COLR for W estinghouse fuel, the Fram atome HTP fuel limit would be set by reducing the Westinghouse fuel limit by 5%. Enclosures 2, 3, and 4 to this submittal provide the corrected version of the proposed license condition, which also takes into account revisions to address RAI-5.

CNL-21-039 E1 4 of 10

Enclosure 1

RAI-5

On page 49 of 63 of Enclosure 1 to the LAR, the proposed license condition states, in part:

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 NRC staff requests the licensee to clarify what is meant by the DNBR limit shall be reduced by. Does this mean that the DNBR safety limit (as specified in TS 2.1.1.1) of 1.14 is actually reduced, or does it mean the values computed by the correlations are reduced by 0.25% or 0.50% (depending on correlation used), and then compared to the limit? In addition, the NRC staff requests the licensee provide an example of how these penalties are applied.

TVA Response to RAI-5 The statement was intended to refer to the calculated margin to the DNBR limit. The DNBR penalty of 0.25% and 0.5% for the WRB-2M and ABB-NV CHF correlations, respectively will be subtracted from the calculated margin to the DNBR limit. The 95/95 correlation limit DNBR is not modified to reflect downstream analytical calculations. Enclosure 2 to this submittal provides the corrected version of the proposed license condition in Section 2.3 and Section 3.2.9 of Enclosure 1 to the LAR. Enclosure 3 to this submittal provides the corrected markups to SQN Unit 1 License Condition 25 and SQN Unit 2 License Condition

18. Enclosure 4 to this submittal provides the corrected final typed SQN Unit 1 License Condition 25 and SQN Unit 2 License Condition 18.

Note that the F NH Condition was also updated to address the possibility of applying a lesser F NH limit to the W estinghouse fuel as described in the response to RAI-4.

The following example dem onstrates the application of a 0.5% DNBR penalty:

Suppose a calculated minimum DNBR of 1.750 is compared to a DNBR limit of 1.25.

The total amount of margin to the DNBR limit is calculated to be 28.5% (1-1.25/1.75).

The 0.5% DNBR penalty is then subtracted from the total available margin of 28.5%

to determine a remaining m argin of 28.0% (28.5%-0.5%).

RAI-6

The proposed license condition has DNBR penalties for the WRB-2M and ABB-NV correlations. However, 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. The NRC staff requests the licensee to explain why there is no penalty applied to the WLOP correlation.

CNL-21-039 E1 5 of 10

Enclosure 1 TVA Response to RAI-6 The W estinghouse Low Pressure Critical Heat Flux (WLOP CHF) correlation is used only for the Hot Zero Power Steamline Break (SLB) accident analysis. A review of the Sequoyah reload transition Hot Zero Power SLB Thermal-Hydraulic DNBR analysis was performed, and it was determined that none of the cases considered were limited in the lower portions of the fuel. As the grid loss coefficients (except bottom nozzle) were higher for the HTP fuel compared to the RFA-2 fuel, the relative flow is increased for RFA-2 fuel above the bottom portions of the fuel. Because there were no cases for Hot Zero Power SLB analysis that were limiting in the lower fuel region, then there is no requirement for a mixed core penalty for the WLOP CHF correlation. 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 is extremely bottom limited.

RAI-7

On page 8 of 63 of Enclosure 1 to the LAR, it states The LOCA analysis considered two different pipe break situations: accumulator (ACC) line break and pressurizer (PZR) surge line break. The staff requests the licensee explain why these breaks were chosen as other breaks were examined in WCAP-9401-P-A that may be more limiting.

TVA Response to RAI-7 WCAP-9401-P-A, Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly, August 1981 was written and approved prior to approval of the Leak-Before-Break (LBB) concept was codified with a change to General Design Criterion 4. Suzanne Black (NRC) letter to Oliver Kingsley (TVA), "Elimination of Primary Loop Pipe Breaks, General Design Criterion 4 (TAC Nos. 72829/72830) - Sequoyah Nuclear Plant Units 1 and 2," dated July 19, 1989, approved LBB for Sequoyah. Therefore, the larger breaks discussed in WCAP-9401-P-A have been excluded from the structural analysis.

The technical basis for the LBB application is presented in UFSAR Section 3.6 Protection Against Effects Associated with the Postulated Rupture of Piping. Table 3.6.2-1 Postulated New Design Basis Break Locations for the LOCA Analysis of the Primary Coolant Loop per WCAP-8172 as Eliminated by LBB per WCAP-12012, summarizes the primary coolant loop break locations remaining after application of LBB. The three break locations are:

1. Residual Heat Removal (RHR) Line/Primary Coolant Loop Connection

2. Accumulator (ACC) Line/Primary Coolant Loop Connection

3. Pressurizer Surge (PS) Line/Primary Coolant Loop Connection.

The first location listed above (RHR line) was not used in determining the LOCA analysis loads, because as the pressure at the time of potential RHR line break is much lower and will not result in more limiting loads.

CNL-21-039 E1 6 of 10

Enclosure 1

RAI-8

Page 45 of 63 of Enclosure 1 to the LAR states While the Framatome HTP fuel remains within compliance with the Framatome methodologies listed in the current TS 5.6.3, these Framatome methods no longer establish core operating limits or PCT [peak cladding temperature] during a LOCA and have been removed. However, this appears inconsistent with page 8 of 45 of Attachment 8 to the LAR which states As such, the existing 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. The NRC staff requests the licensee to clarify if the Framatome methods establish the PCT during a LOCA and address if any other results from the Framatome methods are limiting for the two transition cores.

TVA Response to RAI-8 The Background subsection of LAR Enclosure 1 Section 3.2.8 states, these Framatome methods no longer establish core operating limits or PCT [peak cladding temperature]

during a LOCA and have been removed. However, the Framatome LOCA methodologies remain applicable to the Framatome fuel in transition cores. As such, the statement that the Framatom e methods no longer establish the PCT during a LOCA was not intended. As each fuel type was analyzed using different LOCA methodologies, it is not known which fuel would actually exhibit the PCT during a LOCA event. The other limiting results of core wide oxidation and maximum local oxidation cannot be determined due to the same differences in LOCA methodology. Therefore, TVA will retain both vendors LOCA analyses in parallel and will report both of the PCT values in annual/special 10 CFR 50.46 reporting. Enclosure 2 to this submittal provides the corrected version of the statement in Section 3.2.8 of the LAR (page 45 of 63 of Enclosure 1 to the LAR), with or PCT deleted from this section.

The remainder of the subject statement in the Background subsection of LAR Enclosure 1 Section 3.2.8, that is, that the Framatome methods no longer establish core operating limits, remains valid. The treatm ent of the HTP fuel in the transition cores expressed in this paragraph of the Background subsection of Section 3.2.8 ensures that the W estinghouse methods are appropriate for setting the core operating limits for the transition cores.

Therefore, rem oval of the Framatome m ethods from the list of COLR methods cited in TS 5.6.3, including those for LOCA, is appropriate as a part of this amendment.

CNL-21-039 E1 7 of 10

Enclosure 1

RAI-9

Limitation and Condition #3 of the FSLOCA EM requires, in part, that 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. Page 4 of 46 of Enclosure 2 of the LAR states that a plant-specific initial containment temperature associated with normal full-power operating conditions was modeled to comply with the Limitation and Condition. The NRC staff requests the licensee explain how the plant-specific initial containment temperature that was modeled is expected to reduce the containment pressure, which is required by the Limitation and Condition.

TVA Response to RAI-9 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 m odeled 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.

RAI-10

Page 13 of 46 of Enclosure 2 of the LAR states 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. The LAR also states that, for large-break loss-of-coolant accident (LBLOCA) transients, conditions during blowdown and reflood can be affected by mixed core conditions arising from a hydraulic mismatch, and that the PCT increase for HTP fuel resulting from the hydraulic mismatch was estimated to be 23°F. Staff requests the licensee clarify the following:

(a) Given that the Sequoyah analysis with the FSLOCA EM was performed assuming a full core of RFA-2 fuel only (when the actual core for the two transition cycles is a mixed core of RFA-2 and HTP fuels), and the 23°F PCT penalty is applicable only to HTP fuel, the NRC staff requests the licensee to explain how the 23°F penalty is applied during the transition cores.

(b) The NRC staff requests the licensee confirm whether the 23°F PCT increase for HTP legacy fuel is accounted for in the limiting LBLOCA PCT results for the mixed core shown in Table 4 of Enclosure 2 of the LAR.

(c) The NRC staff requests the licensee clarify which fuel type (RFA-2 or HTP) would result in the limiting LBLOCA PCT for the transition cores. How is this determination made given that the analyses are done with a full core or each fuel type?

CNL-21-039 E1 8 of 10

Enclosure 1 TVA Response TVA Response to RAI-10a The 23°F PCT penalty is applicable to the HTP fuel as a change to the existing Fram atome AOR, which continues to support the HTP fuel during the transition cores. The FSLOCA EM (evaluation m odel) analysis supports Westinghouse RFA-2 fuel during the transition cores as well as future full-core RFA-2 cycles. The 23°F PCT penalty is not applicable to the FSLOCA EM analysis results, as the overall lower loss coefficients of the W estinghouse RFA-2 fuel would not lead to a PCT penalty for the RFA-2 fuel.

TVA Response to RAI-10b The results in Table 4 of LAR Enclosure 2 are only for a full core of RFA-2 as predicted by the analysis using the FSLOCA EM, therefore the 23°F PCT penalty for the HTP fuel is not applicable to Table 4. As described in the UFSAR, the Fram atome Realistic LBLOCA (RLBLOCA) analysis predicts a PCT of 2001°F for the HTP fuel, which becomes 2024°F after the transition core PCT increase is applied.

TVA Response to RAI-10c The FSLOCA EM is used to determine a LBLOCA PCT of 1878°F for a full core of RFA-2 fuel. This result conservatively neglects any flow benefit the RFA-2 fuel would receive when adjacent to HTP fuel. The Framatome RLBLOCA analysis is used to determine a LBLOCA PCT for a full core of HTP fuel. The HTP LBLOCA PCT is increased by 23°F for HTP fuel in mixed cores to account for decreased flow when adjacent to RFA-2 fuel. The above described analyses and evaluations do not establish which fuel type has the limiting LBLOCA PCT, but do establish a conservative LBLOCA PCT for each fuel type in a mixed core.

As each fuel type is analyzed using different LOCA m ethodologies, it is not known which fuel would actually exhibit the limiting PCT. TVA will retain both vendors LOCA analyses and will report both of the PCT values in annual/special 10 CFR 50.46 reporting.

RAI-11

Page 13 of 46 of Enclosure 2 of the LAR states that, for small-break loss-of-coolant accident (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 existing 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. In light of the preceding statements, the NRC staff requests the licensee clarify which fuel type (RFA-2 or HTP) has the limiting SBLOCA PCT for the mixed core, as shown in Table 4 of Enclosure 2 of the LAR, is expected to occur and why.

TVA Response to RAI-11 The FSLOCA EM is used to determine a SBLOCA PCT of 1213°F for a full core of RFA-2 fuel, and the results of this analysis are reported in Table 4 of Enclosure 2 of the LAR. As CNL-21-039 E1 9 of 10

Enclosure 1 described in the UFSAR, the Framatome SBLOCA analysis m ethodology uses the S-RELAP5 code and determined a SBLOCA PCT of 1543°F for a full core of HTP fuel. In a SBLOCA, the effect of the grid loss coefficient differences is 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. That is, neither fuel type is starved in a SBLOCA scenario and the PCTs calculated based on the respective methodologies rem ain valid for mixed core conditions. The above described analyses do not establish which fuel type has the limiting SBLOCA PCT, but do establish a SBLOCA PCT for each fuel type that remains valid in a mixed core.

CNL-21-039 E1 10 of 10

Enclosure 2 Corrected Section 2.3 of Enclosure 1 (pg. E1 6 of 63) to CNL-20-014 Corrected Section 3.2.8 of Enclosure 1 (pg. E1 45 of 63) to CNL-20-014 Corrected Section 3.2.9 of Enclosure 1 (pg. E1 49 of 63) to CNL-20-014 CNL-21-039

Enclosure 1 Evaluation of the Transition to Westinghouse RFA-2 Fuel exemption to 10 CFR 50.46 is also requested in Enclosure 5 consistent with WCAP-12610-P-A

& CENPD-404-P-A, Addendum 1-A. The proposed change would also delete the discussion of Framatome lead test assemblies and report BAW-2328 consistent with the proposed implementation of Westinghouse core safety analysis methodology.

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

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

Renewed License No. DPR-77 and DPR-79 The proposed change also revises the SQN Units 1 and 2 Operating License (OL) to replace OL condition 2.C (25) and 2.C (18) respectively, as a result of the implementation of Westinghouse core safety analysis methodology. Specifically, the following license condition is proposed.

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

HTP fuel assemblies FNH shall be maintained 5% less than 1.61the RFA-2 fuel FNH value.

RFA-2 fuel assemblies margin to the DNBR limit shall be reduced byadjusted by subtracting the following:

0.25% for the WRB-2M critical heat flux correlation 0.50% for the ABB-NV critical heat flux correlation 2.4 Condition Intended to Resolve The proposed change will allow TVA to use Westinghouse core safety analysis methodologies as part of a transition from Framatome-supplied HTP fuel to Westinghouse 17x17 RFA-2 fuel with Optimized ZIRLO cladding. This includes the use the FSLOCA Evaluation Model to evaluate the peak cladding temperatures for LBLOCAs and SBLOCAs for SQN Units 1 and 2. The proposed change will allow TVA to operate with 52 RCCAs and no full-length control rod assembly in core location H-08 for Units 1 and 2 for the fuel transition and subsequent cycles.

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Enclosure 1 Evaluation of the Transition to Westinghouse RFA-2 Fuel RFA-2 fuel will always be limiting and establishes core operating limits for the transition cores. While the Framatome HTP fuel remains within compliance with the Framatome methodologies listed in the current TS 5.6.3, these Framatome methods no longer establish core operating limits or PCT during a LOCA and have been removed.

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

WCAP-8745-P-A (methodology for OTT and OPT Reactor Trip System setpoints in TS 3.3.1)

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

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

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

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

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

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

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

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

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

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

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

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Enclosure 1 Evaluation of the Transition to Westinghouse RFA-2 Fuel Summary and Conclusions The analytical methods used to determine the core operating limits have been previously reviewed and approved by the NRC. This satisfies the stipulation in NRC Generic Letter 88-16 for relocating cycle-specific variables to the COLR. Attachment 8 goes into more detail on the application of these methodologies as well as others, such as PAD5, that arent specifically cited in the COLR.

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

Background

The proposed change also revises the SQN Units 1 and 2 Operating License (OL) to replace OL condition 2.C(25) and 2.C(18) respectively, as a result of the implementation of Westinghouse core safety analysis methodology. Specifically, the following license condition is proposed.

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

HTP fuel assemblies FNH shall be maintained 5% less than 1.61the RFA-2 fuel FNH value.

RFA-2 fuel assemblies margin to the DNBR limit shall be reduced byadjusted by subtracting the following:

0.25% for the WRB-2M critical heat flux correlation 0.50% for the ABB-NV critical heat flux correlation Summary and Conclusions Removal of the current license conditions is appropriate as the conditions were associated with a mixed core DNBR penalty resulting from the Framatome fuel conversion in 1997.

The proposed license conditions are appropriate for the mixed cores until a homogeneous core of Westinghouse RFA-2 fuel exists. Additional information is provided above in subsection 3.2.8.

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

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

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Enclosure 3 Corrections (markups) to SQN Unit 1 License Condition 25 and SQN Unit 2 License Condition 18 CNL-21-039

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

This condition has been deleted.

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

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

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

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

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

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

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

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

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

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

TVA will obtain NRC approval prior to startup for any cycle's coreN that involves a (a) The HTP fuel assemblies FNNH shall be maintained less thanthan 5% less 1.61;the RFA-2 fuel F H value.

reduction in the Hdeparture from nucleate boiling ratio initial transition core penalty (b) The RFA-2 fuel assemblies DNBR limit shall be reduced by:

margin to the DNBR limit shall be adjusted by subtracting the following:

below that value stated in TVA's submittal on Framatome fuel conversion dated

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

April 6, 1997.

2. 0.50% for the ABB-NV critical heat flux correlation 2.

Renewed License No. DPR 77 September 28, 2015

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

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

(17) Surveillance Interval Extension The performance interval for the 36-month surveillance requirements in TS 4.3.2.1.3 shall be extended to May 18, 1996, to coincide with the Cycle 7 refueling outage. The extended interval shall not exceed a total of 50 months for the 36-month surveillances.

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

5% N (a) The HTP fuel assemblies FNH shall TVA will obtain NRCbe maintained approval toless less prior thanthan 1.61; startup theany for RFA-2 fuel Fcore cycle's H value.

that involves a (b) The RFA-2 fuel assemblies margin DNBR to the limit DNBR shall be limit reducedshall be by: adjusted by subtracting reduction in the departure from nucleate boiling ratio initial transition the core following:

penalty 1.

1. 0.25% for the WRB-2M critical heat flux correlation below that value stated in TVA's submittal on Framatome fuel conversion dated 2.

2. 0.50% for the ABB-NV critical heat flux correlation April 6, 1997.

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

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

(21) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy and with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel Renewed License No. DPR 79 September 28, 2015

Enclosure 4 Corrections (final-typed version) to SQN Unit 1 License Condition 25 and SQN Unit 2 License Condition 18 CNL-21-039