L-2021-071, License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt Full Spectrum Tm LOCA Methodology

License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt Full Spectrum Tm LOCA Methodology

ML21105A848
Person / Time
Site:	Turkey Point
Issue date:	04/15/2021
From:	Pearce M Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML21105A847	List: L-2021-071, License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt Full Spectrum Tm LOCA Methodology
References
L-2021-071
Download: ML21105A848 (76)

Text

April 15, 2021 L-2021-071 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington D C 20555-0001 RE: Turkey Point Nuclear Plant, Unit 3 and 4 Docket Nos. 50-250 and 50-251 Renewed Facility Operating Licenses DPR-31 and DPR-41 License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt FULL SPECTRUMTM LOCA Methodology Pursuant to 10 CFR Part 50.90, Florida Power & Light Company (FPL) hereby requests amendments to Renewed Facility Operating Licenses (RFOLs) DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by revising the Loss-of-Coolant Accident (LOCA) methodology to reflect the adoption of WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) (ADAMS Accession No. ML17277A130).

The enclosure to this letter provides FPLs evaluation of the proposed change. Attachment 1 to the enclosure provides for information only, a technical evaluation of the FULL SPECTRUMTM LOCA (FSLOCATM) evaluation model (EM) application to Turkey Point. Attachment 1 contains information that Westinghouse Electric Company LLC (Westinghouse) considers to be proprietary in nature. Pursuant to 10 CFR 2.390(a)(4), FPL requests the proprietary information be withheld from public disclosure. provides a non-proprietary version of the technical evaluation provided in Attachment 1. provides the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-21-5171 affidavit supporting the proprietary withholding request. The request is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission

("Commission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Accordingly, FPL requests that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-21-5171 and be addressed to Camille T. Zozula, Manager, Regulatory Compliance & Corporate Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Suite 165, Cranberry Township, Pennsylvania 16066. Attachment 4 provides the Turkey Point TS pages marked up to show the proposed changes. Attachment 5 provides the Turkey Point TS Bases pages marked up to show the proposed changes. The TS Bases markups are provided for information only and will be incorporated in accordance with the Turkey Point TS Bases Control Program upon implementation of the approved license amendments.

FPL has determined that the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and there are no significant environmental impacts associated with the proposed change. The Turkey Point Onsite Review Group has reviewed the proposed license amendments. In accordance with 10 CFR 50.91(b)(1), a copy of the proposed license amendments is being forwarded to the State designee for the State of Florida.

FPL requests that the proposed change is processed as a normal license amendment request with approval within one year of submittal. Once approved, the amendments shall be implemented on a forward fit basis by no later than the next Unit 3 and Unit 4 reload campaigns, respectively.

Florida Power & Light Company 9760 SW 344th Street, Homestead, FL 33035

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 1 of 127 EVALUATION OF THE PROPOSED CHANGE Turkey Point Nuclear Plant Unit 3 and Unit 4 License Amendment Request 273, Update Listing of Approved LOCA Methodologies to Adopt FULL SPECTRUMTM LOCA Methodology 1.0

SUMMARY

DESCRIPTION ............................................................................................................. 2 2.0 DETAILED DESCRIPTION ............................................................................................................. 2 2.1 System Design and Operation ............................................................................................ 2 2.2 Current Requirements / Description of the Proposed Change ........................................... 2 2.3 Reason for the Proposed Change ...................................................................................... 2

3.0 TECHNICAL EVALUATION

............................................................................................................ 3

4.0 REGULATORY EVALUATION

....................................................................................................... 4 4.1 Applicable Regulatory Requirements/Criteria ..................................................................... 4 4.2 No Significant Hazards Consideration ................................................................................ 5 4.3 Conclusion .......................................................................................................................... 6

5.0 ENVIRONMENTAL CONSIDERATION

.......................................................................................... 6

6.0 REFERENCES

................................................................................................................................. 7

.. - Application of FSLOCATM EM to Turkey Point (proprietary version) ......8 - Application of FSLOCATM EM to Turkey Point (non-proprietary version) .....61 - Westinghouse Affidavit Requesting Withholding Proprietary Information .....114 - Turkey Point Technical Specification Pages (markup) .118 - Turkey Point Technical Specification Bases Pages (markup) .122

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 2 of 127 1.0

SUMMARY

DESCRIPTION Florida Power & Light Company (FPL) requests amendments to Renewed Facility Operating Licenses (RFOL) DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point),

respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by revising the referenced Loss-of-Coolant Accident (LOCA) methodology to reflect the adoption of WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) (Reference 6.1).

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Turkey Point Updated Final Safety Analysis Report (UFSAR), Chapter 14.3, summarizes the comprehensive safety analysis of postulated pipe ruptures within the Reactor Coolant System (RCS) boundary for Turkey Point. The analysis includes cases of the Loss of Coolant Accident (LOCA) resulting from a broad spectrum of small and large pipe ruptures including the Maximum Hypothetical Accident (MHA) case of the double ended guillotine (DEG) rupture of the largest RCS pipe. The objective of the analysis is to determine the condition of the RCS, core, and containment in the event of a postulated LOCA, and to verify that the various Emergency Core Cooling Systems (ECCS) have the capability to control each LOCA, including the MHA. The design requirements of existing plant systems, structures, and components (SSCs) are used as inputs to the Turkey Point LOCA analysis, but do not directly impact the existing design or configuration of any plant SSCs.

2.2 Current Requirements / Description of the Proposed Change TS 6.9.1.7 specifies the requirements for the Core Operating Limits Report (COLR),

including the listing of NRC approved analytical methods for determining the Heat Flux Hot Channel Factor, FQ(Z), the Nuclear Enthalpy Rise Hot Channel Factor, FH, and the K(Z) curve, the normalized FQ(Z) as a function of core height.

The proposed change updates the listing of approved COLR analytical methods for determining FQ(Z), FH and the K(Z) curve by replacing the 1st listed analytical method with WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology). The proposed change deletes the 2nd through 6th listing of approved analytical methods and renumbers the 7th and 8th listed methods to become the 2nd and 3rd. The proposed change additionally deletes the footnote denoted by a double-asterisk (**) applicable to the existing 5th and 6th listed methods proposed for deletion.

2.3 Reason for the Proposed Change The proposed license amendments update the TS listing of approved COLR analytical methods to fulfill an FPL regulatory commitment to submit a license amendment request (LAR) to adopt the FULL SPECTRUMTM LOCA (FSLOCATM) methodology of WCAP-16996-P-A, Revision 1 (Reference 6.1) by April 2021, as established in Reference 6.2.

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 3 of 127

3.0 TECHNICAL EVALUATION

3.1 Update COLR Listing of Approved LOCA Methodologies The proposed change updates the listing of approved analytical methods for determining FQ(Z), FH and the K(Z) curve by replacing the 1st through 6th listed COLR analytical methods of TS 6.9.1.7 with WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) (Reference 6.1). The proposed change thereby deletes the 2nd through 6th listing of approved analytical methods, deletes the footnote denoted by a double-asterisk

(**) applicable to the existing 5th and 6th listed methods and renumbers the 7th and 8th listed methods to become the 2nd and 3rd. These changes are administrative in nature since, as described below, the 1st through 6th analytical methods are being replaced by WCAP-16996-P-A, Revision 1, and renumbering the 7th and 8th analytical methods does not change any applicable requirements. The proposed change reflects the adoption of WCAP-16996-P-A, Revision 1, demonstrating Turkey Point compliance with the ECCS performance criterion of 10 CFR 50.46 subject to the NRCs specified conditions and limitations. The proposed change is necessary to fulfill a regulatory commitment (Reference 6.2) to complete a re-analysis and submit an amendment request adopting the FSLOCATM methodology by April 2021. The commitment requires implementation of the revised methodology on a forward-fit basis by no later than the next reload campaign following implementation of a license amendment adopting the FSLOCATM Evaluation Model (EM) of WCAP-16996-P-A, Revision 1.

The FSLOCATM EM was developed to address the full spectrum of loss-of-coolant accidents (LOCAs) which result from a postulated break in the reactor coolant system (RCS) of a pressurized water reactor (PWR). The break sizes considered in the FSLOCATM EM 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 (DEG) rupture of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as small and large break LOCAs. The FSLOCATM EM encompasses the methodologies addressed in the 2nd through 6th listed analytical methods for determining FQ(Z), FH, and the K(Z) curve, consistent with Table 1 of Reference 6.1. Generic authorization for use of the FSLOCATM EM was provided in Reference 6.1 subject to the conditions and limitations specified therein.

An analysis utilizing the FSLOCATM EM was performed for Turkey Point. Consistent with the NRC endorsed guidance in NEI 96-07, Revision 1 (Reference 6.3) for adopting an approved methodology, consideration was given for the type of analysis and the applicable terms, conditions and limitations in confirming the intended use of the FSLOCATM EM.

Based on the generic approval of WCAP-16996-P-A, Revision 1, for Westinghouse 3-loop plants with cold leg ECCS injection, such as Turkey Point, and satisfaction of the fifteen (15) conditions and limitations specified in the NRCs Safety Evaluation Report (SER) for WCAP-16996-P-A, Revision 1 (Reference 6.1), the 10 CFR 50.59 evaluation concluded that the FSLOCATM EM can be implemented following NRC approval of the revised TS 6.9.1.7 listing of approved COLR methodologies. As such, implementation of the FSLOCATM EM is scheduled on a forward-fit basis pending approval of this amendment request, consistent with the regulatory commitment in Reference 6.2. Attachment 1 provides for information only, the technical evaluation of the application of the FSLOCATM EM to Turkey Point. Attachment 1 contains information that Westinghouse Electric Company LLC considers to be proprietary in nature. Pursuant to 10 CFR 2.390(a)(4), FPL requests the proprietary information be withheld from public disclosure. Attachment 2 provides a non-proprietary version of the technical evaluation provided in Attachment 1.

Attachment 3 provides the Westinghouse Application for Withholding Proprietary

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 4 of 127 Information from Public Disclosure CAW-21-5171 affidavit supporting the proprietary withholding request.

Turkey Point currently utilizes reactor fuel assemblies containing fuel rods which feature Optimized ZIRLOTM cladding. As such, the Zircaloy-4 and ZIRLO cladding fuel properties (i.e., inputs) were not selected for Turkey Points application of the FSLOCATM EM, though these cladding types are approved for use in WCAP-16996-P-A, Revision 1. Should it become desirable to address the presence of Zircaloy-4 or ZIRLO cladding in future Turkey Point LOCA analyses, the change will be evaluated as a separate, independent change in accordance with 10 CFR 50.46. In addition, the TS Bases are amended to reflect that the initial containment pressure and temperature assumed in the Turkey Point FSLOCATM EM analysis are based on limiting normal operating conditions rather than the minimum containment pressure and maximum containment temperature limits specified in the Turkey Point TS, consistent with the WCAP-16996-P-A, Revision 1, methodology.

Attachment 5 to this amendment request provides the Turkey Point TS Bases pages marked up to show the proposed change. The TS Bases changes are provided for information only and will be incorporated in accordance with the Turkey Point TS Bases Control Program upon implementation of the approved license amendments.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.46 establishes acceptance criteria for emergency core cooling systems for light-water nuclear power reactors fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding.

10 CFR 50, Appendix K, specifies documentation requirements for the emergency core cooling performance evaluation models specified in 10 CFR 50.46.

1967 Proposed General Design Criteria (GDC) 15 states that protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features.

1967 Proposed GDC 37 states that engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends.

1967 Proposed GDC 41 states that engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.

1967 Proposed GDC 42 states that engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident.

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 5 of 127 1967 Proposed GDC 44 states that at least two emergency core cooling systems, preferably of different design principles, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary, including the double-ended rupture of the largest pipe.

1967 Proposed GDC 49 states that the containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.

1967 Proposed GDC 52 states that where an active heat removal system is needed under accident conditions to prevent exceeding containment design pressure, this system shall perform its required function, assuming failure of any single active component.

The proposed license amendments comply with the requirements of 10 CFR 50.46, 10 CFR 50, Appendix K, and 1967 Proposed GDCs 15, 37, 41, 42, 44, 49 and 52, consistent with the applicable regulations and regulatory guidelines. Therefore, all requirements will continue to be satisfied as a result of the proposed change.

4.2 No Significant Hazards Consideration The proposed license amendments revise the Turkey Point Technical Specifications (TS) by revising the referenced LOCA methodology to reflect the adoption of WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology). As required by 10 CFR 50.91(a),

FPL has evaluated the proposed change using the criteria in 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration.

An analysis of the issue of no significant hazards consideration is presented below:

(1) Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change neither alters plant equipment nor the manner in which equipment is operated and maintained, and thereby cannot increase the probability of an accident. The proposed change cannot adversely affect the type or amount of effluent that may be released off-site or increase individual or cumulative occupational exposures resulting from any design basis accident, and thereby cannot increase the consequences of any accident.

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

(2) Do the proposed amendments create the possibility of a new or different kind of accident from any accident previously evaluated?

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 6 of 127 Response: No The proposed change neither installs new nor modifies existing plant equipment and thereby cannot introduce new equipment failure modes. The proposed change does not alter safety analysis assumptions, or create new accident initiators or precursors, and thereby cannot introduce a new or different type of accident.

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

(3) Do the proposed amendments involve a significant reduction in a margin of safety?

Response: No The proposed change does not modify any safety limits, limiting safety system settings, or safety analysis assumptions or inputs, and thereby cannot affect plant operating margins. The proposed change does not modify equipment credited in safety analyses, and thereby cannot affect the integrity of any radiological barrier.

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

Based upon the above analysis, FPL concludes that the proposed license amendment does not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92, Issuance of Amendment, and accordingly, a finding of no significant hazards consideration is justified.

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

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendments revise the Turkey Point Technical Specifications (TS) by revising the referenced LOCA methodology to reflect the adoption of WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology). FPL has evaluated the proposed amendments for environmental considerations and determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 7 of 127

6.0 REFERENCES

6.1 Westinghouse Topical Report WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) October 13, 2017 (ADAMS Accession No. ML17277A130) 6.2 Florida Power & Light Letter L-2018-077, Schedule for Re-Analysis of Turkey Point Licensing Basis Analyses Affected by PAD5 Implementation, March 27, 2018 (ADAMS Accession No. ML18086A154) 6.3 Nuclear Energy Institute (NEI) 96-07, Revision 1, Guidelines for 10 CFR 50.59 Implementation, November 2000 (ADAMS Accession No. ML003771157)

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 61 of 127 ATTACHMENT 2 APPLICATION OF WESTINGHOUSE FULL SPECTRUMTM LOCA EVALUATION MODEL TO THE TURKEY POINT NUCLEAR GENERATING PLANT (NON-PROPRIETARY VERSION)

(52 pages follow)

L-2021-071 Enclosure Page 62 of 127 APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO THE TURKEY POINT NUCLEAR GENERATING PLANT

1.0 INTRODUCTION

An analysis with the FULL SPECTRUM' 1loss-of-coolant accident (FSLOCA') evaluation model (EM) has been completed for the Turkey Point Nuclear Generating Plant. This license amendment request (LAR) for Turkey Point Units 3 and 4 requests approval to apply the Westinghouse FSLOCA EM.

The FSLOCA EM (Reference 1) was developed to address the full spectrum of loss-of-coolant accidents (LOCAs) which result from a postulated break in the reactor coolant system (RCS) of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM 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 (DEG) rupture of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as Small and Large Break LOCAs.

The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as Small Break LOCAs (SBLOCAs). Region II includes break sizes that are typically defined as Large Break LOCAs (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 (Reference 2), which explicitly models TCD and is benchmarked to high burnup data in Reference 2. The fuel pellet thermal conductivity model in the WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet TCD.

Three of the Title 10 of the Code of Federal Regulations (CFR) 50.46 criteria (peak cladding temperature (PCT), maximum local oxidation (MLO), and core-wide oxidation (CWO)) are considered directly in the FSLOCA EM. A high probability statement is developed for the PCT, MLO, and CWO that is needed to demonstrate compliance with 10 CFR 50.46 acceptance criteria (b)(1), (b)(2), and (b)(3) (Reference 3) via statistical methods. The MLO is defined as the sum of pre-transient corrosion and transient oxidation consistent with the position in Information Notice 98-29 (Reference 4). The coolable geometry acceptance criterion, 10 CFR 50.46 (b)(4), is assured by compliance with acceptance criteria (b)(1),

(b)(2), and (b)(3), and demonstrating that fuel assembly grid deformation due to combined seismic and LOCA loads does not extend to the in-board fuel assemblies such that a coolable geometry is maintained.

The FSLOCA EM has been generically approved by the Nuclear Regulatory Commission (NRC) for Westinghouse 3-loop and 4-loop plants with cold leg Emergency Core Cooling System (ECCS) injection (Reference 1). Since Turkey Point Units 3 and 4 are Westinghouse designed 3-loop plant with cold leg ECCS injection, the approved method is applicable.

This report summarizes the application of the Westinghouse FSLOCA EM to Turkey Point Units 3 and 4.

The application of the FSLOCA EM to Turkey Point Units 3 and 4 is consistent with the NRC-approved methodology (Reference 1), with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The application of the FSLOCA EM to Turkey Point Units 3 and 4 is consistent with the conditions and limitations as identified in the NRCs Safety Evaluation Report (SER).

A single analysis with the FSLOCA EM was performed for Turkey Point Units 3 and 4.

Both Florida Power & Light Company and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface FULL SPECTRUM and FSLOCA are trademarks of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 63 of 127 processes, along with Westinghouse internal processes for assessing EM changes and errors, are used to identify the need for LOCA analysis impact assessments.

The major plant parameter and analysis assumptions used in the Turkey Point Units 3 and 4 analysis with the FSLOCA EM are provided in Tables 1 through 7.

2.0 METHOD OF ANALYSIS 2.1 FULL SPECTRUM LOCA Evaluation Model Development In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 (Reference 3 and Reference 6) and Appendix K, ECCS Evaluation Models, to permit the use of a realistic EM to analyze the performance of the ECCS during a hypothetical LOCA. Westinghouses previously approved best-estimate LBLOCA EM is discussed in Reference 8. The EM is referred to as the Automated Statistical Treatment of Uncertainty Method (ASTRUM), and was developed following Regulatory Guide (RG) 1.157 (Reference 7).

When the FSLOCA EM was being developed, the NRC issued RG 1.203 (Reference 9) which expands on the principles of RG 1.157, while providing a more systematic approach to the development and assessment process of a PWR accident and safety analysis EM. Therefore, the development of the FSLOCA EM followed the Evaluation Model Development and Assessment Process (EMDAP), which is documented in RG 1.203. While RG 1.203 expands upon RG 1.157, there are certain aspects of RG 1.157 which are more detailed than RG 1.203; therefore, both RGs were used for the development of the FSLOCA EM.

2.2 WCOBRA/TRAC-TF2 Computer Code The FSLOCA EM (Reference 1) uses the WCOBRA/TRAC-TF2 code to analyze the system thermal-hydraulic response for the full spectrum of break sizes. WCOBRA/TRAC-TF2 was created by combining a 1D module (TRAC-P) with a 3D module (based on Westinghouse modified COBRA-TF). The 1D and 3D modules include an explicit non-condensable gas transport equation. The use of TRAC-P allows for the extension of a two-fluid, six-equation formulation of the two-phase flow to the 1D loop components.

This new code is WCOBRA/TRAC-TF2, where TF2 is an identifier that reflects the use of a three-field (TF) formulation of the 3D module derived by COBRA-TF and a two-fluid (TF) formulation of the 1D module based on TRAC-P.

This best-estimate computer code contains the following features:

1. Ability to model transient three-dimensional flows in different geometries inside the reactor vessel

2. Ability to model thermal and mechanical non-equilibrium between phases

3. Ability to mechanistically represent interfacial heat, mass, and momentum transfer in different flow regimes

4. Ability to represent important reactor and plant components such as fuel rods, steam generators (SGs), reactor coolant pumps (RCPs), etc.

A detailed assessment of the computer code WCOBRA/TRAC-TF2 was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena for a LOCA. Modeling of a LOCA introduces additional uncertainties which are identified and quantified in the plant-specific analysis. The reactor vessel and loop

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 64 of 127 noding scheme used in the FSLOCA EM is consistent with the noding scheme used for the experiment simulations that form the validation basis for the physical models in the code. Such noding choices have been justified by assessing the model against large and full scale experiments.

2.3 Compliance with FSLOCA EM Limitations and Conditions The NRCs SER for Reference 1 contains 15 limitations and conditions on the NRC-approved FSLOCA EM. A summary of each limitation and condition and how it was met is provided below.

Limitation and Condition Number 1 Summary The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.

Compliance The analysis for Turkey Point Units 3 and 4 with the FSLOCA EM is only being used to demonstrate compliance with 10 CFR 50.46 (b)(1) through (b)(4).

Limitation and Condition Number 2 Summary 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.

Compliance Turkey Point Units 3 and 4 are Westinghouse-designed 3-loop PWRs with cold-side injection, so they are within the NRC-approved methodology. The analysis for Turkey Point Units 3 and 4 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 (Reference 5) and LTR-NRC-19-6 (Reference 12).

Limitation and Condition Number 3 Summary 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.

Compliance The containment pressure calculation for the Turkey Point Units 3 and 4 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.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 65 of 127 Limitation and Condition Number 4 Summary The decay heat uncertainty multiplier will be [

]a,c 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.

Compliance Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was [

]a,c for the Turkey Point Units 3 and 4 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 11.

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

]a,c respectively.

Compliance The maximum analyzed assembly and rod length-average burnup were less than or equal to [

]a,c respectively, for Turkey Point Units 3 and 4.

Limitation and Condition Number 6 Summary 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.

Compliance PAD5 fuel performance data were utilized in the Turkey Point Units 3 and 4 analysis with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of Reference 2, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 2.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 66 of 127 Limitation and Condition Number 7 Summary The YDRAG uncertainty parameter should be [

]a,c Compliance Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was [

]a,c for the Turkey Point Units 3 and 4 Region I analysis.

Limitation and Condition Number 8 Summary The [

]a,c Compliance Consistent with the NRC-approved methodology, the [

]a,c for the Turkey Point Units 3 and 4 Region I analysis.

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

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

Compliance Turkey Point Units 3 and 4 are Westinghouse-designed 3-loop PWRs, so this Limitation and Condition is not applicable.

Limitation and Condition Number 10 Summary 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 [

]a,c 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.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 67 of 127 Compliance Turkey Point Units 3 and 4 are Westinghouse-designed 3-loop PWRs, so this part of the Limitation and Condition is not applicable.

The minimum sampled break area for the Turkey Point Units 3 and 4 Region II analysis was 1 ft2.

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

1. The [ ]a,c 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 [

]a,c 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, and CWO 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.

Compliance This Limitation and Condition was met for the Turkey Point Units 3 and 4 analysis as follows:

1. The [ ]a,c 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 analyses. The [

]a,c 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 Turkey Point Units 3 and 4 in Table 1.

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

Compliance A bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves (MSSVs) was modeled in the Turkey Point Units 3 and 4 analysis.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 68 of 127 Limitation and Condition Number 13 Summary In plant-specific models for analysis with the FSLOCA EM: 1) the [

]a,c and 2) the a,c

[ ]

Compliance The [

]a,c in the analysis for Turkey Point Units 3 and 4. The [

]a,c in the analysis.

Limitation and Condition Number 14 Summary 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 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.

Compliance For the Turkey Point Units 3 and 4 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%.

Limitation and Condition Number 15 Summary 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 [ ]a,c Compliance The Region II uncertainty analysis for Turkey Point Units 3 and 4 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 [

]a,c

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 69 of 127 3.0 REGION I ANALYSIS 3.1 Description of Representative Transient The small break LOCA transient can be divided into time periods in which specific phenomena are occurring, as discussed below.

Blowdown The rapid depressurization of the RCS coincides with subcooled liquid flow through the break. Following the reactor trip on the low pressurizer pressure setpoint, the pressurizer drains, and safety injection is initiated on the low pressurizer pressure SI setpoint. After reaching this setpoint and applying the safety injection delays, high pressure safety injection flow begins. Phase separation begins in the upper head and upper plenum near the end of this period until the entire RCS eventually reaches saturation, ending the rapid depressurization slightly above the steam generator secondary side pressure near the modeled MSSV setpoint.

Natural Circulation This quasi-equilibrium phase persists while the RCS pressure remains slightly above the secondary side pressure. The system drains from the top down, and while significant mass is continually lost through the break, the vapor generated in the core is trapped in the upper regions by the liquid remaining in the crossover leg loop seals. Throughout this period, the core remains covered by a two-phase mixture and the fuel cladding temperatures remain at the saturation temperature level.

Loop Seal Clearance As the system drains, the liquid levels in the downhill side of the pump suction (crossover leg) become depressed all the way to the bottom elevations of the piping, allowing the steam trapped during the natural circulation phase to vent to the break (i.e., a process called loop seal clearance). The break flow and the flow through the RCS loops become primarily vapor. Relief of a static head imbalance allows for a quick but temporary recovery of liquid levels in the inner portion of the reactor vessel.

Boil-Off With a vapor vent path established after the loop seal clearance, the RCS depressurizes at a rate controlled by the critical flow, which continues to be a primarily high quality mixture of water and steam. The RCS pressure remains high enough such that safety injection flow cannot make up for the primary system fluid inventory lost through the break, leading to core uncovery and a fuel rod cladding temperature heatup.

Core Recovery The RCS pressure continues to decrease, and once it reaches that of the accumulator gas pressure, the introduction of additional ECCS water from the accumulators replenishes the reactor vessel inventory and recovers the core mixture level. The transient is considered over as the break flow is compensated by the injected flow.

3.2 Analysis Results The Turkey Point Units 3 and 4 Region I analysis was performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The transient that produced the analysis PCT result is a cold leg break with a break diameter of 2.1-inches. The most limiting ECCS single failure of one ECCS train is assumed in the analysis as identified in Table 1. Control rod drop is modeled for breaks less than 1 square foot assuming a 2 second

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 70 of 127 signal delay time and a 2.4 second rod drop time. RCP trip is modeled coincident with reactor trip on the low pressurizer pressure setpoint for LOOP transients. When the low pressurizer pressure SI setpoint is reached, there is a delay to account for emergency diesel generator start-up, filling headers, etc., after which safety injection is initiated into the reactor coolant system.

The results of the Turkey Point Units 3 and 4 Region I uncertainty analysis are summarized in Table 8.

The sampled decay heat uncertainty multipliers for the Region I analysis cases are provided in Table 11.

Table 9 contains a sequence of events for the transient that produced the Region I analysis PCT result.

Figures 1 through 13 illustrate the calculated key transient response parameters for this transient.

4.0 REGION II ANALYSIS 4.1 Description of Representative Transient A large-break LOCA transient can be divided into phases in which specific phenomena are occurring. A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the fuel assemblies undergo. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below.

Blowdown - Critical Heat Flux (CHF) Phase In this phase, the break flow is subcooled, the discharge rate of coolant from the break is high, the core flow reverses, the fuel rods go through departure from nucleate boiling (DNB), and the cladding rapidly heats up and the reactor is shut down due to the core voiding.

The regions of the RCS with the highest initial temperatures (upper core, upper plenum, and hot legs) begin to flash during this period. This phase is terminated when the water in the lower plenum and downcomer begins to flash. The mixture level swells and a saturated mixture is pushed into the core by the intact loop RCPs, still rotating in single-phase liquid. As the fluid in the cold leg reaches saturation conditions, the discharge flow rate at the break decreases significantly.

Blowdown - Upward Core Flow Phase Heat transfer is increased as the two-phase mixture is pushed into the core. The break discharge rate is reduced because the fluid becomes saturated at the break. This phase ends as the lower plenum mass is depleted, the fluid in the loops become two-phase, and the RCP head degrades.

Blowdown - Downward Core Flow Phase The break flow begins to dominate and pulls flow down through the core as the RCP head degrades due to increased voiding, while liquid and entrained liquid flows also provide core cooling. Heat transfer in this period may be enhanced by liquid flow from the upper head. Once the system has depressurized to less than the accumulator cover pressure, the accumulators begin to inject cold water into the cold legs.

During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is bypassed around the downcomer and sent out through the break. As the system pressure continues to decrease, the break flow and consequently the downward core flow are reduced. The system pressure approaches the containment pressure at the end of this last period of the blowdown phase.

During this phase, the core begins to heat up as the system approaches containment pressure, and the phase ends when the reactor vessel begins to refill with ECCS water.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 71 of 127 Refill Phase The core continues to heat up as the lower plenum refills with ECCS water. This phase is characterized by a rapid increase in fuel cladding temperature at all elevations due to the lack of liquid and steam flow in the core region. The water completely refills the lower plenum and the refill phase ends. As ECCS water enters the core, the fuel rods in the lower core region begin to quench and liquid entrainment begins, resulting in increased fuel rod heat transfer.

Reflood Phase During the early reflood phase, the accumulators begin to empty and nitrogen is discharged into the RCS.

The nitrogen surge forces water into the core, which is then evaporated, causing system re-pressurization and a temporary reduction of pumped ECCS flow; this re-pressurization is illustrated by the increase in RCS pressure. During this time, core cooling may increase due to vapor generation and liquid entrainment, but conversely the early reflood pressure spike results in loss of mass out through the broken cold leg.

The pumped ECCS water aids in the filling of the downcomer throughout the reflood period. As the quench front progresses further into the core, the PCT elevation moves increasingly higher in the fuel assembly.

As the transient progresses, continued injection of pumped ECCS water refloods the core, effectively removes the reactor vessel metal mass stored energy and core decay heat, and leads to an increase in the reactor vessel fluid mass. Eventually the core inventory increases enough that liquid entrainment is able to quench all the fuel assemblies in the core.

4.2 Analysis Results The Turkey Point Units 3 and 4 Region II analysis was performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The analysis was performed assuming both LOOP and OPA, and the results of both of the LOOP and OPA analyses are compared to the 10 CFR 50.46 acceptance criteria. The most limiting ECCS single failure of one ECCS train is assumed in the analysis as identified in Table 1. The results of the Turkey Point Units 3 and 4 Region II LOOP and OPA uncertainty analyses are summarized in Table 8.

The sampled decay heat uncertainty multipliers for the Region II analysis cases are provided in Table 11.

Table 10 contains a sequence of events for the transient that produced the more limiting analysis PCT result relative to the offsite power assumption. Figures 14 through 27 illustrate the key response parameters for this transient.

The containment pressure is calculated for each LOCA transient in the analysis using the COCO code (References 10 and 11). The COCO containment code is integrated into the WCOBRA/TRAC-TF2 thermal-hydraulic code. The transient-specific mass and energy releases calculated by the thermal-hydraulic code at the end of each timestep are transferred to COCO. COCO then calculates the containment pressure based on the containment model (the inputs are summarized in Tables 2, 3, and 4) and the mass and energy releases, and transfers the pressure back to the thermal-hydraulic code as a boundary condition at the break, consistent with the methodology in Reference 1. The containment model for COCO calculates a conservatively low containment pressure, including the effects of all the installed pressure reducing systems and processes such as assuming all trains of containment spray are operable and assuming fan cooler operation. The containment backpressure for the transient that produced the analysis PCT result is provided in Figure 21.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 72 of 127 5.0 COMPLIANCE WITH 10 CFR 50.46 It must be demonstrated that there is a high level of probability that the following criteria in 10 CFR 50.46 are met:

(b)(1) The analysis PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT is less than 2,200°F, the analysis with the FSLOCA EM confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., Peak Cladding Temperature does not exceed 2,200°F, is demonstrated.

The results are shown in Table 8 for Turkey Point Units 3 and 4.

(b)(2) The analysis 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 10 CFR 50.46 acceptance criterion (b)(2), i.e., Maximum Local Oxidation of the cladding does not exceed 17 percent, is demonstrated.

The results are shown in Table 8 for Turkey Point Units 3 and 4.

(b)(3) The analysis 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 10 CFR 50.46 acceptance criterion (b)(3), i.e., Core-Wide Oxidation does not exceed 1 percent, is demonstrated.

The results are shown in Table 8 for Turkey Point Units 3 and 4.

(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains in a coolable geometry.

This criterion is met by demonstrating compliance with criteria (b)(1), (b)(2), and (b)(3), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. Criteria (b)(1), (b)(2), and (b)(3) have been met for Turkey Point Units 3 and 4 as shown in Table 8.

It is discussed in Section 32.1 of the NRC-approved FSLOCA EM (Reference 1) 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 combined LOCA and seismic loads is not calculated to occur for Turkey Point Units 3 and 4.

(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS.

Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM (Reference 1).

Based on the analysis results for Region I and Region II presented in Table 8 for Turkey Point Units 3 and 4, it is concluded that Turkey Point Units 3 and 4 comply with the criteria in 10 CFR 50.46.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 73 of 127

6.0 REFERENCES

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

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

3. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974.

4. Information Notice 98-29: Predicted Increase in Fuel Rod Cladding Oxidation, USNRC, August 1998.

5. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017, LTR-NRC-18-30, July 2018.

6. Emergency Core Cooling Systems: Revisions to Acceptance Criteria, Federal Register, V53, N180, pp. 35996-36005, September 1988.

7. Best Estimate Calculations of Emergency Core Cooling System Performance, Regulatory Guide 1.157, USNRC, May 1989.

8. Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM), WCAP-16009-P-A, January 2005.

9. Transient and Accident Analysis Methods, Regulatory Guide 1.203, USNRC, December 2005.

10. Westinghouse Emergency Core Cooling System Evaluation Model - Summary, WCAP-8339, June 1974.

11. Containment Pressure Analysis Code (COCO), WCAP-8327, June 1974.

12. U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018, LTR-NRC-19-6, February 2019.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 74 of 127 Table 1. Plant Operating Range Analyzed and Key Parameters for Turkey Point Units 3 and 4 Parameter As-Analyzed Value or Range 1.0 Core Parameters a) Core power 2652 MWt +/- 0% Uncertainty b) Fuel type 15x15 Upgrade Fuel, Optimized ZIRLOTM1Cladding with Intermediate Flow Mixers (IFMs), Integral Fuel Burnable Absorbers (IFBA) or Non-IFBA c) Maximum total core peaking factor (FQ), 2.4 including uncertainties d) Maximum hot channel enthalpy rise peaking 1.65 factor (FH), including uncertainties e) Axial flux difference (AFD) band at 100% -13% / +10%

power f) Maximum transient operation fraction 0.25 2.0 Reactor Coolant System Parameters a) Thermal design flow (TDF) 86,900 gpm/loop b) Vessel average temperature (TAVG) 564°F TAVG 589°F c) Pressurizer pressure (PRCS) 2197 psia PRCS 2303 psia d) Reactor coolant pump (RCP) model and power Model 93, 6000 hp 3.0 Containment Parameters a) Containment modeling Region I: Constant pressure equal to initial containment pressure Region II: Calculated for each transient using transient-specific mass and energy releases and the information in Tables 2, 3, and 4 4.0 Steam Generator (SG) and Secondary Side Parameters a) Steam generator tube plugging level 15%

b) Main steam safety valve (MSSV) nominal set Table 7 pressures, uncertainty and accumulation c) Main feedwater temperature Nominal (422.5°F)

Optimized ZIRLO is a 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.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 75 of 127 Table 1. Plant Operating Range Analyzed and Key Parameters for Turkey Point Units 3 and 4 Parameter As-Analyzed Value or Range d) Auxiliary feedwater temperature Nominal (69.5°F) e) Auxiliary feedwater flow rate 66.7 gpm/SG 5.0 Safety Injection (SI) Parameters a) Single failure configuration ECCS: Loss of one train of pumped ECCS Region II containment pressure: All containment spray trains are available b) Safety injection temperature (TSI) 34°F TSI 105°F c) Low pressurizer pressure safety injection safety 1615 psia analysis limit d) Initiation delay time from low pressurizer 23 seconds (OPA) or 35 seconds (LOOP) pressure SI setpoint to full SI flow e) Safety injection flow Minimum flows assuming 1 HHSI pump in Tables 5 and 5a (Region I) or Table 6 (Region II) 6.0 Accumulator Parameters a) Accumulator temperature (TACC) 80°F TACC 130°F b) Accumulator water volume (VACC) 865 ft3 VACC 920 ft3 c) Accumulator pressure (PACC) 575 psig PACC 700 psig d) Accumulator boron concentration 2300 ppm 7.0 Reactor Protection System Parameters a) Low pressurizer pressure reactor trip signal 2 seconds processing time b) Low pressurizer pressure reactor trip setpoint 1805 psia 8.0 Refueling Water Storage Tank (RWST) / Switchover Parameters a) Usable RWST volume 260,000 gallons b) Interruption time for switchover to cold leg 120 seconds recirculation c) SI flow after switchover to cold leg Table 5a recirculation d) SI temperature after switchover to cold leg 205°F recirculation

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 76 of 127 Table 2. Containment Data Used for Region II Calculation of Containment Pressure for Turkey Point Units 3 and 4 Parameter Value Maximum containment net free volume 1,600,000 ft3 Minimum initial containment temperature at full power operation 80°F Refueling water storage tank (RWST) temperature for containment spray 34°F (TRWST)

Minimum RWST temperature for broken loop spilling SI 34°F Minimum containment outside air / ground temperature 39°F Minimum initial containment pressure at normal full power operation 14.1 psia Minimum containment spray pump initiation delay from containment high 11 seconds (OPA) or pressure signal time 27 seconds (LOOP)

Maximum containment spray flow rate from all pumps 3,520 gpm Maximum number of containment fan coolers in operation during LOCA 2 transient Minimum fan cooler initiation delay time 11 seconds (OPA) or 26 seconds (LOOP)

Maximum heat removal rate per fan cooler as a function of containment Table 3 temperature Maximum number of containment venting lines (including purge lines, 3 pressure relief lines or any others) which can be OPEN at onset of transient at full power operation Maximum effective valve diameter of each containment venting line 48 inches limited to 35° disc angle 54 inches limited to 30° disc angle 2 inches Maximum containment pressure setpoint for venting valve closure 20.7 psia Maximum delay time between reaching containment pressure setpoint and 1 sec start of venting valve closure Maximum venting valve closure time at normal full power operation 5 sec Containment walls / heat sink properties Table 4 SI spilling flows 431.6 lbm/sec

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 77 of 127 Table 3. Fan Cooler Performance Data Used for Region II Calculation of Containment Pressure for Turkey Point Units 3 and 4 Containment Temperature Heat Removal Rate Heat Removal Rate

(°F) (MBTU/hr) (BTU/sec) 110 0.0 0.0 120 1.554 432.0 140 3.770 1047.0 160 6.906 1918.0 180 11.221 3117.0 200 17.353 4820.0 220 25.352 7042.0 240 36.492 10137.0 260 50.880 14133.0 283 73.125 20313.0

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 78 of 127 Table 4. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Turkey Point Units 3 and 4 Wall Area (ft2) Thickness (ft) Material 1 1,850.50 0.5000 Concrete 2 0.5000 Concrete 7,848.53 0.5000 Concrete 3 0.5000 Concrete 27,484.39 0.5000 Concrete 4 0.5000 Concrete 6,079.92 0.5000 Concrete 5 0.5000 Concrete 5,742.24 0.5000 Concrete 6 0.0208 Carbon Steel 16,921.00 0.5000 Concrete 0.5000 Concrete 7 0.0208 Carbon Steel 45,292.66 0.5000 Concrete 0.5000 Concrete 8 0.0400 Carbon Steel 2,037.62 0.5000 Concrete 0.5000 Concrete 9 0.0100 Stainless Steel 881.93 0.5000 Concrete 0.5000 Concrete 10 0.0310 Stainless Steel 66.29 0.5000 Concrete 0.5000 Concrete 11 0.0052 Stainless Steel 5,514.10 0.5000 Concrete 0.5000 Concrete 12 0.0156 Stainless Steel 816.67 0.5000 Concrete 0.5000 Concrete 13 0.0208 Stainless Steel 2,430.58 0.5000 Concrete 0.5000 Concrete

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 79 of 127 Table 4. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Turkey Point Units 3 and 4 Wall Area (ft2) Thickness (ft) Material 14 0.0258 Stainless Steel 107.65 0.5000 Concrete 0.5000 Concrete 15 0.0208 Stainless Steel 184.50 0.5000 Concrete 0.5000 Concrete 16 7,698.81 0.5000 Concrete 17 0.0420 Carbon Steel 473.00 0.5000 Concrete 18 0.0420 Carbon Steel 231.00 0.5000 Concrete 0.5000 Concrete 19 0.0420 Carbon Steel 174.00 0.5000 Concrete 0.5000 Concrete 20 674.50 0.5000 Concrete 21 0.0208 Stainless Steel 16.00 0.5000 Concrete 22 7,311.53 0.0047 Admiralty 23 240.12 0.0125 Aluminum 24 1,464.49 0.0250 Aluminum 25 151,574.63 0.0068 Carbon Steel 26 81,251.54 0.0140 Carbon Steel 27 27,887.12 0.0250 Carbon Steel 28 14,548.27 0.0339 Carbon Steel 29 13,606.81 0.0454 Carbon Steel 30 3,710.52 0.0544 Carbon Steel 31 1,495.49 0.0623 Carbon Steel 32 635.75 0.07442 Carbon Steel 33 1,546.10 0.0879 Carbon Steel 34 3,263.02 0.0972 Carbon Steel

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 80 of 127 Table 4. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Turkey Point Units 3 and 4 Wall Area (ft2) Thickness (ft) Material 35 3,420.47 0.1385 Carbon Steel 36 4,265.44 0.1651 Carbon Steel 37 3,716.66 0.2668 Carbon Steel 38 23,827.86 0.0067 Stainless Steel 39 2,571.30 0.0155 Stainless Steel 40 3,208.46 0.0258 Stainless Steel 41 798.42 0.0367 Stainless Steel 42 1,060.78 0.0493 Stainless Steel 43 83.26 0.0514 Stainless Steel 44 1,043.11 0.0626 Stainless Steel 45 235.38 0.0899 Stainless Steel 46 366.46 0.0998 Stainless Steel 47 4.09 0.1500 Stainless Steel 48 41.45 0.2671 Stainless Steel 49 1,255.16 0.0167 Pooling Carbon Steel 50 230.02 0.0243 Pooling Carbon Steel 51 182.92 0.0993 Pooling Carbon Steel 52 5,713.79 0.0057 Pooling Stainless Steel 53 28.80 0.0375 Pooling Stainless Steel 54 12,037.96 0.0154 Galvanized Steel (Use Carbon Steel) 55 3,259.09 0.0243 Galvanized Steel (Use Carbon Steel) 56 802.88 0.7202 Carbon Steel 57 28.61 0.4500 Stainless Steel 58 7,775.54 0.0061 Copper/Nickel 59 204,691.28 0.0008 Copper

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 81 of 127 Table 5. Safety Injection Flow Used for Region I Calculation for Turkey Point Units 3 and 4 High Head Safety Injection Low Head Safety Injection Pressure (psia) (HHSI) Flow (gpm)

(LHSI) Flow (gpm)

(assuming 1 HHSI pump) 14.7 394 1918 24.7 392 1840 34.7 391 1759 44.7 389 1675 54.7 387 1586 64.7 385 1493 74.7 383 1396 84.7 381 1287 90.0 380 1226 94.7 379 1171 104.7 378 1039 114.7 376 870 134.7 372 384 214.7 357 0 314.7 338 0 414.7 318 0 514.7 299 0 614.7 279 0 714.7 256 0 814.7 231 0 914.7 206 0 1014.7 179 0 1114.7 145 0 1214.7 103 0 1314.7 40 0

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 82 of 127 Table 5a. Safety Injection Flow After Switchover to Cold Leg Recirculation Used for Region I Calculation for Turkey Point Units 3 and 4 High Head Safety Injection (HHSI) Flow (gpm)

Pressure (psia)

(assuming 1 HHSI pump) 14.7 414 34.7 411 54.7 408 74.7 404 94.7 400 114.7 397 134.7 393 154.7 390 214.7 379 314.7 361 414.7 342 514.7 322 614.7 304 714.7 283 814.7 259 914.7 234 1014.7 208 1114.7 177 1214.7 139 1314.7 83

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 83 of 127 Table 6. Safety Injection Flow Used for Region II Calculation for Turkey Point Units 3 and 4 High Head Safety Injection Low Head Safety Injection Pressure (psia) (HHSI) Flow (gpm)

(LHSI) Flow (gpm)

(assuming 1 HHSI pump) 14.7 394 1914 19.7 392 1876 24.7 390 1836 29.7 387.5 1370 34.7 385 942 39.7 383 540 44.7 381 155 46.7 380 7 54.7 376 0 64.7 372 0 74.7 367 0 84.7 363 0 90.0 361 0 94.7 358 0 104.7 354 0 114.7 349 0 214.7 304 0 314.7 257 0 414.7 208 0 514.7 156 0 614.7 103 0 714.7 48 0

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 84 of 127 Table 7. Steam Generator Main Steam Safety Valve Parameters for Turkey Point Units 3 and 4 Stage Set Pressure (psig) Uncertainty (%) Accumulation (%)

1 1085 +/-3 +/-3 2 1100 +/-3 +/-3 3 1105 +/-3 +/-3 4 1105 +/-3 +/-3 Table 8. Turkey Point Units 3 and 4 Analysis Results with the FSLOCA EM Region I Value Region II Value Region II Value Outcome (OPA) (LOOP) 95/95 PCT 1475°F 1947°F 1981°F 95/95 MLO 8.02% 10.91% 11.06%

95/95 CWO 0.06% 0.73% 0.84%

Table 9. Turkey Point Units 3 and 4 Sequence of Events for the Region I Analysis PCT Case Event Time after Break (sec)

Start of Transient 0.0 Reactor Trip Signal 22.8 Safety Injection Signal 35.5 Safety Injection Begins 70.5 Loop Seal Clearing Occurs 778 Top of Core Uncovered 1586 Accumulator Injection Begins 2774 PCT Occurs 2777 Top of Core Recovered 4602

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 85 of 127 Table 10. Turkey Point Units 3 and 4 Sequence of Events for the Region II Analysis PCT Case Event Time after Break (sec)

Start of Transient 0.0 Fuel Rod Burst Occurs 3.7 Safety Injection Signal 4.3 Accumulator Injection Begins 7.5 End of Blowdown 14.5 PCT Occurs 38.5 Safety Injection Begins 39.3 Accumulator Empty 52.5 All Rods Quenched 363 Table 11. Turkey Point Units 3 and 4 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases Region Case DECAY_HT (units of ) DECAY_HT (absolute units)*

PCT + 0.7396 3.71%

Region I MLO + 0.7396 3.71%

CWO + 0.7396 3.71%

PCT + 0.8658 4.10%

Region II (OPA) MLO + 0.4088 2.06%

CWO + 0.3244 1.55%

PCT + 0.8450 4.14%

Region II MLO + 0.9357 4.57%

(LOOP)

CWO + 0.0273 0.13%

• Approximate uncertainty in total decay heat power at 1 second after shutdown as defined by the ANSI/ANS-5.1-1979 decay heat standard for 235U, 239Pu, and 238U assuming infinite operation.

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 86 of 127 Figure 1: Turkey Point Units 3 and 4 Break Flow Void Fraction for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 87 of 127 Figure 2: Turkey Point Units 3 and 4 Total Safety Injection Flow (not including Accumulator Injection Flow) and Total Break Flow for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 88 of 127 Figure 3: Turkey Point Units 3 and 4 RCS Pressure for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 89 of 127 Figure 4: Turkey Point Units 3 and 4 Hot Assembly Two-Phase Mixture Level (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 90 of 127 Figure 5: Turkey Point Units 3 and 4 Peak Cladding Temperature for all Rods for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 91 of 127 Figure 6: Turkey Point Units 3 and 4 Vapor Mass Flow Rate through the Crossover Legs for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 92 of 127 Figure 7: Turkey Point Units 3 and 4 Collapsed Liquid Level for Each Core Channel (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 93 of 127 Figure 8: Turkey Point Units 3 and 4 Accumulator Injection Flow for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 94 of 127 Figure 9: Turkey Point Units 3 and 4 Vessel Fluid Mass for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 95 of 127 Figure 10: Turkey Point Units 3 and 4 Steam Generator Secondary Side Pressure for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 96 of 127 Figure 11: Turkey Point Units 3 and 4 Normalized Core Power Shapes for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 97 of 127 Figure 12: Turkey Point Units 3 and 4 Relative Core Power for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 98 of 127 Figure 13: Turkey Point Units 3 and 4 Vapor Temperature and Void Fraction at Core Outlet (Hot Assembly Channel) for the Region I Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 99 of 127 Figure 14: Turkey Point Units 3 and 4 Peak Cladding Temperature for all Rods for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 100 of 127 Figure 15: Turkey Point Units 3 and 4 Peak Cladding Temperature Elevation (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 101 of 127 Figure 16a: Turkey Point Units 3 and 4 Vessel-Side Break Mass Flow Rate for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 102 of 127 Figure 16b: Turkey Point Units 3 and 4 Pump-Side Break Mass Flow Rate for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 103 of 127 Figure 17: Turkey Point Units 3 and 4 Lower Plenum Collapsed Liquid Level (Relative to Inside Bottom of Vessel) for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 104 of 127 Figure 18: Turkey Point Units 3 and 4 Vapor Mass Flow Rate per Assembly at the Top Cell Face of the Core Average Channel not Under Guide Tubes for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 105 of 127 Figure 19: Turkey Point Units 3 and 4 RCS Pressure for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 106 of 127 Figure 20: Turkey Point Units 3 and 4 Accumulator Injection Flow per Loop for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 107 of 127 Figure 21: Turkey Point Units 3 and 4 Containment Pressure for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 108 of 127 Figure 22: Turkey Point Units 3 and 4 Vessel Fluid Mass for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 109 of 127 Figure 23: Turkey Point Units 3 and 4 Collapsed Liquid Level for Each Core Channel (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 110 of 127 Figure 24: Turkey Point Units 3 and 4 Average Downcomer Collapsed Liquid Level (Relative to the Bottom of the Upper Tie Plate) for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 111 of 127 Figure 25: Turkey Point Units 3 and 4 Safety Injection Flow per Loop (not including Accumulator Injection Flow) for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 112 of 127 Figure 26: Turkey Point Units 3 and 4 Normalized Core Power Shapes for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

L-2021-071 Enclosure Page 113 of 127 Figure 27: Turkey Point Units 3 and 4 Relative Core Power for the Region II Analysis PCT Case

• • • This record was final approved on 9/11/2020 2:16:49 PM. (This statement was added by the PRIME system upon its validation)

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 114 of 127 ATTACHMENT 3 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE (3 pages follow)

Westinghouse Non-Proprietary Class 3 CAW-21-5171 Page 1 of 3 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Camille T. Zozula, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of L-2021-071 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable

Westinghouse Non-Proprietary Class 3 CAW-21-5171 Page 2 of 3 others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

(5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

Westinghouse Non-Proprietary Class 3 CAW-21-5171 Page 3 of 3 (6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

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

01 April 2021 Executed on: _______________ ____________________

C ill T Camille T. Z l M Zozula, Manager Regulatory Compliance & Corporate Licensing

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 118 of 127 ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION PAGES (MARKUP)

(3 pages follow)

L-2021-071 Enclosure Page 119 of 127 This page is provided for information only. No changes are proposed to this page.

ADMINISTRATIVE CONTROLS PEAKING FACTOR LIMIT REPORT 6.9.1.6 The W(Z) function(s) for Base-Load Operation corresponding to a 2% band about the target flux difference and/or a 3% band about the target flux difference, the Load-Follow function Fz(Z) and the augmented surveillance turnon power fraction PT shall be provided to the U.S. Nuclear Regulatory Commission, whenever PT is <1.0. In the event, the option of Baseload Operation (as defined in Section 4.2.2.3) will not be exercised, the submission of the W(Z) function is not required. Should these values (i.e., W(Z), Fz(Z) and PT) change requiring a new submittal or an amended submittal to the Peaking Factor Limit Report, the Peaking Factor Limit Report shall be provided to the NRC Document Control desk with copies to the Regional Administrator and the Resident Inspector within 30 days of their implementation, unless otherwise approved by the Commission.

The analytical methods used to generate the Peaking Factor limits shall be those previously reviewed and approved by the NRC. If changes to these methods are deemed necessary they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.

CORE OPERATING LIMITS REPORT 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

1. Reactor Core Safety Limits for Specification 2.1.1.

2. Overtemperature T, Note 1 of Table 2.2-1 for Specification 2.2.1, determination of values K1, K2, K3, T', P', 1, 2, 3, 4, 5, 6, and the breakpoint and slope values for the f1 (I).

3. Overpower T, Note 3 of Table 2.2-1 for Specification 2.2.1, determination of values for K4, K5, K6, T, 7 and f2 (I).

4. Shutdown Margin - Tavg >200°F for Specification 3/4.1.1.1.

5. Shutdown Margin - Tavg <200°F for Specification 3/4.1.1.2.

6. Moderator Temperature Coefficient for Specification 3/4.1.1.3.

7. Axial Flux Difference for Specification 3.2.1.

8. Control Rod Insertion Limits for Specification 3.1.3.6.

9. Heat Flux Hot Channel Factor - FQ(Z) for Specification 3/4.2.2.

10. All Rods Out position for Specification 3.1.3.2.

11. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

12. DNB Parameters for Specification 3.2.5, determination of values for Reactor Coolant System Tavg and Pressurizer Pressure.

The analytical methods used to determine the AFD limits shall be those previously reviewed and approved by the NRC in:

1. WCAP-10216-P-A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION, June 1983.

2. WCAP-8385, POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES -

TOPICAL REPORT, September 1974.

TURKEY POINT - UNITS 3 & 4 6-17 AMENDMENT NOS. 263 AND 258

L-2021-071 Enclosure Page 120 of 127 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

The analytical methods used to determine FQ (Z), FH and the K(Z) curve shall be those previously reviewed and approved by the NRC in:

1. WCAP-9220-P-A, Rev. 1, Westinghouse ECCS Evaluation Model - 1981 Version, February 1982.

2. WCAP-10054-P-A, (proprietary), Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985.

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

TURKEY POINT - UNITS 3 & 4 6-18 AMENDMENT NOS. 263 AND 258

L-2021-071 Enclosure Page 121 of 127 ADMINISTRATIVE CONTROLS

3. WCAP-10054-P-A, Addendum 2, Revision 1 (proprietary), Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, July 1997.

4. WCAP-16009-P-A, Realistic Large-break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005.

5. USNRC Safety Evaluation Report, Letter from R. C. Jones (USNRC) to N. J. Liparulo (W),

Acceptance for Referencing of the Topical Report WCAP-12945(P) Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Analysis, June 28, 1996.**

6. Letter dated June 13, 1996, from N. J. Liparulo (W) to Frank R. Orr (USNRC), Re-Analysis Work Plans Using Final Best Estimate Methodology.**

2.

7. WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, S. L. Davidson and

3. T. L. Ryan, April 1995.

8. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO', July 2006.

The analytical methods used to determine Overtemperature T and Overpower T shall be those previously reviewed and approved by the NRC in:

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

2. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 The analytical methods used to determine Safety Limits, Shutdown Margin - Tavg > 200°F, Shutdown Margin -

Tavg < 200°F, Moderator Temperature Coefficient, DNB Parameters, Rod Bank Insertion Limits and the All Rods Out position shall be those previously reviewed and approved by the NRC in:

1. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.

The analytical methods used to support the suspension of the measurement of the Moderator Temperature Coefficient in accordance with Surveillance Requirement 4.1.1.3.b shall be those previously reviewed and approved by the NRC in:

1. WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 1997.

2. WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.

3. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004.

4. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology,"

August 2007.

• • As evaluated in NRC Safety Evaluation dated December 20, 1997.

TURKEY POINT - UNITS 3 & 4 6-19 AMENDMENT NOS. 271 AND 266

Turkey Point Nuclear Plant L-2021-071 Docket Nos. 50-250 and 50-251 Enclosure Page 122 of 127 ATTACHMENT 5 PROPOSED TECHNICAL SPECIFICATION BASES PAGES (MARKUP)

(5 pages follow)

L-2021-071 Enclosure Page 123 of 127 REVISION NO.: PROCEDURE TITLE: PAGE:

41 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 72 of 238 PROCEDURE NO.:

0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 53 of 219) 3/4.2 Power Distribution Limits The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) Maintaining the minimum DNBR in the core greater than or equal to the applicable design limit during normal operation and in short-term transients, and (2) Limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200F is NOT exceeded.

The definitions of certain hot channel and peaking factors as used in 10 CFR 50.46 these specifications are as follows:

acceptance criteria are FQ(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; FNH Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and FXY(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 Axial Flux Difference The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F Q(Z) limit defined in the CORE OPERATING LIMITS REPORT times the normalized axial peaking factor is NOT exceeded during either normal operation or in the event of xenon redistribution following power changes.

L-2021-071 Enclosure Page 124 of 127 REVISION NO.: PROCEDURE TITLE: PAGE:

41 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 74 of 238 PROCEDURE NO.:

0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 55 of 219) 3/4.2.1 (Continued)

Operation outside of the target band for the short time period allowed (15 minutes) will NOT result in significant xenon redistribution such that the envelope of peaking factors will change sufficiently to prohibit continued operation in the power region defined above. To assure that there is NO residual xenon redistribution impact from past operation on the Base Load operation, a 24-hour waiting period within a defined range of PT and AFD allowed by RAOC is necessary.

During this period, load changes and rod motion are restricted to that allowed by the Base Load requirement. After the waiting period, extended Base Load operation is permissible.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitoring Alarm.

The computer monitors the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are: 1) Outside the acceptable AFD (for RAOC operation), or 2) Outside the acceptable AFD target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation), or

2) PT (Base Load operation). Penalty deviation minutes for Base Load operation are NOT accumulated based on the short time period during which operation outside of the target band is allowed.

3/4.2.2 10 CFR 50.46 acceptance criteria are NOT exceeded.

3/4.2.3 Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor The limits on Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor ensure that: (1) The design limits on peak local power density and minimum DNBR are NOT exceeded, and (2) In the event of a LOCA the peak fuel clad temperature will NOT exceed the 2200°F ECCS acceptance criteria limit. The LOCA peak fuel clad temperature limit may be sensitive to the number of steam generator tubes plugged.

FQ(Z), Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux.

L-2021-071 Enclosure Page 125 of 127 REVISION NO.: PROCEDURE TITLE: PAGE:

41 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 146 of 238 PROCEDURE NO.:

0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases Note that the FSLOCA EM (Page 127 of 219) analysis models only one SI 3/4.5.2 pump and one RHR pump for

& and remaining 10 CFR the full spectrum of breaks 3/4.5.3 ECCS Subsystems 50.46 acceptance criteria (Region I and Region II).

The OPERABILITY of ECCS components and flowpaths required in MODES 1, 2, and 3 ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming any single active failure consideration. Two SI pumps and one RHR pump operating in conjunction with two accumulators are capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all pipe break sizes up to and including the maximum hypothetical accident of a circumferential rupture of a reactor coolant loop. The integrity of the cold leg injection flowpath can be impacted by the opposite unit if a discharge path is opened in a low pressure condition. This is NOT normally a concern based on the opposite unit operating at 2235 psig maintaining cold leg injection check valves closed. In addition, the RHR subsystem provides long-term core cooling capability in the RECIRCULATION mode during the accident recovery period.

Management of gas voids is important to ECCS OPERABILITY.

Motor Operated Valves (MOVs) 862A, 862B, 863A, 863B are required to take suction from the containment sump via the RHR System.

PC-600 supplies controlling signals to valves MOVs 862B and 863B, to prevent opening these valves if RHR Pump B discharge pressure is above 210 psig. PC-601 provides similar functions to valves MOVs 862A and 863A. Although all four valves are normally locked in position, with power removed, the capability to power up and stroke the valves must be maintained in order to satisfy the requirements for OPERABLE flow paths (capable of taking suction from the containment sump).

When PC-600/-601 are calibrated, a test signal is supplied to each circuit to check operation of the relays and annunciators operated by subject controllers. This test signal will prevent MOVs 862A, 862B, 863A, 863B from opening. Therefore, it is appropriate to tag out the MOV breakers, and enter Technical Specification Action Statement 3.5.2.a. and 3.6.2.1 when calibrating PC-600/-601.

With the RCS temperature below 350F, operation with less than full redundant equipment is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

L-2021-071 Enclosure Page 126 of 127 REVISION NO.: PROCEDURE TITLE: PAGE:

41 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 160 of 238 PROCEDURE NO.:

0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 141 of 219) 3/4.6.1.3 (Continued)

At this point the door is still physically restrained from opening, but the seating pressure against the o-ring seal may have been reduced such that the door seal is in an untested configuration, potentially creating a leakage path. In this configuration, the door is considered closed per the Technical Specifications and would satisfy the interlock test requirements, but the overall air lock leakage requirement may have been invalidated. This configuration would result in an inoperable airlock door since the O-ring seal was NOT properly compressed. As there is NO functional difference between an unsecured door and a leaking door (as far as maintenance of containment integrity is concerned), the unsecured door must be considered inoperable.

In the event the air lock leakage results in exceeding the overall containment leakage rate, the Note directs entry into the ACTION of LCO 3.6.1.2, "Containment Leakage."

3/4.6.1.4 Internal Pressure The limitations on Containment Internal Pressure ensure that: (1) The containment structure is prevented from exceeding its design negative Add new pressure differential of 2.5 psig with respect to the outside atmosphere, paragraph from and (2) The containment peak pressure does NOT exceed the design INSERT A pressure of 55 psig during LOCA conditions.

3/4.6.1.5 Air Temperature The limitations on containment average air temperature ensure that the design limits for a LOCA are NOT exceeded, and that the environmental qualification of equipment is NOT impacted. If temperatures exceed 120°F, but remain below 125°F for up to 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> during a calendar year, NO action is required. If the 336-hour limit is approached, an evaluation may be performed to extend the limit if some of the hours have been spent at less than 125°F. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air Add new temperature.

paragraph from INSERT B

L-2021-071 Enclosure Page 127 of 127 INSERT A - Add to end of TS Bases 3/4.6.1.4 The FSLOCATM EM analysis models a minimum initial containment pressure calculated based on plant data (14.1 psia) rather than the TS minimum containment pressure. This is consistent with the approved methodology.

INSERT B - Add to end of TS Bases 3/4.6.1.5 The FSLOCATM EM analysis assumes a minimum initial containment temperature to model a conservatively low pressure in the analysis, which does not directly correlate with the maximum containment temperatures described in this TS.