ML15002A091

St. Lucie, Unit 2 - Application for Technical Specification Change and Exemption Request Regarding the Transitioning to Areva Fuel

ML15002A091
Person / Time
Site:	Saint Lucie
Issue date:	12/30/2014
From:	Jensen J Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15002A105	List: L-2014-366, ANP-3352NP, Revision 0, St. Luice, Unit 2, Fuel Transition License Amendment Request, Technical Report L-2014-366, St. Lucie, Unit 2 - Application for Technical Specification Change and Exemption Request Regarding the Transitioning to Areva Fuel ML15002A091
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L-2014-366
Download: ML15002A091 (62)

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0 December 30, 2014 L-2014-366 FPL. 10 CFR 50.90 10 CFR 50.12 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 2 Docket No. 50-389 Application for Technical Specification Change and Exemption Request Regarding the Transitioning to AREVA Fuel Pursuant to 10 CFR 50.90 and 50.12, Florida Power & Light Company (FPL) requests to amend Renewed Facility Operating License NPF- 16 and ask for exemption from the regulation for St.Lucie Unit 2.The proposed license amendment request will revise the Technical Specifications (TS) to allow the use of AREVA fuel at St. Lucie Unit 2. Additionally, pursuant to 10 CFR 50.12, FPL requests an exemption from the provisions of 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," and Appendix K to 10 CFR 50, "ECCS Evaluation Models" to allow the use of M5 fuel rod cladding in future core reload applications for St. Lucie Unit 2.Enclosure 1 provides the evaluation of the proposed changes, including the description of the proposed changes, the no significant hazards consideration determination, the existing marked up TS and COLR pages showing the proposed changes, an informational copy of the TS Bases changes, the technical evaluation of the new fuel, and the revised (clean) TS pages. Attachment 1 provides the request for exemption from IOCFR50.46 and Appendix K to 10 CFR 50 requirements for the use of M5 clad for the new fuel. Attachment 2 provides the AREVA affidavit for withholding the AREVA proprietary technical reports from the public. Attachments 3, 4, 5, and 6 provide the supporting AREVA proprietary technical reports referenced in the technical evaluation within Enclosure

1. Attachment 7 provides the non-proprietary version of the AREVA technical reports.FPL requests approval of the proposed amendment by March 31, 2016, with implementation upon the start of the SL2-23 spring 2017 refueling outage to support the AREVA fuel transition project plan.This license amendment proposed by FPL has been reviewed by the St. Lucie Plant Onsite Review Group. In accordance with 10 CFR 50.91(b)(1), a copy of the proposed license amendment is being forwarded to the State Designee for the State of Florida.4oo Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957 L-2014-366 Page 2 Please contact Mr. Eric Katzman, Licensing Manager at 772-467-7734 if there are any questions about this submittal.

I declare under penalty of perjury that the foregoing is true and correct.Executed on December B3O ,2014.rsn ensen Site Vice President St. Lucie Plant

Enclosure:

1. Evaluation of Proposed Changes Attachments:

1. 1 OCFR50.46 and 10 CFR 50 Appendix K Exemption Request 2. AREVA affidavit for withholding proprietary information from the public 3. AREVA proprietary report ANP-3352P, Revision 0, St. Lucie Unit 2 Fuel Transition License Amendment Request Technical Report 4. AREVA proprietary report ANP-3347P, Revision 0, St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report 5. AREVA proprietary report ANP-3345P, Revision 0, St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report 6. AREVA proprietary report ANP-3346P, Revision 0, St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report 7. Non-proprietary versions of the AREVA reports cc: Ms. Cynthia Becker, Florida Bureau of Radiation Control NRC Region II Administrator NRC Site Resident Inspector L-2014-366 Enclosure 1 Page 1 of 53 Enclosure 1 Evaluation of the Proposed Change

Subject:

License Amendment Request AREVA Fuel Transition

2.0 PROPOSED CHANGE

S 3.0 JUSTIFICATION OF PROPOSED CHANGES 4.0 REGULATORY ANALYSIS -No Significant Hazards Consideration Determination

5.0 ENVIRONMENTAL

EVALUATION

6.0 CONCLUSION

S

7.0 REFERENCES

ATTACHMENTS:

1. Technical Evaluation of Changes 2. Technical Specification Markups 3. Technical Specification Bases Markup 4. Core Operating Limits Report (COLR) Markups 5. Word Processed Technical Specifications L-2014-366 Enclosure 1 Page 2 of 53

1.0 DESCRIPTION

St. Lucie Unit 2 is planning on transitioning to the AREVA fuel beginning with Cycle 23.St. Lucie Unit 2 is currently operating in Cycle 21 and uses Westinghouse CE (Combustion Engineering) 16x16 fuel, which is susceptible to grid-to-rod fretting (GTRF)fuel failures, particularly in the core peripheral locations, based on the past operating experience.

The proposed amendment would allow St. Lucie Unit 2 to transition to AREVA CE 16x16 HTPTM fuel design.AREVA fuel design, with the high thermal performance (HTP) grids, has performed well in the industry in CE designed reactor cores with no GTRF failures.

Although AREVA HTP grid fuel design will be the first full reload application for CE 16x16 plants, the grid design is essentially the same as that used for St. Lucie Unit 1 and other 14x14 plants (Millstone Unit 2, Calvert Cliffs Units 1 and 2). Additionally, the HTP grid design has been successfully used in the industry for other fuel types, such as for Westinghouse 15x15 and 17x17 plants. Other than the grid design, all other key fuel design parameters of the AREVA fuel, such as the rod and the pellet diameters, cladding thickness and the active fuel length, remain the same as the current fuel design.Transitioning to AREVA CE 16x16 HTP fuel requires Technical Specifications (TS)changes. The proposed changes include the use of M5 fuel rod cladding, which needs to be included in the specification of fuel assembly design, removal of linear heat rate (LHR) surveillance requirement when operating on the excore detector monitoring system and the inclusion of AREVA NRC approved analysis methods which will be used in the Updated Final Safety Analysis Report (UFSAR) analysis to support operation with the AREVA fuel. Other proposed TS changes will be administrative changes to specify applicability of certain specifications to Westinghouse fuel only.Since the AREVA fuel will use M5 fuel rod cladding, a 10 CFR 50.46 and 10 CFR 50 Appendix K Exemption Request is included as part of this license amendment request, to allow M5 cladding to be implemented for St. Lucie Unit 2.

2.0 PROPOSED CHANGE

S 2.1 Technical Specifications Changes 2.1.1 TS 4.2.1.3 (Surveillance Requirements for LINEAR HEAT RATE (LHR))Delete the following text in bullets d -f.

L-2014-366 Enclosure 1 Page 3 of 53 d. Verifying that the measured linear heat rate LHRM(z), obtained from a previous incore detector power distribution map, meets the following criteria: W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

LHR and W(z) are specified in COLR Figure 3.2-1 and Table 3.2-3, respectively.

e. Operation is limited to the following:

1. The operation using excore detector monitoring system is limited to less<50 10% above the power level corresponding to the power level at which LHRM(z) is determined in Specification 4.2.1.3d.2. Continuous operation using excore detector monitoring system is limited to 31 days from the time of the power distribution map used in Specification 4.2.1.3d.f. The limit specified in Specification 4.2.1.3d above is not applicable in the following core plane regions: 1. Lower core region from 0 to 15%, inclusive 2. Upper core region from 85 to 100%, inclusive 2.1.2 TS 5.3.1 (Fuel Assemblies)

Add M5 as a fuel rod cladding material.Change "Zircaloy or ZIRL 0" to "Zircaloy, ZIRLOTM or M5 ".2.1.3 TS 6.9.1.11 (CORE OPERATING LIMITS REPORT)Revise TS 6.9.1.1 1.b to include AREVA NRC approved methods for Neutronics, Fuel Mechanical, Thermal-Hydraulics, and Safety Analyses.69. EMF-96-029(P)(A), Volumes I and 2, "Reactor Analysis System for PWRs, Volume I Methodology Description, Volume 2 Benchmarking Results," Siemens Power Corporation, January 1997.

L-2014-366 Enclosure 1 Page 4 of 53 70.XN-NF-78-44 (NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Inc.," October 1983.71.XN-75-27(A) and Supplements 1 through 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Report and Supplement 1 dated April 1977, Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P).72.XN-NF-82-06 (P)(A), Rev. I and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company, Inc., October 1986.73.XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, Inc., November 1986.74.ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, December 1991.75. EMF-92-116(P)(A), Rev. 0, "Generic Mechanical Design Criteria for PWR Fuel Design," Siemens Power Corporation, February, 1999.76.BAW-10240(P)(A), Rev.O, "Incorporation of M5 T M Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., May 2004.77.XN-NF-82-21(P)(A), Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations", Exxon Nuclear Company, September 1983.78. EMF-92-153(P)(A), Revision 1, "HTP. Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," January 2005.79. EMF-1961(P)(A), Revision 0, "Statistical/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation, July 2000.80. EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP, Inc., May 2004.

L-2014-366 Enclosure 1 Page 5 of 53 81.XN-75-32(P)(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," October 1983.82. BAW-10231P-A Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004.83. EMF-2103(P)(A)

Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors, "April 2003.84. EMF-2328 (P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 2001.2.1.4 TS License Condition 3.N (FATES3B Safety Analyses)Add "(Westinghouse Fuel Only)" to this Condition and delete the following text from this Condition: "Upon NRC approval of a new long-term fuel evaluation model and associated methods that explicitly account for thermal conductivity degradation (TCD) that is applicable to St. Lucie Unit 2 design, FPL will, within 6 months: (a) Demonstrate that the St. Lucie Unit 2 safety analyses remain conservatively bounded in licensing basis analyses when compared to the NRC-approved new long-term fuel evaluation model that is applicable to St. Lucie Unit 2 design, or (b) Provide a schedule for re-analysis using the NRC-approved new long-term fuel evaluation model that is applicable to St. Lucie Unit 2 design for any affected licensing basis analyses." The TS marked up pages are provided in Attachment

2. The word processed TS pages are provided in Attachment

5.2.2 Technical

Specifications Bases Changes Changes to the TS Bases are needed to be consistent with the proposed TS changes.The marked up pages are provided in Attachment 3 for information only.2.3 Core Operating Limits Report (COLR) Changes The COLR changes to be implemented are consistent with the proposed TS changes.The marked up pages are provided in Attachment 4 for information only.

L-2014-366 Enclosure 1 Page 6 of 53 Additionally, Figure 3.2-1 is revised to change the peak linear heat rate limit from 12.5 kW/ft to 13 kW/ft based on the large break LOCA re-analysis.

The current limit of 12.5 kW/ft was imposed by the current Westinghouse large break LOCA analysis.

All other current UFSAR accident and transient analyses, and those performed by AREVA for the fuel transition, support a 13 kW/ft value for the peak linear heat rate.3.0 JUSTIFICATION OF PROPOSED CHANGES 3.1 TS 4.2.1.3 (Surveillance Requirements for LINEAR HEAT RATE)Delete bullets d -f.Deletion of LHR surveillance with W(z), when monitoring on excore detector system, is consistent with AREVA methods and similar to St. Lucie Unit 1 requirements.

The W(z) curve is the Westinghouse method to protect the operation of the plant, through protection of the LHR COLR limit, if the incore detector system has become inoperable and operation is only monitored by the excore detectors.

AREVA methods perform this same function through the use of the local power density (LPD) limiting condition of operation (LCO) setpoint verification analyses (Reference 1). The use of the LPD LCO setpoint verification in lieu of W(z) curve is consistent with the CE plants currently serviced by AREVA.3.2 TS 5.3.1 (Fuel Assemblies)

Add M5 as a fuel rod cladding material in TS 5.3.1.The use of nuclear fuel cladding material M5 in PWR reactor fuel is approved by the NRC in Reference 2 topical report BAW-10227P-A, Revision 1,"Evaluation of Advanced Cladding and Structural Material (M5e) in PWR Reactor Fuel," and its use in AREVA approved methods is approved in the Reference 3 topical report BAW-10240(P)(A)

Revision 0, "Incorporation of M5 Properties in Framatome ANP Approved Methods." The M5 cladding is an AREVA Proprietary material composed primarily of zirconium (approximately 99%) and niobium (approximately 1%). This composition has demonstrated superior corrosion resistance and reduced irradiation induced growth relative to both standard and low-tin Zircaloy.

The resulting alloy microstructure is highly stable under irradiation and provides improved in-reactor thermal and mechanical performance of any zirconium alloy.Note that ZIRLO is the same as ZIRLOTM. TM is added to specify it is Westinghouse trademark.

L-2014-366 Enclosure 1 Page 7 of 53 A 10 CFR 50.46 and 10 CFR 50 Appendix K Exemption Request for the implementation of M5 fuel rod cladding is provided in this LAR. This request is necessary since both of these regulatory requirements state or assume that either Zircaloy or ZIRLO is to be used as the fuel rod cladding material and the requirements are not applicable to M5 cladding.3.3 TS 6.9.1.11 (CORE OPERATING LIMITS REPORT)The list of COLR methodologies in TS 6.9.1.11.b is revised to include NRC approved AREVA methods for Neutronics, Fuel Mechanical, Thermal-Hydraulics, and Safety Analyses.Methodologies proposed to be listed in TS 6.9.1.11.b as Items 69, 70 and 71 are used, in combination, as part of the neutronics analyses for every reload cycle and support the COLR limits for the following parameters listed in TS 6.9.1.1 1.a: a. Shutdown Margin, b. Moderator Temperature Coefficient, c. DNB Parameters, d. Total Integrated Radial Peaking factor, e. Linear Heat rate, f. CEA Insertion Limits.Methodologies proposed to be listed in TS 6.9.1.11.b as Items 72, 73, 74, and 75 are used, in combination, in the thermal mechanical and mechanical design analyses which support the COLR limits for the following parameters listed in TS 6.9.1.11.a:

a. DNB Parameters, b. Total Integrated Radial Peaking factor, c. Linear Heat rate, Methodologies proposed to be listed in TS 6.9.1.11.b as Items 77, 78, 80 and 81 are used, in combination, in the thermal hydraulics analysis and in the setpoint verification analyses which support the COLR limits for the following parameters listed in TS 6.9.1.11.a:

a. DNB Parameters, b. Total Integrated Radial Peaking factor, c. Linear Heat rate, d. CEA Insertion Limits.

L-2014-366 Enclosure 1 Page 8 of 53 Methodologies proposed to be listed in TS 6.9.1.11.b as Items 80 and 82 are used, in combination, in the non-LOCA safety analyses which support the COLR limits for the following parameters listed in TS 6.9. .1 11.a: a. Shutdown Margin, b. Moderator Temperature Coefficient, c. DNB Parameters, d. Total Integrated Radial Peaking factor, e. Linear Heat rate, f. CEA Insertion Limits.The new methodology in TS 6.9.1.11.b Item 82, COPERNIC Code, accounts for the effects of the thermal conductivity degradation with increasing rod exposures, and is used to generate the fuel thermal-conductivity, heat capacity and fuel pellet-to-clad gap coefficient inputs for the average core and hot spot models for the S-RELAP5 simulations (new TS 6.9.1.11 .b methodology Item 80).Methodologies proposed to be listed in TS 6.9.1.11.b as Items 83 and 84 are used in the small break and large break LOCA analyses, respectively, which support the COLR limits for the following parameters listed in TS 6.9.1.11 .a: a. Total Integrated Radial Peaking factor, b. Linear Heat rate, Justification for BAW-10240(P)(A), Revision 0 (TS 6.9.1.1 l.b, Item 76)The NRC Safety Evaluation (SE) for BAW-10240(P)(A), Revision 0 (TS 6.9.1.11.b Item 76) found the topical report acceptable for referencing in licensing applications for Westinghouse and Combustion Engineering designed PWRs to the extent specified and under the limitations provided in the Technical Report (TR) and Safety Evaluation (SE).The conditions in the SE for the use of M5 in AREVA's approved methods are the following:

1. The corrosion limit, as predicted by the best-estimate model will remain below 100 microns for all locations of the fuel.Response:

The restriction that corrosion limit, as predicted by the best-estimate model, will remain below 100 microns for all locations of the fuel is implemented in the AREVA fuel design processes.

This limit will be verified for each reload as part of cycle specific reload analysis.

L-2014-366 Enclosure 1 Page 9 of 53 2. All of the conditions listed in the NRC SEs for all AREVA methodologies used for M5 fuel analysis will continue to be met, except that the use of M5 cladding in addition to Zircaloy-4 cladding is now approved.Response: Conditions from approved Safety Evaluations are incorporated as restrictions in AREVA design procedures and guidelines that will control the core reload designs for St. Lucie Unit 2. This will be verified for each reload as part of cycle specific reload analysis.3. All AREVA methodologies will be used only within the range for which M5 data was acceptable and for which the verifications discussed in BAW-10240(P)(Reference

3) or BAW-10227P-A (Reference

2) were performed.

Response: Limitations to ensure AREVA methodologies will be used only within the range for which M5 data was acceptable, and for which the verifications discussed in BAW-10240(P) or Reference 2 were performed, are incorporated as restrictions in AREVA design procedures and guidelines that will control the core reload designs for St. Lucie Unit 2. This will be verified for each reload as part of cycle specific reload analysis.4. The burnup limit for implementation of M5 is 62 GWd/MTU.Response:

The burnup limit identified in approved methodologies is contained in AREVA design processes.

This limit will be verified for each reload as part of cycle specific reload analysis.The rationale for including BAW-10240(P)(A), Revision 0, in the list of Core Operating Limits Report (COLR) methodologies in TS 6.9.1.11.b (Item 76) is as follows: 1. NRC Final Safety Evaluation (SE) for Topical Report BAW-10240(P), approved the incorporation of M5 properties in Framatome ANP approved methods, namely the mechanical analysis methodology (TS 6.9.1.11.b Item 75), the Small Break LOCA (SBLOCA) methodology (TS 6.9.1.11.b Item 84)and the non-LOCA methodologies (TS 6.9.1.11.b Items 79 & 80). BAW-10240(P)(A) provides the basis for using these methodologies for the applicable analyses to support the limits for COLR parameters listed in TS 6.9.1.1 1.a, as stated below, for the fuel with M5 cladding.2. The proposed methodologies in TS 6.9.1.11.b Items 75, 79, 80 & 84, are used to verify or establish limits for the following COLR parameters:

1. Shutdown Margin, 2. Moderator Temperature Coefficient, 3. DNB Parameters, L-2014-366.Enclosure 1 Page 10 of 53 4. Total Integrated Radial Peaking factor, 5. Linear Heat rate, 6. CEA Insertion Limits.3. Based on the GL 88-16 guidance, TS 6.9.1.11.b listed all approved methodologies to support cycle-specific COLR limits listed in TS 6.9.1.11.a.

Since the Topical Report, BAW-10240(P)

Revision 0 will be used in conjunction with the above specified COLR methodologies, to support the cycle-specific COLR limits for the parameters stated above, it is considered appropriate to include this Topical Report in the list of COLR methodologies in TS 6.9.1.11.b.

Assessment of Current Methodologies in TS 6.9.1.11.b The following methodologies will be retained in TS 6.9.1.11.b, since these will continue to be used as part of neutronics design for future cycles with AREVA nuclear fuel.1. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietary).

2. NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point & St. Lucie Nuclear Plants," Florida Power & Light Company, January 1995.Since Westinghouse fuel will remain in the core during transition cycles, the following Westinghouse Topical Reports, which are listed in TS 6.9.1.11.b, are retained since they will be applicable to the Westinghouse fuel analyses during the transient cycles.These methodologies may be deleted from TS 6.9.1.11.b after the completion of fuel transition to full core of AREVA fuel.5. CENPD-275-P, Revision 1-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May 1988, & Revision 1-P Supplement 1-P-A, April 1999.8. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 1: CE Calculated Local Power Density and Thermal Margin/Low Pressure LSSS for St. Lucie Unit 1," December 1979.10. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 3: CE Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for St. Lucie Unit 1," February 1980.

L-2014-366 Enclosure 1 Page 11 of 53 11. CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981.12. Letter, J. W. Miller (NRC) to J. R. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of CEN-123(F)-P (three parts) and CEN-191(B)-P).

14. Letter, J. A. Norris (NRC) to J. H. Goldberg (FPL), Docket No. 50-389, "St.Lucie Unit 2 -Change to Technical Specification Bases Sections '2.1.1 Reactor Core' and '3/4.2.5 DNB Parameters' (TAC No. M87722)," March 14, 1994 (Approval of CEN-371(F)-P).

19. CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983.20. CENPD-1 39-P-A, "C-E Fuel Evaluation Model Topical Report," July 1974.21.CEN-161(B)-P-A, "Improvements to Fuel Evaluation Model," August 1989.22.CEN-161(B)-P, Supplement 1-P-A, "Improvements to Fuel Evaluation Model," January 1992.23.CENPD-132, Supplement 3-P-A, "Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," June 1985.24.CENPD-133, Supplement 5-A, "CEFLASH-4A, A FORTRAN77 Digital Computer Program for Reactor Blowdown Analysis," June 1985.25.CENPD-134, Supplement 2-A, "COMPERC-II, a Program for Emergency Refill-Reflood of the Core," June 1985.26. CENPD-1 35-P, Supplement 5, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977.27. Letter, R. L. Baer (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-1 35, Supplement

1. 5," September 6, 1978.28. CENPD-1 37, Supplement 1-P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977.29. CENPD-133, Supplement 3-P, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," January 1977.

L-2014-366 Enclosure 1 Page 12 of 53 30. Letter, K. Kniel (NRC) to A. E. Scherer (CE), "Evaluation of Topical Reports CENPD-1133, Supplement 3-P and CENPD-137, Supplement 1-P," September 27, 1977.31.CENPD-138, Supplement 2-P, "PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977.32. Letter, C. Aniel (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-138, Supplement 2-P," April 10, 1978.33. Letter, W. H. Bohlke (FPL) to Document Control Desk (NRC), "St. Lucie Unit 2, Docket No. 50-389, Proposed License Amendment, MTC Change from -27 pcm to -30 pcm," L-91-325, December 17, 1991.34. Letter, J. A. Norris (NRC) to J. H. Goldberg (FPL), "St. Lucie Unit 2 -Issuance of Amendment Re: Moderator Temperature Coefficient (TAC No. M82517)," July 15, 1992.35. Letter, J. W. Williams, Jr. (FPL) to D. G. Eisenhut (NRC), "St. Lucie Unit No.2, Docket No. 50-389, Proposed License Amendment, Cycle 2 Reload," L-84-148, June 4, 1984.36. Letter, J. R. Miller (NRC) to J. W. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of Methodology contained in L-84-148).42.CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997.43. CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990.48. CEN-396(L)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/KG for St. Lucie Unit 2," November 1989 (NRC SER dated October 18, 1991, Letter J. A. Norris (NRC) to J. H. Goldberg (FPL), TAC No. 75947).49.CENPD-269-P, Rev. 1-P, "Extended Burnup Operation of Combustion Engineering PWR Fuel," July 1984.50.CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2," December 1984 (NRC SER dated December 21, 1999, Letter K. N.Jabbour (NRC) to T. F. Plunkett (FPL), TAC No. MA4523).51. CENPD-1 37, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998.

L-2014-366 Enclosure 1 Page 13 of 53 55. CENPD-387-P-A, Revision 000, "ABB Critical Heat Flux Correlations for PWR Fuel," May 2000.56.CENPD-132, Supplement 4-P-A, "Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model," March 2001.57. CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998.58.CENPD-404-P-A, Rev. 0, "Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," November 2001.60. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control;FQ Surveillance Technical Specification," February 1994.61.WCAP-11397-P-A, (Proprietary), "Revised Thermal Design Procedure," April 1989.62.WCAP-14565-P-A, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.63.WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code," May 2003.64. Letter, W. Jefferson, Jr. (FPL) to Document Control Desk (USNRC), "St. Lucie Unit 2 Docket No. 50-389: Proposed License Amendment WCAP-9272 Reload Methodology and Implementing 30% Steam Generator Tube Plugging Limit," L-2003-276, December, 2003 (NRC SER dated January 31, 2005, Letter B. T. Moroney (NRC) to J. A. Stall (FPL), TAC No. MC1566).66.WCAP-7908-A, Rev. 0, "FACTRAN-A FORTRAN IV Code for Thermal Transients in a U02 Fuel Rod," December 1989.67.WCAP-7979-P-A, Rev. 0, "TWINKLE -A Multi-Dimensional Neutron Kinetics Computer Code," January 1975.Since the UFSAR events not related to the fuel design change, such as the overpressure events, steam generator tube rupture, pressurizer overfill events, containment peak pressure/temperature events, etc., are not currently re-analyzed with AREVA methodology, the following Westinghouse methodologies will continue to maintain the analyses of record in the UFSAR. As such, these methodologies will be retained in TS 6.9.1.11.b until such time as the corresponding events are re-analyzed with AREVA methodology.

L-2014-366 Enclosure 1 Page 14 of 53 52.CENPD-140-A, "Description of the CONTRANS Digital Computer Code for Containment Pressure and Temperature Transient Analysis," June 1976.59.WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology"'

July 1985.65.WCAP-14882-P-A, Rev. 0, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.68.WCAP-7588, Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.3.4 TS License Condition 3.N (FATES3B Safety Analyses)Since the reactor will have fuel from both the fuel vendors (Westinghouse and AREVA) during the transition cycles, adding "Westinghouse Fuel Only" to this Condition clarifies that the restrictive core design limits specified in this Condition apply only to the Westinghouse fuel.Proposed deletion of the text in this Condition eliminates the need to perform re-analyses of the affected licensing basis analyses for the Thermal Conductivity Degradation (TCD) effects with the NRC approved new long-term fuel evaluation model. This is justified since the Westinghouse fuel will be phased out of the core prior to a realistic schedule for the re-analyses.

However, the restrictive core design limits on the core design will continue to be used to cover any adverse impact of TCD on the Westinghouse fuel analyses.Detailed evaluation of the proposed changes is provided in Attachment 1 along with the LAR's referenced AREVA reports.

L-2014-366 Enclosure I Page 15 of 53 4.0 REGULATORY ANALYSIS In 10 CFR 50.36, the NRC established the regulatory requirements related to the content of the Technical Specifications.

Pursuant to 10 CFR 50.36, the TS are required to include items in the following five specific categories related to station operation:

(1) Safety limits, limiting safety system settings, and limiting control settings;(2) Limiting conditions for operation (LCOs);(3) Surveillance requirements (SRs);(4) Design features; and (5) Administrative controls.From the above TS requirements, Items 3, 4 and 5 are potentially affected by this license amendment request. The proposed change to the surveillance requirement for linear heat rate is consistent with the new fuel vendor methodology for analyses used to confirm the linear heat rate limit when operating on excore detector monitoring system.Operation on excore detector monitoring system is restricted to a maximum of 31 days and verification performed by analysis using approved methodology is considered adequate to ensure the linear heat rate limit is not exceeded during this time period.This surveillance requirement, using W(z), is no longer needed and is thus deleted.Other surveillance requirements related to the LHR continue to remain in the TS. The current design features of the fuel assembly are retained in the TS with an addition to allow M5 cladding to cover the new fuel vendor design. Similarly, the methodologies used in the analyses to support the COLR limits, including the AREVA methodologies applicable to the proposed fuel design change, continue to remain in TS 6.9.1.11.b.

None of the other categories are impacted by the proposed amendment.

The acceptance criteria used for the re-analyses performed to support the proposed changes were reviewed to be consistent with the Standard Review Plan (SRP)requirements of applicable sections of Chapter 4 and Chapter 15. Also, the re-analyses of small break and large break LOCA events continue to meet the 10 CFR 50.46 requirements.

Re-qulatory Precedent In the recent past, Calvert Cliffs transitioned from Westinghouse Turbo 14x14 fuel to AREVA Advanced CE-14 HTP fuel (Reference 4). The fuel design for Calvert Cliffs is similar to St. Lucie Unit 1. There is no direct precedent for AREVA CE 16x16 fuel transition on a reload basis. However, AREVA CE16x16 HTPTM fuel design, similar to the one proposed for St. Lucie Unit 2, operated in Palo Verde Unit 1 and San Onofre Unit 2 (SONGS2) as lead assemblies with no issues and excellent fuel performance.

From analyses viewpoint, Calvert Cliffs fuel transition analyses and St. Lucie Unit 1 extended power uprate (EPU) analyses were approved by the NRC in the last 3 to 4 years (References 4 and 5). Lessons learned from these submittals and review process L-2014-366 Enclosure 1 Page 16 of 53 were all implemented in the current St. Lucie Unit 2 fuel transition analyses which use the same methodology as approved for these previous submittals.

The use of M5cladding is recently approved by the NRC for several utilities with St. Lucie Unit 1 getting approval for the use of M5 in 2014 (Reference 6).Determination of No Significant Hazards Consideration The standards used to arrive at a determination that a request for amendment involves a No Significant Hazards Consideration are included in the Commission's regulation, 10 CFR 50.92, which states that No Significant Hazards Considerations are involved if the operation of the facility in accordance with the amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.Pursuant to the requirements of 10 CFR 50.91(a), an analysis of the issue of No Significant Hazards Consideration is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes for St. Lucie Units 2 revise the Technical Specification (TS)5.3.1 to include MS cladding, delete the linear heat rate surveillance requirement with W(z) in TS 4.2.1.3 and include previously approved AREVA Topical Reports in the list of COLR methodologies in TS 6.9.1.11.

Other change is in TS License Condition 3.N which is related to future analysis of the current fuel and is considered an administrative change, all as a result of changing the fuel supplier.The fuel assembly design is not an initiator to any accident previously evaluated.

Therefore, there is no significant increase in the probability of any accident previously evaluated.

However, the fuel design parameters and the correlations used in the analyses supporting the operation of St. Lucie Unit 2 with the new proposed AREVA fuel are dependent on the fuel assembly design. All the analyses, potentially impacted by the fuel design, have been re-analyzed using the correlations and the methodology applicable to the proposed fuel design and previously approved by the NRC for similar applications.

There are no changes to any limits specified in the Technical Specifications.

M5 cladding to be used in the proposed AREVA fuel design has been previously approved by the NRC for PWR applications, including St. Lucie Unit 1. The core design peaking factors remain unchanged from the current analyses values, except for the large break LOCA which is shown to meet all the 10CFR 50.46 criteria with the increased peak linear heat rate limit.

L-2014-366 Enclosure 1 Page 17 of 53 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

No new or different accidents result from utilizing the proposed AREVA CE 16x16 fuel design. Other than the fuel design change, the proposed license amendment does not involve a physical alteration of the plant or plant systems (i.e., no new or different type of equipment will be installed which would create a new or different kind of accident).

The change to the linear heat rate surveillance requirement, when operating on excore detector monitoring system, and the use of M5 cladding do not affect or create any accident initiator.

There is no change to the methods governing normal plant operation and the changes do not impose any new or different operating requirements.

The core monitoring system remains unchanged.

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

3. Does the proposed change involve a significant reduction in a margin of safety?The changes proposed in this license amendment request are related to the fuel design with M5 cladding and the methodology supporting the analysis of accidents impacted by the fuel design change. The analysis methods used are previously approved by the NRC for similar applications.

The change to the surveillance requirement for the linear heat rate does not change any accident analysis requirements.

The fuel design limits related to the DNBR and fuel centerline melt remain consistent with the limits previously approved for the proposed fuel design change. The overpressure limits for the reactor coolant system integrity and the containment integrity remain unchanged.

All the analyses performed to support the fuel design change meet all applicable acceptance criteria.

The LOCA analyses, with the peak linear heat rate limit increase, continue to meet all the applicable 10 CFR 50.46 acceptance criteria, and thus the proposed changes do not affect margin to safety for any accidents previously evaluated.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based on the previous discussion of the amendment request, it is determined that the proposed amendment does not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any previously evaluated; nor (3) involve a significant L-2014-366 Enclosure 1 Page 18 of 53 reduction in a margin of safety; the amendment does not involve a significant hazards consideration.

5.0 ENVIRONMENTAL

EVALUATION A review has determined that this TS amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed changes 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 changes 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 to be prepared in connection with the proposed amendments.

6.0 CONCLUSION

S This evaluation concludes, on the basis of the considerations discussed above, that (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner with the proposed fuel design change, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. EMF-1961(P)(A), Revision 0, "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," July 2000.2. BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel".3. BAW-10240(P)(A), Rev.0, "Incorporation of M5 T M Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., May 2004.4. Letter, Douglas V. Pickett (USNRC) to George H. Gellrich (CCNPP), "Calvert Cliffs Nuclear Power Plant, Units 1 and 2 -Amendment RE: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (TAC Nos. ME2831 and ME2832)," February 18, 2011.

L-2014-366 Enclosure 1 Page 19 of 53 5. Letter, Tracy Orf (USNRC) to Mano Nazar (FPL), "St. Lucie Plant, Unit 1 -Issuance of Amendment Regarding Extended Power Uprate (TAC No. ME5091)," July 9, 2012.6. Letter, Lisa M. Regner (USNRC) to Mano Nazar (NEE), "St. Lucie Plant, Unit 1 -Issuance of Amendment Regarding the Use of M5 Alloy Fuel Cladding in Core Reload Applications (TAC No. MF1817)," March 21, 2014.

L-2014-366 Enclosure 1 Page 20 of 53 Attachment 1 St. Lucie Unit 2 Fuel Transition Technical Evaluation Attachment 1 -Page 1 of 9 L-2014-366 Enclosure 1 Page 21 of 53 Attachment 1 1.0 Summary St. Lucie Unit 2 is a Combustion Engineering 2x4 plant with a 16x16 fuel lattice array.Except for the fuel rod array, St. Lucie Units 1 and 2 are similar with 217 fuel assemblies in the core, each fuel assembly with 4 corner guide tubes and 1 center instrument tube, 8.18 inch assembly pitch and 136.7 inch active fuel length operating at 3,020 MW(th) core power.The changes related to the St. Lucie Unit 2 fuel design and the proposed technical specifications changes have been evaluated with respect to their impact on the design basis analyses.

This evaluation, which shows acceptable results, includes detailed analyses of: " Fuel mechanical design with thermal mechanical compatibility" Neutronics analyses* Thermal hydraulic analysis" Non-Loss of coolant accident (LOCA) analyses" LOCA analyses Testing of fuel assemblies for pressure drops has been performed to study the mixed core effects which have been adequately addressed in all applicable analyses.

The pressure drop testing showed that AREVA fuel has a higher pressure drop than the co-resident Westinghouse fuel. Also, testing of fuel design components has been performed to benchmark the codes used to perform the mechanical analyses.

Details are provided in the AREVA Technical Report (Reference 8)For neutronics design, St. Lucie Unit 2 will continue to use the Westinghouse methodology for the core design as listed in TS 6.9.1.11.b, Item 1, with AREVA neutronics methodology, as listed in the proposed TS 6.9.1.11.b, Item 69, as backup.The physics input to safety will be generated for each reload cycle using Westinghouse neutronics codes, as a primary method. These physics inputs generated with Westinghouse methodology will support the safety analyses performed with the AREVA safety analyses codes. This approach, which involves a combination of Westinghouse physics methods and AREVA safety analysis methods, is currently used effectively for St. Lucie Unit 1 and is planned to be used for St. Lucie Unit 2. Details are in Section 8.0.For this fuel transition project, only the accident analyses for events affected by the fuel design change have been performed.

For other events which are not affected by the fuel design change, the current analyses of record are retained.

Details for each unanalyzed event are provided in the AREVA Chapter 15 Non-LOCA Summary Report (Reference 9).Attachment 1 -Page 2 of 9 L-2014-366 Enclosure 1 Page 22 of 53 Attachment 1 Analyses, such as the radiological consequences analyses, spent fuel pool criticality analyses, containment peak pressure/temperature analysis and post-LOCA long term cooling analysis remain unaffected by the fuel design change as all pertinent fuel design and operating parameters for the AREVA fuel remain the same as those of the current fuel and the inputs used in these analyses remain unchanged.

These analyses are therefore not redone. Details are in Sections 3.0, 4.0, 5.0 and 6.0.The AREVA Technical Report (Reference

8) includes the description of analyses related to mechanical design, nuclear design, thermal & hydraulic design including setpoint verification, and the UFSAR non-LOCA and LOCA events. The mechanical and thermal hydraulics design evaluations include effects of mixed core on thermal hydraulics and structural compatibility.

All the analyses show acceptable results.The AREVA Chapter 15 Non-LOCA Summary Report (Reference

9) contains all of the non-LOCA events analyzed/evaluated for the fuel transition.

The non-LOCA events not affected by the fuel design change are not analyzed, such as the overpressure events, and the current analyses of record (AOR) will continue to support those events until those events are re-analyzed in the future. The re-analysis will use AREVA methodology following the 10 CFR 50.59 process. However, any evaluations or dispositions needed to the current AOR, such as deviations in the valve opening set-pressures, will be performed on the current AOR consistent with the current AOR methodology.

The AREVA SBLOCA Summary Report provides the details of the SBLOCA break spectrum analysis (Reference 10).The AREVA RLBLOCA Summary Report provides the details of the RLBLOCA analysis (Reference 11).Discussed below are the following topics to facilitate the NRC review of this license amendment request: " Application of S-RELAP5 for overpressure analyses" Radiological consequences analyses* New fuel vault and spent fuel pool criticality analyses* Containment peak pressure/temperature analysis* Post-LOCA long term cooling analysis* Impact of mixed core on co-resident fuel* Use of Westinghouse neutronics methodology with AREVA safety analysis Attachment 1 -Page 3 of 9 L-2014-366 Enclosure 1 Page 23 of 53 Attachment 1 2.0 Application of S-RELAP5 for Overpressure Analyses S-RELAP5 methodology (Reference

3) is used in this license amendment request for the analysis of non-LOCA events analyzed for challenges to specified acceptable fuel design limits (SAFDL) for the proposed new fuel design. Although reactor coolant system (RCS) and secondary overpressure analyses with S-RELAP5 are not performed as part of this submittal, FPL plans to use this code in the future for the analysis of non-LOCA overpressure events. The code is considered acceptable for use in the St. Lucie Unit 2 overpressure analysis since it has been successfully applied to the St. Lucie Unit 1 overpressure events and has been approved by the NRC as part of the extended power uprate (EPU) Amendment

1. 213 in Reference

4. St. Lucie Units 1 and 2 have similar plant systems important to the analysis of overpressure events with similar capacities and same opening setpoints for the pressurizer (2400 psia) and steam safety valves (1000 psia 1st bank of valves & 1040 psia 2 nd bank of valves). Also, the RCS geometry and configuration are similar in terms of the modeling for S-RELAP5 analysis (2 hot legs and 4 cold legs). The reactor trip setpoints are similar, with high pressurizer pressure reactor trip setpoint being slightly lower for Unit 2 (2370 psia) as compared to Unit 1 (2400 psia). Since the two Units are operating at the same power level (3020 MWth) with similar system configuration, the transient response for the two Units will be similar and it is therefore considered acceptable to use S-RELAP5 for the analysis of Unit 2 non-LOCA pressurization events.3.0 Radiological Consequences Analyses The radiological consequences analyses, approved as part of EPU in Amendment No.163 (Reference 5), remain unaffected for the following reasons: The key parameters related to the core design used in the radiological consequences analyses include the reactor power level, Core Operating Limits Report (COLR) limit for radial peaking factor, core average/assembly average burnup, fuel enrichment limits, fuel rod density and event specific fuel failure limits. None of these parameters change for the proposed fuel transition.

Additionally, the plant operating parameters and system configurations including actuation setpoints important to the dose analysis, such as the containment spray system, emergency core cooling systems, control room ventilation and filtration system and control room isolation, remain unchanged.

Other parameters, such as the atmospheric dispersion factors, iodine flashing fraction, minimum sump pH, etc. are not affected by the fuel transition.

There are no changes to the steam releases assumed in the dose analyses due to the fuel design change. Therefore, the radiological consequences analyses in the UFSAR will continue to remain applicable for the operation of St. Lucie Unit 2 with the transition to AREVA CE 16x16 HTPTM fuel as proposed in this license amendment request.Attachment 1 -Page 4 of 9 L-2014-366 Enclosure 1 Page 24 of 53 Attachment 1 4.0 New Fuel Vault and Spent Fuel Pool Criticality Analyses The criticality analyses, approved by the NRC in Amendment No. 162 (Reference 6), remain unaffected for the following reasons: The key fuel design and core design parameters used in the criticality analyses include the power level and operating conditions for depletion in the core, fuel rod dimensions and array configuration, pellet density, maximum enrichment and associated tolerances as used in the analysis.

These parameters will continue to remain applicable and are unchanged for the transition fuel design. Other parameters, such as the soluble boron concentration requirements in the core and in the spent fuel pool, are unchanged from their current values. Spent fuel pool and new fuel vault configurations remain the same.The neutron absorbers credited in the criticality analysis and the fuel storage configurations specified in the Technical Specifications are unchanged.

Therefore, the criticality analysis requirements will continue to remain applicable with the proposed change to the fuel design.5.0 Containment Peak Pressure/Temperature Analysis There are no changes to the inputs used in the containment peak pressure/temperature analysis.

The emergency containment coolers and containment spray systems along with the actuation setpoints remain unchanged for the proposed fuel design change.The mass and energy release calculations are unaffected by the fuel design change as the operating parameters including the power level and the reactor trip functions remain unchanged.

Therefore, the containment analysis is not affected due to the proposed fuel design change.6.0 Post-LOCA Long Term Coolinq Analysis Post LOCA long term cooling analysis addresses requirements related to boron precipitation and return to criticality.

The timing for boron precipitation, as determined in the current analysis, depends mainly on the core power (affecting decay heat), emergency core cooling systems configuration

& flows, and the maximum boron concentrations allowed in the boric acid makeup tank and refueling water storage tank along with the contained water volumes. These analysis input parameters are not affected by the fuel design change. The timing requirement for simultaneous hot/cold leg injection, generated in the current analysis, thus remains unchanged.

Post-LOCA return to criticality requirement is verified every cycle to ensure that the sump boron at the time of flushing the core, as determined from the boron precipitation analysis, remains above the critical boron, as determined from the cycle specific core design. The timing requirement for simultaneous hot/cold leg injection remains Attachment 1 -Page 5 of 9 L-2014-366 Enclosure 1 Page 25 of 53 Attachment 1 unchanged, for the proposed fuel design change. The process of verifying the post-LOCA criticality requirements will continue to be followed on a cycle specific basis as is currently done.7.0 Impact of Mixed Core on Co-resident Fuel Based on the pressure drop testing performed and the thermal hydraulics analysis, as documented in Appendix A, it is determined that the AREVA fuel will have a higher pressure drop than the co-resident Westinghouse fuel in the transition core.Westinghouse fuel will thus have a higher flow than the AREVA fuel resulting in a benefit to the departure from nucleate boiling ratio (DNBR). Westinghouse fuel therefore is expected to remain non-limiting during the mixed core operation.

Due to the differences in the DNB correlations for the two fuel designs, a cycle specific verification, using Westinghouse methodology, will be performed as part of the reload process, based on the actual transition cycle core design, to ensure that the Westinghouse fuel continues to meet the respective fuel design limits during the transition cycles.The evaluation of Westinghouse fuel assemblies under seismic/LOCA loadings associated with mixed core configurations of AREVA and Westinghouse fuel demonstrates that the Westinghouse fuel assemblies continue to satisfy the applicable seismic/LOCA design criteria.8.0 Use of Westinghouse Neutronics Methodology with AREVA Safety Analysis Current and Proposed Approach for Neutronics and Safety Analysis Methodology Current approach used for the reload analysis of St. Lucie Unit 2 reload uses Westinghouse Advanced Nodal Code (ANC) code (Reference

7) for the core design and the generation of physics input to safety, which are used in the safety analysis performed with the Westinghouse safety analysis codes listed in TS 6.9.1.11.b (Reference 1). Core monitoring is currently performed with the ANC based BEACON monitoring system, using data from the fixed incore detector system.The proposed approach with the transition to AREVA fuel will remain unchanged for the core design, generation of physics inputs to safety (including axial shapes) and core monitoring system. However, the safety analysis methods supporting the physics inputs will be changed from the current Westinghouse methodology to the AREVA safety analysis methodology, including the setpoint verification analysis.

This approach, although not typically used for plants fueled with AREVA fuel, has been effectively used for St. Lucie Unit 1, which uses AREVA CE 14x14 HTPTM fuel, for more than 10 cycles.Attachment 1 -Page 6 of 9 L-2014-366 Enclosure 1 Page 26 of 53 Attachment 1 Core Design Guidelines for the Placement of Gadolinia-Bearing Rods St. Lucie Unit 2 core design configurations currently use gadolinia as poison in the fuel rods in certain locations around the corner guide tubes and the center instrument tube, mainly to control peaking factors. There are no gadolinia-bearing rods placed face adjacent or diagonally adjacent to other gadolinia-bearing rods. Additionally, gadolinia-bearing rods are not placed on the assembly periphery.

Although the center instrument tube may be instrumented with incore detectors during the plant operation, the uncertainties generated for the core monitoring tool (BEACON) do not place restrictions for using gadolinia-bearing rods around the center instrument tube for the configurations used.Measurement and Trip Processing Uncertainties Since the Westinghouse Methodology of ANC/BEACON will be the primary software for design and monitoring of the reload cores, measurement uncertainties related to physics parameters will continue to be based on ANC/BEACON methods. The key core parameters, such as the power distribution, peaking factors and the axial shape index (ASI) will be monitored and surveilled using the BEACON analysis of the flux map with the data from fixed incore detector system. The limits for these parameters will be verified, as required, using the BEACON flux map analysis.

The excore ASI, which is the parameter used in the limiting condition for operation (LCO) and limiting safety system setting (LSSS) bands, is calibrated to the BEACON ASI obtained from the incore detector flux map. These uncertainties are independent of the fuel design change and the neutronics methods used in the reload cycle design. Using measurement uncertainties for ASI and peaking factors based on BEACON monitoring system, in combination with AREVA safety/setpoint analysis methodology, is thus justified for St.Lucie Unit 2.There are no changes to the LCO and LSSS axial shape bands, so that the setpoints and the associated measurement uncertainties and processing uncertainties remain unchanged from the current values.The uncertainties related to the analysis input parameters, the trip processing uncertainties and setpoint uncertainties, applicable to the current operation, remain unchanged and are used by AREVA in the thermal hydraulics and safety analyses.

Any additional uncertainties, specifically related to the AREVA methodology, as specified in the approved analysis methodology, are separately applied to comply with the respective Topical Report requirements.

Physics Inputs for Fuel Transition Analysis Representative core designs for the transition work and the generation of physics parameter values input to non-LOCA/LOCA analyses, applicable to St. Lucie Unit 2 with AREVA fuel, were performed with AREVA neutronic methods. These parameters are biased to cover cycle-to-cycle changes and any changes due to the differences Attachment 1 -Page 7 of 9 L-2014-366 Enclosure 1 Page 27 of 53 Attachment 1 between AREVA neutronic methods and ANC. These physics inputs, supported by the safety analysis, will be checked every cycle with the design tool used for the cycle specific core design, which will be primarily ANC. If any cycle specific parameter falls outside the safety analysis input value, the respective analyses will be evaluated/re-analyzed, as necessary, to ensure compliance with the respective acceptance criteria.Although the setpoint verification analyses performed for the transition cores are shown to meet all the limits using axial shapes generated with AREVA neutronic codes, these analyses are performed every cycle to ensure that the cycle specific core design satisfies all the limits. These cycle specific inputs and axial shapes will be generated with the cycle specific core design performed with the ANC code. The axial shapes generated with ANC will be analyzed with the AREVA setpoint methodology (Reference

2) using measurement and processing uncertainties along with other methodology specific uncertainties and allowances.

This process will ensure acceptable setpoints verification.

To ensure applicability of the use of AREVA physics methods in the transition analysis for St. Lucie Unit 2, significant benchmarking was performed for previous several cycles. Selected physics parameters generated with the transition core design, developed with the Westinghouse ANC methodology, were checked against those generated with the AREVA methodology.

The results showed reasonable agreement between the two methodologies.

Conclusion The methodology used for the reload analysis ensures that all the neutronic design parameters generated with Westinghouse methodology are covered by the safety and setpoint analyses which are performed with the AREVA NRC approved methodology applying methodology specific allowances/uncertainties as applicable.

The proposed approach for St. Lucie Unit 2 is used effectively for St. Lucie Unit 1.Attachment 1 -Page 8 of 9 L-2014-366 Enclosure 1 Page 28 of 53 Attachment 1 9.0 References

1. St. Lucie Unit 2 Technical Specifications, Amendment 164.2. EMF-1961(P)(A), Revision 0, "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," July 2000.3. EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP, Inc., May 2004.4. Letter, Tracy Orf (USNRC) to Mano Nazar (FPL), "St. Lucie Plant, Unit 1 -Issuance of Amendment Regarding Extended Power Uprate (TAC No. ME5091)," July 9, 2012.5. Letter, Tracy Orf (USNRC) to Mano Nazar (FPL), "St. Lucie Plant, Unit 2 -Issuance of Amendment Regarding Extended Power Uprate (TAC No. ME5843)," September 24, 2012. (ADAMS ML1223/ML12235A463)

6. Letter, Tracy Orf (USNRC) to Mano Nazar (FPL), "St. Lucie Plant, Unit 2 -Issuance of Amendment Regarding New Fuel vault and Spent Fuel Pool Nuclear Criticality Analysis (TAC No. ME8782)," September 19, 2012.7. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietary).

8. AREVA proprietary report ANP-3352P, Revision 0, St. Lucie Unit 2 Fuel Transition License Amendment Request Technical Report 9. AREVA proprietary report ANP-3347P, Revision 0, St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report 10. AREVA proprietary report ANP-3345P, Revision 0, St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report 11. AREVA proprietary report ANP-3346P, Revision 0, St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report Attachment 1 -Page 9 of 9 L-2014-366 Enclosure 1 Page 29 of 53 Attachment 2 Technical Specification Markups Attachment 2 -Page 1 of 7 L-2014-366 Enclosure 1 Page 30 of 53 Attachment 2-7-NRC dated December 9, 2003, and October 29, 2004, in response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.(c) The first performance of the periodic measurement of CRE pressure, Specification 6.15.d, shall be within 36 months in a staggered test basis, plus the 138 days allowed by SR 4.0.2, as measured from November 13, 2006, which is the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

N. FATES3B Safety Analyses (Westinghouse Fuel Only)-FATES3B has been specifically approved for use for St. Lucie Unit 2 licensing basis analyses based on FPL maintaining the more restrictive operational/design radial power fall-off curve limits as specified in Attachment 4 to FPL Letter L-2012-121, dated March 31, 2012 as compared to the FATES3B analysis radial power fall-off curve limits. The radial power fall-off curve limits shall be verified each cycle as part of the Reload Safety Analysis Checklist (RSAC) rocess.SA I IXt. Lu lU t 9cciQ St r I., 2r safety a1 -emaiff T:E / (b) .. ,.h .. ' -fr re enert-ic cinRg thz "ppro -vd , nDG og term fu onluetiR.

modGel thet ic. pplicebl:

to St. L-1i4 U1Un it2 d-zign $Fo en';4. This renewed license is effective as of the date of issuance, and shall expire at midnight April 6, 2043.FOR THE NUCLEAR REGULATORY COMMISSION Original signed by J. E. Dyer. Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A, Technical Specifications

2. Appendix B, Environmental Protection Plan 3. Appendix C, Antitrust Conditions

4. Appendix D, Antitrust Conditions Date of Issuance:

October 2, 2003 Renewed License No. NPF-16 Amendment No. 164 Revised by letter dated December 17, 2012 Attachment 2 -Page 2 of 7 L-2014-366 Enclosure 1 Page 31 of 53 Attachment 2 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

c. Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of COLR Figure 3.2-2, where 100% of maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:

MxN where: 1. M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.

2. N is the maximum allowable fraction of RATED THERMAL POWER as determined by the FT curve of CO R i r 3 -W(2) ic hc tyole dependent functon that accOuntoS for pwerM dictriutioln tranci: encoIAuntred1II du1-Frig noFrmal cperatRion.

LHIR and- W(:) are sepoifiod in COLR Figuro 3.2 1 and Table -.2 3, -r 4pctivcly.

43- Op8ratian ic lim~itod to thc Alc4g 4-1 The operation, u"ing .....e dctctr mon.itoring cystem is limited to ccc _detcmr.e.nd in Spooifioation 4.2.1.3d..Ccntinu'eus.

cpcration ucbing ncSore detecotr mon.itoring is limited It 31 days frcm tIII time of the power distribution maep used in Spocification f- The limit cpeoified inSpecification 4.2.14 abov e ir, not applicable in tho foIIoAA 4-~ Lezr F re regioA e n fo 0 to 15% r naug v 2? Upper carc region fremR 85 to 100%A, nuco-i, 0,1 ST. LUCIE -UNIT 2 3/4 2-2 Amendment No. 4.7. 4, 592,138 Attachment 2 -Page 3 of 7 L-2014-366 Enclosure 1 Page 32 of 53 Attachment 2 DESIGN FEATURES 5.3 REACTOR CORE REPLACE WITH FUEL ASSEMBLIES ZIRLOTV or M56 5.3.1 The react in 217 fuel assemblies.

Each assembly shall consist of a matrix of Zircalo e4PQ lad fuel rods and/or poison rods, with fuel rods having an initial composition o na ural or slightly enriched uranium dioxide (U0 2) as fuel material.Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 91 full-length control element assemblies and no part-length control element assemblies.

5.4 REACTOR

COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b. For a pressure of 2485 pslg, and c. For a temperature of 6501F, except for the pressurizer which is 700 0 F.ST. LUCIE -UNIT 2 5-3 Amendment No. 4, 138 Attachment 2 -Page 4 of 7 L-2014-366 Enclosure 1 Page 33 of 53 Attachment 2 ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

61. WCAP-1 1397-P-A, (Proprietary), 'Revised Thermal Design Procedure," April 1989.62. WCAP-1 4565-P-A, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.63. WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code," May 2003.64. Letter, W. Jefferson Jr. (FPL) to Document Control Desk (USNRC), "St.Lucie Unit 2 Docket No. 50-389: Proposed License Amendment WCAP-9272 Reload Methodology and Implementing 30% Steam Generator Tube Plugging Limit," L-2003-276, December 2003 (NRC SER dated January 31, 2005, Letter B.T. Moroney (NRC) to J.A. Stall (FPL), TAC No. MC1566).65. WCAP-14882-P-A, Rev. 0, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses", April 1999.66. WCAP-7908-A, Rev. 0, "FACTRAN-A FORTRAN IV Code for Thermal Transients in a U02 Fuel Rod", December 1989.67. WCAP-7979-P-A, Rev. 0, "TWINKLE -A Multi-Dimensional Neutron Kinetics Computer Code", January 1975.68. WCAP-7588, Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics JADD -INSERT A >-I Methods", January 1975.c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle on the NRC.ST. LUCIE -UNIT 2 6-20e Amendment No. -05, 448, 4-3, -447, 163 Attachment 2 -Page 5 of 7 L-2014-366 Enclosure 1 Page 34 of 53 Attachment 2 Insert A 69. EMF-96-029(P)(A), Volumes 1 and 2, "Reactor Analysis System for PWRs, Volume 1 Methodology Description, Volume 2 Bencmarking Results," Siemens Power Corporation, January 1997.70. XN-NF-78-44 (NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Inc.," October 1983.71. XN-75-27(A) and Supplements I through 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Report and Supplement 1 dated April 1977, Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated Febnrary 1987 (P).72. XN-NF-82-06 (P)(A), Rev. 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company, Inc., October 1986.73. XN-NF-85-92(P)(A). "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, Inc., November 1986.74. ANF-88-133(P)(A) and Supplement

1. "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, December 1991.75. EMF-92-116(P)(A), Rev. 0, "Generic Mechanical Design Criteria for PWR Fuel Design," Siemens Power Corporation, February, 1999.76. BAW-10240(P)(A), Rev.0, "Incorporation of M5TM Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., May 2004.77. XN-NF-82-21(P)(A), Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, September 1983.78. EMF-92-153(P)(A), Revision 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," January 2005.79. EMF-1961(P)(A),.

Revision 0, "Statistical/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation, July 2000.Attachment 2 -Page 6 of 7 L-2014-366 Enclosure 1 Page 35 of 53 Attachment 2 80. EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP. hic., May 2004.81. X"N-75-32(P)(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," October 1983.82. BAW-10231P-A Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004.83. EMF-2103(P)(A)

Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003.84. EMF-2328 (P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based." March 2001.Attachment 2 -Page 7 of 7 L-2014-366 Enclosure 1 Page 36 of 53 Attachment 3 Technical Specification Bases Markup (Information Only)Attachment 3 -Page 1 of 7 L-2014-366 Enclosure 1 Page 37 of 53 Attachment 3 BMTD N NO.: FGE TIrrLE: TECHNICAL SPECIFICATIONS

2.0 BASES

ATTACHMENT 1 0F ADM-25.04 3 ofl REVISON NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS u I 6 ST. LUCIE UNIT 2 BASES FOR SECTION ZO.2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting will occur. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the A-BB-NV correlation.

The DNB correlation has been developed to predict th DNB heat flux and the ation of DNB for axially uniform and non-uniform heat f distributions.

The I cal DNB heat flux ratio, DNBR, defined as the ratio of the at flux that would c use DNB at a particular core location to the local heat flt, is indicative of he margin to DNB. Rep R e la e it HTP" b The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to the appropriate correlati

ýnJt 4.i S .L _.GB4.NB-,.£ conjunto ihtcntnc L mbi-'tibn~of')..\

4 -M- P'. .".... " - ..-% -.'- n .-ff ...-c- ionof-ey stemparameter DELETE probabidity distribution functions with the APP44M. DNB correlation uncertainties.

Thisvalue correspondsto a 95%probabiiiw a-jd95%-confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

Attachment 3 -Page 2 of 7 L-2014-366 Enclosure 1 Page 38 of 53 Attachment 3 SBZTIJ N NO.: A V TITLE: TECHNICAL SPECI 2.0 BASES ATTACHMENT 1 OFADM-25.04 of -RdEISON NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS o0 s 6 ST. LUCIE UNIT 2 2.1 SAFETY LIMITS (continued)

BASES (continued)

Replace with' HTP" 2.1.1 REACTOR CORE (continued)

The curves of Figure 2.1- show conservative loci of points of THERMAL POWER, Reactor Cog nt System pressure and vessel inlet temperature with four Reactor Cool;'Pumps operating for which the DNB-SAFDL is not violated based on the ABB-NY CHF correlation for the reference 1.55 Chopped Cosine Axial Shape and Design Limit FTlimit shown in Figure B 2.1-1. The dashed line is not a safety limit; however, operation above this line is not possible because of the actuation of the main steam line safety valves which limit the maimum value of reactor inlet temperature.

Reactor operation at THERMAL POWER levels higher than 107% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in Table 2.2-1. The area of safe transient condition is below and to the left of these lines.The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.The Thermal M argirnLow Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the DNB-SAFDL and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operational Occurrences.

Specific verification of the DNB-SAFDL limit using an appropriate DNB correlation ensures that the reactor core safety limit is satisfied.

?.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The Reactor Coolant System components are designed to Section III, 1971 Edition including Addenda to the Summer, 1973, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 11 0% (2750 psia) of design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System was hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

Attachment 3 -Page 3 of 7 L-2014-366 Enclosure 1 Page 39 of 53 Attachment 3 SBDTIJPG E:N TITLE: TECHNICAL SPECIFICATIONS

2.0 BASES

ATTACHMENT 1 OF ADM-25.04 lJoI REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS -o10 6 ST. LUCIE UNIT 2 2.2 LIMITING SAFETY SYSTEM SETTINGS (continued)

BASES (continued)

2.2.1 REACTOR

TRIP SETPOINTS (continued)

Pressurizer Pressure-High The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam line safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor trip. Thistrip's setpoint is at lessthan or equal to 2375 psia which is belowthe nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.T hermal Marginto w Pressure The Thermal M arginrLow Pressure trip is provided to prevent operation w:w the DNBR is less thanLth13 plXo ~iate correlation limit for DNB-SAFD

--DELETE " -The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIM UTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.

In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.The Thermal MarginfLow Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time, measurement uncertainties and processing error. The allowances include: a variable (power dependent) allowance to compensate for potential power measurement error, an allowance to compensate for potential temperature measurement uncertainty; an allowance to compensate

• for pressure measurement error; and an allowance to compensate for the time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.Attachment 3 -Page 4 of 7 L-2014-366 Enclosure 1 Page 40 of 53 Attachment 3 1CTION NO.:PAGE;-, TILE: TECHNICAL SPECIFICATIONS 3/4.2 BASES ATTACHMENT 4 OF ADM-25.04 R *=VSN NOi.: POWER DISTRIBUTION LIMITS 3 ST. LUCIE UNIT 2 BASES F OR SECTION 314.2 3/4.2 POVWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 22000F.Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring Sy stem performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of COLR Figure 3.2-2. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: (1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, (2)the AZIM UTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and (3) the measured linear heat rate obtained from a previous power distribution map using incore detectors meetsthe criteria of Specification Althoeug h liearF he at rate i s coanti nuoausly maonitore d whe~n u--sinM, t he encore D etecto r MoanitorFing Sy st em, th e formFnal mea su remnen of14ý U z snral ma deA un der ste ady st ate- con--Ad~ions.2 Shoul d- the- I nco-re -De-tector Moiorn Sytm Teom ino9pe rable, t he last Me asurement o-f linaPA-Ar heA-at rate, LHR1 (zvuld remain applicable, but only under steady state cond~ioins

-With the Incore Detector Monitoring System inoperable, and using onlyth (E core Det ectOr M on itoriang SysGtem, v ari atioansG i n p owe r d istri buti ons resulIting from'ths steady state power distributon are, hwoever, conservatively calculated by considering a wide range of unit mnaneuvers in normal 71 op eratioan.

Th e max imumF pe akin g faetorF i ncr ea seG oer ste ady st ate vaus calcaulated as a function of core elcyation, :!, is called \A(z-).-To con far pouer- dotiuto tmndentsen--rm-unt~emd duIrig nOrMa Ioper atiaOn, thsetrasrwcrt lmr its far LHR~z) asrz ctbllishcde utilfi2ing the c 7..le de.pefidRzkt ftrfdisnf LIH=R"(U)i st hz nr .a.urs.. LHRi(z) recsc el lay t te.. aI rcs far itm nu f,--uri- rg toeranns and carim4tric unc.rtsinty.

-4 I.Attachment 3 -Page 5 of 7 L-2014-366 Enclosure 1 Page 41 of 53 Attachment 3 S ECTI N NO.: TrTLE: TECHNICAL SPECIFICATIONS PA 314.2 BASES ATTACHMENT 4 OF ADM-25.04

§4 of R EV S N NO.: POWER DISTRIBUTION LIMITS 3 ST. LUCIE UNIT 2 34.2 POWER DISTRIBUTION LIMITS (continued)

E (continued) isprviedi in theCOL d isate core ele\etions.

LHR-~tz)vauiosfor compaOSison to- the- tranden-Bt limitGs em ot epplicable for the folo'R9in 8Xaxl core reg1ios, me8aired in percent of c rn heigh: aI. L9.AM co. re.gi On, fo o 1% i v;. d ba. Upper corn egin frm8Fo10 nld T he8 top end bottom 1 5% o f the AF c rn emeNludedIG fVam the ovehluati On beceuse of the lovwvpmbeabil ity that these region R uS be61 mor li8M iting in the safet Y analYse8 an~d beceu'w of the of making e pre mea.iremt in these regions.If the- tvo9 Mos-9-t rncent LHPR(Z) eveluetions toAG n incees inýA A GR the quantit4y:

IR-)nerclizedtz199%

RATED-THERMIAL PO iR i is not guere.. teed. .that LHR(z) Wl rmaRi '-Ahin the t ranient lRmt 1dring the Whe g I Anrveilance inteor.'a.

Tohrefore LHR(z) isineasd by thpes pealty fartor DE Espeed iniiAlthe inCOrLR dend com pared to the transie tt LH R p(z) li Ifthc relationstip:

s nt estisll el, ainn ply wh thee auirzrnabs of Sipmds tin 3.2.1 far 3.-1. ThR exceeding its liMmit.RReduceTHERM4AL PtA t ER at least!1% for each 1% LHR(z)exceedstho limife The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of CO LR Figure 3.2-1. The setpoints for these alarms include allowances, set in conservative directions, for (1) a measurement-calculational uncertainty factor, (2) an engineering uncertainty factor, (3) an allowance for axial fuel densification and thermal expansion, and (4) a THERMAL POWER measurement uncertainty factor.Attachment 3 -Page 6 of 7 L-2014-366 Enclosure 1 Page 42 of 53 Attachment 3 CTIOTN NO.: ITIFLE: TECHNICAL SPECIFICATIONS PIGpE'.i--

314.2 BASES ATTACHMENT 4 OF ADM-25.04 SEV SC N NO.: POWER DISTRIBUTION LIMITS L b 01 3 ST. LUCIE UNIT 2 W14.2 POWER DISTRIBUTION LIMITS (continued)

BASES (continued)

/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and safety analyses.

The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1e ,--ntr~pJjtnnli for AFD il. "&nc÷in-,ih-"t41 Io'I:TI"PA&DELETroughout each ato'r 7 at ,-m total flow rate is maintained in the LCO. The remaining DNB DELETE parameter limits are cycle-specific and have been relocated to the COLR.These variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. Operating within these limits will result in meeting the DNBR criterion in the event of a DNB lirrited transient.

The 1 2-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18-month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 1 2-hour basis.G3 Attachment 3 -Page 7 of 7 L-2014-366 Enclosure 1 Page 43 of 53 Attachment 4 Core Operating Limits Report (COLR) Markups (Information Only)Attachment 4 -Page 1 of 5 L-2014-366 Enclosure 1 Page 44 of 53 Attachment 4 Table of Contents Description Page 1.0 Introduction 3 2.0 Core Operating Limits 4 21 Moderator Temperature Coefficient 4 2.2 CEA Position -Misalignment

> 15 inches 4 2.3 Regulating CEA Insertion Limits 4 2.4 Linear Heat Rate 4 2.5 TOTAL INTEGRATED RADIAL PEAKING FACTOR 5 2.6 DNB Parameters 5 2.7 Refueling Operations

-Boron Concentration 5 2.8 SHUTDOWN MARGIN -Tavg Greater Than 200 OF 5 2.9 SHUTDOWN MARGIN -Tvg Less Than or Equal To 200 OF 5 3.0 List of Approved Methods 14 List of Tables and Figures Title Paqe Tb32- DNB Margin Limits 6 SII ii .lDELETEI X-2Q222 W(2) Fa~iGM r. a PU48 FntOnf Ccro Hcight ý , DE-T 7 Fig 3.1-1a Allowable Time To Realign CEA vs. Initial r 8 Fig 3.1-2 CEA Group Insertion Limits vs. THERMAL POWER 9 Fig 3.2-1 Allowable Peak Linear Heat Rate vs. Burnup 10 Fig 3.2-2 AXIAL SHAPE INDEX vs. Maximum Allowable Power Level 11 Fig 3.2-3 Allowable Combinations of THERMAL POWER and FT 12 Fig 3.2-4 AXIAL SHAPE INDEX Operating Limits vs. THERMAL POWER 13 St. Lucie Unit 2 Cycle 21 COLR Rev. 1 Page 2 of 18 Attachment 4 -Page 2 of 5 L-2014-366 Enclosure 1 Page 45 of 53 Attachment 4 2.0 CORE OPERATING LIMITS 2.1 Moderator Temperature Coefficient (TS 3.1.1.4)The moderator temperature coefficient (MTC) shall be less negative than -32 pcmtF at RATED THERMAL POWER.2.2 CEA Position -Misalignment

> 15 Inches (TS 3.1.3.1)The time constraints for full power operation with one full-length CEA misaligned from any other CEA in its group by more than 15 inches are shown in Figure 3.1-la.2.3 Regulating CEA Insertion Limits (TS 3.1.3.6)The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits shown on Figure 3.1-2, with CEA insertion between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits restricted to: a. < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval, b. < 5 Effective Full Power Days per 30 Effective Full Power Days, and c. < 14 Effective Full Power Days per calendar year.2.4 Linear Heat Rate (TS 3.2.1)The linear heat rate shall not exceed the limits shown on Figure 3.2-1.The AXIAL SHAPE INDEX power dependent control limits are shown on Figure 3.2-2.Excore Detector Monitoring System During operation, with the linear heat rate (LHR) being monitored by the Excore DEET Detector Monitoring System, the AXIAL SHAPE INDEX shall be maintained withn h li its of Fi ure 3. -2.If the maFrgin to the linear hcat rate hc derrearced ac determin"ed from the laet two flux maps, then the linear heat rato mucst be inrGeaced by an appropriate penalty. A valuc of 2W6 0 choene as the etandard penalty. Lincar heat rote m~argin deorcacca that arc pedid be gMtFr than 2% pr 314 Effective Full Power Dyc, raguire that penaIk fa-r of grae thn % o ed to inrGeace the meacured linear heat rate LI 1 flRm~) The fineramental penalty facto~re (in e)(eeec of 2% margin decroace) arc included inthe W.(z) functio of Table 3.2 3.St. Lucie Unit 2 Cycle 21 COLR Rev. 1 Page 4 of 18 Attachment 4 -Page 3 of 5 L-2014-366 Enclosure 1 Page 46 of 53 Attachment 4 COLR Table 3.2-3 -W(z) Factors as a Function of Core He .2ht~wfz)'-Height 100 EFPH 2000 EFPH 4000 EFPH 7000 EFPH 10000 EFPH/(t} (145 MWDIMTU) (2903 MWDIMTU (5805 MWDIMTU) (10159 MWDIMTU) (14513 MWDI 1.7 1309 1.274 1.249 1.307 1.3A 1.1 1.298 1.263 1,237 1.297 11 2.67-_ 1 283 1.252 1.226 1.285 t.294 4.24 11739 1.24 1.24 1.73/.27 4.4 1.159 2.39 1.256 1.228 1.202 1.259 1.251 2.55 1.145 1.218 1.193 1.246 1.2.71 236 1.209 1186 1.232 1223 2.265 1116 "12U2_____

1 1/9 1." 1.210 3.02 11, 61.195_ 1.173 1.211 _ 1_201 ..5.7 1.25 1.189 1.169 1.206 1.197 3.34 1.199 1.184 1.166 1.210 1.195 3.50 1108.177 1.162 1.1') , 1.192 3.66 1.190 1.170 1.157 189 1 3.82 1.11 1.165 1.151.183 6.7 .16,183_____

1.187____3.98 1.131 1.160 i'i46 1.177 1.183 4.14 1.156 1.141 1.170 1.178 4.30 1.166 N4.150 1.137 1.162 1.173 4.46 1.159 1.'04 1.132 1.154 1.168 4.62 1,1 .1,12 1.127 1 9145 1.161 4.77 1141.241 1,122 1.136 1.1 4.93 1.135 1.125 1.117/ 1.127 1.147 5.09 1.126 1.118 31.1/ 1.28 1.141 5.25 1.116 1.110 1 ;34 1.110 1.135 5.41 1.108 1.103 1 0997 1.103 1.133 5.57 1.102 1099 1.101 1.140 5.73 1.102 1.102 1 .098 1.103 1.153 5.89 1.106 1.109 .7102 1.110 1.170 6.05 1.113 1.115 ,, F .'_1"06 t.121 1,186 6.21 1.121 1.1205 1.1 1.132 1.202 6.37 1.126 1.128# 1, 1 38\ 1.142 1.216 6.52 1.131 1.145_V 1.156 I 1.151 1.229 6.68 1.142 1.1rY1. 1.173___Ný 1',161 1,241&.84 1.158 11.e7 110 " 1.171 1.252 7.00 1.172 /190 1.206 1.182 1,261" 7.16 1.185 / 1.204 1.222 1.269 7.32 1.197 4 -1.217 1.236 1.201 1.275 7.48 1,209 1,229 1.249 ..0.209 1.279 7.64 1.220 j 1239 1.262 1 16 1.282 7.80 1.230 / 1.249 1.272 1,.2 1.282 7.96 1 .2-39 1.257 12122N%112 1.281 8.12 1.24/ 1.265 1.291 1231.279 8.28 1.3 ,2n 1.273 1.300 1.243 1.276 ..8.43 1 .60 1.279 1.307 1.258 1.271 8.59 ..264 1.288 1.311 1.273 1 t274 8-75 j"1.269 1.297 1.318 1ý285 , 1.294 8.91 1.284 1.306 1.338 1.298. 1.311 9.07 /' 1.300 1ý324 1.366 1ý314 9.23 / 1.315 1.351 1 1.395 1.333 X.339 ,, 9.39/ 1.337 1.378 1.423 1.350 1.X59 9.5136 1.403 1.ý49 1.7 1.43N1 1 861.427 1.74 1.401.86 " 1 1 1 1 1 op and Bottom 15% are excluded.divide these values by the power fraction or, alternately, divide the measured LHR by the pwrfat .St. Lucie Unit 2 Cycle 21 COLR Rev. I Page 7 of 18 Attachment 4 -Page 4 of 5 L-2014-366 Enclosure 1 Page 47 of 53 Attachment 4 14.0 13.5 13.0 I m 12.co a. 12.0 11.5 11.0 UNALCEP ABLE OPERATION----~ --I --7 V ~:~~jz 2~J/r4~L'4 ~ ~ Th~.kA.A ~ LAJ.A ~ ..AJA. ~. f\~AC EPT BLE PERArI"ON BOL Cycle Life EOL (Fuel + Clad + Moderator)

FIGURE 3.2-1 Allowable Peak Linear Heat Rate vs. Burnup St. Lucie Unit 2 Cycle 21 COLR Rev. I Page 10 of 18 Attachment 4 -Page 5 of 5 L-2014-366 Enclosure 1 Page 48 of 53 Attachment 5 Word Processed TS Changes Attachment 5 -Page 1 of 6 L-2014-366 Enclosure 1 Page 49 of 53 Attachment 5-7-NRC dated December 9, 2003, and October 29, 2004, in response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.(c) The first performance of the periodic measurement of CRE pressure, Specification 6.15.d, shall be within 36 months in a staggered test basis, plus the 138 days allowed by SR 4.0.2, as measured from November 13, 2006, which is the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

N. FATES3B Safety Analyses (Westinghouse Fuel Only)FATES3B has been specifically approved for use for St. Lucie Unit 2 licensing basis analyses based on FPL maintaining the more restrictive operational/design radial power fall-off curve limits as specified in Attachment 4 to FPL Letter L-2012-121, dated March 31, 2012 as compared to the FATES3B analysis radial power fall-off curve limits. The radial power fall-off curve limits shall be verified each cycle as part of the Reload Safety Analysis Checklist (RSAC) process.4. This renewed license is effective as of the date of issuance, and shall expire at midnight April 6, 2043.FOR THE NUCLEAR REGULATORY COMMISSION Original signed by J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A, Technical Specifications

2. Appendix B, Environmental Protection Plan 3. Appendix C, Antitrust Conditions

4. Appendix D, Antitrust Conditions Date of Issuance:

October 2, 2003 Renewed License No. NPF-16 Amendment No. 164 Revised by letter dated December 17, 2012 Attachment 5 -Page 2 of 6 L-2014-366 Enclosure 1 Page 50 of 53 Attachment 5 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

c. Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of COLR Figure 3.2-2, where 100% of maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:

MxN where: 1 M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.

2. N is the maximum allowable fraction of RATED THERMAL POWER as determined by the FT curve of COLR Figure 3.2-3.ST. LUCIE -UNIT 2 3/4 2-2 Amendment No. 4-7, 7,0 92, 438 Attachment 5 -Page 3 of 6 L-2014-366 Enclosure 1 Page 51 of 53 Attachment 5 DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 217 fuel assemblies.

Each assembly shall consist of a matrix of Zircaloy, ZIRLOTm or M5 clad fuel rods and/or poison rods, with fuel rods having an initial composition of natural or slightly enriched uranium dioxide (UO 2) as fuel material.Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.CONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 91 full-length control element assemblies and no part-length control element assemblies.

5.4 REACTOR

COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b. For a pressure of 2485 psig, and c. For a temperature of 650 0 F, except for the pressurizer which is 700 0 F.ST. LUCIE -UNIT 2 5-3 Amendment No. 8, 45, 438 Attachment 5 -Page 4 of 6 L-2014-366 Enclosure 1 Page 52 of 53 Attachment 5 ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

61. WCAP-1 1397-P-A, (Proprietary), 'Revised Thermal Design Procedure," April 1989.62. WCAP-14565-P-A, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.63. WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code," May 2003.64. Letter, W. Jefferson Jr. (FPL) to Document Control Desk (USNRC), "St.Lucie Unit 2 Docket No. 50-389: Proposed License Amendment WCAP-9272 Reload Methodology and Implementing 30% Steam Generator Tube Plugging Limit," L-2003-276, December 2003 (NRC SER dated January 31, 2005, Letter B.T. Moroney (NRC) to J.A. Stall (FPL), TAC No. MC 1566).65. WCAP-14882-P-A, Rev. 0, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses", April 1999.66. WCAP-7908-A, Rev. 0, "FACTRAN-A FORTRAN IV Code for Thermal Transients in a U02 Fuel Rod", December 1989.67. WCAP-7979-P-A, Rev. 0, "TWINKLE -A Multi-Dimensional Neutron Kinetics Computer Code", January 1975.68. WCAP-7588, Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Methods", January 1975.69. EMF-96-029(P)(A), Volumes 1 and 2, "Reactor Analysis System for PWRs, Volume 1 Methodology Description, Volume 2 Benchmarking Results," Siemens Power Corporation, January 1997.70. XN-NF-78-44 (NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Inc.," October 1983.71. XN-75-27(A) and Supplements 1 through 5, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Report and Supplement 1 dated April 1977, Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P).72. XN-NF-82-06 (P)(A), Rev. 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Bumup," Exxon Nuclear Company, Inc., October 1986.ST. LUCIE -UNIT 2 6-20e Amendment No. 4405,48, 4 33,38, 447,463 Attachment 5 -Page 5 of 6 L-2014-366 Enclosure 1 Page 53 of 53 Attachment 5 ADMINISTRATIVE CONTROLS (continued_

CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

73. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, Inc., November 1986.74. ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, December 1991.75. EMF-92-116(P)(A), Rev. 0, "Generic Mechanical Design Criteria for PWR Fuel Design," Siemens Power Corporation, February, 1999.76. BAW-10240(P)(A), Rev.0, "Incorporation of M5TM Properties in Framatome ANP Approved Methods," Framatome ANP, Inc., May 2004.77. XN-NF-82-21 (P)(A), Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, September 1983.78 EMF-92-153(P)(A), Revision 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," January 2005.79. EMF-1961(P)(A), Revision 0, "Statistical/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation, July 2000.80. EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP, Inc., May 2004.81. XN-75-32(P)(A), Supplements 1, 2, 3, and 4, "Computational Procedure for Evaluating Fuel Rod Bowing," October 1983.82. BAW-1 0231 P-A Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004.83. EMF-2103(P)(A)

Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003.84. EMF-2328 (P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 2001.c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle on the NRC.ST. LUCIE -UNIT 2 6-20ea Amendment No.Attachment 5 -Page 6 of 6 L-2014-366 Attachment 1 Page 1 of 4 10 CFR 50.46 AND 10 CFR 50 APPENDIX K EXEMPTION REQUEST L-2014-366 Attachment 1 Page 2 of 4 10 CFR 50.46 AND 10 CFR 50 APPENDIX K EXEMPTION REQUEST In accordance with 10 CFR 50.12, Specific Exemptions, Florida Power & Light (FPL), requests exemptions from the requirements specified in 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50 Appendix K, ECCS Evaluation Models, paragraph I.A.5, regarding the use of Zircaloy or ZIRLO as a fuel rod cladding material at St. Lucie Unit 2.10 CFR 50.46(a)(1)(i) states in part: "Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system..." 10 CFR 50 Appendix K, Paragraph I.A.5, states in part: "The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation (Baker, L., Just, L. C., "Studies of Metal Water Reactions at High Temperatures, Ill. Experimental and Theoretical Studies of the Zirconium-Water Reaction," ANL-6548, page 7, May 1962)..." Both of these regulatory requirements, either explicitly or implicitly, state or assume that either Zircaloy or ZIRLO is to be used as the fuel rod cladding material.

This exemption request pertains to the proposed use of the M5 zirconium alloy for fuel rod cladding.10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not result in an undue risk to public health and safety, 3) the exemption is consistent with the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.12(a)(2) are present. The requested exemption to allow the use of a zirconium alloy other than Zircaloy or ZIRLO for fuel cladding material at St. Lucie Unit 2 satisfies these requirements as described below.1. The requested exemption is authorized by law.The fuel that will be irradiated at St. Lucie Unit 2 contains cladding material that does not conform to the cladding material designations explicitly defined in 10 CFR 50.46 and 10 CFR 50, Appendix K. However, the criteria of these sections will continue to be satisfied for the operation of the St. Lucie Unit 2 core containing M5 fuel rod cladding material.

Transition to an alternate, but equivalent fuel product is not precluded by law. As stated above, 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR 50, and granting the proposed exemption would not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations.

Therefore, the exemption is authorized by law.

L-2014-366 Attachment 1 Page 3 of 4 2. The requested exemption does not present an undue risk to the public health and safety.The use of the M5 material as fuel rod cladding follows the NRC approved evaluation process for its implementation by the fuel vendor, as done for the zircaloy or ZIRLO fuel rod cladding, to confirm that operation of this fuel product does not increase the probability of occurrence or the consequences of an accident, and does not create any new or different type of accident that could pose a risk to public health and safety.FPL, in conjunction with AREVA NP Inc. (AREVA), will utilize NRC approved methods for the reload design process, for St. Lucie Unit 2 reload cores containing M5 fuel rod cladding, to ensure safety analysis limits are met for operation within the operating limits specified in the Technical Specifications.

Thus, granting of this exemption request will not present an undue risk to the public health and safety.3. The requested exemption will not endanger the common defense and security.The M5 fuel rod cladding is similar in design to the zircaloy fuel rod cladding material used at St. Lucie Unit 2. The special nuclear material in this fuel product will continue to be handled and controlled in accordance with approved plant procedures.

Therefore, the requested exemption for the proposed use of M5 as fuel rod cladding will not endanger the common defense and security.4. Special circumstances are present which necessitate the request of an exemption to the regulations of 10 CFR 50.46 and 10 CFR 50 Appendix K.The 10 CFR 50.46 and 10 CFR 50 Appendix K regulations do not allow the use of M5 fuel rod cladding material.

The chemical composition of the M5 advanced alloy differs from the specifications for either Zircaloy or ZIRLO. Therefore, in the absence of the requested exemption, use of the M5 advanced alloy falls outside the language and intent of 10 CFR 50.46, and 10 CFR 50 Appendix K, Paragraph I.A.5.The M5 advanced fuel rod cladding is designed to accommodate the high fuel rod burnups that are required for today's modern fuel management schemes and core designs. M5 is an alloy composed primarily of zirconium (approximately 99 percent)and niobium (approximately 1 percent) that has demonstrated superior corrosion resistance and reduced irradiation induced growth relative to both standard and low-tin Zircaloy.

The resulting alloy microstructure is highly stable under irradiation and provides superior in-reactor performance of any zirconium alloy. These improvements permit higher burnup of the fuel in conjunction with improved thermal and mechanical performance.

The M5 alloy has been tested in both reactor and non-reactor environments to determine its mechanical and structural properties as described in Reference

1. The M5 alloy is planned to be used at St. Lucie Unit 2 for fuel rod cladding.

When used as fuel rod cladding, the M5 alloy will provide increased performance margins with regard to fuel rod corrosion and fuel rod growth.10 CFR 50.12 states that the Commission will not grant an exemption from L-2014-366 Attachment 1 Page 4 of 4 requirements contained in 10 CFR 50 unless special circumstances, as defined in 10 CFR 50.12(a)(2) are present. The requested exemption meets the special requirements of 10 CFR 50.12(a)(2)(ii) that "application of the subject regulation is not necessary to achieve the underlying purpose of the rule." For the reasons described below, the use of the M5 advanced alloy as a fuel rod cladding material achieves the underlying purposes of 10 CFR 50.46, and 10 CFR 50 Appendix K, Paragraph I.A.5.The underlying purpose of 10 CFR 50.46 is to ensure that facilities have adequate acceptance criteria for the ECCS. Reference 1 demonstrates that the effectiveness of the ECCS will not be affected by a change from Zircaloy fuel rod cladding to M5 fuel rod cladding.

Analysis described in the reference also demonstrates that the ECCS acceptance criteria applied to reactors fueled with Zircaloy clad fuel are also applicable to reactors fueled with M5 fuel rod cladding.Because the underlying purpose of 10 CFR 50.46 is achieved through the use of the M5 advanced alloy as a fuel rod cladding material, special circumstances are present under 10 CFR 50.12(a) (2) (ii) for granting an exemption to 10 CFR 50.46." The underlying purpose of 10 CFR Appendix K, Paragraph I.A.5, is to ensure that cladding oxidation and hydrogen generation are appropriately limited during a LOCA and conservatively accounted for in the ECCS evaluation model.Specifically, Appendix K requires that the Baker-Just equation be used in the ECCS evaluation model to determine the rate of energy release, cladding oxidation, and hydrogen generation.

Appendix D of Reference 1 demonstrates that the Baker-Just model is conservative in all post-LOCA scenarios with respect to the use of the M5 advanced alloy as a fuel rod cladding material.Because the underlying purpose of 10 CFR 50 Appendix K, Paragraph I.A.5 is achieved through the use of the M5 advanced alloy as a fuel rod cladding material, special circumstances are present under 10 CFR 50.12(a)(2)(ii) for granting exemptions to 10 CFR 50 Appendix K, Paragraph I.A.5.In summary, the intent of 10 CFR 50.46 and 10 CFR 50, Appendix K will continue to be satisfied for the planned operation with M5 fuel rod cladding.

Issuance of an exemption from the criteria of these regulations for the use of M5 fuel rod cladding in the St. Lucie Unit 2 reload cores will not compromise the safe operation of the reactors.

Approval of this exemption request will allow the use of M5 advanced alloy and will improve reactor performance at the St. Lucie Unit 2 Nuclear Plant.

Reference:

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

L-2014-366 Attachment 2 Page 1 of 3 AFFIDAVIT STATE OF WASHINGTON

)) ss.COUNTY OF BENTON )1. My name is Alan B. Meginnis.

I am Manager, Product Licensing, for AREVA Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary.

I am familiar with the policies established by AREVA to ensure the proper application of these criteria.3. I am familiar with the AREVA information contained in the Reports ANP-3345P Revision 0, "St. Lucie Unit 2 Fuel Transition Small Break LOCA Summary Report," dated December 2014, ANP-3346P Revision 0, "St. Lucie Unit 2 Fuel Transition Realistic Large Break LOCA Summary Report," dated December 2014, ANP-3347P Revision 0, "St. Lucie Unit 2 Fuel Transition Chapter 15 Non-LOCA Summary Report," dated December 2014, and ANP-3352P Revision 0, "St. Lucie Unit 2 Fuel Transition License Amendment Request," dated December 2014 and referred to herein as "Documents." Information contained in these Documents has been classified by AREVA as proprietary in accordance with the policies established by AREVA for the control and protection of proprietary and confidential information.

4. These Documents contain information of a proprietary and confidential nature and are of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in these Documents as proprietary and confidential.

L-2014-366 Attachment 2 Page 2 of 3 5. These Documents have been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in these Documents be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA's research and development plans and programs or their results.(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA.(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA.The information in these Documents is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.7. In accordance with AREVA's policies governing the protection and control of information, proprietary information contained in these Documents have been made available, L-2014-366 Attachment 2 Page 3 of 3 on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.SUBSCRIBED before me this ____day of 2014. --'%L -NOTARy ;<1:='L~I.Dg II e,., 44 Susan K. McCoy (l"",,'" 'NOTARY PUBLIC, STATE OF WASHNTON MY COMMISSION EXPIRES: 1/14/2016