ML103560174

License Amendment Request for Extended Power Uprate, Attachment 1; Descriptions and Technical Justifications for the Renewed Operating License, Technical Specifications, and Licensing Basis Changes

ML103560174
Person / Time
Site:	Turkey Point
Issue date:	12/14/2010
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
L-2010-113, LAR 205
Download: ML103560174 (59)

Text

Turkey Point Units 3 and 4 EPU LAR Att. 1-1 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Turkey Point Units 3 and 4 LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE ATTACHMENT 1 Descriptions and Technical Justifications for the Renewed Operating License, Technical Specifications, and Licensing Basis Changes This coversheet plus 58 pages

Turkey Point Units 3 and 4 EPU LAR Att. 1-2 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 1.0 PURPOSE AND SCOPE Florida Power and Light Company (FPL) is proposing to amend Renewed Facility Operating License Nos. DPR-31 and DPR-41 for the Turkey Point Nuclear Plant (PTN), Units 3 and 4, respectively.

The proposed license amendment request (LAR) will revise the Renewed Facility Operating Licenses to permit PTN to operate each with a maximum steady-state reactor core thermal power of 2644 megawatts thermal (MWt). The requested increase constitutes an Extended Power Uprate (EPU) that includes a Measurement Uncertainty Recapture (MUR) Uprate to provide greater unit electrical generating capacity. FPL developed this LAR in accordance with the guidance provided in NRC Review Standard (RS)-001, Review Standard for Extended Power Uprate (Reference 1) and Regulatory Issue Summary (RIS) 2002-3, Guidance on the Content of Measurement Uncertainty Recapture Uprate Applications (Reference 2).

The first phase of the uprate modifications is planned to be implemented during the Fall 2010 outage for Unit 3 and the Spring 2011 outage for Unit 4. The remaining modifications are planned before and during the Spring 2012 outage for Unit 3 and the Fall 2012 outage for Unit 4. Subject to approval of this license application, power ascension to 2644 MWt is planned following startup from the Spring 2012 outage for Unit 3 and from the Fall 2012 outage for Unit 4.

FPL has submitted a license amendment request (PTN LAR 196, Alternative Source Term and Conforming Amendment, dated June 25, 2009 (ML092050277)), (Reference 3) to adopt the alternative source term (AST) as allowed by 10 CFR 50.67. The changes proposed in AST LAR 196 as supplemented or modified by responses to NRC Requests for Additional Information (RAIs) are assumed in the EPU analyses and evaluations contained herein.

FPL has reviewed the PTN Renewed Facility Operating Licenses, Technical Specifications and current licensing bases (CLB) and has determined that no revisions to these documents other than those addressed below or in the AST submittal are required to properly control plant operations and configuration under EPU conditions. Mark-ups of the proposed Renewed Facility Operating Licenses and Technical Specification changes are provided in Attachment 2. The proposed mark-ups of the Technical Specification Bases are also provided as Attachment 3 for information only.

Licensing Reports (LRs) prepared in accordance with the RS-001 review templates are provided in Attachment 4. Additional LRs beyond those included in the template are listed in LR Section 1.0, Introduction. LR Section 1.0 also provides a discussion of the key modifications associated with the power uprate. The functional and operational post-modification testing will be performed to verify satisfactory installation and performance. These tests are listed in LR Section 2.12.1, Approach to EPU Power Level and Test Plan. Appendix A to Attachment 4, Safety Evaluation Report Compliance, supplements information provided in LR Section 2.8.5.0, Accident and Transient Analyses-Non-LOCA Analyses Introduction, with respect to the compliance of the codes and methods used in the EPU analyses and the NRC approving Safety Evaluation Reports corresponding to each method. Appendix B, Additional Codes and Methods, identifies the codes used in the EPU analyses and their application that are not part of the current UFSAR.

A listing of the key regulatory commitments made in this LAR is provided in Attachment 6.

Turkey Point Units 3 and 4 EPU LAR Att. 1-3 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 An assessment of the environmental impacts of the EPU is contained in the Supplemental Environmental Report provided here as Attachment 7.

To support the MUR portion of the uprate, FPL is installing new feedwater flow instrumentation, specifically the Cameron/Caldon Leading Edge Flow Measurement (LEFM) CheckPlus' System. LR Section 2.4.4, Measurement Uncertainty Recapture evaluates the proposed LEFM system using the staffs criteria contained in RIS 2002-3. The Cameron Engineering Reports related to the PTN MUR are provided in Attachment 8 and include the test data from the full scale models for the Units 3 and 4 hydraulic geometry and piping arrangements.

On August 5, 2010, FPL submitted License Amendment Request No. 207, Fuel Storage Criticality Analysis to the NRC (Reference 16) containing proposed TS changes and a new supporting criticality analysis for the purpose of revising the current licensing basis criticality analysis for both new fuel and spent fuel pool storage. The new criticality analysis, documented in WCAP-17094-P, Revision 2, Turkey Point Units 3 & 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis (Reference 15), is resubmitted herein as Attachment 10 and forms the basis for the technical evaluation provided in LR Sections 2.8.6.1, New Fuel Storage and 2.8.6.2, Spent Fuel Storage and supports a requested Technical Specification increase in the maximum allowable fuel enrichment to 5.0 wt% U-235.

Instrument uncertainties for the trip setpoints affected by the EPU are based on the methodology described in WCAP-17070-P, Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4 (Power Uprate to 2644 MWt - Core Power) (Reference 5), which is provided as Attachment 12. The guidance of Technical Specification Task Force (TSTF) No. 493, Rev. 4, Clarify Application of Setpoint Methodology for LSSS Functions, is applied to the Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) trip setpoints and surveillance requirements that are impacted by the EPU. The two Technical Specification notes recommended by TSTF-493 specify operability criteria and require that out-of-tolerance conditions detected during surveillances be evaluated before returning the channel to service. These notes are shown on the markups in Attachment 2 for Technical Specification Table 2.2-1 (RTS Instrumentation Trip Setpoints), Table 4.3-1 (RTS Instrumentation Surveillance Requirements), Table 3.3-3 (ESFAS Instrumentation Trip Setpoints) and Table 4.3-2 (ESFAS Instrumentation Surveillance Requirements).

2.0 BACKGROUND

PTN Units 3 and 4 are each currently licensed at a rated reactor core thermal power of 2300 MWt by Renewed Facility Operating Licenses Nos. DPR-31 and DPR-41, respectively.

Approval of this LAR would authorize FPL to operate each unit at 2644 MWt. This represents a net increase in licensed core thermal power of approximately 15%. It includes a 13% power uprate and a 1.7% measurement uncertainty recapture. The net increase is calculated as follows:

(2300 MWt x 1.13) x 1.017 2644 MWt (2644 MWt - 2300 MWt)/2300 MWt 15%

Due to the magnitude of this increase in licensed core thermal power, this power uprate is defined as an EPU.

Turkey Point Units 3 and 4 EPU LAR Att. 1-4 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 FPL has evaluated the impact of the 15% power uprate for the applicable systems, structures, components, and safety analyses at PTN. The EPU Licensing Reports in Attachment 4 contain the results of this evaluation and the details that support the requested Operating License amendments, Technical Specification changes, Licensing Basis changes and plant modifications.

This Attachment works in concert with the other attachments of the LAR to provide a comprehensive evaluation of the effects of the proposed EPU.

3.0 PROPOSED CHANGE

S 3.1 Operating License and Technical Specification Changes The requested changes involve one revision to each Renewed Facility Operating License and changes to the Technical Specifications and Licensing Basis. The following sections summarize the proposed Technical Specification changes provided as markups in Attachment 2 to this LAR.

Each includes the current state of the Technical Specification, the proposed change, a summary of the basis for the change, and a reference to the Licensing Report section(s) where a more detailed discussion and justification may be found. Where practical, the new text or numbers in the Proposed TS sections are highlighted to enhance readability.

3.1.1 Renewed Operating License Condition d and 3.A. (DPR-31 and DPR-41)

Current OL

d. There is reasonable assurance (i) that the facility can be operated at steady state power levels up to 2300 megawatts thermal in accordance with this renewed operating license without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; 3.A. The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).Proposed OL Proposed OL

d. There is reasonable assurance (i) that the facility can be operated at steady state power levels up to 2644 megawatts thermal in accordance with this renewed operating license without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; 3.A. The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).

Basis for the Change: The results of the analyses and evaluations performed and discussed in the EPU Licensing Reports (Attachment 4) demonstrate that the proposed increase in power can be safely and acceptably achieved by satisfying all applicable acceptance criteria, provided the modifications summarized in LR Section 1.0 and more fully described in the referenced LR sections and the regulatory commitments in Attachment 6 are implemented as stated.

Licensing Report Sections: LR Section 1.0, Introduction and LR Section 1.1, Nuclear Steam Supply System Parameters.

Turkey Point Units 3 and 4 EPU LAR Att. 1-5 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.2 Technical Specification 1.0, Definitions, 1.23 Rated Thermal Power (RTP)

Current TS 1.23 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2300 MWt.

Proposed TS 1.23 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2644 MWt.

Basis for the Change: The results of the analyses and evaluations performed and discussed in the EPU Licensing Reports (Attachment 4) demonstrate that the proposed increase in power can be safely and acceptably achieved by satisfying all applicable acceptance criteria, provided the modifications committed to in the individual Licensing Report sections as well as the regulatory commitments in Attachment 6 are implemented as stated.

Licensing Report Section: LR Section 1.0, Introduction, and LR Section 1.1, Nuclear Steam Supply System Parameters.

3.1.3 Technical Specification, 2.1 Safety Limits, Reactor Core Current TS Figure 2.1-1 Reactor Core Safety Limit - Three Loops in Operation (Tavg (°F) vs Power (fraction of nominal))

Proposed TS Figure 2.1-1 Reactor Core Safety Limit - Three Loops in Operation (Highest Loop Average Temperature (°F) vs Core Power (fraction of 2644 MWt) is as shown in Attachment 2 Basis for the Change: Departure from Nucleate Boiling (DNB) analyses were required to define new core limits, axial offset limits, and Condition II accident acceptability to support operation at EPU conditions.

Licensing Report Section: LR Section 2.8.3, Thermal and Hydraulic Design.

3.1.4 Technical Specification Table 2.2-1, RTS Instrumentation Trip Setpoints Current TS (General)

This Technical Specification table contains changes to Nominal Trip Setpoint (NTS) and Allowable Values for RTS setpoints. The NTS values are the Limiting Safety System Setting (LSSS) values that are derived from the analytical values and adjusted to account for the specific instrument uncertainties.

Turkey Point Units 3 and 4 EPU LAR Att. 1-6 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS (General)

The NTS values for the Nominal Trip Setpoints and Allowable Values are proposed to be changed, as shown in Attachment 2 and described in Sections 3.1.4 through 3.1.12 below, for the following functions:

2.a Power Range Neutron Flux-High Setpoint 5.

Overtemperature TNotes 1 and 2 6.

Overpower TNotes 3 and 4

10. Reactor Coolant Flow-Low 11.

Steam Generator Water Level-Low-Low

12. Steam/Feedwater Flow Mismatch Coincident with Steam Generator (SG) Water Level-Low 15.a Turbine Trip-Emergency Trip Header Pressure Basis for the Change: As defined in 10 CFR 50.36, LSSS are settings for automatic protective devices related to those variables having significant safety functions. 10 CFR 50.36 requires that these limiting settings be included in the Technical Specifications.

The NTS values for RTS trip functions in Table 2.2-1 are calculated based on limits from the safety analyses, process limits for the instrumentation, and the instrument loop uncertainties calculated with 95% probability and 95% confidence to industry standard methodology. The methods used to determine the NTS and Allowable Values and summaries of associated calculations are provided WCAP-17070-P (Reference 5).

The NTS values proposed for TS Table 2.2-1 are values for the nominal trip setpoint and are calculated such that there is 95% probability and 95% confidence that the trip will occur prior to the process variable exceeding the established limit under EPU conditions. Therefore, the assumptions of the safety analyses and their results are protected by the proposed LSSS values.

These LSSS values have been evaluated using methods described in LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems, and WCAP-17070-P (Reference 5).

In accordance with TSTF-493 Rev. 4, Option A, Notes (a) and (b) regarding the as-found and as-left tolerances around the Nominal Trip Setpoints are added to the Channel Calibration and analog channel operational test surveillances associated with the above NTS values in TS Table 4.3-1 (Reactor Trip System Instrumentation Surveillance Requirements). Although TSTF-493 Rev. 4 only requires them to be placed against these surveillance requirements, they are also being added to the affected functions in Table 2.2-1. Since the information for each trip function is spread out over three tables, this approach will enhance the ability of the operator to readily recognize the trip functions affected by the TSTF.

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems.

Turkey Point Units 3 and 4 EPU LAR Att. 1-7 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.5 Technical Specification Table 2.2-1, RTS Instrumentation Trip Setpoints, Notes Current TS (Table Notes)

None.

Proposed TS (Table Notes)

Add Notes (a) & (b) to the Trip Setpoints of the following functions in Table 2.2-1:

2.a Power Range Neutron Flux-High Setpoint 5.

Overtemperature TNote 1 (K1) 6.

Overpower TNote 3 (K4)

10. Reactor Coolant Flow-Low 11.

Steam Generator Water Level-Low-Low

12. Steam/Feedwater Flow Mismatch Coincident with Steam Generator (SG) Water Level-Low 15.a Turbine Trip-Emergency Trip Header Pressure.

Notes (a) and (b) specify operability criteria and require that out-of-tolerance conditions detected during surveillance be evaluated before returning the channel to service.

Note (a) states:

If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

Note (b) states:

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTS) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTS are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field settings) to confirm channel performance. The NTS and methodologies used to determine the as-found and the as-left tolerances are specified in WCAP-17070-P.

Basis for the Change: Surveillance limits are established to verify that reactor trip system instrumentation with an LSSS in TS Table 2.2-1 operates within the boundaries of applicable instrument uncertainty calculations. These limits are implemented in plant procedures in accordance with Notes (a) and (b) above which are consistent with the wording provided in TSTF-493 Rev 4. The methods used to determine NTS values and summaries of the associated calculations are described in WCAP-17070-P, Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 & 4 (Power Uprate to 2644 MWt-Core Power).

The implementation of as-left and as-found limits verifies that the instrument loops are performing in accordance with uncertainty calculation assumptions and ensures that

Turkey Point Units 3 and 4 EPU LAR Att. 1-8 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 out-of-tolerance conditions are evaluated. If a channel cannot be set within the as-left tolerance band, the channel is declared inoperable and Note (b) applies.

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Engineered Safety Features and Control Systems.

3.1.6 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 2a, Power Range Neutron Flux - High Current TS Proposed TS Basis for the Change: The EPU accident and transient analyses determined that for some accidents the Safety Analysis Limit (SAL) for the Power Range Neutron Flux - High reactor trip would need to be reduced from the current 118% to 115% Rated Thermal Power (RTP). Although the SAL is decreasing to 115% RTP, the current nominal trip setpoint of 109% RTP has adequate margin to accommodate the New SAL limit and will not change. There is, however, a change required to the Allowable Value to comply with the methodology as described in WCAP-17070-P.

The trip setpoint is considered a nominal value (i.e., expressed as a value without inequalities) for purposes of Channel Operability Test (COT) and Channel Calibration. Notes (a) and (b) are inserted here as well as in Table 4.3-1 to enhance the ability of the operator to readily recognize the trip functions affected by the TSTF.

Licensing Report Sections: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems and LR Section 2.8.5.0, Non-LOCA Analyses Introduction.

3.1.7 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 5, Overtemperature T, Notes 1 and 2 Current TS NOTE 1: OVERTEMPERATURE T Equation variables are defined as follows:

• K1 = 1.24

• K2 = 0.017/°F

• K3 = 0.001/psig ALLOWABLE VALUE TRIP SETPOINT

2. Power Range, Neutron Flux

a. High Setpoint 112.0% RTP**

109.0% RTP**

ALLOWABLE VALUE TRIP SETPOINT

2. Power Range, Neutron Flux

a. High Setpoint 109.6% RTP **

109.0%(a),(b) of RTP**

Turkey Point Units 3 and 4 EPU LAR Att. 1-9 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251

• T' 577.2°F (Nominal Tavg at RATED THERMAL POWER)

• 3 = 0 s

• 6 = 0 s And f1(I) is a function of the indicated difference between top and bottom detectors of the power range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For qt - qb between -50% and +2%, f1(I) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qb exceeds -50%, the T Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of qt - qb exceeds +2%, that T Trip Setpoint shall be automatically reduced by 2.19% of its value at RATED THERMAL POWER.

NOTE 2: The channels maximum trip setpoint shall not exceed its computed setpoint by more than 0.84% of instrument span.

Proposed TS NOTE 1 OVERTEMPERATURE T Equation variables are defined as follows:

• K1 = 1.31 (a), (b)

• K2 = 0.023/°F

• K3 = 0.00116/psi

• T' 583°F (Indicated Loop Tavg at RATED THERMAL POWER)

• 3 = 2s

• 6 = 2s And f1(I) is a function of the indicated difference between top and bottom detectors of the power range neutron ion chambers with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For qt - qb between -18% and +7%, f1(I) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qb exceeds -18%, the T Trip Setpoint shall be automatically reduced by 3.51% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of qt - qb exceeds +7%, that T Trip Setpoint shall be automatically reduced by 2.37% of its value at RATED THERMAL POWER.

Turkey Point Units 3 and 4 EPU LAR Att. 1-10 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 NOTE 2: The Overtemperature T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel, 0.2% T span for the Pressurizer Pressure channel, and 0.4% T span for the f(I) channel. No separate Allowable Value is provided for Tavg because this function is part of the T value.

Basis for the Change: The accident and transient analyses have determined that the analytical limits utilized for Overtemperature T reactor trip function will change for the EPU program. This function provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature and axial power distribution for various transient analyses. Notes (a) and (b) are inserted on K1 consistent with TSTF-493 Rev 4.

Licensing Report Sections: LR Section 2.4.1, Reactor Protection, Safety Features Actuation and Control Systems, and LR Section 2.8.5.0, Non-LOCA Analyses Introduction.

3.1.8 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 6, Overpower T, Notes 3 and 4 Current TS NOTE 3: OVERPOWER T Equation variables are defined as follows:

• K4 1.10

• K5 0.02/°F for increasing average temperature and 0 for decreasing average temperature

• T" 577.2°F (Nominal Tavg at RATED THERMAL POWER)

• 7 10 sec NOTE 4: The channels maximum trip setpoint shall not exceed its computed trip setpoint by more than 0.96% of instrument span.

Proposed TS NOTE 3: OVERPOWER T Equation variables are defined as follows:

• K4 = 1.10 (a),(b)

• K5 0.0/°F for increasing average temperature and 0/°F for decreasing average temperature

• T" 583°F (Indicated Loop Tavg at RATED THERMAL POWER)

• 7 0 sec NOTE 4: The Overpower T function Allowable Value shall not exceed the nominal trip setpoint by more than 0.5% T span for the T channel. No separate Allowable Value is provided for Tavg because this function is part of the T value.

Basis for the Change: The accident and transient analyses have determined that the constant (K5) utilized for the Overpower T reactor trip function will change for the EPU program. This function prevents power density anywhere in the core from exceeding the design power density.

Turkey Point Units 3 and 4 EPU LAR Att. 1-11 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 This provides assurance of fuel integrity under all possible overpower conditions. Notes (a) and (b) are inserted on K4 consistent with TSTF-493 Rev 4.

Licensing Report Sections: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems, and LR Section 2.8.5.0, Non-LOCA Analyses Introduction.

3.1.9 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 10, Reactor Coolant Flow - Low Current TS Proposed TS Basis for the Change: The EPU accident and transient analyses determined that the existing analytical limit (% flow) for the Reactor Coolant Flow - Low reactor trip does not change and the current nominal trip setpoint of 90% of loop design flow will not change. However, the value of loop design flow has changed for EPU. The Allowable Value is changing to comply with the methodology described in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e.,

expressed as a value without inequalities) for purposes of COT and Channel Calibration. It is intended that the NTS be set at the setpoint in Table 2.2-1 within the as-left tolerance required by Note (b).

Licensing Report Sections: Section 2.4.1.2.3.2.3, Reactor Protection, Safety Features Actuation, and Control Systems and LR Section 2.8.5.0, Non-LOCA Analyses Introduction.

3.1.10 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 11, Steam Generator (SG) Water Level Low-Low Current TS ALLOWABLE VALUE TRIP SETPOINT

10. Reactor Coolant Flow-Low 88.8% of loop design flow*

90% of loop design flow*

• Loop design flow = 85,000 gpm ALLOWABLE VALUE TRIP SETPOINT

10. Reactor Coolant Flow-Low 89.6% of loop design flow*

90%(a),(b) of loop design flow*

• Loop design flow = 86,900 gpm ALLOWABLE VALUE TRIP SETPOINT

11. SG Water Level Low-Low 8.15% of narrow range instrument span 10% of narrow range instrument span

Turkey Point Units 3 and 4 EPU LAR Att. 1-12 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS Basis for the Change: The accident and transient analyses have determined that the analytical limit utilized in the Loss of Normal Feedwater/Loss of AC Power events will remain unchanged for the EPU. The steam generator water level low-low reactor trip (and ESFAS initiation) safety analysis limit for a Loss of Normal Feedwater/Loss of AC Power event is 4.0% Narrow Range Span (NRS). To account for instrument channel uncertainties for EPU conditions and to provide additional margin for operator response time, the Technical Specification low-low SG water level setpoint will change from 10.0% to 16% NRS. The Allowable Value is changing to comply with the methodology described in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e., expressed as a value without inequalities) for purposes of COT and Channel Calibration. It is intended that the NTS be set at the setpoint in Table 2.2-1 within the as-left tolerance required by Note (b).

Licensing Report Sections: LR Section 2.4.1.2.3.2.3, Reactor Protection, Safety Features Actuation, and Control Systems and LR Section 2.13.1, Risk Evaluation for EPU.

3.1.11 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 12, Steam/Feedwater Flow Mismatch Coincident with Steam Generator Water Level - Low Current TS Proposed TS Basis for the Change: The current main steam line flow and main feedwater flow transmitters require changes to support EPU. The transmitters are currently calibrated for a range of 0-4.0 x 106 lbm/hr. The main steam and main feedwater flow transmitters will be recalibrated for a range of 0-5.0 x 106 lbm/hr (0% flow to 129% flow). The expanded range meets or exceeds the current range. The NTS does not require a change for EPU. However, the Allowable Value ALLOWABLE VALUE TRIP SETPOINT

11. SG Water Level Low-Low 15.5% of narrow range instrument span 16%(a),(b) of narrow range instrument span ALLOWABLE VALUE TRIP SETPOINT

12. Steam/Feed Flow Mismatch Coincident with Feed Flow 23.9%

below rated steam flow Feed Flow 20% below rated steam flow Steam Generator Water Level-Low 8.15% of narrow range instrument span 10% of narrow range instrument span ALLOWABLE VALUE TRIP SETPOINT

12. Steam/Feed Flow Mismatch Coincident with Feed Flow 20.7%

below rated steam flow Feed Flow 20%(a),(b) below rated steam flow Steam Generator Water Level-Low 15.5% of narrow range instrument span 16%(a),(b) of narrow range instrument span

Turkey Point Units 3 and 4 EPU LAR Att. 1-13 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 changes from 23.9% to 20.7% below rated steam flow to comply with the methodology described in WCAP-17070-P.

The steam generator water level low signal, coincident with steam flow/feedwater flow mismatch, provides a backup reactor trip that is not specifically credited in the safety analyses. The proposed trip setpoint and Allowable Value are established by calculation. To account for instrument channel uncertainties for EPU conditions, the Technical Specification SG water level low setpoint will change from 10.0% NRS to 16.0% NRS and the Allowable Value will change from 8.15% to 15.5% to comply with the methodology as described in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e., expressed as a value without inequalities) for purposes of COT and Channel Calibration. It is intended that the NTS be set at the setpoint in Table 2.2-1 within the as-left tolerance required by Note (b).

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems.

3.1.12 Technical Specification Table 2.2-1 RTS Instrumentation Trip Setpoints Function 15.a, Turbine Trip - Auto Stop Oil Pressure Current TS Proposed TS Basis for the Change: The 300 psi mechanical hydraulic system which controls the turbine will be replaced with an 1800 psi electrical hydraulic control system as part of EPU. The change to the higher pressure Emergency Trip Header Pressure trip is reflective of the new higher pressure system. As per the current licensing basis, the Turbine Trip is not credited in the safety analyses.

Notes (a) and (b) are applied to the turbine trip consistent with TSTF-493 Rev. 4.

Licensing Report Sections: LR Section 2.5.1.2.2, Turbine Generator.

3.1.13 Technical Specification 3/4.1.1.3, Moderator Temperature Coefficient Current TS LCO 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

d.

Less negative than -3.5 x 10-4 k/k/°F for the all rods withdrawn, end of cycle life (EOL),

RATED THERMAL POWER condition.

ALLOWABLE VALUE TRIP SETPOINT

15. Turbine Trip

a. Auto Stop Oil Pressure 42 psig 45 psig ALLOWABLE VALUE TRIP SETPOINT

15. Turbine Trip

a. Emergency Trip Header Pressure 950 psig 1000(a),(b) psig

Turkey Point Units 3 and 4 EPU LAR Att. 1-14 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 SR 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

b.

The MTC shall be measured at any THERMAL POWER and compared to -3.0 x 10-4 k/k/°F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.0 x 10-4 k/k/°F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification 3.1.1.3d., at least once per 14 EFPD during the remainder of the fuel cycle.

Proposed TS LCO 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

d.

Less negative than -4.1 x 10-4 k/k/°F for the all rods withdrawn, end of cycle life (EOL),

RATED THERMAL POWER condition.

SR 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

b.

The MTC shall be measured at any THERMAL POWER and compared to -3.5 x 10-4 k/k/°F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.5 x 10-4 k/k/°F, the MTC shall be re-measured, and compared to the EOL MTC limit of Specification 3.1.1.3d., at least once per 14 EFPD during the remainder of the fuel cycle.

Basis for the Change: The MTC End of Cycle life LCO and Surveillance Tech Spec limits are changed from -3.5 x 10-4 and -3.0 x 10-4 k/k/°F to -4.1 x 10-4 and -3.5 x 10-4 k/k/°F, respectively, to ensure that sufficient margin exists for this parameter for future cycle bounding analyses. MTC values must remain within the bounds of those used in the accident analysis of the UFSAR Chapter 14 and MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

Licensing Report Sections: LR Section 2.8.2, Nuclear Design.

3.1.14 Technical Specification 3/4.1.2.1, Boration Systems, Flow Path-Shutdown Current TS SR 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of the rooms containing flow path components is greater than or equal to 55°F when a flow path from the boric acid tanks is used, and Proposed TS SR 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of the rooms containing flow path components is greater than or equal to 62°F when a flow path from the boric acid tanks is used, and

Turkey Point Units 3 and 4 EPU LAR Att. 1-15 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Basis for the Change: For EPU, the minimum temperature of the rooms containing flow path components when a flow path from the boric acid tanks is used is increased to ensure the solubility of the boron solution associated with the increased boron concentration. The minimum temperature is based on the solubility of the increased maximum concentration of boric acid in the boric acid storage tanks of 4.0%, according to the solubility table in WCAP-1570, plus 5°F for instrument uncertainty. Maintaining the room temperature above the solubility temperature will ensure that the boric acid storage system will remain operable at all EPU concentrations.

Licensing Report Sections: LR Table 2.8.5.6.3.4-4, Boric Acid Solution Solubility Limit.

3.1.15 Technical Specification 3/4.1.2.2, Flow Paths-Operating Current TS SR 4.1.2.2 The above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of the rooms containing flow path components is greater than or equal to 55°F when a flow path from the boric acid tanks is used; Proposed TS SR 4.1.2.2 The above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of the rooms containing flow path components is greater than or equal 62°F when a flow path from the boric acid tanks is used; Basis for the Change: For EPU, the minimum solution temperature when a flow path from the boric acid tanks is used is increased to ensure the solubility of the boron solution associated with the increased boron concentration. The minimum temperature is based on the solubility of the increased maximum concentration of boric acid in the boric acid storage tanks of 4.0%,

according to the solubility table in WCAP-1570, plus 5°F for instrument uncertainty. Maintaining the room temperature above the solubility temperature will ensure that the boric acid storage system will remain operable at all EPU concentrations.

Licensing Report Sections: LR Table 2.8.5.6.3.4-4, Boric Acid Solution Solubility Limit.

3.1.16 Technical Specification 3/4.1.2.4, Borated Water Source-Shutdown Current TS LCO 3.1.2.4 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

2) A boron concentration between 3.0 wt% (5245 ppm) and 3.5 wt% (6119 ppm), and

3) A minimum boric acid tanks room temperature of 55°F.

b.

The refueling water storage tank (RWST) with:

2) A minimum boron concentration of 1950 ppm, and SR 4.1.2.4 The above required borated water source shall be demonstrated OPERABLE:

Turkey Point Units 3 and 4 EPU LAR Att. 1-16 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 a.

At least once per 7 days by:

3) Verifying that the temperature of the boric acid tanks room is greater than or equal to 55°F, when it is the source of borated water.

Proposed TS LCO 3.1.2.4 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

2) A boron concentration between 3.0 wt% (5245 ppm) and 4.0 wt% (6993 ppm), and

3) A minimum boric acid tanks room temperature of 62°F.

b.

The refueling water storage tank (RWST) with:

2) A boron concentration between 2400 ppm and 2600 ppm, and SR 4.1.2.4 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

3) Verifying that the temperature of the boric acid tanks room is greater than or equal to 62°F, when it is the source of borated water Basis for the Change: The EPU boron delivery capability of the Chemical and Volume Control System during cooldown sets new boron concentration limits that are achieved in the RCS with the boric acid tanks and refueling water storage tanks at conservative temperatures and levels.

The resulting minimum achievable Reactor Coolant System boron concentration limits are used in the Reload Safety Analysis Checklist to ensure adequate reactivity shutdown margin is available for any post shutdown time. The maximum boron concentration in the Boric Acid Storage System has been raised from 3.5 wt% to 4.0 wt%, and the minimum concentration in the RWST has been raised from 1950 ppm to 2400 ppm. These increases create acceptable margin to the core design limit and ensure shutdown margin under EPU conditions. The limit on the maximum boron concentration in the RWST of 2600 ppm will preclude boron precipitation in the core.

For EPU, the minimum temperature of the boric acid tanks room is increased to ensure the solubility of the boron solution associated with the increased boron concentration. The minimum temperature is based on the solubility of the increased maximum concentration of boric acid in the boric acid storage tanks of 4.0%, according to the solubility table in WCAP-1570, plus 5°F for instrument uncertainty. Maintaining the room temperature above the solubility temperature will ensure that the boric acid storage system will remain operable at all EPU concentrations.

Licensing Report Sections: LR Section 2.8.5.6.3, Emergency Core Cooling System and Loss-of-Coolant Accidents.

3.1.17 Technical Specification 3.1.2.5, Borated Water Sources - Operating Current TS LCO 3.1.2.5 The following borated water sources shall be OPERABLE:

Turkey Point Units 3 and 4 EPU LAR Att. 1-17 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 a.

A Boric Acid Storage System with:

3) A minimum boric acid tanks room temperature of 55°F.

b.

The refueling water storage tank (RWST) with:

2) A minimum boron concentration of 1950 ppm, Action:

c.

With the boric acid tank inventory concentration greater than 3.5 wt%, verify that the boric acid solution temperature for boration sources and flow paths is greater than the solubility limit for the concentration.

Figure 3.1-2 Boric Acid Tank Minimum Volume (1) - Modes 1, 2, 3, and 4 (Minimum BAT Volume vs BAT Inventory Concentration)

SR 4.1.2.5 Each borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

3) Verifying that the temperature of the boric acid tanks room is greater than or equal to 55°F, when it is the source of borated water.

Proposed TS LCO 3.1.2.5 The following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

3) A minimum boric acid tanks room temperature of 62°F.

b.

The refueling water storage tank (RWST) with:

2) A boron concentration between 2400 ppm and 2600 ppm.

Action:

c.

With the boric acid tank inventory concentration greater than 4.0 wt% verify that the boric acid solution temperature for boration sources and flow paths is greater than the solubility limit for the concentration.

Figure 3.1-2 Boric Acid Tank Minimum Volume (1) - Modes 1, 2, 3, and 4 (Minimum BAT Volume vs BAT Inventory Concentration) is revised as shown in Attachment 2 SR 4.1.2.5 Each borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

3) Verifying that the temperature of the boric acid tanks room is greater than or equal to 62°F when it is the source of borated water.

Basis for the Change: The EPU boron delivery capability of the Chemical and Volume Control System during cooldown sets new boron concentration limits that are achieved in the RCS with the boric acid tanks and refueling water storage tanks at conservative temperatures and levels.

The resulting minimum achievable Reactor Coolant System boron concentration limits are used

Turkey Point Units 3 and 4 EPU LAR Att. 1-18 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 in the Reload Safety Analysis Checklist to ensure adequate reactivity shutdown margin is available for any post shutdown time. The minimum boron concentration in the Boric Acid Storage System, provided in a curve of boric acid tank volume versus concentration, has been increased and the minimum concentration in the RWST has been raised from 1950 ppm to 2400 ppm, with a new maximum concentration specified for 2600 ppm to preclude boron precipitation in the core. These increases create acceptable margin to the core design limit.

For EPU, the minimum temperature of the boric acid tanks room is increased to ensure the solubility of the boron solution associated with the increased boron concentration. The minimum temperature is based on the solubility of the increased maximum concentration of boric acid in the boric acid storage tanks of 4.0%, according to the solubility table in WCAP-1570, plus 5°F for instrument uncertainty. Maintaining the room temperature above the solubility temperature will ensure that the boric acid storage system will remain operable at all EPU concentrations.

ACTION statement c. is being revised to reflect the increase in the maximum boron concentration of the boric acid storage tanks from 3.5 wt% to 4.0 wt%.

Licensing Report Section: LR Section 2.8.5.6.3, Emergency Core Cooling System and Loss-of-Coolant Accidents.

3.1.18 Technical Specification 3.2.5, DNB Parameters Current TS LCO 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Reactor Coolant System Tavg 581.2°F b.

Pressurizer Pressure 2200 psig*, and c.

Reactor Coolant System Flow 264,000 gpm Proposed TS LCO 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Reactor Coolant System Tavg 585°F b.

Pressurizer Pressure 2204 psig* and c.

Reactor Coolant System Flow 270,000 gpm Basis for the Change: The EPU revised limits for 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 accident analyses. The limits are consistent with the UFSAR assumptions and have been analytically demonstrated adequate to maintain a minimum Departure from Nucleate Boiling Ratio (DNBR) above the applicable design limits throughout each analyzed transient.

Licensing Report Sections: LR Section 1.1, Nuclear Steam Supply System Parameters, and LR Section 2.8.3, Thermal and Hydraulic Design.

Turkey Point Units 3 and 4 EPU LAR Att. 1-19 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.19 Technical Specification Table 3.3-1, Reactor Trip System Instrumentation Current TS

15. Turbine Trip a.

Autostop Oil pressure Proposed TS

15. Turbine Trip a.

Emergency Trip Header Pressure Basis for Change: The 300 psi mechanical hydraulic system which controls the turbine will be replaced with an 1800 psi electrical hydraulic control (EHC) system as part of EPU. The Emergency Trip Header Pressure associated with the new EHC system will serve the same function as an anticipatory trip that is not credited in the safety analyses.

Licensing Report Section: LR Section 2.5.1.2.2, Turbine Generator.

3.1.20 Technical Specification Table 4.3-1, RTS Instrumentation Surveillance Requirements, Notes Current TS (Table Notes)

None.

Proposed TS (Table Notes)

Add Notes (a) and (b) in Table 4.3-1 to the applicable Channel Calibration and analog channel operational test surveillance requirements of the RTS functions that are proposed to be changed in Attachment 2 to specify as-found and as-left criteria in accordance with TSTF-493, Rev 4. The Table 4.3-1 functions to which these notes are applied are:

2.a Power Range, Neutron Flux - High Setpoint 5.

Overtemperature T Note 1 (K1) 6.

Overpower T Note 3 (K4)

10. Reactor Coolant Flow-Low 11.

Steam Generator Water Level-Low-Low

12. Steam Generator Water Level-Low Coincident with Steam/Feedwater Flow Mismatch 15.a Turbine Trip - Emergency Trip Header Pressure Note (a) states:

If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

Turkey Point Units 3 and 4 EPU LAR Att. 1-20 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Note (b) states:

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTS) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTS are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field settings) to confirm channel performance. The NTS and methodologies used to determine the as-found and the as-left tolerances are specified in WCAP-17070-P.

Basis for the Change: As defined in 10 CFR 50.36, LSSS are settings for automatic protective devices related to those variables having significant safety functions. 10 CFR 50.36 requires that these limiting settings be included in the Technical Specifications and have appropriate surveillance tests performed. Surveillance requirements are established to verify that reactor trip system instrumentation with an LSSS in TS Table 2.2-1 operates within the boundaries of applicable instrument uncertainty calculations. These instrument tolerances are implemented in plant procedures in accordance with Notes (a) and (b) above which are consistent with the wording in TSTF 493 Rev. 4. The methods used to determine NTS values and summaries of calculations are described in WCAP-17070-P. The implementation of as-left and as-found tolerances verifies that the instrument loops are performing in accordance with uncertainty calculation assumptions and that out-of tolerance conditions are evaluated. If a channel cannot be set within the as-left tolerance band, the channel is declared inoperable and Note (b) applies.

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Engineered Safety Features and Control Systems.

Turkey Point Units 3 and 4 EPU LAR Att. 1-21 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.21 Technical Specification Table 3.3-2 ESFAS Instrumentation Function 1, Safety Injection Current TS Proposed TS Basis for the Change: FPL submitted a license amendment request to the NRC via FPL Letter (L-2009-133), License Amendment Request (LAR) 196 - Alternative Source Term and Conforming Amendment, (Reference 3) to adopt the alternative source term (AST) as allowed by 10 CFR 50.67. As part of the Amendment, credit for the Emergency Containment Filtering System is removed from the licensing basis and SI actuation signals no longer exist for these fans. The remaining list of SI actuated equipment/systems is being removed since they are already identified in the Technical Specification basis and each are controlled under separate specifications. This is also consistent with Standard Technical Specifications (NUREG 1431 Rev 3).

Licensing Report Section: None FUNCTIONAL UNIT TOTAL NO. OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION

1. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Ventilation Isolation, Start Diesel Generators, Containment Phase A Isolation (except Manual SI), Containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water)

FUNCTIONAL UNIT TOTAL NO. OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION

1. Safety Injection

Turkey Point Units 3 and 4 EPU LAR Att. 1-22 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.22 Technical Specification Table 3.3-2 ESFAS Instrumentation Function 5, Feedwater Isolation Current TS Proposed TS Basis for the Change: Proposed Technical Specification 3/4.7.1.7 adds LCO requirements for the new FIVs consistent with NUREG-1431, Standard Technical Specifications Westinghouse Plants FUNCTIONAL UNIT TOTAL NO. OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION

5. Feedwater Isolation a.

Automatic Actuation Logic and Actuation Relays 2

1 2

1, 2 22 b.

Safety-Injection See Item 1. above for all Safety Injection initiating functions and requirements c.

Steam Generator Water Level - High-High 3/steam generator 2/steam generator in any operating steam generator 2/steam generator in any operating steam generator 1, 2, 15 FUNCTIONAL UNIT TOTAL NO. OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION

5. Feedwater Isolation a.

Automatic Actuation Logic and Actuation Relays 2

1 2

1, 2, 3 22 b.

Safety-Injection See Item 1. above for all Safety Injection initiating functions and requirements c.

Steam Generator Water Level - High-High 3/steam generator 2/steam generator in any operating steam generator 2/steam generator in any operating steam generator 1, 2, 3 15

Turkey Point Units 3 and 4 EPU LAR Att. 1-23 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 with applicability to Modes 1, 2 and 3. The current LCO for the ESFAS Feedwater isolation function applies to modes 1 and 2 only. In order to be consistent with new LCO 3/4.7.1.7, the ESFAS modes of applicability are being changed in this table.

Licensing Report Sections: LR Section 2.5.5.4, Condensate and Feedwater and LR Section 2.6.3.2, Mass and Energy Release Analysis for Secondary System Pipe Ruptures.

3.1.23 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Current TS (General)

This Technical Specification table contains changes to Nominal Trip Setpoint (NTS) and Allowable Values for ESFAS setpoints. The NTS values are the LSSS values that are derived from the analytical values and adjusted to account for the specific instrument uncertainties. The Allowable Values and Nominal Trip Setpoints for selected functions in Technical Specification Table 3.3-3 (ESFAS Instrumentation Trip Setpoints) are being revised to reflect changes resulting from EPU and implementation of the methodology of WCAP-17070-P (Reference 5).

Proposed TS (General)

The NTS values for ESFAS Trip Setpoints and Allowable Values are proposed to be changed, as shown in Attachment 2 and described in Sections 3.1.26 through 3.1.33 below, for the following functions:

1.f Safety Injection (SI)-Steam Line Flow-High Coincident with Steam Generator Pressure-Low 4.d Steam Line Isolation-Steam Line Flow-High Coincident with Steam Line Pressure -Low or Tavg-Low 5.c Feedwater Isolation-Steam Generator Water level High-High 6.b Auxiliary Feedwater(3)-Steam Generator Water Level-Low-Low 7.b Loss of Power-480V Load Centers Undervoltage 7.c Loss of Power-480V Load Centers Degraded Voltage Basis for the Change: As defined in 10 CFR 50.36, LSSS are settings for automatic protective devices related to those variables having significant safety functions. 10 CFR 50.36 requires that these limiting settings be included in the Technical Specifications.

The NTS values proposed for TS Table 3.3-3 are values for the nominal trip setpoint and are calculated such that there is 95% probability and 95% confidence that the interlock, permissive or block function will occur prior to the process variable exceeding the established limit and ensures the interlock, permissive or block function will occur in accordance with the assumptions of the analyses. Therefore, the assumptions of the safety analyses and results are protected by proposed LSSS values. The methods used to determine NTS values and summaries of calculations are provided in LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems, and WCAP-17070-P (Reference 5). These LSSS values have been evaluated under EPU conditions using methods described in LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems, and WCAP-17070-P (Reference 5)

Turkey Point Units 3 and 4 EPU LAR Att. 1-24 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Licensing Report Section: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems.

3.1.24 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints, Notes Current TS (Table Notes)

None.

Proposed TS (Table Notes)

Add Notes (a) & (b) to Functions 1.f, 4.d, 5.c, and 6.b (see 3.1.23) in Table 3.3-3 to specify operability criteria and to require that out-of-tolerance conditions detected during the surveillance be evaluated before returning the channel to service. These notes are consistent with TSTF-493, Rev 4.

Note (a) states:

If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

Note (b) states:

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTS) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTS are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field settings) to confirm channel performance. The NTS and methodologies used to determine the as-found and the as-left tolerances are specified in WCAP-17070-P.

Basis for the Change: Surveillance limits are established to verify that engineered safety features actuation system instrumentation with an LSSS in TS Table 3.3-3 operates within the boundaries of applicable instrument uncertainty calculations. In accordance with TSTF-493 Rev. 4, Option A, Notes (a) and (b) regarding the as-found and as-left tolerances around the Nominal Trip Setpoints are added to the Channel Calibration and analog channel operational test surveillances associated with the NTS values for Functions 1.f, 4.d, 5.c and 6.b in TS Table 4.3-2 (Engineered Safety Features Actuation System Instrumentation Surveillance Requirements).

Although TSTF-493 Rev. 4 only requires the notes to be placed against the surveillance requirements, they are also being added to the affected functions in Table 3.3-3. Since the information for each trip function is spread out over three tables, this approach will enhance the ability of the operator to readily recognize the trip functions affected by the TSTF. The determination of as-left setting tolerance and as-found criteria is described in WCAP-17070-P.

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems.

Turkey Point Units 3 and 4 EPU LAR Att. 1-25 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.25 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 1. Safety Injection Current TS Proposed TS Basis for the Change: FPL submitted a license amendment request to the NRC via FPL Letter (L-2009-133), License Amendment Request (LAR) 196 - Alternative Source Term and Conforming Amendment, (Reference 3) to adopt the alternative source term (AST) as allowed by 10 CFR 50.67. As part of the Amendment, credit for the Emergency Containment Filtering System is removed from the licensing basis and SI actuation signals no longer exist for these fans. The remaining list of SI actuated equipment/systems is being removed since they are already identified in the Technical Specification basis and each are controlled under separate specifications. Removal of these items is also consistent with Standard Technical Specifications (NUREG 1431 Rev 3).

Licensing Report Section: None.

ALLOWABLE VALUE TRIP SETPOINT

1. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Ventilation Isolation, Start Diesel Generators, Containment Phase A Isolation (except Manual SI), Containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water)

ALLOWABLE VALUE TRIP SETPOINT

1. Safety Injection

Turkey Point Units 3 and 4 EPU LAR Att. 1-26 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.26 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 1.f, Safety Injection on Steam Line Flow - High Coincident with SG Pressure - Low or Tavg - Low Current TS Proposed TS Basis for the Change: The existing NTS of less than or equal to a P corresponding to 40%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114% steam flow at full load is acceptable for the EPU and is not changing. However, the Allowable Value ALLOWABLE VALUE TRIP SETPOINT

1. Safety Injection

f. Steam Line Flow --

High A function defined as follows: A P corresponding to 44%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 116.5% steam flow at full load A function defined as follows: A P corresponding to 40%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114%

steam flow at full load Coincident with:

Steam Generator Pressure - Low 588 psig 614 psig ALLOWABLE VALUE TRIP SETPOINT

1. Safety Injection

f. Steam Line Flow --

High A function defined as follows: A P corresponding to 41.2%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114.4% steam flow at full load A function defined as follows: A P corresponding to 40%(a),(b) steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114%(a),(b) steam flow at full load Coincident with:

Steam Generator Pressure - Low (4) 607 psig 614(a),(b) psig (4) Time constants utilized in the lead-lag controller for Steam Generator Pressure-Low Steam Line Pressure-Low are t1 50 seconds and t2 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

Turkey Point Units 3 and 4 EPU LAR Att. 1-27 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 must be changed to a P corresponding to 41.2% steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114.4% steam flow at full load to comply with the methodology of WCAP-17070-P.

The accident analyses determined that the analytical limit for the steam line pressure low SI inside containment steam break will need to be revised from the existing value of 432.3 psig to 566.3 psig for the EPU. The current Safety Analysis Limit (SAL) of 432.3 psig remains applicable for the outside containment steam break analysis. Although the SAL is increasing to the 566.3 psig, the current NTS of 614 psig has adequate margin to accommodate the new SAL limit and will not change. There is however, a change required to the Allowable Value to comply with the methodology in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e.,

expressed as a value without inequalities) for purposed of COT and Channel Calibration.

The addition of the lead/lag addressed in the new Note (4) on the steamline pressure input causes the safety injection signal to occur significantly faster, reducing the total mass that enters containment.

Licensing Report Sections: LR Section 2.4.1.2.3.2.4, Reactor Protection, Safety Features Actuation, and Control Systems; and LR Section 2.6.3.2, Mass and Energy Release Analysis for Secondary System Pipe Ruptures.

3.1.27 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 4.d., Steam Line Isolation on Steam Line Flow - High Coincident with Steam Line Pressure - Low or Tavg - Low Current TS ALLOWABLE VALUE TRIP SETPOINT

4. Steam Line Isolation

d. Steam Line Flow --

High A function defined as follows: A P corresponding to 44%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 116.5% steam flow at full load A function defined as follows: A P corresponding to 40%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114%

steam flow at full load Coincident with:

Steam Line Pressure - Low 588 psig 614 psig

Turkey Point Units 3 and 4 EPU LAR Att. 1-28 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS Basis for the Change: The accident analyses determined that the analytical limit for the EPU for the high steam flow function is 60% steam flow at 0% load increasing linearly from 20% load to a value of 129% steam flow at full load. The full load value is an increase from the existing value of 120% steam flow. The existing NTS P corresponding to 40% steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114% steam flow at full load is acceptable for the EPU. However, the Allowable Value must be changed to comply with the methodology of WCAP-17070-P.

The accident analyses determined that the analytical limit for the steam line pressure low SI inside containment steam break will need to be revised from the existing value of 432.3 psig to 566.3 psig for the EPU. The current SAL of 432.3 psig remains applicable for the outside containment steam break analysis. Although the SAL is increasing to 566.3 psig, the current nominal trip setpoint of 614 psig has adequate margin to accommodate the new SAL limit and will not change. There is however, a change required to the Allowable Value as shown above to comply with the methodology described in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e., expressed as a value without inequalities) for purposes of COT and Channel Calibration.

The addition of the lead/lag addressed in the new Note (4) on the steamline pressure input causes the safety injection signal to occur significantly faster, reducing the total mass that enters containment.

Licensing Report Sections: LR Section 2.4.1.2.3.2.4, Reactor Protection, Safety Features Actuation, and Control System and LR Section 2.6.3.2, Mass and Energy Release Analysis for Secondary System Pipe Ruptures.

ALLOWABLE VALUE TRIP SETPOINT

4. Steam Line Isolation

d. Steam Line Flow --

High A function defined as follows: A P corresponding to 41.2%

steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114.4% steam flow at full load A function defined as follows: A P corresponding to 40%(a),(b) steam flow at 0% load increasing linearly from 20% load to a value corresponding to 114%(a),(b) steam flow at full load Coincident with:

Steam Line Pressure - Low (4) 607 psig 614(a),(b) psig (4) Time constants utilized in the lead-lag controller for Steam Generator Pressure-Low Steam Line Pressure-Low are t1 50 seconds and t2 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

Turkey Point Units 3 and 4 EPU LAR Att. 1-29 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.28 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 5.c, Feedwater Isolation - SG Water Level - High-High Current TS Proposed TS Basis for the Change: For operational considerations, the top of the instrument span is assumed with allowances for void fraction (maximum reliable indicated level (MRIL)). The uncertainty analysis is based on maintaining the operating limit below the MRIL of 96.8% narrow range span (NRS). The current trip setpoint of 80% NRS has adequate margin to accommodate the MRIL and will not change. There is however, a change required to the Allowable Value shown above to comply with the methodology described in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e., expressed without inequalities) for purposes of COT and Channel Calibration.

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control System.

3.1.29 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 6.b, Auxiliary Feedwater - Steam Generator Water Level - Low-Low Current TS Proposed TS ALLOWABLE VALUE TRIP SETPOINT

5. Fedwater Isolation

c. Steam Generator Water Level High-High 81.9% of narrow range instrument span 80% of narrow range instrument span ALLOWABLE VALUE TRIP SETPOINT

5. Feedwater Isolation

c. Steam Generator Water Level High-High 80.5% of narrow range instrument span 80%(a),(b) of narrow range instrument span ALLOWABLE VALUE TRIP SETPOINT

6. Auxiliary Feedwater

b. Steam Generator Water Level Low-Low 8.15% of narrow range instrument span 10% of narrow range instrument span ALLOWABLE VALUE TRIP SETPOINT

6. Auxiliary Feedwater

b. Steam Generator Water Level Low-Low 15.5% of narrow range instrument span 16%(a),(b) of narrow range instrument span

Turkey Point Units 3 and 4 EPU LAR Att. 1-30 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Basis for the Change: The accident and transient analyses have determined that the analytical limit utilized in the Loss of Normal Feedwater/Loss of AC Power events will not change for the EPU. The steam generator water level low-low reactor trip (and ESFAS initiation) SAL for a Loss of Normal Feedwater/Loss of AC Power event is 4.0% NRS. To account for instrument channel uncertainties for EPU conditions and to provide additional margin for operator action, the low-low setpoint will change from 10.0% NRS to 16.0% NRS and the Allowable Value will change from 8.15% to 15.5% to comply with the methodology described in WCAP-17070-P. The trip setpoint is considered a nominal value (i.e., expressed as a value without inequalities) for purposes of COT and Channel Calibration.

Licensing Report Sections: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control System, LR Section 2.8.5.2.2, Loss of Non-Emergency AC Power to Station Auxiliaries, and LR Section 2.8.5.2.3, Loss of Normal Feedwater Flow.

3.1.30 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 7.b 480v Load Centers Undervoltage Current TS Proposed TS Allowable Value Trip Setpoint Load Center 3A

[ ]

430V+/-5V (10 Sec +/-1sec delay) 3B

[ ]

438V+/-5V (10 Sec +/-1sec delay) 3C

[ ]

434V+/-5V (10 Sec +/-1sec delay) 3D

[ ]

434V+/-5V (10 Sec +/-1sec delay) 4A

[ ]

435V+/-5V (10 Sec +/-1sec delay) 4B

[ ]

434V+/-5V (10 Sec +/-1sec delay) 4C

[ ]

434V+/-5V (10 Sec +/-1sec delay) 4D

[ ]

430V+/-5V (10 Sec +/-1sec delay)

Allowable Value Trip Setpoint Load Center 3A

[ ]

430V+/-3V (10 Sec +/-1sec delay) 3B

[ ]

438V+/-3V (10 Sec +/-1sec delay) 3C

[ ]

434V+/-3V (10 Sec +/-1sec delay) 3D

[ ]

434V+/-3V (10 Sec +/-1sec delay) 4A

[ ]

435V+/-3V (10 Sec +/-1sec delay) 4B

[ ]

434V+/-3V (10 Sec +/-1sec delay) 4C

[ ]

434V+/-3V (10 Sec +/-1sec delay) 4D

[ ]

430V+/-3V (10 Sec +/-1sec delay)

Turkey Point Units 3 and 4 EPU LAR Att. 1-31 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Basis for the change: The evaluations demonstrate that there are no adverse voltage effects on the safety-related 480 V load center buses protected by degraded voltage relays (327I) and under voltage relays (327H). The bounding steady-state and transient-state voltages remain within acceptable limits. Therefore, the degraded voltage relay and under voltage relay settings are not affected by operation at EPU conditions.

The 480 V system evaluations indicate that the degraded voltage relays (327I), under voltage relays (327H) and the inverse-time degraded voltage relays (327T) have sufficient voltage to pickup within the required time frame following a dropout under transient-state conditions.

The design basis for the undervoltage and degraded voltage relay settings are not affected by operation at EPU conditions. The reduced tolerances for the under voltage and degraded voltage trip setpoints result in increased operating voltage margin at the 480V load centers. The existing

+/- 5V tolerance on the under voltage and degraded voltage relay setpoints results in a setpoint span of 10V on the operating voltage range of the 480V load centers which produces a tight operating voltage margin. Reducing the setpoint tolerance to +/-3V will result in a setpoint span of 6V on the operating voltage range thus increasing the operating voltage margin. This tightened tolerance band is within the capability of the relays.

Licensing Report Sections: LR Section 2.3.3, AC Onsite Power System.

3.1.31 Technical Specification Table 3.3-3 ESFAS Instrumentation Trip Setpoints Function 7.c 480v Load Centers Degraded Voltage Current TS Proposed TS Allowable Value Trip Setpoint Load Center 3A

[ ]

424V+/-5V (60 Sec +/-30 sec delay) 3B

[ ]

427V+/-5V (60 Sec +/-30 sec delay) 3C

[ ]

437V+/-5V (60 Sec +/-30 sec delay) 3D

[ ]

435V+/-5V (60 Sec +/-30 sec delay) 4A

[ ]

430V+/-5V (60 Sec +/-30 sec delay) 4B

[ ]

436V+/-5V (60 Sec +/-30 sec delay) 4C

[ ]

434V+/-5V (60 Sec +/-30 sec delay) 4D

[ ]

434V+/-5V (60 Sec +/-30 sec delay)

Allowable Value Trip Setpoint Load Center 3A

[ ]

424V+/-3V (60 Sec +/-30 sec delay) 3B

[ ]

427V+/-3V (60 Sec +/-30 sec delay) 3C

[ ]

437V+/-3V (60 Sec +/-30 sec delay) 3D

[ ]

435V+/-3V (60 Sec +/-30 sec delay)

Turkey Point Units 3 and 4 EPU LAR Att. 1-32 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Basis for the change: The EPU evaluations demonstrate that there are no adverse voltage effects on the safety-related 480 V load center buses protected by degraded voltage relays (327I) and under voltage relays (327H). The bounding steady-state and transient-state voltages remain within acceptable limits. Therefore, the degraded voltage relay and under voltage relay settings are not affected by operation at EPU conditions.

The 480 V system evaluations indicate that the degraded voltage relays (327I), under voltage relays (327H) and the inverse-time degraded voltage relays (327T) have sufficient voltage to pickup within the required time frame following a dropout under transient-state conditions.

The design basis for the undervoltage and degraded voltage relay settings are not affected by operation at EPU conditions. The design basis for the undervoltage and degraded voltage relay settings are not affected by operation at EPU conditions. The reduced tolerances for the under voltage and degraded voltage trip setpoints result in increased operating voltage margin at the 480V load centers. The existing +/- 5V tolerance on the under voltage and degraded voltage relay setpoints results in a setpoint span of 10V on the operating voltage range of the 480V load centers which produces a tight operating voltage margin. Reducing the setpoint tolerance to +/-3V will result in a setpoint span of 6V on the operating voltage range thus increasing the operating voltage margin. This tightened tolerance band is within the capability of the relays.

Licensing Report Section: LR Section 2.3.3, AC Onsite Power System.

3.1.32 Technical Specification Table 4.3-2 ESFAS Instrumentation Surveillance Requirements, Notes Current TS (Table Notes)

None.

Proposed TS (Table Notes)

Add Notes (a) & (b) in Table 4.3-2 to the applicable Channel Calibration and analog channel operational test surveillance requirements for ESFAS Functions listed below to specify as-found and as-left criteria in accordance with TSTF-493:

1.f Safety Injection (SI)-Steam Line Flow-High Coincident with Steam Generator Pressure-Low 4.d Steam Line Isolation-Steam Line Flow-High Coincident with Steam Line Pressure -Low or Tavg-Low 5.c Feedwater Isolation-Steam Generator Water level High-High 4A

[ ]

430V+/-3V (60 Sec +/-30 sec delay) 4B

[ ]

436V+/-3V (60 Sec +/-30 sec delay) 4C

[ ]

434V+/-3V (60 Sec +/-30 sec delay) 4D

[ ]

434V+/-3V (60 Sec +/-30 sec delay)

Allowable Value Trip Setpoint Load Center

Turkey Point Units 3 and 4 EPU LAR Att. 1-33 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 6.b Auxiliary Feedwater(2)-Steam Generator Water Level-Low-Low Note (a) states:

If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

Note (b) states:

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTS) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTS are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field settings) to confirm channel performance. The NTS and methodologies used to determine the as-found and the as-left tolerances are specified in WCAP-17070-P.

Basis for the Change: As defined in 10 CFR 50.36, LSSS are settings for automatic protective devices related to those variables having significant safety functions. 10 CFR 50.36 requires that these limiting settings be included in the Technical Specifications and have appropriate surveillance test performed. Surveillance requirements are established to verify that engineered safety features actuation system instrumentation with an LSSS in TS Table 3.3-3 operates within the boundaries of applicable instrument uncertainty calculations. These instrument tolerances are implemented in plant procedures in accordance with Notes (a) and (b) above which are consistent with the wording in TSTF 493 Rev. 4. The methods used to determine NTS values and summaries of calculations are described in WCAP-17070-P. The implementation of as-left and as-found tolerances verifies that the instrument loops are performing in accordance with uncertainty calculation assumptions and that out-of tolerance conditions are evaluated. If a channel cannot be set within the as-left tolerance band, the channel is declared inoperable and Note (b) applies.

Licensing Report Section: LR Section 2.4.1, Reactor Protection, Safety Features Actuation, and Control Systems.

Turkey Point Units 3 and 4 EPU LAR Att. 1-34 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.33 Technical Specification Table 4.3-2 ESFAS Instrumentation Surveillance Requirements Function 1. Safety Injection and Function 5.c Steam Generator Water Level--High-High Current TS FUNCTIONAL UNIT CHANNEL Check CHANNEL Calibration Analog Channel Operational Test Trip Actuating Device Operational test Actuation Logic test Modes For Which Surveillance Is Required

1. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Ventilation Isolation, Start Diesel Generators, Containment Phase A Isolation (except Manual SI),

Containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water) 5.a. Automatic Actuation Logic and Actuation Relays 1,2 5.c. Steam Generator Water LevelHigh-High 1,2

Turkey Point Units 3 and 4 EPU LAR Att. 1-35 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS Basis for the Change: FPL submitted a license amendment request to the NRC via FPL Letter (L-2009-133), License Amendment Request (LAR) 196 - Alternative Source Term and Conforming Amendment, (Reference 3) to adopt the alternative source term (AST) as allowed by 10 CFR 50.67. As part of the Amendment, credit for the Emergency Containment Filtering System is removed from the licensing basis and SI actuation signals no longer exist for these fans. The remaining list of SI actuated equipment/systems is being removed since they are already identified in the Technical Specification basis and each are controlled under separate specifications. Removal of these items from the technical specifications is consistent with Standard Technical Specifications (NUREG 1431 Rev 3). Mode 3 is added to the column Modes for Which Surveillance is Required to be consistent with new Technical Specification 3.7.1.7, Feedwater Isolation.

Licensing Report Section: None.

3.1.34 Technical Specification 3.4.2 RCS Safety Valves - Shutdown and Operating Current TS SHUTDOWN LCO 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE* with a lift setting of 2485 psig + 2%, -3%.** ***

OPERATING LCO 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig + 2%, -3%.* **

Proposed TS SHUTDOWN LCO 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE* with a lift setting of 2465 psig + 2%, -3%.** ***

FUNCTIONAL UNIT CHANNEL Check CHANNEL Calibration Analog Channel Operational Test Trip Actuating Device Operational test Actuation Logic test Modes For Which Surveillance Is Required

1. Safety Injection 5.a. Automatic Actuation Logic and Actuation Relays 1,2,3 5.c. Steam Generator Water LevelHigh-High 1,2,3

Turkey Point Units 3 and 4 EPU LAR Att. 1-36 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 OPERATING LCO 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2465 psig + 2%, -3%.* **

Basis for the Change: To meet an RCS pressure limit of 2748.5 psia for a Loss of External Electrical Load/Turbine Trip event at EPU conditions, the pressurizer safety valve lift setting was reduced.

Licensing Report Sections: LR Section 2.8.5.2.1, Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, and Steam Pressure Regulator Failure.

3.1.35 Technical Specification 3.4.9 Figure 3.4-2 Turkey Point Units 3 and 4 Reactor Coolant System Heatup Limitations Current TS Figure 3.4-2 Turkey Point Units 3 and 4 Reactor Coolant System Heatup Limitations (Heatup Rate of 60 and 100°F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors)

Proposed TS Figure 3.4-2 Turkey Point Units 3 and 4 Reactor Coolant System Heatup Limitations (Heatup Rate of 60 and 100°F/hr) Applicable for 48 EFPY (Without Margins for Instrumentation Errors) is as shown in Attachment 2 Basis for the Change: Reactor vessel integrity is potentially impacted by any changes in plant parameters that affect neutron fluence levels or temperature/pressure transients. To account for the EPU impact in development of the Pressure-Temperature heatup limit curves and in the 48 EFPY use projections, the changes in neutron fluence resulting from the proposed EPU have been evaluated to determine the impact on reactor vessel integrity.

The EPU fluence projections were then used to redevelop the P-T limit curves at 48 EFPY. The P-T limit curves were developed based on the most limiting axial flaw (forging) and circumferential flaw (circumferential weld) from the reactor vessels. One set of heatup and cooldown curves was developed, applicable to both units. Each set of heatup and cooldown curves were developed based on the limiting 1/4T and 3/4T adjusted reference temperature (ART). Regulatory Guide 1.99, Revision 2 methodology was used to calculate the ART values.

Licensing Report Section: LR Section 2.1.2, Pressure-Temperature Limits and Upper Shelf Energy.

3.1.36 Technical Specification 3.4.9 Figure 3.4-3 Turkey Point Units 3 and 4 Reactor Coolant System Cooldown Limitations Current TS Figure 3.4-3 Turkey Point Units 3 and 4 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100°F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors)

Turkey Point Units 3 and 4 EPU LAR Att. 1-37 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS Figure 3.4-3 Turkey Point Units 3 and 4 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100°F/hr) Applicable for 48 EFPY (Without Margins for Instrumentation Errors) is as shown in Attachment 2 Basis for the Change: Reactor vessel integrity is potentially impacted by any changes in plant parameters that affect neutron fluence levels or temperature/pressure transients. To account for the EPU impact in development of the Pressure-Temperature cooldown limit curves and in the 48 EFPY use projections, the changes in neutron fluence resulting from the proposed EPU have been evaluated to determine the impact on reactor vessel integrity.

The EPU fluence projections were then used to redevelop the P-T limit curves at 48 EFPY. The P-T limit curves were developed based on the most limiting axial flaw (forging) and circumferential flaw (circumferential weld) from the reactor vessels. One set of heatup and cooldown curves was developed, applicable to both units. Each set of heatup and cooldown curves were developed based on the limiting 1/4T and 3/4T adjusted reference temperature (ART). Regulatory Guide 1.99, Revision 2 methodology was used to calculate the ART values.

Licensing Report Section: LR Section 2.1.2, Pressure-Temperature Limits and Upper Shelf Energy.

3.1.37 Technical Specification 3.4.9.3 Overpressure Mitigating Systems Current TS LCO 3.4.9.3 The high pressure safety injection flow paths to the Reactor Coolant System (RCS) shall be isolated, and at least one of the following Overpressure Mitigating Systems shall be OPERABLE:

a.

Two power-operated relief valves (PORVs) with a lift setting of 468 psig, or Proposed TS LCO 3.4.9.3 The high pressure safety injection flow paths to the Reactor Coolant System (RCS) shall be isolated, and at least one of the following Overpressure Mitigating Systems shall be OPERABLE:

a.

Two power-operated relief valves (PORVs) with a lift setting of 448 psig, or Basis for the Change: The design basis mass injection flow rate has increased slightly at EPU conditions. Additionally, the differential pressure between the reactor vessel and the hot leg pressure transmitters has increased from 57.0 to 57.4 psig, and the existing P/T limits for 48 EFPY have changed for the proposed EPU. There are no significant changes to the RCS volumes to Turkey Point Units 3 and 4 components as part of the proposed EPU. The existing OMS setpoint analysis was revised to include the effects of these changes on the existing OMS PORV setpoint for the proposed EPU. The methodology and computer code used in the analysis are the same as for the existing analysis. The existing OMS PORV setpoint was revised from 468 psig to 448 psig.

Turkey Point Units 3 and 4 EPU LAR Att. 1-38 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Licensing Report Section: LR Section 2.8.4.3, Overpressure Protection During Low Temperature Operation.

3.1.38 Technical Specification 3.5.1 Accumulators Current TS Each accumulator shall be demonstrated OPERABLE:

SR 4.5.1.1.b. At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the solution in the water-filled accumulator is between 1950 and 2350 ppm; Proposed TS Each accumulator shall be demonstrated OPERABLE:

SR 4.5.1.1.b. At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the solution in the water-filled accumulator is between 2300 ppm and 2600 ppm; Basis for the Change: The accumulator minimum boron concentration was increased to ensure the core will remain subcritical following a LOCA under EPU conditions. The analysis also confirmed that the proposed accumulator maximum boron concentration would remain soluble and preclude precipitation in the core.

Licensing Report Section: Section 2.8.5.6.3.4, Post-LOCA Subcriticality and Long Term Cooling.

3.1.39 Technical Specification 3/4.5.4 Refueling Water Storage Tank Current TS LCO 3.5.4 For single Unit operation, one refueling water storage tank (RWST) shall be OPERABLE or for dual Unit operation two RWSTs shall be OPERABLE with:

b.

A minimum boron concentration of 1950 ppm of boron, Proposed TS LCO 3.5.4 For single Unit operation, one refueling water storage tank (RWST) shall be OPERABLE or for dual Unit operation two RWSTs shall be OPERABLE with:

b.

A boron concentration between 2400 ppm and 2600 ppm]

Basis for the Change: The RWST minimum boron concentration value was increased to between 2400 ppm and 2600 ppm. The minimum boron concentration has been raised to create an acceptable margin to the core design limit while a maximum has been added to ensure that boric acid precipitation in the core is precluded. The resulting sump boron concentration, which is calculated as a function of the pre-LOCA RCS boron concentration, is reviewed for each cycle-specific core design to confirm that adequate boron exists to maintain subcriticality in the long-term post-LOCA environment at EPU conditions.

Turkey Point Units 3 and 4 EPU LAR Att. 1-39 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Licensing Report Section: LR Section 2.8.5.6.3.4, Post-LOCA Subcriticality and Long Term Cooling.

3.1.40 Technical Specification Table 3.7-1 Maximum Allowable Power Level with Inoperable Steam Line Safety Valves During Three Loop Operation Current TS Proposed TS Basis for the Change: The analyses related to the effects of EPU on overpressure protection capability during power operation were determined to adequately account for the effects of the proposed EPU and concludes the overpressure protection features will continue to provide adequate protection to meet the requirements of its current licensing basis with changes in maximum allowable power level related to the number of inoperable MSSVs as shown in the table. Lower maximum allowable power levels are required to prevent exceeding a main steam system pressure limit of 1208.5 psia for a loss of external electrical load/turbine trip with less than 4 operable MSSVs per steam generator.

Licensing Report Sections: LR Section 2.8.4.2, Overpressure Protection During Power Operation, and LR Section 2.8.5.2.1, Loss of External Electrical Load, Turbine Trip, Loss of Condenser Vacuum, and Steam Pressure Regulator Failure.

Table 3.7-1 Maximum Allowable Power Level With Inoperable Steam Line Safety Valves During Three Loop Operation Maximum Number of Inoperable Safety Valves on Any Operating Steam Generator Maximum Allowable Power Level (Percent of Rated Thermal Power) 1 53 2

33 3

14 Table 3.7-1 Maximum Allowable Power Level With Inoperable Steam Line Safety Valves During Three Loop Operation Maximum Number of Inoperable Safety Valves on Any Operating Steam Generator Maximum Allowable Power Level (Percent of Rated Thermal Power) 1 44 2

27 3

10

Turkey Point Units 3 and 4 EPU LAR Att. 1-40 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 3.1.41 Technical Specification Table 3.7-2 Steam Line Safety Valves Per Loop Current TS Proposed TS Basis for the Change: The results of the loss-of-electrical-load/turbine-trip analysis demonstrate that the secondary system pressure limits are met at the proposed EPU conditions when the nominal lift settings of the two highest main steam safety valves are lowered. To meet a main steam system pressure limit of 1208.5 psia for a loss of external electrical load/turbine trip event, the nominal lift settings of MSSVs RV 1402, 1403, 1407, 1408, 1412, and 1413 were reduced.

Lower maximum allowable power levels are required to prevent exceeding a main steam system pressure limit of 1208.5 psia for a loss of external electrical load/turbine trip with less than 4 operable MSSVs per steam generator.

Licensing Report Sections: LR Section 2.8.4.2, Overpressure Protection During Power Operation, and LR Section 2.8.5.2.1, Loss of External Electrical Load, Turbine Trip, Loss of Condenser Vacuum, and Steam Pressure Regulator Failure.

3.1.42 Technical Specification 3.7.1.6 Standby Feedwater System, Demineralized Water Storage Tank Current TS LCO 3.7.1.6 Two Standby Steam Generator Feedwater Pumps shall be OPERABLE* and at least 135,000 gallons of water (indicated volume), shall be in the Demineralized Water Storage Tank.**

Steam Line Safety Valves Per Loop Valve Number Lift Setting

(+/- 3%***)

Orifice Size Square Inches Loop A Loop B Loop C 1.

RV1400 RV1405 RV1410 1085 psig 16 2.

RV1401 RV1406 RV1411 1100 psig 16 3.

RV1402 RV1407 RV1412 1115 psig 16 4.

RV1403 RV1408 RV1413 1130 psig 16 Steam Safety Valves Per Loop Valve Number Lift Setting

(+/-3%***)

Orifice Size Square Inches Loop A Loop B Loop C 1.

RV1400 RV1405 RV1410 1085 psig 16 2.

RV1401 RV1406 RV1411 1100 psig 16 3.

RV1402 RV1407 RV1412 1105 psig 16 4.

RV1403 RV1408 RV1413 1105 psig 16

Turkey Point Units 3 and 4 EPU LAR Att. 1-41 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 ACTION:

c.

With less than 135,000 gallons of water indicated in the Demineralized Water Storage Tank restore the available volume to at least 135,000 gallons indicated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or submit a SPECIAL REPORT per 3.7.1.6d.

Proposed TS LCO 3.7.1.6 Two Standby Steam Generator Feedwater Pumps shall be OPERABLE* and at least 145,000 gallons of water (indicated volume), shall be in the Demineralized Water Storage Tank.**

ACTION:

c.

With less than 145,000 gallons of water indicated in the Demineralized Water Storage Tank restore the available volume to at least 145,000 gallons indicated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or submit a SPECIAL REPORT per 3.7.1.6d.

Basis for the Change: The Demineralized Water Storage Tank (DWST) indicated volume increased due to EPU. The DWST capacity is based on the inventory required to support decay heat removal from the reactor for six hours for a single unit or two hours for two units. This timeframe is also sufficient to restore the AFW system or establish make-up to the DWST. Due to the EPU, the supply needed from the DWST is increasing from 65,000 to 77,000 gallons. This results in a change in the minimum indicated volume from 135,000 to 145,000 gallons after accounting for instrument uncertainties, for water deemed unusable because of discharge line location, and vortex formation.

Licensing Report Sections: LR Section 2.5.1.4, Fire Protection 3.1.43 Technical Specification 3/4.7.1.7 Plant Systems - Feedwater Isolation Current TS None Proposed TS 3.7.1.7 Six Feedwater Control Valves (FCVs) both main and bypass and six Feedwater Isolation Valves (FIVs) both main and bypass shall be OPERABLE*.

APPLICABILITY MODES 1, 2, and 3**

ACTION:

a.

With one or more FCVs inoperable, restore operability, or close or isolate the inoperable FCVs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify that the inoperable valve(s) is closed or isolated at least once per 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one or more FIVs inoperable, restore operability, or close or isolate the inoperable FIV(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify that the inoperable valve(s) is closed or isolated at least

Turkey Point Units 3 and 4 EPU LAR Att. 1-42 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 once per 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With one or more bypass valves in different steam generator flow paths inoperable, restore operability, or close or isolate the inoperable bypass valve(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify that the inoperable valve(s) is closed or isolated at least once per 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d.

With two valves in the same steam generator flow path inoperable, restore operability, or isolate the affected flowpath within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

• Separate Condition entry is allowed for each valve.

• • The provisions of specification 3.0.4 and 4.0.4 are not applicable SR 4.7.1.7 Each FCV, FIV and bypass valve shall be demonstrated OPERABLE:

a.

At least every 18 months by:

1.

Verifying that each FCV, FIV and bypass valve actuates to the isolation position on an actual or simulated actuation signal.

b.

In accordance with the Inservice Testing Program by:

1.

Verifying that each FCV, FIV and bypass valve isolation time is within limits.

Basis for Change: FP&L has decided to upgrade its TS by proposing a new Specification titled Plant Systems - Feedwater Isolation. The LCO, Applicability, Actions and Surveillance Requirements are consistent with those found in ITS 3.7.3 for Feedwater Isolation. This new PTN Specification reflects consideration of the steam line break analysis. The current main steam line break (MSLB) analysis credits the closure of the Feedwater Control Valves and its associated bypass for isolation of the Feedwater lines on either a Safety Injection Actuation Signal (SIAS) or a Steam Generator High-High water level signal. The backup Feedwater Isolation Valve (FIV) function is currently provided by closure of the Main Feedwater Pump (MFP) Discharge valves.

For EPU, the MSLB will continue to credit the FCVs for primary isolation but new backup feedwater isolation valves for the main and bypass lines will be provided. The proposed technical specification will ensure that the FCVs and the FIVs in each header will close and isolate the feedwater lines on their actuation signal in the required time assumed in the safety analyses.

Licensing Report Sections: LR Section 2.5.5.4, Condensate and Feedwater and LR Section 2.6.3.2, Mass and Energy Release Analysis for Secondary System Pipe Ruptures.

3.1.44 Technical Specification 3/4.8.1 A.C. Sources - Diesel Generator Current TS (General)

SR 4.8.1.1.2 contains a series of surveillance requirements requiring that the diesel generators be demonstrated to be operable by verifying that each diesel starts, accelerates to, and maintains a generator voltage and frequency of 4160 +/- 420 volts and 60 +/- 1.2 Hz under various conditions including automatic start, load rejection, on ESF actuation (with and without loss of offsite power) and during a 24-hour test.

Turkey Point Units 3 and 4 EPU LAR Att. 1-43 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS (General)

SR 4.8.1.1.2 surveillance requirements are changed to require that the diesel generators be demonstrated to be operable by verifying that each diesel starts, accelerates to, and maintains a generator voltage and frequency of 3950-4350 volts and 60 +/- 0.6 Hz under various conditions including automatic start, load rejection, on ESF actuation (with and without loss of offsite power) and during a 24-hour test. Specifically, the following surveillance requirements are changed from a voltage and frequency tolerance of 4160 +/- 420 volts and 60 +/-1.2 Hz to 3950-4350 volts and 60 +/- 0.6 Hz (see Attachment 2):

SR 4.8.1.1.2.a.4), SR 4.8.1.1.2.g.2), SR 4.8.1.1.2 g.4), SR 4.8.1.1.2.g.5), SR 4.8.1.1.2.g.6).b),

SR 4.8.1.1.2 g.7), Footnote ** to SR 4.8.1.1.2.g.7), and SR 4.8.1.1.2.h Basis for the Change: EDG loading under EPU conditions results in minor load additions due to both EPU and non-EPU related activities. Additionally, a loading increase is identified in the EPU-related EDG frequency and voltage evaluation at the extreme limits of these parameters.

Therefore, worst case EDG loading will increase. The EDGs, however, will remain loaded within their ratings with margin, safety-related equipment will operate within their ratings and sufficient voltages will exist to ensure proper functioning equipment under steady state conditions upon incorporation of the indicated changes to the steady state voltage and frequency tolerances.

Licensing Report Sections: LR Section 2.3.3, AC Onsite Power System 3.1.45 Technical Specification 3.9.1 Refueling Operations - Boron Concentration Current TS LCO 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:

a.

A Keff of 0.95 or less, or b.

A boron concentration of greater than or equal to 1950 ppm.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 16 gpm of a solution containing greater than or equal to 3.0 wt% (5245 ppm) boron or its equivalent until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 1950 ppm, whichever is the more restrictive.

Proposed TS LCO 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:

a.

A Keff of 0.95 or less, or

Turkey Point Units 3 and 4 EPU LAR Att. 1-44 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 b.

A boron concentration of greater than or equal to 2300 ppm.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate boration at greater than or equal to 16 gpm of a solution containing greater than or equal to 3.0 wt%

(5245 ppm) boron or its equivalent until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2300 ppm, whichever is the more restrictive.

Basis for the Change: The increase in the boron concentration from 1950 ppm to 2300 ppm will ensure that the reactor remains subcritical during core alterations and that a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel even assuming a boron dilution incident.

Licensing Report Sections: LR Section 2.8.5.4.5, CVCS Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant.

3.1.46 Technical Specification 3.9.14 Refueling Operations - Spent Fuel Storage Current TS LCO 3.9.14 The following conditions shall apply to spent fuel storage:

a.

The minimum boron concentration in the Spent Fuel Pit shall be 1950 ppm.

ACTION:

a.

With boron concentration in the Spent Fuel Pit less than 1950 ppm, suspend movement of spent fuel in the Spent Fuel Pit and initiate action to restore boron concentration to 1950 ppm or greater.

SR 4.9.14 The boron concentration of the Spent Fuel Pit shall be verified to be 1950 ppm or greater at least once per month.

Proposed TS LCO 3.9.14 The following conditions shall apply to spent fuel storage:

a.

The minimum boron concentration in the Spent Fuel Pit shall be 2300 ppm.

ACTION:

a.

With boron concentration in the Spent Fuel Pit less than 2300 ppm, suspend movement of spent fuel in the Spent Fuel Pit and initiate action to restore boron concentration to 2300 ppm or greater.

SR 4.9.14 The boron concentration of the Spent Fuel Pit shall be verified to be 2300 ppm or greater at least once per month.

Basis for the Change: Although the criticality analysis (Attachment 10) assumed 1600 ppm in the Spent Fuel Pit to assure keff < 0.95 under the worst case accident conditions at EPU, the higher boron concentration of 2300 ppm is proposed because, during refueling, the water volume in the

Turkey Point Units 3 and 4 EPU LAR Att. 1-45 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Spent Fuel Pit, the transfer canal, the refueling canal and the reactor vessel form a single mass.

The proposed value provides significant margin to this requirement.

Licensing Report Sections: LR Section 2.8.6.2 Spent Fuel Storage 3.1.47 Technical Specification 5.5.1 Fuel Storage - Criticality Current TS LCO 5.5.1.1 The spent fuel storage racks shall be maintained with:

b.

A keff less than or equal to 0.95 when flooded with water borated to 650 ppm, which includes an allowance for biases and uncertainties as described in UFSAR Chapter 9.

c.

A nominal 10.6 inch center-to-center distance for Region I and 9.0 inch center-to-center distance for Region II for the two region spent fuel pool storage racks. A nominal 10.1 inch center-to-center distance in the east-west direction and a nominal 10.7 inch center-to-center distance in the north-south direction for the Region I cask area storage rack.

d.

The maximum enrichment loading for fuel assemblies is 4.5 weight percent of U-235.

f.

Fresh or irradiated fuel assemblies not stored in the cask area storage rack shall be stored in accordance with Specification 5.5.1.3 or configurations that have been shown to comply with Specification 5.5.1.la and 5.5.1.1b using the NRC approved methodology in UFSAR Chapter 9.

LCO 5.5.1.2 The racks for new fuel storage are designed to store fuel in a safe subcritical array and shall be maintained with:

b.

Fuel assemblies placed in the New Fuel Storage Area shall contain no more than 4.5 weight percent of U-235.

LCO 5.5.1.3 Credit for burnup and cooling time is taken in determining placement locations for spent fuel in the two-region spent fuel racks. Fresh or irradiated fuel assemblies shall be stored in compliance with the following:

a.

Any 2X2 array of Region I storage cells shall comply with the storage patterns in Figure 5.5-1 and the requirements of Table 5.5-1 and 5.5-2, as applicable. The reactivity rank of fuel assemblies in the 2X2 array (rank determined using Table 5.5-3) shall be equal to or less than that shown for the 2X2 array.

b.

Any 2X2 array of Region II storage cells containing fuel shall:

i.

Comply with the storage patterns in Figure 5.2-2 and the requirements of Table 5.5-1 and 5.5-2, as applicable. The reactivity rank of fuel assemblies in the 2X2 array (rank determined using Table 5.5-3) shall be equal to or less than that shown for the 2X2 array.

c.

Any 2X2 array of Region II storage cells that interface with Region I shall comply with the rules of Figure 5.5-3. Arrays II-E and II-F may interface with Region I without special restriction.

d.

Any 2X2 array of Region II storage cells may adjoin a row of assemblies with a reactivity rank of II-2 (or lower) that is located in the outer row adjacent to the spent fuel pit wall. The outer

Turkey Point Units 3 and 4 EPU LAR Att. 1-46 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 row of reactivity rank II-2 (or lower) fuel assemblies need not contain any Metamic inserts or full length RCCAs, as long as the following additional requirements are met:

i.

Fuel is loaded to comply with the allowable storage patterns defined in Figure 5.5-4, and ii.

Arrays II-E and II-F are loaded without any additional restriction on that 2X2 array. Arrays II-E and II-F do not have empty cells, Metamic inserts, or RCCAs, that restrict the interface with the adjoining rank II-2 (or lower) fuel assemblies.

Table 5.5 Blanketed Fuel-Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)

Table 5.5 Non-Blanketed Fuel-Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)

Table 5.5 Fuel Categories Ranked by Reactivity Figure 5.5 Allowable Region I Storage Arrays Figure 5.5 Allowable Region II Storage Arrays Figure 5.5 Allowable Interfaces Between Region II Storage Arrays Figure 5.5 Allowable Region II Storage Adjacent to Spent Fuel Pit Walls Proposed TS LCO 5.5.1.1 The spent fuel storage racks shall be maintained with:

b.

A keff less than or equal to 0.95 when flooded with water borated to 500 ppm, which includes an allowance for biases and uncertainties as described in UFSAR Chapter 9.

c.

A nominal 10.6 inch center-to-center distance for Region I and 9.0 inch center-to-center distance for Region II for the two region spent fuel pool storage racks. A nominal 10.1 inch center-to-center distance in the east-west direction and a nominal 10.7 inch center-to-center distance in the north-south direction for the cask area storage rack.

d.

The maximum enrichment loading for fuel assemblies is 5.0 weight percent of U-235.

f.

Fresh or irradiated fuel assemblies not stored in the cask area storage rack shall be stored in accordance with Specification 5.5.1.3 or configurations that have been shown to comply with Specification 5.5.1.la and 5.5.1.1b using NRC approved methodology in UFSAR Chapter 9.

LCO 5.5.1.2 The racks for new fuel storage are designed to store fuel in a safe subcritical array and shall be maintained with:

b.

Fuel assemblies placed in the New Fuel Storage Area shall contain no more than 4.7 weight percent of U-235 if the assembly contains no burnable absorber rods and no more than 5.0 weight percent of U-235 if the assembly contains at least 16 IFBA rods (or an equivalent amount of other burnable absorber).

Turkey Point Units 3 and 4 EPU LAR Att. 1-47 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 LCO 5.5.1.3 Credit for burnup and cooling time is taken in determining placement locations for spent fuel in the two-region spent fuel racks. Fresh or irradiated fuel assemblies shall be stored in compliance with the following:

a.

Any 2X2 array of Region I storage cells shall comply with the storage patterns in Figure 5.5-1 and the requirements of Table 5.5-1 and 5.5-2, as applicable. The reactivity rank of fuel assemblies in the 2X2 array (rank determined using Table 5.5-3) shall be equal to or less reactive than that shown for the 2X2 array.

b.

Any 2X2 array of Region II storage cells containing fuel shall:

i.

Comply with the storage patterns in Figure 5.2-2 and the requirements of Table 5.5-1 and 5.5-2, as applicable. The reactivity rank of fuel assemblies in the 2X2 array (rank determined using Table 5.5-3) shall be equal to or less reactive than that shown for the 2X2 array.

c.

Any 2X2 array of Region II storage cells that interface with Region I storage cells shall comply with the rules of Figure 5.5-3. Arrays II-E and II-F may interface with Region I without special restriction.

d.

Any 2X2 array of Region II storage cells may adjoin assemblies that are located in the outer row adjacent to the spent fuel pit wall. The outer row of fuel assemblies facing the pool wall need not contain any Metamic inserts or full length RCCAs, as long as the fuel is loaded to comply with the allowable storage patterns defined in Figure 5.5-4.

All Tables and Figures are replaced as shown in Attachment 2. New titles, as applicable are shown below:

Table 5.5 Blanketed Fuel-Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)

Table 5.5 Non-Blanketed Fuel-Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)

Table 5.5 Fuel categories Ranked by Reactivity Figure 5.5 Allowable Region I Storage Arrays Figure 5.5 Allowable Region II Storage Arrays Figure 5.5-3 -Allowable Interface Restrictions Between Region II and Region I Arrays Figure 5.5 Allowable Exceptions to Region II Storage Arrays When Adjacent to Spent Fuel Pit Walls Basis for the Change: License Amendment Request No. 207 (Reference 16) submitted proposed TS changes and a new criticality analysis to revise the current licensing basis for both new fuel and spent fuel storage. The LAR 207 TS changes included new spent fuel storage patterns that account for both the increase in fuel maximum enrichment from 4.5 wt% U-235 to 5.0 wt% U-235 and the impact on the fuel of the higher power operation proposed under the EPU. It did not, however, actually propose that the fuel enrichment limit of 4.5 wt% U-235 be changed. In this EPU LAR, the proposed TS changes are identical except that changes to

Turkey Point Units 3 and 4 EPU LAR Att. 1-48 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 TS 5.5.1.1.d and 5.5.1.2.b would increase the allowable enrichment in the spent fuel storage racks to 5.0 wt% U-235 and in the New Fuel Storage Area to 4.7% U-235 or up to 5.0 wt% U-235 with 16 or more IFBA rods or an equivalent amount of other burnable absorber. Since the LAR 207 basis and justification for its proposed TS changes included the impact of the higher enrichment and operation under EPU conditions, it also applies to and supports the proposed EPU TS changes.

Licensing Report Sections: LR Section 2.8.2 Nuclear Design, LR Section 2.8.6.1 New Fuel Storage and LR Section 2.8.6.2 Spent Fuel Storage.

3.1.48 Technical Specification Table 5.6-1 Component Cyclic or Transient Limits Current TS COMPONENT CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT Reactor Coolant System 200 heatup cycles at 100°F/h and 200 cooldown cycles at 100°F/h Heatup cycle - Tave from 200°F to 550°F Cooldown cycle - Tave from 550°F to 200°F 200 pressurizer cooldown cycles at 200°F/h Pressurizer cooldown cycle temperature from 650°F to 200°F 80 loss of load cycles, without immediate Turbine or Reactor trip 15% of RATED THERMAL POWER to 0% of RATED THERMAL POWER 40 cycles of loss-of-offsite A.C.

electrical power Loss-of-offsite A.C. electrical ESF Electrical System 80 cycles of loss of flow in one reactor coolant loop Loss of only one reactor coolant pump 400 reactor trip cycles 100% to 0% of RATED THERMAL POWER 150 leak tests Pressurized to 2435 psig 5 hydrostatic pressure tests Pressurized to 3100 psig Secondary Coolant System 6 loss of secondary pressure Loss of secondary pressure 50 leak tests Pressurized to 1085 psig 35 hydrostatic pressure tests Pressurized to 1356 psig

Turkey Point Units 3 and 4 EPU LAR Att. 1-49 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS Basis for the Change: As stated in LR Section 2.2.6, NSSS Design Transients, the transients listed in LR Table 2.2.6-1 and their associated frequencies of occurrence at EPU conditions are unchanged from those in the current design basis. Table 5.6-1 is proposed to be updated, however, to clarify that there is a separate limit for the number of pressurizer cooldown cycles COMPONENT CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT Reactor Coolant System 200 heatup cycles at 100°F/h and 200 cooldown cycles at 100°F/h Heatup cycle - Tave from 200°F to 550°F Cooldown cycle - Tave from 550°F to 200°F 200 pressurizer cooldown cycles at 200°F/h from nominal pressure Pressurizer cooldown cycle temperature from 650°F to 200°F 200 pressurizer cooldown cycles at 200°F/h from 400 psia Pressurizer cooldown cycle temperature from 650°F to 200°F 80 loss of load cycles, without immediate Turbine or Reactor trip 15% of RATED THERMAL POWER to 0% of RATED THERMAL POWER 40 cycles of loss-of-offsite A.C.

electrical power Loss-of-offsite A.C. electrical ESF Electrical System 80 cycles of loss of flow in one reactor coolant loop Loss of only one reactor coolant pump 400 reactor trip cycles 100% to 0% of RATED THERMAL POWER 10 cycles of inadvertent auxiliary spray Spray water temperature differential to 560°F 150 primary to secondary side leak tests Pressurized to 2435 psig 15 primary to secondary side leak tests Pressurized to 2250 psig 5 hydrostatic pressure tests Pressurized to 2485 psig and 400°F Secondary Coolant System 6 loss of secondary pressure Loss of secondary pressure 50 hydrostatic pressure tests Pressurized to 1085 psig 10 hydrostatic pressure tests Pressurized to 1356 psig 15 secondary to primary side leak tests Pressurized to 840 psig

Turkey Point Units 3 and 4 EPU LAR Att. 1-50 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 from nominal pressure and from 400 psia. Transients are also proposed to be added and one to be deleted from Table 5.6-1 to make the selection of transients listed consistent with NUREG-0452.

Licensing Report Section: LR Section 2.2.6, NSSS Design Transients.

3.1.49 Technical Specification 6.8.4.h Containment Leakage Rate Testing Program Current TS The peak calculated containment interval pressure for the design basis loss of coolant accident, Pa, is 49.9 psig.

Proposed TS The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is defined here as the containment design pressure of 55 psig.

Basis for the Change: The limiting containment peak pressure resulting from a LOCA under EPU conditions is 53.4 psig. Assuming that the peak containment pressure is the containment design pressure for purposes of the containment leakage rate testing program is therefore conservative.

Licensing Report Section: LR Section 2.6.1, Primary Containment Functional Design.

3.1.50 Technical Specification 6.9.1.7 Core Operating Limits Report Current TS The analytical methods used to determine the AFD limits shall be those previously reviewed and approved by the NRC in:

1.

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

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

3.

WCAP-10054-P, Addendum 2, Revision 1 (proprietary), Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection in the Broken Loop and Improved Condensation Model, October 1995.*

4.

WCAP-12945-P, Westinghouse Code Qualification Document for Best Estimate LOCA Analysis, Volumes I-V, June 1996**

Turkey Point Units 3 and 4 EPU LAR Att. 1-51 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Proposed TS The analytical methods used to determine the AFD limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP-10216-P-A, Revision 1A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL OF FQ SURVEILLANCE TECHNICAL SPECIFICATION, February 1994.

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

3.

WCAP-10054-P-A, Addendum 2, Revision 1 (proprietary), Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection in the Broken Loop and COSI Condensation Model, July 1997.

4.

WCAP-16009-P-A, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),

January 2005 Basis for the Change: The proposed changes incorporate updated NRC-approved methodologies.

Licensing Report Section: None.

3.2 Licensing Basis Changes Changes to the PTN licensing basis that are not directly reflected in the Technical Specifications, but are consistent with EPU analysis are:

3.2.1 Auxiliary Feedwater Supply The original sizing of the Condensate Storage Tank (CST) for each unit was based on allowing the Auxiliary Feedwater system (AFW) to take each unit from full power to hot standby, maintain hot standby conditions for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and then cool to the Residual Heat Removal entry point of 350°F in four hours or, in the alternative, keep each unit at hot standby for 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. As a result of NRC concerns with the possibility of reactor vessel head voiding during natural circulation cooldown, as expressed in Generic Letter 81-21, FPL implemented a procedure that requires the operator to limit natural circulation cooldown to 25°F per hour while maintaining a 50°F subcooling margin. This process takes about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Before cooling down at this rate, the plant could be kept at hot standby for about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. As a result, the current licensing basis was modified to maintain hot standby conditions for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to commencing a natural circulation cooldown to RHR entry conditions.

The increased core thermal power and resulting higher heat removal requirements would tend to increase the seismic Category 1 inventory requirements from the CST. The licensing basis for the sizing of the CST under EPU conditions, however, will be revised to meet the NRCs requirement in Branch Technical Position (BTP) 5-4 for a Seismic Category I water supply for AFW with sufficient inventory to permit operation at hot standby for at least four hours followed by a cooldown to RHR entry conditions. In conformance FPLs response to GL 81-21, the licensing

Turkey Point Units 3 and 4 EPU LAR Att. 1-52 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 basis will also limit the natural circulation cooldown rate to 25°F per hour with 50°F subcooling.

This would take about 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. The CST inventory requirements will actually be slightly decreased under the BTP 5-4 criteria at EPU conditions and a Technical Specification change is therefore not necessary. See LR Section 2.5.4.5, Auxiliary Feedwater System and LR Section 2.8.7.2, Natural Circulation Cooldown.

3.2.2 Revised Definition of Design Basis Step Load Decrease The definition of the design basis step-load decrease of 50% is revised to mean a rapid load decrease equivalent to 50% of the EPU core thermal power at a maximum turbine unloading rate of 200%/minute. This is a realistic representation of an actual load rejection transient in the plant and is consistent with uprating projects previously performed on other Westinghouse plants. See, LR Section 2.4.2.1, Plant Operability (Margin to Trip).

3.2.3 Credit for Swing High Head Safety Injection Pump during LOCA Standard LOCA methods have been applied to the EPU Turkey Point Analysis which includes a single failure of one emergency diesel generator resulting in the loss of one train of safety injection. This assumption leads to one high head safety injection pump from the affected unit being available to provide flow to the RCS. In addition to the affected units available High Head Safety Injection (HHSI) pump, the Turkey Point EPU analysis credits one swing high head safety injection pump from the unaffected unit being aligned to supply flow to the affected unit. See LR Sections 2.8.5.6.3.3 and 2.8.5.6.3.4.

3.2.4 Measurement Uncertainty Recapture The measurement uncertainty recapture is based on installation of improved feedwater flow and temperature instrumentation using the LEFM technology. A summary of the technical evaluation of the system at PTN is provided in LR Section 2.4.4 and the detailed engineering reports in. The reports demonstrate that the LEFMs will be installed in accordance with the NRC regulatory criteria as contained in NRC Regulatory Issue Summary (RIS) 2002-3, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications (Reference 2).

3.2.5 Analytical Methodologies Appendix A to Attachment 4, Safety Evaluation Report Compliance, supplements the information provided in LR Section 2.8.5.0, Accident and Transient Analyses on NRC-approved codes and methods used in the PTN Extended Power Uprate. The appendix addresses compliance with the limitations, restrictions and conditions specified in the approving safety evaluation of the applicable codes.

The methodologies and computer codes that are being applied at PTN for the first time are identified below:

1.

Calculation of containment response following a postulated main steam line break and LOCA and calculation of the long-term post-reflood releases are analyzed using the GOTHIC computer code (References 7 and 8). See Attachment 4, LR Section 2.6.3.1, Mass and

Turkey Point Units 3 and 4 EPU LAR Att. 1-53 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 Energy Releases for Postulated Loss of Coolant Accidents, and LR Section 2.6.1, Primary Containment Functional Design.

2.

The thermal-hydraulic DNB analysis of the fuel uses the VIPRE-01 computer code (Reference 9). Refer to Attachment 4, LR Section 2.8.3, Thermal and Hydraulic Design. It is used specifically for the following transients: increase in feedwater flow (LR Section 2.8.5.1.2), steam system piping failures from hot zero and full power (LR Section 2.8.5.1.2), loss of reactor coolant flow (LR Section 2.8.5.1.2), locked reactor coolant pump rotor (LR Section 2.8.5.3.2), uncontrolled rod withdrawal from a subcritical condition (LR Section 2.8.5.4.1) and dropped rod cluster control assembly (LR Section 2.8.5.4.3).

3.

RETRAN (Reference 10) is used for analyses of the transient responses at PTN for the first time. See Attachment 4, LR Section 2.8.5, Accident and Transient Analyses. It is used specifically for the following transients: excessive load increase (LR Section 2.8.5.1.1),

increase in feedwater flow (LR Section 2.8.5.1.1), steam system piping failure from hot zero and full power (LR Section 2.8.5.1.2), loss of electrical load (LR Section 2.8.5.2.1), loss of all AC (LR Section 2.8.5.2.2), loss of normal feedwater (LR Section 2.8.5.2.3), loss of reactor coolant flow (LR Section 2.8.5.3.1), locked reactor coolant pump rotor (LR Section 2.8.5.3.2),

and uncontrolled rod cluster control assembly at power (LR Section 2.8.5.4.2).

4.

The loss of offsite AC Power (LR Section 2.8.5.2.2) and the loss of normal feedwater flow analyses (LR Section 2.8.5.4.2.3) use the RETRAN computer code, including the RETRAN thick metal mass heat transfer model.

5.

RETRAN has replaced LOFTRAN in the analysis of the mass and energy releases from steam line breaks inside containment. See LR Section 2.6.3.2, Mass and Energy Releases for Secondary System Pipe Ruptures.

6.

A large-break LOCA best estimate analysis (BE LBLOCA) was used for the Turkey Point 3 and 4 large break LOCAs using the ASTRUM (Reference 11) statistical approach methodology to develop the Peak Cladding Temperature and oxidation results at the 95th percentile. See LR Section 2.8.5.6.3.2, Emergency Core Cooling System and Loss-of-Coolant-Accidents.

7.

For the large break LOCA, the cold leg recirculation interruption time is modeled in the thermal-hydraulic code WCOBRA/TRAC using 10 CFR 50 Appendix K decay heat. See LR Section 2.8.5.6.3.4, Post-LOCA Subcriticality and Long-Term Cooling.

8.

PTN applies the interim boric acid precipitation model consistent with the guidance authorized by the NRC in Reference 13.

3.3 Plant Modification Changes The major plant modifications associated with the EPU are discussed in LR Section 1.0, Introduction to the PTN Extended Power Uprate Licensing Report. These modifications are generally associated with changing piping, valves, pumps and heat exchangers to increase feedwater, condensate, and main steam system flows; secondary BOP instrument and relief valve set point changes; replacement of the HP turbine and condenser for increased steam flow; turbine and drain system control system upgrades; increased capacity for generator,

Turkey Point Units 3 and 4 EPU LAR Att. 1-54 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 transformers and switchyard equipment; reactor coolant, reactor protection, engineered safety features systems instrument set point, relief and control changes; heat exchanger replacement for spent fuel pool and turbine building cooling water systems to increase capacity; increased auxiliary feedwater capacity; increase Technical Support Center (TSC) shielding; improved feedwater isolation capability; increase boron concentrations; and increased fuel enrichment.

All of the modifications that support operation at the uprate power are being implemented pursuant to 10 CFR 50.59 except for the use of the LEFM and fuel enrichment.

As discussed in LR 2.4.4, the Measurement Uncertainty Recapture is being accomplished through the installation of the LEFM system which will reduce feedwater flow uncertainty. Its installation will be accomplished through the 10 CFR 50.59 process but it will not be used for the purpose of determining power levels for adherence to Technical Specification limits.

Use of 5.0 weight percent enriched fuel will be used to optimize fuel efficiency at the increased power level. Lower enrichment would require increased fuel shuffles and operating cost.

Approval is also required for all the setpoint changes associated with each of the requested Technical Specification changes listed in Section 3.1 above.

The safety analyses contained in this LAR assume that the proposed changes, including Technical Specification changes, in the FPLs Alternative Source Term LAR for PTN (Reference 3) have been approved by the NRC and implemented prior to operation at the uprated power level.

4.0 TECHNICAL ANALYSIS

The acceptability of proposed Renewed Facility Operating License, Technical Specification, and Licensing Basis change is addressed in Attachment 4, EPU Licensing Report. Attachment 4 summarizes the evaluations performed to assure acceptable operation at EPU conditions, and provides technical justification for the EPU related changes.

5.0 ENVIRONMENTAL EVALUATION The environmental considerations evaluation is contained in Attachment 7, Supplemental Environmental Report. It concludes that EPU will not result in a significant change in non-radiological impacts on land use, water use, waste discharges, terrestrial and aquatic biota, transmission facilities, or social and economic factors, and will have no non-radiological environmental impacts other than those evaluated in the Supplemental Environmental Report.

The Supplemental Environmental Report further concludes that EPU will not introduce any new radiological release pathways, will not result in a significant increase in occupational or public radiation exposures, and will not result in significant additional fuel cycle environmental impacts.

PTN has determined that operation with the proposed EPU license amendment would neither result in any significant change in the types, or a significant increase in the amounts, of any effluent that may be released offsite nor involve a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed license amendment is eligible for categorical exclusion as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no

Turkey Point Units 3 and 4 EPU LAR Att. 1-55 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 environmental impact statement or environmental assessment is needed in connection with the approval of the proposed license amendment.

6.0 REGULATORY ANALYSIS

6.1 Applicable Regulatory Requirements/Criteria The proposed license amendments will revise the PTN Renewed Facility Operating Licenses Nos. DPR-31 and DPR-41 and the Technical Specifications in order to increase the core thermal power by approximately 15% from 2300 MWt to 2644 MWt. The proposed changes are described in detail in this license amendment request and are also indicated on the marked-up pages of the Renewed Facility Operating License and Technical Specifications contained in. Corresponding Technical Specification bases changes that are being proposed to reflect the changes to the Technical Specifications are provided in Attachment 3, for information only.

As described in detail in Attachment 4, FPL has evaluated the proposed changes and determined that applicable regulations and requirements continue to be met, that exemptions or relief from regulatory requirements are not required, and that they do not affect conformance with any General Design Criterion (GDC) differently than described in the Updated Final Safety Analysis Report (UFSAR).

6.2 No Significant Hazards Consideration FPL has evaluated whether or not a significant hazards consideration is involved with the proposed amendments in Section 3.1 above by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1.

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

Response: No.

Comprehensive analyses and evaluations have been performed to asses the impact of operation under EPU conditions on the nuclear steam supply systems (NSSS) and balance of plant (BOP) systems and components, including the nuclear fuel, safety and dose consequence analyses. They demonstrate that PTN meets applicable design and licensing requirements. These analytical efforts addressed the proposed changes as applicable, including changes to Technical Specifications, licensing basis and systems and components as described in Sections 3.1, 3.2 and 3.3 above.

The fission product barriers -- fuel cladding, reactor coolant pressure boundary, and the containment building - remain unchanged. The spectrum of previously analyzed postulated accidents and transients was evaluated, and effects of EPU on the fuel, the reactor coolant pressure boundary, and the containment were determined. These analyses were performed consistent with the proposed Technical Specification changes. The results demonstrate that existing reactor coolant pressure boundary and containment limits are met and that effects on the fuel are such that dose consequences meet existing criteria at EPU conditions.

Turkey Point Units 3 and 4 EPU LAR Att. 1-56 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 FPL evaluated dose consequences using EPU parameters and determined that doses at the exclusion area and low population zone boundaries and in the control room continue to meet the regulatory limits of 10 CFR 50.67.

The reviews for NSSS and BOP systems and components confirmed that they will function as designed and applicable performance requirements will be satisfied. None of the proposed changes are initiators of any design basis accident or event except for the HP turbine and control system modifications since they can cause a 100% loss of load. However, the turbine modifications provide equipment and control systems with equal or greater performance and reliability. Control system studies demonstrated that plant response to operational transients under EPU conditions does not significantly increase reactor trip frequency, so there will be no significant increase in the frequency of SSC challenges caused by reactor trip.

Therefore, the proposed changes described above, 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?

Response: No.

The EPU does not create new failure modes for existing SSCs. Modified components do not introduce failures different from those of the components in their pre-modified condition.

Consequently, no new or different accident sequences arise from SSC interactions or failures.

The installation and use of the LEFM system will not create the possibility of a new or different kind of accident from any accident previously analyzed. The analyses of record were revaluated and reanalyzed as necessary, to support the measurement uncertainty recapture of 1.7% through use of the LEFM consistent with the guidance provided in RIS-2002-3. The results demonstrate compliance with applicable analysis limits.

Deletion of the pressurizer backup heater actuation on high pressurizer level will not create the possibility of a new or different kind of accident from any accident previously evaluated because pressurizer heater actuation was not credited in the EPU analysis.

The new feedwater isolation replaces the existing isolation valve and re-locates it closer to the feedwater control valve to eliminate a portion of the un-isolatable piping volume and improve isolation response. Therefore this change does not add any new functions or interactions to the system.

The addition of a lead/lag to the steam generator pressure signals does not eliminate or change the protection logic and meets all the system redundancy and design requirements.

The added module provides earlier detection of a MSLB.

Training will be provided to address EPU effects, and the plants simulator will be updated consistent with EPU conditions. Operating procedure changes are minor and do not result in any significant changes in operating philosophy since the existing computer system will be used to control the turbine and Heater Drain Tank Systems. For these reasons, the EPU does not introduce human performance issues that could create new accidents or different

Turkey Point Units 3 and 4 EPU LAR Att. 1-57 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 accident sequences. The increase in power level does not create new fission product release paths. The fission product barriers - fuel cladding, reactor coolant pressure boundary, and the containment building -- remain unchanged.

Consequently, no new accident scenarios, failure mechanisms or single failures are introduced as a result of the proposed changes. All systems, structures and components previously required for the mitigation of an event remain capable of fulfilling their intended design function. The proposed changes will not have an adverse effect on any safety-related system or component, and will not challenge the performance or integrity of any safety related system.

Therefore, none of the proposed changes create the possibility of a new or different kind of accident from any previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Fuel performance evaluations were performed using parameter values appropriate for a reload core operating at EPU conditions. Those evaluations demonstrate that fuel performance acceptance criteria continue to be met. Reload evaluation processes ensure that fuel in the first cycle to be operated at the increased power level, will meet regulatory criteria. LOCA and non-LOCA safety analyses were performed under EPU conditions.

Emergency core cooling system performance was shown to meet the criteria of 10 CFR 50.46. The non-LOCA events identified in the PTN UFSAR Chapter 14 were shown to meet existing acceptance criteria. The LOCA and non-LOCA analyses were performed consistent with the proposed Technical Specification changes which are being implemented to ensure there are no significant reductions in safety margins are maintained. These revisions will not adversely impact plant safety since they will not degrade the ability of systems, structures or components important for the mitigation of a design basis accident or to adversely impact the ability of SSCs to safely shutdown the plant.

The effects of operation at EPU conditions on the reactor coolant pressure boundary (RCPB) have been analyzed. The analyses demonstrate that the RCPB materials will continue to be acceptable following implementation of EPU and will continue to meet the applicable requirements of its current licensing basis.

The containment building response to mass and energy releases was evaluated under EPU conditions. The evaluations showed that temperature and pressure limits were met. No plant changes associated with the EPU reduce the degree of component or system redundancy.

Existing Technical Specification operability and surveillance requirements are not reduced by the proposed changes, thus no margins of safety are reduced.

Structural evaluations performed at EPU conditions demonstrated that calculated loads on affected SSCs remain within their design allowables for all design basis event categories.

ASME Code fatigue limits continue to be met.

The radiological changes will not involve a significant reduction in a margin of safety because PTN compliance with the limits set forth in 10 CFR 20, 10 CFR 50, Appendix I, 40 CFR 190

Turkey Point Units 3 and 4 EPU LAR Att. 1-58 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 and 10 CFR 50.67, as supplemented by Regulatory Guide 1.183, will be maintained under EPU conditions.

An assessment of the cumulative effect of the proposed changes provides reasonable expectation that collectively they will not result in a significant reduction in the overall margin of safety. The results of the analyses demonstrate that the applicable design and safety criteria and regulatory requirements will continue to be met following approval of the proposed changes. An assessment of the proposed changes to Technical Specification limits has determined that that collectively they do not involve a significant reduction in a margin of safety.

Therefore, none of the proposed changes involve a significant reduction in a margin of safety.

Based on the above, FPL concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of - no significant hazards consideration is justified.

6.3 Summary In accordance with the requirements of 10 CFR 50.90, FPL hereby requests amendments to Renewed Facility Operating Licenses DPR-31 and DPR-41 for PTN. The purpose of the proposed LAR is to revise the Renewed Facility Operating Licenses and the Technical Specifications to allow operation at an increased licensed core thermal power of 2644 MWt.

FPL has evaluated the proposed LAR in accordance with 10 CFR 50.91 against the standards in 10 CFR 50.92 and has determined that the operation of PTN Units 3 and 4 in accordance with the proposed LAR presents no significant hazards and therefore, a finding of no significant hazards consideration is justified.

A comprehensive review of accident analyses, component and system analyses, and radiological dose consequences was performed for the EPU. Analyses met the appropriate criteria, as explained in the no significant hazards determination.

Operation of PTN Units 3 and 4 in accordance with the proposed amendments will not result in a significant increase in the probability or consequences of any accident previously analyzed; will not result in a new or different kind of an accident from any accident previously analyzed; and will not result in a significant reduction in margin of safety. Therefore, operation of PTN Units 3 and 4 in accordance with the proposed amendments does not involve a significant hazards consideration.

7.0 REFERENCES

1.

RS-001, Review Standard for Extended Power Uprates, NRC, December 2003.

2.

RIS 2002-3, Guidance on the Content of Measurement Uncertainty Recapture Uprate Applications, NRC, 2002.

3.

PTN LAR 196, Alternative Source Term, June 25, 2009 (ML092050277).

4.

Updated Final Safety Analysis Report (UFSAR) for PTN, Section 14.3.3.

Turkey Point Units 3 and 4 EPU LAR Att. 1-59 Renewed Facility Operating License, Technical Specifications, and Licensing Basis Changes Turkey Point Units 3 and 4 L-2010-113 Docket Nos. 50-250 and 50-251 5.

WCAP-17070-P, Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4.

6.

RIS-2006-17, NRC Staff Position on Requirements of 10 CFR 50.36, Technical Specifications, Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels.

7.

NAI 8907 06, Revision 16, GOTHIC Containment Analysis Package Technical Manual, Version 7.2a, January 2006.

8.

NAI 8907 09, Revision 9, GOTHIC Containment Analysis Package Qualification Report, Version 7.2a, January 2006.

9.

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

10. WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, April 1999.

11. WCAP-16009-P-A, Nissley, M.E., et al, 2005, Realistic Large Break LOCA Evaluation Methodology using the Automated Statistical Treatment of Uncertainty Method (ASTRUM).

12. WCAP-10698-P-A, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, August 1987.

13. SLIDES for the Summary of August 23, 2006 Meeting with the Pressurized Water Reactor Owners Group (PWROG) to Discuss the Status of Program to Establish Consistent Criteria for Post Loss-of-Coolant (LOCA) Calculations, October 3, 2006 (ML062720565).

14. NRC, 1999, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1431, Vol. 1, Addendum 1), Division of Regulatory Improvement Programs, Office of Nuclear Reactor Regulation, August 1999.

15. WCAP-17094-P, Revision 2, Turkey Point Units 3 & 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, July 2010.

16. FPL to U.S. Nuclear Regulatory Commission, License Amendment Request No. 207, Fuel Storage Criticality Analysis, August 5, 2010.