ML17027A078

Issuance of Amendments Regarding Technical Specifications for Inservice Testing Programs (CAC Nos. MF8202 Through MF8209)

ML17027A078
Person / Time
Site:	Saint Lucie, Point Beach, Seabrook, Turkey Point, Duane Arnold
Issue date:	04/07/2017
From:	Audrey Klett Plant Licensing Branch II
To:	Nazar M Nextera Energy
Klett A, NRR/DORL/LPL2-2, 301-415-0489
References
CAC MF8202, CAC MF8203, CAC MF8204, CAC MF8205, CAC MF8206, CAC MF8207, CAC MF8208, CAC MF8209
Download: ML17027A078 (129)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 7, 2017 Mr. Mano Nazar President and Chief Nuclear Officer Florida Power & Light Company NextEra Energy Duane Arnold, LLC NextEra Energy Point Beach, LLC NextEra Energy Seabrook, LLC Mail Stop NT3/JW 15430 Endeavor Drive Jupiter, FL 33478

SUBJECT:

DUANE ARNOLD ENERGY CENTER; POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2; SEABROOK STATION, UNIT NO. 1; ST. LUCIE PLANT, UNIT NOS. 1 AND 2; AND TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATIONS FOR INSERVICE TESTING (CAC NOS. MF8202 THROUGH MF8209)

Dear Mr. Nazar:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the following enclosed amendments: Amendment No. 300 to Renewed Facility Operating License No. DPR-49 for the Duane Arnold Energy Center; Amendment Nos. 259 and 263 to Renewed Facility Operating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2, respectively; Amendment No. 154 to Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1; Amendment Nos. 238 and 189 to Renewed Facility Operating License Nos. DPR-67 and NPF-16 for the St. Lucie Plant, Unit Nos. 1 and 2, respectively, and Amendment Nos. 274 and 269 to Renewed Facility Operating License Nos. DPR-31 and DPR-41 for the Turkey Point Nuclear Generating Unit Nos. 3 and 4, respectively.

In response to the application dated July 28, 2016 (L-2016-137), as supplemented by letter L-2016-219 dated December 15, 2016, from NextEra Energy Resources/Florida Power & Light Company, the amendments revise the Technical Specifications (TSs) consistent with Technical Specifications Task Force Traveler 545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing."

M. Nazar A copy of the Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Audrey L. Klett, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-250, 50-251, 50-266, 50-301, 50-331, 50-335, 50-389, and 50-443

Enclosures:

1. Amendment No. 300 to DPR-49

2. Amendment No. 259 to DPR-24

3. Amendment No. 263 to DPR-27

4. Amendment No. 154 to NPF-86

5. Amendment No. 238 to DPR-67

6. Amendment No. 189 to NPF-16

7. Amendment No. 274 to DPR-31

8. Amendment No. 269 to DPR-41

9. Safety Evaluation cc w/encl.: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY DUANE ARNOLD, LLC DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 300 Renewed License No. DPR-49

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

A. The application for amendment by NextEra Energy Duane Arnold, LLC dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-49 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC, shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Dateoflssuance: April 7, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 300 DUANE ARNOLD ENERGY CENTER RENEWED FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace page 3 of Renewed Facility Operating License No. DPR-49 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

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

Remove Insert 1.1-3 1.1-3 3.1-22 3.1-22 3.1-26 3.1-26 3.4-7 3.4-7 3.5-5 3.5-5 3.5-6 3.5-6 3.5-7 3.5-7 3.5-11 3.5-11 3.5-13 3.5-13 3.6-13 3.6-13 3.6-14 3.6-14 3.6-18 3.6-18 3.6-29 3.6-29 5.0-11 5.0-11

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level NextEra Energy Duane Arnold, LLC is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1912 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC I shall operate the facility in accordance with the Technical Specifications.

(a) For Surveillance Requirements (SRs) whose acceptance criteria are modified, either directly or indirectly, by the increase in authorized maximum power level in 2.C.(1) above, in accordance with Amendment No. 243 to Facility Operating License DPR-49, those SRs are not required to be performed until their next scheduled performance, which is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment No. 243.

(b) Deleted.

(3) Fire Protection Program NextEra Energy Duane Arnold, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated August 5, 2011 (and supplements dated October 14, 2011, April 23, 2012, May 23, 2012, July 9, 2012, October 15, 2012, January 11, 2013, February 12, 2013, March 6, 2013, May 1, 2013, May 29, 2013, two supplements dated July 2, 2013, and supplements dated August 5, 2013 and August 28, 2013) and as approved in the safety evaluation report dated September 10, 2013. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. DPR-49 Amendment No. 300

Definitions 1.1 1.1 Definitions (continued)

CORE OPERATING LIMITS The COLR is the unit specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle.

These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of DOSE EQUIVALENT 1-131 1-131 (microcuries/ml), that alone would produce the same dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989 and FGR 12, "External Exposure to Radionuclides in Air, Water, and Soil," 1993.

END OF CYCLE The EOC RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP time interval from initial signal generation by the TRIP (EOC RPT) SYSTEM associated turbine stop valve limit switch or from when RESPONSE TIME the turbine control valve hydraulic oil control oil pressure drops below the pressure switch setpoint to actuation of the breaker secondary (auxiliary) contact. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

(continued)

DAEC 1.1-3 Amendment 300

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate In accordance

~ 26.2 gpm at a discharge pressure ~ 1150 with the psig. IN SERVICE TESTING PROGRAM SR 3.1.7.7 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel. with the Surveillance Frequency Control Program SR 3.1.7.8 Verify all heat traced piping between storage In accordance tank and pump suction is unblocked. with the Surveillance Frequency Control Program AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1. 7-2 DAEC 3.1-22 Amendment 300

SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 ------------------------------N()TE---------------------------

Not required to be met on vent and drain valves closed during the performance of SR 3.1.8.2.

Verify each SDV vent and drain valve is In accordance open. with the Surveillance Frequency Control Program SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully In accordance closed and fully open position. with the INSERVICE TESTING PR()GRAM SR 3.1.8.3 Verify each SDV vent and drain valve: In accordance with the

a. Closes in:::; 30 seconds after receipt Surveillance of an actual or simulated scram Frequency signal; and Control Program

b. ()pens when the actual or simulated scram signal is reset.

DAEC 3.1-26 Amendment 300

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the SRVs In accordance and SVs are as follows: with the INSERVICE Number of Setpoint TESTING SRVs (~sig} PROGRAM 1 1110 +/- 33.0 1 1120 +/- 33.0 2 1130 +/- 33.0 2 1140 +/- 33.0 Number of Setpoint SVs (~sig}

2 1240 +/- 36.0 Following testing, lift settings shall be within +/-

1%.

SR 3.4.3.2 Verify each SRV actuator strokes when In accordance with manually actuated. the INSERVICE TESTING PROGRAM DAEC 3.4-7 Amendment 300

ECCS- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.2 -------------------------------NOTE-------------------------------------

The low pressure coolant injection (LPCI) system may be considered OPERABLE during alignment and operation for decay heat removal in MODE 3, if capable of being manually realigned and not otherwise inoperable.

---

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

Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS injection/spray subsystem power In accordance operated and automatic valve in the flow path, that is not with the locked, sealed, or otherwise secured in position, is in the Surveillance correct position. Frequency Control Program SR 3.5.1.3 Verify a 100 day supply of nitrogen exists for each ADS In accordance accumulator. with the Surveillance Frequency Control Program SR 3.5.1.4 Verify the following ECCS pumps develop the specified In accordance flow rate against a system head corresponding to the with the specified reactor pressure. INSERVICE TESTING PROGRAM SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray ~ 2718 gpm 1 ~ 113 psig LPCI ~ 4320 gpm 1 ~ 20 psiQ (continued)

DAEC 3.5-5 Amendment 300

ECCS- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.5 ----------------------------NOTE----------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1025 and ~ 940 In accordance psig, the HPCI pump can develop a flow rate with the

~ 2700 gpm against a system head INSERVICE corresponding to reactor pressure. TESTING PROGRAM SR 3.5.1.6 -----------------------------NOTE---------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressures 160 psig, the In accordance HPCI pump can develop a flow rate~ 2700 with the gpm against a system head corresponding to Surveillance reactor pressure. Frequency Control Program SR3.5.1.7 ----------------------------NOTES--------------------------

1. Vessel injection /spray may be excluded.

2. For the LPCI System, the Surveillance may be met by any series of sequential and/or overlapping steps, such that the LPCI Loop Select function is tested.

Verify each ECCS injection/spray subsystem In accordance actuates on an actual or simulated automatic with the initiation signal. Surveillance Frequency Control Program (continued)

DAEC 3.5-6 Amendment 300

ECCS- Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.8 -------------------------NOTE-------------------------------

Va Ive actuation may be excluded.

Verify the ADS actuates on an actual or In accordance simulated automatic initiation signal. with the Surveillance Frequency Control Program SR 3.5.1.9 Verify each ADS valve actuator strokes when In accordance manually actuated. with the INSERVICE TESTING PROGRAM DAEC 3.5-7 Amendment 300

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the specified In flow rate against a system head corresponding to the accordance specified reactor pressure.

with the INSERVICE SYSTEM HEAD TESTING NO. CORRESPONDING PROGRAM OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs 22718gpm 1 2 113 psig LPCI 2 4320 gpm 1 2 20 psig SR 3.5.2.6 -----------------------------------NOTES-------------------------------

1. Vessel injection/spray may be excluded.

2. For the LPCI System, the surveillance may be met by any series of sequential and/or overlapping steps, such that the LPCI Loop Select function is tested.

Verify each required ECCS subsystem actuates on an In accordance actual or simulated automatic initiation signal. with the Surveillance Frequency Control Program DAEC 3.5-11 Amendment 300

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System locations susceptible In accordance to gas accumulation are sufficiently filled with with the water. Surveillance Frequency Control Program SR 3.5.3.2 ----------------------------N()TE---------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each RCIC System power operated and In accordance automatic valve in the flow path, that is not with the locked, sealed, or otherwise secured in position, Surveillance is in the correct position. Frequency Control Program SR 3.5.3.3 -----------------------------N()TE---------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure ::::; 1025 psig and In accordance 2 940 psig, the RCIC pump can develop a flow with the rate 2 400 gpm against a system head INSERVICE corresponding to reactor pressure. TESTING PROGRAM SR 3.5.3.4 ---------------------------N()TE-----------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure::::; 160 psig, the In accordance RCIC pump can develop a flow rate 2 400 with the gpm against a system head corresponding to Surveillance reactor pressure. Frequency Control Program (continued)

DAEC 3.5-13 Amendment 300

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 -------------------------N()TE-----------------------------

Not required to be met when the 18 inch primary containment purge valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

Verify each 18 inch primary containment purge In accordance with va Ive is closed. the Surveillance Frequency Control Program SR 3.6.1.3.2 Verify continuity of the traversing incore In accordance with probe (TIP) shear isolation valve explosive the Surveillance charge. Frequency Control Program SR 3.6.1.3.3 Verify the isolation time of each power In accordance operated automatic PCIV, except for with the MSIVs, is within limits. INSERVICE TESTING PR()GRAM SR 3.6.1.3.4 Perform leakage rate testing for each primary In accordance with containment purge valve with resilient seals. the Surveillance Frequency Control Program

()nee within 92 days after opening the valve (continued)

DAEC 3.6-13 Amendment 300

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each MSIV is In accordance with

> 3 seconds and < 5 seconds. the INSERVICE TESTING PROGRAM SR 3.6.1.3.6 --------------------------NOTE----------------------------

For the MSIVs, this SR may be met by any series of sequential, overlapping, or total system steps, such that proper operation is verified.

Verify each automatic PCIV actuates to the In accordance with isolation position on an actual or simulated the Surveillance isolation signal. Frequency Control Program SR 3.6.1 .3. 7 Verify a representative sample of reactor In accordance with instrumentation line EFCVs actuate on a the Surveillance simulated instrument line break to restrict Frequency Control flow. Program SR 3.6.1.3.8 Remove and test the explosive squib from each In accordance with shear isolation valve of the TIP System. the INSERVICE TESTING PROGRAM (continued)

DAEC 3.6-14 Amendment 300

LLS Valves 3.6.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify each LLS valve actuator strokes when In accordance manually actuated. with the INSERVICE TESTING PROGRAM SR 3.6.1.5.2 --------------------------NOTE--------------------------

Valve actuation may be excluded.

Verify the LLS System actuates on an actual In accordance or simulated automatic initiation signal. with the Surveillance Frequency Control Program DAEC 3.6-18 Amendment 300

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify by administrative means each RHR In accordance suppression pool cooling subsystem with the manual, power operated and automatic Surveillance valve in the flow path that is not locked, Frequency sealed, or otherwise secured in position is in Control Program the correct position or can be aligned to the correct position.

SR 3.6.2.3.2 Verify each RHR pump develops a flow rate In accordance

~ 4800 gpm through the associated heat with the exchanger while operating in the suppression INSERVICE pool cooling mode. TESTING PROGRAM SR 3.6.2.3.3 Verify RHR suppression pool cooling In accordance subsystem locations susceptible to gas with the accumulation are sufficiently filled with water. Surveillance Frequency Control Program DAEC 3.6-29 Amendment 300

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 DELETED.

(continued)

DAEC 5.0-11 Amendment No. 300

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH. LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 259 Renewed License No. DPR-24

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

A. The application for amendment by NextEra Energy Point Beach, LLC (the licensee), dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 4.8 of the Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:

8. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 259, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~~

Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of issuance: April 7, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 263 Renewed License No. DPR-27

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

A. The application for amendment by NextEra Energy Point Beach, LLC (the licensee), dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 3

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 4.B of the Renewed Facility Operating License No. DPR-27 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 263, are hereby incorporated in the renewed operating license. NextEra Point Beach shall operate the facility in accordance with Technical Specifications.

3. This license amendment is effective as of the date of issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of issuance: April 7, 201 7

ATTACHMENT TO LICENSE AMENDMENT NOS. 259 AND 263 POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Replace page 3 of Renewed Facility Operating License No. DPR-24 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Replace page 3 of Renewed Facility Operating License No. DPR-27 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

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

Remove Insert 1.1-2 1.1-2 3.4.10-2 3.4.10-2 3.4.14-3 3.4.14-3 3.5.2-2 3.5.2-2 3.6.3-5 3.6.3-5 3.6.6-3 3.6.6-3 3.7.1-2 3.7.1-2 3.7.2-2 3.7.2-2 3.7.3-2 3.7.3-2 3.7.5-4 3.7.5-4 5.5-6 5.5-6

D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 259, are hereby incorporated in the renewed operating license.

NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-24 Amendment No. 259

C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed source for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.

4. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 263, are hereby incorporated in the renewed operating license.

NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.

C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.

Renewed License No. DPR-27 Amendment No. 263

Definitions 1.1 1.1 Definitions CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (COT) actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.4. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

DOSE EQUIVALENT DOSE EQUIVALENT Xe-133 shall be that concentration of Xe-133 Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body Deep Dose Equivalent as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity.

The determination of DOSE EQUIVALENT Xe-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" or the average gamma disintegration energies as provided in ICRP Publication 38, "Radionuclide Transformations," or similar source.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

Point Beach 1.1-2 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the INSERVICE with the TESTING PROGRAM. Following testing, lift INSERVICE settings shall be within .:t. 1%. TESTING PROGRAM Point Beach 3.4.10-2 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 --------------------------N()TES------------------------

1. Not required to be performed in M()DES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is within the In accordance limits contained in the RCS PIV Leakage with the Program. INSERVICE TESTING PR()GRAM, and 18 months Prior to entering M()DE 2 whenever the unit has been in M()DE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months continued)

Point Beach 3.4.14-3 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify ECCS locations susceptible to gas In accordance accumulation are sufficiently filled with water. with the Surveillance Frequency Control Program SR 3.5.2.3 Verify each ECCS pump's developed head at the In accordance test flow point is greater than or equal to the with the required developed head. INSERVICE TESTING PROGRAM SR 3.5.2.4 Verify each ECCS automatic valve in the flow In accordance path that is not locked, sealed, or otherwise with the secured in position, actuates to the correct Surveillance position on an actual or simulated actuation Frequency signal. Control Program SR 3.5.2.5 Verify each ECCS pump starts automatically on In accordance an actual or simulated actuation signal. with the Surveillance Frequency Control Program SR 3.5.2.6 Verify, by visual inspection, each ECCS train In accordance containment sump suction inlet is not restricted with the by debris and the suction inlet debris screens Surveillance show no evidence of structural distress or Frequency abnormal corrosion. Control Program Point Beach 3.5.2-2 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.3 ---------------------------N()TE--------------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve Prior to entering and blind flange that is located inside M()DE 4 from containment and not locked, sealed, or M()DE 5 if not otherwise secured and required to be closed performed within during accident conditions is closed, except for the previous containment isolation valves that are open under 92 days administrative controls.

SR 3.6.3.4 Verify the isolation time of each automatic power In accordance operated containment isolation valve is within with the INSERVICE TESTING PR()GRAM limits. INSERVICE TESTING PR()GRAM SR 3.6.3.5 Verify each automatic containment isolation In accordance valve that is not locked, sealed or otherwise with the secured in position, actuates to the isolation Surveillance position on an actual or simulated actuation Frequency signal. Control Program Point Beach 3.6.3-5 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.3 Verify each containment fan cooler unit can In accordance achieve a cooling water flow rate within design with the limits with a fan cooler service water outlet valve Surveillance Frequency open.

Control Program SR 3.6.6.4 Verify each containment spray pump's In accordance developed head at the flow test point is greater with the than or equal to the required developed head. INSERVICE TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment spray and In accordance containment fan cooler unit service water outlet with the valve in the flow path that is not locked, sealed, Surveillance or otherwise secured in position, actuates to the Frequency correct position on an actual or simulated Control Program actuation signal.

SR 3.6.6.6 Verify each containment spray pump starts In accordance automatically on an actual or simulated with the actuation signal. Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment fan cooler unit accident In accordance fan starts automatically on an actual or with the simulated actuation signal. Surveillance Frequency Control Program SR 3.6.6.8 Verify proper operation of the accident fan In accordance cooler unit backdraft dampers. with the Surveillance Frequency Control Program (continued)

Point Beach 3.6.6-3 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

MSSVs 3.7.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) -----------N 0 TE-----------

Only required in MODE 1.

B.2 Reduce the Power Range 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Neutron Flux - High reactor trip setpoint to less than or equal to the Maximum Allowable

% RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND OR C.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> One or more steam generators with three or more MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 ---------------------------NOTE-------------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per In accordance Table 3.7.1-2 in accordance with the with the INSERVICE TESTING PROGRAM. Following INSERVICE testing, lift setting shall be within +/-1 %. TESTING PROGRAM Point Beach 3.7.1-2 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

MSIVs and Non-Return Check Valves 3.7.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Verify MSIV and non- Once per 7 days return check valve in the affected flowpath a re closed and the MSIV is de-activated.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C not AND met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 --------------------------NOTE-------------------------

Only required to be performed in MODE 1.

Verify closure time of each MSIV is within limits. In accordance with the INSERVICE TESTING PROGRAM SR 3.7.2.2 --------------------------NOTE-------------------------

Only required to be performed in MODE 1.

In accordance Verify each MSIV actuates to the isolation with the position on an actual or simulated actuation Surveillance signal. Frequency Control Program SR 3.7.2.3 Verify each main steam non-return check valve In accordance can close. with the INSERVICE TESTING PROGRAM Point Beach 3.7.2-2 Unit 1 - Amendment No. 2591 Unit 2 - Amendment No. 263

MFIVs, MFRVs, and MFRV Bypass Valves 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two valves in the same D.1 Isolate affected flow 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flowpath inoperable. path E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify each MFIV, MFRV, and MFRV bypass In accordance valve, actuate to the isolation position on an with the actual or simulated actuation signal. Surveillance Frequency Control Program SR 3.7.3.2 Verify each MFIV, MFRV, and MFRV Bypass In accordance Valve isolation time is within limits. with the INSERVICE TESTING PROGRAM Point Beach 3.7.3-2 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 ---------------------------N()TE---------------------------

AFW pump system(s) may be considered

()PERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

In accordance with the Verify each AFW manual, power operated, and Surveillance automatic valve in each water flow path, and in Frequency both steam supply flow paths to the steam Control turbine driven pump, that is not locked, sealed, or Program otherwise secured in position, is in the correct position.

SR 3.7.5.2 ----------------------------N()TE---------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL P()WER exceeds 2% RTP.

Verify the developed head of each required AFW In accordance pump at the flow test point is greater than or with the equal to the required developed head. INSERVICE TESTING PR()GRAM SR 3.7.5.3 ----------------------------N()TE--------------------------

AFW pump system(s) may be considered

()PERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

In accordance with the Verify each AFW automatic valve that is not Surveillance locked, sealed, or otherwise secured in position, Frequency actuates to the correct position on an actual or Control simulated actuation signal. Program Point Beach 3.7.5-4 Unit 1 - Amendment No. 259 Unit 2 - Amendment No. 263

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 DELETED Point Beach 5.5-6 Unit 1 - Amendment No. 2591 Unit 2 - Amendment No. 263

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY SEABROOK, LLC, ET AL.*

DOCKET NO. 50-443 SEABROOK STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 154 License No. NPF-86

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

A. The application for amendment filed by NextEra Energy Seabrook, LLC, et al.

(the licensee), dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

• NextEra Energy Seabrook, LLC is authorized to act as agent for the: Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, and Taunton Municipal Light Plant and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

Enclosure 4

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-86 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 154, and the Environmental Protection Plan contained in Appendix B are incorporated into the Facility License No. NPF-86.

NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~2!&=-r-Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: April 7, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 154 SEABROOK STATION, UNIT NO. 1 FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace page 3 of Renewed Facility Operating License No. NPF-86 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

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

Remove Insert 1-4 1-4 3/4 0-3 3/4 0-3 3/4 0-4 3/4 0-4 3/4 4-8 3/4 4-8 3/4 4-9 3/4 4-9 3/4 4-12 3/4 4-12 3/4 4-17 3/4 4-17 3/4 4-28 3/4 4-28 3/4 5-6 3/4 5-6 3/4 6-14 3/4 6-14 3/4 6-17 3/4 6-17 3/4 7-1 3/4 7-1 3/4 7-9 3/4 7-9

(4) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein; and (7) DELETED C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level NextEra Energy Seabrook, LLC, is authorized to operate the facility at reactor core power levels not in excess of 3648 megawatts thermal (100% of rated power).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 154*, and the Environmental Protection Plan contained in Appendix B are incorporated into the Facility License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) License Transfer to FPL Energy Seabrook. LLC**

a. On the closing date(s) of the transfer of any ownership interests in Seabrook Station covered by the Order approving the transfer, FPL Energy Seabrook, LLC**, shall obtain from each respective transferring owner all of the accumulated decommissioning trust funds for the facility, and ensure the deposit of such funds and additional funds, if necessary, into a decommissioning trust or trusts for Seabrook Station established by FPL Energy Seabrook, LLC**, such that the amount of such funds deposited meets or exceeds the amount required under 10 CFR 50. 75 with respect to the interest in Seabrook Station FPL Energy Seabrook, LLC**,

acquires on such dates(s).

• Implemented

• • On April 16, 2009, the name "FPL Energy Seabrook, LLC" was changed to "NextEra Energy Seabrook, LLC".

AMENDMENT NO. 154

DEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary to secondary leakage).

INSERVICE TESTING PROGRAM 1.17A The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

MASTER RELAY TEST 1.18 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include, a continuity check of each associated slave relay.

MEMBER(S) OF THE PUBLIC 1.19 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.20 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6. 7.6 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.8.1.3 and 6.8.1.4.

OPERABLE - OPERABILITY 1.21 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s), and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.22 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

SEABROOK - UNIT 1 1-4 Amendment No. 7,-9,- 66, 81, 115, 147, 154

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.4 Entry into a MODE or other specified condition in the Applicability of a Limiting Condition for Operation (LCO) shall only be made when the LCO's Surveillances have been met within their specified frequency, except as provided by Specification 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with Specification 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

4.0.5 Surveillance Requirements for inservice inspection of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. lnservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g),

except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code including applicable Addenda for the inservice inspection activities required by the ASME Boiler and Pressure Vessel Code including applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Code Required frequencies for including applicable Addenda terminology performing service for inservice inspection activities Inspection activities.

Weekly At least once per 7 days Monthly At least once per 31 days Semi-quarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days SEABROOK - UNIT 1 3/4 0-3 Amendment No. 69, 114, 154

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.5 (Continued)

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection activities;

d. Performance of the above inservice inspection activities shall be in addition to other specified Surveillance Requirements; and

e. Deleted.

SEABROOK - UNIT 1 314 0-4 Amendment No. 444, 154

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting* of 2485 psig +/- 3%.**

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

• • Within +/-1 % following pressurizer Code safety valve testing.

SEABROOK - UNIT 1 3/4 4-8 Amendment No. 4-a, 154

REACTOR COOLANT SYSTEM SAFETY VALVES OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting* of 2485 psig +/- 3%.**

APPLICABILITY: MODES 1, 2, and 3 .

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least 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 />.

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

• • Within +/-1 % following pressurizer Code safety valve testing.

1. Entry into this MODE is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform post-modification or post-maintenance testing to verify OPERABILITY of components. ACTION requirements shall not apply until OPERABILITY has been verified.

SEABROOK - UNIT 1 314 4-9 Amendment No. +a, 154

REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of the INSERVICE TESTING PROGRAM, each PORV shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a. Performance of a CHANNEL CALIBRATION, and

b. Operating the valve through one complete cycle of full travel during MODES 3 or4.

4.4.4.2 Each block valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.

SEABROOK - UNIT 1 3/4 4-12 Amendment No. 16, 141, 154

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. In accordance with the Surveillance Frequency Control Program,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve, and

d. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve.*

e. Testing in accordance with the INSERVICE TESTING PROGRAM.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

• Not applicable to RHR Pumps BA and BB suction isolation valves.

SEABROOK - UNIT 1 3/4 4-17 Amendment No. 44, 69, 115, 141, 154

REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 ACTION: (Continued) f) With more than one charging pump capable of injecting into the RCS, immediately initiate action to restore a maximum of one charging pump capable of injecting into the RCS.

SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE when the PORV(s) are being used for overpressure protection by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, in accordance with the Surveillance Frequency Control Program thereafter when the PORV is required OPERABLE; and

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel in accordance with the Surveillance Frequency Control Program; and

c. Verifying the PORV isolation valve is open in accordance with the Surveillance Frequency Control Program.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valve(s) are being used for overpressure protection as follows:

a. For RHR suction relief valve RC-V89 by verifying in accordance with the Surveillance Frequency Control Program that RHR suction isolation valves RC-V87 and RC-V88 are open.

b. For RHR suction relief valve RC-V24 by verifying in accordance with the Surveillance Frequency Control Program that RHR suction isolation valves RC-V22 and RC-V23 are open.

c. Testing in accordance with the INSERVICE TESTING PROGRAM.

4.4.9.3.3 The RCS vent(s) shall be verified to be open in accordance with the Surveillance Frequency Control Program** when the vent(s) is being used for overpressure protection.

• • Except when the vent pathway is provided with a valve(s) or device(s) that is locked, sealed, or otherwise secured in the open position, then verify this valve(s) or device(s) open in accordance with the Surveillance Frequency Control Program.

SEABROOK-UNIT 1 3/44-28 Amendment No. 3, 5, 16, 74, 115, 116, 141, 154

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS -Tavg GREATER THAN OR EQUAL TO 350°F SURVEILLANCE REQUIREMENTS 4.5.2 (Continued)

d. In accordance with the Surveillance Frequency Control Program by:

1) Verifying automatic interlock action of the RHR system from the Reactor Coolant System to ensure that with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 440 psig, the interlocks prevent the valves from being opened.

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.

f. By verifying OPERABILITY of each pump when tested in accordance with the INSERVICE TESTING PROGRAM:

1) Centrifugal charging pump;

2) Safety Injection pump; and

3) RHR pump.

SEABROOK - UNIT 1 3/4 5-6 Amendment No. 33, 74, 83, 141, 154

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST* and automatically transferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 />; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position**, and

2) Verifying Containment Spray locations susceptible to gas accumulation are sufficiently filled with water.

b. By verifying OPERABILITY of each pump when tested in accordance with the INSERVICE TESTING PROGRAM;

c. In accordance with the Surveillance Frequency Control Program during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal, and

2) Verifying that each spray pump starts automatically on a Containment Pressure-Hi-3 test signal.

d. By verifying each spray nozzle is unobstructed following activities that could result in nozzle blockage.

• In MODE 4, when the Residual Heat Removal System is in operation, an OPERABLE flow path is one that is capable of taking suction from the refueling water storage tank upon being manually realigned.

• • Not required to be met for system vent flow paths opened under administrative control.

SEABROOK - UNIT 1 3/4 6-14 Amendment No. 30, 90, 128, 141, 144, 154

CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS

c. Verifying that on a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic containment isolation valve shall be determined to be within its limit when tested in accordance with the INSERVICE TESTING PROGRAM.

SEABROOK - UNIT 1 3/4 6-17 Amendment No. ~. 154

~4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION

3. 7 .1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY: MODES 1, 2, and 3 .

ACTION:

With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed, provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS

4. 7.1.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

1. Entry into this MODE is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform post-modification or post-maintenance testing to verify OPERABILITY of components. ACTION requirements shall not apply until OPERABILITY has been verified.

SEABROOK - UNIT 1 3/4 7-1 Amendment No. 154

PLANT SYSTEMS TURBINE CYCLE MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION

3. 7 .1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 .

ACTION:

MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise 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 />.

MODES 2 and 3:

With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, 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 />.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5.0 seconds when tested in accordance with the INSERVICE TESTING PROGRAM.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

1. Entry into this MODE is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform post-modification or post-maintenance testing to verify OPERABILITY of components. ACTION requirements shall not apply until OPERABILITY has been verified.

SEABROOK - UNIT 1 3/4 7-9 Amendment No. 154

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-335 ST. LUCIE PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 238 Renewed License No. DPR-67

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

A. The application for amendment by Florida Power & Light Company (FPL, the licensee), dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 5

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-67 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 238, are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: - Apr i 1 7, 2o1 7

ATTACHMENT TO LICENSE AMENDMENT NO. 238 ST. LUCIE PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 Replace page 3 of Renewed Facility Operating License No. DPR-67 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

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

Remove Insert I [index] I [index]

1-4 1-4 3/4 1-12 3/4 1-12 3/4 1-13 314 1-13 3/4 1-14 3/4 1-14 3/4 1-15 3/4 1-15 3/4 4-3 3/4 4-3 3/4 5-5 3/4 5-5 3/4 6-15a 3/4 6-15a 3/4 6-19 3/4 6-19 3/4 6-26 3/4 6-26 3/4 7-1 3/4 7-1 3/4 7-5 3/4 7-5 3/4 7-9 3/4 7-9 6-15c 6-15c

applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 3020 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 238, are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

Appendix 8, the Environmental Protection Plan (Non-Radiological), contains environmental conditions of the renewed license. If significant detrimental effects or evidence of irreversible damage are detected by the monitoring programs required by Appendix B of this license, FPL will provide the Commission with an analysis of the problem and plan of action to be taken subject to Commission approval to eliminate or significantly reduce the detrimental effects or damage.

C. Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on March 28, 2003, describes certain future activities to be completed before the period of extended operation. FPL shall complete these activities no later than March 1, 2016, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on March 28, 2003, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed license. Until that update is complete, FPL may make changes to the programs described in such supplement without prior Commission approval, provided that FPL evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

D. Sustained Core Uncovery Actions Procedural guidance shall be in place to instruct operators to implement actions that are designed to mitigate a small-break loss-of-coolant accident prior to a calculated time of sustained core uncovery.

Renewed License No. DPR-67 Amendment No. 238

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS 1.1 Action ............................................................................................................................ 1-1 1.2 Axial Shape Index ......................................................................................................... 1-1 1.3 Azimuthal Power Tilt ..................................................................................................... 1-1 1.4 Channel Calibration ...................................................................................................... 1-1 1.5 Channel Check ............................................................................................................. 1-1 1.6 Channel Functional Test ............................................................................................... 1-2 1.7 Containment Vessel Integrity ........................................................................................ 1-2 1.8 Controlled Leakage ....................................................................................................... 1-2 1.9 Core Alteration .............................................................................................................. 1-2 1.9a Core Operating Limits Report (COLR) .......................................................................... 1-2 1.10 Dose Equivalent 1-131 ................................................................................................... 1-3 1.11 Dose Equivalent Xe-133 ............................................................................................... 1-3 1.12 Engineered Safety Features Response Time ............................................................... 1-3 1.13 Frequency Notation ...................................................................................................... 1-3 1.14 Gaseous Radwaste Treatment System ........................................................................ 1-3 1.15 Identified Leakage ........................................................................................................ 1-4 1.16 Inservice Testing Program ............................................................................................ 1-4 1.17 Member(s) of the Public ................................................................................................ 1-4 1.18 Offsite Dose Calculation Manual (ODCM) .................................................................... 1-4 1.19 Operable - Operability .................................................................................................. 1-5 1.20 Operational Mode - Mode ............................................................................................ 1-5 1.21 Physics Tests ................................................................................................................ 1-5 1.22 Pressure Boundary Leakage ........................................................................................ 1-5 ST. LUCIE - UNIT 1 Amendment No. U, ~. ea, 09,

-%G.~. 238

DEFINITIONS IDENTIFED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the secondary system (Primary-to-secondary leakage).

INSERVICE TESTING PROGRAM 1.16 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

MEMBER(S) OF THE PUBLIC 1.17 MEMBER OF THE PUBLIC means an individual in a controlled or unrestricted area.

However, an individual is not a member of the public during any period in which the individual receives an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 THE OFFS ITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

ST. LUCIE - UNIT 1 1-4 Amendment No. W, W, 99, &+, 4-04, ~.

4-2a,200.~.238

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or high pressure safety injection pump* in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no charging pump or high pressure safety injection pump* OPERABLE, suspend all operations involving CORE AL TE RATIONS or positive reactivity changes** until at least one of the required pumps is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 At least one of the above required pumps shall be demonstrated OPERABLE by verifying the charging pump develops a flow rate of greater than or equal to 40 gpm or the high pressure safety injection pump develops a total head of greater than or equal to 2571 ft.

when tested pursuant to the INSERVICE TESTING PROGRAM.

• The flow path from the RWT to the RCS via a single HPSI pump shall be established only if:

(a) the RCS pressure boundary does not exist, or (b) RCS pressure boundary integrity exists and no charging pumps are operable. In the latter case, all charging pumps shall be disabled.

• • Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

ST. LUCIE - UNIT 1 3/41-12 Amendment No. W, &+, 00, -W4, ++G, 441-,~.4-e3,4+9,~.238

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERA TING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />; restore at least two charging pumps to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying that each pump develops a flow rate or greater than or equal to 40 gpm when tested pursuant to the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 1 3/41-13 Amendment No. 00, ~. 238

REACTIVITY CONTROL SYSTEMS BORIC ACID PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 At least one boric acid pump shall be OPERABLE if only the flow path through the boric acid pump in Specification 3.1.2.1 a above, is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no boric acid pump OPERABLE as required to complete the flow path of Specification 3.1.2.1 a, suspend all operations involving CORE AL TERA TIONS or positive reactivity changes* until at least one boric acid pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required boric acid pump shall be demonstrated OPERABLE by verifying that the pump develops the specified discharge pressure when tested pursuant to the INSERVICE TESTING PROGRAM.

• Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

ST. LUCIE - UNIT 1 3/4 1-14 Amendment No. W, ~. ~.

4-94,238

REACTIVITY CONTROL SYSTEMS BORIC ACID PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3.1.2.2a shall be OPERABLE if the flow path through the boric acid pump in Specification 3.1.2.2a is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one boric acid pump required for boron injection flow path(s) pursuant to Specification 3.1.2.2a inoperable, restore the boric acid pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.6 The above required boric acid pump(s) shall be demonstrated OPERABLE by verifying that the pump(s) develop the specified discharge pressure when tested pursuant to the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 1 3/4 1-15 Amendment No. 00, +ea, 4-94, 238

REACTOR COOLANT SYSTEM SAFETY VALVES -OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of

~ 2422.8 psig and~ 2560.3 psig.

APPLICABILITY: MODES 1, 2, 3, and 4 with all RCS cold leg temperatures > 281°F.

ACTION:

a. With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT STANDBY within 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 next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With two or more pressurizer code safety valves inoperable, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN with all RCS cold leg temperatures~ 281°F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.3 Verify each pressurizer code safety valves is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, as-left lift settings shall be within +/- 1% of 2500 psia.

ST. LUCIE - UNIT 1 3/4 4-3 Amendment No. 00, ~. 4W, 238

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

e. In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1. Verifying that each automatic valve in the flow paths actuates to its correct position on a Safety Injection Actuation Signal.

2. Verifying that each of the following pumps start automatically upon receipt of a Safety Injection Actuation Signal;

a. High-Pressure Safety Injection Pumps.

b. Low-Pressure Safety Injection Pumps.

c. Charging Pumps.

3. Verifying that upon receipt of an actual or simulated Recirculation Actuation Signal: each low-pressure safety injection pump stops, each containment sump isolation valve opens, each refueling water tank outlet valve closes, and each safety injection system recirculation valve to the refueling water tank closes.

f. By verifying that each of the following pumps develops the specified total developed head when tested pursuant to the INSERVICE TESTING PROGRAM.

1. High-Pressure Safety Injection pumps.

2. Low-Pressure Safety Injection pumps.

ST. LUCIE - UNIT 1 3/4 5-5 Amendment No. 2:6, 00, ~. 4-W, 4-e4,

+94.~.~.238

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is positioned to take suction from the RWT on a Containment Pressure -- High High test signal.*

b. By verifying that each spray pump develops the specified discharge pressure when tested pursuant to the INSERVICE TESTING PROGRAM.

c. In accordance with the Surveillance Frequency Control Program by verifying containment spray system locations susceptible to gas accumulation are sufficiently filled with water.

• Not required to be met for system vent flow paths opened under administrative control.

ST. LUCIE - UNIT 1 3/4 6-15a Amendment No. ~. ~. 4-94,

~.~. 238

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued) 4.6.3.1.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:

a. Verifying that on a Containment Isolation test signal, and/or SIAS test signal, each isolation valve actuates to its isolation position.

4.6.3.1.3 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 1 3/4 6-19 Amendment No. W, 449, ~.

~.238

CONTAINMENT SYSTEMS 3/4.6.5 VACUUM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.6.5.1 Two vacuum relief lines shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one vacuum relief line inoperable, restore the vacuum relief line to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.5.1 Verify each vacuum relief line OPERABLE in accordance with the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 1 3/4 6-26 Amendment No. W, ~. ~. ~. 238

3/4.7 PLANT SYSTEMS 3.4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE with lift settings as specified in Table 4.7-1.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each main steam line code safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, as-left lift settings shall be within

+/- 1% of 1000 psia for valves 8201 through 8208, and within +/- 1% of 1040 psia for valves 8209 through 8216 specified in Table 4.7-1.

ST. LUCIE - UNIT 1 3/4 7-1 Amendment No. 90, +eJ, 400, 22Q, 238

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS <Continued)

1. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program during shutdown by:

1. Verifying that each automatic valve in the flowpath actuates to its correct position upon receipt of the Auto Start actuation test signal.

2. Verifying that each auxiliary feedwater pump starts automatically as designed upon receipt of the Auto Start actuation test signal.

c. By verifying the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 when testing the steam turbine-driven AFW pump and this Surveillance must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 3 and prior to entering MODE 2.

ST LUCIE - UNIT 1 3/4 7-5 Amendment No. g+., 9Q, 99, ~. 238

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODE 1 With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, 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 />.

MODES 2 - With one or both main steam isolation valve(s) inoperable, subsequent operation in and 3 MODES 2 or 3 may proceed provided the isolation valve(s) is (are) maintained closed. Otherwise, be in at least 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 COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve that is open shall be demonstrated OPERABLE by verifying full closure within 6.0 seconds when tested pursuant to the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 1 3/4 7-9 Amendment No. QQ, ~. +ad,

rn), 238

ADMINISTRATIVE CONTROLS (continued)

The provisions of T.S. 4.0.2 do not apply to test frequencies in the Containment Leak Rate Testing Program.

The provisions of T.S. 4.0.3 are applicable to the Containment Leak Rate Testing Program.

i. Deleted

j. Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

2. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

a. a change in the TS incorporated in the license; or

b. a change to the updated UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

3. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

4. Proposed changes that meet the criteria of Specification 6.8.4.j.2.a or 6.8.4.j.2.b, above, shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

ST. LUCIE - UNIT 1 6-15c Amendment No. ~. ~. 4-+e, 238

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 189 Renewed License No. NPF-16

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

A. The application for amendment by Florida Power & Light Company, et al.

(FPL, the licensee}, dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 6

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. NPF-16 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 189, are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: April 7, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 189 ST. LUCIE PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace page 3 of Renewed Facility Operating License No. NPF-16 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

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

Remove Insert I [index] I [index]

1-4 1-4 3/4 1-9 3/4 1-9 314 1-10 314 1-10 314 1-11 314 1-11 314 1-12 3/4 1-12 314 4-8 314 4-8 314 4-36 3/4 4-36 3/4 5-5 3/4 5-5 314 6-15a 314 6-15a 314 6-20 3/4 6-20 314 6-26 314 6-26 314 7-1 314 7-1 314 7-5 314 7-5 3/4 7-9 314 7-9 3/4 7-10 3/4 7-10 6-15c 6-15c

neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

D. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission's regulations: 10 CFR Part 20, Section 30.34 of 10 FR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 3020 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 189, are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16 Amendment No. 189

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTION ........................................................................................................................ 1-1 1.2 AXIAL SHAPE INDEX .................................................................................................. 1-1 1.3 AZIMUTHAL POWER TILT ........................................................................................... 1-1 1.4 CHANNEL CALIBRATION ............................................................................................ 1-1 1.5 CHANNEL CHECK ....................................................................................................... 1-1 1.6 CHANNEL FUNCTIONAL TEST ................................................................................... 1-2 1.7 CONTAINMENT VESSEL INTEGRITY ........................................................................ 1-2 1.8 CONTROLLED LEAKAGE ............................................................................................ 1-2 1.9 CORE AL TERA Tl ON .................................................................................................... 1-2 1.9a CORE OPERATING LIMITS REPORT (COLR) ........................................................... 1-2 1.10 DOSE EQUIVALENT 1-131 ........................................................................................... 1-3 1.11 DOSE EQUIVALENT XE-133 ....................................................................................... 1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME ............................................ 1-3 1.13 FREQUENCY NOTATION ............................................................................................ 1-3 1.14 GASEOUS RADWASTE TREATMENT SYSTEM ........................................................ 1-3 1.15 IDENTIFIEDLEAKAGE ................................................................................................ 1-3 1.16 INSERVICE TESTING PROGRAM .............................................................................. 1-4 1.17 MEMBER(S) OF THE PUBLIC ..................................................................................... 1-4 1.18 OFF SITE DOSE CALCULATION MANUAL (ODCM) ................................................... 1-4 1.19 OPERABLE-OPERABILITY ....................................................................................... 1-4 1.20 OPERATIONAL MODE - MODE .................................................................................. 1-4 1.21 PHYSICS TESTS ......................................................................................................... 1-4 1.22 PRESSURE BOUNDARY LEAKAGE ........................................................................... 1-5 1.23 PROCESS CONTROL PROGRAM .............................................................................. 1-5 1.24 PURGE- PURGING .................................................................................................... 1-5 1.25 RATED THERMAL POWER ......................................................................................... 1-5 1.26 REACTOR TRIP SYSTEM RESPONSE TIME ............................................................. 1-5 1.27 REPORTABLE EVENT ................................................................................................. 1-5 1.28 SHIELD BUILDING INTEGRITY ................................................................................... 1-5 1.29 SHUTDOWN MARGIN ................................................................................................. 1-6 1.30 SITE BOUNDARY ......................................................................................................... 1-6 ST. LUCIE - UNIT 2 Amendment No. +tl, ~. +03, 189

DEFINITIONS INSERVICE TESTING PROGRAM 1.16 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

MEMBER($) OE THE PUBLIC 1.17 MEMBER OF THE PUBLIC means an individual in a controlled or unrestricted area.

However, an individual is not a member of the public during any period in which the individual receives an occupational dose.

OEESITE DOSE CALCULATION MANUAL (ODCM) 1.18 THE OFFS ITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1. 7 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

ST. LUCIE - UNIT 2 1-4 Amendment No. 4B, ;u, 4e, e+,

~.~. 189

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or high pressure safety injection pump in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no charging pump or high pressure safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes*.

SURVEILLANCE REQUIREMENTS 4.1.2.3 At least the above required pumps shall be demonstrated OPERABLE by verifying the charging pump develops a flow rate of greater than or equal to 40 gpm or the high pressure safety injection pump develops a total head of greater than or equal to 2854 ft.

when tested pursuant to the INSERVICE TESTING PROGRAM.

• Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

ST. LUCIE - UNIT 2 3/4 1-9 Amendment No. 94-, ~. 189

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERA TING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 200°F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying that each pump develops a flow rate of greater than or equal to 40 gpm when tested pursuant to the INSERVICE TESTING PROGRAM.

4.1.2.4.2 In accordance with the Surveillance Frequency Control Program verify that each charging pump starts automatically on an SIAS test signal.

ST. LUCIE - UNIT 2 3/4 1-10 Amendment No. 8, ~. 94-, -We,

~. 189

REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 At least one boric acid makeup pump shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if only the flow path through the boric acid pump in Specification 3.1.2.1 a is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no boric acid pump OPERABLE as required to complete the flow path of Specification 3.1.2.1 a, suspend all operations involving CORE AL TERA TIONS or positive reactivity changes*.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required boric acid makeup pump shall be demonstrated OPERABLE by verifying that the pump develops the specified discharge pressure when tested pursuant to the INSERVICE TESTING PROGRAM.

• Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

ST. LUCIE - UNIT 2 3/41-11 Amendment No. 9-i, ~. ~. 189

REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid makeup pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3.1.2.2 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump(s) in Specification 3.1.2.2 is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With no boric acid makeup pump required for the boron injection flow path(s) pursuant to Specification 3.1.2.2 operable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 200°F; restore the above required boric acid makeup pump(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.6 The above required boric acid makeup pump(s) shall be demonstrated OPERABLE by verifying that the pump(s) develop the specified discharge pressure when tested pursuant to the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 2 314 1-12 Amendment No. 8, 2a, 4Q, 9+, ~.

~.+w, 189

REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer code safety valves shall be OPERABLE with a lift setting of

~ 2410.3 psig and~ 2560.3 psig.*

APPLICABILITY: MODES 1, 2, 3, and 4 with all RCS cold leg temperatures> 230°F.

ACTION:

a. With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT STANDBY within 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 next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With two or more pressurizer code safety valves inoperable, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN with all RCS cold leg temperatures at~ 230°F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2.2 Verify each pressurizer code safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, as-left lift settings shall be within +/- 1% of 2500 psia.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

ST. LUCIE - UNIT 2 3/4 4-8 Amendment No. 94-, 4-W, ~. 189

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION ACTION: (Continued)

c. In the event either the PORVs, SDCRVs or the RCS vent(s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, SDCRVs or vent(s) on the transient and any corrective action necessary to prevent recurrence.

d. LCO 3.0.4.b is not applicable to PORVs when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. In addition to the requirements of the INSERVICE TESTING PROGRAM, operating the PORV through one complete cycle of full travel in accordance with the Surveillance Frequency Control Program.

ST. LUCIE - UNIT 2 3/4 4-36 Amendment No. ~. 34-, 9+, 4-+G,

~. 189

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

3. Verifying that a minimum total of 173 cubic feet of solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

4. Verifying that when a representative sample of 70.5 .:!:. 0.5 grams of TSP from a TSP storage basket is submerged, without agitation, in 10.0 .:!:. 0.1 gallons of 120 .:!:. 10°F borated water from the RWT, the pH of the mixed solution is raised to greater than or equal to 7 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

f. In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1. Verifying that each automatic valve in the flow paths actuates to its correct position on SIAS and/or RAS test signals.

2. Verifying that each of the following pumps start automatically upon receipt of a Safety Injection Actuation Test Signal:

a. High-Pressure Safety Injection pumps.

b. Low-Pressure Safety Injection pumps.

c. Charging Pumps

3. Verifying that upon receipt of an actual or simulated Recirculation Actuation Signal: each low-pressure safety injection pump stops, each containment sump isolation valve opens, each refueling water tank outlet valve closes, and each safety injection system recirculation valve to the refueling water tank closes.

g. By verifying that each of the following pumps develops the specified total developed head when tested pursuant to the INSERVICE TESTING PROGRAM:

1. High-Pressure Safety Injection pumps.

2. Low-Pressure Safety Injection pumps.

h. By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:

1. During valve stroking operation or following maintenance on the valve and prior to declaring the valve OPERABLE when the ECCS subsystems are required to be OPERABLE.

ST. LUCIE - UNIT 2 3/4 5-5 Amendment No.~. 99, 400,

~.~.~. 189

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is positioned to take suction from the RWT on a Containment Pressure - - High-High test signal.*

b. By verifying that each spray pump develops the specified discharge pressure when tested pursuant to the INSERVICE TESTING PROGRAM.

c. In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on a CSAS test signal.

2. Verifying that upon a Recirculation Actuation Test Signal (RAS), the containment sump isolation valves open and that a recirculation mode flow path via an OPERABLE shutdown cooling heat exchanger is established.

• Not required to be met for system vent flow paths opened under administrative control ST. LUCIE - UNIT 2 314 6-15a Amendment No. 7G, 9+, ~.

~. 4+4, 189

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each automatic containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:

a. Verifying that on a Containment Isolation test signal (CIAS) and/or a Safety Injection test signal (SIAS), each isolation valve actuates to its isolation position.

b. Verifying that on a Containment Radiation-High test signal, each containment purge valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to the INSERVICE TESTING PROGRAM.

ST LUCIE - UNIT 2 3/4 6-20 Amendment No. M, 94-, ~. 189

CONTAINMENT SYSTEMS 3/4.6.5 VACUUM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.6.5 Two vacuum relief lines shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one vacuum relief line inoperable, restore the vacuum relief line to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.5 Verify each vacuum relief line OPERABLE in accordance with the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 2 3/4 6-26 Amendment No. W, 89, 94-, +2a, 189

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE with lift settings as shown in Table 3.7-2.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided that, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each main steam line code safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, as-left lift settings shall be within

+/- 1% of 1000 psia for valves 8201 through 8208, and within +/- 1% of 1040 psia for valves 8209 through 8216 specified in Table 3.7-2.

ST LUCIE - UNIT 2 3/4 7-1 Amendment No. 8, 08, 94, 4--1-0,

+:7-0, 189

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS <Continued)

b. In accordance with the Surveillance Frequency Control Program during shutdown by:

1. Verifying that each automatic valve in the flowpath path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal.

2. Verifying that each pump starts automatically upon receipt of an auxiliary feedwater actuation test signal.

c. Following an extended cold shutdown (30 days or longer) and prior to entering MODE 2, a flow test shall be performed to verify the normal flow path from the condensate storage tank (CST) to the steam generators.

d. By verifying the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 when testing the steam turbine-driven AFW pump and this Surveillance must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 3 and prior to entering MODE 2.

ST. LUCIE - UNIT 2 3/4 7-5 Amendment No. -Me, ~. 189

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

MODE1 - With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, be in at least 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 />.

MODES 2, 3 - With one or both main steam isolation valve(s) inoperable, subsequent and 4 operation in MODES 2, 3 or 4 may proceed provided the isolation valve(s) is (are) maintained closed. Otherwise, be in at least 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 COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 6.75 seconds when tested pursuant to the INSERVICE TESTING PROGRAM.

ST. LUCIE - UNIT 2 3/4 7-9 Amendment No. ~. a2, 9+, 4-+Q, 189

PLANT SYSTEMS MAIN FEEDWATER ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 Four main feedwater isolation valves (MFIVs) shall be OPERABLE.

APPLICABILITY:* MODES 1, 2 and 3, except when the MFIV is closed and deactivated.

ACTION:

a. With one MFIV inoperable in one or more main feedwater lines, OPERATION may continue provided each inoperable valve is restored to OPERABLE status, closed, or isolated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Otherwise, 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 two MFIVs inoperable in the same flowpath, restore at least one of the inoperable MFIVs to OPERABLE status or close one of the inoperable valves within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise, 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 />.

SURVEILLANCE REQUIREMENTS 4.7.1.6.a Each MFIV shall be demonstrated OPERABLE by verifying full closure within 5.15 seconds when tested pursuant to the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

4.7.1.6.b For each inoperable MFIV, verify that it is closed or isolated once per 7 days.

• Each MFIV shall be treated independently.

ST. LUCIE - UNIT 2 3/4 7-10 Amendment No. 8, ~. ~. +-+, 94-, 189

ADMINISTRATIVE CONTROLS <Continued)

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is~ 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are< 0.60 La for the Type Band C tests,~ 0.75 La for Type A tests, and ~ 0.096 La for secondary containment bypass leakage paths.

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage is~ 0.05 La when tested at~ Pa.

2) For each door seal, leakage rate is< 0.01 La when pressurized to~ Pa.

The provisions of T.S. 4.0.2 do not apply to test frequencies in the Containment Leak Rate Testing Program.

The provisions for T.S. 4.0.3 are applicable to the Containment Leak Rate Testing Program.

i. Deleted ST. LUCIE - UNIT 2 6-15c Amendment No. 88, 94-, ~. 189

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 27 4 Renewed License No. DPR-31

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

A. The application for amendment by Florida Power & Light Company (the licensee) dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 7

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-31 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 274, are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: April 7, 201 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT NUCLEAR GENERATING UNIT NO. 4 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 Renewed License No. DPR-41

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

A. The application for amendment by Florida Power & Light Company (the licensee) dated July 28, 2016, as supplemented by letter dated December 15, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

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

Enclosure 8

2. Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-41 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 269, are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULA TORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: April 7, 2017

ATTACHMENT TO LICENSE AMENDMENT NOS. 274 AND 269 TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 RENEWED FACILITY OPERATING LICENSE NOS. DPR-31 AND DPR-41 DOCKET NOS. 50-250 AND 50-251 Replace page 3 of Renewed Facility Operating License No. DPR-31 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Replace page 3 of Renewed Facility Operating License No. DPR-41 with the attached page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

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

Remove Insert i [index] i [index]

1-3 1-3 314 0-3 314 0-3 314 0-4 314 0-4 314 1-10 314 1-10 314 4-7 314 4-7 314 4-8 314 4-8 314 5-5 314 5-5 314 6-12 314 6-12 314 6-17 314 6-17 314 7-1 314 7-1 314 7-10 314 7-10 314 7-13 314 7-13

3 E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:

A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 274, are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.

The licensee shall complete these activities no later than July 19, 2012.

The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Unit 3 Renewed License No. DPR-31 Amendment No. 274

3 E. Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.

3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:

A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 269, are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.

The licensee shall complete these activities no later than April 10, 2013.

The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Unit4 Renewed License No. DPR-41 Amendment No. 269

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS.................................................................................................................... 1.0 1.1 ACTION......................................................................................................................... 1-1 1.2 ACTUATION LOGIC TEST........................................................................................... 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST................................................................ 1-1 1.4 AXIAL FLUX DIFFERENCE.......................................................................................... 1-1 1.5 CHANNEL CALIBRATION............................................................................................ 1-1 1.6 CHANNEL CHECK....................................................................................................... 1-1 1.7 CONTAINMENT INTEGRITY........................................................................................ 1-2 1.8 CONTROLLED LEAKAGE............................................................................................ 1-2 1.9 CORE ALTERATIONS.................................................................................................. 1-2 1.10 CORE OPERATING LIMITS REPORT......................................................................... 1-2 1.11 DIGITAL CHANNEL OPERATIONAL TEST................................................................. 1-2 1.12 DOSE EQUIVALENT 1-131 ........................................................................................... 1-3 1.13 DOSEEQUIVALENTXE-133 ....................................................................................... 1-3 1.14 FREQUENCY NOTATION .. .... .... ........ .... ..... ........ .... .... .... ..... ...... ...... ...... ......... .. ........... 1-3 1.15 GAS DECAY TANK SYSTEM....................................................................................... 1-3 1.16 IDENTIFIED LEAKAGE................................................................................................ 1-3 1.16A INSERVICE TESTING PROGRAM............................................................................... 1-3 1.17 OPERABLE - OPERABILITY....................................................................................... 1-4 1.18 OPERATIONAL MODE - MODE.................................................................................. 1-4 1.19 PHYSICS TESTS .......................................................................................................... 1-4 1.20 PRESSURE BOUNDARY LEAKAGE........................................................................... 1-4 1.21 PURGE - PURGING..................................................................................................... 1-4 1.22 QUADRANT POWER TILT RATIO ............................................................................... 1-5 1.23 RATED THERMAL POWER......................................................................................... 1-5 1.24 SHUTDOWN MARGIN.................................................................................................. 1-5 1.25 SITE BOUNDARY......................................................................................................... 1-5 TURKEY POINT - UNITS 3 & 4 AMENDMENT NOS. 274 AND 269

DEFINITIONS DOSE EQUIVALENT 1-131 1.12 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

DOSE EQUIVALENT XE -133 1.13 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131 m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water. and Soil."

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GAS DECAY TANK SYSTEM 1.15 A GAS DECAY TANK SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System (primary-to-secondary leakage).

INSERVICE TESTING PROGRAM 1.16A The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

TURKEY POINT - UNITS 3 & 4 1-3 AMENDMENT NOS. 274 AND 269

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. If an ACTION item requires periodic performance on a "once per ... " basis, the above frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications.

4.0.3 If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition of Operation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the surveillance is not performed within the delay period, the Limiting Condition of Operation must immediately be declared not met, and the applicable ACTION(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition of Operation must immediately be declared not met, and the applicable ACTION(s) must be entered.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement(s) associated with a Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

4.0.5 Surveillance Requirements for inservice inspection of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. lnservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a.

TURKEY POINT - UNITS 3 & 4 3/4 0-3 AMENDMENT NOS. 274 AND 269

APPLICABILITY SURVEILLANCE REQUIREMENTS <CONTINUEDl

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Code and applicable Addenda Required frequencies for terminology for inservice inspection performing inservice inspection activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection activities.

d. Performance of the above inservice inspection activities shall be in addition to other specified Surveillance Requirements.

e. DELETED

f. Each reactor coolant pump flywheel shall be inspected at least once every 20 years, by either conducting an in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or conduct a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the disassembled flywheel.

4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifications or whenever certain portions of a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes or body of the requirement.

TURKEY POINT - UNITS 3 & 4 3/4 0-4 AMENDMENT NOS. 274 AND 269

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.3 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> or be in at least HOT STANDBY and borated to a boron concentration equivalent to at least the required SHUTDOWN MARGIN at COLD SHUTDOWN at 200°F within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUI REMENTS 4.1.2.3.1 The required charging pumps shall be demonstrated OPERABLE by testing in accordance with the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 and 4.

TURKEY POINT - UNITS 3 & 4 3/4 1-10 AMENDMENT NOS. 274 AND 269

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE* with a lift setting of 2465 psig + 2%, -3%.** ***

APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

• While in MODE 5, an equivalent size vent pathway may be used provided that the vent pathway is not isolated or sealed.

• • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

• • • All valves tested must have "as left" lift setpoints that are within +/- 1% of the lift setting value.

TURKEY POINT - UNITS 3 & 4 3/4 4-7 AMENDMENT NOS. 274 AND 269

REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2465 psig + 2%, -3%.* **

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least 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 />.

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by the INSERVICE TESTING PROGRAM.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

• • All valves tested must have "as left" lift setpoints that are within +/- 1% of the lift setting value.

TURKEY POINT - UNITS 3 & 4 3/4 4-8 AMENDMENT NOS. 274 AND 269

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS component and flow path shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying by control room indication that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position 864A and B Supply from RWST to ECCS Open 862A and B RWST Supply to RHR pumps Open 863A and B RHR Recirculation Closed 866A and B H.H.S.I. to Hot Legs Closed HCV-758* RHR HX Outlet Open To permit positive valve position indication for surveillance or maintenance purposes in the event that continuous valve position indication is unavailable in the control room, power may be restored to these valves for a period not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b. In accordance with the Surveillance Frequency Control Program by:

1) Verifying ECCS locations susceptible to gas accumulation are sufficiently filled with water, and

2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.**

c. By verifying that each SI and RHR pump develops the indicated differential pressure applicable to the operating conditions when tested in accordance with the INSERVICE TESTING PROGRAM.:

1) SI pump 2 1083 psid at a metered flowrate 2 300 gpm (normal alignment or Unit 4 SI pumps aligned to Unit 3 RWST}, or 2 1113 psid at a metered flowrate 2 280 gpm (Unit 3 SI pumps aligned to Unit 4 RWST).

2) RHR pump Develops the indicated differential pressure applicable to the operating conditions in accordance with Figure 3.5-1.

• Air Supply to HCV-758 shall be verified shut off and sealed closed in accordance with the Surveillance Frequency Control Program.

• • Not required to be met for system vent flow paths opened under administrative control.

TURKEY POINT - UNITS 3 & 4 3/4 5-5 AMENDMENT NOS. 274 AND 269

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one Containment Spray System inoperable restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With two Containment Spray Systems inoperable restore at least one Spray System to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore both Spray Systems to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of initial loss or be in at least 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position* and that power is available to flow path components that require power for operation;

b. By verifying that on recirculation flow, each pump develops the indicated differential pressure, when tested in accordance with the INSERVICE TESTING PROGRAM.

Containment Spray Pump ~241.6 psid while aligned in recirculation mode.

c. In accordance with the Surveillance Frequency Control Program by verifying containment spray locations susceptible to gas accumulation are sufficiently filled with water.

• Not required to be met for system vent flow paths opened under administrative control.

TURKEY POINT - UNITS 3 & 4 3/4 6-12 AMENDMENT NOS. 274 AND 269

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS <Continued) 4.6.4.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B isolation valve actuates to its isolation position; and

c. Verifying that on a Containment Ventilation Isolation test signal, each purge, exhaust and instrument air bleed valve actuates to its isolation position.

4.6.4.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested in accordance with the INSERVICE TESTING PROGRAM.

3/4.6.5 DELETED 3/4.6.6 DELETED TURKEY POINT - UNITS 3 & 4 3/4 6-17 AMENDMENT NOS. 274 AND 269

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With (3) reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, and

a. in MODES 1 and 2, with a positive Moderator Temperature Coefficient, operation may continue provided that, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve(s) are restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced to the maximum allowable percent of RATED THERMAL POWER listed In Table 3.7-1; otherwise, be in at least 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or

b. in MODES 1 and 2, with a negative or zero Moderator Temperature Coefficient; or in Mode 3, with a positive, negative or zero Moderator Temperature Coefficient, operation may continue provided that, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve(s) are restored to OPERABLE status or reactor power is reduced to less than or equal to the maximum allowable percent of RATED THERMAL POWER listed in Table 3.7-1; otherwise, be in at least 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

TURKEY POINT - UNITS 3 & 4 3/4 7-1 AMENDMENT NOS. 274 AND 269

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; otherwise 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 />.

MODES 2 and 3:

With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, 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 />.

SURVEI LLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested in accordance with the INSERVICE TESTING PROGRAM. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

TURKEY POINT - UNITS 3 & 4 3/4 7-10 AMENDMENT NOS. 274 AND 269

PLANT SYSTEMS 3/4.7.1.7 FEEDWATER ISOLATION LIMITING CONDITION FOR OPERATION 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 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 paths 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 /> ..

SURVEILLANCE REQUIREMENTS 4.7.1.7 Each FCV, FIV and bypass valve shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program 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.

• Separate Condition entry is allowed for each valve.

• • The provisions of specification 3.0.4 and 4.0.4 are not applicable.

TURKEY POINT - UNITS 3 & 4 3/4 7-13 AMENDMENT NOS. 274 AND 269

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 300 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 AMENDMENT NO. 259 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 AMENDMENT NO. 154 TO FACILITY OPERATING LICENSE NO. NPF-86 AMENDMENT NO. 238 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-67 AMENDMENT NO. 189 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 AMENDMENT NO. 274 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-41 DUANE ARNOLD ENERGY CENTER POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SEABROOK STATION, UNIT NO. 1 ST. LUCIE PLANT, UNIT NOS. 1 AND 2 TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 NEXTERA ENERGY RESOURCES/FLORIDA POWER & LIGHT COMPANY, ET AL.

DOCKET NOS. 50-331, 50-266. 50-301. 50-443. 50-335. 50-389. 50-250. and 50-251

1.0 INTRODUCTION

By application dated July 28, 2016, 1 as supplemented by letter dated December 15, 2016, 2 NextEra Energy Resources/Florida Power & Light Company (the licensee) requested changes to the Technical Specifications (TSs) for Duane Arnold Energy Center (Duane Arnold); Point Beach Nuclear Plant, Units 1 and 2 (Point Beach 1 and 2); Seabrook Station, Unit No. 1 (Seabrook); St. Lucie Plant, Unit Nos. 1 and 2 (St. Lucie 1 and 2); and Turkey Point Nuclear Generating Unit Nos. 3 and 4 (Turkey Point 3 and 4). The TSs are contained in 1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML16214A276; L-2016-137.

2 ADAMS Accession No. ML16350A041; L-2016-219.

Enclosure 9

Appendix A of each plant's facility or renewed facility operating license. The licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STSs) Change Traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal & [and] Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing," dated October 21, 2015. 3 By electronic mail (e-mail) dated November 18, 2016, 4 the NRC sent a request for additional information to the licensee. By letter dated December 15, 2016, the licensee responded to the request. The licensee's response provided clarifying information that did not expand the scope of the application and did not change the staff's original proposed no significant hazards consideration (NSHC) determination, as published in the Federal Regsiter (FR) on October 11, 2016 (81 FR 70180 and 70181).

The licensee's letter dated July 28, 2016, also included a request to use American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, "lnservice Test Frequency," as an alternative to certain ASME OM Code requirements at Point Beach 1 and 2, Seabrook, St. Lucie 1 and 2, and Turkey Point 3 and 4. 5 The U.S. Nuclear Regulatory Commission (NRC) considered this request separately from the proposed license amendment, and authorized the licensee's use of this alternative by letter dated December 15, 2016. 6

2.0 REGULATORY EVALUATION

2.1 Description of lnservice Testing Requirements and TSTF-545 An inservice test is a test to assess the operational readiness of a system, structure, or component after first electrical generation by nuclear heat. The ASME OM Code provides requirements for inservice testing of certain components in light-water nuclear power plants.

The ASME OM Code identifies the components subject to the testing (i.e., pumps, valves, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 50.55a{f),

"lnservice testing requirements," requires that inservice testing of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda. The TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.

The regulation in 10 CFR 50.55a(f)(5)(ii) states that if a revised inservice test program for a facility conflicts with the TSs, the licensee must apply to the Commission for amendment of the TSs to conform the TSs to the revised program. TSTF-545, Revision 3 provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a and the TSs. TSTF-545, Revision 3, proposes elimination of the lnservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the lnservice Testing Program. The elimination of the lnservice Testing Program from the TSs could cause 3 ADAMS Accession No. ML15294A555.

4 ADAMS Accession No. ML16326A009.

5 The NRG approved the use of ASME OM Code Case OMN-20 for Duane Arnold Energy Center on June 9, 2014 (ADAMS Accession No. ML14144A002).

6 ADAMS Accession No. ML16330A118.

confusion about the correct application of these surveillance requirements. Therefore, TSTF-545, Revision 3, also proposes adding a new definition, "INSERVICE TESTING PROGRAM," to the TSs, which would be defined as the licensee program that fulfills the requirements of 10 CFR 50.55a(f). TSTF-545, Revision 3, proposes replacement of existing uses of the term, "lnservice Testing Program," with the defined term, as denoted by capitalized letters, throughout the TSs.

By letter dated December 11, 2015, 7 the NRC found the changes to the STSs proposed in TSTF-545, Revision 3 to be suitable for incorporation into the STSs and announced that licensees could request amending their licenses to adopt TSTF-545, Revision 3. The NRC published a notice of availability of TSTF-545, Revision 3 in the FR on March 28, 2016 (81 FR 17208).

2.2 Proposed Technical Specifications Changes The licensee requested to revise the plants' TSs by deleting the the IST program TSs from the Administrative Controls TS sections for Duane Arnold (TS 5.5.6), Point Beach (TS 5.5.7), and St. Lucie (TS 6.8.4.i), and the inservice testing requirements 8 from TS 4.0.5 for Seabrook and Turkey Point, as follows, with proposed deletions shown as stricken text and additions as bolded text).

Duane Arnold:

DELETEDlnservice Testing Program This program provides controls for inservise testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing Frequensies specified in Sestion XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follmvs:

ASME Boiler and Pressure Vessel Code and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Biquarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days 7 ADAMS Package Accession No. ML15317A071.

8 The Administrative Controls sections of the TSs for Seabrook and Turkey Point do not include a program for inservice testing; rather, Seabrook and Turkey Point TS 4.0.5 contains the requirements regarding inservice testing. Discussion of "inservice testing program TSs" throughout this safety evaluation is intended to include the inservice testing requirements in Seabrook and Turkey Point TS 4.0.5.

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservise testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

Point Beach 1 and 2:

Deletedlnservice Testing Program This program provides controls for inservise testing of ASME Code Class 1, 2, and 3 pumps and valves. The program shall include the following:

a. Testing Frequensies specified in the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) and applicable Addenda are as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice testing inservice testing activities activities Sem iq uarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 24 months

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and other normal and accelerated Frequencies, that do not exceed two years, specified in the lnservice Testing Program for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservise testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

St. Lucie 1 and 2:

Deletedlnservice Testing Program This program provides controls for inservise testing of ASME Code Class 1, 2, and 3 components (pumps and valves). The program shall include the following:

a. Testing Frequensies specified in Sestion XI of the ASME Boiler and Pressure Vessel Code* and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code* and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 4.0.2 are applicable to the above required frequencies for performing inservice testing activities;

c. The provisions of SR 4.0.3 are applicable to inservise testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code* shall be construed to supersede the requirements of any technical specification.

• Where ASME Bieler and Pressure Vessel Code is referenced it also refers to the applicable portions of ASME.tANSI OM Code, "Operation and Maintenance of Nuclear Power Plants," with applicable addenda, to the extent it is referenced in the Code.

Seabrook:

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. lnservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g),

except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).

lnservice testing of ASME Code Class 1, 2, and 3 components shall be performed in accordance with the Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(f), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(f)(6)(i).

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and the ASME OM Code including applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and the ASME OM Code including applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Code and the ASME OM Gede including applicable Required Frequencies for Addenda terminology for performing service inservice inspection and testing Inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Semi-quarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 'days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities;

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and

e. Nothing in the ASME Boiler and Pressure Vessel Code or the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

Turkey Point 3 and 4:

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. lnservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a.

lnservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50, Section 50.55a.

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda and the ASME OM Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Code and the ASME OM Required frequencies for Gede and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days

Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities;

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and

e. DELETEDNothing in the ASME Boiler and Pressure Vessel Code or the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

f. Each reactor coolant pump flywheel shall be inspected at least once every 20 years, by either conducting an in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or conduct a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the disassembled flywheel.

SR 3.0.2 for Duane Arnold and Point Beach and SR 4.0.2 for Seabrook, St. Lucie, and Turkey Point allow an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, SR 3.0.3 for Duane Arnold and Point Beach and SR 4.0.3 for St. Lucie allow the licensee to delay declaring the associated limiting condition for operation not met in order to perform the missed surveillance. The licensee did not request changes to SR 3.0.2 or SR 3.0.3 for Duane Arnold and Point Beach or to SR 4.0.2 and SR 4.0.3 for Seabrook, St. Lucie, and Turkey Point.

The licensee requested to revise the Definitions section of the TSs by adding the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The licensee also requested that all existing occurrences of "lnservice Testing Program" for Duane Arnold, Point Beach, and St. Lucie, and "Specification 4.0.5," as they relate to inservice testing, for Seabrook and Turkey Point in TS SRs be replaced with "INSERVICE TESTING PROGRAM," so that the SRs refer to the new definition in lieu of the deleted program.

For St. Lucie and Turkey Point, the licensee proposed conforming changes to the TSs' index pages denoting the addition of the new definition.

2.3 Regulatory Review The staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes.

Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Paragraph 50.36(c)(3) states that SRs are requirements

relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Paragraph 50.36(c)(5) states that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

The staff's guidance for review of the TSs is in NUREG-0800, "Standard Review Plan,"

Chapter 16, "Technical Specifications," Revision 3, dated March 2010. 9 As described therein, as part of the regulatory standardization effort, the staff has prepared improved STSs for each of the light-water reactor nuclear steam supply systems and associated balance-of-plant equipment systems. The licensee's proposed amendments are based on TSTF-545, Revision 3, which is an NRG-approved change to the improved STSs. The staff's review included consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. The staff gives special attention to TS provisions that depart from the improved STSs, as modified by NRG-approved TSTF Travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met.

lnservice Testing Pursuant to 10 CFR 50.54, "Conditions of licenses," the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f).

Paragraph 50.55a(f) states that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel (BPV) Code and ASME OM Code as specified in the paragraph, and that each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions of 10 CFR 50.55a(f)(1) through (f)(6).

The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements, with conditions, were suitable for incorporation into the NRC's rules.

Paragraph 50.55(a)(f)(5)(ii) of 10 CFR states that if a revised inservice test program for a facility conflicts with the TSs for the facility, the licensee must apply for an amendment of the TSs to conform the TSs to the revised program.

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and lnservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants," Final Report, published October 2013, 10 provides guidance for the inservice testing of pumps and valves.

NUREG-0800, Section 3.9.6, "Functional Design, Qualification, and lnservice Testing Programs for Pumps, Valves, and Dynamic Restraints," Revision 3, dated March 2007, 11 provides 9 ADAMS Accession No. ML100351425.

10 ADAMS Accession No. ML13295A020.

11 ADAMS Accession No. ML070720041.

guidance and acceptance criteria for the staff's review of the inservice testing program for pumps and valves.

3.0 TECHNICAL EVALUATION

The staff evaluated the licensee's application to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5) (i.e.,

provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). The staff also considered whether the TSs, as amended, would assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met per 10 CFR 50.36(c)(3). In making its determination as to whether to amend the license, the staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54. Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.

3.1 Deletion of the lnservice Testing Program TSs For Duane Arnold, Point Beach, and St. Lucie, the lnservice Testing Program TSs are TS 5.5.6, TS 5.5.7, and TS 6.8.4.i, respectively, which are in the Administrative Controls section of the TSs. For Seabrook and Turkey Point, the inservice testing requirements are in TS 4.0.5. The inservice testing program TSs and the second paragraph of Seabrook and Turkey Point TS 4.0.5.a have requirements for inservice testing of ASME Code Class 1, 2, and 3 components (i.e., pumps and valves). Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54. Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained further in this safety evaluation, it is not necessary to have additional administrative controls in the TSs for Duane Arnold, Point Beach, and St. Lucie relating to the inservice testing program to assure operation of the facility in a safe manner. For the reasons explained further in this safety evaluation, it is also not necessary to have additional requirements in TS 4.0.5 for Seabrook and Turkey Point relating to the inservice testing program to assure the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Deletion of TS Jnservice Testing Program Frequency Descriptions The ASME OM Code requires testing to normally be performed within certain time periods.

Duane Arnold TS 5.5.6.a, Point Beach TS 5.5.7.a, St. Lucie TS 6.8.4.i.a, and Seabrook and Turkey Point TS 4.0.5.b set inservice testing frequencies more precisely than those specified in the ASME OM Code and applicable addenda (e.g., "at least once per 31 days" contrasted with "monthly"). However, the staff determined that the more precise inservice testing frequencies

are not necessary to assure operation of the facility in a safe manner or the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Therefore, the staff found these proposed changes acceptable.

Deletion of SR 3.0.214.0.2 Provisions from lnservice Testing Program TSs Duane Arnold TS 5.5.6.b, Point Beach TS 5.5.7.b, St. Lucie TS 6.8.4.i.b, and Seabrook and Turkey Point TS 4.0.5.c allow the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by the plants' inservice testing TSs (and for Point Beach, other normal and accelerated frequencies specified as 2 years or less in the lnservice Testing Program). Similar to these TSs, the NRC authorization 12 of ASME OM Code Case OMN-20 also permits the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent. The staff determined that the TS allowance to extend inservice testing intervals is not needed to assure operation of the facility in a safe manner or the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Therefore, the staff determined that deletion of these TSs as they pertain to inservice testing is acceptable. The deletion of these TSs does not impact the licensee's ability to extend inservice testing intervals using ASME OM Code Case OMN-20, as authorized by the NRC. Therefore, the staff found these proposed changes acceptable.

Deletion of SR 3.0.314.0.3 Provisions from Duane Arnold, Point Beach, and St. Lucie lnservice Testing Program TSs Duane Arnold TS 5.5.6.c, Point Beach TS 5.5.7.c, and St. Lucie TS 6.8.4.i.c allow the licensee to use SR 3.0.3 (or SR 4.0.3 for St. Lucie) when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 3.0.3 (or SR 4.0.3) allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 3.0.3 (or SR 4.0.3) for inservice tests is limited to those inservice tests required by an SR. In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test. Deletion of Duane Arnold TS 5.5.6.c, Point Beach TS 5.5.7.c, and St. Lucie TS 6.8.4.i.c does not change any of these requirements, and SR 3.0.3 (or SR 4.0.3) will continue to apply to those inservice tests required by SRs. Therefore, the staff determined that deletion of these TSs is acceptable.

Deletion of Duane Arnold TS 5. 5. 6. d, Point Beach TS 5. 5. 7. d, St. Lucie TS 6. 8. 4. i. d, and Seabrook and Turkey Point TS 4.0.5.e Duane Arnold TS 5.5.6.d, Point Beach TS 5.5.7.d, St. Lucie TS 6.8.4.i.d, and Seabrook and Turkey Point 4.0.5.e state that nothing in the ASME OM (or BPV) Code shall be construed to supersede the requirements of any TS. However, the regulations in 10 CFR 50.55a(f)(5)(ii) address what to do if a revised inservice testing program for a facility conflicts with the TSs for the facility. These regulations require the licensee to apply for an amendment to the TSs to conform the TSs to the revised program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for a TS stating how to address conflicts between the TSs and the inservice testing program because the regulations specify how conflicts must be resolved. Therefore, the staff finds the deletion of these TSs 12 See footnotes 5 and 6.

acceptable. The staff also finds that for Seabrook and Turkey Point, the deletion of TS 4.0.5.e is acceptable because the regulations in 10 CFR 50.55a(g)(5)(ii) also address what to do if a revised inservice inspection program for a facility conflicts with the TSs for the facility.

Conclusion Regarding Deletion of lnservice Testing Program TSs The NRC staff determined that the requirements currently in the inservice testing program TSs are not necessary to assure operation of the facility in a safe manner or the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Based on this evaluation, the staff concludes that deletion of the inservice testing program TSs from the licensee's TSs is acceptable because the inservice testing program TSs are not required by 10 CFR 50.36(c)(5) or 10 CFR 50.36(c)(3).

3.2 Definition of INSERVICE TESTING PROGRAM and Revision to SRs The licensee proposed to revise the TS Definitions section to include the term, "INSERVICE TESTING PROGRAM," with the following definition: "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f)." The proposed definition is consistent with the definition in TSTF-545, Revision 3. The staff finds the definition acceptable because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).

The licensee requested that all existing references to the "lnservice Testing Program" (or "Specification 4.0.5" if applicable to inservice testing requirements for Seabrook and Turkey Point) in SRs be revised to "INSERVICE TESTING PROGRAM" to reference the new TS defined term in lieu of the deleted inservice testing program TSs. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in SRs with the new definition. The staff verified that for each SR reference to the "lnservice Testing Program," or "Specification 4.0.5," as applicable, the licensee proposed to change the reference to "INSERVICE TESTING PROGRAM." The proposed change does not alter how the SR testing is performed. However, the inservice testing frequencies could change because the TSs will no longer include the more precise test frequencies in Duane Arnold TS 5.5.6.a, Point Beach TS 5.5.7.a, St. Lucie TS 6.8.4.i.a, and Seabrook and Turkey Point 4.0.5.b. As discussed in Section 3.1 of this safety evaluation, the staff determined that the TSs do not need to include the more precise testing frequencies currently in the TSs. Based on its review, the staff determined that revising the SRs to refer to the new definition is acceptable because these SRs will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The staff also determined that, with the proposed changes, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

3.3 Conforming Changes and Variations from TSTF-545 The staff evaluated the following conforming changes and variations from TSTF-545, Revision 3 not previously addressed in this safety evaluation.

a. The TSs for St. Lucie, Seabrook, and Turkey Point have not been converted to the improved STSs on which TSTF-545, Revision 3 is based. As a result, the numbering, format, and content of these TSs vary from TSTF-545, Revision 3. In addition, all the plants' TSs, including the Point Beach TSs, use different numbering than the improved

STSs. The NRC staff finds that the licensee's proposed deviations in numbering, format, and content are editorial in nature and that the licensee's proposed TS changes are consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

b. The index for the St. Lucie and Turkey Point TSs is included as part of the TSs.

Therefore, the licensee included conforming changes to the index resulting from the addition of the new definition. The staff finds that the proposed deviations are editorial in nature and that the licensee's proposed TS changes are consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

c. The inservice testing program TSs for Duane Arnold and St. Lucie refer to testing frequencies specified in Section XI of the ASME BPV Code (St. Lucie's TSs have a note indicating that the ASME OM Code is also applicable to this TS), which varies from TSTF-545, Revision 3. As discussed in the NRC's authorizations for ASME OM Code Case OMN-20 for these plants, 13 the code of record for these plants' inservice testing is the ASME OM Code, and 10 CFR 50.55a(f) requires the inservice testing program to meet the ASME OM Code. Therefore, deletion of these TSs does not create new requirements related to the ASME OM Code for these plants. As discussed in Section 3.1 of this safety evaluation, the staff found the proposed deletion of the inservice testing program TSs, which include these references to Section XI of the ASME BPV Code, acceptable.

d. The licensee proposed to replace the content of lnservice Testing Program TSs for Duane Arnold, Point Beach, and St. Lucie with the word, "Deleted," or "DELETED," and retain the existing numbering sequence. The staff finds that these proposed changes are editorial in nature and consistent with the intent of TSTF-545, Revision 3. Therefore, the staff finds that the licensee's proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the staff notified officials from the States of Iowa, Wisconsin, New Hampshire, Massachusetts, and Florida on January 19, 2017, of the proposed issuance of the amendments. Each State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change inspections or SRs. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding, which was published in the FR on October 11, 2016 (81 FR 70180), that the amendments involve NSHC, and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 13 See footnotes 5 and 6.

10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

Based on the aforementioned considerations, the NRC staff concluded that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Audrey Klett Caroline Tilton Robert Wolfgang Date of issuance: April 7, 2017

M. Nazar

SUBJECT:

DUANE ARNOLD ENERGY CENTER; POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2; SEABROOK STATION, UNIT NO. 1; ST. LUCIE PLANT, UNIT NOS. 1 AND 2; AND TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 - ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATIONS FOR INSERVICE TESTING (CAC NOS. MF8202 THROUGH MF8209) DATED APRIL 7, 2017 DISTRIBUTION:

PUBLIC RidsACRS_MailCTR CTilton, NRR LPL 1 R/F LPL2-2 R/F LPL3 R/F RidsNrrDorlLPL 1 RidsNrrDorlLpl2-2 RidsNrrDorlLpl3 RidsRgn 1MailCenter RidsRgn2MailCenter RidsRgn3MailCenter RidsNrrLALRonewicz RidsNrrLABClayton RidsNrrLASRohrer RidsNrrPMDuaneArnold RidsNrrPMPointBeach RidsNrrPMSeabrook RidsNrrPMStLucie RidsNrrPMTurkeyPoint RWolfgang, NRR RidsNrrDssStsb RidsNrrDeEpnb ADAMS A ccess1on No.: ML17027A078 *B;yema1*1 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DE/EPNB/BC* N RR/DSS/STSB/BC*

NAME AKlett BClayton DAiiey A Klein DATE 3/30/17 3/30/17 3/17/17 3/22/17 OFFICE OGC- NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME BHarris BBeasley AKlett DATE 4/5/17 4/7/17 4/7/17 OFFICIAL RECORD COPY