Text

St. Lucie, Unit 2 - License Amendment Request for Extended Power Uprate, Attachment 3; Renewed Facility Operating License and Technical Specifications Markups and Clean Pages
	ML110730284
Person / Time
Site:	Saint Lucie
Issue date:	02/25/2011
From:	; Florida Power & Light Co
To:	; Office of Nuclear Reactor Regulation
References
	L-2011-021
	Download: ML110730284 (133)
	v • d • e

St. Lucie Unit 2 L-2011-021 Docket No. 50-389 Attachment 3 ATTACHMENT 3 LICENSE AMENDMENT REQUEST EXTENDED POWER UPRATE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS MARKUPS AND CLEAN PAGES FLORIDA POWER & LIGHT ST. LUCIE UNIT 2 This coversheet plus 13 pages __________________________________________________________________________ St. Lucie Unit 2 EPU LAR Att. 3-1 Renewed Facility Operating License and Technical Specifications Markups and Clean Pages St. Lucie Unit 2 L-2011-021 Docket No. 50-389 Attachment 3 LIST OF AFFECTED PAGES St. Lucie Unit 1 EPU LAR Att. 3-2 Renewed Facility Operating License and Technical Specifications Markups and Clean Pages Renewed Facility Operating License

Technical Specifications I

XXI XXII INSERT 1 XXIII 1-3 INSERT 2 1-4 1-5 2-3 (replaced by INSERT 3)

INSERT 3 2-4 2-5 2-6 3/4 1-1 3/4 1-3 3/4 1-8 3/4 1-13 3/4 1-14 3/4 1-15 (replaced by INSERT 4)

INSERT 4 3/4 1-24 3/4 2-14

3/4 2-15 3/4 4-8 3/4 4-25 INSERT 5 3/4 4-27

3/4 4-28

3/4 4-31a 3/4 4-31b 3/4 4-37a

3/4 5-1 3/4 5-3 3/4 5-5 3/4 5-8 3/4 7-3 3/4 8-1

Technical Specifications (continued) 3/4 8-6 INSERT 6 3/4 8-7 INSERT 7 3/4 8-9 3/4 9-1 3/4 9-12 3/4 10-1 3/4 11-15 5-4 INSERT 8 5-4a 5-4b 5-4C (DELETED) 5-4D (DELETED)

5-4E (DELETED)

5-4F (DELETED)

5-4G (DELETED)

5-4h 5-4i 5-4j 5-4k 5-4l 5-4m 5-4n 5-4o 6-15b 6-20 6-20a 6-20c 6-20d 6-20e INSERT 9 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 CFR 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 2700 megawatts (thermal).

Commencing with the startup for Cycle 16 and until the Combustion Engineering Model 3410 Steam Generators are replaced, the maximum reactor core power shall not exceed 89 percent of 2700 megawatts (thermal) if:

a. The Reactor Coolant System Flow Rate is less than 335,000 gpm but greater than or equal to 300,000 gpm, or

b. The Reactor Coolant System Flow Rate is greater than or equal to 300,000 gpm AND the percentage of steam generator tubes plugged is greater than 30 percent (2520 tubes/SG) but less than or equal to 42 percent (3532 tubes/SG).

This restriction in maximum reactor core power is based on analyses provided by FPL in submittals dated October 21, 2005 and February 28, 2006, and approved by the NRC in Amendment No. 145, which limits the percent of steam generator tubes plugged to a maximum of 42 percent (3532 tubes) in either steam generator and limits the plugging asymmetry between steam generators to a maximum of

600 tubes.

B. Technical Specifications

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 157 are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16Revised by letter dated October 24, 2008 3020 XXX XXXXXX XX, XXXX ST. LUCIE - UNIT 2 IAmendment No. 13 , 92 INDEX DEFINITIONS

SECTION PAGE1.0 DEFINITIONS1.1AC TION...........................................................................................................................1-11.2AXIAL SHAPE INDEX....................................................................................................1-11.3AZIMUTHAL POWER TILT............................................................................................1-11.4CHANNEL CALIBRATION.............................................................................................1-11.5CHANNEL CHECK.........................................................................................................1-11.6CHANNEL FUNCTION AL TEST....................................................................................1-21.7CONTAINMENT VESSEL INTEGRITY..........................................................................1-21.8CONTROLLED LEAKAGE.............................................................................................1-21.9CORE ALTERATION......................................................................................................1-21.9aCORE OPERATING LIMITS REPORT (COLR)............................................................1-21.10DOSE EQUIVALENT I-131............................................................................................1-31.11 - AVERAGE DISI NTEGRATION ENERGY................................................................1-31.12ENGINEERED SAFETY F EATURES RESPONSE TIME.............................................1-31.13FREQUENCY NOTATION.............................................................................................1-31.14GASEOUS RADWASTE TREATMENT SYSTEM.........................................................1-31.15IDENTIFIED LEAKAGE..................................................................................................1-31.16LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE....................1-41.17MEMBER(S) OF THE PUBLIC.......................................................................................1-41.18OFFSITE DOSE CALCULATION MANUAL (ODCM)....................................................

1-41.19OPERABLE - OPERABILITY........................................................................................1-4 1.20OPERATIONAL MODE - MODE...................................................................................1-41.21PHYSICS TESTS...........................................................................................................1-4 1.22PRESSURE BOUNDA RY LEAKAGE............................................................................1-51.23PROCESS CONTROL PROGRAM................................................................................1-51.24PURGE - PURGING......................................................................................................1-5 1.25RATED THERMAL POWER..........................................................................................1-51.26REACTOR TRIP SYSTEM RESPONSE TIME..............................................................

1-51.27REPORTABLE EVENT..................................................................................................1-51.28SHIELD BUILDING INTEGRITY....................................................................................1-5 1.29SHUTDOWN MARGIN...................................................................................................1-61.30SITE BO UNDARY..........................................................................................................1-6 E DOSE EQUIVALENT XE-133 DELETED ST. LUCIE - UNIT 2 XXI Amendment No. 8 , 53 , 73 , 92 , 112 , 117 , 154 INDEX LIST OF FIGURES FIGURE PAGE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING...........................................................2-3 2.2-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 1 (FRACTION OF RATED THERMAL POWER VERSUS QR 2)..................................2-7 2.2-2 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 2 (QR 2 VERSUS Y 1).....................................................................................................2-8 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 (Y 1 , VERSUS A 1)......................................................................................................2-9 2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR 1)................................2-10

3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AS A FUNCTION OF STORED BORIC ACID CONCENTRATION................................................3/4 1-15 3.1-1a DELETED..............................................................................................................

3.1-2 DELETED..............................................................................................................

3.2-1 DELETED..............................................................................................................

3.2-2 DELETED..............................................................................................................

3.2-3 DELETED..............................................................................................................

4.2-1 DELETED..............................................................................................................

3.2-4 DELETED..............................................................................................................

3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMITS VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY > 1 Ci/GRAM DOSE EQUIVALENT I-131...........................................................................................3/4 4-28 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE- TEMPERATURE LIMITS FOR 55 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST...........................................................................................3/4 4-31a 47 VS DELETED BAMT ST. LUCIE - UNIT 2 XXII Amendment No. 8 , 29 , 53 , 101 , 112 , 117 , 135 , 154 INDEX LIST OF FIGURES (continued)

FIGURE PAGE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, COOLDOWN AND INSERVICE TEST.............................................................................................................3/4 4-31b

3.4-4 DELETED.......................................................................................................3/4 4-32

4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..............................3/4 7-25

5.1-1 SITE AREA MAP....................................................................................................5-2

5.6-1a REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT and DECAY TIME, REGION II, 1.3 w/o......................................................................5-4B 5.6-1b REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT and DECAY TIME, REGION II, 1.5 w/o......................................................................5-4C 5.6-1c REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT and DECAY TIME, REGION I, 1.4 w/o.......................................................................5-4D 5.6-1d REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT and DECAY TIME, REGION I, 1.82 w/o.....................................................................5-4E 5.6-1e REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT, REGION I, 2.82 w/o.............................................................................................5-4F

5.6-1f REQUIRED FUEL ASSEMBLY BURNUP vs INITIAL ENRICHMENT, REGION II CASK PIT STORAGE RACK.............................................................5-4G

6.2-1 DELETED...............................................................................................................6-3 6.2-2 DELETED...............................................................................................................6-4 47 DELETED INSERT 1 5.6-1 ALLOWABLE REGION 1 STORAGE PATTERNS AND FUEL ARRANGEMENTS-------------------------5-4h

5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 1 of 3) ------------------...5-4i

5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 2 of 3) ------------------...5-4j

5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 3 of 3) ------------------...5-4k

5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 1 of 2)----------------------------5-4l

5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 2 of 2)---------------------------..5-4m

5.6-4 ALLOWABLE CASK PIT STORAGE RACK PATTERNS---------5-4n

INSERT 1 ST. LUCIE - UNIT 2XXIIIAmendment No. 8 , 73 , 74, 86, 104INDEXLIST OF TABLES TABLEPAGE1.1FREQUENCY NOTATION....................................................................................1-81.2OPERATIONAL MODES......................................................................................1-92.2-1REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS..........2-43.1-1DELETED3.2-1DELETED.....................................................................................................3/4 2-11 3.2-2DNB MARGIN LIMITS..................................................................................3/4 2-15 3.3-1REACTOR PROTECTIVE INSTRUMENTATION...........................................3/4 3-2 3.3-2DELETED.....................................................................................................4.3-1REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCEREQUIREMENTS...........................................................................................3/4 3-83.3-3ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION...................................................................................3/4 3-123.3-4ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION TRIP VALUES...........................................................3/4 3-173.3-5DELETED.....................................................................................................4.3-2ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................3/4 3-223.3-6RADIATION MONITORING INSTRUMENTATION.......................................3/4 3-25 4.3-3RADIATION MONITORING INSTRUMENTATION SURVEILLANCEREQUIREMENTS.........................................................................................3/4 3-283.3-8DELETED.....................................................................................................

4.3-5DELETED.....................................................................................................

DELETED 5.6-1 MINIMUM BURNUP COEFFICIENTS..............................................................5-4o ST. LUCIE - UNIT 2 1-3 Amendment No. 105 , 137 , 147 , 152 DEFINITIONS DOSE EQUIVALENT I-131

1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." - AVERAGE DISINTEGRATION ENERGY

1.11 shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. 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.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

FREQUENCY NOTATION

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

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED 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).

E E INSERT 2

INSERT 2 ST. LUCIE - UNIT 21-4Amendment No. 16, 31, 46 , 61 , 63DEFINITIONS LOW TEMPERATURE OVERPRESSURE PROTECTION RANGE - RCS1.16The LOW TEMPERATURE OVERPRESSURE PROTECTION RANGE is that operatingcondition when (1) the RCS cold leg temperature is less than or equal to thatspecified in Table 3.4-3, and (2) the Reactor Coolant System is not vented tocontainment by an opening of at least 3.58 square inches.MEMBER(S) OF THE PUBLIC1.17MEMBER OF THE PUBLIC means an individual in a controlled or unrestrictedarea. However, an individual is not a member of the public during any periodin which the individual receives an occupational dose.OFFSITE DOSE CALCULATION MANUAL (ODCM)1.18THE OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodologyand parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous andliquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of theEnvironmental Radiological Monitoring Program. The ODCM shall also contain(1) the Radioactive Effluent Controls and Radiological EnvironmentalMonitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological EnvironmentalOperating and Annual Radioactive Effluent Release Reports required bySpecifications 6.9.1.7 and 6.9.1.8.OPERABLE - OPERABILITY1.19A system, subsystem, train, component or device shall be OPERABLE orhave 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 itsfunction(s) are also capable of performing their related support function(s).OPERATIONAL MODE - MODE1.20An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusivecombination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2.PHYSICS TESTS1.21PHYSICS TESTS shall be those tests performed to measure the fundamentalnuclear characteristics of the reactor core and related instrumentation and(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisionsof 10 CFR 50.59, or (3) otherwise approved by the Commission.

Deleted ST. LUCIE - UNIT 2 1-5 Amendment No. 9 , 13 , 61 , 137 , 147 DEFINITIONS PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component

body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING

1.24 PURGE or PURGING is the controlled process of discharging air or gas

from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER

1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME

1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power to the CEA drive mechanism is interrupted. 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. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRITY 1.28 SHIELD BUILDING INTEGRITY shall exist when:

a. Each door is closed except when the access opening is being used for normal transit entry and exit;

b. The shield building ventilation system is in compliance with Specification 3.6.6.1, and

c. The sealing mechanism associated with each penetration (e.g.,

welds, bellows or O-rings) is OPERABLE.

3020 ST. LUCIE - UNIT 22-3Amendment No. 8, 138FIGURE 2.1-1:REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINESFOUR REACTOR COOLANT PUMPS OPERATING I 0 24115 pall . . UNACClPTAlL! " TMNSfENT , , CoNDIT\ONS 620 --. --1£,600 1866 pall I ....... i

._ ***

! ... -.... wmVI-J"VE is ' I .................

......... . 1 . ..... i . ... ....... ..........

--. ! seo . AOCIPTAILI TRANSISNT

'I , 5 540 UMIT8 CONTAIN 'NO ALL.O. ,, __ . CONomoNS .-* FOR IN8TRUMINT eRROR OR , I I , ...I t=1.UCTVATIONS . ! 520 VALID FOR TOTAt.INT!GAATeC . l-. RACIAL P&AI<lNG FACTORS LUS THAN SPECtFIIO IN TIiCH SPIC U.3. , MAXIMUM OF 10nr. 1100 I-PU!ACTOR CONDITIONS LlMITID TO pP RATIO THI!RMAL LESS THAN THE 'l'IMPIRATUNS POWlR8VTHS OEFlN!O BV THE ACTUAT/dN OF THe , HIGH POWIR Level. I 480 -SeCONDARY SAPlTVVALVE!S TRIP 'I , I ' '\ I ' . .' I 460 , 0.40 O.eo 0.80 1.00 uo .. MOTION 0 .. FtA TIll) THINlAi. POWIR 460 480 500 520 540 560 580 600 620 6400.400.500.600.700.800.901.001.101.20FRACTION OF RATED THERMAL POWERVESSEL INLET TEMPERATURE (F) SECONDARY SAFETY VALVES1855 psia2415 psia2250 psiaTHERMAL POWER LIMITED TO A MAXIMUM OF 107% OF RATED THERMAL POWER BY THE HIGH POWER LEVEL TRIPUNACCEPTABLE TRANSIENT CONDITIONS ACCEPTABLE TRANSIENT CONDITIONSLIMITS CONTAIN NO ALLOWANCE FOR INSTRUMENT ERROR OR FLUCTUATIONSVALID FOR TOTAL INTEGRATED RADIAL PEAKING FACTORS LESS THAN SPECIFIED IN TECH SPEC 3.2.3.REACTOR CONDITIONS LIMITED TO LESS THAN THE TEMPERATURES DEFINED BY THE ACTUATION OF THE SECONDARY SAFETY VALVES FIGURE 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING

ST. LUCIE - UNIT 2 2-3 Amendment No. 8, 138 INSERT 3 ST. LUCIE - UNIT 22-4Amendment No. 8 , 23 , 60TABLE 2.2-1REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITSFUNCTIONAL UNITTRIP SETPOINTALLOWABLE VALUES1.Manual Reactor TripNot ApplicableNot Applicable2.Variable Power Level - High (1)Four Reactor Coolant PumpsOperating< 9.61% above THERMAL POWER,with a minimum setpoint of15% of RATED THERMAL POWER,and a maximum of < 107.0% ofRATED THERMAL POWER.< 9.61% above THERMAL POWER, anda minimum setpoint of 15% ofRATED THERMAL POWER and a maximumof < 107.0% of RATED THERMAL POWER.3.Pressurizer Pressure - High< 2370 psia< 2374 psia4.Thermal Margin/Low Pressure (1)Four Reactor Coolant Pumps OperatingTrip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4.Minimum value of 1900 psia.Trip setpoint adjusted to notexceed the limit lines of Figures 2.2-3 and 2.2-4.Minimum value of 1900 psia.5.Containment Pressure - High< 3.0 psig< 3.1 psig6.Steam Generator Pressure - Low> 626.0 psia (2)> 621.0 psia (2)7.Steam Generator Pressure (1)Difference - High (Logic in TM/LP Trip Unit)< 120.0 psid< 132.0 psid8.Steam Generator Level - Low> 20.5% (3)> 19.5% (3) 35.0%(3)Low (6), (7)34.1%(3)psia (2)

ST. LUCIE - UNIT 22-5Amendment No. 8 , 60, 131 , 138TABLE 2.2-1 (Continued)REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITSFUNCTIONAL UNITTRIP SETPOINTALLOWABLE VALUES9.Local Power Density - High (5)OperatingTrip setpoint adjusted tonot exceed the limit linesof Figures 2.2-1 and 2.2-2Trip setpoint adjusted tonot exceed the limit linesof Figures 2.2-1 and 2.2-2.10.Loss of Component Cooling Waterto Reactor Coolant Pumps - Low> 636 gpm**> 636 gpm11.Reactor Protection System LogicNot ApplicableNot Applicable12.Reactor Trip BreakersNot ApplicableNot Applicable13.Rate of Change of Power - High (4)< 2.49 decades per minute< 2.49 decades per minute14.Reactor Coolant Flow - Low (1)> 95.4% of design ReactorCoolant flow with four pumps operating*> 94.9% of design ReactorCoolant flow with four pumps operating*15.Loss of Load (Turbine)Hydraulic Fluid Pressure - Low (5)> 800 psig> 800 psig

Design reactor coolant flow with four pumps operating is the minimum RCS flow specified in the COLR Table 3.2-2. ** 10-minute time delay after relay actuation.

For minimum , refer to minimum ST. LUCIE - UNIT 22-6Amendment No. 98TABLE 2.2-1 (Continued)REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITSTABLE NOTATION(1)Trip may be manually bypassed below 0.5% of RATED THERMAL POWER during testing pursuant to Special Test Exception 3.10.3; bypass shallbe automatically removed when Wide Range Logarithmic Neutron Flux power is greater than or equal to 0.5% of RATED THERMAL POWER.(2)Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig.(3)% of the narrow range steam generator level indication.(4)Trip may be bypassed below 10

-4% and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when Wide RangeLogarithmic Neutron Flux power is > 10

-4% and Power Range Neutron Flux power < 15% of RATED THERMAL POWER.(5)Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when Power Range Neutron Flux power isgreater than or equal to 15% of RATED THERMAL POWER.

(6) If the as-found channel setpoint is either outside its predefined as-found acceptance criteria band or is not conservativ e with respect to the Allowable Value, then the channel shall be declared inoperable and shall be evaluated to verify that it is functioning as r equired before

returning the channel to service.

(7) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the Trip Setpoint or a value that is more

conservative than the Trip Setpoint, otherwise that channel shall not be returned to OPERABLE status. The Trip Setpoint and the methodology used to determine the Trip Setpoint, the as-found acceptance criteria band, and the as-left acceptance criteria are specified i n the UFSAR.

ST. LUCIE - UNIT 23/4 1-1Amendment No. 25, 89, 1053/4.1 REACTIVITY CONTROL SYSTEMS3/4.1.1 BORATION CONTROLSHUTDOWN MARGIN - Tavg GREATER THAN 200

°FLIMITING CONDITION FOR OPERATION 3.1.1.1The SHUTDOWN MARGIN shall be within the limits specified in the COLR.APPLICABILITY: MODES 1, 2*, 3 and 4.ACTION:With the SHUTDOWN MARGIN outside the COLR limits, immediately initiate and continueboration at greater than or equal to 40 gpm of a solution containing greater than or equal to1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.SURVEILLANCE REQUIREMENTS 4.1.1.1.1The SHUTDOWN MARGIN shall be determined to be within the COLR limits:a.Within one hour after detection of an inoperable CEA(s) and at least once per12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA is notfully inserted, and is immovable as a result of excessive friction or mechanicalinterference or is known to be untrippable, the above required SHUTDOWNMARGIN shall be verified acceptable with an increased allowance for thewithdrawn worth of the immovable or untrippable CEA(s).b.When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at leastonce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the PowerDependent Insertion Limits of Specification 3.1.3.6.c.When in MODE 2 with Keff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achievingreactor criticality by verifying that the predicted critical CEA position is within thelimits of Specification 3.1.3.6. *See Special Test Exception 3.10.1.

1900 ST. LUCIE - UNIT 23/4 1-3Amendment No. 8 , 25, 105REACTIVITY CONTROL SYSTEMSSHUTDOWN MARGIN - Tavg LESS THAN OR EQUAL TO 200

°FLIMITING CONDITION FOR OPERATION 3.1.1.2The SHUTDOWN MARGIN shall be within the limits specified in the COLR.APPLICABILITY: MODE 5.ACTION:With the SHUTDOWN MARGIN outside the COLR limits, immediately initiate and continueboration at greater than or equal to 40 gpm of a solution containing greater than or equal to1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.SURVEILLANCE REQUIREMENTS 4.1.1.2The SHUTDOWN MARGIN shall be determined to be within the COLR limits:a.Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable CEA(s) and at least once per12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA isimmovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovableor untrippable CEA(s).b.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:1.Reactor coolant system boron concentration,2.CEA position, 3.Reactor coolant system average temperature,4.Fuel burnup based on gross thermal energy generation,5.Xenon concentration, and6.Samarium concentration.c.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , when the Reactor Coolant System is drained belowthe hot leg centerline, by consideration of the factors in 4.1.1.2b and by verifyingat least two charging pumps are rendered inoperable by racking out their motorcircuit breakers.

1900 ST. LUCIE - UNIT 23/4 1-8Amendment No. 8 , 25 , 40, 105REACTIVITY CONTROL SYSTEMSFLOW PATHS - OPERATINGLIMITING CONDITION FOR OPERATION 3.1.2.2At least two of the following three boron injection flow paths shall be OPERABLE:a.One flow path from the boric acid makeup tank(s) with the tank meetingSpecification 3.1.2.8 part a) or b), via a boric acid makeup pump through acharging pump to the Reactor Coolant System.b.One flow path from the boric acid makeup tank(s) with the tank meetingSpecification 3.1.2.8 part a) or b), via a gravity feed valve through a charging pump to the Reactor Coolant System.c.The flow path from the refueling water storage tank via a charging pump to theReactor Coolant System.

ORAt least two of the following three boron injection flow paths shall be OPERABLE:a.One flow path from each boric acid makeup tank with the combined tankcontents meeting Specification 3.1.2.8 c), via both boric acid makeup pumpsthrough a charging pump to the Reactor Coolant System.b.One flow path from each boric acid makeup tank with the combined tankcontents meeting Specification 3.1.2.8 c), via both gravity feed valves through acharging pump to the Reactor Coolant System.c.The flow path from the refueling water storage tank, via a charging pump to theReactor Coolant System.APPLICABILITY: MODES 1, 2, 3 and 4.ACTION:With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System toOPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to aSHUTDOWN MARGIN equivalent to its COLR limit at 200

°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore atleast two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWNwithin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.e.f.

ST. LUCIE - UNIT 23/4 1-13Amendment No. 40, 122REACTIVITY CONTROL SYSTEMSBORATED WATER SOURCES - SHUTDOWNLIMITING CONDITION FOR OPERATION 3.1.2.7As a minimum, one of the following borated water sources shall beOPERABLE:a.One boric acid makeup tank with a minimum borated water volume of3550 gallons of 2.5 to 3.5 weight percent boric acid (4371 to 6119 ppm boron).b.The refueling water tank with:1.A minimum contained borated water volume of 125,000 gallons,2.A minimum boron concentration of 1720 ppm, and 3.A solution temperature between 40

°°°°F and 120°°°

°F.APPLICABILITY: MODES 5 and 6.ACTION:With no borated water sources OPERABLE, suspend all operations involving COREALTERATIONS or positive reactivity changes*.SURVEILLANCE REQUIREMENTS 4.1.2.7The above required borated water source shall be demonstrated OPERABLE:a.At least once per 7 days by:1.Verifying the boron concentration of the water,2.Verifying the contained borated water volume of thetank, andb.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature when itis the source of borated water and the outside air temperature isoutside the range of 40

°°°°F and 120°°°

°F..c.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the Reactor Auxiliary Building airtemperature is less than 55

°°°°F, by verifying that the boric acidmakeup tank solution temperature is greater than 55

°°°°F when thatboric acid makeup tank is required to be OPERABLE. *Plant temperature changes are allowed provided the temperature change is accounted forin the calculated SHUTDOWN MARGIN.

1900 5420 3.1 ST. LUCIE - UNIT 2 3/4 1-14 Amendment No. 8 , 25 , 40 , 105 , 157 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 At least two of the following four borated water sources shall be OPERABLE:

a. Boric Acid Makeup Tank 2A in accordance with Figure 3.1-1.

b. Boric Acid Makeup Tank 2B in accordance with Figure 3.1-1.

c. Boric Acid Makeup Tanks 2A and 2B with a minimum combined contained borated water volume in accordance with Figure 3.1-1.

d. The refueling water tank with:

1. A minimum contained borated water volume of 477,360 gallons,

2. A boron concentration of between 1720 and 2100 ppm of boron, and

3. A solution temperature of between 55 F and 100 F. APPLICABILITY

MODES 1, 2, 3 and 4.

ACTION:

a. With the above required boric acid makeup tank(s) inoperable, restore the tank(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 200F; restore the above required boric acid makeup tank(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.8 At least two required borated water sources shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the boron concentration in the water and

2. Verifying the contained borated water volume of the water source.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature when the outside air temperature is outside the range of 55F and 100 F. c. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the Reactor Auxiliary Building air temperature is less than 55F, by verifying that the boric acid makeup tank solution is greater than 55 F. 1900 and 2200 ST. LUCIE - UNIT 23/4 1-15Amendment No. 40FIGURE 3.1-1ST. LUCIE 2 MIN BAMT(4196 PPM)(4546 PPM)(4895 PPM)(5245 PPM)(5595 PPM)(5944 PPM)(6294 PPM)STORED BAMT CONC (wt % boric acid)2.42.62.833.23.43.6 3 4 5 6 7 8 9 10VOLUME vs STORED BAMT CONCENTRATIONMIN BAMT VOLUME (gallons)

(thousands)UNACCEPTABLE OPERATION (8650)(7500)(6650)(5950)(5350)ACCEPTABLE OPERATION See INSERT 4 - New

Fi g ure 3.1-1

FIGURE 3.1-1MINIMUM BAMT VOLUME vs STORED BORIC ACID CONCENTRATION400050006000700080009000100003.003.103.203.303.403.50STORED BAMT CONC (wt % Boric Acid, and PPM Boron)MINIMUM BAMT VOLUME (gallons) ACCEPTABLEOPERATIONUNACCEPTABLEOPERATION(8750)(8250)(7550)(6119 PPM)(5594 PPM)(5420 PPM)(5769 PPM)(5944 PPM)(5245 PPM)

ST. LUCIE - UNIT 2 3/4 1-15 Amendment No. 40 INSERT 4 ST. LUCIE - UNIT 2 3/4 1-24 Amendment No. 8 , 38 , 158 REACTIVITY CONTROL SYSTEMS CEA DROP TIME

LIMITING CONDITION FOR OPERATION

3.1.3.4 The individual full-length (shutdown and regulating) CEA drop time, from a fully withdrawn position, shall be less than or equal to 3.2 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90% insertion position with:

a. T avg greater than or equal to 515 F, and b. All reactor coolant pumps operating.

APPLICABILITY

MODES 1 and 2.

ACTION: a. With the drop time of any full-length CEA determined to exceed the above limit:

1. If in MODE 1 or 2, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or

2. If in MODE 3, 4, or 5, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal and installation of the reactor vessel head,

b. For specifically affected individuals CEAs following any main- tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At least once per 18 months.

3.25 individual ST. LUCIE - UNIT 23/4 2-14Amendment No. 89, 145POWER DISTRIBUTION LIMITSDNB PARAMETERSLIMITING CONDITION FOR OPERATION 3.2.5The following DNB-related parameters shall be maintained within the limits shown onTable 3.2-2:a.Cold Leg Temperature b.Pressurizer Pressure c.Reactor Coolant System Total Flow Rate d.AXIAL SHAPE INDEXAPPLICABILITY: MODE 1.ACTION:With any of the above parameters exceeding its limit, restore the parameter to within its limit within2 hours or reduce THERMAL POWER to < 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .SURVEILLANCE REQUIREMENTS 4.2.5.1Each of the parameters of Table 3.2-2 shall be verified to be within their limits byinstrument readout at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .4.2.5.2The Reactor Coolant System total flow rate shall be determined to be within its limit bymeasurement* at least once per 18 months. *Not required to be performed until THERMAL POWER is > 80% of RATED THERMAL POWER.

of the COLR:

DNB-related

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% per minute of RATED THERMAL POWER or a THERMAL

POWER step increase of greater than 10% of RATED THERMAL POWER.

- - 90%

ST. LUCIE - UNIT 23/4 2-15Amendment No. 8 , 92, 131 , 138, 145TABLE 3.2-2DNB MARGINLIMITSPARAMETERFOUR REACTORCOOLANT PUMPSOPERATINGCold Leg Temperature (Narrow Range)Within the limits specified in theCOLR Table 3.2-2Pressurizer Pressure*Within the limits specified in theCOLR Table 3.2-2Reactor Coolant Flow Rate**> 335,000 gpm and > the limit specifiedin the COLR Table 3.2-2AXIAL SHAPE INDEXCOLR Figure 3.2-4 *Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATEDTHERMAL POWER or a THERMAL POWER step increase of greater than 10% of RATEDTHERMAL POWER.**Commencing with the startup for Cycle 16 and until the Combustion Engineering Model 3410Steam Generators are replaced, Reactor Coolant Flow Rate will also be limited in accordancewith Renewed Operating License Paragraph 3.A.

DELETED ST. LUCIE - UNIT 23/4 4-8Amendment No. 91, 110REACTOR COOLANT SYSTEMOPERATINGLIMITING CONDITION FOR OPERATION 3.4.2.2All pressurizer code safety valves shall be OPERABLE with a lift setting of> 2435.3 psig and < 2535.3 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 theinoperable valve to OPERABLE status within 15 minutes or be in HOTSTANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .b.With two or more pressurizer code safety valves inoperable, be in HOTSTANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN with all RCS cold legtemperatures at < 230

°°°

°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS 4.4.2.2Verify each pressurizer code safety valve is OPERABLE in accordance with theInservice 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 nominaloperating temperature and pressure.

2410.3 2560.3 ST. LUCIE - UNIT 23/4 4-25Amendment No. 44REACTOR COOLANT SYSTEM3/4.4.8 SPECIFIC ACTIVITYLIMITING CONDITION FOR OPERATION 3.4.8The specific activity of the primary coolant shall be limited to:a.Less than or equal to 1.0 microcurie/gram DOSE EQUIVALENTI-131, andb.Less than or equal to 100/ microcuries/gram.APPLICABILITY: MODES 1, 2, 3, 4 and 5.ACTION:MODES 1, 2 and 3

a.With the specific activity of the primary coolant greater than 1.0 microcurie/gram DOSE EQUIVALENT I-131 for more than 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months sduring one continuous time interval or exceeding the limit lineshown on Figure 3.4-1, be in at least HOT STANDBY with T avg lessthan 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .b.With the specific activity of the primary coolant greater than 100/ microcuries/ gram, be in at least HOT STANDBY with T avg lessthan 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .MODES 1, 2, 3, 4 and 5

With the specific activity of the primary coolant greater than1 microcurie/gram DOSE EQUIVALENT I-131 or greater than100/ microcuries/gram, perform the sampling and analysis require-ments of item 4 a) of Table 4.4-4 until the specific activity ofthe primary coolant is restored to within its limits.SURVEILLANCE REQUIREMENTS 4.4.8The specific activity of the primary coolant shall be determined to bewithin the limits by performance of the sampling and analysis program ofTable 4.4-4. *With T avg greater than or equal to 500

°F.E E E 518.9DOSEEQUIVALENTXE

-133 and INSERT 5

a. With the specific activity of the primary coolant > 1.0 Ci/gram DOSE EQUIVALENT I-131, verify DOSE EQUIVALENT I-131 is 60.0 Ci/gram once per four hours.

b. With the specific activity of the primary coolant > 1.0 Ci/gram DOSE EQUIVALENT I-131, but 60.0 Ci/gram DOSE EQUIVALENT I-131, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to

restore DOSE EQUIVALENT I-131 to within the 1.0 Ci/gram limit.

Specification 3.0.4 is not applicable.

c. With the specific activity of the primary coolant > 1.0 Ci/gram DOSE EQUIVALENT I-131 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time

interval, or > 60.0 Ci/gram DOSE EQUIVALENT I-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the primary coolant > 518.9 Ci/gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the

518.9 Ci/gram DOSE EQUIVALENT XE-133 limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the primary coolant > 518.9 Ci/gram DOSE EQUIVALENT XE-133 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous

time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

INSERT 5 ST. LUCIE - UNIT 23/4 4-27Amendment No. 25TABLE 4.4-4PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLEAND ANALYSIS PROGRAMTYPE OF MEASUREMENTAND ANALYSISSAMPLE AND ANALYSISFREQUENCYMODES IN WHICH SAMPLEAND ANALYSIS REQUIRED1.Gross Activity DeterminationAt least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s1, 2, 3, and 42.Isotopic Analysis for DOSEEQUIVALENT I-131 Concentration1 per 14 days 13.Radiochemical for Determination1 per 6 months*

14.Isotopic Analysis for IodineIncluding I-131, I-133, and I-135a)Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,whenever the specificactivity exceeds1 micro-Ci/gram, DOSE EQUIVALENT I-131or 100/ micro-Ci/gram, and1#, 2#, 3#, 4#, 5#b)One sample between2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months followinga THERMAL POWERchange exceeding 15% of the RATED THERMALPOWER within a 1-hourperiod.1, 2, 3 #Until the specific activity of the primary coolant system is restored within its limits.*Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactorwas last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

E E and DOSE EQUIVALENT XE-133 3.I-132, I-134, 1 per 7 da y s MINIMUM ST. LUCIE - UNIT 23/4 4-28FIGURE 3.4-1DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit VersusPercent of RATED THERMAL POWER with the Primary Coolant SpecificActivity > 1.0

µCi/gram Dose Equivalent I-131PERCENT OF RATED THERMAL POWER2030405060708090100 250 200 150 100 50 0DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT (u Ci/gm)UNACCEPTABLE OPERATIONACCEPTABLE OPERATION DELETED FIGURE 3.4-2ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST 0 500 1000 1500 2000 25000100200300400500 T C - INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, O F P RCS - INDICATED PRESSURIZER PRESSURE, PSI A 50 OF/HR*ISOTHERMAL*CORE CRITICAL*MIN. BOLTUP TEMP.

80 O F*544.3 PSIA*INSERVICE HYDROSTATIC TEST*ALLOWABLE HEATUP RATESRATE, OF/HR TEMP. LIMIT O F50 AT ALL TEMPERATURESFLANGE LIMITTEMPERATURE170 O FLOWESTSERVICETEMPERATURE160 O F* - Includes Instrument Uncertainty

ST. LUCIE - UNIT 2 3/4 4-31a Amendment No. 37, 46,112,154 47 FIGURE 3.4-3ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 55 EFPY, COOLDOWN AND INSERVICE TEST 0 500 1000 1500 2000 25000100200300400500 T C - INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, O F PRCS - INDICATED PRESSURIZER PRESSURE, PSI AMINIMUM BOLTUP TEMPERATURE 80 OF* ALLOWABLE COOLDOWN RATESRATE, OF/HR TEMP. LIMIT O F100 10575 < 105LOWESTSERVICETEMPERATURE 160 O FCOOLDOWNUP TO 100 OF/HR*544.3 PSIA*INSERVICE HYDROSTATIC TEST*COOLDOWN UP TO 75 OF/HR*FLANGE LIMIT TEMPERATURE170 o F* - Includes Instrument Uncertainty

ST. LUCIE - UNIT 2 3/4 4-31b Amendment No. 37, 46,112,154 47 ST. LUCIE - UNIT 2 3/4 4-37a Amendment No. 31 , 46 , 112 , 154 TABLE 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Cold Leg Temperature, F Operating Period, EFPY During Heatup During Cooldown < 55 < 246 < 224

TABLE 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP

Operating Period EFPY T cold , F During Heatup T cold , F During Cooldown < 55 80 132 °F Cold Leg Temperature °F 47 ST. LUCIE - UNIT 23/4 5-1Amendment No. 40, 58, 96 , 1003/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3/4.5.1 SAFETY INJECTION TANKS (SIT)LIMITING CONDITION FOR OPERATION 3.5.1Each Reactor Coolant System safety injection tank shall be OPERABLE with:a.The isolation valve open,b.A contained borated water volume of between 1420 and 1556 cubic feet,c.A boron concentration of between 1720 and 2100 ppm of boron, andd.A nitrogen cover-pressure of between 500 and 650 psig.APPLICABILITY: MODES 1, 2 and 3*.ACTION:a.With one SIT inoperable due to boron concentration not within limits, or due to aninability to verify the required water volume or nitrogen cover-pressure, restore theinoperable SIT to OPERABLE status with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at leastHOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within thefollowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .b.With one SIT inoperable due to reasons other than those stated in ACTION-a,restore the inoperable SIT to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be inat least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN withinthe following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS 4.5.1.1Each safety injection tank shall be demonstrated OPERABLE:a.At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:1.Verifying that the borated water volume and nitrogen cover-pressure in the tanks are within their limits, and2.Verifying that each safety injection tank isolation valve is open. *With pressurizer pressure greater than or equal to 1750 psia. When pressurizer pressure is lessthan 1750 psia, at least three safety injection tanks shall be OPERABLE, each with a minimumpressure of 235 psig and a maximum pressure of 650 psig and a contained water volume ofbetween 1250 and 1556 cubic feet with a boron concentration of between 1720 and 2100 ppmof boron. With all four safety injection tanks OPERABLE, each tank shall have a minimumpressure of 235 psig and a maximum pressure of 650 psig and a contained water volume of between 833 and 1556 cubic feet with a boron concentration of between 1720 and 2100 ppm ofboron.1900 and 2200 1900 and 2200 ST. LUCIE - UNIT 23/4 5-3Amendment No. 106 , 119EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.2Two independent Emergency Core Cooling System (ECCS) subsystems shall beOPERABLE with each subsystem comprised of:a.One OPERABLE high pressure safety injection pump, b.One OPERABLE low pressure safety injection pump, and c.An independent OPERABLE flow path capable of taking suction from therefueling water tank on a Safety Injection Actuation Signal and automaticallytransferring suction to the containment sump on a Recirculation ActuationSignal, andd.One OPERABLE charging pump.APPLICABILITY: MODES 1, 2, and 3*.ACTION:a.1.With one ECCS subsystem inoperable only because its associated LPSItrain is inoperable, restore the inoperable subsystem to OPERABLEstatus within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .2.With one ECCS subsystem inoperable for reasons other than conditiona.1., restore the inoperable subsystem to OPERABLE status within72 hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .b.In the event the ECCS is actuated and injects water into the Reactor CoolantSystem, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The currentvalue of the usage factor for each affected safety injection nozzle shall beprovided in this Special Report whenever its value exceeds 0.70. *With pressurizer pressure greater than or equal to 1750 psia.

3****** One ECCS subsystem charging pump shall satisfy the flow path requirements of Specification 3.1.2.2.a or 3.1.2.2.d. The second ECCS subsystem charging

pump shall satisfy the flow path requirements of Specification 3.1.2.2.b or

3.1.2.2.e.

ST. LUCIE - UNIT 23/4 5-5Amendment No. 91, 99, 106, 136EMERGENCY CORE COOLING SYSTEMSSURVEILLANCE REQUIREMENTS (continued) 2.A visual inspection of the containment sump and verifying that thesubsystem suction inlets are not restricted by debris and that the sumpcomponents (trash racks, screens, etc.) show no evidence of structuraldistress or corrosion.3.Verifying that a minimum total of 173 cubic feet of solid granular trisodiumphosphate dodecahydrate (TSP) is contained within the TSP storagebaskets.4.Verifying that when a representative sample of 70.5 + 0.5 grams of TSPfrom a TSP storage basket is submerged, without agitation, in 10.0 + 0.1gallons of 120 + 10

°°°°F borated water from the RWT, the pH of the mixedsolution is raised to greater than or equal to 7 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .f.At least once per 18 months, during shutdown, by:1.Verifying that each automatic valve in the flow path actuates to its correctposition on SIAS and/or RAS test signals.2.Verifying that each of the following pumps start automatically upon receiptof a Safety Injection Actuation Test Signal:a.High-Pressure Safety Injection pump.

b.Low-Pressure Safety Injection pump.3.Verifying that upon receipt of an actual or simulated Recirculation ActuationSignal: each low-pressure safety injection pump stops, each containmentsump 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 totaldeveloped 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 positionstop for the following ECCS throttle valves:1.During valve stroking operation or following maintenance on the valveand prior to declaring the valve OPERABLE when the ECCS subsystemsare required to be OPERABLE.

paths pumps c. Charging Pumps Delete underline "_"

ST. LUCIE - UNIT 2 3/4 5-8 Amendment No. 157 EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER TANK

LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with:

a. A minimum contained borated water volume 477,360 gallons, b. A boron concentration of between 1720 and 2100 ppm of boron, and

c. A solution temperature of between 55 F and 100 F. APPLICABILITY

MODES 1, 2, 3 and 4.

ACTION: With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN

within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.4 The RWT shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the contained borated water volume in the tank, and

2. Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature when the outside air temperature is less then 55F or greater than 100 F. 1900 and 2200 ST. LUCIE - UNIT 23/4 7-3Amendment No. 8 , 68, 110TABLE 3.7-2STEAM LINE SAFETY VALVES PER LOOPVALVE NUMBERLIFT SETTING (+ 1% to - 3%)

Header A Header Ba.8201 8205> 955.3 psig and < 995.3 psig b.8202 8206> 955.3 psig and < 995.3 psig c.8203 8207> 955.3 psig and < 995.3 psig d.8204 8208> 955.3 psig and < 995.3 psig e.8209 8213> 994.1 psig and < 1035.7 psig f.8210 8214> 994.1 psig and < 1035.7 psig g.8211 8215> 994.1 psig and < 1035.7 psig h.8212 8216> 994.1 psig and < 1035.7 psig

- +/-3% for valves a through d and +2%/-3% for valves e through h 1015.3 1046.1 42,500 ST. LUCIE - UNIT 23/4 8-6Amendment No. 39ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS (Continued) 4.Simulating a loss-of-offsite power by itself, and:a.Verifying deenergization of the emergency busses andload shedding from the emergency busses.b.Verifying the diesel starts on the auto-start signal,****energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connectedshutdown loads through the load sequencer and operatesfor greater than or equal to 5 minutes while itsgenerator is loaded with the shutdown loads. Afterenergization, the steady-state voltage and frequencyof the emergency busses shall be maintained at4160 + 420 volts and 60 + 1.2 Hz during this test.5.Verifying that on an ESF actuation test signal (withoutloss-of-offsite power) the diesel generator starts****on the auto-start signal and operates on standby forgreater than or equal to 5 minutes. The steady

-stategenerator voltage and frequency shall be 4160 + 420 voltsand 60 + 1.2 Hz within 10 seconds after the auto-startsignal; the generator voltage and frequency shall bemaintained within these limits during this test.

6.Simulating a loss-of-offsite power in conjunction with anESF actuation test signal, anda)Verifying deenergization of the emergency busses and load shedding from the emergency busses.b)Verifying the diesel starts on the auto-start signal,****energizes the emergency busses with permanentlyconnected loads within 10 seconds, energizes the auto-connected emergency (accident) loads throughthe load sequencer and operates for greater than orequal to 5 minutes while its generator is loaded withthe emergency loads. After energization, thesteady-state voltage and frequency of the emergencybusses shall be maintained at 4160 + 420 volts and60 + 1.2 Hz during this test. ****This test may be conducted in accordance with the manufacturer'srecommendations concerning engine prelube period.

210 0.6 INSERT 6 210 0.6

5. Verifying that on an ESF actuation test signal (without loss-of-offsite power) the diesel generator starts**** on the auto-start signal, and:

a) Within 10 seconds, generator voltage and frequency shall be 4160 +/- 420 volts and 60 +/- 1.2 Hz.

b) Operates on standby for greater than or equal to 5 minutes.

c) Steady-state generator voltage and frequency shall be 4160 +/- 210 volts and 60 +/- 0.6 Hz and shall be maintained throughout this test.

INSERT 6 ST. LUCIE - UNIT 23/4 8-7Amendment No. 39, 60, 78 , 89ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS (Continued) c)Verifying that all automatic diesel generator trips, except engineoverspeed and generator differential, are automatically bypassed uponloss of voltage on the emergency bus concurrent with a safety injectionactuation signal.7.Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**** During thefirst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a loadband of 3800 to 3985 kW

and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, thediesel generator shall be loaded within a load band of 3450 to 3685 kW

.The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2Hz within 10 seconds after the start signal; the steady-state generator voltageand frequency shall be maintained within these limits during this test.8.Verifying that the auto-connected loads to each diesel generator do notexceed the 2000-hour rating of 3935 kW.9.Verifying the diesel generator's capability to:a)Synchronize with the offsite power source while the generator is loadedwith its emergency loads upon a simulated restoration of offsite power.b)Transfer its load to the offsite power source, andc)Be restored to its standby status.10.Verifying that with the diesel generator operating in a test mode (connectedto its bus), a simulated safety injection signal overrides the test mode by(1) returning the diesel generator to standby operation and (2) automaticallyenergizes the emergency loads with offsite power.11.Verifying that the fuel transfer pump transfers fuel from each fuel storage tankto the engine-mounted tanks of each diesel via the installed cross connection lines. #This band is meant as guidance to avoid routine overloading of the engine. Variations in load inexcess of this band due to changing bus loads shall not invalidate this test.****This test may be conducted in accordance with the manufacturer's recommendations concerningengine prelube period.

INSERT 7

7. Verifying that the diesel operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s****.

a) Within 10 seconds, generator voltage and frequency shall be 4160 +/- 420 volts and 60 +/- 1.2 Hz.

b) Steady-state generator voltage and frequency shall be 4160 +/- 210 volts and 60 +/- 0.6 Hz and shall be maintained throughout this test.

c) During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a load band of 3800 to 3985 kW

, and d) During the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded within a load band of 3450 to 3685 kW

. INSERT 7 ST. LUCIE - UNIT 23/4 8-9Amendment No. 39, 78, 122ELECTRICAL POWER SYSTEMSA.C. SOURCESSHUTDOWNLIMITING CONDITION FOR OPERATION 3.8.1.2As a minimum, the following A.C. electrical power sources shall be OPERABLE:a.One circuit between the offsite transmission network and the onsite Class 1Edistribution system, andb.One diesel generator with:1.Two engine-mounted fuel tanks containing a minimum volume of200 gallons of fuel,2.A fuel storage system containing a minimum volume of 40,000 gallons offuel, and3.A fuel transfer pump.APPLICABILITY: MODES 5 and 6.ACTION:With less than the above minimum required A.C. electrical power sources OPERABLE,immediately suspend all operations involving CORE ALTERATIONS, operations involvingpositive reactivity additions that could result in loss of required SHUTDOWN MARGIN or boronconcentration, movement of irradiated fuel, or crane operation with loads over the fuel storagepool, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System through a greaterthan or equal to 3.58 square inch vent. In addition, when in MODE 5 with the reactor coolantloops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vesselflange, immediately initiate corrective action to restore the required sources to OPERABLEstatus as soon as possible.SURVEILLANCE REQUIREMENTS 4.8.1.2.1The above required A.C. electrical power sources shall be demonstratedOPERABLE by the performance of each of the Surveillance Requirements of4.8.1.1.1 and 4.8.1.1.2 (except for requirement 4.8.1.1.2a.5).

42,500 ST. LUCIE - UNIT 23/4 9-1Amendment No. 923/4.9 REFUELING OPERATIONS3/4.9.1 BORON CONCENTRATIONLIMITING CONDITION FOR OPERATION 3.9.1With the reactor vessel head closure bolts less than fully tensioned or with the headremoved, the boron concentration of all filled portions of the Reactor Coolant System andthe refueling cavity shall be maintained within the limit specified in the COLR.APPLICABILITY: MODE 6*.ACTION:With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue borationat greater than or equal to 40 gpm of a solution containing 1720 ppm boron or greater to restoreboron concentration to within limits.SURVEILLANCE REQUIREMENTS 4.9.1.1The boron concentration limit shall be determined prior to:a.Removing or unbolting the reactor vessel head, andb.Withdrawal of any full length CEA in excess of 3 feet from its fully inserted positionwithin the reactor pressure vessel.4.9.1.2The boron concentration of the reactor coolant system and the refueling canal shall bedetermined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . *The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the reactorvessel head closure bolts less than fully tensioned or with the head removed.

1900 ST. LUCIE - UNIT 23/4 9-12Amendment No. 101REFUELING OPERATIONS3/4.9.11 SPENT FUEL STORAGE POOLLIMITING CONDITION FOR OPERATION 3.9.11The Spent Fuel Pool shall be maintained with:a.The fuel storage pool water level greater than or equal to 23 ft over the top of irradiated fuel assemblies seated in the storage racks, andb.The fuel storage pool boron concentration greater than or equal to 1720 ppm.APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel storage pool.ACTION:a.With the water level requirement not satisfied, immediately suspend all movement of fuel assemblies and crane operations with loads in the fuel storageareas and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .b.With the boron concentration requirement not satisfied, immediately suspend all movement of fuel assemblies in the fuel storage pool and initiate action torestore fuel storage pool boron concentration to within the required limit.c.The provisions of Specification 3.0.3 are not applicable.SURVEILLANCE REQUIREMENTS 4.9.11The water level in the spent fuel storage pool shall be determined to be at least itsminimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.4.9.11.1Verify the fuel storage pool boron concentration is within limit at least once per 7 days.1900 ST. LUCIE - UNIT 23/4 10-1Amendment No. 263/4.10 SPECIAL TEST EXCEPTIONS3/4.10.1 SHUTDOWN MARGINLIMITING CONDITION FOR OPERATION 3.10.1The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may besuspended for measurement of CEA worth, MTC, and SHUTDOWN MARGIN providedreactivity equivalent to at least the highest estimated CEA worth is availablefor trip insertion from OPERABLE CEA(s).APPLICABILITY: MODES 2 and 3*.ACTION:a.With any full-length CEA not fully inserted and with less than theabove reactivity equivalent available for trip insertion, immedi-ately initiate and continue boration at greater than or equal to40 gpm of a solution containing greater than or equal to 1720 ppmboron or its equivalent until the SHUTDOWN MARGIN required bySpecification 3.1.1.1 is restored.b.With all full-length CEAs inserted and the reactor subcriticalby less than the above reactivity equivalent, immediately initiateand continue boration at greater than or equal to 40 gpm of asolution containing greater than or equal to 1720 ppm boron or itsequivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.SURVEILLANCE REQUIREMENTS 4.10.1.1The position of each full-length CEA required either partially orfully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .4.10.1.2Each CEA not fully inserted shall be demonstrated capable of fullinsertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the SHUTDOWN MARGIN to less than the limits ofSpecification 3.1.1.1. *Operation in MODE 3 shall be limited to 6 consecutive hours.

1900 ST. LUCIE - UNIT 23/4 11-15Amendment No. 13RADIOACTIVE EFFLUENTSGAS STORAGE TANKSLIMITING CONDITION FOR OPERATION 3.11.2.6The quantity of radioactivity contained in each gas storage tankshall be limited to less than or equal to 285,000 curies noble gases(considered as Xe-133).APPLICABILITY: At all times.ACTION:a.With the quantity of radioactive material in any gas storage tankexceeding the above limit, immediately suspend all additions ofradioactive material to the tank.b.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.SURVEILLANCE REQUIREMENTS 4.11.2.6The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per24 hours when radioactive materials are being added to the tank when reactorcoolant system activity exceeds .

E 100 202,500 518.9 Ci/g ram DOSE EQUIVALENT XE-133.

ST. LUCIE - UNIT 25-4Amendment No. 7 , 95, 101 , 135, 138DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION5.5.1The meteorological tower shall be located as shown on Figure 5.1-1.5.6 FUEL STORAGECRITICALITY5.6.1a.The spent fuel storage racks are designed and shall be maintained with:1.A keff equivalent to less than 1.0 when flooded with unborated water, including a conservative allowance for biases and uncertainties asdescribed in Section 9.1 of the Updated Final Safety Analysis Report.2.A keff equivalent to less than or equal to 0.95 when flooded with watercontaining 520 ppm boron, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.3.A nominal 8.96 inch center-to-center distance between fuel assembliesplaced in the spent fuel pool storage racks and a nominal 8.80 inch center-to-center distance between fuel assemblies placed in the cask pit storage rack.4.The cask pit storage rack shall contain neutron absorbing material (Boral)between stored fuel assemblies when installed in the spent fuel pool.b.Fuel placed in Region I of the spent fuel storage racks shall be stored in aconfiguration that will assure compliance with 5.6.1 a.1 and 5.6.1 a.2, above, withthe following considerations:1.Fresh fuel shall have a nominal average U-235 enrichment of less than orequal to 4.5 weight percent.2.The reactivity effect of CEAs placed in fuel assemblies may be considered.3.The reactivity equivalencing effects of burnable absorbers may beconsidered.4.The reactivity effects of fuel assembly burnup and decay time may beconsidered as specified in Figures 5.6-1c through 5.6-1e.c.Fuel placed in Region II of the spent fuel storage racks shall be placed in aconfiguration that will assure compliance with 5.6.1 a.1 and 5.6.1 a.2, above, withthe following considerations:1.Fuel placed in the Region II spent fuel pool storage racks shall meet theburnup and decay time requirements specified in Figure 5.6-1a or 5.6-1b.Fuel placed in the Region II cask pit storage rack shall meet the burnuprequirements specified in Figure 5.6-1f.2.The reactivity effect of CEAs placed in fuel assemblies may be considered.3.The reactivity equivalencing effects of burnable absorbers may beconsidered.

500 8.965 INSERT 8

4. For storage of enriched fuel assemblies, requirements of Specification 5.6.1.a.1 and 5.6.1.a.2 shall be met by positioning fuel in the spent fuel pool storage racks consistent with the requirements of Specification 5.6.1.c or in configurations that have been shown to comply with Specifications 5.6.1.a.1 and 5.6.1.a.2 using the methodology as described in Section 9.1 of the Updated Final Safety Analysis Report. 5. Fissile material, not contained in a fuel assembly lattice, shall be stored in accordance with the requirements of Specifications 5.6.1.a.1 and 5.6.1.a.2.

b. The cask pit storage rack shall contain neutron absorbing material (Boral) between stored fuel assemblies when installed in the spent fuel pool.

c. Loading of spent fuel pool storage racks shall be controlled as described below.

1. The maximum initial planar average U-235 enrichment of any fuel assembly inserted in a spent fuel pool storage rack shall be less than or equal to 4.6 weight percent. 2. Fuel placed in Region 1 of the spent fuel pool storage racks shall comply with the storage pattern definitions of Figure 5.6-1 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions) 3. Fuel placed in Region 2 of the spent fuel pool storage racks shall comply with the storage pattern definitions or allowed special arrangement definitions of Figure 5.6-2 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions) 4. The 2x2 array of fuel assemblies that span the interface between Region 1 and Region 2 of the spent fuel pool storage racks shall comply with the storage pattern definitions of Figure 5.6-3 and the minimum burnup requirements as defined in Table 5.6-1. The allowed special arrangements in Region 2 as shown in Figure 5.6-2 shall not be placed adjacent to Region 1.

(See Specification 5.6.1.c.7 for exceptions) 5. Fuel placed in the cask pit storage rack shall comply with the storage pattern definitions of Figure 5.6-4 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions) 6. The same directional orientation for Metamic inserts is required for contiguous groups of 2x2 arrays where Metamic inserts are required. 7. Fresh or spent fuel in any allowed configuration may be replaced with non-fuel hardware, and fresh fuel in any allowed configuration may be replaced with a fuel rod storage basket containing fuel rod(s). Also, storage of Metamic inserts or control rods, without any fissile material, is acceptable in locations designated as completely water-filled cells.

INSERT 8 a maximum p lanar avera g e 4.6 1491 1716 ST. LUCIE - UNIT 25-4BAmendment No. 101FIGURE 5.6-1aRequired Fuel Assembly Burnup vs Initial Enrichment and Decay TimeRegion II, 1.3 w/o 0 10000 20000 30000 40000 500001.52.02.53.03.54.04.55.0 Initial U-235 Enrichment (w/o)

Fuel Assembly Average Burnup (MWD/MTU)

Acceptable Burnup 20 years 0 years 5 years 15 years 10 years Pages 5-4C through 5-4F (Amendment 101) and page 5-4G (Amendment 135) have been deleted from the Technical Specifications. The next page is 5-4h.

b ST. LUCIE - UNIT 25-4CAmendment No. 101FIGURE 5.6-1bRequired Fuel Assembly Burnup vs Initial Enrichment and Decay TimeRegion II, 1.5 w/o 0 10000 20000 30000 400001.52.02.53.03.54.04.55.0 Initial U-235 Enrichment (w/o)Fuel Assembly Average Burnup(MWD/MTU)

Acceptable Burnup 20 years 0 years 5 years 10 years 15 years ST. LUCIE - UNIT 25-4DAmendment No. 101FIGURE 5.6-1cRequired Fuel Assembly Burnup vs Initial Enrichment and Decay TimeRegion I, 1.4 w/o 0 10000 20000 30000 40000 50000 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)Fuel Assembly Average Burnup(MWD/MTU)

Acceptable Burnup 20 years 0 years 5 years 10 years 15 years ST. LUCIE - UNIT 25-4EAmendment No. 101FIGURE 5.6-1dRequired Fuel Assembly Burnup vs Initial Enrichment and Decay TimeRegion I, 1.82 w/o 0 5000 10000 15000 20000 25000 30000 350001.52.02.53.03.54.04.55.0 Initial U-235 Enrichment (w/o)

Fuel Assembly Average Burnup (MWD/MTU)

Acceptable Burnup 20 years 0 years 5 years 10 years 15 years ST. LUCIE - UNIT 25-4FAmendment No. 101FIGURE 5.6-1eRequired Fuel Assembly Burnup vs Initial EnrichmentRegion I, 2.82 w/o 5000 10000 15000 01.52.02.53.03.54.04.55.00 yearsAcceptable Burnup > - 484.92*E^3 + 4504.5*E^2 - 6086.6*E - 7783.1Fuel Assembly Burnup (MWD/MTU)Initial U-235 Enrichment (w/o)

Unacceptable Burnup DomainAcceptable Burnup Domain 4.5 4.0 3.5 3.0 2.5 2.0Initial Fuel Enrichment, wt% U-235 Bu = -16.3401 + 10.7922*E + 0.186428*E 2(Valid from 2% to 4.5% Enrichment) 40 35 30 25 20 15 10 5Average Fuel Burnup, MWD/KgU (G/TECH SPEC/U2-F5-4G-R0)

Allowable Storage Patterns (See Notes 1 and 2)

Pattern "A" Pattern "B" Pattern "C" See Definition 1 See Definition 2 See Definition 3 F WH 1 1 2 2 WH F CEA 1 1 WH 2 DEFINITIONS:

1. Allowable pattern is Fresh Fuel or fuel of lower reactivity checkerboarded with completely water-filled cells. Diagram is for illustration only, where F represents Fresh Fuel and WH represents a completely water-filled cell.

2. Allowable pattern is placement of fuel assembly type 1 or fuel of lower reactivity in each 2x2 array location with at least one full-length full-strength CEA or equivalent (5 absorber rods) placed in any cell. Minimum burnup for fuel assembly type 1 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

3. Allowable pattern is placement of fuel assembly type 2 or fuel of lower reactivity in three of the 2x2 array locations in combination with one completely water-filled cell. Minimum burnup for fuel assembly type 2 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only. NOTES: 1. The storage arrangements of fuel within a rack module may contain more than one pattern. There are no interface limitations within Region 1 between rack modules or within racks; however, each assembly must meet the burnup requirements of each 2x2 array that it resides within.

2. Open cells within any pattern are acceptable.

FIGURE 5.6-1 Allowable Region 1 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT 2 5-4h Amendment No. xxx NEW FIGURE 5.6-1 ALLOWED SPECIAL ARRANGEMENTS (See Notes 1 and 2) Fresh Fuel Assemblies in Region 2 Racks See Definition SR1 See Definition SR2 See Definition SR3

F F F F M M M M M M F F M M M F F M M M M M M M M M F F M M M F F M M M

= Fresh Fuel Assembly = Fuel Assembly with Metamic Insert F M = Empty Cell = Fuel Assembly per Definition

ALLOWABLE STORAGE PATTERNS (See Notes 1 and 2)

Pattern "D" Pattern "E" Pattern "F" See Definition 1 See Definition 2 See Definition 3 3 WH 4 4 5 5 3 4 5 4 3 5 Pattern "G" Pattern "H" See Definition 4 See Definition 5 7 CEA 7 6 6 CEA 6 7 7 6 CEA FIGURE 5.6-2 (Sheet 1 of 3) Allowable Region 2 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT 2 5-4i Amendment No. xxx NEW FIGURE 5.6-2 SHEET 1 of 3 DEFINITIONS for Figure 5.6-2

1. Allowable pattern is fuel assembly type 3 or fuel of lower reactivity in three of the 2x2 array locations in combination with one completely water-filled cell. Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

2. Allowable pattern is fuel assembly type 4 or fuel of lower reactivity in each of the 2x2 array locations with at least two Metamic inserts placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration

only.

3. Allowable pattern is fuel assembly type 5 or fuel of lower reactivity in each of the 2x2 array locations with at least one Metamic insert placed anywhere in the 2x2 array.

Minimum burnup for fuel assembly type 5 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration

only. 4. Allowable pattern is fuel assembly type 6 or fuel of lower reactivity in each of the 2x2 array locations with at least two full-length, full strength 5 finger CEAs or equivalent placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 6 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

5. Allowable pattern is fuel assembly type 7 in each of the 2x2 array locations with at least one full-length, full strength 5 finger CEA or equivalent placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 7 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

SR1. Allowable pattern is up to four fresh fuel assemblies or fuel of lower reactivity placed

in a 3x3 array in combination with Pattern "D" placed outside the 3x3 array. Fresh fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly of type 3 or fuel of lower reactivity shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

SR2. Allowable pattern is up to four fresh fuel assemblies or fuel of lower reactivity placed in a 3x3 array in combination with Pattern "E" placed outside the 3x3 array. Fresh fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly of type 4 or fuel of lower reactivity with a Metamic insert shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

FIGURE 5.6-2 (Sheet 2 of 3) Allowable Region 2 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT 2 5-4j Amendment No. xxx NEW FIGURE 5.6-2 SHEET 2 of 3 SR3. Allowable pattern is up to four fresh fuel assemblies or fuel of lower reactivity placed in a 3x3 array in combination with Pattern "F" placed outside the 3x3 array. Fresh fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly of type 5 or fuel of lower reactivity with a Metamic insert shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 5 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

NOTES 1. The storage arrangements of fuel within a rack module may contain more than one pattern. There are no interface limitations within Region 2 between rack modules or within racks. However, each assembly must meet the burnup requirements of each 2x2 array or allowed special arrangement that it resides within.

2. Open cells within any pattern are acceptable.

FIGURE 5.6-2 (Sheet 3 of 3) Allowable Region 2 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT 2 5-4k Amendment No. xxx NEW FIGURE 5.6-2 SHEET 3 of 3 Allowed Region 2 to Region 1 Fuel Alignments (See Note 1)

Region 2

Interface of Region 2 -----------------------------------------------------------------

3 3 3 3 Pattern "D" with Region 1 See Definition 1

R egion 1 3 WH 3 WH R

egion 2 4 4 4 Interface of Region 2 ------------------------------------------------------------------

Pattern "E" with Region 1

See Definition 1 Region 1

Region 2 Interface of Region 2 -----------------------------------------------------------------

4 4 5 4 4 4 5 5 5 Pattern "F" with Region 1

See Definition 1 Region 1 5 5 5 5

Region 2 Interface of Region 2 -----------------------------------------------------------------

6 6 CEA 6 6 CEA Pattern "G" with Region 1 See Definition 1 Region 1 6 CEA 6 6 CEA 6

R egion 2 Interface of Region 2 -----------------------------------------------------------------

7 7 7 7 Pattern "H" with Region 1 See Definition 1 R

egion 1 7 CEA 7 7 CEA 7 FIGURE 5.6-3 (Sheet 1 of 2)

Interface Requirements between Region 1 and Region 2 ST. LUCIE - UNIT 2 5-4l Amendment No. xxx NEW FIGURE 5.6-3 SHEET 1 of 2 DEFINITION:

1. Each 2x2 array that spans Region 1 and Region 2 shall match one of the Region 2 allowable storage patterns as defined by Specification 5.6.1.c.3. Any required Metamic inserts must be placed into the fuel assemblies in Region 2. Locations of water-filled cells or CEAs may be in either Region 1 or Region 2. For interface assemblies, the requirements of Specifications 5.6.1.c.2 and Specification 5.6.1.c.3 shall be followed within Region 1 and Region 2, respectively. The Diagrams are for illustration only.

NOTES: 1. Open cells within any pattern are acceptable.

FIGURE 5.6-3 (Sheet 2 of 2) Interface Requirements between Region 1 and Region 2

ST. LUCIE - UNIT 2 5-4m Amendment No. xxx NEW FIGURE 5.6-3 SHEET 2 of 2 Allowable Storage Patterns (See Notes 1 and 2)

Pattern "A" Pattern "B" See Definition 1 See Definition 2 F WH 8 8 WH F WH 8 DEFINITIONS:

1. Allowable pattern is Fresh Fuel or lower reactivity fuel checkerboarded with completely water-filled cells. Diagram is for illustration only, where F represents Fresh Fuel and WH represents a completely water-filled cell.

2. Allowable pattern is placement of fuel assembly type 8 or fuel of lower reactivity in three of the 2x2 array locations in combination with one completely water-filled cell in any location. Minimum burnup for fuel assembly type 8 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

NOTES: 1. The storage arrangements of fuel within a rack module may contain more than one pattern. There are no interface limitations within the cask pit storage rack, however, each assembly must meet the burnup requirements of each 2x2 array that it resides within.

2. Open cells within any pattern are acceptable.

FIGURE 5.6-4 Allowable Cask Pit Storage Rack Patterns

ST. LUCIE - UNIT 2 5-4n Amendment No. xxx NEW FIGURE 5.6-4 TABLE 5.6-1 Minimum Burnup Coefficients Coefficients Fuel Type Cooling Time (Years) A B C 1 0 -33.4237 25.6742 -1.6478 2 0 -25.3198 14.3200 -0.4042 3 0 -23.4150 16.2050 -0.5500 0 -33.6414 25.0670 -1.5551 2.5 -32.3764 23.9988 -1.5075 5 -30.9234 22.9382 -1.4372 10 -28.4951 21.1511 -1.3029 15 -27.2024 20.2802 -1.2479 4 20 -25.2009 18.6218 -1.0364 0 -24.8402 23.5991 -1.2082 2.5 -23.0170 21.6493 -1.0298 5 -21.9293 20.6257 -0.9730 10 -20.0813 19.0808 -0.9022 15 -19.5503 18.5429 -0.9129 5 20 -18.7485 17.7308 -0.8390 0 -32.4900 25.3077 -1.5518 2.5 -31.1598 23.9185 -1.4435 5 -29.2169 22.5424 -1.3274 10 -26.8886 20.6662 -1.1425 15 -25.5703 19.7629 -1.1129 6 20 -24.5754 18.9056 -1.0147 0 -24.6989 24.1660 -1.2578 2.5 -23.0399 22.3047 -1.0965 5 -21.3290 20.7413 -0.9613 10 -20.0836 19.4780 -0.8949 15 -19.2480 18.5880 -0.8685 7 20 -18.6424 18.1241 -0.8950 8 0 -47.5000 12.5000 0.0000 NOTES: 1. To qualify in a fuel type, the calculated burnup of a fuel assembly must exceed the "minimum burnup" determined for the "cooling time" and "maximum initial planar enrichment" of the fuel assembly. The "minimum burnup" for any fuel type

is determined from the following polynomial function:

BU = A + B*E + C*E 2 , where:

BU = Minimum Burnup (GWD/MTU)

E = Maximum Initial Planar Average Enrichment (weight percent U-235)

A, B, C = Coefficients for each fuel type

2. Interpolation between values of cooling time is not permitted.

ST. LUCIE - UNIT 2 5-4o Amendment No. xxx NEW TABLE 5.6-1 43.48 which ST. LUCIE - UNIT 26-20Amendment No. 13, 25, 61 , 92 , 105, 138ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (continued)6.9.1.9At least once every 5 years, an estimate of the actual population within 10 miles ofthe plant shall be prepared and submitted to the NRC.6.9.1.10At least once every 10 years, an estimate of the actual population within 50 miles ofthe plant shall be prepared and submitted to the NRC.6.9.1.11CORE OPERATING LIMITS REPORT (COLR)a.Core operating limits shall be established prior to each reload cycle, or prior toany remaining portion of a reload cycle, and shall be documented in the COLRfor the following:Specification 3.1.1.1Shutdown Margin - T avg Greater than 200

°°°°FSpecification 3.1.1.2Shutdown Margin - T avg Less Than or Equal to 200

°°°

°FSpecification 3.1.1.4Moderator Temperature CoefficientSpecification 3.1.3.1Movable Control Assemblies - CEA Position Specification 3.1.3.6Regulating CEA Insertion Limits Specification 3.2.1Linear Heat Rate Specification 3.2.3Total Integrated Radial Peaking Factors - F r TSpecification 3.2.5DNB Parameters Specification 3.9.1Refueling Operations - Boron Concentrationb.The analytical methods used to determine the core operating limits shall bethose previously reviewed and approved by the NRC, as described in thefollowing documents or any approved Revisions and Supplements thereto:1.WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear DesignSystem for Pressurized Water Reactor Cores," June 1988 (WestinghouseProprietary).2.NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of TurkeyPoint & St. Lucie Nuclear Plants," Florida Power & Light Company,January 1995.3.CENPD-199-P, Rev. 1-P-A, "C-E Setpoint Methodology: CE Local PowerDensity and DNB LSSS and LCO Setpoint Methodology for AnalogProtection Systems," January 1986.4.CENPD-266-P-A, "The ROCS and DIT Computer Code for NuclearDesign," April 1983.5.CENPD-275-P, Revision 1-P-A, "C-E Methodology for Core DesignsContaining Gadolinia-Urania Burnable Absorbers," May 1988.6.CENPD-188-A, "HERMITE: A Multi-Dimensional Space - Time KineticsCode for PWR Transients," July 1976., & Revision 1-P Supplement

1-P-A, April

1999.DELETED ST. LUCIE - UNIT 26-20aAmendment No. 92ADMINISTRATIVE CONTROLS (Continued)

CORE OPERATING LIMITS REPORT (COLR) (Continued)b.(Continued)7.CENPD-153-P, Rev. 1-P-A, "Evaluation of Uncertainty in the Nuclear PowerPeaking Measured by the Self-Powered, Fixed Incore Detector System,"May 1980.8.CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 1:CE Calculated Local Power Density and Thermal Margin/Low PressureLSSS for St. Lucie Unit 1," December 1979.9.CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 2:Combination of System Parameter Uncertainties in Thermal Margin Analysesfor St. Lucie Unit 1," January 1980.10.CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 3:CE Calculated Departure from Nucleate Boiling and Linear Heat RateLimiting Conditions for Operation for St. Lucie Unit 1," February 1980.11.CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for CalvertCliffs Units 1 and 2," December 1981.12.Letter, J.W. Miller (NRC) to J.R. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of CEN-123(F)-P (three parts) and CEN-191(B)-P).13.CEN-371(F)-P, "Extended Statistical Combination of Uncertainties,"

July 1989.14.Letter, J.A. Norris (NRC) to J.H. Goldberg (FPL), Docket No. 50-389, "St. Lucie Unit 2 - Change to Technical Specification Bases Sections

'2.1.1Reactor Core' and '3/4.2.5 DNB Parameters' (TAC No. M87722),"

March 14,1994 (Approval of CEN-371(F)-P).15.CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986.16.CENPD-162-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies withStandard Spacer Grids Part 1, Uniform Axial Power Distribution,"

April 1975.17.CENPD-207-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies withStandard Spacer Grids Part 2, Non-uniform Axial Power Distribution," December 1984.18.CENPD-206-P-A, "TORC Code, Verification and Simplified Modeling Methods," June 1981.

DELETED ST. LUCIE - UNIT 26-20cAmendment No. 92, 105ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR) (continued)b.(continued)34.Letter, J.A. Norris (NRC) to J.H. Goldberg (FPL), St. Lucie Unit 2 -Issuance of Amendment Re: Moderator Temperature Coefficient(TAC No. M82517), July 15, 1992.35.Letter, J.W. Williams, Jr. (FPL) to D.G. Eisenhut (NRC), St. Lucie UnitNo. 2, Docket No. 50-389, Proposed License Amendment, Cycle 2 Reload,L-84-148, June 4, 1984.36.Letter, J.R. Miller (NRC) to J.W. Williams, Jr. (FPL), Docket No. 50-389,Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16and SER), November 9, 1984 (Approval to Methodology contained in L-84-148).37.Letter, A.E. Scherer Enclosure 1-P to LD-82-001, CESEC-DigitalSimulation of a Combustion Engineering Nuclear Steam Supply System,December 1981.38.Safety Evaluation Report, CESEC Digital Simulation of a CombustionEngineering Steam Supply System (TAC No.: 01142), October 27, 1983.39.CENPD-282-P-A, Volumes 1, 2 and 3, and Supplement 1, TechnicalManual for the CENTS Code, February 1991, February 1991, October1991, and June 1993, respectively.40.CEN-121(B)-P, CEAW, Method of Analyzing Sequential Control ElementAssembly Group Withdrawal Event for Analog Protected Systems,November 1979 (NRC SER dated December 21, 1999, LetterK. N. Jabbour (NRC) to T.F. Plunkett (FPL), TAC No. MA4523).41.CEN-133(B), FIESTA, A One Dimensional, Two Group Space-TimeKinetics Code for Calculating PWR Scram Reactivities, November 1979(NRC SER dated December 21, 1999, Letter K. N. Jabbour (NRC) toT.F. Plunkett (FPL), TAC No. MA4523).42.CEN-348(B)-P-A, Supplement 1-P-A, Extended Statistical Combination ofUncertainties, January 1997.43.CEN-372-P-A, Fuel Rod Maximum Allowable Gas Pressure, May 1990.44.CENPD-183-A, C-E Methods for Loss of Flow Analysis, June 1984.45.CENPD-190-A, C-E Method for Control Element Assembly EjectionAnalysis, July 1976.

DELETED ST. LUCIE - UNIT 26-20dAmendment No. 92, 105 , 118, 138ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR) (continued)b.(continued)46.CENPD-199-P, Rev. 1-P-A, Supplement 2-P-A, "CE Setpoint Methodology,"

June 1998.47.CENPD-382-P-A, "Methodology for Core Designs Containing ErbiumBurnable Absorbers," August 1993.48.CEN-396(L)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of60 MWD/KG for St. Lucie Unit 2," November 1989 (NRC SER datedOctober 18, 1991, Letter J.A. Norris (NRC) to J.H. Goldberg (FPL), TAC No. 75947).49.CENPD-269-P, Rev. 1-P, "Extended Burnup Operation of CombustionEngineering PWR Fuel," July 1984.50.CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit2," December 1984 (NRC SER dated December 21, 1999, Letter K. N. Jabbour (NRC) to T.F. Plunkett (FPL), TAC No. MA4523).51.CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CESmall Break LOCA Evaluation Model," April 1998.52.CENPD-140-A, "Description of the CONTRANS Digital Computer Code forContainment Pressure and Temperature Transient Analysis," June 1976.53.CEN-365(L), "Boric Acid Concentration Reduction Effort, Technical Basesand Operational Analysis," June 1988 (NRC SER dated March 13, 1989, Letter J.A. Norris (NRC) to W.F. Conway (FPL), TAC No. 69325).54.DP-456, F.M. Stern (CE) to E. Case (NRC), dated August 19, 1974,Appendix 6B to CESSAR System 80 PSAR (NRC SER, NUREG-75/112, Docket No. STN 50-470, "NRC SER - Standard Reference System, CESSAR System 80," December 1975).55.CENPD-387-P-A, Revision 000, "ABB Critical Heat Flux Correlations forPWR Fuel," May 2000.56.CENPD-132, Supplement 4-P-A, "Calculative Methods for the CE NuclearPower Large Break LOCA Evaluation Model," March 2001.57.CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CESmall Break LOCA Evaluation Model," April 1998.58.CENPD-404-P-A, Rev. 0, "Implementation of ZIRLO TM Cladding Material inCE Nuclear Power Fuel Assembly Designs," November 2001.59.WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"July 1985.60.WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial OffsetControl; FQ Surveillance Technical Specification," February 1994.

DELETED DELETED ST. LUCIE - UNIT 2 6-20e Amendment No. 105 , 118 , 133 , 138 , 147 ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR)

(continued)

b. (continued)

61. WCAP-11397-P-A, (Proprietary), 'Revised Thermal Design Procedure," April 1989.

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

63. WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code," May 2003.

64. 30% SGTP PLA Submittal and the SER.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety

analysis are met.

d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle on the NRC.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection of the replacement SGs performed in accordance with Specification 6.8.4.l.1. The report shall include: a. The scope of inspections performed on each SG, b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, h. The effective plugging percentage for all plugging in each SG.

INSERT 9

64. Letter, W. Jefferson Jr. (FPL) to Document Control Desk (USNRC), "St. Lucie Unit 2 Docket No. 50-389: Proposed License Amendment WCAP-9272 Reload Methodology and Implementing 30% Steam Generator Tube Plugging Limit," L- 2003-276, December 2003 (NRC SER dated January 31, 2005, Letter B.T.

Moroney (NRC) to J.A. Stall (FPL), TAC No. MC1566).

65. WCAP-14882-P-A, Rev. 0, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," April 1999.

66. WCAP-7908-A, Rev. 0, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a UO2 Fuel Rod," December 1989.

67. WCAP-7979-P-A, Rev. 0, "TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," January 1975.

68. WCAP-7588, Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Methods,"

January 1975.

INSERT 9 St. Lucie Unit 2 L-2011-021 Docket No. 50-389 Attachment 3 St. Lucie Unit 2 EPU LAR Renewed Facility Operating License and Technical Specifications Markups and Clean Pages

ATTACHMENT 3 LICENSE AMENDMENT REQUEST EXTENDED POWER UPRATE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS CLEAN PAGES FLORIDA POWER & LIGHT ST. LUCIE UNIT 2

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 CFR 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. XXX are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16Revised by letter dated XXXXXX XX, XXXX

ST. LUCIE - UNIT 2 I Amendment No. 13 , 92 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 ALTERATION....................................................................................................1-2

1.9a CORE OPERATING LIMITS REPORT (COLR)...........................................................1-2

1.10 DOSE EQUIVALENT I-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................................................................................................1-3

1.16 DELETED.....................................................................................................................1-4 1.17 MEMBER(S) OF THE PUBLIC.....................................................................................1-4

1.18 OFFSITE DOSE CALCULATION MANUAL (ODCM)...................................................1-4

1.19 OPERABLE - OPERABILITY.......................................................................................1-4 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 XXI Amendment No. 8 , 53 , 73 , 92 , 112 , 117 , 154 INDEX LIST OF FIGURES FIGURE PAGE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING...........................................................2-3 2.2-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 1 (FRACTION OF RATED THERMAL POWER VERSUS QR 2)..................................2-7 2.2-2 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 2 (QR 2 VERSUS Y 1).....................................................................................................2-8 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 (Y 1 , VERSUS A 1)......................................................................................................2-9 2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR 1)................................2-10 3.1-1 MINIMUM BAMT VOLUME VS STORED BORIC ACID CONCENTRATION...3/4 1-15

3.1-1a DELETED....................................................................................................................... 3.1-2 DELETED....................................................................................................................... 3.2-1 DELETED.......................................................................................................................

3.2-2 DELETED.......................................................................................................................

3.2-3 DELETED.......................................................................................................................

4.2-1 DELETED....................................................................................................................... 3.2-4 DELETED....................................................................................................................... 3.4-1 DELETED.......................................................................................................................

3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE- TEMPERATURE LIMITS FOR 47 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST...........................................................................................3/4 4-31a ST. LUCIE - UNIT 2 XXII Amendment No. 8 , 29 , 53 , 101 , 112 , 117 , 135 , 154 INDEX LIST OF FIGURES (continued)

FIGURE PAGE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 47 EFPY, COOLDOWN AND INSERVICE TEST.............................................................................................................3/4 4-31b

3.4-4 DELETED.......................................................................................................3/4 4-32

4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..............................3/4 7-25

5.1-1 SITE AREA MAP....................................................................................................5-2

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5.6-1b DELETED....................................................................................................................

5.6-1c DELETED....................................................................................................................

5.6-1d DELETED....................................................................................................................

5.6-1e DELETED....................................................................................................................

5.6-1f DELETED.................................................................................................................... 5.6-1 ALLOWABLE REGION 1 STORAGE PATTERNS AND FUEL ARRANGEMENTS................................................................................................5-4h 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 1 of 3)..........................................................................5-4i 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 2 of 3)..........................................................................5-4j 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 3 of 3).........................................................................5-4k 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 1 of 2).........................................................................................................5-4l 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 2 of 2).......................................................................................................5-4m 5.6-4 ALLOWABLE CASK PIT STORAGE RACK PATTERNS.....................................5-4n 6.2-1 DELETED...............................................................................................................6-3

6.2-2 DELETED...............................................................................................................6-4

ST. LUCIE - UNIT 2 XXIII Amendment No. 8 , 73 , 74 , 86 , 104 INDEX LIST OF TABLES TABLE PAGE 1.1 FREQUENCY NOTATION......................................................................................1-8

1.2 OPERATIONAL MODES........................................................................................1-9 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS..........2-4

3.1-1 DELETED ................................................................................................................... 3.2-1 DELETED.......................................................................................................3/4 2-11

3.2-2 DELETED.......................................................................................................3/4 2-15

3.3-1 REACTOR PROTECTIVE INSTRUMENTATION.............................................3/4 3-2 3.3-2 DELETED....................................................................................................................

4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................................................................3/4 3-8

3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................................................................3/4 3-12

3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES.............................................................3/4 3-17 3.3-5 DELETED....................................................................................................................

4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................3/4 3-22

3.3-6 RADIATION MONITORING INSTRUMENTATION.........................................3/4 3-25

4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................................................................................3/4 3-28

3.3-8 DELETED.................................................................................................................... 4.3-5 DELETED....................................................................................................................

5.6-1 MINIMUM BURNUP COEFFICIENTS..................................................................5-4o

ST. LUCIE - UNIT 2 1-3 Amendment No. 105 , 137 , 147 , 152 DEFINITIONS DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." DOSE EQUIVALENT XE-133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (Ci/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-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 III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. 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. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED 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).

ST. LUCIE - UNIT 2 1-4 Amendment No. 16 , 31 , 46 , 61 , 63 DEFINITIONS 1.16 Deleted 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 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 6.8.4 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 1-5 Amendment No. 9 , 13 , 61 , 137 , 147 DEFINITIONS PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component

body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING

1.24 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER

1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3020 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME

1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power to the CEA drive mechanism is interrupted. 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. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRITY

1.28 SHIELD BUILDING INTEGRITY shall exist when:

a. Each door is closed except when the access opening is being used for normal transit entry and exit;

b. The shield building ventilation system is in compliance with Specification 3.6.6.1, and

c. The sealing mechanism associated with each penetration (e.g.,

welds, bellows or O-rings) is OPERABLE.

FRACTION OF RATED THERMAL POWERVESSEL INLET TEMPERATURE (F) FIGURE 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING

ST. LUCIE - UNIT 2 2-4 Amendment No. 8 , 23 , 60 TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 1. Manual Reactor Trip Not Applicable Not Applicable

2. Variable Power Level - High (1) Four Reactor Coolant Pumps

Operating

< 9.61% above THERMAL POWER, with a minimum setpoint of 15% of RATED THERMAL POWER, and a maximum of < 107.0% of RATED THERMAL POWER.

< 9.61% above THERMAL POWER, and a minimum setpoint of 15% of RATED THERMAL POWER and a maximum of < 107.0% of RATED THERMAL POWER.

3. Pressurizer Pressure - High < 2370 psia

< 2374 psia 4. Thermal Margin/Low Pressure (1) Four Reactor Coolant Pumps Operating Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4.

Minimum value of 1900 psia. Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4.

Minimum value of 1900 psia.

5. Containment Pressure - High <

3.0 psig

< 3.1 psig 6. Steam Generator Pressure - Low

> 626.0 psia (2) > 621.0 psia (2) 7. Steam Generator Pressure (1) Difference - High (Logic in TM/LP Trip Unit)

< 120.0 psid

< 132.0 psid 8. Steam Generator Level - Low (6), (7) > 35.0%(3) > 34.1%(3)

ST. LUCIE - UNIT 2 2-5 Amendment No. 8 , 60 , 131 , 138 TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 9. Local Power Density - High (5) Operating Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2 Trip setpoint adjusted to

not exceed the limit lines of Figures 2.2-1 and 2.2-2.

10. Loss of Component Cooling Water to Reactor Coolant Pumps - Low

> 636 gpm**

> 636 gpm

11. Reactor Protection System Logic Not Applicable Not Applicable

12. Reactor Trip Breakers Not Applicable Not Applicable 13. Rate of Change of Power - High (4) < 2.49 decades per minute

< 2.49 decades per minute

14. Reactor Coolant Flow - Low (1) > 95.4% of minimum Reactor Coolant flow with four pumps operating*

> 94.9% of minimum Reactor Coolant flow with four

pumps operating* 15. Loss of Load (Turbine) Hydraulic Fluid Pressure - Low (5) > 800 psig

> 800 psig

For minimum reactor coolant flow with four pumps operating, refer to COLR Table 3.2-2. ** 10-minute time delay after relay actuation.

ST. LUCIE - UNIT 2 2-6 Amendment No. 98 TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATION (1) Trip may be manually bypassed below 0.5% of RATED THERMAL POWER during testing pursuant to Special Test Exception 3.10.3; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is greater than or equal to 0.5% of RATED THERMAL POWE R. (2) Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig.

(3) % of the narrow range steam generator level indication.

(4) Trip may be bypassed below 10

-4% and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is >

10-4% and Power Range Neutron Flux power < 15% of RATED THERMAL POWER.

(5) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when Power Range Neutron Flux power is greater than or equal to 15% of RATED THERMAL POWER.

(6) If the as-found channel setpoint is either outside its predefined as-found acceptance criteria band or is not conservative with respect to the Allowable Value, then the channel shall be declared inoperable and shall be evaluated to verify that it is functioning as required before returning the channel to service. (7) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the Trip Setpoint or a value that is more conservative than the Trip Setpoint, otherwise that channel shall not be returned to OPERABLE status. The Trip Setpoint and the methodology used to determine the Trip Setpoint, the as-found acceptance criteria band, and the as-left acceptance criteria are specified in the UFSAR.

ST. LUCIE - UNIT 2 3/4 1-1 Amendment No. 25 , 89 , 105 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tavg GREATER THAN 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be within the limits specified in the COLR.

APPLICABILITY

MODES 1, 2*, 3 and 4.

ACTION: With the SHUTDOWN MARGIN outside the COLR limits, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1900 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be within the COLR limits:

a. Within one hour after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable. If the inoperable CEA is not fully inserted, and is immovable as a result of excessive friction or mechanical interference or is known to be untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable CEA(s).

b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.

c. When in MODE 2 with Keff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

ST. LUCIE - UNIT 2 3/4 1-3 Amendment No. 8 , 25 , 105 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tavg LESS THAN OR EQUAL TO 200 F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be within the limits specified in the COLR.

APPLICABILITY

MODE 5.

ACTION: With the SHUTDOWN MARGIN outside the COLR limits, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1900 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be within the COLR limits:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. CEA position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration.

c. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , when the Reactor Coolant System is drained below the hot leg centerline, by consideration of the factors in 4.1.1.2b and by verifying at least two charging pumps are rendered inoperable by racking out their motor circuit breakers.

ST. LUCIE - UNIT 2 3/4 1-8 Amendment No. 8 , 25 , 40 , 105 REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. One flow path from the boric acid makeup tank(s) with the tank meeting Specification 3.1.2.8 part a) or b), via a boric acid makeup pump through a charging pump to the Reactor Coolant System.

b. One flow path from the boric acid makeup tank(s) with the tank meeting Specification 3.1.2.8 part a) or b), via a gravity feed valve through a charging pump to the Reactor Coolant System.

c. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System.

OR At least two of the following three boron injection flow paths shall be OPERABLE:

d. One flow path from each boric acid makeup tank with the combined tank contents meeting Specification 3.1.2.8 c), via both boric acid makeup pumps through a charging pump to the Reactor Coolant System.

e. One flow path from each boric acid makeup tank with the combined tank contents meeting Specification 3.1.2.8 c), via both gravity feed valves through a charging pump to the Reactor Coolant System.

f. The flow path from the refueling water storage tank, via a charging pump to the Reactor Coolant System.

APPLICABILITY

MODES 1, 2, 3 and 4.

ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ST. LUCIE - UNIT 2 3/4 1-13 Amendment No. 40 , 122 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN

LIMITING CONDITION FOR OPERATION

3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:

a. One boric acid makeup tank with a minimum borated water volume of 3550 gallons of 3.1 to 3.5 weight percent boric acid (5420 to 6119 ppm boron).

b. The refueling water tank with:

1. A minimum contained borated water volume of 125,000 gallons,

2. A minimum boron concentration of 1900 ppm, and

3. A solution temperature between 40 F and 120F.

APPLICABILITY

MODES 5 and 6.

ACTION:

With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes*.

SURVEILLANCE REQUIREMENTS

4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the boron concentration of the water,

2. Verifying the contained borated water volume of the tank, and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature when it is the source of borated water and the outside air temperature is outside the range of 40 F and 120F. .

c. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the Reactor Auxiliary Building air temperature is less than 55F, by verifying that the boric acid makeup tank solution temperature is greater than 55 F when that boric acid makeup tank is required to be OPERABLE.

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

ST. LUCIE - UNIT 2 3/4 1-14 Amendment No. 8 , 25 , 40 , 105 , 157 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 At least two of the following four borated water sources shall be OPERABLE:

a. Boric Acid Makeup Tank 2A in accordance with Figure 3.1-1.

b. Boric Acid Makeup Tank 2B in accordance with Figure 3.1-1.

c. Boric Acid Makeup Tanks 2A and 2B with a minimum combined contained borated water volume in accordance with Figure 3.1-1.

d. The refueling water tank with:

1. A minimum contained borated water volume of 477,360 gallons,

2. A boron concentration of between 1900 and 2200 ppm of boron, and

3. A solution temperature of between 55 F and 100 F.

APPLICABILITY

MODES 1, 2, 3 and 4.

ACTION:

a. With the above required boric acid makeup tank(s) inoperable, restore the tank(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 200F; restore the above required boric acid makeup tank(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.8 At least two required borated water sources shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the boron concentration in the water and

2. Verifying the contained borated water volume of the water source.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature when the outside air temperature is outside the range of 55F and 100 F. c. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the Reactor Auxiliary Building air temperature is less than 55F, by verifying that the boric acid makeup tank solution is greater than 55 F.

ST. LUCIE - UNIT 2 3/4 1-15 Amendment No. 40 FIGURE 3.1-1 MINIMUM BAMT VOLUME vs STORED BORIC ACID CONCENTRATION

ST. LUCIE - UNIT 2 3/4 1-24 Amendment No. 8 , 38 , 158 REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION

3.1.3.4 The individual full-length (shutdown and regulating) CEA drop time, from a fully withdrawn position, shall be less than or equal to 3.25 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90% insertion position with:

a. T avg greater than or equal to 515 F, and b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION: a. With the drop time of any full-length CEA determined to exceed the above limit:

1. If in MODE 1 or 2, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or

2. If in MODE 3, 4, or 5, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal and installation of the reactor vessel head,

b. For specifically affected individual CEAs following any main- tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At least once per 18 months.

ST. LUCIE - UNIT 2 3/4 2-14 Amendment No. 89 , 145 POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-2 of the COLR:

a. Cold Leg Temperature

b. Pressurizer Pressure*

c. Reactor Coolant System Total Flow Rate

d. AXIAL SHAPE INDEX

APPLICABILITY

MODE 1.

ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to < 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the DNB-related parameters shall be verified to be within their limits by instrument readout at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement** at least once per 18 months.

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% per minute of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER. ** Not required to be performed until THERMAL POWER is > 90% of RATED THERMAL POWER.

ST. LUCIE - UNIT 2 3/4 2-15 Amendment No. 8 , 92 , 131 , 138 , 145

DELETED ST. LUCIE - UNIT 2 3/4 4-8 Amendment No. 91 , 110 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 > 230F. 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

230 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 4-25 Amendment No. 44 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. Less than or equal to 1.0 microcurie/gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 518.9 microcuries/gram DOSE EQUIVALENT XE-133.

APPLICABILITY

MODES 1, 2, 3, and 4 ACTION: a. With the specific activity of the primary coolant > 1.0 Ci/gram DOSE EQUIVALENT I-131, verify DOSE EQUIVALENT I-131 is 60.0 Ci/gram once per four hours.

b. With the specific activity of the primary coolant > 1.0 Ci/gram DOSE EQUIVALENT I-131, but 60.0 Ci/gram DOSE EQUIVALENT I-131, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT I-131 to within the 1.0 Ci/gram limit. Specification 3.0.4 is not applicable.

c. With the specific activity of the primary coolant > 1.0 Ci/gram DOSE EQUIVALENT I-131 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or > 60.0 Ci/gram DOSE EQUIVALENT I-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With the specific activity of the primary coolant > 518.9 Ci/gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 518.9 Ci/gram DOSE EQUIVALENT XE-133 limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the primary coolant > 518.9 Ci/gram DOSE EQUIVALENT XE-133 for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

ST. LUCIE - UNIT 2 3/4 4-27 Amendment No. 25 TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS MINIMUM FREQUENCY MODES IN WHICH SAMPLE AND ANALYSIS REQUIRED

1. DOSE EQUIVALENT XE-133 Determination 1 per 7 days 1, 2, 3, and 4 2. Isotopic Analysis for DOSE EQUIVALENT I-131 Concentration 1 per 14 days 1 3. Isotopic Analysis for Iodine Including I-131, I-132, I-133, I-134, and I-135 a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds 1 micro-Ci/gram, DOSE EQUIVALENT I-131, and 1#, 2#, 3#, and 4#

b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period. 1, 2, 3 # Until the specific activity of the primary coolant system is restored within its limits.

ST. LUCIE - UNIT 2 3/4 4-28 Amendment No. xxx

DELETED ST. LUCIE - UNIT 2 3/4 4-31a Amendment No. 37 , 46 , 112 , 154 FIGURE 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 47 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST

ST. LUCIE - UNIT 2 3/4 4-31b Amendment No. 37 , 46 , 112 , 154 FIGURE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 47 EFPY, COOLDOWN, AND INSERVICE TEST

ST. LUCIE - UNIT 2 3/4 4-37a Amendment No. 31 , 46 , 112 , 154 TABLE 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Cold Leg Temperature, F Operating Period, EFPY During Heatup During Cooldown < 47 < 246 < 224

TABLE 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Cold Leg Temperature, F Operating Period EFPY During Heatup During Cooldown < 47 80 132 ST. LUCIE - UNIT 2 3/4 5-1 Amendment No. 40 , 58 , 96 , 100 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS (SIT)

LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System safety injection tank shall be OPERABLE with:

a. The isolation valve open,

b. A contained borated water volume of between 1420 and 1556 cubic feet,

c. A boron concentration of between 1900 and 2200 ppm of boron, and

d. A nitrogen cover-pressure of between 500 and 650 psig.

APPLICABILITY

MODES 1, 2 and 3*.

ACTION: a. With one SIT inoperable due to boron concentration not within limits, or due to an inability to verify the required water volume or nitrogen cover-pressure, restore the inoperable SIT to OPERABLE status with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one SIT inoperable due to reasons other than those stated in ACTION-a, restore the inoperable SIT to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each safety injection tank shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1. Verifying that the borated water volume and nitrogen cover-pressure in the tanks are within their limits, and

2. Verifying that each safety injection tank isolation valve is open.

With pressurizer pressure greater than or equal to 1750 psia. When pressurizer pressure is less than 1750 psia, at least three safety injection tanks shall be OPERABLE, each with a minimum pressure of 235 psig and a maximum pressure of 650 psig and a contained water volume of between 1250 and 1556 cubic feet with a boron concentration of between 1900 and 2200 ppm of boron. With all four safety injection tanks OPERABLE, each tank shall have a minimum pressure of 235 psig and a maximum pressure of 650 psig and a contained water volume of between 833 and 1556 cubic feet with a boron concentration of between 1900 and 2200 ppm of boron.

ST. LUCIE - UNIT 2 3/4 5-3 Amendment No. 106 , 119 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE high pressure safety injection pump,

b. One OPERABLE low pressure safety injection pump, and

c. An independent OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal, and

d. One OPERABLE charging pump*.

APPLICABILITY

MODES 1, 2, and 3**.

ACTION: a. 1. With one ECCS subsystem inoperable only because its associated LPSI train is inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. With one ECCS subsystem inoperable for reasons other than condition a.1., restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

One ECCS subsystem charging pump shall satisfy the flow path requirements of Specification 3.1.2.2.a and 3.1.2.2.d. The second ECCS subsystem charging pump shall satisfy the flow path requirements of Specification 3.1.2.2.b or 3.1.2.2.e. ** With pressurizer pressure greater than or equal to 1750 psia.

ST. LUCIE - UNIT 2 3/4 5-5 Amendment No. 91 , 99 , 106 , 136 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 +

10F 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f. At least once per 18 months, 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 5-8 Amendment No. 15 7 EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with:

a. A minimum contained borated water volume 477,360 gallons, b. A boron concentration of between 1900 and 2200 ppm of boron, and

c. A solution temperature of between 55 F and 100 F. APPLICABILITY

MODES 1, 2, 3 and 4.

ACTION: With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.4 The RWT shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1. Verifying the contained borated water volume in the tank, and

2. Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWT temperature when the outside air temperature is less then 55F or greater than 100 F.

ST. LUCIE - UNIT 2 3/4 7-3 Amendment No. 8 , 68 , 110 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING* Header A Header B a. 8201 8205 > 955.3 psig and <

1015.3 psig b. 8202 8206 > 955.3 psig and <

1015.3 psig c. 8203 8207 > 955.3 psig and <

1015.3 psig d. 8204 8208 > 955.3 psig and <

1015.3 psig e. 8209 8213 > 994.1 psig and <

1046.1 psig f. 8210 8214 > 994.1 psig and <

1046.1 psig g. 8211 8215 > 994.1 psig and <

1046.1 psig h. 8212 8216 > 994.1 psig and <

1046.1 psig

+/-3% for valves a through d and +2%/-3% for valves e through h ST. LUCIE - UNIT 2 3/4 8-1 Amendment No. 25 , 39 , 78 , 115 , 123 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and

b. Two separate and independent diesel generators, each with:

1. Two separate engine-mounted fuel tanks containing a minimum volume of 200 gallons of fuel each,

2. A separate fuel storage system containing a minimum volume of 42,500 gallons of fuel, and

3. A separate fuel transfer pump.

APPLICABILITY

MODES 1, 2, 3, and 4.

ACTION: a. With one offsite circuit of 3.8.1.1.a inoperable, except as provided in Action f. below, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Restore the offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one diesel generator of 3.8.1.1.b inoperable, demonstrate the OPERABILITY of the A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and if the EDG became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE EDG by performing Surveillance Requirement 4.8.1.1.2a.4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless it can be confirmed that the cause of the inoperable EDG does not exist on the remaining EDG*; restore the diesel generator to OPERABLE status within 14 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Additionally, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the discovery of concurrent inoperability of required redundant feature(s) (including the steam driven auxiliary feed pump in MODE 1, 2, and 3), declare required feature(s) supported by the inoperable EDG inoperable if its redundant required feature(s) is inoperable.

If the absence of any common-cause failure cannot be confirmed, this test shall be completed regardless of when the inoperable EDG is restored to OPERABILITY.

ST. LUCIE - UNIT 2 3/4 8-6 Amendment No. 39 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

4. Simulating a loss-of-offsite power by itself, and:

a. Verifying deenergization of the emergency busses and load shedding from the emergency busses.

b. Verifying the diesel starts on the auto-start signal,**** energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +

210 volts and 60 + 0.6 Hz during this test.

5. Verifying that on an ESF actuation test signal (without loss-of-offsite power) the diesel generator starts****

on the auto-start signal, and:

a) Within 10 seconds, generator voltage and frequency shall be 4160 +/-

420 volts and 60 +/- 1.2 Hz.

b) Operates on standby for greater than or equal to 5 minutes.

c) Steady-state generator voltage and frequency shall be 4160 +/- 210 volts and 60 +/- 0.6 Hz and shall be maintained throughout this test.

6. Simulating a loss-of-offsite power in conjunction with an ESF actuation test signal, and

a) Verifying deenergization of the emergency busses and load shedding from the emergency busses.

b) Verifying the diesel starts on the auto-start signal,****

energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +

210 volts and 60 + 0.6 Hz during this test.

- - - This test may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.

ST. LUCIE - UNIT 2 3/4 8-7 Amendment No. 39 , 60 , 78 , 89 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a safety injection actuation signal.

7. Verifying the diesel generator operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s**** .

a) Within 10 seconds, generator voltage and frequency shall be 4160 +/- 420 volts and 60 +/- 1.2 Hz.

b) Steady-state generator voltage and frequency shall be 4160 +/- 210 volts and 60 +/- 0.6 Hz and shall be maintained throughout this test.

c) During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a load band of 3800 to 3985 kW

, and d) During the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded within a load band of 3450 to 3685 kW

. 8. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3935 kW.

9. Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power.

b) Transfer its load to the offsite power source, and

c) Be restored to its standby status.

10. Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power.

11. Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the engine-mounted tanks of each diesel via the installed cross connection lines.

This band is meant as guidance to avoid routine overloading of the engine. Variations in load in excess of this band due to changing bus loads shall not invalidate this test.

- - - This test may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.

ST. LUCIE - UNIT 2 3/4 8-9 Amendment No. 39 , 78 , 122 ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class 1E distribution system, and

b. One diesel generator with:

1. Two engine-mounted fuel tanks containing a minimum volume of 200 gallons of fuel,

2. A fuel storage system containing a minimum volume of 42,500 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY

MODES 5 and 6.

ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, operations involving positive reactivity additions that could result in loss of required SHUTDOWN MARGIN or boron concentration, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System through a greater than or equal to 3.58 square inch vent. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.8.1.2.1 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 (except for requirement 4.8.1.1.2a.5).

ST. LUCIE - UNIT 2 3/4 9-1 Amendment No. 92 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling cavity shall be maintained within the limit specified in the COLR.

APPLICABILITY

MODE 6*.

ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 40 gpm of a solution containing 1900 ppm boron or greater to restore boron concentration to within limits.

SURVEILLANCE REQUIREMENTS 4.9.1.1 The boron concentration limit shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full length CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.

ST. LUCIE - UNIT 2 3/4 9-12 Amendment No. 101 REFUELING OPERATIONS 3/4.9.11 SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 The Spent Fuel Pool shall be maintained with:

a. The fuel storage pool water level greater than or equal to 23 ft over the top of irradiated fuel assemblies seated in the storage racks, and

b. The fuel storage pool boron concentration greater than or equal to 1900 ppm.

APPLICABILITY

Whenever irradiated fuel assemblies are in the spent fuel storage pool.

ACTION: a. With the water level requirement not satisfied, immediately suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With the boron concentration requirement not satisfied, immediately suspend all movement of fuel assemblies in the fuel storage pool and initiate action to restore fuel storage pool boron concentration to within the required limit.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the spent fuel storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

4.9.11.1 Verify the fuel storage pool boron concentration is within limit at least once per 7 days.

ST. LUCIE - UNIT 2 3/4 10-1 Amendment No. 26 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth, MTC, and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY

MODES 2 and 3*.

ACTION: a. With any full-length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immedi-ately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1900 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full-length CEAs inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1900 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the SHUTDOWN MARGIN to less than the limits of

Specification 3.1.1.1.

Operation in MODE 3 shall be limited to 6 consecutive hours.

ST. LUCIE - UNIT 2 3/4 11-15 Amendment No. 13 RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 202,500 curies noble gases (considered as Xe-133).

APPLICABILITY

At all times.

ACTION: a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank when reactor coolant system activity exceeds 518.9 Ci/gram DOSE EQUIVALENT XE-133.

ST. LUCIE - UNIT 2 5-4 Amendment No. 7 , 95 , 101 , 135 , 138 DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1 a. The spent fuel storage racks are designed and shall be maintained with:

1. A k eff equivalent to less than 1.0 when flooded with unborated water, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.

2. A k eff equivalent to less than or equal to 0.95 when flooded with water containing 500 ppm boron, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.

3. A nominal 8.965 inch center-to-center distance between fuel assemblies placed in the spent fuel pool storage racks and a nominal 8.80 inch center- to-center distance between fuel assemblies placed in the cask pit storage

rack. 4. For storage of enriched fuel assemblies, requirements of Specification 5.6.1.a.1 and 5.6.1.a.2 shall be met by positioning fuel in the spent fuel pool storage racks consistent with the requirements of Specification 5.6.1.c or in configurations that have been shown to comply with Specifications 5.6.1.a.1 and 5.6.1.a.2 using the methodology as described in Section 9.1 of the Updated Final Safety Analysis Report. 5. Fissile material, not contained in a fuel assembly lattice, shall be stored in accordance with the requirements of Specifications 5.6.1.a.1 and 5.6.1.a.2. b. The cask pit storage rack shall contain neutron absorbing material (Boral) between stored fuel assemblies when installed in the spent fuel pool. c. Loading of spent fuel pool storage racks shall be controlled as described below. 1. The maximum initial planar average U-235 enrichment of any fuel assembly inserted in a spent fuel pool storage rack shall be less than or equal to 4.6 weight percent. 2. Fuel placed in Region 1 of the spent fuel pool storage racks shall comply with the storage pattern definitions of Figure 5.6-1 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions) 3. Fuel placed in Region 2 of the spent fuel pool storage racks shall comply with the storage pattern definitions or allowed special arrangement definitions of Figure 5.6-2 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions)

ST. LUCIE - UNIT 2 5-4a Amendment No. 7 , 101 , 135 DESIGN FEATURES (continued)

CRITICALITY (continued) 5.6.1 c. (continued)

4. The 2x2 array of fuel assemblies that span the interface between Region 1 and Region 2 of the spent fuel pool storage racks shall comply with the storage pattern definitions of Figure 5.6-3 and the minimum burnup requirements as defined in Table 5.6-1. The allowed special arrangements in Region 2 as shown in Figure 5.6-2 shall not be placed adjacent to Region 1. (See Specification 5.6.1.c.7 for exceptions)

5. Fuel placed in the cask pit storage rack shall comply with the storage pattern definitions of Figure 5.6-4 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions)

6. The same directional orientation for Metamic inserts is required for contiguous groups of 2x2 arrays where Metamic inserts are required.

7. Fresh or spent fuel in any allowed configuration may be replaced with non-fuel hardware, and fresh fuel in any allowed configuration may be replaced with a fuel rod storage basket containing fuel rod(s). Also, storage of Metamic inserts or control rods, without any fissile material, is acceptable in locations designated as completely water-filled cells.

d. The new fuel storage racks are designed for dry storage of unirradiated fuel assemblies having a maximum planar average U-235 enrichment less than or equal to 4.6 weight percent, while maintaining a k eff of less than or equal to 0.98 under the most reactive condition.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.

CAPACITY 5.6.3 The spent fuel pool storage racks are designed and shall be maintained with a storage capacity limited to no more than 1491 fuel assemblies, and the cask pit storage rack is designed and shall be maintained with a storage capacity limited to no more than 225 fuel assemblies. The total Unit 2 spent fuel pool and cask pit storage capacity is limited to no more than 1716 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

ST. LUCIE - UNIT 2 5-4b Amendment No. 101

Pages 5-4C through 5-4F (Amendment 101) and page 5-4G (Amendment 135) have been deleted from the Technical Specifications.

The next page is 5-4h.

ST. LUCIE - UNIT 2 5-4h Amendment No. xxx Allowable Storage Patterns (See Notes 1 and 2)

Pattern "A" Pattern "B" Pattern "C" See Definition 1 See Definition 2 See Definition 3 DEFINITIONS:

1. Allowable pattern is Fresh Fuel or fuel of lower reactivity checkerboarded with completely water-filled cells. Diagram is for illustration only, where F represents Fresh Fuel and WH represents a completely water-filled cell.

2. Allowable pattern is placement of fuel assembly type 1 or fuel of lower reactivity in each 2x2 array location with at least one full-length full-strength CEA or equivalent (5 absorber rods) placed in any cell. Minimum burnup for fuel assembly type 1 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only. 3. Allowable pattern is placement of fuel assembly type 2 or fuel of lower reactivity in three of the 2x2 array locations in combination with one completely water-filled cell. Minimum burnup for fuel assembly type 2 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

NOTES: 1. The storage arrangements of fuel within a rack module may contain more than one pattern. There are no interface limitations within Region 1 between rack modules or within racks; however, each assembly must meet the burnup requirements of each 2x2 array that it resides within.

2. Open cells within any pattern are acceptable.

FIGURE 5.6-1 Allowable Region 1 Storage Patterns and Fuel Arrangements WH F WH F 1 1 1 2 2 2 WH CEA 1 ST. LUCIE - UNIT 2 5-4i Amendment No. xxx ALLOWED SPECIAL ARRANGEMENTS (See Notes 1 and 2) Fresh Fuel Assemblies in Region 2 Racks See Definition SR1 See Definition SR2 See Definition SR3 F F F F M M M M M M F F M M M F F M M M M M M M M M F F M M M F F M M M F = Fresh Fuel Assembly M= Fresh Assembly with Metamic Insert

= Empty Cell = Fresh Assembly per Definition ALLOWABLE STORAGE PATTERNS (See Notes 1 and 2)

Pattern "D" Pattern "E" Pattern "F" See Definition 1 See Definition 2 See Definition 3 Pattern "G" Pattern "H" See Definition 4 See Definition 5 FIGURE 5.6-2 (Sheet 1 of 3) Allowable Region 2 Storage Patterns and Fuel Arrangements 6 6 CEA 6 CEA 7 7 7 CEA 7 6 WH 3 3 4 4 4 5 5 5 5 4 3 ST. LUCIE - UNIT 2 5-4j Amendment No. xxx DEFINITIONS for Figure 5.6-2

1. Allowable pattern is fuel assembly type 3 or fuel of lower reactivity in three of the 2x2 array locations in combination with one completely water-filled cell. Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

2. Allowable pattern is fuel assembly type 4 or fuel of lower reactivity in each of the 2x2 array locations with at least two Metamic inserts placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

3. Allowable pattern is fuel assembly type 5 or fuel of lower reactivity in each of the 2x2 array locations with at least one Metamic insert placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 5 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

4. Allowable pattern is fuel assembly type 6 or fuel of lower reactivity in each of the 2x2 array locations with at least two full-length, full strength 5 finger CEAs or equivalent placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 6 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

5. Allowable pattern is fuel assembly type 7 in each of the 2x2 array locations with at least one full-length, full strength 5 finger CEA or equivalent placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 7 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

SR1. Allowable pattern is up to four fresh fuel assemblies or fuel of lower reactivity placed in a 3x3 array in combination with Pattern "D" placed outside the 3x3 array. Fresh fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly of type 3 or fuel of lower reactivity shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

SR2. Allowable pattern is up to four fresh fuel assemblies or fuel of lower reactivity placed in a 3x3 array in combination with Pattern "E" placed outside the 3x3 array. Fresh fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly of type 4 or fuel of lower reactivity with a Metamic insert shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only. FIGURE 5.6-2 (Sheet 2 of 3) Allowable Region 2 Storage Patterns and Fuel Arrangements

ST. LUCIE - UNIT 2 5-4k Amendment No. xxx SR3. Allowable pattern is up to four fresh fuel assemblies or fuel of lower reactivity placed in a 3x3 array in combination with Pattern "F" placed outside the 3x3 array. Fresh fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly of type 5 or fuel of lower reactivity with a Metamic insert shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 5 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only. NOTES 1. The storage arrangements of fuel within a rack module may contain more than one pattern. There are no interface limitations within Region 2 between rack modules or within racks. However, each assembly must meet the burnup requirements of each 2x2 array or allowed special arrangement that it resides within.

2. Open cells within any pattern are acceptable.

FIGURE 5.6-2 (Sheet 3 of 3) Allowable Region 2 Storage Patterns and Fuel Arrangements

ST. LUCIE - UNIT 1 5-4l Amendment No. xxx Allowed Region 2 to Region 1 Fuel Alignments (See Note 1)

Region 2 Interface of Region 2 Pattern "D" with Region 1 -----------------------------------------------------------------

See Definition 1 Region 1 Region 2 Interface of Region 2 Pattern "E" with Region 1 -----------------------------------------------------------------

See Definition 1 Region 1 Region 2 Interface of Region 2 Pattern "F" with Region 1 -----------------------------------------------------------------

See Definition 1 Region 1 Region 2 Interface of Region 2 Pattern "G" with Region 1 -----------------------------------------------------------------

See Definition 1 Region 1 Region 2 Interface of Region 2 Pattern "H" with Region 1 -----------------------------------------------------------------

See Definition 1 Region 1 FIGURE 5.6-3 (Sheet 1 of 2) Interface Requirements between Region 1 and Region 2 3 3 3 3 3 WH 3 WH 4 4 4 4 4 4 4 4 5 5 5 5 5 5 5 5 6 6 CEA 6 6 CEA 6 CEA 6 6 CEA 6 7 7 7 7 7 CEA 7 7 CEA 7 ST. LUCIE - UNIT 2 5-4m Amendment No. xxx DEFINITION:

1. Each 2x2 array that spans Region 1 and Region 2 shall match one of the Region 2 allowable storage patterns as defined by Specification 5.6.1.c.3. Any required Metamic inserts must be placed into the fuel assemblies in Region 2. Locations of water-filled cells or CEAs may be in either Region 1 or Region 2. For interface assemblies, the requirements of Specifications 5.6.1.c.2 and Specification 5.6.1.c.3 shall be followed within Region 1 and Region 2, respectively. The Diagrams are for illustration only.

NOTES:

1. Open cells within any pattern are acceptable.

FIGURE 5.6-3 (Sheet 2 of 2)

Interface Requirements between Region 1 and Region 2

ST. LUCIE - UNIT 2 5-4n Amendment No. xxx Allowable Storage Patterns (See Notes 1 and 2)

Pattern "A" Pattern "B" See Definition 1 See Definition 2 DEFINITIONS:

1. Allowable pattern is Fresh Fuel or lower reactivity fuel checkerboarded with completely water-filled cells. Diagram is for illustration only, where F represents Fresh Fuel and WH represents a completely water-filled cell.

2. Allowable pattern is placement of fuel assembly type 8 or fuel of lower reactivity in three of the 2x2 array locations in combination with one completely water-filled cell in any location. Minimum burnup for fuel assembly type 8 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

NOTES: 1. The storage arrangements of fuel within a rack module may contain more than one pattern. There are no interface limitations within the cask pit storage rack, however, each assembly must meet the burnup requirements of each 2x2 array that it resides within.

2. Open cells within any pattern are acceptable.

FIGURE 5.6-4 Allowable Cask Pit Storage Rack Patterns WH F WH F 8 8 8 WH ST. LUCIE - UNIT 2 5-4o Amendment No. xxx TABLE 5.6-1 Minimum Burnup Coefficients Coefficients Fuel Type Cooling Time (Years) A B C 1 0 -33.4237 25.6742 -1.6478 2 0 -25.3198 14.3200 -0.4042 3 0 -23.4150 16.2050 -0.5500 0 -33.6414 25.0670 -1.5551 2.5 -32.3764 23.9988 -1.5075 5 -30.9234 22.9382 -1.4372 10 -28.4951 21.1511 -1.3029 15 -27.2024 20.2802 -1.2479 4 20 -25.2009 18.6218 -1.0364 0 -24.8402 23.5991 -1.2082 2.5 -23.0170 21.6493 -1.0298 5 -21.9293 20.6257 -0.9730 10 -20.0813 19.0808 -0.9022 15 -19.5503 18.5429 -0.9129 5 20 -18.7485 17.7308 -0.8390 0 -32.4900 25.3077 -1.5518 2.5 -31.1598 23.9185 -1.4435 5 -29.2169 22.5424 -1.3274 10 -26.8886 20.6662 -1.1425 15 -25.5703 19.7629 -1.1129 6 20 -24.5754 18.9056 -1.0147 0 -24.6989 24.1660 -1.2578 2.5 -23.0399 22.3047 -1.0965 5 -21.3290 20.7413 -0.9613 10 -20.0836 19.4780 -0.8949 15 -19.2480 18.5880 -0.8685 7 20 -18.6424 18.1241 -0.8950 8 0 -47.5000 12.5000 -0.0000 NOTES: 1. To qualify in a fuel type, the calculated burnup of a fuel assembly must exceed the "minimum burnup" determined for the "cooling time" and "maximum initial planar enrichment" of the fuel assembly. The "minimum burnup" for any fuel type is determined from the following polynomial function:

BU = A + B*E + C*E 2 , where: BU = Minimum Burnup (GWD/MTU) E = Maximum Initial Planar Average Enrichment (weight percent U-235) A, B, C = Coefficients for each fuel type 2. Interpolation between values of cooling time is not permitted.

ST. LUCIE - UNIT 2 6-15b Amendment No. 61 , 88 , 130 , 140 , 146 ADMINISTRATIVE CONTROLS than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR

Part 190.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

g. Radiological Environmental Monitoring Program

A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of

the accuracy of the effluent monitoring program and modeling of the environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the

following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in

the ODCM.

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

h. Containment Leakage Rate Testing Program

A program to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program is in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance Containment Leak-Test Program," as modified by Bechtel Topical Report, BN-TOP-1 or ANS 56.8-1994 (as recommended by R.G. 1.163) which will be used for type A testing. The peak calculated containment internal pressure for the design basis loss of coolant accident P a, is 43.48 psig. The containment design pressure is 44 psig.

The maximum allow containment leakage rate, L a , at P a, shall be 0.50% of containment air weight per day.

ST. LUCIE - UNIT 2 6-20 Amendment No. 13 , 25 , 61 , 92 , 105 , 138 ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (continued) 6.9.1.9 At least once every 5 years, an estimate of the actual population within 10 miles of the plant shall be prepared and submitted to the NRC.

6.9.1.10 At least once every 10 years, an estimate of the actual population within 50 miles of the plant shall be prepared and submitted to the NRC.

6.9.1.11 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

Specification 3.1.1.1 Shutdown Margin - Tavg Greater than 200 F Specification 3.1.1.2 Shutdown Margin - Tavg Less Than or Equal to 200 F Specification 3.1.1.4 Moderator Temperature Coefficient Specification 3.1.3.1 Movable Control Assemblies - CEA Position Specification 3.1.3.6 Regulating CEA Insertion Limits Specification 3.2.1 Linear Heat Rate Specification 3.2.3 Total Integrated Radial Peaking Factors - F r T Specification 3.2.5 DNB Parameters Specification 3.9.1 Refueling Operations - Boron Concentration

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in the following documents or any approved Revisions and Supplements thereto:

1. WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietary).

2. NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point & St. Lucie Nuclear Plants," Florida Power & Light Company, January 1995.

3. DELETED

4. DELETED

5. CENPD-275-P, Revision 1-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May 1988, & Revision 1-P Supplement 1-P-A, April 1999.

6. DELETED ST. LUCIE - UNIT 2 6-20a Amendment No. 92 ADMINISTRATIVE CONTROLS (Continued)

CORE OPERATING LIMITS REPORT (COLR) (Continued)

b. (Continued)

7. DELETED

8. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 1: CE Calculated Local Power Density and Thermal Margin/Low Pressure LSSS for St. Lucie Unit 1," December 1979.

9. DELETED

10. CEN-123(F)-P, "Statistical Combination of Uncertainties Methodology Part 3: CE Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for St. Lucie Unit 1," February 1980.

11. CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981.

12. Letter, J.W. Miller (NRC) to J.R. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of CEN-123(F)-P (three parts) and CEN-191(B)-P).

13. DELETED

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

15. DELETED

16. DELETED

17. DELETED

18. DELETED ST. LUCIE - UNIT 2 6-20c Amendment No. 92 , 105 ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR)

(continued)

b. (continued)

34. Letter, J.A. Norris (NRC) to J.H. Goldberg (FPL), "St. Lucie Unit 2 -

Issuance of Amendment Re: Moderator Temperature Coefficient (TAC No. M82517)," July 15, 1992.

35. Letter, J.W. Williams, Jr. (FPL) to D.G. Eisenhut (NRC), "St. Lucie Unit No. 2, Docket No. 50-389, Proposed License Amendment, Cycle 2 Reload

," L-84-148, June 4, 1984.

36. Letter, J.R. Miller (NRC) to J.W. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval to Methodology contained in L-84-148).

37. DELETED

38. DELETED

39. DELETED

40. DELETED

41. DELETED

42. CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997.

43. CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990.

44. DELETED

45. DELETED ST. LUCIE - UNIT 2 6-20d Amendment No. 92 , 105 , 118 , 138 ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR)

(continued)

b. (continued)

46. DELETED

47. DELETED

48. CEN-396(L)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/KG for St. Lucie Unit 2," November 1989 (NRC SER dated October 18, 1991, Letter J.A. Norris (NRC) to J.H. Goldberg (FPL), TAC No. 75947).

49. CENPD-269-P, Rev. 1-P, "Extended Burnup Operation of Combustion Engineering PWR Fuel," July 1984.

50. CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2," December 1984 (NRC SER dated December 21, 1999, Letter K. N. Jabbour (NRC) to T.F. Plunkett (FPL), TAC No. MA4523).

51. CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998.

52. CENPD-140-A, "Description of the CONTRANS Digital Computer Code for Containment Pressure and Temperature Transient Analysis," June 1976.

53. DELETED

54. DELETED

55. CENPD-387-P-A, Revision 000, "ABB Critical Heat Flux Correlations for PWR Fuel," May 2000.

56. CENPD-132, Supplement 4-P-A, "Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model," March 2001.

57. CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998.

58. CENPD-404-P-A, Rev. 0, "Implementation of ZIRLO TM Cladding Material in CE Nuclear Power Fuel Assembly Designs," November 2001.

59. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

July 1985.

60. WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control; FQ Surveillance Technical Specification," February 1994.

ST. LUCIE - UNIT 2 6-20e Amendment No. 105 , 118 , 133 , 138 , 147 ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR)

(continued)

b. (continued) 61. WCAP-11397-P-A, (Proprietary), 'Revised Thermal Design Procedure," April 1989. 62. WCAP-14565-P-A, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999. 63. WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code," May 2003. 64. Letter, W. Jefferson Jr. (FPL) to Document Control Desk (USNRC), "St. Lucie Unit 2 Docket No. 50-389: Proposed License Amendment WCAP-9272 Reload Methodology and Implementing 30% Steam Generator Tube Plugging Limit," L-2003-276, December 2003 (NRC SER dated January 31, 2005, Letter B.T. Moroney (NRC) to J.A. Stall (FPL), TAC No. MC1566) 65. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses", April 1999. 66. WCAP-7908-A, "FACTRAN-A FORTRAN IV Code for Thermal Transients in a UO2 Fuel Rod", December 1989. 67. WCAP-7979-P-A, "TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code", January 1975. 68. WCAP-7588, Rev. 1-A, "An Evaluation of The Rod Ejection Accident In Westinghouse Pressurized Water Reactors Using Special Kinetics Methods", January 1975. c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met. d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle on the NRC.

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