Text

Technical Specification Pages
	ML12191A236
Person / Time
Site:	Saint Lucie
Issue date:	07/09/2012
From:	Orf T; Plant Licensing Branch II
To:	Nazar M; Florida Power & Light Co
	Orf, T J
Shared Package
ML12191A220	List: ML12156A208; ML12181A019; ML12191A236;
References
	TAC ME5091
	Download: ML12191A236 (66)
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ATTACHMENT TO LICENSE AMENDMENT NO. 213 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 Replace Pages 3 and 5 of Renewed Operating License DPR-67 with the attached Pages 3 and

5. Insert the attached Page 6 of Renewed Operating License DPR-67.

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

Remove Pages I

III VIII 1-3 1-4 1-6 2-2 2-4 3/41-1 3/4 1-3 3/41-8 3/41-9a 3/41-10 3/4 1-12 3/41-16 3/4 1-17 3/41-18 3/4 1-19 3/42-13 3/42-14 3/4 3-7 3/4 3-8 3/44-17 3/44-19 3/44-20 3/44-21 3/44-22 3/44-23a 3/44-23b 3/44-23c 3/44-59 3/44-60 3/45-1 3/4 5-3 3/4 5-5 Insert Pages I

III VIII 1-3 1-4 1-6 2-2 2-4 3/41-1 3/4 1-3 3/4 1-8 3/41-9a 3/41-10 3/4 1-12 3/4 1-16 3/41-17 3/41-18 3/4 1-19 3/42-13 3/42-14 3/43-7 3/4 3-8 3/44-17 3/44-19 3/44-20 3/44-21 3/44-22 3/44-23a 3/44-23b 3/44-23c 3/44-59 3/44-60 3/45-1 3/4 5-3 3/4 5-5

- 2 3/4 5-8 3/46-12 3/4 7-2 3/4 7-3 3/4 7-6 3/48-1 3/4 8-6 3/4 8-7 3/49-1 3/49-11 3/49-16 3/4 10-1 3/4 11-15 5-4 5-5 5-6 5-6b 5-6c 5-6d 5-6e 6-15b 6-19 6-19a 6-19b 3/4 5-8 3/46-12 3/4 7-2 3/4 7-3 3/4 7-6 3/48-1 3/4 8-6 3/4 8-7 3/49-1 3/49-11 3/49-16 3/4 10-1 3/4 11-15 5-4 5-5 5-6 5-6b 5-6c 5-6d 5-6e 5-6f 5-69 6-15b 6-19 6-19a 6-19b

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

A.

Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts (thermal).

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 213 are hereby incorporated in the renewed license.

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

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

C.

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

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

D.

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

Renewed License No. OPR-67 Amendment No. 213

-5 (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders H.

Control Room HabitabililV Upon implementation of Amendment No. 205, adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 4.7.7.1.e, in accordance with TS 6.8.4.m, the assessment of CRE habitability as required by Specification 6.8A.m.c. (ii), and the measurement of CRE pressure as required by Specification 6.8.4.m.d. shall be considered met Following implementation:

(a)

The first performance of SR 4.7.7.1.e, in accordance with Specification 6.BAm.cO), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 4.0.2, as measured from September 2003, the date of the most recent successful tracer gas test.

as stated in FPL letters to NRC dated December 9, 2003, and October 29, 2004, in response to Generic Letter 2003-01.

(b)

The first performance of the periodic assessment of CRE habitability, Specification 6.8.4.m.c(ii), shall be within 3 years, plus the 9-month allowance of SR 4.0.2, as measured from September 2003, the date of the most recant successful tracer gas test, as stated in FPL letters to NRC dated December 9, 2003, and October 29, 2004, in response to Generic Letter 2003..{)1, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c)

The first performance of the periodic measurement of CRE pressure, Specification 6.SA.c.d, shall be within 36 months in a staggered test basis, plus the 138 days allowed by SR 4.0.2, as measured from June 30, 2006, which is the date of the most recen1 successful pressure measurement test, or within 138 days if not performed previously.

I.

RQDEX2 Safety Analyses RODEX2 has been specifically approved for use for S1. Lucie Unit 1 licensing basis analyses. Upon NRC's approval of a generic supplement to the RODE.X2 code and associated methods that accounts for thermal conductivity degradation (TeO). FPL will within six months:

(a)

Demonstrate that St lucie Unit '1 safety analyses remain conservatively bounded in licensing basis analyses when compared to the NRC approved generic supplement to the RODEX2 methodology. or (b)

Provide a schedule for the re-analysis using the NRC-approved generic supplement to the RODEX.2 methodology for any of the affected licensing basis analyses.

Renewed License No. DPR-67 Amendment No. 213

- 6

4.

This renewed license is effective as of the date of issuance and shall expire at midnight on March 1, 2036.

FOR THE NUCLEAR REGULATORY COMMISSION ORIGINAL SIGNED BY J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Appendix A, Technical Specifications

2. Appendix B, Environmental Protection Plan Renewed License No. DPR-67 Amendment No. 213 Revised by letter dated July 9,2012

INDEX SECTION 1.0 DEFINITIONS 1.1 Action............................................................................................................................1-1 1.2 Axial Shape Index.........................................................................................................1-1 1.3 Azimuthal Power Tilt......................................................................................................1-1 1.4 Channel Calibration......................................................................................................1-1 1.5 Channel Check.................................,.,.........................................................................1-1 1.6 Channel Functional Test............................................................................................,.. 1-2 1.7 Containment Vessel Integrity..................... '..............,................................................... 1-2 1.8 Controlled Leakage................................................,...................................................... 1-2 1.9 Core Alteration..........,.........................,.....................................,................................... 1-2 1.9a Core Operating Limits Report (COLR).........................................................................1-2 1.10 Dose Equivalent 1-131....................................................... _.......................................... 1-3 1.11 Dose Equivalent Xe-133...............................................................................................1-3 1.12 Engineered Safety Features Response Time...............................................................1-3 1.13 Frequency Notation......................................................................................................1-3 1.14 Gaseous Radwaste Treatment System.......................................................................1-3 1.15 Identified Leakage........................................................................................................1-4 1.16 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...................................................................................... "..........1-5 1.20 Operational Mode - Mode............................................................................................1-5 1.21 PhYSics Tests...............................,................................................................................ 1-5 1.22 Pressure Boundary Leakage........................................................................................1-5 ST. LUCIE* UNIT 1 Amendment No. -;Y, ~, 66, &9,

.:/W,213

INDEX LIMITING CONDlTJO~~ FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY................................................................................................ 3/40-1 Boron Dilution................................................ :.................................................... 3/41-4 Minimum Temperature for Criticality...................................................................3/41-7 Charging Pumps - Shutdown........... _..................................................

3/41-1'2 Charging Pumps _ Operating...........................................................................3/41-13 Borated Water Sources - Shutdown................................................................. 3/41-16 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL....................................................................................... 3/41-1 Shutdown Margin - T""9 > 200"F......................................................................314 1-1 Shutdown Margin - Tavg 5. 200°F........................................................................ 3/4 1-3 Moderator Temperature Coefficient.................................................................... 3/4 1-5 3/4.1.2 BORATION SYSTEMS.......................................................................................314 1-8 Flow Paths - Shutdown......................................................................................3/4 1-8 Flow Paths - Operating....................................................................................314 1-10 Boric Acid Pumps - Shutdown......................................................................... 3/4 1-14 Boric Acid Pumps - Operating....................................................................... 3/4 1-15 Borated Water Sources - Operating.................................................................3/4 1-17 3/4.1.3 MOVABLE CONTROL ASSEMBLIES........................................................... 314 1-20 Full Length CEA Position..................................................................................3/41-20 Shutdown CEA Insertion limit..........................................................................3/41-27 Regulating CEA Insertion Limits.......................................................................3/41-28 Position Indicator Channels............................................................................3/4 1-24 CEA Drop Time................................................................................................. 3/4 1-26.

ST. LUCIE - UNIT 1 III Amendment No. 37,443.434. 213

LIMITING CONDITIONS FOR OPERATION MID SURvEILLANCE REQUIREMENTS SECTION PAGE

.;}/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.............................................................................. 3/49-1 3/4.9.2 INSTRUMENTATION.........................................................................................3/49-2 3/4.9.3 DECAY TIME......................................................................................................3/49-3 3/4.9,4 CONTAINMENT-PENETRATIONS.....................................................................3/49-4 3/4.9.5 COMMUNICATIONS..........................................................................................3/49-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY...................................................... :..... 3/4 9-6 3/4.9.7 DELETED...........................................................................................................3/49-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION................................. 314 9-8 3/4.9.9 CONTAINMENT ISOLATION SySTEM.............................................................. 3/4*9-9 3/4.9.10 WATER LEVEL - REACTOR VESSEL............................................................ 3/4 9-10 3/4.9.11 SPENT FUEL STORAGE POOL..................................................................... 3/4 9-11 3/4.9.12 FUEL POOL VENTILATION SYSTEM FUEL STORAGE.............................. 3/4 9-12 3/4.9.13 DELETED......................................................................................................... 3/49-15 3/4.9.14 DELETED......................................................................................................... 3/49-16 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN................................................................................... 3/410-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.........................................................................................................3/410-2 3/4.10.3 DELETED......................................................................................................... 3/410-3 3/4.10.4 DELETED.........................................................................................................3/410-4 3/4.10.5 CENTER CEA MISALIGNMENT......................................................................3/410-5 ST. LUCIE - UNIT 1 VIII Amendment No.4, :M, :!ii, ~. 56, 4W,213

DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 ().lCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 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.M DOSE EQUIVALENT XE*133 1.11 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (I1Ci/gram) that alone would produce the same"acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131 m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-:138 actually present. If a specifIC noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-l33 shall be performed using effective dose conversion factors for air SUbmersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and SoiL" ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES REPONSE 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 measu red 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 effJuef.lts 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.

ST. LUCIE UNIT 1 1-3 Amendment No. :;g.. 69.~. we, dOO.

213

IDENTIFED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

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

b.

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

c.

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

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 DOSECALCULAllON 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 AlarmfTrip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitor ing Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

ST. LUCIE - UNIT 1 1-4 Amendment No. &9, 00. 99, 3+. -+04, -RJ.

m.~.213

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 arid 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.

SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.30 SITE BOUNDARY means that line beyond which the land or property is not owned,leased, or otherwise controlled by the licensee.

SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

ST. LUCIE - UNIT 1 1-6 Amendment No.:;g., ~, ea, ea, 89,~,

-+ea,213

1500 UNACCEPTABLE OPERATION 580

~

""t,CEPTABLE OPERATION PRESSURE IN PSIA

~

uJ' 560 ex:

S

\\,1

~ n 2400 REACTOR OPERATION LIMITED TO lESS THAN w

580°F BY ACTUATION OF THE MAIN STEAM SAFETY VALVES

! 540 I

REACTOR COOLANT SYSTEM TOTAL FLOW RATE

~

Q 520 SPECifiED IN LCO 3.2.5 e

LIMITS CONTAIN NO ALLOWANCE FOR INSTRUMENT ERROR OR FLUCTUATIONS i

IE 500 i

BASED ON THE AXIAL SHAPE ON FIGURE 62.1-1 480 ACCEPTABLE OPERATION

~ 2225 Q 2000 h 1750 THERMAL POWER LIMITED TO A MAXIMUM OF 112% OF RATED THERMAL POWER BY THE HIGH POWER TRIP LEVEL 460LI------------------------------------------------~------------------~

o 0.2 0.4 0.5 0.8 1

1.2 1.4 1.6 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY L1MIT FOUR REACTOR COOLANT PUMPS OPERATING ST. LUCIE - UNIT 1 2-2 Amendment No. 43. ~, 44&,~, 213

TABLE 2.2-"

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT ALLOWABLE VALUES TRIP SETPOINT

-~-.---

Not Applicable Not Applicable

1. Manual Reactor Trip

2. Power Level-High (1)

Four Reactor Coolant Pumps Operating

, 9.61% above THERMAL POWER, and a minimum setpolnt of 15% of RATED

.! 9.61% above THERMAL POWER, with a minimum setpolnt of 15% of RATED THERMAL POWER, and a maximum of THERMAL POWER and a maximum of !

-< 107.0% of RATED THERMAL

.5. 107.0% of RATED THERMAL POWER.

POWER.

3, Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps Operating ::. 95% of minimum reactor coolant flow

. 95% of minimum reactor coolant flow with 4J~umps operating..

with 4 QumQs oQeratlng.,

-"~---

, 2400 psia

.! 2400 psis

4. Pressurizer Pressure - High r--~~

< 3,3 psig

5. ContaInment Pressure - High

~~~: ~ J~sig

> 600 psia

6. Steam Generator Pressure -low (2)

>~6~OO psi a

. 35.0% Water level - each steam

. 35.0% Water level-each steam generator

7. Steam Generator Water level - low generator

~

.-,.~

local Power Density - High (3)

Trip setpoint adjusted to not exceed the Trip set point adjusted to not exceed the limit lines of Figures 2.2-1 and 2,2-2, limit lines of Figures 2.2-1 and 2,2-2.

'11 For minimum reactor coolant flow with 4 pumps operating, refer to Technical Specification lCO 3.2.5, ST, LUCIE - UNIT 1 2-4 AJnendmentNo.3,~,32, 4i.+Oe,~,44i,~

213

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORA TION CONTROL SHUTDOWN MARGIN

Tam > 200 of LIMITING CONDlJION 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:

Wrth the SHUTDOWN MARGIN not within limits immediately initiate and continue boration at ~ 40 gpm of 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 increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

b.

When in MODES 1 or 2', 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# at least once during CEA withdrawal and at least once per hour thereafter until the reactor is critical.

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel IQading, by consideration of the factors of e below, with the CEA groups at the Power Dependent Insertion Limits of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

With K.,. ;'! 1.0.

1. Wrth ~ < 1.0.

ST. LUCIE* UNIT 1 3141*1 Amendment No.;J;i., 4.5, ~, Ii, ~.

+14.213

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN" T~ ~ 200 of LIMITING CONDITION fOR 0PSRATIQH 3.1.1.2 The SHUTDOWN MARGIN shall be:

Within the limits specified in the COLR, and in addition with the Reactor Coolant System drained below the hot leg centerline, one charging pump shall be rendered inoperable.*

APPLICABILITY: MODE 5.

ACTION:

If the SHUTDOWN MARGIN requirements cannot be met. immediately initiate and continue boration at.:: 40 gpm of greater than or equal to 1900 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEIL! ANCE REQllIREMENTS 4.1.1.2 The SHUTDOWN MARG1N requirements of Specification 3.1.1.2 shall be determined:

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 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 coo/ant system boron concentration,

2. CEA position,

3. Reactor coolant system average temperature,

4. Fuel bumup 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 4.1.1.2.b and by verifying at least one charging pump is rendered inoperable.*

Breaker racked-out.

ST. LUCIE* UNIT 1 3/41*3 Amendment No. MI, ill, 474.213

REACTIVITY CONTROL SYSTEMS 314.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITIOt.j FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

a. A flow path from the boric acid makeup tank via either a boric acid pump or a gravity feed connection and any charging pump to the Reactor Coolant System if only the boric acid makeup tank in Specification 3.1.2. 7a is OPERABLE, or

b. The flow path from the refueling water tank via either a charging pump or a high pressure safety injection pump" to the Reactor Coolant System if only the refueling water tank in Specification 3.1.2.7b is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes"" until at least one injection path is restored to OPERABLE status.

SURVEILLANCE !3EQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position. is in its correct position.

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

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

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

ST. LUCIE - UNIT 1 Jf41-8 Amendment No. SQ, 84-, 00, 94, 4G3,

-W4, #,~, m, 213

DELETED ST. LUCIE - UNIT 1 314 1-9a Amendment No. M-, -W4, +4+, +00,213

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 make the reactor subcritical within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and borate to a SHUTDOWN MARGIN equivalent to the requirements of Specification 3.1.12 at 200"F; 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 1 3141-10 Amendment No. 43. 00, 00, 94, +7-1-, 213

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

APPLICABILITY: MODES 5 and 6.

ACTION:

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

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

when tested pursuant to the (nservice Testing Program.

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

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

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

ST. LUCIE - UNIT 1 3141-12 Amendment No. 00, 8-1-, W. 4Q4. +W.

l.~.4Q.m. 213

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONPITION 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 3650 gallons of 3.0 to 3.5 weight percent boric acid (5245 to 6119 ppm boron).

b. The refueling water tank with:

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

2. A minimum boron concentration of 1900 ppm, and

3. A minimum solution temperature of40 G F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water sources OPERABLE, suspend all operations involving positive reactivity changes* until at least one borated water source is restored to.OPERABLE st-atus.

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 water level 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 mborated water and the site ambient air temperature is <: 40"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 5S"F by verifying that the Boric Acid Makeup Tank solution temperature is greater than SS"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.

51. LUCIE - UNIT 1 3141.16 AmendmentNo.~.~.m. 213

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 At !east two of the foHowing four borated water sources shall be OPERABLE:

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

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

c.

Boric Acid Makeup Tanks 1 A and 1 B 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 volume of 477,360 gallons of water.

2. A minimum boron concentration of 1900 ppm,

3. A maximum solution temperature of 100"F.

4.

A minimum solution temperature of 55°F when in MODES 1 and 2. and

5.

A minimum solution temperature of 4QoF when in-MODES 3 and 4.

APPLICABILITY: MODES 1. 2. 3 and 4.

ACTION:

With only one borated water source OPERABLE, restore at least two borated water sources to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or make the reactor subcritical within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and borate to a SHUTDOWN MARGIN equivalent to the requirements of Specification 3.1.1.2 at 200"F; restore at least two borated water sources 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 .

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

a. At least once per 7 days by:

1.

Verifying the boron concentration ofthe water source, ST. LUCIE - UN!T 1 3141-17 Amendmen1 No. 28. 43, 89, 1M*.:t39.

4+4,2W.213

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIB.ENj[;NTS (Continued)

2.

Verifying the water level in each 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.

c.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that ihe Boric Acid Makeup Tank solution temperature is greater than 55°F when the Reactor Auxiliary Building air temperature is below 55°F.

ST. LUClE - UNIT 1 3141-18 Amendment No. 94, 213

FIGURE 3.1*1 ST. LUCIE 1 MIN BAMT VOLUME VS STORED BAMT CONCENTRATION (MODES " 2, 3 and 4)

~..

~

'i.e w

J

..I 0

OPERATION (6800) r m

6000 z

~

(8700) 8000 7000 (7500)

I UNACCEPTABLE OCOO~*------------~-----------r----------~r---------~

40ID~----------~----------~~----------~----------~

3.00 3.25 3.50 (5682 PPM)

(5245 PPM) 16119 PPM)

STORED BAtH CONe (wt% bork: ackf)

ST. LUCIE: - UNIT 1 3141-19 Amendment No. 2+.94, 4.ilG, 213

POWER DISTRIBUTION LIMITS ONB PARAMETERS LIMITING CONDITION FOR OeERATION 3.2.5 The following DNB related parameters shall be maintained within the limits:

a.

Cold Leg Temperature as shown on Table 3.2-1 of the COLR.

b.

Pressurizer Pressure* as shown on Table 3.2-1 of the COLR, C.

Reactor Coolant System Total Flow Rate - greater than or equal to 375.000 gprn, and

d.

AXIAL SHAPE INDEX as shown on Figure 3.24 of the COLR.

APPLICABILITY: MODE 1.

ACTION:

INith any of the above parameters exceeding its limit, restore the parameter to within its rllTlit 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 shaH 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 1 3/42-13 Amendment No. 2+. 410, 213

Relocated to the COLR ST. LUCIE - UNIT 1 3/42-14 Amendmenl No.~. <IB, ~, ~.

~,-l93. 213

TABL.E 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT

1.

Manual Reactor Trip

2.

Power Level-High

a. Nuclear Power

b. !!.T Power

3.

Reactor Coolant Flow ~ Low

4.

Pressurizer Pressure - High

5.

Containment Pressure High

e.

Steam Generator Pressure - Low

7.

Steam Generator Water Level - Low B.

Local Power Density - High

9. Thermal Margin/Low Pressure 9a. Steam Generator Pressure Difference - High

10. Loss of Turbine -- Hydraulic Fluid Pressure - Low

11. Wide Range Logarithmic Neutron Flux Monitor

12. Reactor Protection System Logic

13. Reactor Trip Breakers ST. LUCIE - UNIT 1 CHANNE;L CHECK N/A CHANNEL CALIBRATION NA CHANNEL FUNCTIONA1.

TEST S/U(1)

MODES IN WHICH SURVEI1.1.ANCE REQUIRED N/A S

S S

D(2), M(3), 0(5) 0(4), Q R

R R

M M

M(6,7) 1,2 1,2 1,2 1, 2 1,2 1,2 S

S R

R M

M 1

1,2 S

R M

1.2 NA NA S/U(1)

NA S

N.A.

NA N.A.

N.A.

S/U(1)

M and S/U(1)

M 1,2,3,4, 6 and

1,2 and

1,2 and" 3/43*7 Amendment No. ::1+, ~, 213

TABLE 4.3-1 (Continued)

TABLE NOTATION With reactor trip breaker closed.

(1)

If not performed in previous 7 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER; adjust "Nuclear Power Calibrate" potentiometer to null "Nuclear Pwr - AT Pwr." During PHYSICS TESTS, these daily calibrations of nuclear power and IlT power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(3)

Above 15% of RATED THERMAL POWER. recalibrate the excore detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POVVER during subsequent operations to ~ 90% of the maximum allowed THERMAL POWER level with the existing Reactor Coolant Pump combination.

(4) -

Adjust "AT Pwr Calibrate" potentiometers to make IlT power signals agree with calorimetric calculation.

(5)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(6) -

If the as-found 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 ;s 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 Field Trip Setpoint. otherwise that channel shall not be returned to OPERABLE status. The Field Trip Setpoint and the methodology used to determine the Field Trip Setpoint, the as-found acceptance criteria band, and the as-left acceptance criteria are speCified in the UFSAR Section 7.2.

ST. LUCIE - UNIT 1 3/43-8 Amendment No. 213

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITIQN FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a.

~ 1.0 /-lCi/gram DOSE EQUIVALENT 1-131, and

b. ~ 518.9 /-lCi/gram DOSE EQUIVALENT XE-133.

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

ACTION:

a. With the' specific activity of the primary coolant >1.0 !lCi/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 is ~ 60.0 /-lCi/gram once per four hours.

b. With the specific activity of the primary coolant> 1.0 /-lCi/gram DOSE EQUIVALENT 1-131. but~ 60.0 J..lCi/gram DOSE EQUIVALENT 1-131, 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 1-131 to within the 1.0 /-lCi/gram limit. Specification 3.0.4 is not applicable.

c.

With the specific activity of the primary coolant> 1.0 J..lCi/gram. DOSE EQUIVALENT 1-13*1 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 j.1Ci/gram DOSE EQUIVALENT 1-131, 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 /-lCi/gram DOSE EQUIVALENT XE-1'33, 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 /-lCi/gram DOSE EQUIVALENT XE-133 limit. Specification 3.0.4 is not applicable.

e.

With the specific activity of the primary coolant> 518.9 /-lCi/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 ofthe primary coolant shall be determined to be within the limits by peliormance of the sampling and analysis pro gram of Table 4.4-4.

ST. LUCIE* UNIT 1 3/44-17 Amendment No. 00, m, 213

TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVIIY SAMPLE AND ANALY§IS PROGRAM TYPE OF MEASUREMENT MINIMUM AND ANALYSIS FREQUENCY

1.

DOSE EQUIVALENT XE-133 Determination 1 per 7 days'

2.

Isotopic Analysis for DOSE 1 per 14 days EQUIVALENT 1-131 Concentration

3. Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Including 1-131, 1-132, 1-133,1-134, whenever the DOSE and 1-135 EQUIVALENT 1-131 exceeds 1.0 fJ.Ci/gram, and 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 percent of the RATED THERMAL POWER within a one hour period.

Until the specific activity of the primary coolant system Is restored within its ST. LUCIE - UNIT 1 3/44-19 MODES !N WHICH SAMPLE AND ANALYSIS REQUIRED 2,3 and 4

.2,3. and 4 1.2,3 Amendment No. 213

DELETED ST. LUCIE - UNIT 1 3/44-20 Amendment No. 213

REACTOR COOLANT SYSTEM 314.4.9 PRESSUREITEMPERA TURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2a and 3.4-2b during heatup. cooldown. criticality. and inservice leak and hydrostatic testing.

APPLICABILITY: At aU times."

ACTION:

With any of the above limits exceeded. restore the temperature and/or pressure to within the limits within 30 minutes; perform an analysis to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations 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 reduce the RCS T avg to less than 200°F within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months in accordance with Figure 3.4-2b.

.. During hydrostatic testing operations above system design pressure, a maximum temperature change in anyone hour period shall be limited to 5°F.

ST. LUCIE* UNIT 1 3/44*21 Amendment No. 4; 84. 4-04,~. 213

REACTOR COOLANT SYSTEM SURVEil I ANCE REQUIREMENTS 4.4,9.1

a.

The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

b.

The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.

c.

The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in malerial properties as required by 10 CFR 50 Appendix H. The results of these examinations shall be used to update Rgures 3.4-2a and 3.4-2b.

ST. LUCIE - UNIT 1 3144-22 Amendment No. 84,.:143, 213

FIGURE 3.4-2a ST. LUCIE UNIT 1 PIT LIMITS, 54 EFPY HEATUP AND CORE CRITICAL t,

1'-

-~-

-i 2000 f---- f-

--ff

-I f-ISOTHERMAl f-I H,

1 I

rLLOWEST SERVICE TEM?ERA, URE 155 F 1---+1-- -(--+---------j------i------t U1

~,

' i N

rr:

(j)

U)

. CORE CRITICAL U1 a:

n.

o

--~-

--I

--70"FiHR

'-1J r

"-f '.'..-.

~

o

--+

~~-I-

-i

-r-f---/--

+----i o

i::!55iJ.a ?s1A z

I

~

500

n.

--t--

+--

+

ALLOWABLE HEATLP RAlt. ;

MIN.BOLTUPTEMP.80'F r-TO'FiHR E

o o

100 200 300 400 500 Tc -INDICATED Res TEMPERATURE.OF Limiting Material: Lower Shell Axial Welds (Ht. #305424)

Limiting ART Values at 54 EFPY:

1/4T,210°F 3/4T, 156°F ST. LUCIE - UNIT 1 3/44-23a Amendment No. -'1+. ;m. ~.-W4.

~.~. 213

FIGURE 3.4-2b ST. LUCIE UNIT 1 PIT LIMITS. 54 EFPY COOLDOWN AND INSERVICE TEST 2000

~

Ci5 l00"flHR TO ISOTHERMAL a..

uJ rr.

J

~ 1500 UJ 0:::

a..

LOWEST SERVlCE TEMPERA-TURE 16S"F 0:::

u.I N cr CJ) r.I) w 1000 a::

a..

o 5

w ISOTHERMAl..

o z 13 500 a::

a..

a,

MIN. BOLTUP TEMPERA'TURE 8O"F AlLOWABLE COOLOOWN RATES RATE. *FIHR TEMP. LIMIT. 'F 2(l

<125 30 125-145

~o 145-161) 50 16().100 100

>180 o

100 200 300 400 500 Tc -INDICATED Res TEMPERATURE,oF Limiting Material: Lower Shell Axial Welds (Ht #305424)

Limiting ART Values at 54 EFPY:

1/4T.21Oa F 3/4T.156°F ST. LUCIE - UNIT 1 3144-23b Amendment No. 41-, 28, St,

'IQ4, 444,400,213

DELETED ST. LUCIE - UNIT 1 3f44-23c Amendment No. +7, ~, 8+,.w4,

+4,.;.ga, 213

REACTOR COOLANT SYSTEM POWER OPERATED RELIEF VALVES 3.4.13 Two power operated relief valves (PORVs) shall be OPERABLE, with their setpoints selected to the low temperature mode of operation as follows:

a.

A setpoint of less than or equal to 350 psia shall be selected during heatup, cooldown and isothermal conditions when the temperature of any ReS cold leg is less than or equarlo 200*F.

b.

A setpoinl of less than or equal to 530 psia shall be selected during heatup. cooldown and isothermal conditions when the temperature of any ReS cold leg is greater than 200"F and less than or equal 10 300"F.

APPLICABIU1Y:

MODE 4 when the temperature of any ReS cold leg is less than or equa/to 300"F. MODE 5, and MODE 6 when the head is on the reactor vessel; and the RCS is not venled through greater than a 1.75 square inch vent.

ACTION:

a.

With one PORV inoperable in MODE 4, reslore the inoperable PORV to OPERABLE status within 7 days; or depressurize and vent the ReS through greater than a 1.75 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With one PORV inoperable in UODES 5 or 6, either (1) restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or (2) complete depressurization and venting of1he Res through greater than a 1.75 square inch vent within a total of 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months .

c.

With both PORVs inoperable. restore at least one PORV 10 operable status or complete depressurization and venting of the ReS through greater than a 1.75 SQuare inch vent within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

With the RCS vented per ACTIONS a, b, or c, verify Ihe vent pathway at least once per 31 days when the pathway is provided by a valve(s) that is locked, sealed. or otherwise secured in the open position; otherwise. verify the vent pathway everj12 hours.

e.

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

f.

The provisions of Specification 3.0.4 are not applicable.

SURVEil LANGE REQUIREMENTS 4.4.13 Each PORV shall be demonstrated OPERABLE by:

a.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

b.

Performance of a CHANNEL FUNCTION TEST, but excluding valve operation, at least once per 31 days; and

c.

Performance of a CHANNEL CALIBRATION at least once per 16 months.

ST. LUCIE - UNIT 1 3/44-59 Amendment No. 00. 84, ~,~. 213

REACTOR COOLANT SYSTEM REACTOR COOLANT PUMP - STARTING LIMITING CONDITIOIII FOR OPERATION.

3.4.14 If the steam generator temperature exceeds the primary temperature by more than 3OCF, the first idle reactor coolant pump shall not be started.

APPLICABILITY: MODES 4# and 5.

ACTION:

If a reactor coolant pump is started when the steam generator temperature exceeds primary temperature by more than 3OCF, evaluate the subsequent transient to determine compliance with Specification 3.4.9.1.

SURVEILLANCE REQUIREMENTS 4.4.14 Prior to starting a reactor coolant pump, verify that the steam generator temperature does not exceed primary temperature by more than 30°F.

Reactor Coolant System Cold Leg Temperature is less than 300"F.

ST. LUCIE - UNIT 1 3/44-60 Amendment NQ. 00, 3:\\., cW4, 213

3/4.5 EMERGENCY CORE COOL1NG SYSTEMS (ECCSI SAFETY INJECTION TANKS (Sin LIMITING CONDITION EQR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with:

a.

The isolation valve open,

b.

Between 1090 and 1170 cubic feet of borated water,

c.

A minimum boron concentration of 1900 ppm, and

d.

A nitrogen cover-pressure of between 230 and 280 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 .

SURVEILLAf\\.ICE REQUIREMENTS 4.5.1 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.

Wth pressurizer pressure.:: 1750 psia.

ST. LUCIE - UNIT 1 3145-1 Amendment No.~. 213

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - OPERATING LlMITINQ CONDITION FOR{lEERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high-pressure safety injection (HPSI) pump.

b.

One OPERABLE low-pressure safety injection pump.

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 ECGS 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 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s*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 condiUon 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.

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.

- With pressurizer pressure ~ 1750 psia.

ST. LUCIE - UNIT 1 3145-3 Amendment No.~, 43S, 464,~. 213

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

e.

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 a Safety Injection Actuation Signal.

2.

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

a.

High-Pressure Safety Injection Pumps.

b.

Low-Pressure Safety Injection Pumps.

c.

Charging Pumps.

3.

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

f.

By verifYing that each of the following pumps develops the specified total

.developed head when tested pursuant to the Inservice Testing Program.

1.

High-Pressure Safety Injection pumps.

2.

Low-Pressure Safety Injection pumps.

ST. LUCIE - UNIT 1 3145-5 AlnendrnentNo.lS.9Q,~.460.4G4.

~. 213

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER TANK lIMITll'jG CQI'jDITIOtLEOR OPERATIQN 3.5.4 The refueling water tank shall be OPERABLE with:

a.

A minimum contained volume 477,360 gallons of borated

water,

b.

A minimum boron concentration of 1900 ppm, C.

A maximum watertemperature of 100°F,

d.

A minimum water temperature of 55°F when in MODES 1 and 2,aOO

e.

A minimum water temperature of 40°F when in MODES 3 and 4 APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Wrth the refueling waler 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 SI:::IUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLAN<;E REQUIREMENTS 4.5.4 The RWT shall be demonstrated OPERABLE:

a.

Al least once per 7 days by:

1.

Verifying the water level 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, ST. LUCIE - UNIT 1 3145-8 Amendment No. ;,l8,:i!OO. 213

CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment intemal pressure shall be maintained between -0.7 and +0.5 psig.

APPLICABI LlTY: MODES 1, 2, 3 and 4.

ACTION:

With the containment internal pressure outside of the limits above, restore the intemal pressure to wrthin the limits 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 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT 1 3146-12 Amendmem No. 213

TABL.E 3.1-1 MAXIMUM AL.LQWABLE POWER LEVEL*HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETYVAL.YES DURING OPERATION WITH BOTH STEAM GENERATORS Maximum Allowable Power Maximum Number of Inoperable Safety L.ev!!I-Hfgh TrIp Setpolnt Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER) 88.5 2

79.8 3

66.5 ST. LUCIE - UNIT 1 3/47-2 Amendment No. 213

TABLE 4.7.1 STEAM LlN~, SAFETY VALVES PER LOOP VALVE NUMBER

.b.IFT SETTING*

Header A Header S

a.

8201 8205

~ 955.3 psig and:S 1015.3 psig

b.

8202 8206

~ 955.3 psig and ~ 1015.3 psig

c.

8203 8207

~ 955.3 psig and ~ 1015.3 pslg

d.

8204 8208

~ 95~.3 psig and ~ 1015.3 pslg

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 pslg

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 1 3/47*3 Amendment No. m, 213

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATiON 3.7.1.3 The condensate storage tank shall be OPERABLE with a minimum contained volume of 153,400 gallons.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the condensate storage tank. inoperable, restore the condensate storage tank. to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 .

SURYEILLANCE REQUIREMENTS 4.7.1.3 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water level.

ST. LUCIE - UNIT 1 3147-8 Amendment No. 213

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 onsile Class 1E distribution system, and

b.

Two separate and independent diesel generator sets each with:

1.

Engin~mounted fuel tanks containing a minimum of 152 gallons of fuel,

2.

A separate fuel storage system containing a minimum of 19,000 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 SHUTOOVVN 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 Jeast 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 4.8.1.1.2.a.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 SHUTDOVVN 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 auxJ1iary 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 OPERABIUTY.

ST. LUCIE - UNIT 1 3148-1 Amendment No..tOO, 442, -m,.:t:7Q, f80,213

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 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.

4.

Verifying that on an ESF actuation test signal (without Ioss-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.

5.

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 auto-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.

c)

Verifying that all automatic diesel generator trips.

except engine overs peed and generator differential.

are automatically bypassed upon loss of voltage on the emergency bus concurrent with a safety injection signal.

- - - This test may be conducted in accordance with the manufacturer's

. recommendations concerning engine prelube period.

ST. LUCIE - UNIT 1 3148-6 Amendment No. ea,400, 213

ELECTRICAL POWER SYSTEMS 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 set with:

1.

Engine-mounted fuel tanks containing a minimum of 152 gal/ons of fuel,

2.

A fuel storage system containing a minimum of 19,000 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 ALrERATIONS, operations involving positive reactivity additions that could result in*!oss of required SHlITDOWN MARGIN or boron concentration, movement of irradiated fuel, or crane operation with loads ov,er the wei storage pool. In addition, when in MODE 5 with the reactor coolant loops not filled, or inMOOE 6 with the water level less than 23 feet above the top of irradiated fuel assemblies seated within the reactor vessel, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.B.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 1 3/48-7 Amendment No. 4Qa,~, 4+9, 213

314.9 REFUELING OPERATIONS BORON CONCENTRA TTON LIMITING CONDITION FOR OPERATION 3.9.1 Wrth the reactor vessel head unbolted or 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 ~ 40 gpm of greater than or equal to 1900 ppm boron or its equivalent 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.

4.9.1.2 The boron concentration of The refueling cavity shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

ST. LUCIE - UNIT 1 3/49-1 Amendment No. 400, -tW.~. 213

REFU EUNG OPERATIONS SPENT FUEL STORAGE POOL lIM1I1NG CONDITION FOR OPERArlQN 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.

Wrth 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 the 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 lev.el 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 1 3149-11 Amendment No. m. 213

DELETED ST. LUCIE - UNIT 1 3/49-16 Amendment No.;14, 4Q. 213

3/4.10 SPECIAL TEST EXCEPTIONS 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 and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY: MODE 2.

ACTION:

8.

With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at 2: 40 gpm of 1900 ppm boric acid solution 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 2:. 40 gpm of 1900 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURYEILL ANCE 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.

ST. LUCIE - UNIT 1 31410-1 Amendment No. Y, 8+, 213

RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shaH be limited to less than or equal to 165,000 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 'imit, 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 518.9llCi/gram DOSE EQUIVALENT XE-133.

ST. LUCIE* UNIT 1 31411-15 Amendment No. 69, 00. 213

.DESIGN FEATURES 2.1.2 SHIELD BUILDING

a.

Minimum annular space;::: 4 feet

b.

Annulus nominal volume;::: 543,000 cubic feet

c.

Nominal outside height (measured from top of foundation base to the top of the dome) ;;:: 230.5 feet

d.

Nominal inside diameter = 148 feet

e.

Cylinder wall minimum thickness;;; 3 feet

f.

Dome minimum thickness =2.5 feet

g.

Dome inside radius = 112 feet DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment vessel is designed and shall be maintained for a maximum internal pressure of 44 psig and a temperature of 264°F.

PENETRATIONS 5.2.3 Penetrations through the containment structure are designed and shall be maintained in accordance with the original design provisions contained in Sections 3.8.2.1.10 and 6.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy clad fuel rods and/or poison rods, with fuel rods having an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.

5.3.1.1 Except for special test as authorized by the NRC, all fuel assemblies under control element assemblies shall be sleeved with a sleeve design previously approved by the NRC.

ST. LUCIE - UNIT 1 Amendment No. 34, 44, ~, -144,

~,213

CONTROL ELEMENT ASSEMBLIES j

5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies. The control element assemblies shall be deSigned and maintained in accordance with the original design provisions contained in Section 4.2.3.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requ irements.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b.

For a pressure of 2485 psig, and

c.

For a temperature of 650°F. except for the pressurizer which is 700°F.

5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.6 FU EL STORAGE CRITICALITY 5.6.1.a The spent fuel pool and spent fuel storage racks shall be maintained with:

1.

kelf less than 1.0 when fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.

2.

A nominal 1 0.12 inches center to center distance between fuel assemblies in Region 1 of the spent fuel pool storage racks, a nominal 10.30 inches center to center distance between fuel assemblies in the Region 1 cask pit storage rack, and a nominal 8.86 inches center to center distance between fuel assemblies in Region 2 of the spent fuel pool storage racks.

3.

A ketr less than or equal to 0.95 when flooded with water containing 500 ppm boron, including an allowance for biases and uncertainties as described in Section 9.1 ofthe Updated Final Safety Analysis Report.

ST. LUCIE - UNIT 1 5-5 Amendment No. ~, ~, 75, 94-,

446, 4i3, 4-93, 213

DESIGN FEATURES CRITICALITY (Continued)

4.

For storage of enriched fuel assemblies, requirements of Criteria in 5.6.1.a.1 and 5.6.1.a.3 shall be met by positioning fuel in the spent fuel storage racks consistent with the requirements of Specification5.6.1.c.

5.

Vessel Flux Reduction Assemblies (VFRAs), as defined in Section 9.1 of the Updated Final Safety Analysis Report, may be placed in any allowable fuel storage location.

6.

Fissile material, not contained in a fuel assembly lattice, shall be stored in accordance with the requirements of Criteria in 5.6.1.a.1 and 5.6.1.a.3.

7.

The Metamic neutron absorber inserts shall have a 1°B areal density greater than or equal to 0.015 grams 1°B/cmz.

b. The Region 1 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 storage racks shall be controlled as described below. Criteria in 5.6.1.c.2, 5.6.1.c.3, 5.6.1.c.5 and 5.6.1.c.6 do not apply to the Region 1 cask pit storage rack.

1.

The maximum initial planar average U-235 enrichment of any fuel assembly inserted in a spent fuel 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 patterns and alignment restrictions of Figure 5.6-1 and the minimum burnup requirements of Table 5.6-1.

3.

Fuel placed in Region 2 of the spent fuel pool storage racks shall comply with the storage patterns or allowed special arrangements of Figure 5.6-2 and the minimum burnup requirements of Table 5.6-1. The allowed special arrangement for fresh fuel may be repeated, provided the applicable interface requirements specified by the safety analysis are met.

4.

Any fuel satisfying criteria 5.6.1.c.1, including fresh fuel, may be placed in the Region 1 cask pit storage rack.

5.

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

6.

Any 2x2 array of Region 2 storage cells that interface with Region 1 shall comply with the rules of Figure 5.6-3. The allowed special arrangement in Region 2 as shown in Figure 5.6-2 shall not be placed adjacent to Region 1.

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 ken of less than or equal to 0.98 under the most reactive condition.

ST. LUCIE - UNIT 1 5-6 Amendment No. 4+, ~, :!4, 34,

+a. 9-1-, 92, ~,m3, 213

Allowable Checkerboard Storage Patterns (See Notes 4 and 5)

Pattem "A" See Note 1 Pattem "S" See Note 2 Pattem "C" See Note 3 L:JEJ DO EJL:J DO NOTES:

1.

F represents Fresh Fuel. WH represents an empty cell. Allowable Pattem is Fresh Fuel checkerboarded with empty cells. Diagram is for illustration only.

2.

Numbering denotes fuel assembly type. Minimunrbumup for fuel assembly type 1 is defined in Table 5.6-1. Allowable pattem is at least one insert [either Metamic or full-length full-strength-CEA] in anyone of the 2x2 array locations. Diagram is for illustration only.

3.

Numbering denotes fuel assembly type. WH represents an empty cell. Minimum burnup for fuel assembly type 2 is defined in Table 5.6-1. Allowable pattern is at least one empty cell in any of the 2x2 array locations. Diagram is for illustration only.

4.

The storage arrangements of fuel within a rack module may contain more than one pattem.

Each cell is a part of up to four 2x2 arrays, and each cell must simultaneously meet the requirements of all those arrays of which it is a part.

5.

Empty cells within any pattem are acceptable.

FIGURE 5.6-1 Allowable Region 1 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT 1 Amendment No. ~. 493. 213

ALLOWED SPECIAL ARRANGEMENT Fresh Fuel Assemblies in Pattern "C n

, "0", or "E" Racks

[!] =

FRESH FUEL ASSEMBLY

=

EMPTY CELL ALLOWABLE CHECKERBOARD STORAGE PATTERNS (See Notes 4 and 5)

Pattern *C*

Pattern *D~

Pattern*P' See Note 1 See Note 2 See No1e3 OR OR FIGURE 5.6*2 (Sheet 1 of 2)

Allowable Region 2 Storage Patterns and Fuel Alignments ST. LUCIE - UNIT 1 Amendment No. 493,213

NOTES to Figure 5.6-2 NOTES:

1.

Numbering denotes fuel assembly type. WH represents an empty cell. Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1. Allowable pattern is at least one empty cell in any of the 2x2 array locations. Diagram is for illustration only.

2.

Numbering denotes fuel assembly type. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1. Allowable pattern is at least two inserts, (either Metamic or full-length, full-strength CEA) in the 2:x2 array. Diagrams are for illustration only.

3.

Numbering denotes fuel assembly type. Minimum bumup for fuel assembly type 5 is defined in Table 5.6-1. Allowable pattern is one insert, (either Metamic or full-length, full-strength CEA) in the 2x2 array. Diagrams are for illustration only.

4.

The storage arrangements of fuel within a rack module may contain more than one pattern.

Each cell is a part of up to fOUf 2x2 arrays, and each cell must simunaneously meet the requirements of all those arrays of which it is a part.

5.

Empty cells within any pattern are acceptable.

AGURE 5.6--2 (Sheet 2 of 2)

Allowable Region 2 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT t Amendment No. 493, 213

Allowed Region 2 to Region 1 Fuel Alignments (See Notes 4 and 5)

DDDD Region 2 D@][:J@]

Interface of Region 2 Pattern "C~ with Region 1 See Note 1 Region 1 DDDD wDwD O~[J~

Region 2

[]O[][J Interface of Region 2 Pattern "0" with Region 1 See Note 2 Region 1 DDDD

~[]~[]

rJDDD Region 2

~D~D Interface of Region 2 Pattern "E" with Region 1 See Note 3 rlDDD Region 1

~[]~[]

FIGURE 5.6-3 (Sheet 1 of 2)

Region 2 Interface Requirements with Region 1 ST. LUCIE - UNIT 1 Amendment No. :j.Q3, 213

NOTES to Figure 5.6-3 NOTES:

1.

WH represents an empty cell. For the interface of Pattern *C* with Region 1, the empty cell must be on the rack periphery facing Region 1 racks. ~iagrams are for illustration only.

2.

For the interface of pattern "0" with Region 1. at least one cell on the rack periphery facing Region 1 rack must contain an insert (either Metamic of full-length full-strength CEA) in the 2x2 array. If the insert is Metamic. the insert must be oriented so that the corner of 1he L shape is located closest to the Region 1 rack. Diagram is for illustration only.

3.

For the interface of Pattern -E* with Region 1, the insert must be on the rack periphery facing the Region 1 rack. The insert may be either a Metamlc of full-length full strength CEA If the insert is Metamic, the insert must be oriented so that the corner of the L-shape is located closest to the Region 1 rack. Diagram is for illustration only.

4.

Empty cells with any pattern are acceptable.

5.

There are no interface requirements within Region 1. Any Pattern within Region 1 may be used for the interface. Pattern'S" was used only as an illustration.

FIGURE 5.6-3 (Sheet 2 of 2)

Region 2 Interface requirements with Region 1 ST. LUCIE - UNIT 1 5.-Qf Amendment No. 213

TABLE 5.6-1 Minimum Burnup as a Function of Enrichment Cooling Time Coefficients Fuel Type (Years)

A B

C 1

0

-36.6860 I

22.4942

-1.4413 2

I 0

-36.1742 16.6000

-D.S958 I

3 0

-34.7091 I

23.1361

-1.6204 0

-24.5145 i

21.3404

-1.2444 2.5

-26.8311 22.5246

-1.5029 5

-24.7233 I

20.9763

-1.3246 4

10

-2362a 19.9541

-1.2505 15

-23.545 19.9336

-1.3180 20

-22.4382 19.2460

-1.2629 0

I

-8.1856 14.5275

-0.0719 i

2.5 I

-11.8506 16.1475

-0.3969 5

I

16.5196 18.5309

-0.7837 5

10

-13.6831 16.3475

-0.5844 15

-12.5819 i

15.6175

-0.5656 i

I 20

-12.6469 16.4575

-0.5906 NOTES:

1.

To qualify in a "fuel type: the burn up of a fuel assembly must exceed the minimum bumup "BU" calculated by inserting the *coefficients* for the associated "fuel type' and "cooling time" into the polynomial function:

BU = A+ 8*E +

> where' BU Minimum Burnup (GWO/MTU)

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

A. B. C ::: Coefficients

2.

Interpolation between values ofcooljng time is not permitted.

ST. LUCIE - UNrT 1 5-69 Amendment No.213

ADMlNJSTRATIVE CONTROLS (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 jf 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-Based Containment Leak-Test Program," as modified by the following exception{s):

a)

Bechtel Topical Report. BN-TOP-l or ANS 56.8-1994 (as recommended by R.G. 1.163) will be used for type A testing.

b)

The first Type A test performed after the May 1993 Type A test shall be no tater than May 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident Pa, is 42.8 pSig. The containment design pressure is 44 psig.

The maximum allowed containment leakage rate, La. at Pa, shall be 0.50% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

Containment leakage rate acceptance criterion is:::.. 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type Band C tests,

.. 0.75 La for Type A tests, and:::.. 0.096 La for secondary containment bypass leakage paths.

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is:::.. 0.05 La when tested at ~ Pa.

2)

For the personnel air lock door seal. leakage rate is < 0.01 La when pressurized to ~ 1.0 Pa.

3)

For the emergency air lock door seal. leakage rate is < 0.01 La when pressurized to ~ 10 psig.

ST. LUCIE - UNIT 1 6-15b Amendment No.~. 36,423. 449, 43+,49+,200.213

ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (continued) 6.9.1.9 At [east 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 Full Length CEA Position Misalignment> 15 inches Specification 3.1.3.6 Regulating CEA Insertion Limits SpeCification 3.2.1 Linear Heat Rate T

Specification 3.2.3 Total Integrated Radial Peaking Factor - Fr 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, approved Revisions and Supplements as specified in the COLR.

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 & Ught Company, January 1995.

3.

XN-75-27(A) [also issued as XN~NF-75~27(A)], "Exxon Nuclear Neutronic(s) Design Methods for Pressurized Water Reactors"

4.

DELETED

5.

XN~NF~2-21(P)(A). "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations*

ST. LUCIE* UNIT 1 6*19 Amendment No. SQ, 99, oo.. ~,

400.463, m, 494, 213

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

6.

DELETED

7.

XN-75-32(P){A). *Computational Procedure for Evaluating Fuel Rod Bowing"

8.

DELETED

9.

XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors"

10. XN-NF-621 (P){A).

Exxon Nuclear DNB Correlation for PWR Fuel Designs"

11. DELETED

12. XN-NF-82*06(P)(A), "Qualification of EXXon Nuclear Fuel for Extended Bumup'

13. ANF-88-133(P)(A), "Qualification of Advanced Nudear Fuels' PWR Design Methodology for Rod Bumups of 62 GWd/MTU'

14. XN-NF-85-92 (P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results*

15. DELETED

16. DELETED

17. EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Design"

18. EMF-92-153(P)(A). "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel"

19. EMF-96-o29(P)(A). Volumes 1 and 2. "Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2 - Benchmarking Results" ST. LUCIE - UNIT 1 6-19a Amendment No. ~. 466,~. 213

ADMINiSTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

20. EMF-1961 (P)(A), "Statistical SetpointfTransient Methodology for Combustion Engineering Type Reactors"

21. EMF-2310(P)(A). nSRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors: Revision 1, as supplemented by ANP-3000{P). *St. Lucie Unit 1 EPU - Information to Support License Amendment Request: Revision O.

22. EMF-2328(P)(A), *PWR Small Break LOCA Evaluation Model, S-RELAP5 Based,' Revision 0, as supplemented by ANP-3000(P). -St. Lucie Unit 1 EPU - Information to Support License Amendment Request: Revision O.

23. EMF-21 03(P)(A), "Realistic Large Break LOCA Methodology for Pressurized water Reactors: Revision 0, as supplemented by ANP-2903(P), *St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA summary Report with Zr-4 Fuel Cladding.* Revision 1.

ST. LUCIE - UNIT 1 6-19b Amendment No. 493, 4+4. 494, 213

v t e Letter Sequence Other
TAC:ME5091, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Supplement Administration Questions 2 May 2012 19 April 2011 Answers 9 July 2012 Results Approval, Approval, Approval Other: L-2012-309, FPL Review Comments on Extended Power Uprate Safety Evaluation, L-2012-382, Extended Power Uprate Cycle 24 Startup Report, ML11119A084, ML11119A085, ML112360230, ML12053A345, ML12165A511, ML12191A236, ML12194A287, ML12195A168
MONTHYEAR2011-09-30 [Table View]

ML12191A236

Text

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