Text

DPR-67 Appendix a
	ML032731661
Person / Time
Site:	Saint Lucie
Issue date:	10/02/2003
From:	Anand R; NRC/NRR/DRIP/RLEP
To:	Stall J; Florida Power & Light Co
	Anand R NRR/DRIP/RLEP, 415-1146
Shared Package
ML032870058	List: ML032731661; ML032800813; ML032800846; ML032800895; ML032800909; ML032800961; ML032870081;
References
	Download: ML032731661 (321)
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ST. LUCIE PLANT UNIT I TECHNICAL SPECIFICATIONS APPENDIX "A" TO LICENSE NO. DPR-67 P~oAim A

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INDEX DEFINTONS SECTION PAGE 1.0 DEFINITIONS 1.1 Action ..................................................... 1-1 1.2 Axial Shape Index .1-1 1.3 Azimuthal Power ilt .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 E Average Disintegration Energy .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 Low Temperature RCS Overpressure Protection Range ..... ............ 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 1X I Amcnent No. 2, A, 66,

INDEX DEFINITIONS SECTION PAGE 1.23 Process Control Program (PCP) ............................ 1-5 1.24 Purge - Purging .................... 1-5 1.25 Rated Thernmal Power ..... 1-6 1.26 Reactor Trip System Response Time.... 1-6 1.27 Reportable Event ..... 1-6 1.28 Shield Building Integrity ..... 1-6 1.29 Shutdown Margin ..... 1-6 1.30 Site Boundary ..... 1-6 1.31 Source Check ..... 1-6 1.32 Staggered Test Basis ..... 1-7 1.33 Thermal Power . . . ......................... . 1-7 1.34 Unidentified Leakage . . . ................... 1-7 1.35 Unrestricted Area ..... 1-7 1.36 Unrodded Integrated Radial Peaking Factor - Fr........... 1-7 1.37 DELETED 1-7 TABLE 1.1 TABLE 1.2 Frequency Notation ...... .......... - - ............

Operational Modes.....................................

1a 1-9 I Ia Amendment No. 1 34 r9,60,69,T4 9Q, ST. LUCIE - UNIT I

INDEX SAFETY LIMITS AND LIMING SAFETY SYSTEM SETINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core ............................................................................................................ 2 1 Reactor Coolant System Pressure.2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoints .2-3 ST. LUCIE - UNIT 1 1l Amendment No. 176

w e [o In E-LIMITING CONDITIONS FOR OPERATTON A4ND SURVEJLL~ CE TEOU..EMENTS SECTION PAGE 3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS S 3/4.1.1 BORATION CONTROL . . . . . . . . . . . . 3/4 1-1 Shutdown Margin - Tee > 2006F . . . . . 3/4 1-1 I

Shutdown Margin - Tea - 2008F . 3/4 1-3 Boron Dilution . . . . . . . . . . . . . 3/4 1-4 Moderator Temperature Coefficient . . . . 3/4 1-5 Minimum Temperature for Criticality 3/4 1-7 3/4.1.2 BORATION SYSTEMS . . . . . . . . . . . . 3/4 1-8 Flow Paths - Shutdown . . . . . . . . . . 3/4 1-8 Flow Paths - Operating . . . . . . . . . 3/4 1-10 Charging Pumps - Shutdown . . . . . . . . 3/4 1-12 Charging Pumps - Operating . . . . . . . 3/4 1-13 Boric Acid Pumps - Shutdown . . . . . . . 3/4 1-14 Boric Acid Pumps - Operating . . . . . . 3/4 1-15 Borated Water Sources - Shutdown . 3/4 1-16 Borated Water Sources - Operating . .34 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES . . . . . . . 3/4 1-20 Full Length CEA Position . . . . . . . .*.. . . . . . . . . 3/4 1-20 Position Indicator Channels . . . . . . . . . . . . . . . . 3/4 1-24 CEA Drop Time . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-26 Shutdown CEA Insertion Limit . . . . . . . . . . . . . . . 3/4 1-27 Regulating CEA Insertion Limits . . . . . . . . . . . . . . 3/4 1-28 ST. LUCIE - UNIT 1 III Amendment No. 77, A44; 134

IND4EX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE .................................................. 3/42-1 3/4.2.2 DELETED ................................................ 3/4 2-6 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR - Fr 3........................./......

34 2-9 3/4.2.4 AZIMUTHAL POWER TILT - Tq ........................................... 3/4 2-11 3/4.2.5 DNB PARAMETERS ........................................... 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION .......................... ......... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION....................................................................................... 3/43-9 3/4.3.3 MONITORING INSTRUMENTATION .......................... 314 3-21.

Radiation Monitoring .......................... 3/4 3-21 Remote Shutdown Instrumentation .......................... 3/4 3-33 Accident Monitoring Instrumentation............................................................... 3/4 3-41 314.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION ............ ......... 314 4-1 3/4.4.2 DELETED ................................................ 3/4 l 4-2....

3/4.4.3 SAFETY VALVES - OPERATING ............................................... 3/4 4-3 ST. LUCIE - UNIT 1 IV Amendment No. 27,3Z, 3X , ,5

&9, 449, 4a, 434, -*S, 4w, 44z, 1 66

I'IJ DEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS _

SECTION PAGE 3/4.4.4 PRESSURIZER........................................... 3/4 4-4 3/4.4.5 STEAM GENERATORS....................................... 3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE......................... 3/4 4-12 Leakage Detection Systems .............................. 3/4 4-12 Reactor Coolant System Leakage......................... 3/4 4-14 3/4.4.7 CHEMISTRY.............................................. 3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY...................................... 3/4 4-17 3/4.4.9 PRESSURE/TEMPERATURE LIMITS............................ 3/4 4-21 Reactor Coolant System................................. 3/4 4-2t Pressurizer........................................... 3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY................................... 3/4 4-26 3/4.4.11 ASME Code Class 1, 2, and 3 Components................. 3/4 4-26 DELETED................................................ 3/4 4-56 I

3/4.4.12 PORY BLOCK VALVES...................................... 3/4 4-58 3/4.4.13 POWER OPERATED RELIEF VALVES........................... 3/4 4-59 3/4.4.14 REACTOR COOLANT PUMP - STARTING........................ 3/4 4-60 3/4.4.15 REACTOR COOLANT SYSTEM VENTS........................... 3/4 4-61 314.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tayg > 325 0F......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 325*F......................... 3/4 5-7 3/4.5.4 REFUELING WATER TANK................................... 314 5-8 ST. LUCIE-UNIT 1 V Amendment No. ZB9 00, 4ft.8%

134

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT VESSEL ................................... 3/4 6-1 Containment Vessel Integrity. . 3/4 6-1 Containment Leakage .................................. 3/4 6-2 Containment Air Locks ................................ 3/4 6-10 Internal Pressure ..................... 3/4 6-12 Air Temperature ...................... 3/4 6-13 Containment Vessel Structural Integrity . . 3/4 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ................. 3/4 6-15 Containment Spray and Cooling Systems. ........... 3/4 6-15 Spray Additive System ................................ 3/4 6-16a 3/4.6.3 CONTAINMENT ISOLATION VALVES ......................... 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL .............................. 3/4 6-23 Hydrogen Analyzers ................................. 3/4 6-23 Electric Hydrogen Recombiners - W ................... 3/4 6-24 3/4.6.5 VACUUM RELIEF VALVES ................................. 3/4 6-26 3/4.6.6 SECONDARY CONTAINMENT ................................. 3/4 6-27 Shield Building Ventilation System . ............ 3/4 6-27 Shield Building Integrity............................ 3/4 6-30 Shield Building Structural Integrity . . 3/4 6-31 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ........ ......................... 3/4 7-1 Safety Valves................................ . 3/4 7-1 Auxiliary Feedwater System ........................... 3/4 7-4 Condensate Storage Tank .................. 3/4 7-6 Activity............................................. 314 7-7 Main Steam Line Isolation Valves . . 3/4 7-9 I

ST. LUCIE - UNIT I VI Amendment No.ff4.'1,134

IND-EX LIMMN(: CONOmONS FOR OPFRATION AND SURVEII LANCE REOUIRFMFS SECQLN PAGE 34.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION .....................3a/4 7-13 34.7.3 COMPONENT COOLING WATER SYSTEM.3/4 7-14 3/4.7.4 INTAKE COOLING WATER SYSTEM .........................................34 7-16 3/4.7.5 ULTIMATE HEAT SINK ..... 47-18 3.4.7.6 DELETED 314.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ..............................

3/4 7-20 3/4.7.8 ECCS AREA VENTILATION SYSTEM .........................................

347-24 314.7.9 SEALED SOURCE CONTAMINAON ......................................... 47-27 3/4.7.10 SNUBBERS ......................................... /47-29 34.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES .... 48-1 Operating ... 3/48-1 Shutdown . ... . 34 8-7 34.82 ONSITE POWER DISTRIBUTION SYSTEMS ... 348-A.C. Distribution - Operating .. . 3 /4&-8 A.C. Distribution - Shutdown. 4... . 348 D.C. Distribution - Operating . . . 348-10 D.C. Distribution - Shutdown . ... 348-13 ST. WCIE - UNIT 1 Vii Amendment No. SE, 4466, 484.4, 152

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLAUICE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ............. ..... .... . 3/4 9-1 3/4.9.2 INSTRUMENTATION ............. .. .... .......... 314 9-2 3/4.9.3 DECAY TIME ....... ;, ...................... ........ 314 9-3 3/4.9.4 CONTAINMENT PENETRATIONS ....... .................. 3/4 9-4 3/4.9.5 COVUHUNICATIONS .... ....... - - .- ................ 314 9-S 3/4.9.6 MANIPULATOR CRANE OPERABILITY. ..... 314 9-6 3/4.9.7 CRANE TRAVEL-- SPENT FUEL STORAGE POOL BUILDING. 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULA TIO.. .........o.. 3/4 9-8 I 3/4.9.9 CONITAINMENT ISOLATION SYSTEM ......................... _ 314 9-9 3/4.9.10 WATER LEVEL - REACTOR VESSEL. . ... 3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL ....... ; ......... 3/4 9-11 3/4.9.12 FUEL POOL VENTILATION SYSTEM - FUEL STORAGE . .3/4 9-12 3/4.9.13' SPENT FUEL CASK CRANE... . ... 3/4 9-15 3/4.9.14 DECAY TIME - STORAGE POOL ...... ............... 3/4 9-16 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN{......................................... 3/4 tC-l 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS ......................................... 314 10-2 3/4.10.3 DELETED . ........ 3/4 10-3' 3/4.10.4 DELETED .............. 314 10-4 3/4.10.5 CENTER CEA 4ISALIGNMENT ....... 3/4 10-5 ST. LUCIE - Ut:IT I VEIII Amendment No. !4. ZA, Z5. 77. 5 6

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2.5 EXPLOSIVE GAS MIXTURE .......................... 3/4 11-14 3/4.11.2.6 GAS STORAGE TANKS ................. 3/4 11-15 ST. LUCIE - UNIT I VIIIa Amendment No. S9,123

INDEX ST. LUCIE - UNIT I IX Amendment No. 24i, i. 434. 176

INDEX ST. LUCIE - UNIT 1 X Amendment No. a, N. w, 6g, 84, 4 3 4 ;176

INDEX ST. LUCIE- UNIT 1 Xi Amendmen :g.26,44, 66,

INDEX ST. LUCIE - UNIT 1 MIl Amendment No.4, 24, 2,63, 4344,76

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area ................................. 5-1 Low Population Zone ............................................... 5-1 5.2 CONTAINMENT Configuration .5-1 Design. Pressure and Temperature. 5-4 Penetrations. .................................................. 5-4 5.3 REACTOR CORE Fuel Assemblies .5-4

- Control Element Assemblies ................................. s-5.4 REACTOR COOLANT SYSTEM

~- - Design Pressure and Temperature .5-5 Volume ................................................ ....... 5-5 5.5 EMERGENCY CORE COOLING SYSTEMS .5-5 5.6 FUEL STORAGE Criticality .5-5 Drainage. ...................................................... 5-6 5.7 SEISMIC CLASSIFICATION ........................................ 5-6 5.8 METEOROLOGICAL TOWER LOCATION .5-6a 5.9 COMPONENT CYCLE OR TRANSIENT LAlMUlI. 5-6a ST. LUCIE-LUNIT 1 Xi II AmendmentfNo.+S4, 142

NDEX ADMINISTRATIVE CONTROLS SECTION - t PAGE 6.1 RESPONSIBILITY ........................... 6-1 6.2 ORGANIZATION 6.2.1 ONSITE ADo OFFSITE ORGANIZATION .......... 6-1 6.2.2 UNIT STAFF .......... 6-2 6.2.3 SHIFT TECHNICAL ADVISOR .......... 6-5 6.3 UNIT STAFF QUALIFICATIONS...................... 6-6 6.4 TRAINING .......... 6-6 6.5 REVIEW AND AUDIT 6.5.1 FACILITY REVIEW GROUP Function.................................................... 6-6 Composition................................................. 6-6 Alternates................................................. 6-7 Meeting Frequency .. 6-7 Quorum..................................................... 6-7 Responsibilities........................................... 6-7 Authority.................................................. 6-8 Records..................................................... 6-8 6.5.2 COMPANY NUCLEAR REVIEW BOARD Function.................................................... 6-9 Composition................................................. 6-9 Alternates.................................................. 6-9 Consultants. ...................................... 6-10 Meeting Frequency .6-10 Quorum...................................................... 6-10 Review...................................................... 6-10 Audits...................................................... 6-11 Authority................................................... 6-12 Records..................................................... 6-12 ST. LUCIE - UNIT I XIV Amendment No. OJ1,34

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION .......................... 6-12 6.7 SAFETY LIMIT VIOLATION .......................... 6-12 6.8 PROCEDURES AND PROGRAMS .................................................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ..................... 6-15d Startup Report ..................... 6-15d Annual Reports ................... 6-16 Monthly Operating Reports .............................................. ......... 6.....................

6-16a Annual Radioactive Effluent Release Report ......... 6-17 Annual Radiological Environmental Operating Report ......... 6-18 Core Operating Limits Report (COLR) ......... 6-19 6.9.2 SPECIAL REPORTS ......... 6-19c 6.10 RECORD RETENTION ......... 6-20 6.11 RADIATION PROTECTION PROGRAM ......... 6-21 6.12 HIGH RADIATION AREA ......... 6-22 6.13 PROCESS CONTROL PROGRAM ...................................... 2.....................3.......

623 6.14 OFFSITE DOSE CALCULATION MANUAL ......... 6-23 ST. LUCIE - UNIT 1 XV Amendment No. 27, 3, 59, 69, 4-3 434, 450, 4-74. 176

SECTION 1.0 DEFINITIONS

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1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

AXIAL SHAPE INDEX 1.2 The AXIAL SHAPE INDEX (YE) is the power level detected by the lower excore nuclear instrument detectors (L) less the power level detected by the upper excore nuclear instrument detectors (U) dividied by the sum of these power levels. The AXIAL SHAPE INDEX (YI) used for the trip and pretrip signals in the reactc- protection system is the above value (YE) modified by an appropriate multiplier (A) and a constant (B) to determine the true core axial power distribution for that channel.

Y =L-U YI AYE + B AZIMUTHAL POWER TILT - T 1.3 AZIMUTHAL POWER TILT shall be the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core.

AZIMUTHAL POWER TILT = Power in any core quadrant (uLper or lower)

AZIMTHA POWR TLT maxAveagepower of all quadrants (upper or lower)J CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK -

1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include,where possible, comparison of the channel indication and/or status with other indica-tions and/or status derived from independent instrument channels measuring the same parameter.

ST. LUCIE - UNIT I 1 -1 Amendment No. Z3J0J$, 69

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into

- the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT VESSEL INTEGRITY 1.7 CONTAINMENT VESSEL INTEGRITY shall exist when:

a. All containment vessel penetrations required to be closed during accident conditions are either

1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed position except for valves that are open on an intermittent basis under administrative control.

b. All containment vessel equipment hatches are closed and sealed,

c. Each containment vessel air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows or O-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be the seal water flow supplied from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Exceptions to the above include shared (4 fingered) control element assemblies (CEAs) withdrawn into the upper guide structure (UGS) or evolutions performed with the UGS in place such as CEA latching/unlatching or verification of latchingf unlatching which do not constitute a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT (COLRI 1.9a The COLR is the unit-specific document that provides cycle specific parameter limits for the current operating reload cycle. These cycle-specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11. Plant operation within these limits is addressed in individual Specifications.

ST. LUCIE - UNIT 1 1-2 Amer d~ent No. 6G, 446.

DEFINITIONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (pCi/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 ICRP-30, Supplement to Part 1, pages 192-212, Tables entitled, wCommitted Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity(Sv/Bq)."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E 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 tota; non-iodine activity in the coolant.

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.

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.

ST. LUCIE - UNIT 1 1-3 Amendment No. ;. 69, 163

DEFINITIONS 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.

LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE 1.16 The LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE is that operating condition when (1) the cold leg temperature is < 304'F during heatup or

< 281'F during cooldown and (2) the Reactor Coolant System has pressure boundary integrity. The Reactor Coolant System does not have pressure boundary integrity when the Reactor Coolant System is open to containment and the minimum area of the Reactor Coolant System opening is greater than 1.75 square inches.

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 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. 69,60,69,81,194,443, 125

DEFINITIONS 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 functionfs) are also capable of performing their related support function(s).

OPERATIONAL 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.

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube 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.

ST. LUCIE - UNIT I 1-5 Amendment No. 2T69, 123

DEFINITIONS 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 is interrupted to the CEA drive mechanism.

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.27XJ9.0f

DEFINITIONS STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.35 Unrestricted area means an area, access to which is neither limited nor controlled by the licensee.

UNRODDED INTEGRATED RADIAL PEAKING FACTOR - F r

1.36 The UNRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the peak pin power to the average pin power in an unrodded core, excluding tilt.

ST. LUCIE - UNIT I 1-7 Amendment No. 60,6, 109, 125

TABLE 1.1 I FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months w At least once per 7 days 4/M* At least 4 per month at intervals of no greater than 9 days and a minimum of 48 per year M At least once per 31 days Q At least once per 92 days SA At least once per 184 days R At least once per 18 months I'

S/U Prior to each reactor startup p*. Completed prior to each release N.A. Not applicable

For Radioactive Effluent Sampling

- For Radioactive Batch Releases Only II ST. LUCIE - UNIT 1 1-8 Amendment No.,Vt1/,69 l

TABLE 1.2 I OPERATIONAL MODES REACTIVITY %RATED AVERAGE COOLANT MODE CONDITION, Keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION > 0.99 , 5% > 3258 F

2. STARTUP > 0.99 < 5% > 325 0 F

3. HOT STANDBY s 0.99 0 > 325 0F

4. HOT SHUTDOWN 4 0.99 0 325 0F > T avg

, 200 0F

5. COLD SHUTDOWN < 0.99 0 < 200'F

6. REFUELING** < 0.95 0 < 140'F

Lxcluding decay beat.

- Fuel in the-reactor vessel with the vessel head closure bolts less than I fuTly tensioned or with the head removed.

ST. LUCIE - UNIT 1 1 -9 Amendment No. Y 69

/r SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.

2.0 SAFETY IKITS APO LIMITING SAFETY SYSTEo SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and taxi- I mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the coctination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES I and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psla.

be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

Ane^._nert rNO. '-J :1 5 ST. LUCIE - UNIT 1 2-1

( C (

600 I I I I I I I UNACCEPTABLE OPERATION 550 06LE OPERATION LL v- 560 SURE IN PSIA a

0 0:

e 540 ZD REACTOR OPERATION LIMITED TO LESS THAN 580 F BY ACTUATION OF THE I-J MAIN STEAM LINE SAFETY VALVES

.4-_

520 VESSEL FLOW LESS MEASUREMENT 0 UNCERTAINTIES - 365.000 GPM E LIMITS CONTAIN NO ALLOWANCE FOR INSTRUMENT ERROR OR FLUCTUATIONS 500 x BASED ON THE AXIAL SHAPE ON FIGURE B 2.1 -1 \ _ _

0 l1750 450 THERMAL POWER LIMITED TO-ACCEPTABLE OPERATION A MAXIMUM OF 112% OF RATED THERMAL POWER BY THE HIGH POWER TRIP LEVEL 460 I I- I k .. .

.XI I

.0 .2 .4 .6 .6 1.0 1.2 1.4 1.G Fraction of Rated Thermal Power (D/OPS/TS(UNIT I 2*2)no)

FIGURE 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY LIMIT -

FOUR REACTOR COOLANT PUMPS OPERATING ST. LUCIE - UNIT I 2-2 Amendment No. 48, 48, +45, 163

.7 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1.

ACTION:

With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement require-ment of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

ST. LUCIE - UNIT 1 2-3

( (

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. Power Level - High (1)

Four Reactor Coolant Pumps Operating c 9.61% above THERMAL POWER, with <9.61% above THERMAL POWER, and a minimum setpoint of 15% of RATED a minimum setpoint of 15% of RATED THERMAL POWER, and a maximum of THERMAL POWER and a maximum of

< 107.0% of RATED THERMAL < 107.0% of RATED THERMAL POWER. POWER.

3. Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps Operating 2 95% of design reactor coolant flow with > 95% of design reactor coolant flow with 4 pumps operating 4 pumps operating *

4. Pressurizer Pressure - High < 2400 psia <2400 psia

5. Containment Pressure - High < 3.3 psig <3.3 psig

6. Steam Generator Pressure - Low (2) 600 psia 600 psia

7. Steam Generator Water Level - Low > 20.5% Water Level - each steam 2 19.5% Water Level - each steam generator generator

8. 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.

Design reactor coolant flow with 4 pumps operating is 365,000 gpm.

ST. LUCIE - UNIT I 2-4 Amendment No. 3, 4, ae, ,45,405,a43. 45, 163

( ( (

TABLE 2.2-1 (Continued)

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

9. Thermal Margin/Low Pressure (1)

Four Reactor Coolant Pumps Operating Trip setpoint adjusted to not exceed the limit Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4. lines of Figures 2.2-3 and 2.2-4.

9a. Steam Generator Pressure Difference High (1) < 135 psid < 135 psid (logic In TM/LP)

10. Loss of Turbine - Hydraulic Fluid Pressure - Low (3) > 800 psig > 800 psig

11. Rate of Change of Power - High (4) < 2.49 decades per minute < 2.49 decades per minute TABLE NOTATION (1) Trip may be bypassed below 1%of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is > 1%of RATED THERMAL POWER.

(2) Trip may be manually bypassed below 685 psig; bypass shall be automatically removed at or above 685 psig.

(3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when Power Range Neutron Flux power Is

>_15% of RATED THERMAL POWER.

(4) Trip may be bypassed below 10 % and above 15% of RATED THERMAL POWER.

ST. LUCIE - UNIT 1 2-5 Amendment No. 2, 43, 45,9159

of of 1.0 I .. . . . . . . . . .

0.8 .. .... . . - . . .

- .1. -.... L OPE 0.6 2- .- .

OR 2 6r_/

0.4

-. ---. - I . -~~~~~~~~Z 7 t - -

02 0

0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER FIGURE 2.2-1 Local Power Density - High Trip Setpoint Part 1 (Fraction of RATED THERMAL POWER Versus QR2 )

ST. LUCIE - UNIT 1 2-6

1.4 -

1.3 UMACCEZABU MIACCETABLE 1.2 - OPER t.,O. 1.17 ) OpZiON 1.11 8

god . 0) z 0.4 0 .

0.9 ~~~ ~~~~~~~~~~~~~

0.3~ r0 00 0.2 0.1 0--

-0.7 -0.5 -0.3 -0.1 0.1 0.3 0.5 0.7 MAXIL SHAPE INDEX. YI FIGURE 2.2-2 W4CAL P013k DEIRSITY- HIGE TRIP SETPOXN! PART 2 (QR2 Versus Y3L)

ST. LUCIE - UNIT I 2-7 Amendment No. Z7,;7,90, 106

oh'sTO~rNM 1.5 PVAR 2061 Al

QR. + 15.85 To - 8950 1.4 1.3 1.2 1.1 . \.

1. 0

-0.6 0.5 -0.4 -0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 0.5 0.6 AXIAL SHAPE INDEX, Yj FIGURE 2.2-3 Thermal Margin/Low Pressure Trip Setpoint ST. LUCIE - UNIT I 2-8 Amendment- ;o. 7 X0

PVAR = 2061 Al QR1 + 15185 TIN - 8950 QR1 FUNCTION 1.2 1.0 0.8 QR1 0.6.

0.4 0.2 0

0 0.4 0.6 0.8 .2 A. FRACTION OF RATED THERMMAL POWER FIGURE 2.2 Thermal Margin/Low Pressure Trip Setpoint Part 2 (Fraction of RATED TItEMAL PON7ER Versus QR 1 )

ST. LUC1E - UNIT 1 2-9 Amendment No. Z7, 4 8

SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation (LCO) contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation (LCO) and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation (LCO) is not met, except as provided in the associated ACTION requirements, within I hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

'here corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the LCO. Exceptions 4 to these requirements are stated in the individual specifications.

This specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL 'lODE or other specified applicability condition shall not be made when the conditions of the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION statements. Exceptions to these requirements are stated in the individual specifications.

ST. LUCIE - UNIT 1 3/4 0-l Amendment No. #0, f, 103

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

4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

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

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

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

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

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

a. Inservice inspection of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g) (6) (i).

b. deleted ST. LUCIE - UNIT 1 3/4 0-2 Amendment No. 25, 40, 90, 95, 4og. 452, 186

APPLICABILITY SURVEIL LANCF RFQOJIRFMENTS (Contiaued) 4.0.5 (Continued)

c. deleted I

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

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

ST. LUCIE - UNIT I 3140-3 Amendment No. 99, 4-14, 153

314.1 REACTIVITY CONTROL SYSTEMS 314.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tar > 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 not within limits immediately initiate and continue boration at > 40 gpm of greater than or equal to 1720 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 2e 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 loading, 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 Keff > 10.

With Keff < 1.0.

ST. LUCIE - UNrr 1 314 1-1 Amendment No. 2E.45.6. ;3 . 45. 1 71

REACTIVITY CONTROL SYSTEMS

~SURVEILLANCE REQUIREMENTS (Continued)

e. When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by con-sideration 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.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1000 pcm at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

For Modes 3 and 4, during calculation of shutdown margin with all CEA's verified fully inserted, the single CEA with the highest reactivity worth need not be assumed to be stuck in the fully withdrawn position.

ST. LUCIE - UNIT I 3/4 1-2 Amendment No. AA. 86

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - Tam < 200 'F LIMITING CONDITION FOR OPERATION 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 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN 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 coolant 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 3/4 1-3 Amendment No. 4, 8,6 1 71

gEACTiVl y CONTROL SYSTEMS BORON DILUTIQN

[IMmNG CONDITION FOR OPFRAMON 3.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be k 3000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.

APPLICABILITY: ALL MODES.

ACTION:

With the flow rate of reactor coolant to the reactor pressure vessel < 3000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

5URVFILLANQF REOUIRFMETS 4.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be determ-ined to be 2 3000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by eiter.

a. Verifying at least one reactor coolant pump is in operation, or

b. Verifying that at least one low pressure safety injection pump is in operation and supplying 2 3000 gpm to the reactor pressure vessel.

ST. LUCIE - UNrT 1 3(41'4 Amendnrt No 1 52-

REACTIVMTY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMING CONDION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be maintained within the limits specified in the COLR. The maximum positive limit shall be: I

a. Less positive than +7 pcmr/F whenever THERMAL POWER is < 70% of RATED THERMAL POWER, and

b. Less positive than +2 pcnPF whenever THERMAL POWER is > 70%MO of RATED THERMAL POWER.

APPLICABILITY: MODES 1 and 2*#.

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

With Ke > 1.0.

See Special Test Exception 3.10.2.

ST. LUCIE - UNIT I 3t4 1-5 Amendment No. a, 69, 86 150

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

a. Prior to initial operation above 5% of RATED THERMAL POWER, after each refueling.

b. At any THERMAL POWER, within 7 EFPD after initially reaching a RATED THERMAL POWER equilibrium boron concentration.

c. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 30O ppm.

ST. LUCIE - UNIT 1 3/4 1-6 Amendment No. 2 7

I I REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be > 515 0 F when the reactor is critical.

APPLICABILITY: MODES 1 and 2#.

I ACTION:

With a Reactor Coolant System operating loop temperature (Tava) <5150 F, restore T to within its limit within 15 minutes or be in HOT STANDBY within thaV~ext 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.5 The Reactor Coolant System temperature (Tavg ) shall be determined to be > 515 0F. avg

a. Within 15 minutes prior to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System temperature (Tavg) is < 5250 F.

With Keff > 1.0

>1.0.

I ST. LUCIE - UNIT 1 314 1-7 Amendment No. 4

REACTIVITY CONTROL SYSTEMS 314.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMING CONDION 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 REQUIREMENTS 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 RCS pressure boundary does not exist, or (b) RCS pressure boundary integrity exists and no charging pumps are operable. In the latter case: 1) all charging pumps shall be disabled;

2) heatup and cooldown rates shall be limited in accordance with Figure 3.1-1b; and 3) at RCS temperatures below 151F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed:

Hich Pressure Header Auxiliary Header HCV-3616 HCV-3617 HCV-3626 HCV-3627 HCV-3636 HCV-3637 HCV-3646 HCV-3647 Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

ST. LUCIE - UNIT 1 3/4 1-8 Amendment No. 60, 84, 90,

94. 443. 404.444. 42. 179

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , when0 the Reactor Auxiliary Building air temperature is less than 55 F, by verifying that the Boric Acid MakeupTank solution temperature is greater than 55OF, when the flowpath from the Boric Acid Makeup Tank is required to be OPERABLE.

ST. LUCIE - UNIT 1 3/4 1-9 Amendment No. 94

a -- I KM I I

I

. ..:.1 II I

..... I..;.1 I .

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-4

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1:

. ..... -4 180 200 220 Xc - IHDICATED REACTOR COOLANT TEMPERATURE, eF FIGURE 3.1 -lb MAXIMUM ALLOWABLE HEATUP AND COOL[OWN RATES, SINGLE HPSI PUMP INOPERATnON (23.6 EFPY)

ST. LUCIE - UNIT I 314 1-9a Amendment No. 0s. 4.e*, 141

THIS PAGE INTENTIONALLY LEFT BLANK 3S.

ST. LUCIE - UNIT 1 3/4 1-9b Amendment No. 81

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMING 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 pait 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:

a. 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.

b. 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.

c. 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 3.1.1.2 at 2001F; 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 3/41t-10 Amendment No. 48,86, 90, 4, 1 71

,REACTIVITY

._ CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS I

4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

I

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

b.. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation Signal.

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 below 55'F by verifying that the solution temperature of the Boric Acid Makeup Tank(s) is above 550F.

ST. LUCIE - uNrT 1 3/4 1-1 1 Amendment No. t0, 94

REACT1VITY CONTROL SYSTEMS CHARGING PUMPS -SHUTDOWN LlMING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or high pressure safety injection pump* in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

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

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

when tested pursuant to the Inservice Testing Program.

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

(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: 1) all charging pumps shall be disabled;

2) heatup and cooldown rates shall be limited in accordance with Figure 3.1-lb; and 3) at RCS temperatures below 1150F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed:

High Pressure Header Auxiliarm Header HCV-3616 HCV-3617 HCV-3626 HCV-3627 HCV-3636 HCV-3637 HCV-3646 HCV-3647

- Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN. I ST. LUCIE - UNIT 1 3/4 1-12 Amendment No. 60.84. 89 404,44G. 444.,462. 46a, 179

REACTIVTY CONTROL SYSTEMS CHARGING PUMPS - OPERATING IMMING MsNDMO FOR OPFRATION -- - - -___

3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hourv; restore at least two charging pumps to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

S!JRVEIL I ANQF RFOUtIREMENTS 4.12.4 At least two charging pumps shall be demonstrated OPERABLE by verifying that each pump develops a flow rate of greater than or equal to 40 gpm when tested pursuant to the Inservice Testing Program.

I ST. LUCIE - UNIT 1 3/4 1-13 ~~~~~Amnendment ST.

LUCIE

-UNIT 1 3'41-13 No.GG 153

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

APPLICABILITY: MODES 5 and 6.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required boric acid pump shall be demonstrated OPERABLE by verifying that on recirculation flow, the pump develops a discharge pressure of > 75 psig when tested pursuant to the Inservice Testing Program.

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

ST. LUCIE - UNIT I 314 1-14 Amendment No. XO, 453.179

REACilVIlY CONTROL SYSTEMS BORIC ACID PUMPS - OPERATING LIMMNG CONbW1MON EOR OPERA11O)N 3.1.2.6 At least the boric acid pump(s) in the 'oron injection flow path(s) required OPERABLE pursuant to Specification 3.1 2.2a shall be OPERABLE if the flow path through the boric acid pump in Specification 3.1.22a is OPERABLE.

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

With one boric acid pump required for the boron injection flow path(s) pursuant to Specification 3.1 .2.2a inoperable, restore the boric acid pump to OPERABLE status within 72 hours8.333333e-4 days 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVFILLIANOF RFOUIRFMFNTS 4.1.26 The above required boric acid pump(s) shall be demonstrated OPERABLE by verifying that on recirculation flow, the pump develops a discharge pressure of > 75 psig when tested pursuant to the Insenrice Testing Program.

I ST. LUCIE - UNIT 1 3/4 1-15 Amendment No. 9, 153

REACTMTY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDMON 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 2.5 to 3.5 weight percent boric acid (4371 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 1720 ppm, and

3. A minimum solution temperature of 40 0F.

APPLICABILITY: MODES 5 and 6.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.1.2.7 The above required berated 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 leve; 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 site ambient air temperature is < 400F.

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 551F by verifying that the Boric Acid Makeup Tank solution temperature is greater than 550 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. I ST- LUCIE - UNIT I 3/4 1-16 Amendment No. 2, 94, 179

FIGURE 3.1-1 ST. LUCIE 1 MIN BAMT VOLUME VS STORED BAMT CONCENTRATION minq 9,00 u F - I -. - rn-I -

I.

I y

//

X7 (8700)

/ ACCEPTABLE OPERATION 8,00 0-1 0

wm W 7,00 l - --- ---- I za 6,00 UNACCEPTABLE I /

OPERATION 1 (5950_

r, n In b l ] i l }_ A5400) 2.4 2.6 2.8 3 3.2 3.4 3.6 (4546 PPM) (5245 PPM) (5944 PPM)

(4196 PPM) (4896 PPM) (5595 PPM (6294 PPM)

STORED BAMT CONC (wt % boric acid)

ST. LUCIE - UNIT 1 314 1-X7 Amendment No. V, 0# s 129

REACTMOTY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDMON FOR OPERATION 3.1.2.8 At least two of the following 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 IB in accordance with Figure 3.1-1.

c. Boric Acid Makeup Tanks 1A and lB 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 401,800 gallons of water,

2. A minimum boron concentration of 1720 ppm,

3. A maximum solution temperature of 100°F,

4. A minimum solution temperature of 550F when in MODES 1 and 2, and

5. A minimum solution temperature of 400F 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 0F; restore at least two borated I 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 of the water source, ST. LUCIE- -UNIT 1 314 1-18 Amendment No. 28, 48, 86,94, 4.2. 171

REACTIVITY CONTROL SYSTEMS I

SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying the water level in each water source.

I

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 the Boric Acid Makeup Tank solution temperature is greater than 55@F when the Reactor Auxiliary Building air temperature is below 550F.

. LUCIE - UNIT 1 3/4 1-19 Amendment No.94

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES FULL LENGTH CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 The CEA Block Circuit and all full length (shutdown and regulating)

CEAs shall be OPERABLE with each CEA of a given group positioned within 7.5 inches (indicated position) of all other CEAs in its group.

APPLICABILITY: MODES 1* and 2*.

ACTION:

a. With one or more full length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the CEA Block Circuit inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

1. With one CEA position indicator per group inoperable, take action per Specification 3.1.3.3, or I

2. With the group overlap and/or sequencing interlocks inoper-able, maintain CEAs in groups 3, 4, 5 and 6 fully withdrawn and withdraw the CEAs in group 7 to less than 5X insertion and place and maintain the CEA drive system mode switch in either the "Manual" or "Off" position, or

3. Be in at least HOT STANDBY.

c. With one full length CEA inoperable due to causes other than addressed by Action a above, but within its above specified align-ment requirements and either fully withdrawn or within the long term steady state insertion limits if in CEA group 7, operation in MODES 1 and 2 may continue.

d. With one or more full length CEAs misaligned from any other CEAs in its group by more than 7.5 inches but less than 15 inches, operation in MODES 1 and 2 may continue, provided,that within one hour the misaligned CEA(s) is either:

1. Restored to OPERABLE status within its above specified alignment requirements, or

See Special Test Exceptions 3.10.2 and 3.10.5.

ST. LUCIE - UNIT 1 3/4 1-20 Amendment No. Z7, 71

REACTIVIfY CONTROL SYSTEMS FULL LENGTH CEA POSmON (continued)

I IMmNc& CONDITION FOR OPERATION (Mntinuej)

2. Declared inoperable and satisfy SHUTDOWN MARGIN requirements of Specification 3.1.1.1. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 for up to 7 days per occurrence with a total accumulated time of s 14 days per calendar year provided all of the following conditions are met a) Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the CEAs in the group with he inoperable CEA shall be aligned to within 7.5 inches of the inoperable CEA while maintaining te allowable CEA sequence and insertion limits shown on COIR Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 .

e. With one funleng CEA misaligned from any other CEA in its group by 15 or more inches, operation in MODES 1 and 2 may continue provided that the misaligned CEA is positioned within 7.5 inches of other CEAs in its group in accordance with the time constraints shown in COLR Figure 3.1-la.

f. With one full-ength CEA misaligned from any other CEA in its group by 15 or more inches beyond the time constraints shown in COLR Figure 3.1-la, reduce power to s 70%h of RATED THERMAL POWER prior to completing ACTION f.1 or f.2.

1. Restore the CEA to OPERABLE status within its specified alignment requirements, or

2. Declare the CEA inoperable and satsy the SHUTDOWN MARGIN requirements of Specification 3.1.1.1. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:

a) Withn hour, the remainderof the CEAs in the group wit the inoperable CEA shall be aligned to within 7.6 inches of te inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on COLR Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1..6 during subsequent operation.

ST. LUCIE - UNIT 1 3J4 1-2 AMrnert No. 4.469,1 52

I REACTIVITY CONTROL SYSTEMS FULL LENGTH CEA PosmoN (gntinpgm

, MIT1IN CONDTON FOR PERAllON (conlinued b) The SHUTDOWN MARGIN requiremrent of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 .

g. With more than one full length CEA inoperable or misaligned from any other CEA in its group by 15 inches (indicated position) or more, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

h. With one full-length CEA inoperable due to causes other than addressed by ACTION a above, and inserted beyond the long term steady state insertion limits but within its above specied alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6.

URVFILLANnF REQUIREMPNTS 4.1.3.1.1 The position of each fulgength CEA shall be determined to be within 7.5 inches (indicated position) of all other CEAs in its group at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Deviation Circuit and/or CEA Block Circuit are inoperable, then verify the indrvidual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full length CEA not fully inserted sal be deternined to be OPERABLE by inserting it at least 7.5 inches at least once per 92 days.

4.1.3.1.3 The CEA Block Circuit shall be demonstrated OPERABLE at least once per 92 days by a functional test which verifies that the circuit prevents any CEA from being misaligned from all other CEAs in its group by more than 7.5 inches (indicated position).

4.1.3.1.4 The CEA Block Circuit shall be demonstrated OPERABLE by a functional test which verifies that the circuit maintains the CEA group overlap and sequencing requirements of Specification 3.1.3.6 and that the circuit prevents the regulating CEAs from being inserted beyond the Power Dependent Insertion Limit of COLR Figure 3.12:

a. Prior to each entry into MODE 2 from MODE 3, except that such verification need not be performed more often than once per 92 days, and

b. At least once per 6 months.

The licensee shall be excepted from compliance during the startup test program for an entry into MODE 2 from MODE 3 made in association with a measurement of power defect ST. UJCE- UNIT1 (41-22 Ablnrterd No. 44, A. ;.

44,469 ,152

DELETED ST. LUCIE - UNIT 1 3/4 1-23 Amendment No. ; 150

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS LIMITING CONDITION FOR OPERATION 3.1.3.3 All shutdown and regulating CEA reed switch position indicator channels and CEA pulse counting position indicator channels shall be OPERABLE and capable of determining the absolute CEA positions within

+ 2.25 inches.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. Deleted.

b. With a maximum of one reed switch position indicator channel per group or one (except as permitted by ACTION item d. below) pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel partially inserted, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

1. Restore the inoperable position indicator channel to OPERABLE status, or

2. Be in HOT STANDBY, or

3. Reduce THERMAL POWER to < 70% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. Operation at or below this reduced THERMAL POWER level may continue provided that within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a) The CEA group(s) with the inoperable position indi-cator is fully withdrawn while maintaining the withdrawal sequence required by Specification 3.1.3.6 and when this CEA group reaches its fully withdrawn position, the "Full Out" limit of the CEA with the inoperable position indicator is actuated and verifies this CEA to be fully withdrawn. Subsequent to fully withdrawing this CEA group(s), the THERMAL POWER level may be returned to a level consistent with all other applicable specifications; or ST. LUCIE - UNIT I 3/4 1-24 Amendment No. 27

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS (Continued)

LIMITING CONDITION FOR OPERATION b) The CEA group(s) with the inoperable position indi-cator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawal sequence and THERMAL POWER level required by Speci-fication 3.1.3.6 and when this CEA group reaches its fully inserted position, the "Full In" limit of the CEA with the inoperable position indicator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6.

c. With a maximum of one reed switch position indicator channel per group or one pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or (N. fully withdrawn position, operation may continue provided:

1. The position of this CEA is verified immediately and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its "Full In" or "Full Out" limit (as applicable),

2. The fully inserted CEA group(s) containing the inoperable position indicator channel is subsequently maintained fully inserted, and

3. Subsequent operation is within the limits of Specifica-tion 3.1.3.6.

d. With one or more pulse counting position indicator channels inoperable, operation in MODES 1 and 2 may continue for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided all of the reed switch position indicator channels are OPERABLE.

SURVEILLANCE REQUIREMENTS 4.1.3.3 Each position indicator channel shall be determined to be OPERABLE by verifying the pulse counting position indicator channels and the reed switch position indicator channels agree within 4.5 inches at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Deviation circuit is inoperable, then compare the pulse counting position indicator and reed switch position indicator channels at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ST. LUCIE - UNIT 1 3/4 1-25

REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be < 3.1 seconds from when electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a. Tavg > 515'F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 3.

ACTION:

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

l ST. LUCIE - UNIT 1 3/4 1-26 Amendment No. 32

REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LMTNCODTO FOOPRTO LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to at least 129.0 inches.

APPLICABILITY: MODES 1 and 2*#.

ACTION:

With a maximum of one shutdown CEA withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, to less than 129.0 inches, within one hour either:

a. Withdraw the CEA to at least 129.0 inches, or

b. Declare the CEA inoperable'and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to- be withdrawn to at least 129.0 inches:

a. Within 15 minutes prior to-withdrawal of any CEAs in regulating groups during an approach to reactor criticality, and

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

See Special Test Exception 3.10.2.

With Keff ' 1.0.

ST. LUCIE - UNIT 1 314 1-27 Amendment ND. 20

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION UMITS LIMMNG CONDMON FOR OPERATION 3.1.3.6 The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits specified in the COLR (regulating CEAs are considered to be fully withdrawn when withdrawn to at least 129.0 inches) with CEA insertion between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Umits restricted to:

a. 5 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval,

b. s 5 Effective Full Power Days per 30 Effective Full Power Day interval, and

c. s 14 Effective Full Power Days per calendar year.

APPLICABILITY: MODES V and 2#.

ACTION:

a. With the regulating CEA groups inserted beyond the Power Dependent Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either

1. Restore the regulating CEA groups to within the limits, or

2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position and insertion limits specified in the COLR.

b. With the regulating CEA groups inserted between the Long Term Steady State Insertion Umits and the Power Dependent Insertion Umits for intervals

> 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, except during operation pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, operation may proceed provided either.

1. The Short Term Steady State Insertion Limits are not exceeded, or

2. Any subsequent increase in THERMAL POWER is restricted to : 5% of RATED THERMAL POWER per hour.

See Special Test Exceptions 3.10.2 and 3.10.5.

Withl Kenl > 1*0-ST. LUCIE - UNIT I 3/4 1-28 Amendment No. 2q 150

REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Continued)

LIMITING CONDITION FOR OPERATION

c. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Power Dependent Inser-tion Limits for intervals > 5 EFPD per 30 EFPD interval or

> 14 EFPD Der calendar year, except during operations pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, either:

1. Restore the regulating groups to within the tong Term Steady State Insertion Limits within two hours, or

2. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURV'EILLAN'CE REPUIREIE?1ITS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Power Dependent Insertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once-per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The accumulated times during which the regu-lating CEA groups are inserted between the Long Term Steady State Insertion Limits and the Power Dependent Insertion Limits shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..

ST. LUCIE - UNIT 1 3/4 1-29

DELETED ST. LUCIE- UNIT 1 3(4 1-30 Amendment No. 2, 48,1 50

314.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDMION FOR OPERATION 32.1 The linear heat rate shall not exceed the limits specified in the COLR.

APPUCABILITY: MODE 1.

ACTION:

With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of the power dependent control limits of COLR Figure 3.2-2, within 15 minutes initiate corrective action to reduce the linear heat rate to l within the limits and either.

a. Restore the linear heat rate to within its limits within one hour, or

b. Be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 42.1.1 The provisions of Specification 4.0A are not applicable.

4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring system.

42.1.3 Excore Detector Monitoring System - The excore detector monitoring system may be used for monitoring the linear heat rate by: 1

a. Verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Umit of Specification 3.1.3.6.

b. Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on COLR Figure 3.2-2. 1 ST. LUCIE - UNIT I 3J4 2-1 Amendment No. Sa, 150

POWER DISTRIBUTION UMITS SURVEILLANCE REQUIREMENTS (continued)

c. Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of COLR Figure 32-2, where 100 percent of maximum allowable power represents the maximum THERMAL POWER allowed by the following I

expression:

M xN where:

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

2. N is the maximum allowable fraction of RATED THERMAL POWER as determined by the FTr curve of COLR Figure 3.2-3.

42.1.4 Incore Detector Monitoring System" - The incore detector monitoring system may be used for monitoring the linear heat rate by verifying that the incore detector Local Power Density alarrns:

a. Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operation in MODE 1.

b. Have their alarm setpoint adjusted to less than or equal to the limits shown on COLR Figure 3.2-1. I

If the incore system becomes inoperable, reduce power to M x N within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and monitor linear heat rate in accordance with Specification 42.1.3.

ST. LUCIE - UNIT 1 3/4 2-2 Amendment No. 4, 2-, 8a, 6a. 66, 9 40Q, 434. 486, 150

Pages 3/4 2-4 (Amendment 106), 314 2-5 (Amendment 63), and 3/4 2-6 through 3/4 2-8 (Amendment 109) have been deleted from the Technical Specifications.

The next page is 3/4 2-9.

ST. WCIE - UNIT 1 314 2-3 Amendment No. 2, 8a, 48, X, u, 4, 1 50

POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR - Fr I IMIING MONDMON FOR OPFRATION 32.3 The calculated value of FT shall be within the limits specified inthe COLR.

APPLUABILAlY: MODE 1*.

ACTION:

With FT not within limits, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either.

a. Be in at least HOT STANDBY, or

b. Reduce THERMAL POWER to bring the combination of THERMAL POWER and FT to within the limits of COLR Figure 32-3 and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Units of Specification 3.1.3.6. The THERMAL POWER limit determined from COLR Figure 32-3 shall then be used to establish a revised upper THERMAL POWER level limit on COLR Figure 32-4 (truncate Figure 32-4 at the allowable fraction of RATED THERMAL POWER determined by COLR Figure 32-3) and subsequent operation shall be maintained within the reduced acceptable operation region of COLR Figure 32-4.

SLIRVFWItANCE REOUIFEMErNTS 42.3.1 The provisions of Specification 4.0.4 are not applicable.

42.32 FT shall be calculated by the expression FT =Fr(1+Tq) when Fr is calculated with a non-ull core power distribution analysis code and shal be calculated as FT = F when calculations are performed with a full core power distrbution analysis code. FT shall be determined to be within its limit at the following intervals.

a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading.

b. At least once per 31 days of accumulated operation in MODE 1, and C. Within four hours ifthe AZIMUTHAL POWER TILT (Tq) is > 0.03.

See Special Test Exception 3.102.

ST. LUCIE - UNIT 1 3142.9 Arnd tsNo. 2,ea,48.66S, 46,152

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.3 F shall be determined each time a calculation of FT is red by using tge incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination.

4.2.3.4 Tq shall be determined each time a calculation of FT is made using a non-full core power distribution analysis code. The value of Tq used to determine FT in this case shall be the measured value of Tq. .1 l

ST. LUCIE - UNIT 1 3/4 2-10 Amendment No. Z7, 65

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - Tq LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (Tq) shall not exceed 0.03.

APPLICABILITY: MODE 1*

ACTION:

a. With the indicated AZIMUTHAL POWER TILT determined to be

> .030 but 1 0.10, either correct the power tilt within two hours-or determine within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once b.

per subsequent 8 tours, that the TOTAL INTEGRATED RADIAL PEAKING FACTOR (F,) is within the limits of Specification 3.2.3.

With the indicated AZIMUTHAL POWER TILT determined to be > 0.10, I

operation may proceed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided that the TOTAL INTEGRATED RADIAL PEAKING FACTOR (FT) is within the limits of Specification 3.2.3. Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to < 20X of the maximum allowable THERMAL POWER level for -the existing Reactor Coolant Pump combination.

SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

4.2.4.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by:

a. Calculating the tilt at least once per 7 days when the Subchannel Deviation Alarm is OPERABLE, See Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 2-11 Amendment No. A, g7,A, 109,

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

b. Calculating the tilt at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the Subchannel Deviation Alarm is inoperable, and

c. Using the incore detectors to determine the AZIMUTHAL POWER TILT at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when one excore channel is inoperable and THERMAL POWER is > 75% of RATED THERMAL POWER.

27 ST. LUCIE - UNIT 1 314 2-12 Amendment No. I

POWER DISTRIBUTION LIMITS PNB PARAMETERS LIMITING CONDMON FOR OPERATION 32.5 The following DNB related parameters shall be maintained within the limits shown on Table 32-1:

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 parameters of Table 3.2-1 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.

I

Not required to be performed until THERMAL POWER is > 90% of RATED THERMAL POWER.

ST. W01E - Mr 1 3/4 2-13 Amendment No. a, 1 50

TABLE 3.2-1 DNB MARGIN LIMITS Parameter Four Reactor Coolant Pumps Operating Cold Leg Temperature < 5490 F Pressurizer Pressure > 2225 psia

Reactor Coolant Flow Rate > 365,000 gpm I

AXIAL SHAPE INDEX COLR Figure 3.2-4

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

ST. LUCIE - UNIT 1 314 2-14 Amendment No. 0, 46, 13,

.46,

+Se, 163

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE. j APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing.

Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the Total No. of Channels column of Table 3.3-1.

ST. LUCIE - UNIT 1 3/4 3-1 Amenedment No . 128

IC ( I r

-_4 TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION I.

I-I Z MINIMUM

'-f TOTAL NO. CHANNELS CHANNELS APPL ICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I

-a

1. Manual Reactor Trip 2 21() 2 1, 2Wand

1

2. Power Level - High 4 2(a) 3(f) 1, 2 2#

3. Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG 1, 2 (e) 2#' .

4. Pressurizer Pressure - High 4 2 3 1, 2 2#

S. Containment Pressure.- High 4 2 3 1, 2 2# .

2

6. Steam Generator Pressure - Low 4/SG .2(b)/SG 3/SG 1, 2 I,

7. Steam Generator Water Level - Low 4/SG 9/SG 3/SG 1, 2 2#

I r11

8. Local Power Density - High 4 Z(c) 3 1 2#

0b

9. Thermal Margin/Low Pressure 4 2(a) 3 1t 2 (e) 2#

W 9a, Steam Generator Pressure

'SI Difference - High 4 2(a). 3 1, 2 (e)

10. Loss of Turbine--Hydraulic Fluid Pressure - Low 4 2(c) 3 1

( " r I

(I iI II LO

-4 TABLE 3.3-1 (Continued) I REACTOR PROTECTIVE INSTRUMENTATION I MINIMUM b-.

..TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT )F CHANNELS TO TRIP OPERABLE MODES ACTION

11. Wide Range Logarithmic Neutron Flux Monitor

a. Startup and Operating--

Rate of Change of Power -

High I 2(d) 3 1, 2 and

2#

b. Shutdown 0 2 3, 4, 5 3

12. Reactor Protection System 4 2 4 1, 2* 4 Logic I-(a) 13. Reactor Trip Breakers 4 2 4 . 1, 2* 4 w

(a a.

0

TABLE 3.3-1 (Continued)

TABLE NOTATION With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

The provisions of Specification 3.0.4 are not applicable.

(a) Trip may be bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is > 1% of RATED THERMAL POWER. I (b) Trip may be manually bypassed below 685 psig; bypass shall be automatically removed at or above 685 psig.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when Power Range Neutron Flux power is > 15% of RATED THERMAL POWER. I (d) Trip may be bypassed below 10 % and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when Wide Range Logarithmic Neutron Flux power is > 10 %

and Power Range Neutron Flux power S 15% of RATED THERMAL POWER.

(e) Deleted.

(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

ACTION STATEMENTS ACION I - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

ST. LUCIE - UNIT 1 3/4 3-4 Amendment No. 45, 4, 45p 15 9 40e

TABLE 3.3-1 (Continued)

ACTION STATEMENTS

b. Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel.

c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while performing tests and maintenance on than channel provided the other inoperable channel is placed in the tripped condition.

ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing per Specification 4.3.1.1.1.

ST. LUCIE - UNIT I 314 3-5 Amendment'No. 19, 2?

DELETED ST. LUCIE - UNIT 1 3/4 3-6 Amendment No. 27, X1, 128

t.. :{..' .8:,.1, j. . ;.',. iS'~~~~ve~.

I. TABLE 4.:S-ti

(

-I REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS I .

n I CHANNEL MODES INWHICH

. I CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRUD z

0-4

-i 1. Manual Reactor Trip N.A. tI N.A. S/U(1) N.A.

2. Power Level - High

a. Nuclear Power S M(3),q(5) M 1, 2

b. AT Power S D4),: Q M 1; I

3. Reactor Coolant Flow - Low S R M 1,*'2

4. Pressurizer Pressure - High S R M 1, 2

5. Containment Pressure - High S R M 1, 2

6. Steam.Generator Pressure - Low S R M 1, 2 f

7. Steam Generator Water Level - Low S R M4 1,2

8. Local Power Density - High S R M 1,2

.31

9. Thermal Margin/Low Pressure SI R M 2 I;

9a. Steam Generator Pressure Difference - High S R M 1, 2

10. Loss of Turbine-.Hydraulic Fluid Pressure - Low N.A. N.A. S/U(1) N.A.

11. Wide Range Logarithmic Neutron *S N.A. S/U(l ) . 1, 2, 3, 4, 2 Flux Monitor 5 and *

12. Reactor Protection System Logic N.A. N.A. M and S/U(l) 1, 2 and

0

13. Reactor Trip Breakers N.A. N.A. M 1, 2 and

r(

4,

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 AT power may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next mdjorotest power plateau.

(3) - Above 15% of RATER THERMAL POWER, recalibrate the excore detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER 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 AT power signals agree with calorimetric calculation.

(5) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

ST. LUCIE - UNIT I 3/4 3-8

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel Inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the "Total No. of Channels" Column of Table 3.3-3.

ST. LUCIE - UNIT 1 Amendment No. 128 3/4 3-9

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TOTAL NO. MINIMUM OF CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT CHANNELS TO TRIP OPERABLE MODES ACTION

1. SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) 2 1 2 1.2.3,4 8

b. Containment Pressure -

High 4 2 3 1.2.3 9#

c. Pressurizer Pressure -

Low 4 2 3 1. 2,3(a) 9#

2. CONTAiNMENT SPRAY (CSAS)

a. Manual (Trip Buttons) 2 1 2 1.2,3,4 8

b. Containment Pressure -

High-High 4 2(b) 3 1,2,3 10a#, 10b#, 10c I

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) 2 1 2 1.2,3.4 8

b. Containment Pressure -

High 4 2 3 1,2,3 9#

c. Containment Radiation -

High 4 2 3 1.2,3.4 9#

d. SIAS (See Functional Unit 1 above) - -

4. MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) 2/steam 1/steam 2/operating 1,2,3,4 8 generator generator steam generator

b. Steam Generator 4/steam 2/steam 3/steam 1, 2. 3(c)

Pressure - Low generator generator generator ST. LUCIE -UNIT I 314 3-10 Amndment No.46.,P 188

TABLE 3.3-3 (Coritinued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS (Trip Buttons). 2 2 2 1, 2, 3,4 a

b. Refueling Water Tank - Low 4 2 3 1,2,3 13 I

6. LOSS OF POWER

a. 4.16 kv Emergency Bus Under-voltage (Loss of Voltage) 2/Bus 2/Bus 1/Bus 1,2,3 12

b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) 2/Bus 2/Bus 1/Bus 1,2,3 12

c. 480 V Emergency Bus Under-voltage (Degraded Voltage) 2/Bus 2/Bus 1/Bus 1.2,3 12

7. AUXILIARY FEEDWATER (AFAS)

a. Manual (Trip Buttons) 4/SG 2ISG 4ISG 1,2,3 11

b. Automatic Actuation Logic 4ISG 2/SG 3/SG 1.2,3 11 I

c. SG Level (IA1B) - Low 4/SG 2/SG 3ISG 1,2,3 14a#, 14b#, 14c I

8. AUXILIARY FEEDWATER ISOLATION

a. SG lA-SGQ B Differential Pressure 4ISG 2/SG 3/SG 1,2, 3 14a#, 14b#, 14c I

b. Feedwater Header IA- 1B Differential Pressure 4/SG 2/SG 3/SG 1,2,3 14a#, 14c I ST. LUCIE -UNIT I 3/4 3-11 Amendmer No. 45. 3. U. 3. 4U. 42 188

TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is < 1725 psia; bypass shall be automatically removed when pressurizer pressure is > 1725 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed in this MODE below 685 psig; bypass shall be automatically removed at or above 685 psig.

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 8 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the

-' bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purposes of testing and maintenance, the inoperable chanmel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

b. Within one hour. all functional units receiving an

- _ input from the inoperable channel- are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel.

c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

ST. LUCIE - UNIT I 3/4 3-12 Amendment No. A& 45

TABLE 3.3-3 (continued)

TABLE NOTATION ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel Is placed in the bypassed or tripped condition and the Minimum Channels OPERABLE requirement is demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Ifthe inoperable channel can not be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , then place the Inoperable channel In the tripped condition.

b. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , all functional units receiving an Input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as required by a. above for the inoperable channel.

c. With the number of channels OPERABLE one less than the Minimum Channels OPERABLE, operation may proceed provided one of the inoperable channels has been bypassed and the other inoperable channel has been placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Restore one of the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 11 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the Inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed Inthe tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ST. LUCIE - UNfl r 1 314 3-13 Amendment No. 44, 3;, fi. 72, 188

TABLE 3.3-3 (continued)

TABLE NOTATION ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or Upped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If OPERABILITY can not be restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

ACTION 14 - With the number of channels OPERABLE one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If the inoperable channel can not be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , then both AFAS-1 and AFAS-2 in the inoperable channel shall be placed in the bypassed condition. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Specification 6.5.1.6m. The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN.

b. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the Inoperable channel.

c. With the number of channels OPERABLE one less than the Minimum Channels OPERABLE, operation may proceed provided one of the Inoperable channels has been bypassed and the other inoperable channel has been placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Restore one of the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ST. LUCIE - UNIT 1 314 3-13a Amendment No. 188

I i

TABLE 3.3-4 a.

CA ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES C)# ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES 0-4

1. SAFETY INJECTION (SIAS)

-4

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure - High < 5 psig < 5 psig a

c. Pressurizer Pressure - Low > '1000 psia > 1600 psia

2. CONTAINMENT SPRAY (CSAS)

-A

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure - High-High 'i 10 psig . 10 psig

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Containment Pressure - High < 5 psig < 5 psig a

c. Containment Radiation - High < 10 R/hr < 10 R/hr

-Ph

d. SIAS ..--- (See FUNCTIONAL UNIT 1 above) ----

4. MAIN STEAM LINE ISOLATION (MSIS)

-1.1 a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Steam Generator Pressure - Low > 585 psig > 585 psig

5. CONTAINMENT SUMP RECIRCULATION (RAS) *I

a. Manual RAS (Trip Buttons) Not Applicable Not Applicable

b. Refueling Water Tank - Low 48 inches above I 48 inches above tank bottom tank bottom

(

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ALLOWABLE FUNCTIONAL UNIT TRIP VALUE VALUES

6. LOSS OF POWER

a. 4.16 kv Emergency Bus Undervoltage > 2900 volts with a > 2900 volts with a (Loss of Voltage) 1 + .5 second time delay 1 + .5 second time delay

b. 4.16 kv Emergency Bus Undervoltage > 3831 volts with a > 3831 volts with a (Degraded Voltage) 18 + 2 second time delay 18 + 2 second time delay

c. 480 volts Emergency Bus Undervoltage > 415 volts with a > 415 volts with a (Degraded Voltage) < 9 second time delay 5 9 second time delay

7. AUXILIARY FEEDWATER (AFAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

c. SG IA & IB Level Low

19.0% > 18.0%

8. AUXILIARY FEEDWATER ISOLATION

a. Steam Generator AP - High < 275 psid 89.2 to 281 psid

b. Feedwater Header High AP c 150.0 psid 56.0 to 157.5 psid ST. LUCIE - UNIT I 3/4 3-15 Amendment No. 3X, 8, 2, 42. 405,424. 177 I

DELETED ST. LUCIE. - UNIT I 3/4 3-16 Amendment No. 77. 477 jM, 71, 128

DELETED

- ST. LUCIE - UNIT 1 3/4 3-17 Amendment No. 77, Z7, 9J, 7ZM 7F, SA*.

128

C (

TABLE 4.3-2 V)

C, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS n

C.,

I-CHANNEL MODES INWHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST- REQUIRED

-q .

.-I

1. SAFETY INJECTIONI(SIAS)

a. Manual (Trip Buttons) N.A. ' N.A. R N.A. I

b. Containment Pressure - High S R M 1, 2, 3

c. Pressurizer Pressure - Low S R M 1,-2, 3

d. Automatic Actuation Logic N.A. N.A. MMl 1, 2, 3

2. CONTAINMENT SPRAY (CSAS)

a. Manual (Trip Buttons) N.A. NA. R N.A.

b. Containment Pressure --

W High - High .S R Mi 1, 2, 3 c, Automatic Actuation Logic N.A. N.A. 1, 2, 3 W-

3. CONTAINMENT ISOLATION (CIS)

a. Manual (Trip Buttons) N.A. N.A. R N.A.

b. Containment Pressure - High S I*R M 1, 2, 3

c. Containnient Radiation - High S R M 1, 2, 3, 4

d. Automatic Actuation Logic N.A. N.A. M(l) 1, 2, 3

e. SIAS Nf.A. N.A. n N.A.

I

4. ..MAIN STEAM LINE ISOLATION (MSIS)

a. Manual (Trip Buttons) N.A. N.A. R N.A.

b. Steam Generator Pressure - Low S R M 1, 2, 3 r C. Automatic Actuation Logic N.A. N.A. 14(1) 1, 2,.3 F 5. CONTAINMENT SUMP RECIRCULATION (RAS)

-.4 a. Manual RAS (Trip Buttons). N.A. N.A. R N.A.

a b. Refueling Water Storage w

Tank - Low S R M 1, 2, 3 C. Automatic Actuation Logic N.A. N.A. M(l) 1, 2, 3

(. ( Cn

--I (A,

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-0 Ca MODES IN WHICH 0-4 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK - ALUMRAJIUN .L51.... KLUVUI1LO 6-a

6. LOSS OF POWER

a. 4.16 kv Emergency Bus Under-voltage (Loss of Voltage) S R M 1, 2, 3

b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) S R M 1, 2, 3

c. 480 V Emergency Bus Under- S R M 1, 2, 3 I

voltage (Degraded Voltage)

7. AUXILIARY FEEDWATER (AFAS)

-ab

a. Manual (Trip Buttons) N.A. N.A. R 1, 2, 3

b. SG Level (A/B) - Low S R M I, 2, 3

c. Automatic Actuation Logic N.A. N.A. M 1, 2, 3

8. AUXILIARY FEEDWATER ISOLATION

a. SG Level (A/B) - Low and SG Differential Pressure (BtoA/AtoB) - High N.A. R M 1, 2, 3

b. SG Level (A/B) - Low and I- Feedwater Header Differential Pressure (BtoA/AtoB) - High N.A. R M 1, 2, 3

TABLE 4.3-2 (Continued)

TABLE NOTATION (1) The logic circuits shall be tested manually at least once per 31 days.

ST. LUCIE - UNIT 1 3/4 3-20

INSTRUMENTATION 314.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3 3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

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

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3.

ST. LUCIE - UNIT 1 314 3-21 Amendment No. 40. 177

( ( (

TAfLE -3.3-

-4 RADIATION MONITORING INSTUUMENTA11pN I- MINIMUM CHANNELS APPLICABLE ALARM MEASUREMENT IE QPERALtE MODES SETPOINT 88GE RA

-4 1. AREA MONITORS I-$

a. Fuel Storage Pool Area 1

g 15 mR/hr IO 10' mR/hr 13

b. Containment (CIS) 3 6 s 90 mR/hr I - IO,mR/hr 16

c. Containment Area - Hi 1 1,2, 3 & 4 S 10 R/hr I - 101 R/hr 15 Range

2. PROCESS MONITORS

a. Containment MIA I. Gaseous Activity (A)

RCS Leakage Detection I 1, 2, 3 & 4 Not Applicable 10 - 106 cpm 14 ii. Particulate Activity RCS Leakage Detection 1 1, 2, 3 & 4 Not Applicable 10 - 106 cpm 14

b. Fuel Storage Pool Area Ventilatlon System I. Gaseous Activity 1 ** 10" - 105 pCi/cc 12 ii. Particulate Activity 1 ** I - 106 cpm 12 WCA)

With fuel In the storage pool or building.

0 **With Irradiated fuel In the storage pool or whenever there Is fuel movement within the pool or crane operation with loads over the storage pool. pool

- - The Alarm Setpoints are determined and set inaccordance with requirements of the Offslte Dose Calculation Manual.

( (

ABLE 3.3-6 (Continued)

-I (A RADIATION MONITORING INSTRUMENTATION I-a MINIMUM CHANNELS APPLICABLE ALARM MEASUREMENT INSTRUMENT OPERABLE MODES- SETPO-NT -M NGE

~0- 2. PROCESS MONITORS (Continued)

c. Noble Gas Effluent Monitors

i. Radwaste Building Exhaust System (Plant Vent Exhaust Monitor) I 1, 2, 3 & 4 lo" - 10 pCi/cc 15 ii. Steam Generator W

Blowdown Treatment Facility Building - lo-2 pCi/cc Wk (A, Exhaust System I 1, 2, 3 & 4 15 IA iii. Steam Safety Valve Discharge 1/Header 1, 2, 3 & 4 l1o - lo" pCi/cc 15 iv. ECCS Exhaust I/Train 1, 2, 3 & 4 10-' - 0l pCi/cc 15 i.

- - The Alarm Setpoints are determined and set in accordance with the requirements of the Offsite Dose Calculation Manual. I CD 0

-d

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 12 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12. I ACTION 13 - With the number of channels OPERABLE less than required by-the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 14 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. X ACTION 15 - With the number of channels OPERABLE less than required by the Minimum, Channels OPERABLE requirement, either restore the inoperable Channel(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

1) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and

2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 16 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.

ST. LUCIE - UNIT 3/4 3-23 Amendment No. 59

('

TABLE 4.3-3

._I r- RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS rtA ni CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST _ REQUIRED C:

a,

-4 1. AREA MONITORS I-J

a. Fuel Storage Pool Area S n M *

b. Containment (CIS) S R M 6

c. Containment Area - High Range S R M 1, 2, 3 & 4 I

2. PROCESS MONITORS wp

a. Fuel Storage Pool Area -

.P Ventilation System 4>

i. Gaseous Activity S R M **

ii. Particulate Activity S R M

b. Containment

i. Gaseous Activity RCS Leakage Detection t S R M 1, 2, 3 & 4 ii. Particulate Activity RCS Leakage Detection S R M 1, 2, 3 & 4

.I

With fuel in the storage pool or building.

- With irradiated fuel in the storage pool.

n)

Dt I 0:

U'

., ,.. dlb,

(

TABLE 4.3-3 (Continued)

I-RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS C-)

m CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE I- INSTRUMENT CHECK CALIBRATION TEST IS REQUIRED

2. PROCESS MONITORS (Continued)

c. Noble Gas Effluent Monitors

i. Radwaste Building Exhaust System Plant Vent Monitor S R M 1, 2, 3 A 4 ii. Steam Generator Blowdown Treatment Building Exhaust 4:.I Sys tem S R M 1, 2, 3 & 4 '2 fl, iii. Steam Safety Valve Discharge R M 1, 2, 3 & 4 iv. ECCS Exhaust S R M 1, 2, 3 & 4 CL C,'

(.0 A., ,. . 4X b

PAGE 314 3-26 (ORIGINAL) HAS BEEN DELETED FROM THE TECHNICAL SPECIFICATIONS. THE NEXT PAGE IS 3/4 3-27.

ST. LUCIE - UNIT 1 3/4 3-25 Amendment No. 4*. 66, 33, 64, 136

Pages 3/4 3-28 through 3/4 3-32 have been DELETED.

The next page is 34 3-33.

ST. LUCIE - UNIT 1 314 3-27 Amerdment No. *e, 4w 147

INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either:

a. Restore the inoperable channel to OPERABLE status within 30 days, or I

.A b. Be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

. . iI

- e, SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

ST. LUCIE - UNIT 1 3/4 3-33

iS:c ..

'i'P, v saws4..

1I...

N..

i:.

C.I % IX6 TABLE 3.3-9 cn

-4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION I-MINIMUM READOUT MEASUREMENT CHANNELS INSTRUMENT LOCATION RANGE OPERABLE

1. Reactor Trip Breaker Indication SWGR OPEN-CLOSE 1/trip breaker P1_ 2. Pressurizer Pressure Hot Shutdown Panel 1500-2500 psia 1

3. Pressurizer Level Hot Shutdown Panel 0-100% I WI

4. Main Steam Pressure Hot Shutdown Panel 0-1200 psig 1/steam generator iII II i

5. Steam Generator Level Hot Shutdown Panel 0-100% I/steam generator

6. Cold Leg Temperature Hot Shutdown Panel

0-600 0F I

,,,;,: Amy

. . .., . ;. ..i . ..

.. 1.K.. I I

TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS I-C-,

CHANNEL CHANNEL a._ INSTRUMENT CHECK CALIBRATION

1. Reactor Trip Breaker Indication M N.A.

2. Pressurizer Pressure M R

3. Pressurizer Level M R

4. Steam Generator Level M R

5. Main Steam Pressure M R (4

6. Cold Leg Temperature M R (4*

(4 UI S,

i

THIS PAGE HAS BEEN DELETED ST. LUCIE - UNIT 1 3/4 3-36 Amendment No.*4 57 4

Pages 3/4 3-38 through 3/4 3-40 (Amendment No. 115) have been deleted from the Technical Specifications. The next page is 3/4 3-41. I ST. LUCIE - UNIT 1 3/4 3-37 Amendment No. 6, 49, l I, 134

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.

I APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. Actions per Table 3.3-11.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

ST LUCIE - UNIT 1 3/4 3-41 Amendment No. 37, 118

( (. C.

TABLE 3.3-11 ACCIDENT MONITORING INSTRUMENTATION MIN FIMUM TOTAL NO. CHAI NNELS INSTRUMENT OF CHANNELS OPEI RABLE ACTION

1. Pressurizer Water Level 2 I 1,6 I

2. Auxiliary Feedwater Flow Rate 1/pump 1/itpump 7

3. RCS Subcooling Margin Monitor 2 I 1.6

4. PORV Position Indicator I/valve /alve 2 Acoustic Flow Monitor

5. PORV Block Valve Position 1/valve Ih'alve 2 Indicator

6. Safety Valve Position 1/valve mvalve 3 Indicator

7. Incore thermocouples 4/core 2/ core 1, 6 quadrant quEidrant

8. Containment Sump Water Level 1* 1* 4, 5 (Narrow Range)

9. Containment Sump Water Level 2 4 5 (Wide Range)

10. Reactor Vessel Level Monitoring 2** 1**

4, 5 System

11. Containment Pressure 2 1, 6 The non-safety grade containment sump water level Instrument may be substituted.

Definition of OPERABLE: A channel is composed of eight (8) sensors in a probe, of which four (4)sensors must be OPERABLE.

ST. LUCIE - UNIT 1 3/4 3-42 Amendment No. 3W, M. 44, A4d 177

TABLE 3.3-11 (continued)

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels less than the Total No. of Channels shown in Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY within I

the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 - With position indication inoperable, restore the inoperable indicator to OPERABLE status or close the associated PORV block valve and remove power from its operator within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 3 - With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information once per shit to determine valve position.

ACTION 4 - With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3.3-11, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to the specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 5 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-1 1, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or

1. Initiate an alternate method of monitoring the reactor vessel inventory; and

2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and

3. Restore the Channel to OPERABLE status at the next scheduled refueling.

ACTION 6 - With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be at least in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 7 - With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be at least in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT I 3J4 3-43 Amendment No. a, 79, 165

C CA TABLE 4.3-7

-I I- ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS rti CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

-I

1. Pressurizer Water Level M R:

-a

2. Auxiliary Feedwater Flow Rate H R

3. Reactor Coolant System Subcooling Margin Monitor H R

4. PORV Position Indicator H R S. PORV Block Valve Position Indicator M R

-PS Wi 6. Safety Valve Position Indicator H R I I

7. Incore Thermocouples H R

8. Containment Sump Water Level (Narrow Range) H R

9. Containment Sump Water Level H R

10. Reactor Vessel Level Monitoring System H R

11. Containment Pressure H R o.3

3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION F

STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. :

APPLICABILITY: MODES 1 and 2.

ACTION:

With less than the above required reactor coolant pumps in operation. be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I ST LUCIE - 'Vll,.'IT I 3/4 4-1 Amendment No. 70, Z7, fM, 5 X

REACTOR COOLANT SYSTEM HOT STANDBY LIMING CONDMON FOR OPERATION 3.4.1.2 The reactor coolant loops listed below shall be OPERABLE and at least one of these reactor coolant loops shall be in operation.*

a. Reactor Coolant Loop A and its associated steam generator and at least one associated reactor coolant pump.

b. Reactor Coolant Loop B and its associated steam generator and at least one associated reactor coolant pump.

APPLICABILITY: MODE 3.

ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.1 and within one (1) hour initiate corrective action to return the required reactor coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one reactor coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.2.3 The required steam generators shall be determined OPERABLE by verify-ing the secondary side water level to be > 10% of narrow range indication at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.1 and (2) core outlet temperature is maintained at least 100F below saturation temperature.

ST. LUCIE - UNIT I 314 -1a Amendment No. 6. 179

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one reactor coolant or shutdown cooling loop shall be in operation.*

a. Reactor Coolant Loop A and its associated steam generator and at least one associated reactor coolant pump,

b. Reactor Coolant Loop B and its associated steam generator and at least one associated reactor coolant pump,

c. Shutdown Cooling Loop A, -

d. Shutdown Cooling Loop B.

APPLICABILITY: MODE 4.

ACTION:

a. With less than the above required reactor coolant or shutdown cooling loops OPERABLE, within one (1) hour initiate corrective action to return the required loops to OPERABLE status. If the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With no reactor coolant or shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.1 and within one (1) hour initiate corrective action to return the required reactor coolant loop to operation.

All reactor coolant pumps and shutdown cooling pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause introduction into the RCS, c:o.lant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.1 and (2) core outlet temperature is maintained at least 1OOF below saturation temperature.

ST. LUCIE - UNIT 1 314 4-1b Amendment No. a, 44 179

IREACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE REQUIREMENTS -

4.4.1.3.1 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker align-ments and indicated power availability.

4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be > 10o of narrow range indica-tion at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.3 At least one reactor coolant or shutdown t cooling loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT 1 3/4 4-1 c Amendment No. 5 6

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDION FOR OPERATION 3.4.1.4.1 At least one shutdown cooling loop shall be OPERABLE and in operation and either

a. One additional shutdown cooling loop shall be OPERABLE# or

b. The secondary side water level of at least two steam generators shall be greater than 10% of narrow range indication.

APPLICABILITY: MODE 5 with reactor coolant loops filled ACTION:

a. With less than the above required loops OPERABLE or with less than the required steam generator level, within one (1) hour initiate corrective action to return the required loops to OPERABLE status or to restore the required level.

b. With no shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.2 and within one (1) hour initiate corrective action to return the required shutdown loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.2 and

2) core outlet temperature is maintained at least I 00 F below saturation temperature.

One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other shutdown cooling loop is OPERABLE and in operation.

" A reactor coolant pump shall not be started with two idle loops unless the secondary water temperature of each steam generator is less than 300 F above each of the Reactor Coolant System cold leg temperatures.

ST. LUCIE - UNIT 1 3/4 4-1d Amendment No. 66, &. 43, 179

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two shutdown cooling loops shall be OPERABLE# and at least one shut-down cooling loop shall be in operation*.

APPLICABILITY: MODE 5 with reactor coolant loops not filled.

ACTION:

a. With less than the above required loops OPERABLE, within one (1) hour initiate corrective action to return the required loops to OPERABLE status.

b. With no shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet SHUTDOWN MARGIN of Technical Specification 3.1.1.2 and within one (1) hour initiate corrective action to return the required shutdown cooling loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other shutdown cooling loop is OPERABLE and in operation.

The shutdown cooling pump may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.1.1.2 and 2) core outlet temperature is maintained at least 10F below saturation temperature.

ST. LUCIE - UNIT I 3f4 4-le Amendment No. 6, 403,179

DELETED ST. LUCIE - UNIT 1 3/4 4-2 Amendment No. 90, 453,166

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

> 2422.8 psig and < 2560.3 psig.

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

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 < 281OF within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3 Verify each pressurizer code safety valves is OPERABLE in accordance with the Inservice Testing Program. Following testing, as-left lift settings shall be within

+/- 1% of 2500 psia.

ST. LUCIE - UNIT 1 314 4-3 Amendment No. 00, 453,166

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION a..umsssssa***** . z-m ~

amumu~s uw---- aam------------ ma ainus 3.4.4 The pressurizer shall be OPERABLE with a steam bubble, and with at least 150 kw of pressurizer heaters capable of being supplied by emergency power.-

APPLICABILITY: MODES 1 and 2.

ACTION:

With the pressurizer inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS inazinaazzainuinmmainsssstauuinumzmm.............

4.4.4 In accordance with 4.8.1.1.2.

ST LUCIE - UNIT I 3/4 4-4 Amendment No. 4r 112

REACTOR COOLANT SYSTEM I STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more steam generators inoperable, restore the Inoperable generatores) to OPERABLE status prior to increasing Tayg above 2000 F.

SURVEILLANCE REQUIREMENTS 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, Inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.

The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 apd the inspected tubes shall be verified acceptable per the acceptanee criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3%of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

'a. Whtte-experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least SOS of the tubes inspected shall be from these critical areas.

b. The first inservice inspection (subsequent to the preservice inspection) of.each steam generator shall include:

1. All nonplugged tubes that previously had detectable wall penetrations (.20), and -.

2. Tubes in those areas where experience has indicated potential problems.

ST. LUCIE - UNIT I 3/4 4-5 Amendment No. 108.

(1 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

c. The second and third inservice inspections may be less than a full tube inspection by concentrating (selecting at least 50%

of the tubes to the inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1X of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecu-tive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

ST. LUCIE - UNIT 1 3/4 4-6

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a third sampl-e inspection is not required.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with Ehe first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions.

1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,

2. A seismic occurrence greater than the Operating Basis Earthquake,

3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or

4. A main steam line or feedwater line break.

4.4.5.4 Acceptance Criteria

a. As used in this Specification:

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing.indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

3. Degraded Tube means a tube containing imperfections >20%

of the nominal wall thickness caused by degradation.

4. % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

ST. LUCIE - UNIT I 3/4 4-7

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIRENNTS (Continued)

5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

6. Plugging-Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.

7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8. Tube Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Renorts

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged:

in each steam generator shall be reported: to the Commission in a special report pursuant to Specification 6.9.2.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a special report pursuant to Specification 6.9.2 within 12 months following completion of the inspection. This special report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

ST. LUCIE - UNIT I 3/4 4-8 Amendment No. ;a, I10, I18

REACTOR COOLANT SYSTEM lll SURVEILLANCE REQUIREMENTS (Continued)

I I

I ST. LUCIE - UNIT 1 3/4 4-9 Amendment NO. 69

(

( /

H TABLE 4.4-1 C=

MINIMUM NUMBER OF STEAM GENERATORS TO BE P-0 INSPECTED DURING INSERVICE INSPECTION

--- ____- -__ __a' I Preservice Inspection aa No II Yes No. of Steam Generators per Unit Two I Three IFour Two Three First Inservice Inspection All One Two

.0k.

Second & Subsequent Inservice Inspections So & b uts cnp. oOO11 Onel 11One 2

.-a II One2 0~

Table Notation:

1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate all steam generators are performing in a like manner. Note that under some circumstances, the operating conditionsthatin one or. more steam generators may be found to be more severe than those in other steam generators. Under such circum-stances the sample sequence shall be modified to inspect the most severe conditions.

2. The other steam generator not inspected during the first inservice inspection shall ie inspected. The third and subsequent inspections should follow the instructions described In 1 above.

3. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent Inspections shall follow the instructions described in 1 above.

( ( (

TABLE 4.4-2

'A

-4 STEAM GENERATOR TUDE INSPECTION C:

1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Rlequired A minimum of C-1 None N/A N/A N/A N/A

-4 S Tubes per

-A S. G.

C-2 Plug detective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-? Plug defective tubes 4S tubes in this S. G.

Perform action for C-3 C-3 result of first sample w Perform action for C-3 C-3 result of first N/A N/A sample

.-a C-3 Inspect all tubes In All other this S. G., plug de. S. G.s ate None N/A N/A fective tubes and C- 1I_ _ _

Inspect 23 tubes in Some S. G., Perform action for N/AN/

each other S. G.

C-2 but no C-2 result of second N/A I additional sample S. G. are C-3 I I Additional Inspect all tubes in S. G. is C-3 each S. G. and plug defective tubes.

N/A N/A

~0

. I 3N % Where N is the number of steam generators In the unit, and n Is the number of steam generators inspected S

n during an Inspection

t The requiretent to Inspect all tubes may be relaxed for Cycle 5 'Refuelivgsince an engineering evaluation has sOoin that the condition(s) has been adequotely bounded by Inspection.

REACOR OOLNTSYSTEM 3M4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDMON FOR OPERATION 3.4.6.1 The following RCS leakage detection systems shall be OPERABLE:

a. The reactor cavity sump inlet flow monitoring system; and

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

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

ACTION:

a. With the required reactor cavity sump inlet flow monitoring system inoperable, perform a RCS water inventory balance at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and restore the sump inlet flow monitoring system to OPERABLE status within 30 days; 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the required radioactivity monitor inoperable, analyze grab samples of the containment atmosphere or perform a RCS water inventory balance at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and restore the required radioactivity monitor to OPERABLE status within 30 days; 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With all required monitors inoperable, enter LCO 3.0.3 immediately.

d. The provisions of Specification 3.0.4 are not applicable if at least one of the required monitors is OPERABLE.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The RCS leakage detection instruments shall be demonstrated OPERABLE by:

a. Performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION of the required containment atmosphere radioactivity

- monitor at the frequencies specified in Table 4.3-3.

b. Performance of the CHANNEL CALIBRATION of the required reactor cavity sump inlet flow monitoring system at least once per 18 months.

ST. LUCIE - UNIT 1 3t44t12 Amendment No. 144

DELETED ST. LUCIE - UNIT 1 3/4 4-13 Amendment No. 144

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. I GPM total primary-to-secondary leakage through steam generators,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for each Reactor Coolant System Pressure Isolation Valve identified in Table 3.4.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, 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 .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and Reactor Coolant System Pressure Isolation Valve leakage, reduce the leak-age rate to within limits 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 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the-following 3D hours.

c. With any Reactor Coolant System Pressure Isolation Valve leakage

-greater than thexlimit inz3.4.6-.2e above-reactor-operation may continue provided that at -east-two valves, including check valves, in each high pressure line having a non-functional valve are in and remain in the mode corresponding to the isolated con-dition. Motor operated valves shall be placed in the closed posi-tion, and power supplies deenergized. (Note, however, that this may lead to ACTION requirements for systems involved.) Otherwise, reduce the leakage rate to within limits 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 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 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous and particulate radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST LUCIE - UNIT 1 3/4 4-14

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation except when operating in the shutdown cooling mode,

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

e. Verifying each Reactor Coolant System Pressure Isolation Valve leakage (Table 3.4.6-1) to be within limits:

1. Prior to entering MODE 2 after refueling,

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

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

4. The provision of Specification 4.0.4 is not applicable for entry into MODE 3 or 4.

f. Whenever integrity of a pressure isolation valve listed in Table 3.4.6-1 cannot be demonstrated the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in each high pressure line having a leaking valve shall be recorded daily.

ST LUCIE - UNIT 1 3/4 4-14a Amendment No. 133

TABLE 3.4.6-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Check Valve No.

V3227 V3123 V3217 V3113 V3237 V3133 V3247 V3143 V3124 V3114 V3134 V3144 NOTES (a) Maximum Allowable Leakage (each valve):

1. Leakage rates less than or equal to 1.0 gpm are acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between previous measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50X or greater.

4. Leakage rates greater than 5.0 gpm are unacceptable.

(b) To satisy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

(c) Mimimum test differential pressure shall not be less than 150 psid.

ST LUCIIE - UNIT I 3/4 4-14b Order dated 4/20/81

REACTOR COOLANT SYSTEM CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1.

APPLICABILITY: ALL MODES. -

ACTION:

MODES 1, 2, 3 and 4

a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 .

-Y" < b. With any one or more chemistry parameter in excess of its Transient Limit, 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 .

MODES 5 and 6

With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer pressure to < 500 psia, if applicable, and perform an analysis to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.

ST. LUCIE - UNIT 1 314 4-15

TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN < 0.10 ppm* c 1.0 ppm*

CHLORIDE < 0.15 ppm < 1.50 ppm FLUORIDE < 0.10 ppm < 1.00 ppm Limit not applicable with Tavg < 250'F.

KI)

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS MINIMUM MAXIMUM TIME PARAMETER SAMPLING FREQUENCIES BETWEEN SAMPLES DISSOLVED OXYGEN 3 times per 7 days* 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CHLORIDE 3 times per 7 days 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months FLUORIDE 3 times per 7 days 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Not required with Tavg < 250'F.

ST. LUCIE - UNIT 1 3/4 4-16

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

a. < 1.0 uCi/gram DOSE EQUIVALENT I-131, and

b. _ 1ooiE uCi/gram.

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

ACTION:

MODES 1, 2 and 3*:

a. With the specific activity of the primary coolant >1.0 uCi/gram DOSE EQUIVALENT 1-131 for more than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months during one con-tinuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T ag c500°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 >lO/E uCi/gram, be in HOT STANDBY with Taq <500F 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 >1.0 uCi/gram DOSE EQUIVALENT 1-131 or >100/E uCi/gram, perform the sampling and analysis requirement of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

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 pro-gram of Table 4.4-4.

With T~v > 500°F.

ST. LUCIE - UNIT 1 314 4-17 Amendment No. 00, 101

REACTOR COOLANT SYSTEM DELETED ST. LUCIE - UNIT I 3/4 4-18 Amendment No. lol

.* . ."' F 0fs

(

TABLE 4.4-4 CA PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE

-4 AND ANALYSIS PROGRAM

'-4 C TYPE OF MEASUREMENT MINIMUM MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED 1-4

-a4

1. Gross Activity Determination 3 times per 7 days with a 1, 2, 3 and 4 maximum time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples

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

3. Radiochemical for E Determination 1 per 6 months 1

-8*

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1#, 2#. 3#, 4# and 5#

Including 1-131, 1-133, and 1-135 whenever the DOSE EQUIVALENT 1-131 exceeds 1.0 pCi/gram, and b) One sample between 2 1, 2, 3 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 limits.

After at least 2 EFPD and at least 20 days since the last shutdown of longer than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

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ST. LUCIE - UNIT 1 3J4 4-20

REACTOR COOLANT SYSTEM 3(4*.49 PRESSURErrEMPERATURE LIMITS REACTO)R COOLANT SYSTEM l MrnNG CONDION FOR OPPRAlION 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, 3A-2b and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

APPLICABILIY: At all times. *#

ACTION:

With any of the above limits exceeded, restore the temperature andlor pressure to within the lEmits within 30 minutes; perform an analysis to determine the effects of the out-of-imit 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 Tavg to less than 200DF within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months in accordance with Figures 3.4-2b and 3.44.

When the flow path from the RWT to the RCS via a single HPSI pump isestabished per 3.12.1 or 3.1.2-3 and RCS pressure boundary integrity exists, the heatup and cooldown rates shall be established in accordance with Fig. 3.1-lb.

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

ST. LLOE - UNIT I Y(4421 AmencknetNo.4,6,404, 152

REACTOR COOLANT SYSTEM SURVEILLANCE 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 material properties as required by 10 CFR 50 Appendix H. The results of these examinations shall be used to update Figures 3.4-2a, 3.4-2b and 3.4-3.

ST. LUCIE - UNIT I 3/4 4-22 Amendment No. 84., 113,

FIGURE 3.4-2a ST. LUCIE UNIT 1 PIT LIMITS, 23.6 EFPY HEATUP AND CORE CRmCAL L.. ~~~ISOTHERMAL' -

-+ - TO 509F/HR - =mm

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_1IGOL1TIP TEMP. 604zF=_

0 100 200 300 400 500 TC - INDICATED RCS TEMPERATURE, F ST. LUCIE - UNIT 1 3Y4 6-23a Amendment No. 4-. 0-. e4, 404, 141

FIGURE 3.4-2b ST. LUCIE UNIT 1 PIT UMITS, 23.6 EFPY COOLDOWN AND INSERVICE TEST 2500 SR VICE INSERVIC

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404, 141

FIGURE 3A43 ST. LUCIE UNIT 1, 23.6 EFPY I MAXIMUM ALLOWABLE COOLDOWN RATES

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- ,** .=.............t. ...... . . -.

100

-- RATE, OFIHHR TEMP. LIMIT, °F -

20 C125 _ _

30 125.145 80 - 40 145.165 -. -

50 165.185 cc 75 185-195

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a 8oI 100 120 140 160 180 200 Tc . INDICATED REACTOR COOLANT TEMPERATURE, °F NOTE: A MAXIMUM COOLDOWN RATE OF 100°F/HR IS ALLOWED AT ANY TEMPERATURE ABOVE 1950F ST. LUCIE - UNIT 1 3%4 4-23C Amendment No. +9,e0-. 6+.

40, 141

DELETED ST. LUCIE - UNIT I 314 4-24 Amendment No. 4-0,11l

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100°F in any one hour period,

b. A maximum cooldown of 200'F in any one hour period, and

c. A maximum Reactor Coolant System spray water temperature differential of 350eF.

APPLICABILITY: At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes;

(, perform an analysis to determine the effects of the out-of-limit con-dition on the fracture toughness properties of the pressurizer; deter-mine that the pressurizer remains acceptable for continued operation 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 pressurizer pressure to less than 500 psia within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady state operation.

ST. LUCIE - UNIT 1 3/4 4-25

REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL- INTEGRITY -

ASME CODE CLASS 1, 2, AND 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components (except steam generator tubes) shall be maintained in accordance with Specification 4.4.10.1.

APPLICABILITY: All MODES.

ACTION:

a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the. Reactor Coolant System temperature more than 50°F above the minimum temperature required by NDT considerations.

b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200°F.

c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.

ST. LUCIE - UNIT I 3/4 4-26 Amendment No. 0p, 90

i'l.. - - -;-

i I

I SURVEILLANCE REQUIREMENTS (Continued) 9 I

Pages 3/4 4-28 through 3/4 4-55 (Amendment No. 90), and Pages 3/4 4.56 through 3/4 4-57 (Amendment No. 80) have been deleted from the Technical Specifications. The next page is 3/4 4.58.

ST. LUCIE - UNIT 1 3/4 4-27 Amendment No. .90i, 134

REACTOR COOLANT SYSTEM PORY BLOCK VALVES LIMITING CONDITION FOR OPERATION 3.4.12 Each Power Operated Relief Valve (PORY) Block Valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one or more block valve(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.12 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.

'ST LUCIE - UNIT 1 3/4 4-58 Amendment No. A7, 4P, 86

REACTOR COOLANT SYSTEM POWER OPERATED RELIEF VALVES LIMITING CONDITION FOR OPERATION 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:

1. During cooldown when the temperature of any RCS cold leg is less than or equal to 215*F and

2. During heatup and isothermal conditions when the temperature of any RCS cold leg is less than or equal to 1930F.

b. A setpoint of less than or equal to 530 psia shall be selected:

1. During cooldown when the temperature of any RCS cold leg is greater than 215-F and less than or equal to 281-F.

2. During heatup and isothermal conditions when the temperature of any RCS cold leg is greater than or equal to 193-F and less than or equal to 304'F.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 3040F MODE 5, and MODE 6 when the head is on the reactor vessel; and the RCS is not vented through greater than a 1.75 square inch vent.

ACTION:

a. With one PORV inoperable in KODE 4, restore the inoperable PORV to OPERABLE status within 7 days; or depressurize and vent the RCS 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 PORY inoperable in NMES 5 or 6, either (1) restore the inoperable PORY 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 of the RCS 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 PORYs inoperable, restore at least one PORV to operable status or complete depressurization and venting of the RCS 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 the 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 every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e. In the event either the PORYs or the RCS 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 PORYs 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.

SURVEILLANCE REQUIREMENTS 4.4.13 Each PORY shall be demonstrated OPERABLE by:

a. Verifying the PORY 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 18 months.

ST. LUCIE - UNIT I 3/4 4-59 Amendment .E iHF 1 32

REACTOR COOLANT SYSTEM REACTOR COOLANT PUMP - STARTING LIMITING CONDITION FOR OPERATION f 3,4.14 If the steam generator temperature exceeds the primary temperature by more than 300F, the first idle reactor coolant pump shall not be started.

APPLICABILITY: MODES 4E and 5. .

I ACTION:

If a reactor coolant pump is started when the steam generator temperature exceeds primary temperature by more than 300F, evaluate the subsequent transient to determine compliance with Specification 3.4.9.1.

I 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 300F. I

\,i1#Reactor Coolant System Cold Leg Temperature is less than 304'F.

I ST. LUCIE - UNIT I 3/4 4-60 Amendment No. UZ,0Z. 104

IL REACTOR COOLANT SYSTEM 3/4.4.15 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.15 At least one Reactor Coolant System vent path consisting of two vent valves and one block valve powered from emergency buses shall be OPERABLE and closed at each of the following locations:

a. Pressurizer steam space, and

b. Reactor vessel head.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or be in 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 .

b. With both Reactor Coolant System vent paths inoperable, maintain the inoperable vent paths closed with power removed from the valve actuators of all .the vent-valves and block valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or begin 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 4.4.15 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

1. Verifying all manual isolation valves in each vent path are locked in the open position.

2. Cycling each vent valve through at least one complete cycle of full travel from the control room.

3. Verifying flow through the Reactor Coolant System vent paths during venting.

ST. LUCIE-UNIT 1 3/4 4-61 Amendment No. 68

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

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. Between 1090 and 1170 cubic feet of borated water,

c. A minimum boron concentration of 1720 PPM, and

d. A nitrogen cover-pressure of between 200 and 250 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 within 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 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 I

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

With pressurizer pressure Ž 1750 psia.

ST. LUCIE - UNIT 1 314 5-1 Amendment No. 157

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of > 1% of tank volume by verifying the boron concentration of the safety injection tank solution. This lafter surveillance is not required when the volume increase makeup source is the RWT and the RWT has not been diluted since verifying that the RWT boron concentration is equal to or greater than the safety injection tank boron concentration limit.

c. At least once per 31 days when the RCS pressure is above 1750 psia, by verifying that power to the isolation valve operator is removed by maintaining the breaker open under administrative control.

d. At least once per 18 months by verifying that each safety injection tank isolation valve opens automatically under each of the following conditions:

1. When the RCS pressure exceeds 350 psia, and

2. Upon receipt of a safety injection test signal.

ST. LUCIE - UNIT I 314 5-2 Amendment No.4sh-, 157

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - OPERATING LIMITING CONDMON FOR OPERATION 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, 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.

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. Inthe 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.

With pressurizer pressure > 1750 psia.

ST. LUCIE - UNrr 1 3/4 5-3 Amendment No. 28, 438, 4U, 177

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem 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 verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position

1. V-3659 1. Mini-flow 1. Open isolation

2. V-3660 2. Mini-flow 2. Open isolation

b. At least once per 31 days by:

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

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. At least once daily of the areas affected within containment by the containment entry and during the final entry when CONTAINMENT INTEGRITY is established. I

d. At least once per 18 months by:

1. Verifying proper operation of the open permissive interlock (OPI) and the valve open/high SDCS pressure alarms for isolation valves V3651, V3652, V3480, V3481.

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.

ST. LUCIE - UNIT 1 3/4 5-4 Amendment No. gg, fid 133

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTSicontinued)

e. At least once per 18 months, during shutdown, by:

1. Verifying that each automatic valve in the flow path 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 Pump.

b. Low-Pressure Safety Injection Pump.

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 on recirculation flow when tested pursuant to the Inservice Testing Program.

1. High-Pressure Safety Injection pumps: greater than or equal to 2571 ft.

2. Low-Pressure Safety Injection pumps: greater than or equal to 350 ft.

ST. LUCIE - UNIT 1 3/4 5-5 Amendment No. X6, GO, 4S3, 4M, 164 4r

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

THIS PAGE INTENTIONALLY LEFT BLANK ST. LUCIE - UNIT 1 3/4 5-6 Amendment No. 90.

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - SHUTDOWN LIMlITNG CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. In MODES 3* and 4 ,one ECCS subsystem composed of one OPERABLE high pressure safety injection pump and one OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a sump recirculation actuation signal.

b. Prior to decreasing the reactor coolant system temperature below 270°F a maximum of only one high pressure safety injection pump shall be OPERABLE with its associated header stop valve open.

c. Prior to decreasing the reactor coolant system temperature below 2360 F all high pressure safety injection pumps shall be disabled and their associated header stop valves closed except as allowed by Specifications 3.1.2.1 and 3.1.2.3.

APPLICABILITY: MODES 3* and 4.

MODES 5 and 6 when the Pressurizer manway cover is in place and the reactor vessel head is on.

ACTION:

a. With no ECCS subsystems OPERABLE in MODES 3* and 4X immediately restore one ECCS subsystem to OPERABLE status or be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With RCS temperature below 2700F and with more than the allowed high pressure safety injection pump OPERABLE or injection valves and header isolation valves open, immediately disable the high pressure safety injection pump(s) or close the header isolation valves.

c. 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.

SURVEILLANCE REOUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

4.5.3.2 The high pressure safety injection pumps shall be verified Inoperable and the associated header stop valves closed prior to decreasing below the above specified Reactor Coolant System temperature and once per month when the Reactor Coolant System is at refueling temperatures.

With pressurizer pressure < 1750 psia.

REACTOR COOLANT SYSTEM cold leg temperature above 2500F.

ST. LUCIE - UNIT 1 314 5.7 Amendment No. 98.a6, 84, 4O4,4.4, 177

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with:

a. A minimum contained volume 401,800 gallons of borated water,

b. A minimum boron concentration of 1720 ppm,

c. A maximum water temperature of 100 0 F,

d. A minimum water temperature of 55'F when in MODES I and 2, and

e. A minimum water temperature of 40'F when in MODES 3 and 4 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 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 3/4 5-8 Amendment No. 28

314.6 CONTAINMENT SYSTEMS 314.6.1 CONTAINMENT VESSEL CONTAINMENT VESSEL INTEGRITY UMITING CONDION FOR OPERATION 3.6.1.1 CONTAINMENT VESSEL INTEGRITY shall be maintained.

APPLICABIUTY: MODES 1, 2, 3 and 4.

Without CONTAINMENT VESSEL INTEGRITY, restore CONTAINMENT VESSEL INTEGRITY within one hour 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 .

el IMUFII

@ifWPII a~ IN 5.5flN I AL11%C 12CM110c"CLITC

... ,,.~IS 4.6.1.1 CONTAINMENT VESSEL INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that:

1. All containment vessel penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, 2.

or deactivated automatic valves secured in their positions, except for valves that are open on an intermittent basis under administrative control, and All containment vessel equipment hatches are closed and sealed.

I

b. By verifying that each containment vessel air lock is OPERABLE per Speclfication 3.6.1.3.

O Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

ST. WCIE - UNIT I 314 6-1 Amendment No. 44,4 149

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited in accordance with the Containment Leakage Rate Testing Program.

APPUCABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment leakage rate exceeding the acceptance criteria of the Containment Leakage Rate Testing Program, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate action to 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 . Restore the overall leakage rate to less than that specified by the Containment Leakage Rate Testing Program prior to increasing the Reactor Coolant System temperature above 2000F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the required test schedule and shall be determined in conformance with the criteria specified in the Containment Leakage Rate Testing Program.

ST. LUCIE - UNIT 1 3/4 6-2 Amendment No. 08, 0* 4Q; 149

, CONTAINMENT SYSTEMS SURVI L[ANCF RFaOUIFMFNTS (Mntinted)

Pages 3/4 6-4 through 314 6-9 have been DELETED.

Page 34 6-10 is the next valid page.

.1 ST. LLKCIE- UNrT 1 4 6-3 Amern~erd No. is, 4z;.4 15

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMMNG CONDmON FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate in accordance with the Containment Leakage Rate Testing Program.

APPUCABIIUTY: MODES 1. 2, 3 and 4.

ACTION:

a. With one containment air lock door inoperable-:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be dosed at least once per 31 days.

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

4. The provisions of Specification 3.0A are not applicable.

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 .

SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

If the inner air lock door is inoperable, passage through the OPERABLE outer air lock door is permitted to effect repairs to the inoperable inner air lock door. No more than one airlock door shall be open at any time.

ST. LUCIE - UNIT 1 3/4 6-10 Amendment No. 06 14 9

CONTAINMENT SYSTEMS CONTAINMENT AIR lOCKS SURVEILLANCE REQUIREMENTS (continued)

a. By verifying leakage rates and air lock door seals in accordance with the Containment Leakage Rate Testing Program; and

b. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

ST. WGIE - UNIT I 3% 6-11 Amendment No. 86, 4t 149

272 CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between -0.7 and 2.4 PSIG.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment internal pressure outside of the limits above, restore the internal pressure to within 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 3/4 6-12

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120"F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment average air temperature > 120'F, reduce the average air temperature to within the limit within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months si 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 .

i.

SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at three of the following locations and shall IDe determined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Location

a. Containment fan cooler No. IA air intake, elevation 45 feet.

b. Containment fan cooler No. lB air intake, elevation 45 feet.

c. Containment fan cooler No. IC air intake, elevation 62 feet.

d. Containment fan cooler No. lD air intake, elevation 45 feet.

ST. LUCIE - UNIT 1 3/4 6-13

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILlTY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined, in accordance with the containment Leakage Rate Testing Program, by a visual inspection of the accessible interior and exterior surfaces of the vessel and verifying no apparent changes in appearance of the surfaces or other abnormal degradation.

ST. LUCIE - uNrr i W4 6-14 Amendment No. 6S, 14.9

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains sball be OPERABLE.

APPIThIIT: Containment spray System: NODES 1, 2, and MODE 3 with Pressurizer Pressure P 17SO psL.

Containment Cooling System: MODES 1,2, and 3.

aCTION:

1. -fodes 1. 2. and 3 with Pressuri~tr Pressure a 1750 DsLas

a. With one containment spray train inoperable, restore the inoperable spray train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and within 10 days from initial discovery of failure to meet the LCOI otherwise be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within the following S4 hours.

b. With one containment cooling train inoperable, restore the inoperable cooling train to OPERABLE status within 7 days and within 10 days from initial discovery of failure to meet the LCO; otherwise be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one containment spray train and one containment cooling train inoperable, concurrently implement ACTIONS a. and b. The completion intervals for ACTION a. and ACTION b. shall be tracked separately for each train starting from the time each train was discovered inoperable.

d. with two containment cooling trains inoperable, restore one cooling train to OPERABLZ status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e* with two containment spray trains inoperable or any combination of three or macre trains inoperable, enter LCO 3.003 iimediately.

2. Mode 3 with Pressurizer Pressure c 17SO usias

a. With one containment cooling train inoperable, restore the inoperable cooling train to OPERABLS status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise be in MODE 4 within ths next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b* With two containment cooling trains inoperable, enter LCO 3 0.3 Jmmediately.

ST. LUCIE - UNIT 1 3/4 6-15 Amendment No*. 0,131

,qURYEFIILANQF REjOUIRFMFNMTS 4.62.1 Each containment spray system 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 postion, is positioned to take suction from the RWT on a Containment Pressure

- - High High test signal.

b. By verifying that on recirculation flow, each spray pump develops a discharge pressure of a 200 psig, when tested pursuant to the InseMce Testing Program.

ST. LUCIE - UNIT 1 314 6- 1 a ST.

LUCIE-1 3'4 UNIT kAemrnmnn o.4G1, 153 8-i~~~a

CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

c. At least once per 18 months, during shutdown, by:

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

2. Verifying that each spray pump starts automatically on a CSAS test signal.

3. Verifying that upon a recirculation actuation signal, the containment sump isolation valves open and that a recirculation mode flow path via an OPERABLE shutdown cooling heat exchanger m

is established.

d. At least once per 10 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

4.6.2.1.1 ach containment cooling train shall be demonstrated OPERALS '

a. At least once per 31 days bys

1. starting each cooling train fan unit frot the control room and verifying that each unit operates for at least 15 minutes, and

2. Verifying a cooling water flow rate of greater than or equal to 1200 go to each cooling unit.

b. At least once per 18 months, during shutdown, by verifying that each containment cooling train starts automatically on an Se- test signal.

ST. LUCIE - UNIT 1 3/4 6-16 Amendment No. .90-, 44-, 131

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with:

a. A spray additive tank containing a volume of between 4010 and 5000 gallons of between 28.5 and 30.5% by weight NaOH solution, and

b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a containment spray system pump flow.

APPLICABILITY: MODES 1, 2 and 3.*

ACTION:

With the spray additive system inoperable, restore the 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 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the spray additive system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.2 The spray additive system 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.

b. At least once per 6 months by:

1. Verifying the contained solution volume in the tank, and

2. Verifying the concentration of the NaOH solution by chemical analysis.

c. At least once per 18 months, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a CSAS test signal.

Applicable when pressurizer pressure is > 1750 psia.

T. LUCIE - UNIT 1 3/4 6-16a Amendment No. Rk. 94

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

d. At least once per 5 years by verifying a minimum sodium hydroxide (NaOH) flow-rate of 10.5 gpm from the spray additive tank to a drain connection immediately downstream of the tank outlet valve, and a demineralized water flow rate of 18 +/- 1.5 gpm from that same drain connection to each containment spray pump.

ST. LUCIE - UNIT I 3/4 6-16b Amendment No. op. 78

DELETED ST. LUCIE - UNIT 1 3/4 6-1 7 PAnendment No. ZE, 131

CONTAINMENT SYSTEMS 314.6.3 CONTAINMENT ISOLATION VALVES LIMIING CONDmON FOR OPERATION K>s 3.6.3.1 The containment isolation valves shall be OPERABLE:

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more of the isolation valve(s) inoperable, either

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange; or

d. 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 4.6.3.1.1 The isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of the cycling test, and verification of isolation time.

ST. LUCIE - UNIT I 314 6-18 ArSendment No. XG 149

CONTAINMENT SYSTEMS SUR VFILLANCE REOUIRFMENTS (ontinupd*

4.6.3.12 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by.

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

4.6.3.1.3 The isolation time of each power operated or automatic containment isolation Valve shall be determined to be within its limits when tested pursuant to the Inservice Testing Program.

ST. LUCIE- UNT 1 314 6-19 Amendment No. ,449, 153

Pages 3/4 6-21 through 3/4 6-22 have been DELETED.

Page 3/4 6-23 is the next valid page.

I ST. LUClE - UNIT 1 3/4 6-20 3(4 AkwmerBntNO.

6-20 06; 44444B,

~152

CONTAINMENT SYSTEMS 3!4.6.4 'OMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE.

APPLICABILITY:. MODES 1 and 2.

ACTION:

With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or demonstrate within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the grab sample system of the inoperable hydrogen analyzer has the capability to draw a sample of the containment atmosphere into the grab sample canister..

Verify this capability of the grab sample system at least once per 30 days thereafter. Return the inoperable hydrogen analyzer to OPERABLE status within an additional 60 days. 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 .

SURVEILLANCE REQUIREMENTS.

4.6.4.1.1 Each hydrogen analyzer shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing:

a. Nominally one volume percent hydrogen, balance nitrogensand oxygen.

b. Nominally four volume percent hydrogen, balance nitrogen, and oxygen.

ST. LUCIE - UNIT 1 314 6-23 Amendment No. 89

CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS - W LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent containment hydrogen recombiner systems shall be OPERABLE.

ADDI TAPTI TrY-a I. MODES 1 and 2.

Ir1 Iza ACTION:

I I With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 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 .

SURVEILLANCE REQUIREMENTS 4.6.4.2 Each hydrogen recombiner system shall be demonstrated OPERABLE:

. I_4.7

-1

a. At least once per 6 months by verifying during a recombiner system functional test that the minimum heater sheath temper-ature increases to > 7001F within 90 minutes and is maintained for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

b. At least once per 18 months by:

1. Performing a CHANNEL CALIBRATION of all recombiner instru-mentation and control circuits.

2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiners (i.e., loose wiring or structural connections, deposits of foreign materials, etc.).

ST. LUCIE - UNIT 1 3/4 6-24

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3. Verifying during a recombiner system functional test that the heater sheath temperature increases to > 1200OF within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and is maintained for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. Verifying the integrity of the heater electrical circuits by performing a continuity and resistance to ground test following the above required functional test. I The resistance to ground for any heater phase shall be

> 10,000 ohms.

ST. LUCIE - UNIT I 314 6-25 Amendment No. 169

CONTAINMENT SYSTEMS 314.635 VACUUM RELIEF VALVES LMMING CONDMON OR OPERAON 3.6.5.1 Two vacuum relief lines shalt be OPERABLE.

APPUCABILITY: MODES 1. 2, 3 and 4.

ACTION:

With one vacuum reflef raie inoperable, restore the vacuum relief line 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.5.1 Verify each vacuum relief lne OPERABLE in accordance with the Inservice Testing Program.

ST. LUCIE - UNIT I 3146-26 Amendment No. 9X. 452. 453. 182

CONTAINMENT SYSTEMS 3/4.6.6 SECONDARY CONTAINMENT SHIELD BUILDING VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.1. Two independent shield building ventilation systems shall be OPERABLE.

APPLICABLILITY: MODES 1, 2, 3 and 4.

ACTION:

With one shield building ventilation system inoperable, restore the inoperable system 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.6.1 Each shield building ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying.that the train operates for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the heaters on.

b. At least once per 18 months or (l) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ven-tilation zone communicating with the system by:

1. Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm + 10%.

2. Verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N5IO-1975 while operating the ventilation system at a flow rate of 6000 cfm + 10%.

3T. LUCIE - UNIT 1 3/4 6-27 Amendihent No. 27

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 97.5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989 (300 C, 70% RH). The carbon samples not obtained from test canisters shall be prepared by either a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 6000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by either

1. Verifying that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 97.5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989 (300C, 70% RH); or

2. Verifying that a laboratory analysis of at least two carbon samples demonstrate a removal efficiency of > 97.5% for radio-active methyl iodide when the samples are tested in accordance with ASTM D3803-1989 (300C, 70% RH) and the samples are pre-pared by either a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

ST. LUCIE - UNIT I 314 6-28 Amendment No. 167

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demon-strated OPERABLE by also:

a) Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm + 10%, and b) Verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm + 10%.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6.15 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm + 10%.

2. Verifying that the air flow distribution is uniform within 20% across HEPA filters and charcoal adsorbers when tested in accordance with ANSI N510-1975.

3. Verifying that the filtration system starts automatically on a Containment Isolation Signal (CIS).

4. Verifying that the filter cooling makeup air and cross con-nection valves can be manually opened.

5. Verifying that each system produces a negative pressure of

> 2.0 inches W.G. in the annulus within 2 minutes after a Cntainment Isolation Signal (CIS).

6. Verifying that the main heaters dissipate 30 + 3 kw and the auxiliary heaters dissipate.1.5 + 0.25 kwiwhen tested in accordance with ANSI N510-197-5. -

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 92% of the DOP when they are tested in-place in accordance witF ANSI N510-1975 while operating the filtration system at a flow rate of 6000 cfm + 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the filtration system at a flow rate of 6000 cfm + 10%.

ST. LUCIE - UNIT 1 3/4 6-29 Amendment No.27

CONTAINMENT SYSTEMS SHIELD BUILDING INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.2 SHIELD BUILDING INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without SHIELD BUILDING INTEGRITY, restore SHIELD BUILDING INTEGRITY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 .

SURVEILLANCE REQUIREMENTS 4.6.6.2 SHIELD BUILDING INTEGRITY shall be demonstrated at least once per 31 days by verifying that the door in each access opening is closed except when the access opening is being used for normal transit entry and exit.

ST. LUCIE - UNIT 1 3/4 6-30

CONTAINMENT SYSTEMS SHIELD BUILDING STRUCTURAL INTEGRITY UIMING CONDITION FOR OPERATION 3.6.6.3 The structural integrity of the shield building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.6.3.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the shield building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.6.3 The structural integrity of Me shield building shall be determined, in accordance with the Containment Leakage Rate Testing Program, by a visual inspection of the accessible interior and exterior surfaces of the shield building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation.

ST. LUCIE - UNIT I 3'4 6-31 Amendment No. sot 49

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

APPUCABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 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 .

b. The provisions of Specification 3.0A are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each main steam line code safety valve is OPERABLE in accordance with the Inservice Testing Program. Following testing, as-left lift settings shall be within +1-1% of 1000 psia for valves 8201 through 8208, and within +/- 1% of 1040 psia for valves 8209 through 8216 specified in Table 4.7-1.

ST. LUCIE - UNIT I 3/4 7-1 Amendment No. 8X, 463, 166

/( (

TABLE 3.7-1 tn MAXIMUM ALLOWABLE POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE H STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS I

C,

'-4 Maximum Allowable Power

'ii Maximum Number of Inoperable Safety Level-High Trip Setpoint Valves onAny Operating Steam Generator (Percent of RATED THERMAL POWER) 1 93.2 2 79.8 3 66.5

(-A 4*

I 1.

C C C TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (+ 1%to - 3%)

I Header A Header B

a. 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 c 995.3 psig

e. 8209 8213 > 994.1 psig and < 1035.7 psig

f. *8210 8214 > 994.1 psig and < 1035.7 psig

9. 8211 8215 > 994.1 psig andc 1035.7 psig

h. 8212 8216 > 994.1 psig and < 1035.7 psig ST. LUCIE - UNIT I 314 7-3 Amendment No.1 66

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor driven feedwater pumps, and

b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two motor driven pumps and one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying that each motor driven pump-develops a discharge pressure of > 1342 psig on recirculation flow.

I I

2. Verifying that the steam turbinfe driven pump develops a discharge pressure of > 1342 psig on recirculation flow. *

H.fhen not in MODES 1, 2 or 3, this surveillance shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after entering MODE 3 and prior to entering MODE 2.

ST. LUCIE - UNIT I 3/4 7-4 Amendment No. ;f, 90

PLANT SYSTEMS SURVEILLANCE WEOUIREMENTS (Continued) l

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

b. At least once per 18 months during shutdown by:

I

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

2. Verifying that each auxiliary feedwater pump starts automatically as designed upon receipt of the Auto Start actuation test signal. I ST. LUCIE - UNIT 1 3/4 7-5 Amendment No. P, 90

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 116,000 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 .

SURVEILLANCE 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 I ;."'4 7-6

PLANT SYSTEMS ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be

< 0.10 pCi/gram DOSE EQUIVALENT I-131.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the specific activity of the secondary coolant system > 0.10 uCi/

gram DOSE EQUIVALENT I-131, 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 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-2.

ST. LUCIE - UNIT 1 314 7-7

TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM AND ANALYSIS FREQUENCY

1. Gross Activity Determination 3 times per 7 days with a maximum time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months between samples

2. Isotopic Analysis for DOSE a) 1 per 31 days, when-EQUIVALENT I-131 Concentration ever the gross activity determination indicates iodine concentrations I

greater than 10% of the allowable limit.

b) 1 per 6 months, whenever the gross activity deter-mination indicates iodine concentrations below 10%

of the allowable limit.

ST. LUCIE - UNIT 1 3/4 7-8 Amenckr.ent no. 86

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES IMMINCI C0NDQMN FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODE 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or dosed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCF RFQUIRFMFNS 4.7.1.5 Each main steam line isolation valve that is open shall be demonstrated OPERABLE by verifying full closure within 6.0 seconds when tested pursuant to the Inservice Testing Program.

i ST. WCIE - UNIT 1 3%4 7-9 Amendment No. QG-, 44,153

to~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~p S.

Pages 3/4 7-11 through 3/4 7-12 (Amendment No. 86) have been deleted from the Technical Specifications. The next page is 3/4 7-13. l ST. LUCIE - UNIT 1 314 7-10 Amendment No. qi86 134

PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LIMITING CONDITrON FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 701F when the pressure of either cool-ant in the steam generator is > 200 psig. _

APPLICABILITY: ALL MODES.

ACTION:

With the requirements of the above specification not satisfied:

a. Reduce the steam generator pressure of the applicable side to c 200 psig within 30 minutes, and

b. Perform an analysis to determine the effect of the overpres-surization on the structural integrity of the steam generator.

( Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 2000 F.

SURVEILLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generators shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant in the steam generators is

< 701F.

ST. LUCIE - UNIT 1 3/4 7-13

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITINiG CONDITION FOR OPERATION 3.7.3.1 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed or otherwise secured in position, is I

in its correct position.

b. At least once per 18 months during shutdown by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection Actuation Signal. I ST. LUCIE - UNIT 1 3/4 7-14 Amendment No. go

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

II I

i I

THIS PAGE INTENTIONALLY LEFT BLANK ST. LUCIE - UNIT 1 3/4 7-15 Amendment No. 90

IPLANT SYSTEMS 3/4.7.4 INTAKE COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.' At least two independent intake cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one intake cooling water loop OPERABLE, restore at least two loops 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 COLD SHUTDOVIN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.4.1 At least two intake cooling water loops shall be demonstrated OPERABLE:

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

b. At least once per 18 months during shutdown by-verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection Actuation signal.

ST. LUCIE - UNIT 1 3/4 7-16 Amendment No. 90

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

THIS PAGE INTENTIONALLY LEFT BLANK ST. LUCIE - UNIT 1 3/4 7-17 Arnendment No. 90

PLANT SYSTEM4S 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5.1 The ultimate heat sink shall be OPEPRABE with:

a. Cooling water from the Atlantic Ocean providing a water level above

-10.5 feet elevation, Mean Low Water, at the plant intake structure,,

and

b. Two OPERABLE valves in the barrier dam between Big Hud Creek and the intake structure. l APPLICABILITY: At all times.

ACTION:

a. With the water level requirement of the above Specification not satisfied, be in at least HOT STANDBY within six hours and provide cooling water from Big Mud Creek within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one isolation valve in the barrier dam. between Big Mud Creek aviJ the intake structure inoperable, restore the inoperable valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or, within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , install a temporary flow barrier and open the barrier dam isola-tion valve. The availability of the onsite equipment capable of reinovi-ng the barrier shall be veri-ie-d at least once per seven days thereafter.

C. Uith both of the isolation valves in the barrier dam between the in-take structure and Big Mud Creek inoperable, within 24-hours either:

1) Install both temporary flow barriers and manually open both barrier dam isolation valves. The availability of the onsite equipment capable of removing the barriers shall be verified at least once per seven days thereafter, or

2) 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 .

- d. The provisions of Specification 3.0.4 are not applicable.

ISURVEILL.-;^E REQUJIREJIENTS 4.7.5.1.1 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the average water level to be within limits.

4.7.5.1.2 The isolation valves in the barrier wasi between the intake structure and Bix t,1i Creek shall be demonstrated OPERABLE at least once per six months by cycling each valve through at least one comlpete cycle of full travel.

ST. LUCIr - "'.1IT I 3/4 7-18 Amendment No. 54

N-Page Intentionally Blank l ST. WCIE - UNrr 1 3/ 7-1 9 Amendment No. i, 142

PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7.1 The control room emergency ventilation system shall be OPERABLE with:

a. Two booster fans,

b. Two isolation valves in each outside air intake duct,

c. Two isolation valves in the toilet area air exhaust duct,

d. One filter train,

e. At least two air conditioning units, and

f. Two isolation valves in the kitchen area exhaust duct.

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

a. With one booster fan inoperable, restore the inoperable fan to OPE.RABLE 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one isolation valve per air duct inoperable, operation may continue provided the other isolation valve in the same duct is maintained closed; 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the filter train inoperable, restore the filter train to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 .

d. With only one air conditioning unit OPERABLE, restore at least two air conditioning units 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ST. LUCIE - UNIT 1 3J4 7-20 Amendment No. 160

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.7.1 The control room emergency ventilation system 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 verifying that the control room air temperature is < 1200 F.

b. At least once per 31 days by:

1. Initiating flow through the HEPA filter and charcoal adsorber train and verifying that each booster fan operates for at least 15 minutes.

2. Starting (unless already operating) each air conditioning unit and verifying that it operates for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housing, or (2) following painting, fire or chemical release in any venti-lation zone communicating with the system by.

1. Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10%.

2. Verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N51 0-1 975 while operating the ventilation system at a flow rate of 2000 cfm + 10%.

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 97.5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989 (30-C, 70% RH). The carbon samples not obtained from test canisters shall be prepared by either:

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or ST. LUCIE - UNIT I 314 7-21 Amendment No. 167

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 2000 cfm +/- 10% during system operation when tested in accordance with ANSI N51 0-1975.

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by either

1. Verifying that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 97.5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989 (301C, 70% RH); or

2. Verifying that a laboratory analysis of at least two car-bon samples demonstrate a removal efficiency of > 97.5%

for radioactive methyl iodide when the samples are tested in accordance with ASTM D3803-1989 (300C, 70% RH) and the samples are prepared by either a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrat-ed OPERABLE by also:

a) Verifying that the charcoal adsorbers remove > 99%

of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm +/- 10%, and b) Verifying that the HEPA filter banks remove > 99%

of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10%.

ST. LUCIE -UNIT 1 3/47-22 Amendment No. 167

PLANT SYSTE44S SURVEILLANCE REQUIREMENTS (Continued)

e. At least once per 18 months by:

1. Verifying that the pressure drop across the combined ffEPA filters and charcoal adsorber banks is < 4.15 inches Water Gauge while operating the ventilation system at a flow rate of 2000 cfrn + 10%.

2. Verifying that on a containment isolation signal the system automatically isolates the control room within 35 seconds and switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

3. Verifying that the system maintains the control room at a positive pressure af > 1/8 inch W.G. relative to the Outside atmosphere during system operation with < 450 cfrm outside air intake.

. Afler each complete or Dartial replac.emnt of a HIEPA filter bank by verifying that. the HEPA filter banks remove > 99, of the DOP when they are tested in-place in accordance with ANSI NH10-1975 while operating the ventilation system at a flow rate of 2000 cfm - 10%.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 99%_ of a halogenated hydrocarbon refrigerant. test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 2000 cfm + 10%.

ST. LUCIE - U'OT 1 3/4 7-233

- Pime nd ;e nt No. Z', 57'

PLANT SYSTEMS 3/4.7.8 ECCS.AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8.1 Two independent ECCS area exhaust air filter trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one ECCS area exhaust air filter train inoperable, restore the inoperable train 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.8.1 Each ECCS area exhaust air filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiat-ing, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

2. Verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10'?.

ST. LUCIE - UNIT 1 3/4 7-24

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 97.5% for radio-active methyl iodide when the sample is tested in accordance with ASTM D3803-1989 (30 0C, 70% RH). The carbon samples not obtained from test canisters shall be prepared by either a) Emptying one entire bed from a removed adsorber tray, mbdng the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4 Verifying a system flow rate of 30,000 cfm + 10% during system operation when tested in accordance with ANSI N51 0-1975.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by either.

1. Verifying that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of > 97.5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989 (300 C, 70% RH); or

2. Verifying that a laboratory analysis of at least two carbon samples demonstrate a removal efficiency of > 97.5% for radioactive methyl iodide when the samples are tested in accordance with ASTM D3803-1989 (30C, 70% RH) and the samples are prepared by either.

a) Emptying one entire bed from a removed adsorber tray, mixng the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixng the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

ST. LUCIE - UNIT 1 314 7-25 Amendment No. 167

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrat-ed OPERABLE by also:

a) Verifying that the charcoal adsorbers remove > 99%

of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%, and b) Verifying that the HEPA filter banks remove > 99%

of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

d. At least once per 18 months:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4.15 inches Water Gauge while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

2. Verifying that the air flow distribution is uniform within 20% across HEPA filters and charcoal adsorbers when tested in accordance with ANSI N510-1975.

3. Verifying that the filter train starts on a Safety Injec-tion Actuation Signal.

e. After each complete or partial replacement of a HEPA fi.lter bank by verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 99% of a halogenated hydrocarbon refrigerant test gas when they. are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 30,000 cfm + 10%.

ST. LUCIE - UNIT 1 3/4 7-26

PLANT SYSTEMS 3/4.7.9 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.9.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of > 0.005 microcu-ries of removable contamination. -

APPLICABILITY: At all times.

ACTION:

a. Each sealed source with removable contamination in excess of the above limit shall be immediately withdrawn from use and:

1. Either decontaminated and repaired, or

2. Disposed of in accordance with Commission Regulations.

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

SURVEILLANCE REQUIREMENTS 4.7.9.1.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.9.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequencies described below.

a. Sources in use (excludingstartup sources previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive material:

ST. LUCIE - UNIT 1 3/4 7- 27

. 40116..

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1. With a half-life greater thin 30 days (excluding Hydrogen 3), and

2. In any form other than gas.

b. Stored sources not in use - Each sealed source shall-be tested prior to use or transfer to another licensed'-unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c. Startup sources - Each sealed startup source shall be tested within 31 days prior to being subjected to core flux and following repair or maintenance to the source.

4.7.9.1.3 Reports - A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days if source leakage tests reveal the presence of > 0.005 microcuries of removable.

contamination.

ST. LUCIE - UNIT 1 3/4 7-28

PLANI -5YSIEK 3/4 7.10 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.10 All safety related snubbers shall be OPERABLE.

APPLICABSILTY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES).

ACTI :

With one or more safety related snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber(s) to OPERABLE status or declare the supported system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 4.7.10 Each snubber shall be demonstrated OPERABLE by perfarmance of the following augmented inservice inspection program.

a. Inspection Types As used in this specification, 'type of snubber' shall mean snubbers of the same design and manufacturer, irrespective of capacity.

b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these categories (inaccessible or accessible) may be inspected independently according to the schedule determined by Table 4.7-3. The visual inspection interval for each category of snubber shall be determined based upon the criteria providje in Table 4.7-3 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Amendment ST. LUCIE - UNIT I 3/4 7-29 Amendment No. 24.,- 44 , 8a, 110,

PLANT SYSI SURVEILLANCE REQUIREMENTS (Continued) u.uuuauunasuiuuuuUUusUna~ mamawmmm.umu.auauauua n

c. Visual Insoection Acceptance Criteria Visual Inspections shall verify that (1)the snubber has no visible indications of damage or impaired OPERABILITY, (2)attach-merits to the foundation or supporting structure are functional, and (3) fasteners for the attachment of the snubber to the compo-nent and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (i) the cause of the rejection is clearly estab-lished and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (ii)the affected snubber is functionally tested in the as-found condition and determined OPERABLE per specifica-tions 4.7.10.e and 4.7.10.f. as applicable. All snubbers found connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable for determining the next Inspection interval. A review and evaluation shall be performed and docu-mented to justify continued operation with an unacceptable snub-ber. If continued operation cannot be justified, the snubber shall be declared inoperable and the ACTION requirements shall be met.

TABLE 4.7-3 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABLE SNUBBERS Population Column A Column B Column C or Category Extend Interval Repeat Interval Reduce Interval ftf4nac 61 v I

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1 1 0 0 1 80 0 0 2 100 0 1 4 150 0 3 8 200 2 5 13 300 5 12 25 400 8 18 36 500 12 24 48 750 20 40 78 1000 or greater 29 56 109 ST. LUCIE - UNIT 1 3/4 7-29a Amendment No. 110,

SURVEILLANCE REQUIREMENTS (Continued)

.muauauusauaaaaauUUU*UUUUsnUUU3UU3UUEUUUUUUinUU Note 1: The next visual inspection interval for a snubber population or category size shall be determined based upon the previous inspec-tion interval and the number of unacceptable snubbers found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessi-ble. These categories may be examined separately or Jointly.

However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to determine the next inspection interval for that category.

Note 2: Interpolation between population or category sizes and the number of unacceptable snubbers is permissible. Use next lower integer for the value of the limit for Columns A, 8, or C if that integer includes a fractional value of unacceptable snubbers as determined by interpolation.

Note 3: If the number of unacceptable snubbers is equal to or less than the number in Colua* A, the next inspection interval may be twice the previous interval but not greater than 48 months.

Note 4: If the number of unacceptable snubbers is equal to or less than the number in Column B but greater than the number in Column A, the next inspection interval shall be the same as the previous inter-val.

Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unac-ceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one-third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Columns B and C.

Note 6: The provisions of Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months.

ST. LUCIE - UNIT I 3/4 7-29b AWMendment No. 110,

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. Functional Tests At least. once per 18 months during shutdown, i representative sample (10% of the-safety related snubbers) shall be functionally tested eithet in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of Specification 4.7.10.e or 4.7.10.f, an additional 10% of that type of snubber shall be functionally tested. Functional test shall continue until no additional snubbers are found inoperable or all safety related snubbers have been tested.

The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers.

Snubbers identified as *Especially Difficult to Removee or in 'High Exposure Zones During Shutdown' shall also be Included in the representative sample.* Safety related hydraulic snubber listings and safety related mchanical snubber listings may be used jointly or separately as the basis for the sapling plan.

In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period. If a spare snubber has been installed in place of a failedi snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested. Test results of these snubbers shall not result in additional functional testing due to failure.

Permanent or other exemptions from the functional testing for individual snubbers in these categories may be granted by the Commission only if justifiable basis for exemption is presented and/or snubber life destructive testing was performed to qualify snubber operability for all design conditions at either the completion of their fabrication or at a subsequent date.

Amendment No. #I, 44 v.83 110, ST. LUCIE - UNIT I 3/4 7-30

SURVEILLANCE REQUIREMENTS (Continued)

M. -- -- OM O M M M --- M- -

If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency, all snubbers of the same design subject to the same defect shall be functionally tested.

This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional test acceptance criteria.

e. Hydraulic Snubbers Functional Test Acceptance Criteria The hydraulic snubber functional test shall verify that:

1. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

2. Snubber bleed, or release rate, where required, is within the specified range in compression or tension.

f. Mechanical Snubbers Functional Test Acceptance Criteria The mechanical snubber functional test shall verify that:

1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.

2. Activation (restraining action) is achieved in both tension and compression.

g. Snubber Service Life Monitorino A record of the service life of each snubber, the date at which the designated service life commences and the Installation and maintenance records on which the designed service life Is based shall be maintained as required by Specification 6.10.2.1.

Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each safety related snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded by more than 10% prior to the next scheduled snubber service life review. If the indicated service life will be exceeded by more than 10% prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. The results of the reevaluation may be used to justify a change to the service life of the snubber. This reevaluation, replacement or reconditioning shall be indicated in the records.

ST. LUCIE - UNIT I 3/4 7-31 Amendment No. 344r-B3,

.- 110

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314.8 ELECTRICAL POWER SYSTEMS 314.8.1 A.C. SOURCES OPERATING LIMITING CONDION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical mcwer 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 generator sets each with:

1. Engine-mounted fuel tanks containing a minimum of 152 gallons of fuel,

2. A separate fuel storage system containing a minimum of 16,450 gallons of fuel, and

3. A separate fuel transfer pump.

APPLICABILlTY: 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 4.8.1.1.2.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 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 1 3J4 8-1 Amendment No. 403, 44.438,4.0,180

ELECTRICAL POWER SYSTEMS ACTION (continued)

c. With one offsite A.C. circuit and one diesel generator inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within one hour 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 .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 at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 . Restore the other A.C. power source (offsite circuit or diesel generator) to OPERABLE status in accordance with the provisions of Section 3.8.1.1 ACTION Statement a or b, as appropriate, with the time requirement of that ACTION Statement based on the time of the initial loss of the remaining inoperable A-C. power source.

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.

d. With two of the required offsite A-C. circuits inoperable, restore one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 . Following restoration of one offsite source, follow ACTION Statement a. with the time requirement of that ACTION Statement based on the time of the initial loss of the remaining inoperable offsite A.C. circuit 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 1 314 8-2 Amendment No. 403. . 442. 180

ELECTRICAL POWER SYSTEMS ACTQIN (continued)

e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits 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 one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-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 . Following restoration of one diesel generator unit, follow ACTION Statement b. with the time requirement of that ACTION Statement based on the time of initial loss of the remaining inoperable diesel generator.

f. With one Unit 1 startup transformer (1A or IB) inoperable and with a Unit 2 startup transformer (2A or 2B) connected to the same A or B offsite power circuit and administratively available to both units, then should Unit 2 require the use of the startup transformer administratively available to both units, Unit I shall 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 inoperable startup transformer 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 .

SURVEILLANCE REQUiREMENTS 4.8.1.1.1 Each of the above required Independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:

a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availabiitry, and

b. Demonstrated OPERABLE at least once per 18 monfs by transferring (manually and automatically) unit power supply from the auxiliary transformer to the startup transformer.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1. Verifying fuel level in the engine-mounted fuel tank,

2. Verifying the fuel level in the fuel storage tank,

3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the engine-mounted tank, ST. LUCIE - UNIT I 314 83 Amenent No. ies,. ++,138

ELECTRICAL POWER SYSTEMS SU L (continued)

4. Verifying the diesel starts from ambient condition and accelerates to approximately 900 rpm in less than or equal to 10 seconds".

The generator voltage and frequency shall be 4160 +/- 420 volts and 60

1.2 Hz within 10 seconds after the start signal". The diesel generator shall be started for this test by using one of the following signals:

a) Manual'Local b) Simulated loss-of-offsite power by itself.

c) Simulated loss-of-offsite power in conjunction with an ESF actuation test signal.

d) An ESF actuation test signal by itself.

5. Verifying the generator is synchronized, loaded to greater than or equal to 3500 kW in accordance with the manufacturer's recommendations and operates within a load band of 33D0 to 3500 kW*" for at least an additional 60 minutes, and

6. Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

b. By removing accumulated water.

1. From the engine-mounted fuel tank at least once per 31 days and after each occasion when the diesel is operated for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

2. From the storage tank at least once per 92 days.

The diesel generator start (10 sec.) from ambient conditions shall be performed at least once per 184 days in these surveillance tests. ARl other diesel generator starts for the purposes of this surveillance testing may be preceded by an engine prelube period and may also include warmup procedures (e.g., gradual acceleration) as recommended by the manufacturer so that mechanical stress and wear on the diesel generator is minimized.

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

ST. LUCIE - UNIT 1 3/4 8-4 Amendment No. 4f. 66 e,+ie ,138

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. By sampling new fuel in accordance with ASTM D4057-81 prior to addition to the storage tanks and:

1. By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:

a) API Gravity within 0.3 degrees at 600 F or a specific gravity of within 0.0016 at 601600 F,when compared to the supplier's certificate or an absolute specific gravity at 60/600F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity of 60°F of greater than or equal to 27 degrees but less than or equal to 39 degrees.

b) A kinematic viscosity at 40°C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification.

c) A flash point equal to or greater than 1251F, and d) A clear and bright appearance with proper color when tested in accordance with ASTM D4176-82.

2. By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.

d. At least once every 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-83 and verifying that total particulate contamination is less than 10 mg/liter when checked in accordance with ASTM D2276-83, Method A, or Annex A-2.

e. At least once per 18 months during shutdown by:

1. DELETED

2. Verifying generator capability to reject a load of greater than or equal to 600 hp while maintaining voltage at 4160 + 420 volts and frequency at 60 + 1.2 Hz.

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

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

ST. LUCIE - UNIT 1 3/4 8-5 Amendment No. 13, 403,168

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 + 420 volts and 60 + 1.2 Hz during this test.

4. Verifying that on an ESF actuation test signal (without loss-of-offsite power) the diesel generator starts*** on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The steady state generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 10 seconds after the auto-start signal; the generator voltage and frequency shall be maintained within these limits during 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 + 420 volts and 60 + 1.2 Hz during this test.

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 signal.

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

ST. LUCIE - UNIT 1 3/4 8-6 Amendment No. "i, 1O3

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

6. 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****. 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 3960 kWt and 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 3300 to 3500 kW#. The generator voltage and frequency shall be 4160 i 420 volts and 60 i 1.2 Hz within 10 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test.

7. Verifying that the auto-connected loads do not exceed the 2000-hour rating of 3730 kW.

8. 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 loads to the offsite power source, and c) Be restored to its standby status.

9. 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.

10. 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.

11. Verifying that the automatic load sequence timers are operable with the interval between each load block within t 1 second of its design interval.

f. At least once per ten years or after any modification which could affect diesel generator independence by starting** the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to approximately 900 rpm in less than or equal to 10 seconds.

MThis 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 I 3/4 8-6a. Amendment No. god 112

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

g. At least once per ten years by.

1. Draining each fuel storage tank, removing the accumulated sediment and cleaning the tank using an appropriate cleaning compound, and

2. Performing a pressure test of those portions of the diesel fuel oil system designed to USAS B31.7 Class 3 requirements in accordance with the Inservice Inspection Program. I 4.8.1.1.3 Reports - (Not Used) 4.8.1.1.4 The Class IE underground cable system shall be demonstrated OPERABLE within 30 days after the movement of any loads in excess of 80% of the ground surface design basis load over the cable ducts by pulling a mandrel with a diameter of at least 80% of the ducts inside diameter through a duct exposed to the maximum loading (duct nearest the ground's surface) and verifying that the duct has not been damaged.

ST. LUCIE - UNIT 1 3/4 8-6b Amendment No. 49,44a, 44a. 438181

TABLE 4.81 DIESEL GENERATOR TEST SCHEDULE (NOT USED)

ST. LUCIE - UNIT I 364 8 c Amendment No. 4eG. 138

ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONIDmON 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 gallons of fuel,

2. A fuel storage system containing a minimum of 16,450 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. 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 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 4.8.1.2.1 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 1 3/4 8-7 Amendment No. 403.43,. 179

- i

.. f

...N ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generator -sets:-

Pr 4160 volt Emergency Bus 1A3 4160 volt Emergency Bus IB3 480 volt Emergency Bus 1A2 480 volt Emergency Bus 1B2 480 volt Emergency MCC Busses lAS, 1A6, 1A7

.4 480 volt Emergency MCC Busses 1B5, 186, 1B7 120 volt A.C. Instrument Bus IMA 120 volt A.C. Instrument Bus 1MB 120 volt A.C. Instrument Bus 1MC 120 volt A.C. Instrument Bus IMD APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-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 .

SURVEILLANCE REQUIREMENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying indicated power availability.

ST. LUCIE - UNIT 1 3/4 8-8

._2fb-ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following A.C. electrical busses shall be OPERABLE and energized from sources of power other than a diesel gene-

'rator set but aligned to an OPERABLE.diesel generator set:

1 - 4160 volt Emergency Bus 1 - 480 volt Emergency Bus 3 - 480 volt Emergency MCC Busses 2 - 120 volt A.C. Instrument Busses APPLICABILITY: MODES 5 and 6 ACTION:

i-With less than the above complement of A.C. busses OPERABLE and energized, establish CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.8.2.2 The specified A.C. busses shall be determined OPERABLE and energized from A.C. sources other than the diesel generators at least once per 7 days by verifying indicated power availability.

I ST. LUCIE - UNIT 1 3/4 8-9

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.3 As a minimum the following D.C. electrical sources shall be OPERABLE:

a. 125-volt D.C. bus No. lA, 125-volt Battery bank No. 1A and a full capacity charger.

b. 125-volt D.C. bus No. lB, 125-volt Battery bank No. 1B and a full capacity charger.

APPLICABILITY: MDDES 1, 2, 3 and 4.

ACTION:

a. With one of tte required battery banks or busses inoperable, restore the inoperable battery bank or bus to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-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 .

- b. With one of the required full capacity chargers inoperable, demonstrate the OPERABILITY of its associated battery banks by performing Surveillance Requirement 4.8.2.3.2.a.1 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. If any Category A limit in Table 4.8-2 is not met, declare the

, . battery inoperable.

SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized at least once per 7 days by verifying indicated power availability.

4.8.2.3.2 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The parameters in Table 4.8-2 meet the Category A limits,-

and

2. The total battery terminal voltage is greater than or equal to 129-volts on float charge.

ST. LUCIE - UNIT 1 314 8-10 Amendment No. 61

ELECTRICAL POS4ER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 92 days and within 7 days after a battery dis-charge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:

1. The parameters in Table 4.8-2 meet the Category B limits,

2. There is no visible corrosion at either terminals or connec-tors, or the connection resistance of these items is less than 150 x 10:6 ohms, and

3. The average electrolyte temperature of 10% (60 cells total) of connected cells is above 50'F.

c. At least once per 18 months by verifying that:

1. The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,

2. The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material,

3. The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10o6 ohms, and

4. The battery charger will supply at least 300 amperes at 140 volts for at least 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads:for the design duty cycle when the battery is subjected to a battery service test.

e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance dis-charge test may be performed in lieu of the battery service test required by Surveillance Requirement 4.8.2.3.2.d.

f. Annual performance discharge tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

ST. LUCIE-.- UNIT I 3/4 8-11 Amendment No. 61

TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENT CATEGORY A(1 ) CATEGORY B(2)

Parameter Limits for each Limits for each Allowable(3) designated pilot connected cell value for each cell connected cell Electrolyte >Minimum level >Minimum level Above top of Level indication mark, indication mark, plates and not and < 1/4" above and < 1/4" above overflowing maximum level maxiimum level indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts(c) *> 2.07 volts Specific > 1.195() > 1.190 Not more than Gravity(a) .020 below the Average of all average of all connected cells connected cells

> 1.200 Average of all connected cells

> 1.190(b)

.,~

(a) Corrected for electrolyte temperature and level.

(b) Or battery charging current is less than 2 amps when on charge.

(c) Corrected for average electrolyte temperature.

(1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months all the Category B measurements are taken and found to be within their allowable values; and provided all Category A and B parameter(s) are restored to within limits within the next 6 days.

(2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within their allowable values and provided the Category B parameter(s) are restored to within limits within 7 days.

(3) With any Category B parameter not within its allowable value, declare the battery inoperable.

ST. LUCIE - UNIT 1 3/4 8-12 Amendment No. 61

ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE:

1 - 125-volt D.C. bus, and 1 - 125-volt battery bank and charger supplying the above D.C. bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above complement of D.C. equipment and bus OPERABLE,9 establish CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. bus shall be determined OPERABLE and energized at least once per 7 days by verifying indicated power availability.

4.8.2.4.2 The above required 125-volt battery bank and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

ST. LUCIE - UNIT 1 3/4 8-13

314.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDION FOR OPERATION 3.9.1 With 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 1720 ppm boron or its equivalent to restore boron concentration to within limits. I 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 I 314 9-1 Amendment No. 442,450. 171

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDION FOR OPERATION 3.9.2 As a minimum, two wide range logarithmic neutron flux monitors shall be operating, each with continuous visual indication in the con-trol room and one with audible indication in the containment.

APPLICABILITY: MODE 6.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.2 Each wide range logarithmic neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL FUNCTIONAL TEST at least once per 7 days.

b. A CHANNEL FUNCTIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the start of CORE ALTERATiONS, and

c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

ST. LUCIE - UNIT 1 3/4 9-2 Amendment No.179

I REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

APPLICABILITY: During movement of irradiated pressure vessel.

fuel in the reactor I

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , suspend all operations involving movement of irradiated fuel in the reactor pres-sure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

UCIE - UNIT I ST. LI 3/4 9-3 Amendment No. 102

REFUELING OPERATIONS CONTAINMENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a. The equipment door closed and held in place by a minimum of four bolts.

b. A minimum of one door in each airlock is closed.

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed by isolation valve, blind flange, or manual valve except for valves that are open on an intermittent basis under administrative control, or

2. Be capable of being closed by an OPERABLE automatic containment isolation valve, or

3. Be capable of being closed by an OPERABLE containment vacuum relief valve.

Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During movement of recently irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of recently irradiated fuel in the containment. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment penetrations shall be determined to be either in its closedlisolated condition or capable of being closed by an OPERABLE automatic containment isolation valve within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and at least once per 7 days during movement of recently irradiated fuel in the containment by:

a. Verifying the penetrations are in their closedfisolated condition, or

b. Testing of containment isolation valves per the applicable portions of Specifications 4.6.3.1.1. and 4.6.3.1.2.

ST. LUCIE - UNIT 1 314 9-4 Amendment No. 47, 454, 47,184

REFUELING OPERATIONS COMMUNI CATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY: During CORE ALTERATIONS.

ACTION:

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.

The provisions of Specification 3.0.3 are not applicable. I SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

ST. LUCIE - UNIT 1 3/4 9-5 Amendment No. 102

REFUELING OPERATIONS MANIPULATOR CRANE OPERABILMTY

[,IMrTING CONDMON FOR OPFRATlON 3.9.6 The manipulator crane shall be used for movement Of CEAs or fuel assemblies and shall be OPERABLE with:

a. A minimum capacity of 2000 pounds, and

b. An overload cut off limit of s 3000 pounds.

APPLICABILIY: During movement of CEAs orfuel assembries whin th reactor pressure vessel.

ACTTQN:

With the requirements for crane OPERABILITY not sat sfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel. j SURVFILt ANCE RFOUIRFMENTS 4.9.6 The manipulator crane used for movement of CEAs or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of such operations by performing a load test of at least 2500 pounds and demonstrating an automatic load cut off before the crane load exceeds 3000 pounds.

ST. LLUE - UNIT 1 Y8"6 ST.

WCIE-UNIT I 3'49.6 No.~~~~~~mwaennet 152-

REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING I

LIMITING CONDITION FOR OPERATION e 3.9.7 Loads in excess of 2000 pounds shall be prohibited from travel-over irradiated fuel assemblies in the storage pool.

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

With the requirements of the above specification.not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable.

SURVEILLAIICE REQUIREMENTS Z-4.9.7 Crane int'erlocks and physical stops which prevent crane travel with loads in excess of 2000 pounds over fuel assemblies shall be -

demonstrated OPERABLE within-7 days prior to crane use and at least-once per 7 days thereafter during crane operation.

ST. LUCIE - UNIT I 3/4 9-7

REFUELING OPERATIONS SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL LIMING CONDITION FOR OPERATION 3.9.8.1 At least one shutdown cooling loop shall be OPERABLE and in operation*.

APPLICABILITY: MODE 6 when the water level above the top of irradiated fuel assemblies seated within the reactor pressure vessel is greater than or equal to 23 feet.

ACTION:

a. With less than one shutdown cooling loop in operation, suspend all operations involving an increase in reactor decay heat load or operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one shutdown cooling loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The shutdown cooling loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of reactor pressure vessel hot legs, provided no operations are permitted that would cause introduction into the RCS, coolant with boron concentration less than required to meet the SHUTDOWN MARGIN of Technical Specification 3.9.1.

ST. LUCIE - UNIT 1 314 9-8 Amendment No. 66.179

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION_

3.9.8.2 Two independent shutdown cooling loops shall be OPERABLE and at least one shutdown cooling loop shall be in operation.

APPLICABILITY: MODE 6 when the water level above the top of irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet.

ACTION:

a. With less than the required shutdown cooling loops OPERABLE, within one (1) hour 1) initiate corrective action to return the required loops to OPERABLE status, or 2) establish greater than or equal to 23 feet of water above irradiated fuel assemblies seated within the reactor pressure vessel.

b. With no shutdown cooling loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of Technical Specification 3.9.1. and within one (1) hour initiate corrective action to return the required shutdown cooling loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one shutdown cooling loop shall be verified to be in opera-tion and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT I 314 98a Amendment No. 66, 403179

REFUELING OPERATIONS CONTAINMENT ISOLATION SYSTEM LIMrITNG CONDITION FOR OPERATION 3.9.9 The containment isolation system shall be OPERABLE.

APPLICABILITY: During movement of recently irradiated fuel assemblies within containment.

ACTION:

With the containment isolation system inoperable, close each of the penetrations providing direct access from the containment atmosphere to the outside atmosphere.

SURVEILLANCE REQUIREMENTS 4.9.9 The containment isolation system shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and at least once per 7 days during movement of recently irradiated fuel assemblies by verifying that containment isolation occurs on manual initiation and on a high radiation signal from two of the containment radiation monitoring instrumentation channels.

ST. LUCIE - UNrr 1 314 9-9 Amendment No. 40,403,448, 184

REFUELING OPERATIONS WATER LEVEL- REACTOR VESSEL UMING CONDIhION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel.

APPLICABILITY: During CORE ALTERATIONS.

During movement of irradiated fuel assemblies within containment ACTION:

With the requirements of the above specifications not satisfied, immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment, and immediately initiate action to restore refueling cavity water level to within limits.

SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during CORE ALTERATIONS and during movement of irradiated fuel assemblies within containment.

ST. LUCIE - UNIT I 3/4 9-10 Amenftent No. 148

REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 As a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. --

ACTION:

With the requirement of the specification not satisfied, 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 . The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQ'JIREMENTS 4.9.11 The water level in the 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.

ST. LUCIE - UNIT 1 3/4 9-11

REFUELING OPERATIONS FUEL POOL VENTILATION SYSTEM - FUEL STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 At least one fuel pool ventilation system shall be OPERABLE.

APPLICABILITY: Whenever recently irradiated fuel is in the spent fuel pool.

ACTION:

a. With no fuel pool ventilation system OPERABLE, suspend all operations involving movement of recently irradiated fuel within the spent fuel pool or crane operation with loads over the recently irradiated spent fuel until at least one fuel pool Ventilation system is restored to OPERABLE status.

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

SURVEILLANCE REQUIREMENTS 4.9.12 The above required fuel pool ventilation system shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by.

ST. LUCIE - UNIT I 314 9-12 Amendment No. 184

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

1. Verifying that the charcoal adsorbers remove > 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfm + 10%.

2. Verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfmr 10%.

3. Verifying that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of > 85% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1 989 (300C, 95% RH). The carbon samples not obtained from test canisters shall be prepared by either a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

4. Verifying a system flow rate of 10,350 cfmn + 10% during system operation when tested in accordance with ANSI N510-1975.

ST. LUCIE - UNIT 1 3/4 -13 Amendment No. -167

a --

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

c. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4.15 inches Water Gauge while operating the ventilation system at a flow rate of 10,350 cfm + 10%. - -

2. Verifying that the air flow distribution is uniform within 20% across HEPA filters and charcoal adsorbers when tested in accordance with ANSI N510-1975.

3. Verifying that the ventilation system maintains the spent fuel storage pool area at a negative pressure of > 1/8 inches Water Gauge relative to the outside atmosphere during system operation.

d. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99X of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfm + 10%.

e. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 99% of a halogenated hydrocarbon refrigerant test gas when They are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 10,350 cfm + 10%.

ST. LUCIE - UNIT 1 3/4 9-14

REFUELING OPERATIONS SPENT FUEL CASK CRANE LIMITING CONDITION FOR OPERATION 3..3Temxmmla hc a ehnldb h pn ulcs rn 3.9.13 The maximum load which may be handled by the spent fuel cask crane shall not exceed 25 tons.

APPLICABILITY: Whenever irradiated fuel assemblies arerin the storage pool .

ACTION:

With the requirements of the above specification not satisfied, place load in a safe condition. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.13 The loaded weight of a spent fuel assembly cask shall be verified to not exceed 25 tons prior to attaching it to the spent fuel cask crane.

ST. LUCIE - UNIT 1 3/4 9-15

IEFUELING OPERAT IONS 3/4.9.14 DECAY TIME - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.14 The irradiated fuel assemblies in the fuel storage pool shall have decayed for at least 1180 hours0.0137 days 0.328 hours 0.00195 weeks 4.4899e-4 months , unless more than one-third core is placed into the pool, in which case the irradiated fuel assemblies shall have ecayed for 1490 hours0.0172 days 0.414 hours 0.00246 weeks 5.66945e-4 months .

APPLICABILITY: Prior to movement of the spent fuel cask into the fuel cask compartment.

ACTION:

With irradiated fuel assemblies having a decay time of less than 1180 hours0.0137 days 0.328 hours 0.00195 weeks 4.4899e-4 months .

r 1490 hours0.0172 days 0.414 hours 0.00246 weeks 5.66945e-4 months in the case of more than one-third core discharge, suspend all activities involving movement of the spent fuel cask into the fuel cask compartment. .-The provisions of Specification .3;0.3 are- not applicable.

SURVEILLANCE REQUIREMENTS

.1 4.9.14. The irradiated fuel assemblies in the fuel storage pool shall have been determined to have decayed for at least 1180 hours0.0137 days 0.328 hours 0.00195 weeks 4.4899e-4 months , or 1490 hours0.0172 days 0.414 hours 0.00246 weeks 5.66945e-4 months in the case of more than one-third core discharge, by verification of the date and time from the most recent subcriticality prior to movement of the spent fuel cask into the fuel cask compartment.

ST LUCIE - UNIT I 3/4 9-16 Amendment No. Z-, 40

i3/4.10 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:

a. 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 > 40 gpm of 1720 ppm boric acid solution or its equivalent untiT 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 > 40 gpm, of 1720 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLA t VERE U~fi<E.I -:^e C 4 > fft kt' 4.10.1.1 The position of each full length CEA required either partially or fully withdrawn shall be detenmined 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 I 3/4 10-1 Amendment No. i, 87

rs SPECIAL TEST EXCEPTIONS GROUP HEIGh+T. INSERTION AND POWER DISTRIBUTION LIMITS

[IjMm0Nf, CQUOM0N FOR OPER&11ON 3.10.2 The group height, insertion and power distribution rimits of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5,3.1.3.6,332.3 and 32.4 may be suspended during the performance of PHYSICS TESTS provided:

I

a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and

b. The limits of Specification 32.1 are maintained and determined as specified in Specification 4.1022 below.

APPLICABILITY: MODES 1 and 2.

ACTION:

With any of the limits of Specification 32.1 being exceeded while te requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3 and 3.2.4 are suspended, either. I

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 32.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVFIL LANCE RFOUIRFMFIJNTS 4.102.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Spedfications 3.1.1.4,3.1.3.1, 3.1.3.5,3.1.3.6,3.2.3, or 32.A are suspended and shall be verified to be within the test power plateau.

4.1022 The linear heat rate shall be determined to be within the Emits of Specification 32.1 by monitoring it continuously with the Incore Detector Monitoring Systfm pursuant to the requirements of Specifications 41.4A during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1 A.3.1 £.1, 3.1.3.5, 3.1.3.6, 32.3. or 32.A are suspended.

ST. LUCIE -UNTIi 3g4 10.2 Affwenmerzt NO.B, 46G, 44f,'

4S6, 152

SPECIAL TEST EXCEPTIONS PRESSURE/TEMPERATURE LIMITATION - REACTOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.3 This specification deleted.

SURVEILLANCE REQUIREMENTS 4.10.3 This specification deleted.

II I

ST. LUCIE - UNIT 1 3/4 10-3 Amendment No. 4

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.4 This specification deleted.

SURVEILLANCE REQUIREMENTS 4.10.4 This specification deleted.

ST. LUCIE - UNIT I 314 10-4 Amendment No. 2 7

SPECIAL TEST EXCEPTIONS CENTER CEA MISALIGNMENT LIMMNG COND1ON FOR OPERATION 3.10.5 The requirements of Specifications 3.1.3.1 and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS to determine the isothermal temperature coefficient and power coefficient provided:

a. Only the center CEA (CEA #1) is misaligned, and

b. The limits of Specification 32.1 are maintained and determined as specified in Specification 4.10.5.2 below.

APPLICABILITY: MODES I and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.3.1 and 3.1.3.6 are suspended, either.

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.5.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended and shall be verified to be within the test power plateau.

4.10.5.2 The linear heat rate shall be determined to be within the limits of Specification 32.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specification 4.2.1.4 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended.

ST. LUCIE - UNIT 1 3/4 10-5 Amendment No. 1 36

It Pages 3/4 11-2 through 3/4 11-13 (Amendment No. 123) have been deleted from the Technical Specifications. The next 3/4 11-14.

page is I

ST. LUCIE - UNIT 1 3/4 11.1 AMENDMENT w011 134

RAIACTIVE EWERUE EXPLOSIVE GAS MIXTURE I IMMING CONDITON FOR OPFRA-T1ON -

3.11.2.5 The concentration of oxygen in the waste gas decay tanks shal be limited to less than or equal to 2%h by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABIL!X: At all times.

ACTION:

a. With the concentration of oxygen in the waste gas decay tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. With the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2%h by volume, immediately suspend all additions of waste gases to the system and immediately commence reduction of the concentration of oxygen to less than or equal to 2% by volume.

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

StJRVFII I ANCF RFOUIREMENTS 4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall be determined to be within 4.112.5.2 the above limits by continuously' monitoring the waste gases in the on service waste gas decay tank.

With the oxygen concentration in the on service waste gas decay tank greater than 2% by I

volume as determined by Specification 4.11.2.5.1, the concentration of hydrogen in the waste gas decay tank shagl be determined to be within the above limits by gas partitioner sample at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

When continuous monitoring capability`is inoperable, waste gases shall be monitored in accordance with the actions specified for the Waste Gas Decay Tanks Explosive Gas Monitoring System in Chapter 13 of the Updated Final Safety Analysis Report. I ST. LUCIE -UNIT 1 3!4 11-14 Amendment No. 69, 446,4i, 156

RADZOACT'ITE ErFLUENTS 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 285,000 curies noble gases (considered as Xe-133)..

APPLICABILITY: At all times.

ACTI ON:

a. With the quantity of radioactive material in any gEr s 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 REOUIREMENTS 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 systemi activity exceeds 100 .

ST. LUCIE - UNIT I 3/4 11-15 Amendment No. V9'69

SECTION 5.0 DESIGN FEATURES

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area is shown on Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone is shown on Figure 5.1-1.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The containment structure is comprised of a steel containment vessel, having the shape of a right circular cylinder with a hemispherical dome and ellipisoidal bottom, surrounded by a reinforced concrete shield building. The radius of the shield building Is at least 4 feet greater than the radius of circular cylinder portion of the containment vessel at any point.

5.2.1.1. CONTAINMENT VESSEL

a. Nominal inside diameter = 140 feet.

b. Nominal inside height = 232 feet.

c. Net free volume = 2.5 x 106 cubic feet.

d. Nominal thickness of vessel walls = 2 inches.

e. Nominal thickness of vessel dome = 1 inch.

f. Nominal thickness of vessel bottom = 2 inches.

II ST. LUCIE - UNIT I 5-1 Amendrnent No. a, 142

%4j EXCLUSION AREA (0.17 me)

AND LOW POPULATION ZONE (I mm -

NOTES:

1) L-Liquid Radwaste Release Point

2) Due to the scale of the Figure the Exclusion Area Radius (0.97 mile) and the 4k Low Population Zone (1 mile) are shown as being the same FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT 0 1/2 I SCALE IN MILES SITE AREA MAP FIGURE 5.1-1 ST. LUCIE - UNIT 1 5-2 Amendment No. j5g,125

FIGURE 5.1-2 (Deleted)

ST. LUCIE - UNIT 1 5-3 Amendment No. 5 9 fI

.DPESKAN FFATURFSq----

2.12 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 wal minimum thickness = 3 feet

f. Dome minimum thickness = 2.5 feet

g. Dome inside radius - 112 feet DESIGN PRESSURE AND TEMPERATURE 522 The containment vessel is designed and shal be maintained for a maxdmum 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 3.82-1.10 and 62.4 of the FSAR with allowance for noimal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maxdmum of 176 fuel rods dad with Zircaloy-4. Each fuel rod shahave a nominal active fuel length of between 134.1 and 136.7 incWs. Individual fuel assemblies shall contain fuel rods of the same nominal acte fuel length. Fuel assemblies shall be limited to those designs that have been analyzed using NRC approved methodology and shown by tests or analyses to comply with fuel design and safety criteria. The Initial core loading shall have a maximum enrchment of 2.83 weight percent U-235. Reload fuel shall be similar in physical design to the initial core bading.

53.1.1 Except for special test as authrized by the NRC, all fuel asserblies under control element assemblies shall be sleeved with a sleee design previously approved by the NRC.

ST. LLWCE-UNIT1 54 A-rrt N 04i44, 76 4"'152

DESIGN FEATURES CONTROL ELEMENT ASSEMBLIES 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 Requirements.

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 52 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 6500F, except for the pressurizer which is 700TF VOLUME 5.42 The total water and steam volume of the reactor coolant system is 11,100 + 180 cubic feet at a nominal T., of 5670F, when not accounting for steam generator tube plugging.

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

1. k., less than or equal to 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.

ST. LUCIE - UNIT 1 5-5 Amendegnt No. X, ff. ts, e+,

DESIGN FEATURES CRITICALITY (Continued)

2. A nominal 10.12 inches center to center distance between fuel assemblies in Region 1 of the storage racks and a nominal 8.86 inches center to center distance between fuel assemblies in Region 2 of the storage racks.

3. A boron concentration greater than or equal to 1720 ppm.

4. Neutron absorber (boraflex) installed between spent fuel assemblies in the storage racks in Region 1 and Region 2.

b. Region 1 of the spent fuel storage racks can be used to store fuel which has a'J-235 enrichment less than or equal to 4.5 weight percent. Region 2 can be used to store fuel which has. achieved sufficient burnup such that storage in Region 1 is not required. The initial enrichment vs. burnup requirements of Figure 5.6-1 shall be met prior to storage of fuel assemblies in Region 2. Freshly discharged fuel assemblies may be moved temporarily into Region 2 for purposes of fuel assembly inspection and/or repair, provided that the configuration is maintained in a checkerboard pattern (i.e., fuel assemblies and empty locations aligned diagonally). Following such inspection/repair activities, all such fuel assemblies shall be removed from Region 2 and the requirements-of Figure 5.6-1 shall be met for fuel storage.

c. The new fuel storage racks are designed for dry storage of unirradiated fuel assemblies having a U-235 enrichment less than or equal to 4.5 weight percent, while maintaining a k ef of less than or equal to l 0.98 under the most reactive condition.

DRAINAGE 5.6.2 The fuel 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 poo- is. designed and shall be maintained with a storage capacity limited to no more than 1706 fuel assemblies.

5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as seismic Class I in Section 3.2.1 of the FSAR shall be designed and maintained to the original design provisions contained in Section 3.7 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirement.

ST,. LUCIE - UNIT 1 5-6 Amendment No. 77,ZZ.ZJ*.

7F, ML, 92

nere~f-a IrrAnn~rCr UEZIUM MtAIUKLZ 5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.

5.9 COMPONENT CYCLE OR TRANSIENT LIMITS 5.9.1 The components identified in Table 5.9-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.9-1.

ST. LUCIE - UNIT 1 5-6a Amendment No. 91

ACCEPTAB LE E 25 BURNUP DOMAIN z

w 20 52 UNACCEPTABLE 30' 5 15 BURNUP DOMAIN 10 1 La0 L5 3.0 35 4.0 4.5 INITIAL ENRICHMENT. WT % U-2A FIGURE 5.6-1 INITIAL ENRICKMENT VS SURNUP REBUIRENENTS FOR STORAGE OF FUEL ASSEMBLIES IN REGION 2.

ST. LUCIE PLANT UNIT I 5-6b Amendnent No. 91

(

TABLE 5.9-1 (n

I COMPONENT CYCLIC OR TRANSIENT LIMITS C,

n COMPONENT CYCLIC OR TRANSIENT LIMITS DESIGN CYCLE OR TRANSIENT

-I Reactor Coolant System 40 Cycles of loss of load without 100% to 0% RATED THERMAL

-J immediate reactor trip POWER 40 cycles of loss of offsite 100% to 0%RATED THERMAL A.C. electrical power POWER

. 400 reactor trips 100% to 0% RATED THERMAL POWER 16 inadvertent auxiliary Spray line 650 0F to 1200F spray cycles in 1.5 seconds 1: 200 leak tests Pressure > 2235 psig 10 hydrostatic pressure tests Pressure > 3110 psig Secondary System 5 steam line breaks Complete loss of secondary pressure 200 leak tests Pressure > 985 psig 10 hydrostatic pressure tests Pressure > 1235 psig

SECTION 6.0 ADMINISTRATIVE CONTROLS

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant General Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor, or during his absence from the control room a designated individual, shall be responsible for the control room command function.

A management directive to this effect, signed by the Site Vice President, shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION ONSITE AND OFFSITE ORGANIZATION 6.2.1 An onsite and an offsite organization shall be established for unit operation and corporate management. This onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organizational charts.

These organizational charts will be documented in the Topical Quality Assurance Report and updated in accordance with 10 CFR 50.54(a)(3).

b. The Chief Nuclear Officer shall be responsible for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

c. The Plant General Manager shall be responsible for overall safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant.

d. Although the individuals who train the operating staff and those who carry out the quality assurance functions may report to the appropriate manager onsite, they shall have sufficient organizational freedom to be independent from operating pressures.

e. Although health physics individuals may report to any appropriate manager onsite, for matters relating to radiological health and safety of employees and the public, the Health Physics Supervisor shall have direct access to that onsite individual having responsibility for overall unit management. Health physics personnel shall have the authority to cease any work activity when worker safety is jeopardized or in the event of unnecessary personnel radiation exposures.

ST. LUCIE - UNIT 1 6-1 Amendment No. 25, 69, Q9, 44", 426, 4. 178

6.0 ADMINISTRATIVE CONTROLS 6.2 ORGANIZATION (continued)

UNIT STAFF 6.2.2 The unit organization shall be subject to the following:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the reactor is in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator shall be in the control room.

c. A health physics technician# shall be on site when fuel is in the reactor.

d. Either a licensed SRO or licensed SRO limited to fuel handling who has no concurrent responsibilities during this operation shall be present during fuel handling and shall directly supervise all CORE ALTERATIONS.

e. Deleted. I

The health physics technician may be less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

ST. LUCIE - UNIT 1 62 Amendment No. 40. 25, 5C, 69,03,44,446,. 7

=^

DELETED ST. LUCIE - UNIT 1 6-3 Amendment No. 93

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSmON TWO UNITS WITH TWO SEPARATE CONTROL ROOMS WITH UNIT 2 IN MODE 5 OR 6 OR DEFUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3 or 4 MODE 5 or 6 a

SS (SRO) 1 1.a SRO 1 None RO 2 1 AO 2 2 STA* I None WITH UNIT 2 IN MODE 1, 2,3, or 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION

_____ _ MODE 1, 2, 3 or 4 MODE 5 or 6 SS (SRO) la 1a SRO 1 None RO 2 1 AO 2 1 STA* 1 None SS - Shift Supervisor with a Senior Reactor Operator's License on Unit 1 SRO - Individual with a Senior Reactor Operator's License on Unit 1 STA - Shift Technical Advisor RO - Individual with a Reactor Operator's License on Unit 1 AO - Auxiliary Operator Except for the Shift Supervisor, the Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composi-tion to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SRO license shall be designated to assume the Control Room command function. During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 5 or 6, an individual with a valid SRO or RO license shall be designated to assume the Control Room command function.

a/ Individual may fill the same position on Unit 2.

b/ One of the two required individuals may fill the same position on Unit 2.

cl If STA position is filled by an STA qualified Shift Supervisor or dedicated STA, then the individual may fill the same position on Unit 2.

A single, onsite STA position shall be manned in Mode 1, 2, 3, and 4 unless the Shift Supervisor meets the qualifications for the STA as required by Technical Specification 6.3.1 or an individual on each unit with a Senior Reactor Operator's license meets the qualifications for the STA as required by Technical Specification 6.3.1.

ST. LUCIE - UNIT 1 6-4 Amendment No. 34. 6S,6aC,474,i7V

6.0 -ADMINISTRATIVE CONIROLS

f. Administrative procedures shall be developed and implemented to limit the working hours of personnel who perform safety-related functions (e.g., licensed senior reactor operators (SROs), licensed reactor operators (ROs), health physicists, auxiliary operators, and key maintenance personnel). The administrative procedures shall include guidelines on working hours that ensure that adequate shift coverage shall be maintained without routine heavy use of overtime.

Any deviation from the above guidelines shall be authorized by the Plant General Manager or the Plant General Manager's designee, in accordance with approved administrative procedures, and with documentation of the basis for granting the deviation. Controls shall be included in the procedures to require a periodic independent review be conducted to ensure that excessive hours have not been assigned. Routine deviation from the working hour guidelines shall not be authorized.

g. The Operations Supervisor shall hold a Senior Reactor Operator license.

SHIFT TECHNICAL ADVISOR 6.2.3 The Shift Technical Advisor function is to provide on shift advisory technical support in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit ST. LUCIE - UNIT I 65 Amendment No. SQ 93,4C.4.

422.4i, 185

61- ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI / ANS-3.1-1978 for comparable positions, except for (1) the Health Physics Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have specific training in plant design and plant operating characteristics, including transients and accidents, and any of the following educational requirements:

- Bachelor's degree in engineering from an accredited institution; or

- Professional Engineer's (PE) license obtained by successful completion of the PE examination; or

- Bachelor's degree in engineering technology from an accredited institution, including course work in the physical, mathematical, or engineering sciences, or

- Bachelor's degree in physical science from an accredited institution, induding course work in the physical, mathematical, or engineering sciences.

(3) the Multi-Discipline Supervisors who shall meet or exceed the following requirements:

a. Education: Minimum of a high school diploma or equivalent

b. Experience: Minimum of four years of related technical experience, which shall include three years power plant experience of which one year is at a nuclear power plant.

c. Training: Complete the Multi-Discipline Supervisor training program.

6.4 TRAINING 6A.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI I ANS-3.1 - 1978 and 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure I of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

6.5 REVIEW AND AUDIT 6.5.1 FACILITY REVIEW GROUP (FRG)

FUNCTION 6.5.1.1 The Facility Review Group shall function to advise the Plant General Manager on all matters related to nuclear safety.

COMPOSION 6.5.1.2 The FRG shall have voting members composed of individuals from each of the following disciplines:

Operations Electrical Maintenance Reactor Engineering Mechanical Maintenance Health Physics Technical Support Chemistry Quality Assurance I Control Uicensing Services Instrument and Control The Plant General Manager shall appoint the FRG members, in writing, and from this membership shall designate, in writing, a FRG Chairman.

Members shall meet or exceed the qualifications required for Managers; Supervisors, or Professional-Technical, as appropriate, pursuant to Specification 6.3.1.

ST. LUCIE - UNIT 1 6-6 Amendment No. 25, 3S, e,173

4. 4.4,

6.0 ADMINISTRATIVE CONTROLS ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the FRG Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in FRG activities at any one time.

MEETING FREQUENCY 6.5.1.4 The FRG shall meet at least once per calendar month and as convened by the FRG Chairman or his designated alternate.

QUORUM 6.5.1.5 The quorum of the FRG necessary for the performance of the FRG responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.

RESPONSIBILMES 6.5.1.6 The Facility Review Group shall be responsible for.

a. Review of (1) all new procedures required by Specification 6.8 and all procedure changes that require a written 50.59 evaluation, (2) all programs required by Specification 6.8 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Plant General Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed changes to Appendix A Technical Specifications.

d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chief Nuclear Officer and to the Chairman of the Company Nuclear Review Board.

f. Review of all REPORTABLE EVENTS.

g. Review of unit operations to detect potential nuclear safety hazards.

h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant General Manager or the Company Nuclear Review Board.

ST. LUCIE - UNIT 1 6-7 Amendment No. aQ, a9, 83, 40.426,458.4,. 178

6.0 ADMINISTRATIVE CONTROLS

i. Not Used.

j. Not Used.

k. Review of every unplanned on-site release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Chief Nuclear Officer and to the Company Nuclear Review Board.

I. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL and RADWASTE TREATMENT SYSTEMS.

m. Review and documentation of judgment concerning prolonged operation in bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last FRG meeting.

n. Review of the Fire Protection Program and implementing procedures and submittal of recommended changes to the Company Nuclear Review Board.

AUTHORITY 6.5.1.7 The Facility Review Group shall:

a. Recommend in writing to the Plant General Manager, approval or disapproval of items considered under Specifications 6.5.1.6.a through d above.

b. Render determinations in writing with regard to whether or not each item considered under Specifications 6.5.1.6.a, b, d, and e above requires NRC approval pursuant to 10 CFR 50.59.

c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Chief Nuclear Officer and the Company Nuclear Review Board of disagreement between the FRG and the Plant General Manager, however, the Plant General Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.

RECORDS 6.5.1.8 The Facility Review Group shall maintain written minutes of each FRG meeting that, at a minimum, document the results of all FRG activities performed under the responsibility and authority provisions of these Technical Specifications. Copies shall be provided to the Plant General Manager, Chief Nuclear Officer and the Chairman of the Company Nuclear Review Board.

ST. LUCIE - UNIT 1 6-8 Amendment No. , X2, 99, 49;, 442,448,46C,44,468.,

4z 178

ADMINISTRATIVE CONTROLS 6.5.2 COMPANY NUCLEAR REVIEW BOARD (CNRB)

FUNCTION 6.5.2.1 The Company Nuclear Review Board shall function to provide indepen-dent review and audit of designated activities in the areas of

a. nuclear power plant operations

b. nuclear engineering

c. chemistry and radiochemistry

d. metallurgy

e. instrument and control

f. radiological safety

g. mechanical and electrical engineering

h. quality assurance practices COMPOSMON 6.5.22 The Chief Nuclear Officer shall appoint, in writing, a minimum of five members to the CNRB and shall designate from this membership, in writing, a Chairman. The membership shall function to provide independent review and audit in the areas listed in Specification 6.5.2.1. The Chairman shall meet the requirements of ANSIIANS-3.1-1987, Section 4.7.1.

The members of the CNRB shall meet the educational requirements of the ANSI/ANS-3.1-1987, Section 4.7.2, and have at least 5 years of professional level experience in one or more of the fields listed in Specification 6.5.2.1. CNRB members who do not possess the educational requirements of ANSI/ANS-3.1-1987, Section 4.7.2 (up to a maximum of 2 members) shall be evaluated, and have their membership approved and documented, in writing, on a case-by-base basis by the Chief Nuclear Officer, considering the alternatives to educational requirements of ANSIIANS-3.1-1987, Sections 4.1.1 and 4.1.2.

ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the CNRB Chair-man to serve on temporary basis; however, no more than two alternates shall participate as voting members in CNRB activities at any one time.

ST. LUCIE - UNIT 1 6-9 Amendment No. 48. 65,6 9. 88,9.

4. 407. 178

ADMINISTRATIVE CONTROLS CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.

MEETING FREQUENCY 6.5.2.5 The CNRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter and as convened by the CNRB Chairman or his designated altemate.

QUORUM 6.5.2.5 The quorum of the CNRB necessary for the performance of the CNRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least a majority of CNRB members including altemates. No more than a minority of the quorum shall have line responsibility for operation of the facility.

REVIEW 6.5.2.7 The CNRB shall review:

a. The evaluations for (1) changes to procedures, equipment, or systems and (2)tests or experiments completed under the provisions of Section 50.59, 10 CFR, to verify that such actions did not require NRC approval pursuant to 10 CFR 50.59.

b. Proposed changes to procedures, equipment, or systems which require NRC approval pursuant to 10 CFR 50.59.

c. Proposed tests or experiments which require NRC approval pursuant to 10 CFR 50.59.

d. Proposed changes to Technical Specifications or this Operating License.

e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety.

ST. LUCIE - UNIT I 6-10 Amendment No. 4 ,66.6. 402. 177

ADMINISTRATIVE CONTROLS

g. All REPORTABLE EVENTS.

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.

i. Reports and meeting minutes of the Facility Review Group.

AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the CNRB.

These audits shall encompass:

a. The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions.

b. The performance, training and qualifications of the entire unit staff.

c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or method of operation that affect nuclear safety.

d. The performance of activities required by the Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50.

e. Any other area of unit operation considered appropriate by the CNRB or the Chief Nuclear Officer.

f. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA personnel.

g. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant An outside independent fire protection consultant shall be used at least every third year.

h. The radiological environmental monitoring program and the results thereof.

i. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures.

j. The PROCESS CONTROL PROGRAM and implementing procedures for dewatering of radioactive bead resin.

ST. LUCIE - UNIT 1 6-11 Amendment No. 25,69.69.

107.4 178

ADMINISTRATiVE CONTROLS AUTHORITY 6.5.2.9 The CNRB shall report to and advise the Chief Nuclear Officer on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of CNRB activities shall be prepared, approved and distrib-uted as indicated below:

a. Minutes of each CNRB meeting shall be prepared, approved and forwarded to the Chief Nuclear Officer within 14 days following each meeting.

b. Reports of reviews encompassed by Specification 6.5.2.7 above, shall be prepared, approved and forwarded to the Chief Nuclear Officer within 14 days following completion of the review.

c. Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the Chief Nuclear Officer and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the FRG, and the resuits of the review shall be submitted to the CNRB, and the Chief Nuclear Officer.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Chief Nuclear Officer and the CNRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the FRG. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

ST. LUCIE - UNIT I 6-12 Amendment No. aQ, 03, 402. A, 178

6.0 ADMINISTRAliVE CONTROLS

c. The Safety Umit Violation Report shall be submitted to the Commission, the CNRB, and the Chief Nuclear Officer within 14 days of the violation.

d. Critical operation of the unit shall not be resumed until authorized by the Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below.

a. The applicable procedures recommended in Appendix A" of Regulatory Guide 1.33, Revision 2, February 1978, and those required for implementing the requirements of NUREG 0737.

b. Refueling operations.

c. Surveillance and test activities of safety-related equipment.

d. Not Used.

e. Not Used.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Control Program for effluent monitoring, using the guidance in Regulatory Guide 1.21, Revision 1, June 1974.

j. Quality Control Program for environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975.

6.8.2 REVIEW AND APPROVAL OF PROCEDURES Each new procedure of Specification 6.8.1a. through i. above shall be independently reviewed by an individual or group from the appropriate discipline(s), and shall be reviewed by the FRG. New procedures shall be approved by the Plant General Manger or individuals designated in writing by the Plant General Manager prior to implementation. Each procedure of Specification 6.8.1 shall be reviewed periodically as set forth in administrative procedures.

ST. LUCIE - UNIT 1 6-13 Amendment No. 25, a%.6., .3, 4, 42,44,48, 1.78

6.0 ADMINISTRATIVE CONTROLS 6.8.3 CHANGES TO PROCEDURES

a. Each revision to the procedures of Specification 6.8.1a. through i. above shall be independently reviewed by an individual or group from the appropriate discipline(s), and revisions that require a written evaluation pursuant to 10 CFR 50.59 shall be reviewed by the FRG. Procedure revisions shall be approved by the Plant General Manag- or individuals designated in writing by-the Plant General Manager prior to implementation.

b. Temporary changes to procedures of Specification 6.8.1a. through i. above may be made provided:

1. The intent of the original procedure is not altered.

2. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

3. The change is documented and, if appropriate, incorporated in the next revision of the affected procedure pursuant to Specification 6.8.3.a.

6.8.4 The following programs shall be established, implemented, maintained, and shall be audited under the cognizance of the CNRB:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the Shutdown Cooling System, High Pressure Safety Injection System, Containment Spray System, and RCS Sampling. The program shall include the following:

(i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radioiodine Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

(i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment ST. LUCIE - UNIT 1 6-14 Amendment No. 6a, 4X 444, 4N, 177

6.0 ADMINISTRATIVE CONTROLS

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

(i) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, ST. LUCIE - UNIT I 6-14a Amendment No. 1 58

ADMINISTRATIVE CONTROLS (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all off-control point chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Backup Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

(i) Training of personnel, and (ii) Procedures for monitoring.

e. DELETED

f. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reason-ably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall in-clude remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1) Umitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to ten times the concentration values in 10 CFR 20.1001 -

20.2401, Appendix B, Table 2, Column 2.

ST. LUCIE - UNIT 1 6-15 Amendment No. 68, 86, 43, 426, 174

ADMINISTRATIVE CONTROLS _

3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commit-ment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appro-priate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas at or beyond the SITE BOUNDARY shall be limited to the following:

a) For noble gases: Less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin, and b) For Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ;

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part SD,

9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides-in particulate form with half-lives greater 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 due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

g. Radiological Environmental Monitoring Procram A program shall be provided to monitor the radiation and radio-nuclides 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 environmental exposure pathways. The program shall (1)be contained in the ODCM, ST. LUCIE - UNIT 1 6-15a Amendment No. 69, 86,4&G 125

ADMINISTRATIVE 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 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 Testinq 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-1 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 later than May 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident Pa, is 39.6 psig. The containment design pressure is 44 psig.

The maximum allowed containment leakage rate, L8 , at Pa, shall be 0.50%h 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 < 6.0 L8 for the Type B and C tests,

< 0.75 La for Type A tests, and c 0.27 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. 69, 8W, 4ma,449 187

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

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

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

a. Testing frequencies specified in Section Xi of the ASME Boiler and Pressure Vessel Code*

and applicable addenda as follows:

ASME Boiler and Pressure Vessel Code*

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

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

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities; and

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

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

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

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

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

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

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

4. Proposed changes that meet the criteria of Specification 6.8Aj.2.a or 6.8Aj.2.b, above.

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

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

ST. LUCIE - UNIT 1 6-15c Amendment No. 44k, 443.176

ADMINISTRATIVE CONTROLS Econtinueca 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the NRC.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment of the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant ST. LUCIE - UNIT I 6-15d Amendment No. 176

ADMINISTRATIVE CONTROLS 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a compari-son of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS 16.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be subnitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) for whom monitoring was required receiving Annual Deep Dose Equivalent exposures greater than 100 mrem/yr and their associated man-rem l exposure according to work and job functionsJ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total Deep Dose Equivalent received from external sources should be assigned to specific major work functions.

Annual reports shall also include the results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be intluded:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first saMple in which the limit was exceeded; (2) Results of the last 1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

2/ This tabulation supplements the requirements of 20.2206 of 10 CFR Part 20.

ST. LUCIE - UNIT I 6-16 Amendment No. ZAp, A iff, 125 f,

ADMINISTRATIVE CONTROLS isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while the limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the NRC, no later than the 15th of each month following the calendar month covered by the report.

ST. LUCIE - UNIT 1 6-16a Amendment No. 101

ADMINISTRATIVE CONTROLS ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT*

6.9.1.7 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous 12 months of operation shall be submitted within 60 days after January 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) con-sistent with the objectives outlined in the O0CM and PCP and (2) in confor-mance with 10 CFR 50.36a and Section IW.B.1 of Appendix I to 10 CFR Part 50.

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

ST. LUCIE - UNIT I 6-17 Amendment No. ES , 69, 1 2 3

ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT**

6.9.1.8 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May I of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sec-tions IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

- A single submittal may be made for a multiple unit station.

ST. LUCIE - UNIT 1 6-18 Amendment No. A9, 69,1 23

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 - Tang Greater Than 2000 F Specification 3.1.1.2 Shutdown Margin - Ta., Less Than or Equal to 2000 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 Specification 3.2.3 Total Integrated Radial Peaking Factor - FT Specification 3.2.5 DNB Parameters r 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-1 1596-P-A, "Qualification of the PHOENIX-PIANC 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. XN-75-27(A) and Supplements 1 through 5, [also issued as XN-NF-75-27(A)], "Exxon Nuclear Neutronic(s) Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Inc. I Advanced Nuclear Fuels Corporation, Report and Supplement 1 dated April 1977, Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P)

4. ANF-84-73(P)(A) Revision 5, Appendix B, & Supplements 1 and 2, "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors:

Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, October 1990

5. XN-NF-82-21(P)(A) Revision 1, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Inc., September 1983

6. a) ANF-84-93(P)(A) and Supplement 1, [also issued as XN-NF-84-93(P)(A)], "Steamline Break Methodology for PWRs,"

Advanced Nuclear Fuels Corporation, March 1989 ST. LUCIE - UNIT I 6-19 Amendment No. 9, aQ, 6, 123, 4w,46a. 171

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

6. b) EMF-84-093(P)(A) Revision 1, "Steam Line Break Methodology for PWRs," Siemens Power Corporation, February 1999 (This document is a Revisior to ANF-84-93)

7. XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, Inc., October 1983.

8. Siemens Power Corporation Small Break LOCA methodology as defined by:

a) XN-NF-82-49(P)(A) Revision 1, 'Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model," Advanced Nuclear Fuels Corporation, April 1989 b) XN-NF-82-49(P)(A) Revision 1 Supplement-1, "Exxon Nuclear Company Evaluation Mode: Revised EXEM PWR Small Break Model," Siemens Power Corporation, December 1994

9. XN-NF-78-44(NP)(A), nA Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, Inc., October 1983

10. XN-NF-621 (P)(A) Revision 1," Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, Inc., September 1983

11. EXEM PWR Large Break LOCA Evaluation Model as defined by.

a) 1. XN-NF-82-20(P)(A) Revision 1 Supplement 2, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, Inc., February 1985

2. XN-NF-82-20(P)(A) Revision 1 and Supplements 1, 3 and 4, "Exxon Nuclear Company Evaluation Model EXEMI/PWR ECCS Model Updates," Advanced Nuclear Fuels Corporation, January 1990.

b) XN-NF-82-07(P)(A) Revision 1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, " Exxon Nuclear Company, Inc., November 1982 c) 1. XN-NF-81 -58(P)(A) Revision 2, and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Inc., March 1984

2. ANF-81-58(P)(A) Revision 2 Supplement 3, and Supplement 4, ORODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," Advanced Nuclear Fuels Corporation, June 1990 ST. LUCIE - UNIT 1 6-1 ga Amendment No. 469, 163

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

11. d) XN-NF-85-16(P)(A) Volume 1, and Supplements 1, 2 and 3; Volume 2, Revision 1 and Supplement 1, NPWR 17x17 Fuel Cooling Test Program," Advanced Nuclear Fuels Corporation, February 1990 e) XN-NF-85-105(P)(A) and Supplement 1, "Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs,"

Advanced Nuclear Fuels Corporation, January 1990.

f) EMF-2087(P)(A) Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation, June 1999.

12. XN-NF-82-06(P)(A) Revision 1, and Supplements 2, 4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Bumup," Exxon Nuclear Company, Inc., October 1986

13. ANF188-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Bumups of 62 GWd/MTU,"

Advanced Nuclear Fuels Corporation, December 1991

14. XN-NF-85-92 (P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, Inc., November 1986

15. ANF-89-151 (P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, May 1992

16. XN-NF-507(P)(A), Supplements 1 and 2, "ENC Setpoint Methodology for C. E. Reactors: Statistical Setpoint Methodology," Exxon Nuclear Company, Inc., September 1986

17. EMF-92-116(P)(A), Revision 0, "Generic Mechanial Design Criteria for PWR Fuel Design," Siemens Power Corporation, February 1999.

18. EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation, March 1994.

19. EMF-96-029(P)(A) Volumes I and 2, "Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2- Benchmarking Results," Siemens Power Corporation, January 1997.

20. EMF-1 961 (P), Revision 0, "Statistical SetpointiTransient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation, December 1998.

ST. LUCIE - UNIT I St19b Amendment No. 463, 1 71

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

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 to the NRC.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.

ST. LUCIE - UNIT I 6-19C Amendment No. 171

ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of unit operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c. All REPORTABLE EVENTS.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of changes made to the procedures required by Specification 6.8.1.

f. Records of radioactive shipments.

g. Records of sealed source and fission detector leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the unit Operating License:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of reactor tests and experiments.

d. Records of radiation exposure for all individuals entering radia-tion control areas.

e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those unit components identified in Table 5.9-1.

ST. LUCIE - UNIT 1 6-20 Amendment No. fA,/go, 69

ADMINISTRATIVE CONTROLS

g. Records of training and qualification for current members of the unit staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the FRG and the CNRB.

1. Records of the service lives of all safety related snubbers including the date at which the service life commences and associated installation and maintenance records.

m. Records of secondary water sampling and water quality.

n. Annual Radiological Environmental Operating Reports and records of analyses transmitted to the licensee which are used to prepare the Annual Radiological Environmental Monitoring Report.

O. Meteorological data, summarized and reported in a format consistent with the recommendation of Regulatory Guides 1.21 and 1.23.

p. Records of audits performed under the requirements of Specifications 6.5.2.8 and 6.8.4.

q. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

ST. LUCIE - UNIT 1 6-21 Amendment No. 44,59,r697 ,123

ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or Oalarm signal* required by paragraph 20.1601(a) of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr measured at a distance of 30 cm (12 in) shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels in excess of 1000 mrem/hr at 30 cm (12 in) and less than 500 rads/hr at 1 meter shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels in excess of 1000 mrem/hr at 30 cm (12 in) and less than 500 rads/hr at 1 meter that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification ofthe RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may

'be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise fol-lowing plant radiation protection procedures for entry into high radiation areas.

ST. LUCIE - UNIT 1 6-22 Amendment No. / $;,

125

ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2q. This documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change will maintain the overall conformance of the solidified waste product to existing re-quirements of Federal, State, or other applicable regula-tions.

2. Shall become effective after review and acceptance by the Facility Review Group and the approval of the Plant General Manager.

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2q. This documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall become effective after review and acceptance by the Facility Review Group and the approval of the Plant General Manager.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., monthtyear) the change was implemented.

ST. LUCIE - UNIT 1 6-23 Amendment No. 69. 6,986, 43, 42 1

DELETED ST. LUCIE - UNIT I 6-24 AMENDMENT NO. gs,123

APPENDIX B - PART I DELETED ST. LUCIE - UNIT I Amendment No. 50, 67

ML032731661

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