ML17219A494

Proposed Tech Specs,Removing Outdated Matl,Making Minor Text Changes & Correcting Typos

ML17219A494
Person / Time
Site:	Saint Lucie
Issue date:	03/31/1987
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17219A493	List: ML17219A492 ML17219A494
References
NUDOCS 8704070028
Download: ML17219A494 (46)

Text

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN-T GREATER THAN 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to APPLICABILITY:

MODES 1, 2", 3 and 4.ACTION: oooo pc~continue boration at greater than or equal to 40 gpm of a solution con-taining greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal a.Within one hour after detection of an inoperable CEA(s)and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s)is inoperable.

If the inoperable CEA is immovable or untrippable,;he above required SHUTDOWN MARGIN shall be verified acceptable with an increased allawance for the withdrawn worth of, the immovable or untrippable CEA(s).When in MODE 1 or MODE 2 with K f greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.When in MODE 2 with K less than 1.0, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor critI/ality by verifyi ng that the predicted critical CEA position is within the limits of Specification 3.1.3.6.See Special Test Exception 3.10.1.870407a028 8703331 PDR ADOCN 05000389 P PDR ST.LUCIE-UNIT 2 3/4 l-l REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued d.Prior to initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors of e.below, with the CEA groups at the Po~er Dependent Insertion Limits of Specification 3.1~3.6.e.When in MODE 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of at least the following factors: l.2.3..4 5.6.Reactor coolant system boron concentration, CEA position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

fOoo p~m 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within+at least once per 31 Effective Full Power-Days (EFPD).This comparison shal consider at least those factors stated in Specification 4.1.l.1.le., above.The predicted reactivity values shall be adjusted (normalized) to, correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPDs after each fuel loading.ST.LUCIE-UNIT 2 3/4 1"2 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN-T LESS THAN OR E UAL TO 200 F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to APPLICABILITY:

MODE 5.ACTION: With the SHUTDOWN MARGIN less than continue boration at greater than greater than or equal to 1720 ppm SHUTDOWN MARGIN is restored.immediately initiate and or equal to 40 gpm of a solution containing boron or equivalent until the required 3OQQ~tel SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal a.Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable CEA(s)and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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'he withdrawn worth of the'immovable or untrippable CEA(s).b.At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by'consideration of the following factors: 1.2.3.4 5.6.Reactor coolant system boron concentration, CEA position, Reactor coolant system average temperature,.Fuel burnup based on gross thermal energy generation,)(enon concentration,and Samarium concentration.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, when the Reactor Coolant System is drained below the hot leg centerline, by consideration of the factors in 4.l.1.2b.and by verifying at least two charging pumps are rendered inoperable by racking out their motor circuit breakers.ST.LUCIE" UNIT 2 3/4 1-3 Amendment No.8

'REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC)shall be: joF a.Less positive than-4 at<70K RATED THERMAL P POWER, q 3~~bF b.Less pos i ti ve than POWER, and$7pssn/op'.

Less negative than POWER.APPLICABILITY:

MODES 1 and 2"¹ACTION: "4 at>7DX RATED THERMAL at RATED THERMAL With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.SURVEILLANCF.

RE UIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its l.imits by confirmatory measurements.

MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle: as b.C.Prior to initial operation above 5X of RATED THERMAL POWER, after each fuel loading.At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 800 ppm..At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.ith K ff greater than or equal to 1.0.¹See Special Test Exceptions 3.10.2 and 3.10.5.ST.LUCIE" UNIT 2 3/4 1-5 Amendment No.14 REACTIVITY CONTROL SYSTEMS FLOW PATHS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three b6ron injection flow paths and one associated heat tracing circuit shall be OPERABLE: a~b.Two flow paths from.the boric acid makeup tanks via either a boric acid makeup pump or a gravity feed connection, and a charging pump to the.Reactor Coolant System, and The flow path from the refueling water tank via a charging pun to the Reactor Coolant System.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION: With only one of'he 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 21122 Ittt I*I;t1 t 11 paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.oooo pew SURVEILLANCE RE UIREMENTS 4.1.2.2 At.least two of the above required flow paths shall be demonstrated OPERABLE: a.At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid makeup tanks is above the temperature limit line shown on Figure 3.1-1.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.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'an SIAS test signal.d.At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a delivers at least 40 gpm to the Reactor'oolant System.ST.LUCIE-'UNIT 2 3/4 1-8 Amendment No.8 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall.be OPERABLE.APPLICABILITY:

MODES l, 2, 3 and 4.ACTION: 3OOO~With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a HH II H 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />;restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURYEILLANCE RE UIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by.-verifying that each pump develops a flow rate of greater than or equal to 40 gpm when tested pursuant to Specification 4.0.5.4.1.2.4.2 At least once per 18 months verify that each charging pump starts automatically on an SIAS test signal.ST.LUCIE" UNIT 2 3/4 1" 10 Amendment No.8 REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS'-OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid makeup pump(s)in the boron injection flow path(s)required OPERABLE pursuant to Specification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump(s)in Specification 3.1.2.2a is OPERABLE.APPLICABILITY:

MODES 1, 2, 3 and 4." ACTION: Mith one boric acid makeup pump required for the boron injection flow path(s)pursuant to Specification 3.1, 2.2a inoperable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOMN MARGIN: equivalent to at least I~~+>>at 200 F;restore the above required boric acid makeup pump(s)OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.to next'066 Pc+SURVEILLANCE RE UIREMENTS 4.1.2.6 The above required boric acid makeup pump(s)shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump(s)develop a discharge pressure of greater than or equal to 90 psig when tested pursuant to Specification 4.0.5.ST.LUCIE" UNIT 2 3/4 1"12 Amendment No.

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES-'PERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE: a.At least one boric acid makeup tank and at least one associated heat tracing circuit per tank with the contents of the tank in accordance with Figure 3.1-1, and b.The refueling water tank with: l.A minimum contained borated water volume of 417, 100 gallons, 2.A boron concentration of between 1720 and 2100 ppm of boron, and 3.A solution temperature between 55 F and 100 F.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION:+400 Pcwl'.b.With the above required boric acid makeup tank inoperable, restore the tank to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN 01 1 tt.t1~200'0;0 0 required boric acid makeup tank to OPERABLE status within the next 7 days or be in COLC SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.1.2.8 a.b.Each borated water source shall be demonstrated OPERABLE: At least once per 7 days by: l.Verifying the boron concentration in the water, 2.Verifying the contained borated water volume of the water source, and 3.Verifying the boric acid makeup, tank solution temperature.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWT temperature when the outside air temperature is outside the range of 55 F and 100'F.ST.LUCIE-'UNIT 2 3/4 1-14 Amendment No.8 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION INSTRUMENT 1.AREA HONITORS HINIMUM CHANNELS OPERABLE APPLICABLE MODES ALARM/TRI P SETPOINT MEASUREMENT RANGE ACTION a.Fuel Storage Pool Area i.Criticality and Yentilation System Isolation Monitor<20 mR/hr 10-10 mR/hr 22 b.Containment Isolation c.Control Room Isolation 1 per ALL MODES intake<90 mR/hr 1-10 mR/hr 7 25-7-2<2x background 10-10 pCi/cc 26 d.Containment Area-Hi Rang""*1, 2, 3&4 Not Appl i cable 1-10 R/hr 7 27 2.PROCESS MONITORS a.Fuel Storage Pool Area Yentilation System i.Gaseous Activity ii.Particulate Activity b.Containment Gaseous Activity RCS Leakage Detection Particulate Activity RCS Leakage Detection 1, 2, 3&4 Not Applicable 1, 2, 3&4 Not Applicable 10-10 pCi/cc 23 1-10 cpm 6 23 10-10 pCi/cc 24'-10 cpm 6 24 Mith fuel in the storage pool or building.""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."*"The Alarm/Trip Setpoints are determined and set in accordance with the requirements of Specification 3.3.3.10.oo 8 TABLE 3.3-6 (Continued)

RADIATION MONITORING INSTRUHENTATION I n M P1 M INSTRUMENT PROCESS HONITORS (Continued) c.Noble Gas Effluent Monitors MINIM UH CHANNELS APPLICABLE OPERABLE HODES ALARM/TRIP SETPOINT HEASUREHENT RANGE ACTION Reactor Auxiliary Building Exhaust System (Plant Vent Low Range Monitor)1, 2, 3,&4 10-10 pCi/cc 27 iv.1, 2, 3,&4 1, 2, 3,&4 1/steam 1, 2, 3,&4 header Reactor Auxiliary Bui'lding Exhaust Sys-tem (Plant Vent igh Range Monitor""" 1 Steam Generator Blowdown Treatment Facility Building Exhaust System 1 Steam Safety Valve Discharge¹ 10-10 pCi/cc 27-2 5 10-10 pCi/cc 27-1 3 10-10 pCi/cc 27 v.Atmospheric Steam Dump Valve Dis-charge vi.ECCS Exhaust/steam 1, 2, 3,&4 header 1, 2, 3,&4 10-10 pCi/cc 27 10-10 pCi/cc 27-7 5 The Alarm/Trip Setpoints are determined and set in accordance with the requirements of Specification 3.3.3.10.¹The steam safety valve discharge monitor and the atmospheric steam dump valve discharge monitor are the same monitor.

TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS'nstrumentatson ocated]n St.Lucre Unst 1 INSTRUMENTS AND SENSOR LOCATIONS 1.STRONG MOTION TRIAXIAL ACCELEROGRAPHS a.SMR-42-1 CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK CALIBRATION TEST SA b.SMR-42"2 c.SMR-42-3 d.SMR-42"4 e.SMR-42-5 2.PEAK RECORDING ACCELEROGRAPHS a.SMR-42-6 b.SMR-42"7 c.SMR-42-8 3.PEAK SHOCK RECORDERS a.SMR-42-9 b.SMR-42-10 4.EARTHQUAKE FORCE MONITOR a.SMI-42-11 5.SEISMIC SWITCH a.SMS-42"12 uA NA NA NA NA R'R SA SA SA SA NA~NA~RA SA SA Except seismic trigger ST.LUCIE" UNIT 2 3/4 3-34 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION INSTRUMENT READOUT LOCATION CHANNELS RANGE REQUIRED OF NUMBER CHANNELS MINIMUM CHANNELS OPERABLE'.

2.3.4.5.6.7.8.9.10.11.12.13.Power Range Neutron Flux Reactor Trip Breaker Indication Reactor Coolant Temperature

-Cold T Pressurizer Pressure Pressurizer level Steam Generator Pressure Steam Generator Level Shutdown Cooling Flow Rate Shutdown Cooling Temperature Diesel Generator Voltage Diesel Generator Power Atmospheric Dump Valve Pressure Charging Flow/Pressure Hot Shutdown Hot Shutdown Hot Shutdown Hot Shutdown Hot Shutdown Panel Panel Panel Panel Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Hot Shutdown Panel Reactor Trip Switch Gear{RB)'x 10-8X-200K OPEN-CLOSE 0-600 F 0-3000 psia 0-100X level 0-1200 psia 0-100K le'vel 0-5000 gpm 0-350 F 0-5250 V 0-5000 kw 0-l200 psig 0-150 gpm/0-3000 psia 1/trip breaker 1/tr ip breaker~2 steam generator 2/steam generator 1 1 1 l/steam generator 1/steam generator l/steam generator 2 1/steam generator 2 1 1/diesel generator 1/diesel generator 1/diesel generator 1/diesel generator CONTROLS/ISOLATE SWITCHES 2.3.5.6.7.Atmospheric Stm Dump Controllers Aux.Spray Valves Charging Pump Controls Letdown Isol Valve.AFM Pump/Valve Controls AFM Pump Steam Inlet Valve Pzr Heater Controls Hot Shutdown Panel/RAB431 Hot Shutdown Panel/RAB431 Hot Shutdown Panel/RAB431 Hot Shutdown Panel/RAB431 Hot Shutdown Panel/RAB431 Hot Shutdown Panel/RAB431 Hot Shutdown Panel/RAB431 H.A.N.A.N.A.N.A.H.A.N.A.H.A.2/steam generator 1/steam generator 1 2 2 2 INST RUMEH TAT I OH RADIOACTIVE GASEOUS EFFLUEHT MONITORING IHSTRUMEHTATION LIMITIHG CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded.The alarm/trip setpoints of these channels shall be determined in accordance with the methodology and parameters in the ODCM.APPLICABILITY:

As shown in Table 3.3-13 ACTION: a.b.C.With a radioactive gaseous effluent monitoring inst'rumentation channel alarm/trip setpoint less conservative than required by the above Specification, immediately suspend the release of radio-active gaseous effluents monitored by the affected channel or declare the channel inoperable.

• With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3"13.If the inoperable instruments are not returned to operable status within 30 days, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, SURVEILLANCE RE UIREMENTS 4.3.3 1 Each radioactive gaseous effluent monitoring instrumentation channel shall e demonstrated OPERABLE by per formance of the CHANNEL CHECK, SOURCE CHEC CHANNEL CALIBRATION and.CHANNEL FUNCTIONAL TEST operations at the fre encies shown in Table 4.3-9.ST.LUCIE-.UNIT 2 3/4 3"53 Amendment No.I3 TABLE 4.3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS ,I n m INSTRUMENT 1.WASTE GAS DECAY, TANKS a.Noble Gas Activity Monitor-Providing Alarm and Automatic Termination of Release 2.WASTE GAS DECAY TANKS EXPLOSIVE GAS MONITORING SYSTEM CHANNEL CHECK SOURCE CHANNEL CHECK CALIBRATION

.R(3)CHANNEL MODES IN WHICH FUNCTIONAL SURVEILLANCE T EttRE Q(l)a.Oxygen Monitor b.Oxygen Monitor (alternate) 3.CONDENSER EVACUATION SYSTEM a.Noble Gas Activity Monitor N.A.N.A.Q(P)QV R(3)Q(2)4 PLANT VENT SYSTEM a.Noble Gas Activity Monitor b.Iodine Sampler c.Particulate Sampler d.Sampler Flow Rate Monitor N.A.N.A.N.A.R(3)N.A.N.A.Q(2)N.A.N.A.N.A.

I n fA TYPE OF MEASUREMENT AND ANALYSIS TABLE 4.4-4 PRIHARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM SAMPLE AND.ANALYSIS FRE UENCY MODES IN WHICH SAHPLE AND ANALYSIS RE UIRED 2.3.'ross Activity Determinati on Isotopic Analysis for DOSE EQUIVALENT I-131 Concentration Radiochemical for E Determination 4.Isotopi c Analysis for Iodi ne Including I-131, I-133, and I-135 At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1 per 14 days 1 per 6 months~a}Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds/gram, DOSE E UIVALENT I-131 or 100/E/gram, and b}One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding'15K of the RATED THERMAl POWER within a 1-hour period.1, 2, 3, 4 18, 2P, 3P, 4P, 50'2 3~Until the specific activity of the, primary coolant system is restored within its limits.~" Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.-

REACTOR COOLANT SYSTEM 3/4.4.10 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.10 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 sp'ace, and b.Reactor vessel head.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION: a 0 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.b.With 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.4.10.1 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.I 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 2 3/4 4"38 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued) 2.At least once per 18 months.Valve Number LPSI S stem a.VC 3616/3617 a.MCV 3615 b.HVC 3626/3627'.

HCV 3625 c.HVC 3636/3637 c.HCV 3635 d.V 3646/3647 d.'CV 3645 e.523/V3540 By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics.

The test shall measure the individual leg flow rates and.pump total developed head to verify the following conditions:

1.HPSI Pump 2A: The sum of the three lowest cold leg flow rates shall be greater than or equal to 476 gpm with'total developed head greater than or equal to 1150 ft but less than or equal to 1290 ft.2.HPSI Pump 2B: The sum of the three lowest cold leg flow rates shall be greater than or equal to 484 gpm with total developed head greater than or equal to 910 ft but less than or equal to 1040 ft.3.With the system operating in hot/cold leg injection mode, the hot leg flow shall be greater than or equal to 317 gpm and within 10K of the cold.leg header flow and: HPSI Pump 2A: E-The pump shall be producing total developed head greater than or equal to 1297 ft but.less than or equal to 1500 ft, HPSI Pump 2B: The pump shall be producing total developed head greater than or equal to 1042 ft but less than 1250 ft.4.LPSI System-Each Pump: The flow through each injection leg shall be greater than or equal to 1763 gpm at a total developed head greater than or equal to 298 ft but less'than'or equal to 337 ft.+(6 ST.LUCIE." UNIT 2 3/4 5-6 1 TABLE 3.6-1 CONTAINMENT LEAKAGE PATHS Penetration 7 8~Sstem Makeup Water Station Air~-V-IS.I u~o~Valve Ta Number T e I-HCV-15-1 Globe I-V-15-328 Check I-V-18-794 Globe I-V-18-797 Globe I-HCV-18-2 Glob**Location to Containment Service Test T e*Outside Inside Outside Inside Annulus Outside Station Air Supply Primary Makeup Water BYPASS/TYPE C BYPASS/~TYPE C Instrument Air I-HCV-18-1 Globe I-V-18-195 Check Outside Inside Instrument Air Supply BYPASS/.TYPE C 10 I ll 14 23 A)24 O 25 26 Containment Purge Containment Purge Maste Management Component Cooling Component Cooliiig I-FCV-25-5 B'FLY I-FCV-25-4 B'FLY I-FCV-25-2 B'FLY I-FCV-25-3 B'FLY V-6741 Globe V-6792 Check I-HCV-14-7 B'FLY I'-HCV-14-'1 B'FLY I-HCV-14-6 B'FLY I-HCV-14-2 B'FLY CVCS I-V-2516 Globe I-V-2522 Globe f'uel Transfer Tobe Double Gasket Flange Annulus Inside Annulus Inside Outside Inside Outside Inside Outside Inside Inside Inside Outside Containment Purge Exhaust Containment Purge Supply N2'upply to Safety Inj.Tanks RC Pump Cooling Water Supply RC Pump Cooling Mater Return Fuel Transfer Letdown Line TYPE C TYPE C BYPASS/TYPE C BYPASS/TYPE C BYPASS/TYPE C BYPASS/TYPE C BYPASS/TYPE C CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS COHTAINMEHT SPRAY SYSTEM LIMITIHG CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWT on a Containment Spray Actuation Signal and automatically transferring suction to the containment sump on a Sump Recirculation Actuation Signal.Each spray system flow, path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger.

APPLICABILITY:

MODES 1, 2, and 3".ACTION: With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />;restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.6.2.1 a 0 b.C.Each containment spray system shall be demonstrated OPERABLE: 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 positioned to take suction from the RWT on a Containment Pressure-High-High test signal.By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 200 psig when tested pursuant to Specification 4.0.5.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 upon a Recirculation Actuation Test Signal (RAS), the containment sump isol tion valves open and that a recircula-tion mode flow path via o OPERABLE shutdown cooling heat exchanger's establishe

.4t.h 3/4 6"15 Applicable only when pressurizer pressure is>1750 psia.C~CA'T.LUCIE" UNIT 2 Valve Ta Number A)Containment Isolation I-HCV-15-1 I Hcd-I8 I-HCV-18-1 I-FCV-25-5,4 I-FCV-25-2,3 V-6741 I-HCV-14-7 I-HCV-14-1 I-HCV-14-6 I-HCV-14-2 I-HCV-2516 I-HCV-2522 I-SE-05-1A,18, 1C,10,1E I-V-5200 I-V-5203 I-V-5204 I-V-5201 I-V-5205 I-V-5202 Penetration Number 7 8 9 10 23 24 26 28A 28B 29A 29B TABLE 3.6-2 CONTAINMENT ISOLATION VALVES Function Primary Makeup Water (CIS)&ale~Ai~$~~4 Instrument Air Supply (CIS)Containment Purge Exhaust (CIS)Containment Purge Makeup (CIS)Nitrogen Supply to Safety Injection Tanks (CIS)Reactor Coolant Pump Cooling Water Supply (SIAS)Reactor Coolant Pump Cooling Water Return (SIAS)Letdown Line (CIS)Safety Injection Tank Sample Reactor Coolant System Hot Leg Sample (CIS)Pressurizer Surge Sample (CIS)Pressurizer Steam Sample (CIS)Testable Ouring Plant 0 eration Yes Y'es" No No No Yes No No No Yes Yes Yes Yes Maximum Isolation~Time Sec 5 5 S~

O foal~V)I N~B)Hanual CR Remote Hanual Penetration Number TABLE 3.6-2 (Continued)

CONTAINHENT ISOLATION VALVES Function Haxiaua Testable Ouring Isolation P~li~Yi 5 I-V-18-R4 Pi'7 8 I-V-18-m~y,7o Station Air Supply (Hanual)p~k>Air~ugly (Clec.kp Yes i'-V-3463 41 Safety Injection Tank Test Line (Hanual)Yes I-V-07-206 I-V-07-189 I-V-07-170 I-V-07-188 I-FSE-27-8,9,10, 48 11,15,16 I-FSE-27-12,13,14, 51 D,18 46 47 Fuel Pool Cleanup (Inlet)(Hanual)Fuel Pool Cleanup (Outlet)(Hanual)Hq Saapling (Remote Hanual)Hq Sampling (Reaote Hanual)Yes Yes Yes Yes O CD I-V-00-140 I-V-00"143 I-V-00-139 I-V-00-144 I-V-00-101 520 5)E 54 ILRT (Hanual)ILRT (Hanual)ILRT (Hanual)Yes Yes Yes~~-become I-V l8-1279-upo

',P' PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION b.One feedwater pump capab supply system.APPLICABILITY:

MODES 1, 2, and 3.ACTION'.7.

1.2 At least three independent steam gyp ator auxiliary feedwater pumps'nd associated flow paths shall be OPERABLE~with: 4 a.Two feedwater pumps, each capable of being powered from separate OPERABLE emergency busses, and le of being powered from an OPERABLE steam a~With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within.the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.c.With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status.SURVEILLANCE RE UIREMENTS 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 greater than or equal to l270 psig on recirculation f 1 ow.2.Verifying that the turbine-driven pump develops a discharge pressure of greater than or equal to 1260 psig on recirculation flow when the secondary steam supply pressure is greater than 50 psig.The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.3.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.ST.LUCIE-UNIT 2 3/4 7-4 PLANT SYSTEMS 3/4.7.12 FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.12 Al 1 f ire r ated assemb1 i es (wal 1 s, floor/ceilings, and other fi r e barriers)separating safety-related fire areas or separating portions o'redundant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors, fire dam ers, cable, piping, and ventilation duct penetration seals)shall be OPERABLE.~

APPLICABILITY:

At al 1 times.ACTION: With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either establish a continuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of the fire detectors on at least one side of the inoperable assembly and establish.

an hourly fire watch patrol.b.The provisions of Specitications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.12.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by: a.Performing a visual inspection of the exposed surfaces of each fire rated assembly.b.Performing a-visual inspection of each fire damper and-associated hardware.Performing a visual inspection of at least 10K of each type of sealed penetration.

If apparent changes in appearance or abnormal degrada-tions are found, a visual inspection of an additional 10K of each type of sealed penetration shall be made.This inspection process shall continue until a 10K sample with no apparent changes in appear-ance or abnormal degradation is found.Samples shall be selected such that each penetration seal will'-be inspected every 15 years.ST.LUCIE-UNIT 2 3/4 7-39 3/4.8 ELECTRICAL'OWER SYSTEMS 3/4.8.3.A.C.SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1's a minimum, the'following A.C..electrical, power sources shall be OPERABLE: a.b.Two physically independent circuits between the offsite transmission network and the onsite Class lE distribution system,'nd Two separate and independent diesel generators, each with: 1.Two separate engine-mounted fuel tanks containing a minimum volume of 200 gallons of fuel each, 2.'separate fuel s orage system containing a minimum volume of 40,000 gallons of fuel, and 3.A separate fuel transfer pump.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION: a~b.C.With either an offsite circuit other than the conditions delineated in Action 3.8.1.1f.or diesel generator of the above required A.C.electrical power sources in'operable, demonstrate the OPERABILITY of the remaining A.C.sources by performing Surveillance Requirements 4.8.1.l.la.and 4.8;1.1.2a.4.within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.With one offsite circuit and one diesel generator of the above required A.C.electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.sources by per'forming Surveillance Requirements 4.8.1.l.la.and 4.8.1.1.2a.4.within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from.the time of'initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.With one diesel generator inoperable in addition to ACTION a.or b.above, verify that: l.All required systems, subsystems, trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and 2.When a MODE 1, 2, or 3, the steam-driven auxiliary feed pump is OPER ST.LUCIE" UNIT 2 ELECTRICAL POWER SYSTEMS AGTION,(Continued)

If these conditions are not satisfied within 2 hour~be in at least HOT STANDBY.within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.d.With two of the above required offsite A.C.circuits inoperable, demonstrate the OPERABILITY.of two diesel generators by performing Surveillance Requirement 4.8.1.l.2a.4.

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the fol lowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.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,l.l.la.

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore at least one of the inoperable

'diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.Restore at least two diesel generators to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.With one Unit 2 startup transformer (2A or 2B)inoperable and with a Unit 1 startup transformer (lA or 1B)connected to the same A or B offsite power circuit and administratively available to both units, then should Unit 1 req'uire the use of the startup transformer administratively available to both units, Unit 2 shall demonstrate the operability of the remaining A.C.'sources by performing Surveil-lance Requirements 4.8.1.l.la.and 4.8.l.1.2a.4.within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter; restore the inoperable startup transformer to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.8.l.l.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 al,ignments, indicated power availablity, and b.Demonstrated OPERABLE at least once per 18 months by transferring (manually and automatically) unit power supply from the normal circuit to the alterhate circuit.ST.LUCIE-, UNIT 2 3/4 8"2 REFUELING OPERATIONS 3/4.9.5 COMMUNI CAT IONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall e maintained between the control room and personnel at the refueling station." APPLICABILITY:

Dur ing CORE ALTERATIONS.

ACTION: When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend'all CORE ALTERATIONS.

SURVEILLANCE RE UIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.

ST.LUCIE-, UNIT 2 3/4 9-5 RADIOACTIVE EFFLUEHTS LI UID WASTE TREATMENT LIMTTIHG CONDITION FOR OPERATION 3.11.1.3 The liquid radyaste treatment system Shall be OPERABLE.'he appro-priate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site to UNRESTRICTED AREAS (see Figure 5.1-1}when averaged over.31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.APPLICABILITY:

At all tiwes.ACT1ON: 'th the liquid radwaste t eatment system inoperable for more than days with radioactiv

'waste being discharged without reatment and,in excess of the above limits, prepare and submit to the Coaxnission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:

l.Identification of the inoperable equipment or subsystems and the reason for inoperability, 2.hction(s}taken to restore the inoperable equipment to OPERABLE status, and 3.Summary description of action(s)taken to prevent a recurrence.

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

SURVEILLANCE RE UIREMEHTS 4.11.1.3.1 Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with the OOCM unless the liquid radwaste treatment system is being used.4.11.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 30 minutes at least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days., ST.LUCIE-UNIT 2 3/4 11-6 Amendment No.

TABLE 4.11-2 Continued)

TABLE NOTATION b.Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15K of RATED THERMAL POWER within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> unless (1)analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3;and (2)the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.Samples shal l be changed at least 4 times a month and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler).Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, star tup or THERMAL POWER change exceeding 15K of RATED THERMAL POWER in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing if (1)analysis shows that the DOSE E(VIVALENT I-131 concentration in the primary coolant has increased more than a factor of 3;and (2)the noble gas activity monitor shows that effluent activity has increased by more than a factor 3.When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LL be increased by a factor of 10.The ratio of the sample flow rate to e sampled stream flow rate shall be-known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.e.The principal gamma emitters for which the LLD specification applies exclusively're the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

This list does not mean that only these nuclides are to be detected and reported.Other.peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported.ST.LUCIE*-UNIT 2 3/4 11-10 BASES 4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an.OPERATIONAL MODE or other applicable condition.

The intent of this provision is to ensure that surveil-lance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.

Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components and i nservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a.Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications.

This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda.This clarification is provided to ensure consistency in surveillance intervals t ughout these Technical Specifications and to remove any ambiguities elative to the frequencies for performing the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda.For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation.-

And for example, the Technical Specification defi nition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared inoperable.

ST.LUCIE'-UNIT 2 B 3/4 0"3 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN oooo pew A sufficient SHUTDOWN MARGIN ensures that 1)the reactor can be made subcritical from all'operating conditions, 2)the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3)the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.'HUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T.The most restrictive avg'ondition occurs at EOL, with T at no load operating temperature, and is avg associated with a postulated steam line break accident.and resulting uncon-trolled RCS cooldown.In the analysis.of this accident, a minimum SHUTDOWN MARGIN of is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requiremeht is based u'pon this limiting condition and is consistent with FSAR safety analysis'ssumptions.

At earlier times in core life, the minimum SHUTDOWN MARGIN'equired for the most restric-tive conditions is less than With T less than or equal to 200'F av the reactivity transients resulting from any postulated accident are minimal and a.3%delta.k/k SHUTDOWN MARGIN provides.adequate.protection.

3/4.1~1.3 BORON DILUTIOH A minimum flow rate of at least 3000 gpm provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System.A flow rate of at least 3000 gpm will circulate an equivalent Reactor Coolant System volume of 10,931 cubic feet in approximately 26 minutes.The reactivity change rate associated with boron concentration reductions will therefore be within the capability of operator recognition and control.3/4.l.l.4 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator, temperature coefficient (MTC)are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel'ycle.

The surveillance requirements for measurement of the MTC during'each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup..The confirmation that the measured MTC,value is within its limit provides assurances that the coef-ficient will be maintained within acceptable values throughout~

each fuel cycle.ST.LUCIE-UNIT 2 8 3/4 1"1 Amendment Ho.8 0 REACTIVITY CONTROL SYSTEMS BASES 300 b pe~3/4.l.1.5 HINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor wi 11 not be made critical with the Reactor Coolant System average temperature less than 515'F.This limitation is required to ensure (1)the moderator temperature coefficient is within its analyzed temperature range, (2)the protective instrumentation is within its normal operating range, (3)the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4)the reactor pressure vessel is*above its minimum RTNDT temperature.

3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation.

The components required to'perform this function include (1)borated water sources, (2)charging pumps, (3)separate flow paths, (4)boric acid makeup pumps, (5)associated heat tracing systems, and (6)an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in%he event an assumed failure renders one of'the systems inoperablte.

Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repai r period.The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions o after xenon decay and cooldown to 200'F.The, maximum expec e ora>on capa>sty requirement occurs at EOL from full power equi librium xenon conditions and requires boric acid solution from the boric acid makeup tanks in the allowable concentrations and volumes of Specification 3.1.2.8 or 72,000 gallons of 1720 ppm-2100 ppm borated water from the refueling water tank.With the RCS temperature below 200'F one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The boron capability required below 200 F is based upon providing a SIIUTODflN NARGIN f d y d ld f 200'F 1 0 This condition requires either 4,150 gallons of 1720 ppm-2100 ppm borated water from the refueling water tank or boric acid solution from the boric acid makeup tanks in accordance with the requirements of Specification 3.1.2.7.The conta ined water, volume 1 imi ts includes allowance for water not available because of discharge line location and other physical characteristics.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.The limits on contained water volume and boron concentration of the RWT also ensure a pH value of between 7.0 and 11.0 for the solution recirculated within containment after a LOCA.This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

WT I lit'TF-IIMTT B 3/4 1-2 Amendment No.'8 REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued)

Overpower margin is provided to protect the core in the event.of a large misalignment

(>15 inches)of a CEA.However, this misalignment would cause disto'rtion of the core power distribution.

This distribution may, in turn, have a significant effect on (1)the available SHUTDOWN MARGIN, (2)the time-dependent long-term power distributions relative to those used in generating LCOs and LSSS setpoints, and (3)the ejected CEA worth used in the safety'nalysis.Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt realignment of the misaligned CEA.The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA.Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LCO:and LSSS setpoints.

However, extended operation with CEAs significantly inserted in the core may lead to perturbations in (1)local burnup, (2)peaking factors, and (3)available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.

Therefore, ttme limits have been imposed on operation with inoperable CEAs to preclu'de such adverse conditions from developing.

The r quirement to reduce power in certain time limits depending uoon the previous Fis to eliminate a potential nonconservatism for situations when a CEA has been declared inoperable.

A worst-case analysis has shown that a DNBR SAFDL violation.may occur during the second hour after the CEA misalignment if this requirement is not met.This'potential DNBR SAFDL violation is eliminated by limiting the time operation'is permitted at full power before power reductions are required.These reductions will be necessary once the deviated CEA has been declared inoperable.

This time allowed continued operation at a reduced power level can be permitted for the folio ing asons: 1.The margin calculations which support the Technical Speci@cat>ons are based on a steady-state radial peak of F=1.7 0.1 2.Mhen the actual F<1.70, significant additional margin exists.T 3.This additional margin can be credited to offset the increase in Fwith time that can occur following a CEA misalignment.

4.This increase in F is caused by xenon.redistribution.

r 5.The present analysis can support allowing a misalignment to exist for-up to 63 minutes without correction if the initial Fr (1.54 R 0/4.1-4 3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: (1)the CEA insertion limits of Specifications 3.l.3.5 and 3.1.3.6 are satisfied (2)the'lux peaking augmentation factors are as shown in Figure 4.2-1, (3]the AZIMUTHAL POHER TILT restrictions of Specifica-tion 3.2.4 are satisfied, and (4)the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2.The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1.The setpoints for these alarms include allowances, set in the conservative directions, for (1)flux peaking augmentation factors as shown in Figure 4'-1, (2)a measurement-calculational uncertainty factor of 1.062, (3)an engineering uncertainty factor of 1.03, (4)an allowance of F 01 for axial fuel densification and thermal expansion, and (5)a THERMAL POWER measurement uncertainty factor of 1.02.3/4.2.2 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS-F AND F AND MUTHAL POWER TILT-T The limitations on F and T are provided to ensure that the assumptions xy q used in the analysis for establishing the Linear Heat Rate and Local Power Density-High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits.The limitations on Fand T are provided to ensure that the assumptions used in the analysis establishing the DNB Margin LCO, the Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits.If F , F or T exceed their basic limitations, operation may continue under xy'q the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the ST.LUCIE-UNIT 2 8 3/4 2-1

'NSTRUMENTATIOH BASES 3/4.3.3.10 RADIOACTIVE GAS OUS EFFLUENT MONITORING IHSTRUMENTATIOH The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm/trip setpoints for these instruments shall be calculated in ac'cordance

'ith the methodology in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system.The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.

Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.

ST.LUCIE-UNIT 2 B 3/4 3-4 ADMINISTRATIVE CONTROLS FUNCTION (ContInEEnd) e.instrumentation and control g.radi o 1 ogi ca 1 sa f e ty mechanical and electrical engineering quality assurance practices XTISar~V~VISag ti's4 Csae n~4~e')COMPOSITION 6.5.2.2 The CNRB shall be composed of the following members: er: Member: Member: Hember: Member: Member: Member.er: Vice President, Advanced Systems and Tech Chief Engineer, Power Plant En in Yfce President, Nuc ergy Yice Pres r.Operations Director ua1>rance er, Nuclear Fuel Power Plant Engineering Principal Power Plant Engineering Senior Project Hana The Chairman shall be a member of the CNRB and shall be designated in writing.ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing.by the CNRB Chairman to serve on a temporary basis;however, no more than two alternates shall participate as voting members in CNRB activities at any one time.CONSULTANTS 6.5.2.4 Consultants shall be uti lfzed as determined by the CNRB Chairman to provide expert advice to the CNRB.E IIII 6.5.2.5 The CNRB shall Neet 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 alternate.

QUORUM 6.5.2.6 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 four CNRB members including alternates.

No more than a minority of the quorum shall have line responsibility for operation of the unit.ST.LUCIE-UNIT 2 6-10 Amendment No.13 Member: Group Vice President Member: Group Vice President-Nuclear Energy Member: Vice President-Engineering, Projects Construction Member: Vice President-Nuclear Operations Member: Director-Nuclear Licensing Member: Director-Quality Assurance Member: Chief Engineer-Power Plant Engineering Member: Member: Member: Manager-Nuclear Energy Services Manager-Nuclear Fuel I Senior Project Manager-Power, Plant Engineering ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) r c.Secondar Water Chemistr 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 mea'sure 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, (iv)Procedures for the recording and management of data, (v)(vi)Procedures defining corrective actions for all off-control point chemistry conditions, and 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.Backu Method for Determinin Subcoolin Mar in 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,'nd (ii)Procedures for monitoring.

Post-accident Sam lin A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following: (i)Training of personnel, (ii)Procedures for sampling and analysis, and (iii)Provisions for maintenance of sampling and analysis equipment.

ST.LUCIE-UNIT 2 6-15 ADMINISTRATIVE CONTROLS 6.9 REPORTING RE UIREHENTS ROUTINE REPORTS: 6.9.1 In addition to the applicable reporting requirements of Title 10, Code 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 to 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.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 comparison of these values with design predictions and specifications.'ny 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 conmencement of commercial power operation, or (3)9 months following initial criticality, whicheve~is ear)iest.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-6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted 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.A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures-A single submittal say be made for a multiple unit station.The submittal 1/should combine those sections that are common to all units at the station.C ST.LUCIE-UNIT 2 6-16'Amendment No.

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) greater than 100 mrems/yr and their associated man-rem exposure according to work and job functions,-e.g., reactor ope'rations and 2/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 2(C of the individual total dose need not be accounted for.In the aggregate, at least 80K of the total whole body dose received from external sources should be assigned to specific major work functions.

HONTHLY 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.lo cPL 5o f-This tabulation supplements the requirements of f20.407 of 10 CFR Part 20.2.)ST.LUCIE-UNIT 2 6"17 Amendment No.13 ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) locations specified in the Table and Figures in the ODCM, as we11.as summariied and tabulated results of these analyses and measurements'n the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.In the event that some individual results are not available for inclusion with the report, the.report shall be submitted noting and explaining the reasons for the missing results.The missing data shall be submitted as soon as possible in a supplementary report.The reports shall also include the following:

a seminary description of the radiological environmental monitoring program;at least two legible maps" covering all sampling locations keyed to a table giving distances and direc-tions from the centerline of one reactor;the results of the Interlaboratory Comparison Program, required by Specification 3.12.3;discussion of all deviations from the sampling schedule of Table 3.12-1;and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.

6.9.1.9 At least once every 5 years, an estimate of the actual population p1 h11 the NRC.6.9.1.10 At least once every 10 years, an estimate of the actual population the NRC.SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each report.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 5 years: a.b.C.d.Records and logs of unit operation covering time interval at each power level.Records and logs of p~incipal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.All REPORTABLE EVENTS:.Records of surveillance activities; inspections and calibrations required by these Technical Specifications.

Records of changes sade to the procedures required by Specification 6.8.1.One map shall cover stations near the SITE BOUNDARY;a second shall include the sore distant stations.'ST.LUCIE-'UNIT 2 6-20.Amendment No.

ADHINIST TIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM PCP Licensee initiated changes to the PCP: l.Shall be submitted to.the Comnission in the Semiannual Radioactive Effluent Release Report for the period in,which the change(s)was made.This submittal shall contain: a.Sufficiently detailed information to totally suport the rationale for the change without benefit of additional or supplemental information; b.A determination that the change did not reduce the overall conformance of the dewatered bead resin to existing criteria for radioactive wastes;and c.Documentation of the fact that the change has been reviewed and found acceptable by the FRG.2.Shall become effective upon review and acceptance by the FRG.6.14 OFFSITE DOSE CALCULATION MANUAL ODCH~M&Licensee initiated changes to the ODCH: l.Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s)was aade effective.

This submittal shall contain: a.Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supple-.mental information.

Information submitted should consist of a package of those pages of the ODCH to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations)ustifying the change(s);

b.A determination that the change will not reduce the accuracy or reliability of dose calculations.or setpoint determinations; and c.Documentation of the fact that the change has been reviewed and found acceptable by the FRG.2.Shall become effective upon revie~and acceptance by the FRG.ST.LUCIE-UNIT 2 643 Amendment Np.

ATTACHMENT 2 SAFETY EVALUATION This proposed amendment represents a broad administrative update of the St.Lucie Unit 2 Technical Specifications.

The 40 affected pages can be put, into six categories:

1.Change terminology from"4 Delta-k/k" to"pcm" (10 pages).2.Remove outdated material (11 pages).3.Correct typographical errors (15 pages).4.Incorporate Generic Letter 84-13 on snubbers (10 pages)~5.CNRB update to reflect current organization (1 page).6.Change addressee for certain reports to reflect revisions to 10 CFR 50.4 (3 pages).Cate or 1 The following ten pages in Technical Specification Section 3/4.1,"Reactivity Control Systems," are revised to change units of Delta-k/k" to the equivalent"pcm": 3/4 1-1 3/4 1-2 3/4 1-,3 3/4 1-5 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-14 B 3/4 1-1 B 3/4 1-2 Cate or 2 The following eleven pages are revised to remove one-time requirements that have been satisfied.

These requirements were effective for some period of time that is now past, i.e., prior to initial criticality, prior to 5~power, prior to the first refueling outage, prior to installing a specific plant modification, and prior to receiving NRC approval of certain programs: 3/4 3-25 3/4 3-26 3/4 4-38 3/4 7-4 3/4 7-39 3/4 9-5 Cate or 3 The following 15 pages are revised to correct typographical errors: 3/4 3-34 3/4 3-39 3/4 3-53 3/4 3-57 3/4 4-27 3/4 5-6 3/4 6-15 3/4 8-1 3/4 8-2 3/4 11-6 3/4 11-10 B 3/4 0-3 B 3/4 1-4 B 3/4 2-1 B 3/4 3-4 Cate or 4 Page 6-10 is revised to expand the composition of the Company Nuclear Review Board (CNRB)from eight to ten people, and to revise position titles to conform with the current FPL organization.

Cate or 5 A recent change to 10 CFR 50.4, effective January 5, 1987, directs that Part 50 reports be addressed to the Document Control Desk, Washington, DC, 20555.The following three pages are re-vised accordingly:

6-16 6-17 6-20 ATTACHMENT 3 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission's regulations, l0 CFR 50.92, which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (I)involve a significant increase in the probability or consequences of an accident previoiusly evaluated; or (2)create the possibility of a new or different kind of accident from any accident previously evaluated or (3)involve a significant reduction in a margin of safety.Each standard is discussed as follows: (I)Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The changes being proposed are administrative; they do not affect assumptions contained in plant safety analyses nor do they affect Technical Specifications that preserve safety analysis assumptions.

Therefore, the proposed changes do not affect the probability or consequences of accidents previously analyzed.(2)Use of the modified specification would not create the possiblity of a new or different kind of accident from any accident previously evaluated.

The changes being proposed are administrative; they will not lead to material procedure changes or to physical modifications.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident.(3)Use of the modified specification would not involve a significant reduction in a margin of safety.The changes being proposed are administrative; they do not relate to, or modify, the safety margins defined in and maintained by the Technical Specifications.

Therefore, the proposed changes do not involve any reduction in a margin of safety.Based on the above, we have determined that the amendment request does not (I)involve a significant increase in the probability or consequences of an accident previously evaluated, (2)create the probability of a new or different kind of accident from any accident previously evaluated, or (3)involve a significant reduction in a margin of safety;and therefore does not involve a significant hazards consideration.

E JW4/0 I 2/2 W il A S~I~W II, s I I s IS A s Wilt~s s s Ih II st'I I s a~tl II.~I~I ,~W I II's s~I S I I s a~s/~W I I JI I tl I~s I--~A s I W I!'~I il I Ii s S I I~I II S I,~.b 4 II I II I s I ss gs IW'W s I I IW Sli" hi'I I li~I s t W's