Text

Proposed Tech Specs Incorporating Administrative Changes That Improve Consistency Throughout TSs & Related Bases
	ML17229A360
Person / Time
Site:	Saint Lucie
Issue date:	05/29/1997
From:	; FLORIDA POWER & LIGHT CO.
To:	;
Shared Package
ML17229A359	List: ML17229A358; ML17229A360;
References
	NUDOCS 9706040203
	Download: ML17229A360 (71)
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0 aam<m ~14 7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION . .. 3/4 7-13 3/4.7.3 COMPONENT COOLING WATER SYSTEM .. . 3/4 7-14 3/4.7.4 INTAKE COOLING WATER SYSTEM .. . 3/4 7-16 3/4.7.5 ULTIMATEHEAT SINK 3/4 7-18 3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATIONSYSTEM .. .. 3/4 7-20 3/4.7.8 ECCS AREA VENTILATIONSYSTEM 3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION 3/4 7-27 3/4.7.10 SNUBBERS 3/4 7-29 3/4.8.1 A.C. SOURCES 3/4 8-1 Operating 3/4 8-1 Shutdown 3/4 8-7 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS 3/4 8-8 A.C. Distribution - Operating .. 3/4 8-8 A.C. Distribution - Shutdown 3/4 8-9 D.C. Distribution - Operating ~ . ~.... 3/4 8-10 D.C. Distribution - Shutdown . 3/4 8-13 ST. LUCIE - UNIT 1 Vll Amendment No. 85, 44, 66 9706040203 970529 PDR ADQCK 05000335 P PDR

LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Pressure-Low (Continued) 0" to interfere with normal operation, but still high enough 'to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of + 22 psi in the accident analyses.

Steam Generator Mater Level - Low The Steam Generator Mater Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded due to loss of steam generator heat sink. The specified !

setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide ore b~

Local Power Densit -Hi h before reactor coolant system subcooling is lost."

'he local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power ensity in the fuel which corresponds to fuel centerline melting will not occur as a consequence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowab1e limits of Figure 2.2-2. The AXIAL SHAPE INDEX is calculated from the pper and lower ex-core neutron detector channels. The calculated etpoints are generated as a function of THERHAL POMER level with the llowed CEA group position being inferred from tlie THERHAL POMER level.

he trip is automatically bypassed below 15 percent power.

The maximum AZINUTHAL POMER TILT and maximum CEA misalignment per-itted for continuous operation are assumed in generation of the set-oints. In addition, CEA group sequencing in accordance with the pecifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum nsertion of CEA banks which can occur during any anticipated operational ccurrence prior to a Power Level-High trip is assumed.

ST. LUCIE UNIT 1 B 2-6 Amendment No.

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LIMITING SAFETY SYSTEM SET~ iNGS BASES Loss of Turbine A Loss of Turbine trip causes a direct, reactor trip when operating above 15~ of RATED THERMAL POMER. This +rip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protec-tion System.

Rate of Chan e of Power-High The Rate of Change of Power-High trip is provided to protect the core during startup operations and its use serves as a backup to the. administra-eP - . - '~h~eei+M tively enforced startup rate limit. ~s-t~-se+po+n~ees-net-co~and d

I The trip is not crcditcd in any design basis accident evaluated in UFSAR Chapter 15; however, thc trip is I considered in thc safety analysis in that the presence of this trip function prccludcd the need for specific analyses of other events initiated from subcritical conditions.

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ST. LUCIE - UNIT 1 B 2-8 , Amendment No.

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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORATION CONTROL SHUTDOWN. MARGIN - T > 200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 3600 pcm.

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

ACTION:

Mith the SHUTDOMN MARGIN < 3600 pcm, imnediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOMN MARGIN shall be determined to be > 3600 pcm:

Mithin 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 inoperabl .+oy+

gg74 If the ino rable CEA is immovabl or untri able the above required SHUTDOWN MARGIN shall be increased by an amount at.

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least equal to the withdrawn worth of the imnovable or un-

~d> trippable CEA{s). ca ~ yesw 4 ef egcessNe,4 te4ioe oL" eLachmv6eoL ivL+zy emeca Mhen in MODES 1 or 2, at least once per y v that CEA group withdrawal is within the Power t}ependent Insertion Limits of Specification 3.1.3.6.

Ci When least in MODE 2, at thereafter once per hour least once during CEA withdrawal and at until the reactor is critical.

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 Power Dependent Insertion Limits of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

Mith K > 1.0.

kith K < 1.0.

ff ST. LUCIE - UNIT 1 3/4 1-1 Amendment No. 87.

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REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR'PERATION I

3.1;1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be > 3000 gpm whenever a reduction in Reactor Coolant System boron concen ation is being made.

APPLICABILITY: ALL MODES.

ACTION:

Mith the flow rate of reactor coolant to the reactor pressure vessel

( 3000 gpm, imediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

SURVEILLANCE REQUIREMENTS 4.1.1.3 The flow rate of reactor coolant to the reactor pressure vessel shall be determined to be > 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 either:

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 > 3000 gpm to the reactor pressure vessel.

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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 fiow 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 Specificatio 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.

MODES 5 and 6.

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

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 othetwise 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 RCSpressureboundarydoesnotexist,or(b)~ ocha in um sareo erable. Inthe latter case: 1) all charging pumps shall be disablecf; 2) heatup and cooldown rates shall be limited in accordance with Figure 3.1-1 b; and 3) at RCS temperatures below 115'F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed:

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HCV-3646 HCV-3647 ST. LUCIE - UNIT 1 3/4 1W Amendment No. 66, 8+, 98,

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3.1.2.3 At least one charging pump or high pressure safety injection pump'n the boron injection fiow path required OPERABLE pursuant to Specificatio 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

MODES 5 and 6.

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

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 Specification 4.0.5.

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

(a) the RCS pressure boundary does not exist, or (b) o char in um s are o erahle. In the latter case: 1) all charging pumps shall be disabled; 2) heatup and cooldown rates shall be limited in accordance with Figure 3.1-1 b; and 3) at RCS temperatures below 115'F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed:

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EACTIVITYCO YSTEMS FULL LENGTH CEA POSWON continued LIMmNG CONDITION FOR OPERATION continued

2. Declared inoperable and satisfy SHUTDOWN MARGIN requirements of SpeciTication 3.1.1.1. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of SpeciTication 3.1.3,6 for up to 7 days per occurrence with a total accumulated time of < 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 the inoperable CEA shall be aligned to within 7.5 inches of the 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 3.1.3,6 during subsequent operation. tv ten"- rt'ed .

b) The SHUTDOWN MARGIN requiremen f 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 the>wise-,be-ii-t-aHeast 0&7-STANBBY-eithi OQMftvLQc ll< eely+ Heft $ TAPQsf b4'L444A 4'AW ~ ~<8

e. With one full length C 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 3.1-1a.

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

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

. 2 Declare the CEA inoperable and satisfy 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) 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 the inoperable CEA shall be aligned to within 7.5 inches of the 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 3.1.3.6 during subsequent operation.

ST. LUCIE - UNIT 1 3f4 1-21 Amendment No. V4~

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EACTIVITYCONTR Y TEMS L LENG CEA POSWON continued LIMmNG CONDITION FOR OPERATION continued b) The SHUTDOWN MARGlN requirement of Specification 3.1.1.1 is determined lea once ours

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g. ith 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 specified alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length 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 individual 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 shall be determined 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 OPERABl E 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 Specmcation 3.1.3.6 and that the circuit prevents the regulating CEAs from being inserted beyond the Power Dependent Insertion Limit of COLR Figure 3.1-2:

'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. LUCIE - UNIT 1 3I4 1-22 Amendment No. 44, K, V4,

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POWE DIST IB ON LIMITS TAL NTEGRATED R I L P A ING F CTOR - Fi LIMITINGCONDITION FOR OPERATION 3.2.3 The calculated value of F, shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1'.

ACTION:

With F, 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 F, to within the limits of COLR Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined from COLR Figure 3.2-3 shall then be used to establish a revised upper THERMAL POWER level limit on COLR Figure 3Z-4 (truncate Figure 3.2A at the allowable fraction of RATED THERMAL POWER determined by COLR Figure 3.2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of COLR Figure 3.2A.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F, shall be calculated by the expression F, = F,(1 + 7 ) when F, is calculated with a non-full core power distribution analysis code and shall be calculated as F, = F, when calculations are performed with a full core power distribution analysis code. F, 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 if the AZIMUTHALPOWER TILT (Tg is > 0.03.

See Special Test Exception 3,10.2.

ST. LUCIE - UNIT 1 3/4 2-9 endment No. 87; 88, 48, C:

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System {except the pressurizer}, temperature and pressure shall be limited in accordance with the limit lines shown on 'Figures 3.4-2a, 3.4-2b"and 3.4-3 during heatup, cooldown, criticality, and inservice., L leak and hydrostatic testing.

APPLICABILITY: At all times.*f ACTION:

Mith any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an analysis to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tav to less than 200'F within .the'following 30'hours in accordance with' Figures 3.4-2b and 3.4-3.

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~(g ) mrs'hen the flo ath from the RMT to the RCS via a single HPSI pump is established pe .l.2. the heatup and cooldown rates shall be established in accordance with 'Fig. 3.$ -1b.

4During hydrostatic testing operations above system design pressure, a maximum temperature change in any one hour period shall be limited to 5'F.

ST LUCIE - UNIT 1 3/4 4-21 Amendment No. $ , Q,

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, CONTAINMENT SYST SURVEILLANCE REQUIREMENTS contInued Pages 3/4 through 3/4 6-9 have been OELETEO.

Page 3/4 6-10 Is the next valid page.

ST. LUCIE ~ UNIT 1 Amendment No. 88, 4&7

Pages 3/4 through 3/4 6-22 have been DELETED.

Page 3/4 6-23 Is the next valid page.

3/4 6-20 Amendment No. 96, 448

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CONTAINMENT SYSTEMS 3/4.6.5 VACUUM RELIFF VALVES LIMITING CONDITION FOR OPERATION 3.6.5.1 E

The OPERABLE E E 2.25 + 0.25 inches Mater Gauge E

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containment vessel to annulus vacuum differential.

relief valves shall APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Mith one containment vessel to annulus vacuum relief valve inoperable, restore the valve 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 SHUTDOMN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.5.1 No additional Surveillance Requirements other than those required by Specification 4.0.5 and at least once per 3 years verify that the vacuum relief valves open fully within 8 seconds at 2.25 + 0.25 inches Mater Gauge differ.'ential.

ST. LUCIE - UNIT l 3/4 6-26 Amendment No.

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Ej 4, e REFUELING OPERATIONS CONTAINMENT PENETRAT IONS 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, and

c. Each penetrationPmx~pt-abbrev-Wed-H~

SpecAA~~>~~ providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed by an isolation vaive, blind flan e r nua valve, or QgOe Va, M@5 89LT 4lYC, 4P oA o.u. t~ ~&+a+'balls ulacÃl

2. Be capable of being closed by an OPERABLE automatic ~<~<~4t<~e containment isolation valve, or ~~not,

3. Be capable of being closed by an OPERABLE containment vacuum relief valve.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immedi-ately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment. The provisions of Specification 3.0.3 are not applicable.

SURVE ILLANCE REQUIREMENTS 4.9.4 Each of the above required containment penetrations shall be determined to be either in its closed/isolated 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 CORE ALTERATIONS or movement of irradiated fuel in the containment by:

a. Yerifying the penetrations are in their closed/isolated condition, or

b. Testing the containment isolation valves per the applicable portions of Specifications 4.6.3.1.1 and 4.6.3.1.2.

ST . LUCIE - UNIT 1 3/4 9-4 Amendment No+7

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REFUELING OPERATIONS MANIPULATOR CRANE OPERABILITY LIMITING CONDITION FOR OPERATION 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 < 3000 pounds.

APPLICABILITY: During movement of CEAs or fuel assemblies within the reactor pressure vessel.

ACTEON:

Mith the requirements for crane OPERABILITY not satisfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel.

W~t'ov SURVEILLANCE REQUIREMENTS 4.9.5 The manipulator crane used for movement of CEAs or fuel assem-blies 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 l.oad cut off ~ the crane load exceeds 3000 pounds.

'I ST. LUCIE - UNIT 1 3/4 9-6

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3.10.2 group height, insertion and power distribution limits 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 may be suspended during the performanc ofP YSIC TESTSprovided:

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a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMALPOWER, and

b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10Z2 below.

hQX1QH'he MODES 1 and 2.

With any of the limits of Specification 3.2.1 being exceeded while the requirements of SpeciTications 3.1.1.4, 3.1.3.1,, 3D.3.$ 3.1.3.6, 3.2.3 and 32.4 are suspended, either.

a. Reduce THERMAI 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 .

4.10.2.1 The THERMAI POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3, or 3.2A are suspended and shall be veriTied to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3Z.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.4 during PHYSICS TESTS above 5% of RATED THERMALPOWER in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.3, or 3,2.4 are suspended.

ST. LUClE - UNiT 1 3/4 10-2 Amendment No. BP, ~,

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RADIOACTIVE FF U S EXPLOS V M R LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to Zf by volume whenever the hydrogen concentra-tion exceeds 4% by volume.

~A>>t Bt ITY: At 1l t'CTION:

a. Mith the concentration of oxygen in the waste gas decay tank greater than 2% by volume but less than or equal to 4X 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. kith the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2X by volume, immediately suspend all additions of waste gases to the system and immediately caamence reduction of the concentration of oxygen to less'han or equal to 2X by voluine.

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

e SURVEILLANCE REQUIREMENTS 4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall be determined to be within the above limits by continuously monitoring the waste gases in the on service waste gas decay tank

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~~h gen-aeter ~ed 4.11.2.5.2 Mith the oxygen concentration in the on service waste gas decay tank greater than 2X by volume as determined by Specification 4.11.2.5.1, the concentration of hydrogen in the waste gas decay tank shall 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 .

ST. LUCIE - UNIT 1 3/4 11-14 Amendment Ho. 59+

REACTIYITY CONT..OL SYSTEMS BASES 3/4.1.2 BORATION SYSTEl!S Continued The boron addition caoabili ty after the plant has been placed in NODES 5

'nd 6 requires either 3650 gallons of 2.5 to 3.5 weight percent boric acid solution (4371 to 6119 ppm boron) from the boric acid tanks or 11,900 gallons of 1720 yam borated water from the refueling water tank to makeup for contraction of the primary coolant that could occur 200'F to 140'F.,

if the temperature is lo<<ered from The restrictions associated with the establishing of the flow path

.rom the RMT to the RCS via a single HPSI pump provide assurance that l'~k~

Aopendix G pressure/temperature limits will not be exceeded in the case of any inadvertent pressure transient due to a mass addition to the RCS.

3/4.1. 3 NOVABLE CONTROL ASSEMBLIES The specificai distribution limits IfRCS pressure boundary integrity docs not exist as dcfincd in Specification 1.16, maintained, and (3) thcsc restrictions are not required. Additionally, a limiton thc maximum number of limited to acceptah opcrablc HPSI pumps is not ncccssaiy when thc prcssurizcr manway cover or the reactor vessel head is removed.

Th ACTION t requSr ementS are aC ... ~ - IVIIQI I"Cbbl Ilail.IVII'CClllrllCll>UC't 4IIQle the original criteria are met.

The ACTION statements applicable to an immovable or untrippable CEA and to a large misalignment (>> 15 ihches) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOMN NARGIN e For small misalignments (< 15 inches} of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed. in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect-on the time dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on, the available SHUTDOMN MARGIN, and 4) a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to r estore the CEA to within its alignment requirements prior to initiating a reduction in THERMAL POMER. The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs, and (3) minimize the effects of xenon redistribution.

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 istortion of the core power distribution. This distr ibution may, in turn, have a significant, effect on (1) the available SHUTDOMN MARGIN, (2) the time-dependent kong-term power distributions relative to those used in generating

. LCOs and LSSS setpoints, and (3) the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with the large mis-L alignment of the CEA requires a prompt realignment of the misaligned CEA.

ST. LL'CIE - UNIT 1 B 3/4 1-3 Amendment Nn. 27.77.27/~

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REACTOR COOLANT SYSTEM BASES "3 4.4.13 POMER OPERATED RELIEF VALVES and 3 4.4.14 REACTOR COOLANT PUMP STARTING The low temperature overpressure protection system (LTOP) is designed to prevent RCS overpressurization above the 10 CF Appendix G operating limit curves (Figures 3.4-2a and 3.4-2b) at RCS temp atures at or below 304 F during heatup and 281 F during cooldown. The TOP system is based on the use of the pressurizer power-operated relief valve (PORVs) and .the implementation of administrative and operational controls.

The PORYs aligned to the RCS with the low pressure setpoints of 350 and 530 psia, restrictions on RCP starts, limitations on heatup and cooldown rates, and disabling of non-essential components provide assurance that Appendix G P/T limits will not be exceeded during normal operation or design basis overpressurization events due to mass or energy addition to the RCS. The LTOP system APPLICABILITY, ACTIONS, and SURVEILLANCE REQUIREMENTS are consistent with the resolution of Generic Issue 94, "Additional Low-Temperature Overpressure Protection for Light-Mater Reactors," pursuant to Generic Letter 90-06.

3 4.4.15 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The redundancy design of the Reactor Coolant System vent systems serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System vent system are consistent with the requirements of Item II.b.l of NUREG-0737, "Clarification of THI Action Plan Requirements," November 1980.

ST. LUCIE UNIT 1 B 3/4 4-15 Amendment No. ~,68;8+,~,

REFUELI PERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitation on minimum boron concentration ensures that: 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel. The limitation on K, is sufficient to prevent reactor criticality with all full length rods (shutdown and regulating) fully withdrawn.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the wide range logarithmic range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcnticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressunzation potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 MANIPULATORCRANE OPERABIUTY The OPERABILITY requirements of the cranes used for movement of fuel assemblies ensures that: 1) each crane has sufficient load capacity to lift a fuel element, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

In accordance with Generic Letter 91-08, Removal of Component Lists from the Tcchnical Specifications, the of locked or sealed closed containment isolation valves on an intermittent basis under administrative 'pening control includes the following considerations: (1) stationing an operator, who is in constant communication

'ith thc control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions willnot preclude access to close the valves and that

'his action willprevent the release of radioactivity outside thc containment.

ST. LUCIE- UNIT1 B 3/4 9-1 Amendment No. 66

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DESIGN FEATURES 2.1 2 SHIELD B Minimum annular space = 4 feet.

b. Annulus nominal volume = 543,000 cubic feet.

Co Nominal outside height (measured from top of foundation hase to the top of the dome) = 230.5 feet.

d. Nominal inside diameter = 148 feet.

e. Cylinder wall minimum thickness 3 feet.

Dome minimum thickness = 2.5 feet.

go Dome inside radius - 112 feet.

DESIGN PRESSURE AND T ERATUR 5.2.2 The containment vessel is designed and shall he maintained for a maximum internal pressure of 44 psig and a temperature of 264oF.

P ENETRATION 52.3 Penetrations through the containment structure are designed and

.all be maintained in accordance with the original design provisions ontained in Sections 3.8.2.1.10 and 6.2.4 of the FSAR with allowance for nc"mal degradation pursuant to the applicable Surveillance Requirements.

FUEL ASSEMBLI 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of between 134.1 and 136.7 inches. Individual fuel assemblies shall contain fuel rods of the same nominal active 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 enrichment of 2.83 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading-5.3. Except for special test as authorised by the NRC, all fuel assemb ies under control element assemblies shall be sleeved with a sleeve esign previously approved hy the NRC.

ST LUCIE - UNIT 1 5-4 Amendment No.

St. Lucie Unit 1 and Unit 2 Docket Nos. 50-335 and 50-389 Proposed License Amendments ST ~ LUCIE UNIT 2 MARKED-UP TECHNICAL SPECIFICATION PAGES Page XVIII Page XIX Page B 2-1 Page B 2-4 Page B 2-5 Page B 2-6 Page 3/4 1-1 Page 3/4 1-19 Page 3/4 2-14 Page 3/4 3-26 Page 3/4 6-3 Page 3/4 6-21 Page 3/4 6-26 Page 3/4 8-7 Page 3/4 9-6 Page 3/4 11-14 Page B 3/4 2-2

NO EX I

ADMINISTRATIVE CONTROLS 0'ECTION PAGE

6. 1 RESPONSIB I LITY 6-1

6. 2 ORGANIZA ION......

6.2. 1 ONSITE AND OFFSITE ORGANIZATIOK.. 6-1 6.2.2 UNIT STAFF....... 6-2 BEPENB EElHN~ROU .

FUN 6-6 COMPOSITION............ 6-6 RESPONS IB ILITIES.. 6-6 A ~ o ~ ~ ~ 6-6 ECORD ~ ~ ~ o ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ ~ ~ ~ \ ~

6. 2 SHIFT TECHNICAL ADVISOR.......... 6-6 5

6o 3 UNIT STAFF UALIFICATIONS 6-6

6. 4 TRAINING.. 6-7 6.5 REVIEW AND AUDIT. 6-7 6.5. 1 FACILITY REVIEW GROUP 6-7 F UNCTION. \ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-7 COMPOSITION.. 6-7 ALTERNATESo o ~ o o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~

o o ~ ~ 6-7 MEETING FREQUENCY. 6-8 QUORUM.; 6-8 RESPONSIBILITIES. 6-8 AUTHORITY 6-9 RECORDS 6-9 ST. LUCIE - UNIT 2 Amendnent No.~

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 COMPANY NUCLEAR REVIEM BOARD... 0 ~ ~ ~ ~ 6 9 FUNCTION ...........-....----- ~ ~ ~ ~ e ~ o 6 9 COMPOSITION...................... 6-10 ALTERNATES.............. 6-10 CONSULTANTS........ 6-10 MEETING FREQUENCY........ 6-10 QUORUM............ 6-10 REVIEM. 6-11 AUDITS............ ~ ~ ~ tI ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-11 AUTHORITY............. 6-12 RECORDS.. -1 XSceeecaL. gellaH SZSPCedeiSiWV~SS 0 4 S ~ i ~ 4 ~ Oa gg ~ ~ ~ ~ ~ ~ ~ ~ 4 -IN.

6.6 REPORTABLE EVENT ACTION..................................... 6-13

6. 7 SAFETY LIMIT VIOLATION.... 6-13 6.8 PROCEDURES AND PROGRAMS.. 6-13 6.9 REPORTING RE UIREHENTS . 6-16 6.9.1 ROUTINE REPORTS....... 6-16 STARTUP REPORT............................................. 6-16 ANNUAL REPORTS... 6-16 MONTHLY OPERATING REPORTS. 6-17 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT. 6-18 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT... 6-19 6.9.2 SPECIAL REPORTS 6-20

6. 10 RECORD RETENTION. .. . . . . ... ............ ...... 6-20 6.11 RADIATION PROTECTION PROGRAM........................ 6-21 ST. LUCIE UNIT 2 XIX Amendment No. k3,

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2.1 SAFETY L IMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting will occur. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling {ONB) and the resultant sharp reduction in heat transfer coefficient. ONB is not a directly measurable parameter during operation and therefore THERMAL POMER and Reactor Coolant Temperature and Pressure have been related to ONB through the CE-I correlation. The CE-I DNB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal o er ional transients, and anticipated transients is limited to the DNB-SAFOL o . in conjunction with the Extended Statistical Combination of Uncertain-ties (ESCU). This value is derived through a statistical combination of the system parameter probability distribution functions with the CE-I ONB correla-tion uncertainty. This value corresponds to a 95K probability at a 95X confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figure 2.1-1 show conservative loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the DNB-SAFOL is not violated for the family of axial shapes and corresponding radial peaks shown in Figure B 2.1-1. The limits in Figure 2.1-1 were calculated for reactor. coolant inlet temperatures less than or equal to 580'F. The dashed line at 580 F coolant inlet temperature is not a safety limit; however, operation above 580 F is not possible because of the actuation of. the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POMER levels higher than 112'f RATED THERMAL POMER is prohibited by the high power level trip setpoint specified in Table 2.1-1. The area of safe operation is below and to the left of these lines.

The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.

The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the Specified Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operational Occurrences.

ST. LUCIE - UNIT. 2 B 2-1 Amendment No. 8 4S R

L f SAFETY LIMITS AND LIMI ING SAFETY SYSTEM SETTINGS BASES Vari ble Power vel - Hi h A Reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure Trip.

The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.61K above the indicated THERMAL POWER level. Operator action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL POWER decreases. The trip setpoint has a maximum value of 107.0X of RATED THERMAL POWER and a minimum setpoint of 15.0X of RATED THERMAL POWER. Adding to this maximum v'alue the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 112X of RATED THERMAL POWER, which is the value used in the safety analysis.

Pressurizer Pressure-Hi h The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam line safety valves, provides Reactor Coolant System pro-tection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2375 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.

Thermal Mar in Low Pressure l. 28 The Thermal Margin/Low Pressure t is provided to prevent operation when the DNBR is less than the DNB-SAFDL of . , in conjunction with ESCU methodology.

The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of hT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continu-ous operation are assumed in the generation of this trip function. In additi'on, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

o, The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment tits*

all owance dp ERMA~eWER to compen-include: .

or poten ial power m asurement error; an allowance WM+V-F to compensate sa e for potential temperature easurement uncertainty; and Mw4her allowance to compensate for p essure measurement error end time delay associated with of~

providing effective te nation of the occurrence that exhibits the most rapid decrease in margin to t e safety limit. wance-i~nade-u~~

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ST. LUCIE UNIT 2 B Amen ment No. 8 ,50,

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Containment Pressure-Hi h The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to or concurrently with a safety i+ection (SIAS). This also provides assurance that a reactor trip is initiated prior to or concurrently with an MSIS.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. Yhe setpoint of 620 psia is sufficiently below the full load operating point of approximately 885 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of 30 psi in the sa'fety analyses.

Steam Generator Level-Low The Steam Generator Level-Low trip provides protection against a loss of feedwater Row incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to loss of the steam generator heat trip also limits (SAFDL)

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sink. This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to provide bA protects against violation of the specified acceptable fuel design for ONBR, offsite dose and the loss of shutdown margin for Ifi~

asymmetric steam generator transients such as the opening of a main steam safety valve or atmospheric dump valve.

Local Power Densit -Hi h sufficient time for any operator action to initiate auxiliary feedwatcr

.'The Local Power Densi: bc fore reactor coolant system subcooling is lost.""

monitoring, is provided to <

fuel which corresponds to fuei cenxertine meting will not occur as a consequence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.

The AXIAL SHAPE INDEX is calculated from the upper and lour excore neutron detector channels. The calculated setpoints are generated as a function of THERMAL POMER level with the allowed CEA group positiion being inferred from the THERMAL POMER level. The trip is automatically,.bypassed below 15% power.

The maximum A1IMUTHAL POMER TIL'T and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.S is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip i s assumed.

ST. LUCIE>> UNIT 2 B 2-5 Amendment No.

i SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS BASES RCP Loss of Co onent Cool in Mater A loss of component cooling water to the reactor coolant pumps causes a delayed reactor trip. This trip provides protection to the reactor coolant pumps by ensuring that plant operation is not continued without cooling water available. The trip is delayed 10 minutes following a reduction in flow to below the trip setpoint and the trip does not occur if flow is restored before 10 minutes elapses. No credit was taken for this trip in the safety analysis.

Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protective System.

Rate of Chan e of Power Hi h The Rate of Change of Power-High trip is provided to protect the core .

during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit.. Ms-t~p-setpem~o~o~orrespon~

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~'y~. tt The Reactor Coolant Flow - Low trip provides core protection against ONB in the event of a sudden significant decrease in RCS flow. The reactor trip setpoint on low RCS flow is calculated by a relationship between steam generator differential pressure, core inlet temperature, instrument errors and response times. Nen the calculated RCS flow falls below the trip setpoint an automatic reactor trip signal is initiated. The trip setpoint and allowable values ensure that for a degradation of RCS flow resulting from expected transients, a reactor trip occurs to prevent violation of local power density or ONBR safety limits.

'he trip is not credited in any design basis accident cvaluatcd in UFSAR Chapter 15; however, the trip is

'onsidered in the safety analysis in that the presence of this trip function precluded the need for specific analyses of other events initiated from subcritical conditions.

.ST. LUCIE - UNIT 2 B 2-6 Amendment Ho. p~

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'/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.l. 1 BORATION CONTROL SHUTDOWN MARGIN - T GREATER THAN 2004F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTO(M MARGIN shall be greater than or equal to 5000 pcm.

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

ACTION:

With the SHUTDSA MARGIN less than 5000 pcm, immediately initiate. an6 continue boration at greater than or equal to 40 gps or a solution con-taining greater than or equal to 1720 ppm boron or equivalent until the required SHllTDQS MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTOOMN MARGIN shall be determined to be greater than or equal to 5000 pcm:

Mithin one hour after detection of an inoperable CEA(s) and at least

<.w +lip 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 th i5~+

ino erable CEA is immovable or untri able "he above re uired ol S~ll'~~ SHUTDOWN MARGIN shall be erified acceptable with an increased ' +

~~+ CS all~ce for the withdrawn rth of the imeo le or untrippable CEA(s). mc ca.1~ f'of+cesslvc. ra'c&en 4Y naeCA~NAi~ In&~~ren~

b. 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifyfng that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3. 1.3.6.

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

See Special Test Exception 3.10.1.

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

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e. Mith one full-'engtn CEA misaligned rom any otiier CEA .:n its g. ouv by more than 15 inches beyond the time constraints shown '"..".gure 3.l-la, reduce power to ( 70. of RATED THERl@L POWER prior to completing ACT.ON e.l or e.2. I I

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

2. Declare the CEA inoperable and satisfy SHUTOOMH i~lARGIN requ re-ment of Specification 3.1.1.1. After declaring the CEA inoper-able, operation in NODES 1 and 2 may continue pursuant o the requirements of Specification 3. 1.3.6 provided:*

1 a) Mithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the remainde. of the CEAs in the group with the inoperable CEA shall be aligned to within 7.0 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THEPJCL POMER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

b) The SHUTDOWN MARGIN requirement o Specification 3.1.1.1 is

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O tt'wtsa> i> at eosf gpss ~i~~a~e. Next 4 ho~vs 0 e e . er iEAs in its group by more than 7.0 inches but less than or equal to 15 inches, operation in NODES 1 and 2 may continue, provided that wi.hin ] hour the misaligned CEA{s) is either:

l. Restored to OPERABLE status within its above specified alignmen requirenents, or

2. Declared inoperable and the SHUTDOWN MARGIN requiremen of Specifica-tion 3.1.1.1 is satisfied. After declaring +he CEA inoperabIe, opera-tion in NODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:

a) Mithin 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 the inoperable CEA shall be aligned to within 7.9 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THER'~PL PO'MER level shall be restric ed pursuant to Speci ication 3.1.3.6 during subsequent operation.

b) The SHUTDOMN MARGIN requirerent of Specificat on 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 wi hin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . +a >+~

g. Mith 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 bu+ within its above specified alignment requiren-nts, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6.

l

<<If the pre-misalignment ASI was more negative than -0.15, reduce "ower to <70" of RATED THERMAL POWER or 70>> of the THERMAL POWER leveI prior to the misaTign-ment, whichever is less, prior to completing ACTION e.2.a) and e.2.b).

ST. LUCIE-UNIT 2 3/4 1-19 mendmene No.~

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POMER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-.2:

a. Cold Leg Temperature

b. Pressurizer Pressure

c. Reactor Coolant System Total Flow Rate

d. AXIAL SHAPE INDEX APPLICABILITY: MODE l.

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 POMER to < SX of RATED THERMAL POMER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-2 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

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c. Noble Gas Effluent Monitors

i. Reactor Auxiliary Building Exhaust System (Plant Vent Low Range Honitor) 1,2,38(4 10 - 10 QCi/cc 27 ii. Reactor Auxiliary Building Exhaust Sys-tem (Plant Vent High Range Honitor) 1,2,38(4 10 - 10 pCi/cc 27 Steam Generator Blowdown Treatment Facility Building Exhaust System I, 2, 38(4 10 - 10

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pCi/cc 27 iv. Steam Safety Valve Discharge¹ 1/steam 1, 2, 3 8 4 10 10 QCi/cc 27 header vo Atmospheric Steam .

Dump Valve Dis- I/steam I, 2, 3 8 4 10 10 pCi/cc 27 charge¹ header I vi. ECCS Exhaust 1,2,38(4 10 - 10 QCi/cc 27 e arm r p etpoints are determined and set in accordance with the requirements of the Offsite Dose Calculation Hanual.

¹ The steam safety valve discharge monitor and the atmospheric steam dump valve discharge monitor are the same monitor.

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SURVEILLANCE REQUIREMENTS continued Pages 3/4 through 3/4 6-8 have been DELETED.

Page 3/4 6-9 Is the next valid page.

ST. LUCIE - UNIT 2 3/4 6-3 Amendment No. 86, 66

ry Pages 3/4 hrough 314 6-23 have been DELETED.

Page 3/4 6-24 fs the next vatld page.

ST. LUCIE - UNIT 2 3/4 6.21 Amendment No. 26, 68

(OMTAIMMEMT SYSTSMS 3 4.6.5 VACUUM REL EF VALVES LIMITING CONDITION FOR OPERATION 3.6.5 The primary containment vessel to annulus vacuum relief valves shall be OPERABLE with an actuation setpoint of Ress-the 0.35 inches water gauge.

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ACTION:

With one primary containment vessel to annulus vacuum rel ief valve inoperable, restore the valve 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 SHUTDOMN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.5 No additional Surveillance Requirements other than those required by Specification 4.0.5.

ST. LUCIE - UNIT 2 3/4 6-26 Amendment No.

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I f a c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a safety injection actuation signal.

7. Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a load band of 3800 to 3985 kW'nd during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded within a load band of 3450 to 3685 kW'. The generator voltage and frequency shall be 4160 x 420 volts and 60 ~ 12 Hz within 10 seconds after the start signal; the steadywtate generator voltage and frequency shall be maintained within these limits during this test.

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

9. Verifying the diesel generator's capability to:

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

b) Transfer its load to the offsite power source, and c) Be restored to its standby status.

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

'2) automatically energizes the emergency loads with offsite power.

Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the engin~ounted tanks of each diesel via the installed cross connection lines.

¹ This band is meant as guidance to avoid routine overloading of the engine. Variations in load in excess of this band due to changing bus loads shall not invalidate this test.

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

ST. LUCIE - UNlT2 3f4 8-7 Atnendment No. 99,68

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REFUELING OPERATIONS 3/4. 9. 6 MANIPULATOR CRANE LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane shall be used for movement of fuel assemblies, with or without CEAs, and shall be OPERABLE with:

a. A minimum capacity of 2000 pounds, and

b. An overload cut off limit of less than or equal to 3000 pounds.

APPLICABILITY: During movement of fuel assemblies, with or without CEAs, within the reactor pressure vessel.

ACTION:

Mith the requirements for crane OPERABILITY not satisfied, suspend use of any inoperable manipulator crane from operations involving the movement of CEAs and fuel assemblies within the reactor pressure vessel.

SURVEILLANCE RE UIREMENTS 4.9.6 The manipulator crane used for movement of fuel assemblies, with or without CEAs, 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 2000 pounds and demonstrating an automatic load cut off

~~eWthe crane load exceeds 3000 pounds.

ST. LUCIE - UNIT 2 3/4 9-6 Fgteugm ggv'o

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RAOIOACTIV FF U NTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall be limited to less than or equal to 2X by volume whenever the hydrogen concentration exceeds 4X by volume.

APPLICABILITY: At all times.

ACTION:

Mith the concentration of oxygen in the waste gas decay tank greater than 2X by volume but less than or equal to 4X 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. Mith the concentration of oxygen in the waste gas decay tank greater than 4% by volume and the hydrogen concentration greater than 2X by volume, immediately suspend all additions of waste gases to the system and immediately comence reduction of the concentration of oxygen to less than or equal to 2X by volume.

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

SURVEILLANCE REOUIREHENTS 4.11.2.5.1 wpERAB t

The determined to be within the above

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concentration of oxygen in the waste gas decay tank shall be limits by continuously monitoring the waste t t '

4.11.2.5.2 Mith the oxygen concentration in the on service waste gas decay tank greater than 2X by volume as determined by Specification 4.11.2.5.1, the concen-tration of hydrogen in the waste gas decay tank shall 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 .

ST. LUCIE UNIT 2 3/4 11-14 Amendment No.

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POWER DISTRIBUTION LIMITS BASES assumptions used in establishing the Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid.

An AZIMUTHAL POWER TILT ) 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The requirement that the measured value of T be multiplied by the calculated values of F and F. to determine F' anl F's applicable only when F and F are calculated with a non-full core power cfistribution analysis code. Zen monitoring a reactor core power distribution, F or F with a full core power distribution analysis code the azimuthal ti/t is explicitly accounted for as part of the radial power distribution used to calculate F and F.

The Surveillance Requirements for verifying that F , F' and T are within their limits provide assurance that the actual values oF F , F an) T do not exceed the assumed values. Yerifying F and F after each fuel loading prior to exceeding 75X of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

3 4.2.5 DNB PARAMETERS.

The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and safety analyses. The limits are consistent with the safety analyses assumptions and have b n analytically demonstrated adequate to maintain a minimum DNBR of > . in conjunctio th ESCU methodology throughout each analyzed transient. I.28 The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18-month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12-hour basis.

ST. LUCIE UNIT 2 3/4 2-2 No. 4-B RCI-et-tAmendment

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