Proposed tech specs incorporating administrative changes ...leastIf once per 12 hours thereafter while the 'CEA(s) is inoperabl.+oy+ the ino rable CEA is immovabl or untri able the

0aam<m ~14

7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION .

3/4.7.3 COMPONENT COOLING WATER SYSTEM ..

3/4.7.4 INTAKECOOLING WATER SYSTEM ..

3/4.7.5 ULTIMATEHEAT SINK

3/4.7.6 DELETED

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATIONSYSTEM ..

3/4.7.8 ECCS AREA VENTILATIONSYSTEM

3/4.7.9 SEALED SOURCE CONTAMINATION

3/4.7.10 SNUBBERS

.. 3/4 7-13

. 3/4 7-14

. 3/4 7-16

3/4 7-18

.. 3/4 7-20

3/4 7-24

3/4 7-27

3/4 7-29

3/4.8.1 A.C. SOURCES

Operating

Shutdown

3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS

A.C. Distribution - Operating

A.C. Distribution - Shutdown

D.C. Distribution - Operating ~ . ~....

D.C. Distribution - Shutdown

3/4 8-1

3/4 8-1

3/4 8-7

3/4 8-8

.. 3/4 8-8

3/4 8-9

3/4 8-10

. 3/4 8-13

ST. LUCIE - UNIT 1 Vll Amendment No. 85, 44, 66

9706040203 970529PDR ADQCK 05000335P PDR

LIMITING SAFETY SYSTEM SETTINGS

BASES

Steam Generator Pressure-Low (Continued)0"

to interfere with normal operation, but still high enough 'to provide therequired protection in the event of excessively high steam flow. Thissetting was used with an uncertainty factor of + 22 psi in the accidentanalyses.

Steam Generator Mater Level - Low

The Steam Generator Mater Level-Low trip provides core protectionby preventing operation with the steam generator water level below theminimum volume required for adequate heat removal capacity and assures

!

that the design pressure of the reactor coolant system will not beexceeded due to loss of steam generator heat sink. The specifiedsetpoint provides allowance that there will be sufficient water inventoryin the steam generators at the time of trip to provide oreb~

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

'he

local Power Density-High trip, functioning from AXIAL SHAPEINDEX monitoring, is provided to ensure that the peak local powerensity in the fuel which corresponds to fuel centerline melting will

not occur as a consequence of axial power maldistributions. A reactortrip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowab1elimits of Figure 2.2-2. The AXIAL SHAPE INDEX is calculated from thepper and lower ex-core neutron detector channels. The calculatedetpoints are generated as a function of THERHAL POMER level with thellowed 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 thepecifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximumnsertion of CEA banks which can occur during any anticipated operationalccurrence prior to a Power Level-High trip is assumed.

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

'V

I

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~ 4 .~ s*= $ p <g p+ ~ I tk WA oQe'

es

~ h A~A « ~ ~ ~ k 4 4( t = A p A

LIMITING SAFETY SYSTEM SET~ iNGS

BASES

Loss of Turbine

A Loss of Turbine trip causes a direct, reactor trip when operating above15~ of RATED THERMAL POMER. This +rip provides turbine protection, reducesthe severity of the ensuing transient and helps avoid the lifting of the mainsteam line safety valves during the ensuing transient, thus extending theservice life of these valves. No credit was taken in the accident analysesfor operation of this trip. Its functional capability at the specified tripsetting 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 coreduring startup operations and its use serves as a backup to the. administra-tively enforced startup rate limit. ~s-t~-se+po+n~ees-net-co~and

d

eP - . - '~h~eei+MI 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 ofthis trip function prccludcd the need for specific

analyses ofother events initiated from subcritical conditions.

C«

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

C

J

4 ~

ptWee~ '0 'h

4

«t4~0'

fl

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 continueboration at > 40 gpm of 1720 ppm boron or equivalent until the requiredSHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS

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

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Ci

d.

Mithin one hour after detection of an inoperable CEA(s) and atleast once per 12 hours thereafter while the 'CEA(s) is inoperabl .+oy+If the ino rable CEA is immovabl or untri able the aboverequired SHUTDOWN MARGIN shall be increased by an amount at.least equal to the withdrawn worth of the imnovable or un-trippable CEA{s). ca ~ yesw 4 ef egcessNe,4 te4ioe oL"

eLachmv6eoL ivL+zy emecaMhen in MODES 1 or 2, at least once per y vthat CEA group withdrawal is within the Power t}ependentInsertion Limits of Specification 3.1.3.6.

When in MODE 2, at least once during CEA withdrawal and atleast once per hour thereafter until the reactor is critical.Prior to initial operation above 5X RATED THERMAL POWER aftereach fuel loading, by consideration of the factors of ebelow, with the CEA groups at the Power Dependent InsertionLimits of Specification 3.1.3.6.

~ (~ I,

See Special Test Exception 3.10.1.Mith K > 1.0.kith K ff < 1.0.

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

)

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>< /f01 f'Pl O'afe P95xs ~ pf~zsv m p~>l >~f4>-Isq t>t Jr/ JaavcorJ>-'4A~ E + +f

REACTIVITY CONTROL SYSTEMS

BORON DILUTION

LIMITING CONDITION FOR'PERATION

I

3.1;1.3 The flow rate of reactor coolant to the reactor pressure vesselshall be > 3000 gpm whenever a reduction in Reactor Coolant System boronconcen 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 inboron concentration of the Reactor Coolant System.

SURVEILLANCE REQUIREMENTS

4.1.1.3 The flow rate of reactor coolant to the reactor pressure vesselshall be determined to be > 3000 gpm within one hour prior to the startof and at least once per hour during a reduction in the Reactor CoolantSystem 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 pumpis in operation and supplying > 3000 gpm to the reactor pressurevessel.

ST. LUCIE - UNIT 1 3/4 1-4 gggeAIe(n& ~e~

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1

3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLEand capable of being powered from an OPERABLE emergency power source.

a. A fiowpath from the boric acid makeup tank via either a boric acid pump or agravity feed connection and any'charging pump to the Reactor Coolant Systemifonly 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 ahigh pressure safety injection pump to the Reactor Coolant System if only therefueling water tank in Specification 3.1.2.7b is OPERABLE.

MODES 5 and 6.

AQllQH:

With none of the above flow paths OPERABLE, suspend all operations involving COREALTERATIONSor positive reactivity changes until at least ona injection path is restored toOPERABLE 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 operatedor automatic) in the flow path that is not locked, sealed, or othetwise secured inposition, 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. Inthelatter case: 1) all charging pumps shall be disablecf; 2) heatup and cooldown rates shall belimited in accordance with Figure 3.1-1 b; and 3) at RCS temperatures below 115'F, any twoof the following valves in the operable HPSI header shall be verified closed and have theirpower removed:

HCV-3616

HCV-3626

HCV-3636

HCV-3646

HCV<617

HCV-3627

HCV<637

HCV-3647

pcsP"'dW~+

7~y I$$$

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

t~ ~ ~

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

MODES 5 and 6.

AQI1QH:

With no charging pump or high pressure safety injection pump OPERABLE, suspend alloperations involving CORE ALTERATIONSor positive reactivity changes until at least one of therequired pumps is restored to OPERABLE status.

4.1.2,3 At least one of the above required pumps shall be demonstrated OPERABLE byverifying the charging pump develops a flow rate of greater than or equal to 40 gpmor the high pressure safety injection pump develops a total head of greater than orequal 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 thelatter case: 1) all charging pumps shall be disabled; 2) heatup and cooldown rates shall belimited in accordance with Figure 3.1-1 b; and 3) at RCS temperatures below 115'F, any twoof the following valves in the operable HPSI header shall be verified closed and have theirpower removed:

HCV<616

HCV-3626

HCV-3636

HCV-3646

HCV-3617

HCV-3627

HCV-3637

HCV-3647

y$gf~

~gg cL

ST. LUCIE- UNIT1 3f4 1-12 Amendment No. 66, e+, 96,+94, +%,~Q

~"

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t<1 +'$aj „~f p"gQ~Q g gg

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EACTIVITYCO YSTEMS

FULL LENGTH CEA POSWON continued

LIMmNG CONDITION FOR OPERATION continued

2. Declared inoperable and satisfy SHUTDOWN MARGIN requirements ofSpeciTication 3.1.1.1. After declaring the CEA inoperable, operation inMODES 1 and 2 may continue pursuant to the requirements ofSpeciTication 3.1.3,6 for up to 7 days per occurrence with a totalaccumulated time of < 14 days per calendar year provided all of thefollowing conditions are met:

a) Within 1 hour, the remainder of the CEAs in the group with theinoperable CEA shall be aligned to within 7.5 inches of theinoperable CEA while maintaining the allowable CEA sequenceand insertion limits shown on COLR Figure 3.1-2; the THERMALPOWER level shall be restricted pursuant to Specification 3.1.3,6during subsequent operation. tv ten"- rt'ed .

b) The SHUTDOWN MARGIN requiremen f Specification 3.1.1.1 isdetermined at least once per 12 hours the>wise-,be-ii-t-aHeast0&7-STANBBY-eithi

OQMftvLQc ll< eely+ Heft $TAPQsf b4'L444A 4'AW ~ ~<8e. 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 themisaligned CEA is positioned within 7.5 inches of other CEAs in its group inaccordance 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 15or more inches beyond the time constraints shown in COLR Figure 3.1-1a,reduce power to s 70% of RATED THERMALPOWER prior to completingACTION f.1 or f.2.

. 2

Restore the CEA to OPERABLE status within its specified alignmentrequirements, or

Declare the CEA inoperable and satisfy the SHUTDOWN MARGINrequirements of Specification 3.1.1.1. After declaring the CEAinoperable, operation in MODES 1 and 2 may continue pursuant to therequirements of Specification 3,1.3.6 provided:

a) Within 1 hour the remainder of the CEAs in the group with theinoperable CEA shall be aligned to within 7.5 inches of theinoperable CEA while maintaining the allowable CEA sequenceand insertion limits shown on COLR Figure 3,1-2; the THERMALPOWER level shall be restricted pursuant to Specification 3.1.3.6during subsequent operation.

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

~ ~

» ~ I

7 I tli&i

EACTIVITYCONTR Y TEMS

L LENG CEA POSWON continued

LIMmNG CONDITION FOR OPERATION continued

b) The SHUTDOWN MARGlN requirement of Specification 3.1.1.1 isdetermined lea once ours

<<~<~p M i4 W l+LS4 HOT PfAHO$ lutkki44ae. KcPc+ ~ ~M.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 HOTSTANDBY within 6 hours.

h. With one full-length CEA inoperable due to causes other than addressed byACTION a above, and inserted beyond the long term steady state insertionlimits but within its above specified alignment requirements, operation inMODES 1 and 2 may continue pursuant to the requirements of Specification3.1.3.6.


4.1.3.1.1

4.1.3.1.2

4.1.3.1,3

4.1.3.1.4

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 hoursexcept during time intervals when the Deviation Circuit and/or CEA Block Circuitare inoperable, then verify the individual CEA positions at least once per 4 hours.

Each full-length CEA not fully inserted shall be determined to be OPERABLE byinserting it at least 7.5 inches at least once per 92 days.

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

The CEA Block Circuit shall be demonstrated OPERABLE by a functional testwhich verifies that the circuit maintains the CEA group overlap and sequencingrequirements of Specmcation 3.1.3.6 and that the circuit prevents the regulatingCEAs from being inserted beyond the Power Dependent Insertion Limit of COLRFigure 3.1-2:

'a. Prior to each entry into MODE 2 from MODE 3, except that such verificationneed 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 entryinto 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,~90

~,4

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POWE DIST IB ON LIMITS

TAL NTEGRATED R I L P A ING F CTOR - FiLIMITINGCONDITION 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 hours either:

a. Be in at least HOT STANDBY, or

b. Reduce THERMAL POWER to bring the combination of THERMALPOWERand F, to within the limits of COLR Figure 3.2-3 and withdraw the fulllength CEAs to or beyond the Long Term Steady State Insertion Limits ofSpecification 3.1.3.6. The THERMALPOWER limit determined from COLRFigure 3.2-3 shall then be used to establish a revised upper THERMALPOWER level limit on COLR Figure 3Z-4 (truncate Figure 3.2A at theallowable fraction of RATED THERMALPOWER determined by COLRFigure 3.2-3) and subsequent operation shall be maintained within thereduced acceptable operation region of COLR Figure 3.2A.


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 calculatedwith a non-full core power distribution analysis code and shall be calculated asF, = F, when calculations are performed with a full core power distributionanalysis code. F, shall be determined to be within its limit at the followingintervals.

a. Prior to operation above 70 percent of RATED THERMALPOWER aftereach 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

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

(~

I

REACTOR COOLANT SYSTEM

3/4.4.9 PRESSURE TEMPERATURE LIMITS



3.4.9.1 The Reactor Coolant System {except the pressurizer}, temperature andpressure shall be limited in accordance with the limit lines shown on 'Figures3.4-2a, 3.4-2b"and 3.4-3 during heatup, cooldown, criticality, and inservice., Lleak and hydrostatic testing.

APPLICABILITY: At all times.*f

ACTION:

Mith any of the above limits exceeded, restore the temperature and/orpressure to within the limits within 30 minutes; perform an analysis todetermine the effects of the out-of-limit condition on the fracturetoughness properties of the Reactor Coolant System; determine that theReactor Coolant System remains acceptable for continued operations or bein at least HOT STANDBY within the next 6 hours and reduce the RCS Tavto less than 200'F within .the'following 30'hours in accordance with'Figures 3.4-2b and 3.4-3.

~ ( g ) 'F ++g ~5'gravure. bound~ry skk@rdy

mrs'hen

the flo ath from the RMT to the RCS via a single HPSI pump isestablished pe .l.2. the heatup and cooldown rates shall be establishedin accordance with 'Fig. 3.$ -1b.

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

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

4

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, CONTAINMENTSYST

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.


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

4y

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CONTAINMENT SYSTEMS

3/4.6.5 VACUUM RELIFF VALVES


3.6.5.1 The containment vessel to annulus vacuum relief valves shallE OPERABLE E E E P E P~2.25 + 0.25 inches Mater Gauge differential.

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 hours or be in at leastHOT STANDBY within the next 6 hours and in COLD SHUTDOMN within thefollowing 30 hours.


4.6.5.1 No additional Surveillance Requirements other than those requiredby Specification 4.0.5 and at least once per 3 years verify that thevacuum relief valves open fully within 8 seconds at 2.25 + 0.25 inchesMater Gauge differ.'ential.

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

0 l~

0

J

f

Ej

4, e

REFUELING OPERATIONS

CONTAINMENT PENETRAT IONS


3.9.4 The containment penetrations shall be in the following status:.

a. The equipment door closed and held in place by a minimum offour 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 containmentatmosphere to the outside atmosphere shall be either:

1. Closed by an isolation vaive, blind flan e r nuavalve, 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<~econtainment isolation valve, or ~~not,

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

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

ACTION:

With the requirements of the above specification not satisfied, immedi-ately suspend all operations involving CORE ALTERATIONS or movement ofirradiated fuel in the containment. The provisions of Specification3.0.3 are not applicable.

SURVE ILLANCE REQUIREMENTS

4.9.4 Each of the above required containment penetrations shall bedetermined to be either in its closed/isolated condition or capable ofbeing closed by an OPERABLE automatic containment isolation valve within72 hours 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/isolatedcondition, or

b. Testing the containment isolation valves per the applicableportions 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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~+ 4i yaH z) rir~Sm


MANIPULATOR CRANE OPERABILITY


3.9.6 The manipulator crane shall be used for movement of CEAs or fuelassemblies 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 thereactor pressure vessel.

ACTEON:

Mith the requirements for crane OPERABILITY not satisfied, suspenduse of any inoperable manipulator crane from operations involving themovement of CEAs and fuel assemblies within the reactor pressure vessel.W~t'ov


4.9.5 The manipulator crane used for movement of CEAs or fuel assem-blies within the reactor pressure. vessel shall be demonstrated OPERABLEwithin 72 hours prior to the start of such operations by performing a

load test of at least 2500 pounds and demonstrating an automatic l.oadcut off~ the crane load exceeds 3000 pounds.

'I

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

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3.10.2

hQX1QH'he

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 theperformanc ofP YSIC TESTSprovided:g~a. The THERMALPOWER 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 inSpecification 4.10Z2 below.

MODES 1 and 2.

With any of the limits of Specification 3.2.1 being exceeded while the requirements ofSpeciTications 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 ofSpecification 3.2.1, or

b. Be in HOT STANDBYwithin 6 hours.

4.10.2.1

4.10.2.2

The THERMAI POWER shall be determined at least once per hour duringPHYSICS 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 thetest power plateau.

The linear heat rate shall be determined to be within the limits of Specification 3Z.1by monitoring it continuously with the Incore Detector Monitoring System pursuant tothe requirements of Specifications 4.2.1.4 during PHYSICS TESTS above 5% ofRATED 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, ~,+%+6

RADIOACTIVE FF U S

EXPLOS V M R


3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall belimited 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 greaterthan 2% by volume but less than or equal to 4X by volume, reduce theoxygen concentration to the above limits within 48 hours.

b. kith the concentration of oxygen in the waste gas decay tank greaterthan 4% by volume and the hydrogen concentration greater than 2X byvolume, immediately suspend all additions of waste gases to thesystem and immediately caamence reduction of the concentration ofoxygen 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


4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall bedetermined to be within the above limits by continuously monitoring the wastegases in the on service waste gas decay tank ~~h gen-aeter ~ed+PERAB~~N ~&~m&6pecHieat-ion-8-.3~.4.11.2.5.2 Mith the oxygen concentration in the on service waste gas decaytank greater than 2X by volume as determined by Specification 4.11.2.5.1, theconcentration of hydrogen in the waste gas decay tank shall be determined tobe within the above limits by gas partitioner sample at least once per24 hours.

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'nd 6 requires either 3650 gallons of 2.5 to 3.5 weight percent boricsolution (4371 to 6119 ppm boron) from the boric acid tanks or 11,900of 1720 yam borated water from the refueling water tank to makeup forof the primary coolant that could occur if the temperature is lo«ered200'F to 140'F.,

NODES 5acidgallonscontractionfrom

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 caseof any inadvertent pressure transient due to a mass addition to the RCS.

3/4.1. 3 NOVABLE CONTROL ASSEMBLIES

The specificaidistribution limitsmaintained, and (3)limited to acceptah

IfRCS pressure boundary integrity docs not exist as dcfincd in Specification 1.16,thcsc restrictions are not required. Additionally, a limiton thc maximum number ofopcrablc HPSI pumps is not ncccssaiy when thc prcssurizcr manway cover or the

Th ACTION t reactor vessel head is removed.

requSr ementS are aC „... ~ - IVIIQI I"Cbbl Ilail.IVII'CClllrllCll>UC't 4IIQlethe original criteria are met.

The ACTION statements applicable to an immovable or untrippable CEAand to a large misalignment (» 15 ihches) of two or more CEAs, require aprompt shutdown of the reactor since either of these conditions may beindicative of a possible loss of mechanical functional capability of theCEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOMNNARGINe

For small misalignments (< 15 inches} of the CEAs, there is 1) a smalldegradation in the peaking factors relative to those assumed. in generatingLCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect-onthe time dependent long term power distributions relative to those used ingenerating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a smalleffect on, the available SHUTDOMN MARGIN, and 4) a small effect on the ejectedCEA worth used in the safety analysis. Therefore, the ACTION statementassociated with the small misalignment of a CEA permits a one hour timeinterval during which attempts may be made to r estore the CEA to within itsalignment requirements prior to initiating a reduction in THERMAL POMER. Theone hour time limit is sufficient to (1) identify causes of a misaligned CEA,(2) take appropriate corrective action to realign the CEAs, and (3) minimizethe effects of xenon redistribution.

Overpower margin is provided to protect the core in the event of a largemisalignment (> 15 inches) of a CEA. However, this. misalignment would causeistortion 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 safetyanalysis. Therefore, the ACTION statement associated with the large mis-

Lalignment 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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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 toprevent RCS overpressurization above the 10 CF Appendix G operating limitcurves (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 useof the pressurizer power-operated relief valve (PORVs) and .the implementationof administrative and operational controls.

The PORYs aligned to the RCS with the low pressure setpoints of 350 and 530psia, 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 basisoverpressurization events due to mass or energy addition to the RCS. The LTOPsystem APPLICABILITY, ACTIONS, and SURVEILLANCE REQUIREMENTS are consistentwith the resolution of Generic Issue 94, "Additional Low-TemperatureOverpressure Protection for Light-Mater Reactors," pursuant to Generic Letter90-06.

3 4.4.15 REACTOR COOLANT SYSTEM VENTS

Reactor Coolant System vents are provided to exhaust noncondensible gasesand/or steam from the primary system that could inhibit natural circulationcore cooling. The OPERABILITY of at least one Reactor Coolant System ventpath from the reactor vessel head and the pressurizer steam space ensures thecapability exists to perform this function.

The redundancy design of the Reactor Coolant System vent systems serves tominimize the probability of inadvertent or irreversible actuation whileensuring that a single failure of a vent valve, power supply, or controlsystem does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor CoolantSystem vent system are consistent with the requirements of Item II.b.l ofNUREG-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 remainsubcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintainedfor reactivity control in the water volumes having direct access to the reactor vessel. Thelimitation on K,„ is sufficient to prevent reactor criticality with all full length rods (shutdown andregulating) fully withdrawn.

3/4.9.2 INSTRUMENTATION

The OPERABILITYof the wide range logarithmic range neutron flux monitors ensures thatredundant monitoring capability is available to detect changes in the reactivity condition of thecore.

3/4.9.3 DECAY TIME

The minimum requirement for reactor subcnticality prior to movement of irradiated fuelassemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow theradioactive decay of the short lived fission products. This decay time is consistent with theassumptions used in the accident analyses.

3/4.9.4 CONTAINMENTPENETRATIONS

The requirements on containment penetration closure and OPERABILITYensure that arelease of radioactive material within containment will be restricted from leakage to theenvironment. The OPERABILITYand closure restrictions are sufficient to restrict radioactivematerial release from a fuel element rupture based upon the lack of containmentpressunzation potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS

The requirement for communications capability ensures that refueling station personnel can bepromptly informed of significant changes in the facility status or core reactivity condition duringCORE ALTERATIONS.

3/4.9.6 MANIPULATORCRANE OPERABIUTY

The OPERABILITYrequirements of the cranes used for movement of fuel assemblies ensuresthat: 1) each crane has sufficient load capacity to lift a fuel element, and 2) the core internalsand pressure vessel are protected from excessive lifting force in the event they areinadvertently engaged during lifting operations.

In accordance with Generic Letter 91-08, Removal ofComponent Lists from the Tcchnical Specifications, the'peningof locked or sealed closed containment isolation valves on an intermittent basis under administrative

control includes the followingconsiderations: (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 ofradioactivity outside thc containment.

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

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DESIGN FEATURES

2.1 2

b.

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SHIELD B

Minimum annular space = 4 feet.Annulus nominal volume = 543,000 cubic feet.Nominal outside height (measured from top of foundation haseto the top of the dome) = 230.5 feet.Nominal inside diameter = 148 feet.Cylinder wall minimum thickness 3 feet.Dome minimum thickness = 2.5 feet.Dome inside radius - 112 feet.

DESIGN PRESSURE AND T ERATUR

5.2.2 The containment vessel is designed and shall he maintained for amaximum 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 allowancefor nc"mal degradation pursuant to the applicable SurveillanceRequirements.

FUEL ASSEMBLI

5.3.1 The reactor core shall contain 217 fuel assemblies with eachfuel assembly containing a maximum of 176 fuel rods clad withZircaloy-4. Each fuel rod shall have a nominal active fuel length ofbetween 134.1 and 136.7 inches. Individual fuel assemblies shallcontain fuel rods of the same nominal active fuel length. Fuelassemblies shall be limited to those designs that have been analyzedusing NRC approved methodology and shown by tests or analyses to complywith fuel design and safety criteria. The initial core loading shallhave a maximum enrichment of 2.83 weight percent U-235. Reload fuelshall be similar in physical design to the initial core loading-5.3. Except for special test as authorised by the NRC, all fuelassemb ies under control element assemblies shall be sleeved with asleeve esign previously approved hy the NRC.

ST LUCIE - UNIT 1 5-4 Amendment No.

St. Lucie Unit 1 and Unit 2Docket Nos. 50-335 and 50-389Proposed License Amendments

ST ~ LUCIE UNIT 2 MARKED-UP TECHNICAL SPECIFICATION PAGES

PagePagePagePagePagePagePagePagePagePagePagePagePagePagePagePagePage

XVIIIXIXB 2-1B 2-4B 2-5B 2-63/4 1-13/4 1-193/4 2-143/4 3-263/4 6-33/4 6-213/4 6-263/4 8-73/4 9-63/4 11-14B 3/4 2-2

I

ADMINISTRATIVE CONTROLS

NO EX

0'ECTION

6. 1 RESPONSIB ILITY

PAGE

6-1

6. 2 ORGANIZA ION.''.....

6.2. 1 ONSITE AND OFFSITE ORGANIZATIOK..

6.2.2 UNIT STAFF.......

6-1

6-2

BEPENB

FUN

COMPOSITION............RESPONS IBILITIES..A

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6-6

6-66-6

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

6o 3 UNIT STAFF UALIFICATIONS

6-6

6-6

6. 4 TRAINING.. 6-7

6.5 REVIEW AND AUDIT.

6.5. 1 FACILITY REVIEW GROUP

FUNCTION. \ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

COMPOSITION..

LTERNATESo o ~ o o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~

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MEETING FREQUENCY.

QUORUM.;

RESPONSIBILITIES.

AUTHORITY

RECORDS

6-7

6-76-76-76-76-86-86-86-96-9

ST. LUCIE - UNIT 2 Amendnent No.~

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ADMINISTRATIVE CONTROLS

INDEX

SECTION

6.5.2 COMPANY NUCLEAR REVIEM BOARD...

PAGE

0 ~ ~ ~ ~ 6 9

FUNCTION ...........-....-----COMPOSITION......................

~ ~ ~ ~ e ~ o 6 9

6-10

6-10

6-10

6-10

6-10

6-11

6-11

6-12

-14 -IN.

6-13

6-13

6-13

6-16

6-16

6-16

ALTERNATES..............CONSULTANTS........

MEETING FREQUENCY........

QUORUM............REVIEM.

AUDITS............ ~ ~ ~ t I ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

AUTHORITY.............RECORDS..

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6. 7 SAFETY LIMIT VIOLATION....6.8 PROCEDURES AND PROGRAMS..

6.9 REPORTING RE UIREHENTS .

6.9.1 ROUTINE REPORTS.......STARTUP REPORT.............................................

6.6 REPORTABLE EVENT ACTION.....................................

ANNUAL REPORTS...

MONTHLY OPERATING REPORTS.

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT...

6-16

6-17

6-18

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 fuelcladding and possible cladding perforation which would result in the releaseof fission products to the reactor coolant. Overheating of the fuel isprevented by maintaining the steady-state peak linear heat rate below thelevel at which centerline fuel melting will occur. Overheating of the fuelcladding is prevented by restricting fuel operation to within the nucleateboiling regime where the heat transfer coefficient is large and the claddingsurface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime couldresult in excessive cladding temperatures because of the onset of departurefrom nucleate boiling {ONB) and the resultant sharp reduction in heat transfercoefficient. ONB is not a directly measurable parameter during operation andtherefore THERMAL POMER and Reactor Coolant Temperature and Pressure have beenrelated to ONB through the CE-I correlation. The CE-I DNB correlation hasbeen developed to predict the DNB heat flux and the location of DNB foraxially uniform and non-uniform heat flux distributions. The local DNB heatflux ratio, DNBR, defined as the ratio of the heat flux that would cause DNBat a particular core location to the local heat flux, is indicative of themargin to DNB.

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

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

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

The Thermal Margin/Low Pressure and Local Power Density Trip Systems, inconjunction with Limiting Conditions for Operation, the Variable OverpowerTrip and the Power Dependent Insertion Limits, assure that the SpecifiedAcceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceededduring normal operation and design basis Anticipated Operational Occurrences.

ST. LUCIE - UNIT. 2 B 2-1 Amendment No. 8—4SR

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 coreduring rapid positive reactivity addition excursions which are too rapid to beprotected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure Trip.

The Variable Power Level High trip setpoint is operator adjustable and can beset no higher than 9.61K above the indicated THERMAL POWER level. Operator actionis required to increase the trip setpoint as THERMAL POWER is increased. Thetrip setpoint is automatically decreased as THERMAL POWER decreases. The tripsetpoint has a maximum value of 107.0X of RATED THERMAL POWER and a minimumsetpoint of 15.0X of RATED THERMAL POWER. Adding to this maximum v'alue thepossible variation in trip point due to calibration and instrument errors, themaximum actual steady-state THERMAL POWER level at which a trip would be actuatedis 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 safetyvalves and main steam line safety valves, provides Reactor Coolant System pro-tection against overpressurization in the event of loss of load without reactortrip. This trip's setpoint is at less than or equal to 2375 psia which is belowthe nominal lift setting 2500 psia of the pressurizer safety valves and itsoperation 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 theDNBR is less than the DNB-SAFDL of . , in conjunction with ESCU methodology.

B

The trip is initiated whenever the Reactor Coolant System pressure signaldrops below either 1900 psia or a computed value as described below, whicheveris higher. The computed value is a function of the higher of hT power or neutronpower, reactor inlet temperature, the number of reactor coolant pumps operatingand the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, themaximum 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 isassumed. Finally, the maximum insertion of CEA banks which can occur during anyanticipated operational occurrence prior to a Power Level-High trip is assumed.

o, The Thermal Margin/Low Pressure trip setpoints are derived from the coresafety limits through application of appropriate allowances for equipmenttits* dp

include: all owance . ERMA~eWER to compen-sa e or poten ial power m asurement error; an allowance WM+V-F to compensatefor potential temperature easurement uncertainty; and Mw4her allowance of~

to compensate for p essure measurement error end time delay associated withproviding effective te nation of the occurrence that exhibits the most rapiddecrease in margin to t e safety limit. wance-i~nade-u~~55-p~-pressure-measu men~Aewance-an e-dA'vy-a-Howanee.

Qa a owns ;~nl~yppg e ~08r Je+Hdenf/ 'adi A)~~BlC.

ST. LUCIE — UNIT 2 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 tripis initiated prior to or concurrently with a safety i+ection (SIAS). Thisalso provides assurance that a reactor trip is initiated prior to orconcurrently with an MSIS.

Steam Generator Pressure-Low

The Steam Generator Pressure-Low trip provides protection against anexcessive rate of heat extraction from the steam generators and subsequentcooldown of the reactor coolant. Yhe setpoint of 620 psia is sufficientlybelow the full load operating point of approximately 885 psia so as not tointerfere with normal operation, but still high enough to provide the requiredprotection in the event of excessively high steam flow. This setting was usedwith 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 offeedwater Row incident and assures that the design pressure of the ReactorCoolant System will not be exceeded due to loss of the steam generator heatsink. This specified setpoint provides allowance that there will be sufficientwater inventory in the steam generator at the time of the trip to provide ~

C bA ~d~ ''. Ifitrip also protects against violation of the specified acceptable fuel designlimits (SAFDL) for ONBR, offsite dose and the loss of shutdown margin forasymmetric steam generator transients such as the opening of a main steamsafety valve or atmospheric dump valve.

Local Power Densit -Hi h

sufficient time for any operator action to initiate auxiliary feedwatcr.'The Local Power Densi: bcfore reactor coolant system subcooling is lost.""

monitoring, is provided to <

fuel which corresponds to fuei cenxertine meting will not occur as aconsequence of axial power maldistributions. A reactor trip is initiatedwhenever 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 neutrondetector channels. The calculated setpoints are generated as a function ofTHERMAL POMER level with the allowed CEA group positiion being inferred fromthe THERMAL POMER level. The trip is automatically,.bypassed below 15% power.

The maximum A1IMUTHAL POMER TIL'T and maximum CEA misalignment permittedfor continuous operation are assumed in generation of the setpoints. Inaddition, CEA group sequencing in accordance with the Specifications 3.1.3.5and 3.1.3.S is assumed. Finally, the maximum insertion of CEA banks which canoccur during any anticipated operational occurrence prior to a Power Level-Hightrip is 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 adelayed reactor trip. This trip provides protection to the reactor coolantpumps by ensuring that plant operation is not continued without cooling wateravailable. The trip is delayed 10 minutes following a reduction in flow tobelow the trip setpoint and the trip does not occur if flow is restored before10 minutes elapses. No credit was taken for this trip in the safety analysis.Its functional capability at the specified trip setting is required to enhancethe 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~4-Sary-Limi4-and-no-credit-was-taken-~he-safety-analyses-for-opet ~~

W ~ p&H ~ N p f H ~ ttLl~y- f .h~ I 'y~.Reactor Coolant Flow - Low

The Reactor Coolant Flow - Low trip provides core protection againstONB in the event of a sudden significant decrease in RCS flow. The reactortrip setpoint on low RCS flow is calculated by a relationship between steamgenerator differential pressure, core inlet temperature, instrument errorsand response times. Nen the calculated RCS flow falls below the tripsetpoint an automatic reactor trip signal is initiated. The trip setpointand allowable values ensure that for a degradation of RCS flow resultingfrom expected transients, a reactor trip occurs to prevent violation of localpower 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 specificanalyses ofother 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


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. an6continue 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 therequired SHllTDQS MARGIN is restored.


4.1.1.1.1 The SHUTOOMN MARGIN shall be determined to be greater than or equalto 5000 pcm:

<.w +lipolS~ll'~~~~+ CS

b.

Mithin one hour after detection of an inoperable CEA(s) and at leastonce per 12 hours thereafter while the CEA(s) is inoperable. If thino erable CEA is immovable or untri able "he above re uired i5~+SHUTDOWN MARGIN shall be erified acceptable with an increased ' +all~ce for the withdrawn rth of the imeo le or untrippableCEA(s). mc ca.1~ f'of+cesslvc. ra'c&en 4Y

naeCA~NAi~ In&~~ren~When in MODE 1 or MODE 2 with K f greater than or equal to 1.0, atleast once per 12 hours by verifyfng that CEA group withdrawal iswithin the Power Dependent Insertion Limits of Specification 3. 1.3.6.

c. When in MODE 2 with K less than 1.0, within 4 hours prior toachieving reactor crifQality by verifying that the predictedcritical 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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Mith one full-'engtn CEA misaligned rom any otiier CEA .:n its g. ouvby more than 15 inches beyond the time constraints shown '„"..".gure3.l-la, reduce power to ( 70. of RATED THERl@L POWER prior tocompleting ACT.ON e.l or e.2. I

I

1. Restore the CEA to OPERABLE status within its specifiedalignment 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 therequirements of Specification 3. 1.3.6 provided:*

1

a) Mithin 1 hour the remainde. of the CEAs in the group with theinoperable CEA shall be aligned to within 7.0 inches of theinoperable CEA while maintaining the allowable CEA sequenceand insertion limits shown on Figure 3.1-2; the THEPJCL POMER

level shall be restricted pursuant to Specification 3.1.3.6during subsequent operation.

b) The SHUTDOWN MARGIN requirement o Specification 3.1.1.1 is/O tt'wtsa> i> at eosf gpss ~i~~a~e. Next 4 ho~vs

0 e e . er iEAs inits 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 ] hourthe 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 ofSpecification 3.1.3.6 provided:a) Mithin 1 hour the remainder of the CEAs in the group with the

inoperable CEA shall be aligned to within 7.9 inches of theinoperable CEA while maintaining the allowable CEA sequenceand insertion limits shown on Figure 3.1-2; the THER'~PL PO'MER

level shall be restric ed pursuant to Speci ication 3.1.3.6during subsequent operation.

b) The SHUTDOMN MARGIN requirerent of Specificat on 3.1.1.1 isdetermined at least once per 12 hours.

Otherwise, be in at least HOT STANDBY wi hin 6 hours. +a >+~

Mith one full-length CEA inoperable due to causes other than addressedby ACTION a., above, and inserted beyond the Long Term Steady StateInsertion Limits bu+ within its above specified alignment requiren-nts,operation in MODES 1 and 2 may continue pursuant to the requirements ofSpecification 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


3.2.5 The following DNB-related parameters shall be maintained within thelimits 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 parameterto within its limit within 2 hours or reduce THERMAL POMER to < SX of RATEDTHERMAL POMER within the next 4 hours.


4.2.5.1 Each of the parameters of Table 3.2-2 shall be verified to be withintheir limits by instrument readout at least once per 12 hours.

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

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PROCESS MONITORS (Continued)

c. Noble Gas EffluentMonitors

T~AB .6 (c ti di

ADI TION 0 ITO G I IO

HINIHUH.CHANNELS APPLICABLE ALARH/TRIP MEASUREMENTOPllilL~ILE ~D> .KPiUiiL

i. Reactor AuxiliaryBuilding ExhaustSystem (Plant VentLow Range Honitor)

ii. Reactor AuxiliaryBuilding Exhaust Sys-tem (Plant Vent HighRange Honitor)

Steam GeneratorBlowdown TreatmentFacility BuildingExhaust System

iv. Steam Safety ValveDischarge¹

1,2,38(4

1,2,38(4

I, 2, 38(4

1/steam 1, 2, 3 8 4header

10 - 10 QCi/cc

10 - 10 pCi/cc

10 - 10 ~ pCi/cc

10 — 10 QCi/cc

27

27

27

27

vo Atmospheric Steam .

Dump Valve Dis-charge¹

I/steam I, 2, 3 8 4header

10 — 10 pCi/cc 27

I vi. ECCS Exhaust 1,2,38(4

e arm r p etpoints are determined and set in accordance withCalculation Hanual.

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

10 - 10 QCi/cc 27

the requirements of the Offsite Dose

dump valve discharge

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CONTAINMENTSYSTEMS

SURVEILLANCE REQUIREMENTS continued

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


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


3.6.5 The primary containment vessel to annulus vacuum relief valves shall beOPERABLE with an actuation setpoint of Ress-the ~u&-ho 9.85 t0.35 inches water gauge.

A~PPL CAB IV: MODES 1, 2. 3 d 4.

ACTION:

With one primary containment vessel to annulus vacuum rel ief valve inoperable,restore the valve to OPERABLE status within 4 hours or be in at least HOTSTANDBY within the next 6 hours and in COLD SHUTDOMN within the following 30hours.


4.6.5 No additional Surveillance Requirements other than those required bySpecification 4.0.5.

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

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7.

c) Verifying that all automatic diesel generator trips, except engineoverspeed and generator differential, are automatically bypassedupon loss of voltage on the emergency bus concurrent with a safetyinjection actuation signal.

Verifying the diesel generator operates for at least 24 hours,During the first 2 hours of this test, the diesel generator shall be loadedwithin a load band of 3800 to 3985 kW'nd during the remaining 22 hoursof this test, the diesel generator shall be loaded within a load band of3450 to 3685 kW'. The generator voltage and frequency shall be4160 x 420 volts and 60 ~ 12 Hz within 10 seconds after the start signal;the steadywtate generator voltage and frequency shall be maintainedwithin these limits during this test.

8. Verifying that the aut~nnected loads to each diesel generator do notexceed 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 isloaded with its emergency loads upon a simulated restoration ofoffsite 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 testmode 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 storagetank to the engin~ounted tanks of each diesel via the installed crossconnection lines.

¹ This band is meant as guidance to avoid routine overloading of the engine. Variations inload 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 recommendationsconcerning engine prelube period.

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

4

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3/4. 9. 6 MANIPULATOR CRANE


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 anyinoperable manipulator crane from operations involving the movement of CEAsand fuel assemblies within the reactor pressure vessel.


4.9.6 The manipulator crane used for movement of fuel assemblies, with orwithout CEAs, within the reactor pressure vessel shall be demonstratedOPERABLE within 72 hours prior to the start of such operations by performing aload 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


3.11.2.5 The concentration of oxygen in the waste gas decay tanks shall belimited to less than or equal to 2X by volume whenever the hydrogenconcentration exceeds 4X by volume.

APPLICABILITY: At all times.

ACTION:

b.

C.

Mith the concentration of oxygen in the waste gas decay tank greaterthan 2X by volume but less than or equal to 4X by volume, reduce theoxygen concentration to the above limits within 48 hours.

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

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREHENTS

4.11.2.5.1 The concentration of oxygen in the waste gas decay tank shall bedetermined to be within the above limits by continuously monitoring the waste

t * t t t t Attt~~ t '

wpERAB t H t4.11.2.5.2 Mith the oxygen concentration in the on service waste gas decay tankgreater 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 withinthe above limits by gas partitioner sample at least once per 24 hours.

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/LowPressure 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 toidentify the cause of this unexpected tilt.

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

The Surveillance Requirements for verifying that F , F'„ and T are withintheir limits provide assurance that the actual values oF F , F an) T do notexceed the assumed values. Yerifying F and F„ after each fuel loading priorto exceeding 75X of RATED THERMAL POWER provides additional assurance that thecore was properly loaded.

3 4.2.5 DNB PARAMETERS.

The limits on the DNB-related parameters assure that each of theparameters are maintained within the normal steady-state envelope of operationassumed in the transient and safety analyses. The limits are consistent withthe safety analyses assumptions and have b n analytically demonstratedadequate to maintain a minimum DNBR of > . in conjunctio th ESCU

methodology throughout each analyzed transient. I.28The 12-hour periodic surveillance of these parameters through instrument

readout is sufficient to ensure that the parameters are restored within theirlimits following load changes and other expected transient operation. The18-month periodic measurement of the RCS total flow rate is adequate to detectflow degradation and ensure correlation of the flow indication channels withmeasured flow such that the indicated percent flow will provide sufficientverification of flow rate on a 12-hour basis.

ST. LUCIE — UNIT 2 B 3/4 2-2 Amendment No. 4-RC—I-et-t

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Documents

Proposed tech specs incorporating administrative changes ...leastIf once per 12 hours thereafter while the 'CEA(s) is inoperabl.+oy+ the ino rable CEA is immovabl or untri able the