attachment st. unit technical specifications · marked-up st. lucie unit 1 technical specifications...

ATTACHMENT 1

Marked-up St. Lucie Unit 1 Technical Specifications page:

XXI of the Index3/4 0-33/4 1-123/4 3-413/4 4-1b3/4 4-83/4 4-143/4 6-53/4 6-203/4 6-213/4 8-6b3/4 11-23/4 11-103/4 12-10B 3/4 1-4B 3/4 7-5B 3/4 9-3

6-24

92082b0009 920821PDR ADOCK 05000335P,

PDR

ile ~

1

I~ ~ INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMEHTS

SECTION PAGE

3 4.0 APPLICABILITY...~......................'..... 3/4 0-1

3 .4'.1 REACTHfITY CONTROL SYSTEMS

3/4.1.1 GORATION CONTROL'.

Shutdown Margin - T-- ~ 200'F.'................ .....Shutdown. Margin - T „ ~ 200'F........ ... ..........Bororen 0$ 1uQOn'ooeoooooooeoeeoeoo ~ oeooooo ~ oeooooooeo

nderator- Temperature: Coefficient ..................8fnimum Temperature for CrktfcaTjty ........

3/4 1-1

3/4 1-1

3(41-3'/4

1-43/4 1-53'-1-7

3'/4 To2; HORATIO@: SYSTEMS .F foe Pa'ths' Shutdem........8 tof PathS OperaQngeeeeoeeeeeo eo eeoc eoooeo eeoc coo e

Cherg7ng P SShntdawe.....'..'hergIng

Pumpe - OperntIng....'.....................Borfe Acfd.. Pumps: - Shutdeer...................Eorfe Aci4 Pumps, - Operating;....................Berated. Mater Sources - Shutdowr .................Borated Mater Sources. - Operating ...................

3/4 1-8

3/4 1-8

we I-Io A>43/4 1-12.

3(4 1-13

3(41-14'/4

1-15

3/4 1-16

3/4 1-4

3/4.T.3 MOVABLE CONTROL ASSEMBLIES --.~......Foll Length CEA Posftfon '..........................Posftjon; Indicator Channels......................

EA, DrOp Tjtt|ee e o o ~ e o o ~ ~ e ~ o ~ o o o o o o o o ~ o o o ~ ~ o o e o e o o o e o ~ oC

Shutdee CEA Insertion Limit..;........,..............Pegulatlng CEA Insertjon Limits;........'.............

3/4 1-20

3/4 1-20

3/4 1-24

3/4 1-26

3(4 1-27

3/4 1-28

ST. LUCIE - UNIT 1 Amendment No37

APPLICABILITY

SURVEILLANCE RE UIREHENTS (Continued)

4.0.5 (Continued)

ASNE Boiler and PressureVessel Code and applicableAddenda terminology forinservice inspection and

testin activities

Required frequenciesfor performing inserviceinspection and testing

activities

MeeklyMonthly

Quarterly or every 3 monthsSemiannually or every 6 months

Yearly or annually

At least once per 7 daysAt least once per 31 daysAt least once per 92 daysAt least once per 184 daysAt least once per 366 days

C ~ The provisions of Specification 4.0e2 are applicable to the aboverequired frequencies for performing inservice inspection and testing (Qactivities.

d.

e.

Performance of the above inservice inspection and testing activi tiesshall be in addition to other specified Surveillance Requirement ~

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

ST. LUC IE - UNIT 1 3/4 0~3 Amendment Ho. 90

~ g

A

l

4

(4

~ QI~%

REACTIVITY CONTROL SYSTEMS

CHARGING PUM 8 SHUTDOWN

LIMITING CONDITION FOR OPERATION

3.1.2.3 At least one charging pump or one high pressure safety injectionpump* in the boron injection flow path required OPERABLE pursuant to Specifi-cation 3.1.2.1 shall be OPERABLE and capable of being powered from anOPERABLE emergency bus.

APPLICABILITY: MODES 5 and 6.

ACTION:

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

SURVEILLANCE RE UIREMENTS

4.1.2.3 At least one of the above required pumps shall be demonstratedOPERABLE by verifying the charging pump develops a flow rate of greaterthan or equal to 40 gpm or .,the high pressure safety injection pump developsa total head of greater than or equal to 2571 ft. when tested pursuant toSpecification 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) no charging pumps are operable. In this case, all charging pumps

shall be disabled and heatup and cooldown rates shall be limited inaccordance with Fig. 3.1-lb.

HCV-3616HCV-3626HCV-3636HCV-3646

At RCS temperatures below 115'F, any two of the following valves in theoperable HPSI header shall be verified closed and have their power removed:

Hi h Pressure Header AUxiliar HeaderHCV-361HCV-3627HCV-3637HCV-3647

ST. LUCIE - UNIT 1 3/4 1-12 Amendment No. 5 O~Nf, gg, 104.

0INSTRUMENTATION

ACCIDENT MONITORING INSTRUMENTATION


3.3.3.8 The accident monitorin instrumentation channels shown in Table 3.3-11shall be OPERABLE , booAPPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. Actions per Table 3.3-11.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMEWTS

4.3.3.8 Each accident monitoring instrumentation channel shall be demonstratedOPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operationsat the frequencies shown in Table 4.3-7.

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

0REACTOR COOLANT SYSTEt)

HOT SHUTDOWN


,3.4.1.3 At 1east two of the loops, listed-'belc'~ shall be: OPERABLE and at leastone reactor coolant or shutdown cooling loop shall be. in operation.«

a Reactor Coo1ant Loop A and its assocfated steam generator and a<1'east: one associated reactor-coolant pump

b Reactor Coolant'Loop 3 and. fts assocfated steam gener atorTeast one assocfateL reactor coolant pum . D~W~e. /g

c Shutdown Coo1fng, Loop A

d Shutdown Cool fng Loop 8;

\

APPLICABILITY NODE 4

ACTION

gD P-E,p&~ QY7H~/

a..* Mi th. Tess than the- above requfred: re ator coolant. or shutdowncoolf~ Toops OPERABLE within. one (1) hour fnftfate correctfveactfoa to return the requfret. loops. to OPERABLE status. If they'enafnfng OPERABL'F Toop- fs. a, shutdown cooling loop be.fn COLDGitQTIN wfthfa 30'ours.

fifth no reactor coolant or." shutdown: coolfng. Toop fn'peration.uspenC alT operatfons'nvolvfng, a rsiuctfoa. fn boron concentra-

Lfoe of the Reactor Coolant System and wfthfn one (1) hour initiatecorrective action.- to.return the r equired coolant loop to'peration.

«ATT reactor coolant pumps. and shutdown cool fng pumps. may be de-energized for~ up to 1 hour provided (1). no: operations are permitted Rat would cause dilu-

tion of the Peactor Coolant Systia boron concentratforr. and (2) core outlettaaperature is maintained at'least. 10.'F below saturatfon temperature.

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

of Sse.amFo MK COmPLCT»~F EACH lesKpQicE. j45p&qlud

SVcTHlsE5 DkE NQ~9E~ Q'p ~+~ p~ +graf)j~

%P~~~~ C ~~~~ToZ, 5uVu e,a g.mO~l~ y4, ~~ <o~W 5 bNl&5lbN lN 1« ~«~m Puu.sHAhrT V~ 5PMivl~ir~ I',.9.g

REACTOR COOLANT SYSTEM

SURVEILLANCE RE UIREMENTS Contfnued

5. Defect means an imperfection of such severity that it exceedstepTuggtng Ikmkt. A tuba containing a da1'act $ s dai'acttvaAny tube which does not permit the passage of the eddy-current inspection probe shall be doomed a defective tube.

6. PIu fn Limit means the imperfection depth at or beyondw c t e tu e shall be removed from servfce because ft maybecome unserviceable prior to the next fnspectfon and isequal to 44% of the nomfnal tube wall thickness.

7. Unserviceable describes the condition of a tube ff it leaks~tat tIra ght title I tfntegrfty fn the event of an Operatfng Basfs Earthquake, aloss-of-coolant accident, or a steam lfne or feedwater Ifnebreak as specfffed fn 4.4.5.3.c, above.

8. Tube Ins ection means an inspectfon of the steam generatoru e rom t e point of entry (hot Ieg sfde) completely around

the U-bend to the top support of the cold leg.

b. The steam generator shall. be determfned OPERASLK after completingthe cor respondfng actfons (plug all tubes exceeding the pIuggfngIfmft and all tubes contafnfng through-wall cracks) requfred byTable 4.4-2.

4.4.5.5 ~Ra orts

a. Following each fnservfce fnspictfon of steer generator tubes, thenumber of tubes plugged tn each stems generator shall be reportedto the Caaafssfon wfthfn IS days.

b. The complete results of the steam generator tube fnservfce fnspac-tfon shall be included fn the Annual Operatfng Report for theperfod fn whfch thfs fnspectfon was completed. This report shallfnclude:

Number and extent of tubes fnspected.

2. Locatfon and percent of wall-thickness penetration for eachfndfcatfon of an imperfection.

3. Identfffcatfon of tubes plugged.

fKlK76. ANQ MPLgCETHE. 4 oggi~LEVE NG5ua'g Of-Yea STa~ C CVMmep. TunE, lNSERuiCE iaSOECTi~d

To '5 PE@|

SHAl-L SC Su8ehtTlgQ To TH6 Commi&~50& tM A '5~H.tW| p EP T Pu 5OC 4 HA~TRl p'lcATiow 0 I Q MMLH i ~bg~5 F0LLb44ht46 C,o~pgc'giga/ OF ~5LQ~P&TAto . THL5 '5PKCigg REPORT +gag,

ST. LUCIK - UNIT I 3/4 4-8 AmenMnt No 18 108

1I ~

REACTOR COOLANT SYSTEM

REACTOR COOLANT SYSTEM LEAKAGE

LIMITIHG CONDITION FOR OPERATION

3.4.6.2

.a.

b.

C ~

Reactor Coolant System leakage shall be limited to:

Ho PRESSURE BOUHDARY LEAKAGE,

GPM UNIDENTIFIED LEAKAGE,

1 GPM total primary-to-secondary leakage through steam generators,

d. 10 GPH IDEHTIfI » ~ .YAGE from the Reactor Cool ant System, and

e. Leakage as spCcified in Table 3.4.6-1 for each Reactor CoolantSystem Pressure Iso.. -~alve identified in Table 3.4.6-1.

APPLICABILITY: MCOES 1, 2, 3 and 4. V+

ACTIOI(:

a.

ll

b.

c

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBYwithin 6 hours and in COLO SHUTDOWN within the following 30 hours.

With any Reactor Coolant System leakage greater than any,.one ofthe above limits, excluding PRESSURE BOUNDARY LEAKAGE and ReactorCoolant System Pressure Isolation Valve leakage, reduce the leakagerate to within limits within 4 hours or be in at least HOT STANDBY

within 6 hours and in COLD SHUTDOWN within the following 30 hours.

With any Reactor Coolant System Pressure Isolation Valve leakagegreater than the limit in 3.4.6.2.e above reactor operation maycontinue provided that at least two valves, including check valves,in each high pressure line having a non-functional valve are inand remain in the mode corresponding to the isolated condition.Motor operated valves shall be placed in the closed position, andpower supplies deenergized. (Note, however, that this may leadto ACTION requirements for systems involved.) Otherwise, reducethe leakage rate to within limits within 4 hours or be in atleast HOT STANDBY within 6 hours and in COLO SHUTDOWN within thefollowing 30

4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be withineach of the above limits by:

a. Monitoring the containment atmosphere. gaseous and particulateradioactivity at least once per 12 hours.

ST LUCIE - UNIT 1 3/4 4-14 Order dated 4/20/81

Penetration ~Ssten

COtlTAItBENT LEAKAGE PATHS

Valve Ta NumberLocation

to Containment ServiceTest~Ts

Hakeup Hater

8 Station AirPELTATE. PsQ EF Piece wtTH

g@Q C ~ <- l%" 1 1 "j),~(W-v- ><™gq(.g

Gate (I-NV-15-1)Check (I-.V-15328)

I lobe (I-V-18-947)+Gl obe I-V-18- ++Check I-V-181118)w++Globe ( I-SH-18797)+++

OutsideInsideOutsideOutsideInsi e+++Annulus~++

Station Air Supply Bypass

Primary Makeup Pater Bypass

0

~s

10

14

23

25

26

28

Instrument Air

ContainmentPurge

ContainmentPurge

Hastelianagement

ComponentCooling

ComponentCoolingFuel TransferTube

CVCS

Sampling

Ga te ( I-11V-18-1)Check ( I-V-18195)

Butterfly (I-FCV-25-4)Outterfly (I-FCV-25-5)

Butterfly (I-FCV-25-3)Butterfly ( I-FCV-25-2)

Globe (V-6741Check (V-6779)

Butterfly (I-HCV 14-7)Butterfly (I-HCV-14-1)

Butterfly (I-HCV-14-6)Butterfly (I-HCV-14- )

Double Gasket Flange

Globe (V-2515)Globe (V-2516)

Globe (V-5200)Globe (V-5203)Globe (I-FCV-03-lE)Globe ( I-FCV-03-1F)

OutsideInside

InsideOutside

InsideOutside

OutsideOutside

OutsideOutside

OutsideOutside

Inside

InsideInsideOutsideOutsideOutsideOutside

Containment PurgeExhaust

Containment PurgeSupply

Nitrogen supply toSI Tanks

RC Pump CM Supply

Type C

1

Type C

Bypass

Bypass

RC Pump CH ReturnI

Fuel TranSfer

Letdown Line

Reactor CoolantSampleSI Tank SampleSI Tank Sample

Bypass

Bypass

Bypass

Bypass

Bypass

Instrument Air Supply Bypass

+To become I-HCV-18-2 upon completion of the modification described in L-B7-123.++To become I-V-18795 upon completion of the modification described in L-87-123.

t++To become effective upon completion of the modification described in L-87-123.

TASLE 3.6-2

CONTAINMENT ISOLATION VALVES

Valve Ta Number

A. CONTAINMENT ISOLATION

1. I-FCV-25-4,5

2. I-FCV-25-2,3

3. I-MV-15-1

4. I-MV-18-1

5. V-6741

6. I-HCV-14-1 8 7

7. I-HCV-14-6 S 2

8. V-2515,2516

9. V-5200,5203

10. V-5201,5204

ll. V-5202,5205

12. V-6554,6555

13. I-LCV-07-11A,118

PenetrationNumber

10

ll7

9

14

23

24

26

28

29

29

31

42

Function

Containment purge air exhaust, CIS

Containment purge supply, CIS

Primary makeup water, CIS

Instrument air supply, CIS

Nitrogen supply to safety injectiontanks, CIS

Reactor coolant pump cooling watersupply, SIAS

Reactor coolant pump cooling waterreturn, SIAS

Letdown line, CIS, SIAS

Reactor coolant sample, CIS

Pressur izer surge line sample, CIS

Pressurizer steam space sample, CIS

Containment vent header, CIS

Reactor cavity sump pump discharge,CIS

Testable DuringPlant 0 ration

No

No

Yes

No

Yes

No

No

No

Yes

Yes

Yes

Yes

Yes

IsolationTime Sec

5

5

19

28

5

5

5

5

5

10

14. V-6301,6302

15. V-2505

16. I-SE-01-1

43

44

Reactor drain tank pump suction, CIS

Reactor coolant pump controlledbleedoff, CIS

Reactor coolant pump controlledbleedoff, CIS

Yes

No

17.+ I-HCV-18-2 8. Station air supply, CIS

~item 17 to become effective upon completion of the modification describedin L-87-173.

Yes

Valve Ta Number

B- HANUAL OR REHOTE

MANUAL

PenetrationNumber

TABLE 3.6-2 Continued

FunctionTestable OuringPlant 0 ration

Isol a tionTime Sec

DEMT~ Am ERA~~cd&

X-'9-l%- 1 g t

1. I-V-18-947+

2. I-V-25-11,12

3. I-V-25-13,14,15,16

4. V-3463

5. I-V-07009

6. V-07206, V-07189

7. V-07170, V-07188

57 8 58

41

46

47

Station air supply, Hanual

Hydrogen purge outside air make-up, Hanual (NC)

Hydrogen purge exhaust, Manual(NC)

Safety in)ection tank test line,Hanual (NC)

Safety in)ection tank test line,Hanual (NC)

Refueling cavity purification flowinlet, Hanual (NC)

Refueling cavity purification flowoutlet, Hanual (NC)

Yes

Yes

Yes

Yes

Yes

Yes

Yes

1

NA

NA*

NA*

NA

o

8. I-FSE-27-1,2,3,4,8,11

9. I-FSE-27-5,6,7,9,10

4&a 5 Hydrogen sampling line, Remote4&c manual

5la 5 Hydrogen sampling line, Remote51c manual

Yes

Yes

NA*

NA*

~To become I-V-1&795 upon completion of the modification described in L-&7-123.

TR A P W R SYST H

URV AN R R H NT ontinued

QE LET&. A~0Q.EPLACC

Qcc u,MULATc0

g. At least once per ten years by:

2.

Draining each fuel storage tank, removing the accummulatedsediment and cleaning the tank using an appropr>a ecleaning compound, and

Performing a pressure test of those portions of the dieselfuel oil system designed to USAS B31.7 Class 3 requirementsat a test pressure equal to 110% of the system designpressure.

4.8.1.,1.3 ~R~r - All diesel generator failures, valid or non-valid, shallbe reported to the Commission pursuant to Specification 6.9.2. Reports ofdiesel generator failures shall include the information recommended inRegulat'ory Position C.3.b of Regulatory Guide 1. 108, Revision 1, August 1977.If the number of failures in the last 100 valid tests (on a per nuclearunit basis) is greater than or equal to 7, the report shall be supplementedto include the additional information recommended in Regulatory Position C.3.bof Regulatory Guide 1. 108, Revision 1, August 1977. /

4.8. 1. 1.4 The Class 1E underground cable system shall be demonstratedOPERABLE within 30 days after the movement of any loads in excess of80K of the ground surface design basis load over the cable ducts bypulling a mandrel with a diameter of at l'east 80% of the duct's inside .

diameter through a duct exposed to the maximum loading (duct nearestthe ground's surface) and verifying that the duct has not been damaged.

ST. LUCIE - UN?T 1 3/4 8-6b Amendment No. 444 112

~ J.

T.ABLE»".. 11-1

RAGIOACT:V L:GU:O '~ASTE SAMPL'NG ANO n,'iAL",SIS ?ROGRAM

L i qui d .=,el e as e

TypeSamolingFrequency

MinimumAralysisFr quency

Tvpe o ActivityAnalys i s

Lower L imi tor Oetection

(LLO)(pCi/ml )

A. Batch WasteReleaseTanks,

P P

Each Batch Each Batch Principal Gamma Sxlo"7

Emittal s

P

One Ba ch/M

P'

bEach Batch Composite

I-131

Oissolved andEntrained Gases(Gamma Emitters)

H-3R

Gross Aloha

1 x10

1 xl0

1 xl 0

1 x10

P

Each BatchQ

Ccrocsit bS~e9,

Sr-90'e-=5

Sxlo

1xlo

B. CantinuopyReleases " Gaily Composite

«7Principal Gamma 5xlo

ecml tters

O '/MGrab Samole ComoositeGaily

I-131

Oissolved andEntrained Gases(Gamma Emi- ers)

lxlo

1 xl0

GailyM

CompositeH-3

Gross Alpha

lxlo

lxl0

C. Set:1'ngSas in

Gaily

Grab Sample

'QComoos i te

Sr-89, Sr" 90

Fe" 55

Principal Gammae

Emi t"ers

I 131

-S5x10

1xl0

Sx'0

~m'0 "

ST. LUC IE - UW IT 1 3/ -11-2 Amendment 1'Io. 5 9

.I

yg

F

~t


TABLE NOTATION

b.

c ~

Sampling and analysis shall also be'performed following shutdown, startup,or' THERMAL POWER change exceeding 15$ of RATED THERMAL POWER withinI hour unless (I) analysis shows that the OOSE EQUIVALENT I-131 concentra-tion in the primary coolant has not increased more than a factor of 3; and(2) the noble gas activity monitor shows that effluent activity has notincreased by more than a factor of 3.

Samples shall be changed at least 4 times a month and analyses shall becompleted within 48 hours after changing (or after removal from sampler).Sampling shall also be performed at least once per 24 hours for at least7 days following each shutdown, startup or THERMAL. POWER change exceeding155 of RATED THERMAL POWER in I hour and analyses shall be completedwithin 48 hours of changing if (1) analysis shows that the OOSE EQUIVALENTI-131 concentration in the primary coolant has incr'eased more than afactor of 3; and (2) the noble gas activity monitor shows that effluentactivity has increased by more than a factor of 3. When samples collectedfor 24 hours are analyzed, the corresponding LLDs may be increased by afactor of '10.

e.

The ratio of the sample flow rate to the sampled stream flow rate shallbe known for the time period covered by each dose or dose rate calculationmade in accordance with Specifications 3.11.2.I, 3.11.2.2 and 3.11.2.3.

The principal gamma emitters for which the LLD specification appliesexclusively are t'e following radionuclides: Kr-87, Kr-88, Xe-133,Xe-133m,,Xe-135, and .Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58,Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulateemissions. This list does not mean that only 'des are to bedetected and reported. Other peaks which ar mea bl nd identifiable,together with the above nuclides, shall also be identified and reported.

PpPP~ QLTH,

We<5+t'~Q(~'T.

LucrE - UN@' 3/4 11-10 Amendment No. 5 3


TABLE NOTATION

It shooid be recognized that the LLO is defined as an a ~riori (beforethe fact) limit representing the capability of a measurement system and

h f i li~i«ment. Analyses shall be performed in such a manner that the stated LLDswill be achieved under routine conditions. Occasionally backgroundfluctuations, unavoidable small sample 'sizes, the presence of interferingnucl ides, or other uncontrollable circumstances may render these LLDsunachievable. In such cases, the contributing factors shall be identifiedand described in the Annu iolo ical Environmental Operating Reportpursuant to Specification 6.9.1.11.

dLLD for drinking water samples. If no drinking water pathway exists, theLLD of gamma isotopic analysis may be used.

An equilibrium mixture of the parent and daughter isotopes which correspondsto 15 pCi/a of the parent isotope.

OE,LETE Qtdo ~P ~~(A>Thd

ST. LUG I E - UNIT 1 3/4 12-10 Amendment No. 5 3

REACTIVITY CONTROL SYSTEMS

BASES

3/4.1.3 MOVABLE CONTROL ASSEMBLIES Continued

The ACTION statements applicable to misaligned or inoperable CEAsinclude requirements to align the OPERABLE CEAs in a given group with theinoperable CEA. Conformance with these alignment requirements brings thecore, within a short period of time, to a configuration consistent withthat assumed in generating LCO and LSSS setpoints. However, extendedoperation with CEAs significantly inserted in the core may lead to pertur-bations in 1) local burnup, 2) peaking factors, and 3) available shutdownmargin which are more adverse than the conditions assumed to exist in thesafety analyses and LCO and LSSS setpoints determination. Therefore, timelimits have been imposed on operation with inoperable CEAs to precludesuch adverse conditions from developing.

The requi~ement to reduce power in certain time limits, dependinq uponthe previous F , is to eliminate a potential nonconservatism for situationswhen a CEA has been declared inoperable. A worst case analysis has shownthat a DNBR SAFDL violation may occur during the second hour after the CEAmisalignment if this requirement is not met. This potential DNBR SAFDLviolation is eliminated by limiting the time operation is permitted at FULLPOWER before power reductions are required. These reductions will benecessary once the deviated CEA has been declared inoperable. The timeallowed to continue operation at a reduced power level can be permitted forthe following reasons:

1. The margin calculations that support the Technical Specificare based on a steady-state radial peak of F„ 1.70. Qg4 an

2. When the actual F„ < 1.70, significant additional margin exists.

3. T(is additional margin can be credited to offset the increase inF with time that can occur following a CEA misalignment.r

4. This increase in F .is caused by xenon redistribution.r5. The present analysis can support allowing a misalignment to

e~ist for up to 60 minutes without correction, if the initialF < 1.67.

Operability of the CEA position indicators (Specification 3.1.3.3) isrequired to determine CEA positions and thereby ensure compliance with theCEA alignment and insertion limits and ensures proper operation of the rodblock circuit. The CEA "Full In" and "Full Out" limits provide- an addi-tional independent means-for determining the CEA positions wheF'the CEAs areat either their fully inserted or fully withdrawn positions. Therefore, theACTION statements applicable to inoperable CEA position indicytors permitcontinued operations when the positions of CEAs with inoperable positionindicators can be verified by the "Full In" or "Full Out" limits.

ST. LUCIE - UNIT 1 8 3/4 1-4 Amendment No. 18.82~

j„f4

thf

PLANT SYSTEMS

BASES

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM Continued

for operations personnel during and following all credible accident condi-tions. The OPERABILITY of this system in conjunction with control room

design provisions is based on limiting the radiation exposure to personneloccupying the control room to 5 ran or less whole body, or its equivalent.This limi tion is consistent with the r equirements of General DesignCriteria 10 of Appendix "A"., 10 CFR 50.

OELETE AMo EEPW~3 4.7.8 ECCS AREA VENTILATION SYSTEM

The OPERABILITY o f the ECCS area ventilation sys tern ensur es tha t radio-active materials leaking from the ECCS equipment following a LOCA are filteredprior to reaching the environment. The operation of this system and theresultant effect on offsite dosage calculations was assumed in the accidentanalyses.

3/4.7. 9 SEALED SOURCE CONTAMINATION

The limitations on sealed source removable contamination ensure that thetotal body or individual organ irradiation does not exceed allowable limitsin the event of ingestion or inhalation of the probable leakage from the sourcematerial. The limitations on removable contamination for sources requiringleak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits forplutonium. guantities of interest to this specification which are exempt fromthe leakage testing are consistent with the criteria of 10 CFR Parts 30.11-20and 70.19. Leakage from sources excluded from the requirements of thisspecification is not likely to represent more than one maximum permissiblebody burden for total body irradiation if the source material is inhaled oringested.

3 4.7.10 SNUBBERS

All safety related snubbers are required to be OPERABLE to ensure thatthe structural integrity of the reactor coolant system and all other safetyrelated systems is maintained during and following a seismic or other eventinitiating dynamic loads. Snubbers excluded from this inspection program are

those installed on nonsafety-related systems and then only if their failureor failure of the system on which they are installed would have no adverse

effect on any safety-related system.

The visual inspection fr equency is based upon maintaining a constant levelof snubber protection to systems. Therefore, the required inspection intervalvaries inversely with the observed snubber failures and is determined by thenumber of inoperable snubbers found during an inspection. Inspections performed

ST. LUCIE - UNIT 1 B 3/4 7-5 "Amendment No. AA. g'f ~ 83

~ J~

~1fi

4.)i ~ji

V

vE

I

t'lP '

REFUELING OPERATIONS

BASES

3/4.9.12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE

The limitations on the fuel handling building ventilation systemensures that all radioactive material released from an irradiated fuelassembly will be filtered thr ough the HEPA filters and charcoal adsorberprior to discharge to the atmosphere. The OPERABILITY of this systemand the resulting iodine removal capacity are consistent with the assump-tions of the accident analyses.

3/4.9.13 SPENT FUEL CASK CRANE

The maximum load which may be handled by the spent fuel cask craneis limited to a loaded single element cask which is equivalent to approxi- ~

mately 25 tons . This restriction is provided to ensure the structuralintegrity of the spent fuel pool in the event of opped cask accident.Structural damage caused by dropping a load i exc of a loaded single

lement cask could cause leakage from the spent ue pool in excess of themaximum makeup capability. P I= PLUM

3/4.9.14 'OECAY TIME - STORAGE POOL

The minimum requirements for decay of the irradiated fuel assembliesin the entire spent fuel storage pool rior to movement of the spent fuelcask into the fuel cask compartmen insur that sufficient time has ela sedto allow radioactive decay of the fission ro ucts. The decay time of 1180hours is based upon one-third of a core placed in the spent fuel pool eachyear during refueling until the pool is filled. The decay time of 1490 hour sis based upon one-third of a core being placed in the spent fuel pool eachyear during refueling following which an entire core is placed in the poolto fill it. The cask drop analysis assumes that all of the irradiated fuelin the filled pool (7 2/3 cores) is ruptured and follows Regulatory Guide 1.25methodology, except that a Radial Peaking Factor of 1.0 is applied to allirradiated assemblies.

ST. LUCIE - UNIT 1 B 3/4 9-3 Amendment No. 2l, 89, 91

ADMINISTRATI VE CONTROLS

6.15 MAJOR CHANGES TO RADIOACTIVE LI UID, GASEOUS AND SOLID WASTE TREATMENT

V

6.15.1 Licensee initiated major changes to the radioactive waste systems(liquid, gaseous and solid):

1. Shall be reported to the Commission in the Semiannual RadioactiveEffluent Release Report for the period in which the evaluation wasreviewed by the Facility Review Group. The discussion of eachshall contain:

a.

b.

C.

d.

e.

A summary of the evaluation that led to the determination thatthe change could be made in accordance with 10 CFR 50.59;

Sufficient detailed information to totally support the reason 0for the change without benefit of additional or supplemental ~

information;- ReP

A detailed description of the equipment, components andprocesses involved and the interfaces with other plant syst

rAn evaluation of the change which shows the predicted releasesof radioactive materials in liquid and gaseous effluents and/orquantity of solid waste that differ from those previouslypredicted in the license application and amendments thereto;

An evaluation of the change which shows the expected maximumexposures to individuals in the UNRESTRICTED AREA and to thegeneral population that differ from those previously estimatedin the license application and amendments thereto;

AQiLl/y

f. A comparison of the predicted releases of radioactive materialsin liquid and gaseous effluents and in solid waste to the actualreleases for the period prior to when the changes are to be made;

9. An estimate of the exposure to plant operating personnel as aresult of the change; and

h. Documentation of the fact that the change was reviewed andfound acceptable by the FRG.

2. Shall become effective upon review and acceptance by the FRG.

Licensees may choose to submit the information called for in this Specifica-tion as part of the annual FSAR update.

ST. LUCIE - UNIT 1 6-24 Amendment No. 69

ATTACHMENT 2

Marked-up St. Lucie Unit 2 Technical Specifications page:

2-42-5

3/4 2-93/4 3-23/4 3-133/4 6-23/4 6-263/4 8-73/4 9-93/4 11-103/4 12-10

6-25

TABLE 2.2"1

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIHITSIC:nMm

C

FUNCTIONALUNIT'.

Manual Reactor Trip

2. Variable. Power Level - High

Four Reactor Coolant PumpsOperating

3. Pressurizer Pressure - High

4. Thermal Hargin/Low Pressure (0

Four Reactor Coolant PumpsOperating

TRIP SETPOINT

Not Applicable

< 9.61'bove THERMAL POWER,

with a minimm setpoint of15K of RATED THERMAL POWER,

and a maximum of < 107.0X ofRATED THERMAL POQER.

< 2370 psia

Trip setpoint adjusted to notexceed the- limit lines ofFigures 2.2-3 and 2.2-4.Minimum value of 1900 psia.

ALLOWABLE VALUES

Not Applicable

< 9.61% above THERMAL POWER, aod ~a minimum setpoint of 15X ofRATED THERMAL POWER and a maximmof < 107.0X of RATED THERMAL PMER.

< 2374 psia

Trip setpoint adjusted to notexceed the limit lines ofFigures 2.2-3 and 2.2-4.Hiniaum value of 1900 psia.

~ 5. Containaent Pressure - High

Steam Generator Pressure - Low

Steam Generator PressureDifference - High(Logic in TH/LP Trip Unit)

6.

f'+&. Steam Generator Level - Low

3 0 psig

> 626.0 psia (2)

< 120.0 psid

> 20.5X (3)

< 3.l Psig

> 621.0 psia (2)

< 132.0 psid

> 19.5X (3)

TABID 2.2-1 (CooL~loood

ttfACTOlt PROTECTIVE INSTRtHIENIATIDN TRIl'ETPOINT l IHI1S

FUNCTlONnt. UNlf

9. - Local l'ower DensiLy - lligl}(5)

10. Loss of C(}}apo}}ent Coolin(J Matertu lteacto} t:ool ant Pui}}l}s-Low

ll. Iteactov Protection Sysle}}} l.ogic

12. lteact(}r 'trip ltreakevs

13. ltate ol Cl}all(J» of l'ower - lligh

14. Ilo iolor Cooloot Flow - l.o (1)

t5. l.oss of l.oad (Turl}l(}e)llyd}.aul }c I-l(}td I'}ess(}re - low

1 ltlp SETPOI AT

lrip setpoinL adjust.ed Lonut exceed t,l}e li(}}it"li(}eso f F i gures 2. 2-1 and 2. 2-2.

> 636 gl}}}}""

Not Applicable

Not Applicable

< 2.49 decades per (}}i(}ute

> 95.4X of design ReacLorCoo)a(}L flow witl} four

. pu}}}ps operating"

> 800 psig

ALLOWABLE VALUES

Trip set.point adjusted tonot. exceed t.he li}}}itlinesof Figures 2.2-1 and 2.2-2.

> 636 gp(a

Not Appl icable.

Not. Appl icable

< 2.49 decades per a}inute

> 94.9X of (lesig}} Reacto}Coolant. flow witl} fo(}rpu((}ps ol)e}'at ing"

> 800 psi(J

Design reactor cooia}}L flow with fouv pu}t}ps ol}orating is 363,000 (J)}le.

10-}1}i»}}L(}Li}t}e delay after relay act(}atio».

POWER GIS RIBUTION LIMITS

TOTAL INTEGRATED RADIAL P<DKING FACTOR - F„T

LIMITING CONG ITION FOR OPERATiON

3, 2. 3 The calculated value of F„, shall be limited to < l. 70.T

APPLICABILITY: MOOE l".

ACTEON:

With F > l.70, within 6 hours either:Tr

a ~ Be in at least HOT STANDBY, or

Reduce THERMAL POWER to bring the comoination of THERMAL POWER ana

F to ~ithin the limits or Figure 3.2-3 and withdraw the rull-lengthCEAs to or oeyond the Long Term Steady State Insertion Limits orSpecification 3. l.3.6. The THERMAL POWER limit determined fromFigure 3.2-3 shall then be used to establish a revised uoper THERMALPOWER level 'limit on Figure 3.2-4 (truncate Figure 3.2-4 at the

'allowable frac=ion of RATED THERMAL POWER determined by Figure 3.2-3)and subsequent ooeration shall be maintained within the reducedacceptable operation region of Figure 3.2-a.

SURVEILLANCE REOUIREHENTS

4.2.3.1 The provisions of Soecification 4.0.4 aro not aooiicable.

4.2.3.2 F„. shall be calculated by the expression -'. = .=„",i-: ) when F„is calculated with a non-iull core power distr'but:on ana'ivs s code and

shall be calculated as F' when calculations are oerformed with ar r.

.ful 1 cor e "owe. di stribution analysis code. F„shall be determined -o

be within its limit a" the Following inter/ais:

a. Prior toloadin .

opera 'bove IG: of RA ED .HERBAL .=rJ~ER =f:ar ~= " =":DELHI, Awg Q.a wc- ~n"q 4 caw~q "

b. At least once per 31 days or accumulated ooeration in .'~00K ', and

c. Ai hi n '. ilours 1 tile A2 'UTHAL POWER TILT ( ) i s ~ O.

'ee

.oec'. al: est Kxceot. on 3. O. 2.

ST. LUCIE - JiIIT 2 3 If Amendment .'Io.;

TABLE 3.3-1

REACTOR PROTECTIVE INSTRUMENTATION~ I

Mfoal

FUNCTIONAL UNIT

1. Manual Reactor Trip

2.

3.

4.

5.

6.

Variable Power Level — High

Pressurizer Pressure - High

Thermal Margin/Low Pressure

Containment Pressure - High

Steam Generator Pressure - Low

TOTAL NO.OF CHANNELS

44

4

4

4

4/SG

CHANNELSTO TRIP

22

2(a)(d)nDD,

2 d)

2

2/SG(b)

MINIMUMCHANNELSOPERABLE

44

3

3

3

3

3/SG

APPLICABLEMODES

1, 23A 4A 5A

1, 2

1, 2

1, 2

1, 2

1, 2

ACTION

15

2¹

2¹

2¹

2¹

2¹

7. Steam Generator PressureDifference - High

8. Steam Generator Level - Low

9. Local Power Density - High

10. Loss of Component Cooling Waterto reactor Coolant Pumps

ll. Reactor. Protection System Logic

4/SG

12. Reactor Trip Breakers

13. Wide Range Logarithmic NeutronF lux Moni tora. Star tup and Operating-

Rate of Change of Power-High

b. Shutdown

14. Reactor Coolant Flow — Low .. 4/SG

15. Loss of Load (TurbineHydraulic Fluid Pressure - Low) 4

2(a)(d)2/SG

2(c)(d)'3/SG

3

3

3

2(f)

2(c)

2(e)(g) hag0 8 rg 2

2/SG d) 3/SG

1, 2

1, 2

.1

1, 2

1 23* 4* 5A

1, 23* 4* 5'A

1, 23, 4, 5

1, 2

2¹

2¹

2¹

2¹

2¹5

5

2¹3

2¹

~ . IJ ~


ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

IC

Mm

C

4/steamgenerator

4

TOTAL NO.FUNCTIONAL UNIT OF CHANNELS

4. MAIN STEAH LINE ISOLATION (HSIS)a. Hanual (T 2

Button ) AODb. Steam Generator

Pressure - Low

c. Containment Pressure-High

d. Automatic Actuation 2Logic

5. CONTAINMENT SUMP

RECIRCULATION (RAS)a. Manual RAS (Trip

Buttons) 2

b. Refueling Mater StorageTank - Low

c. Automatic ActuationLogic 2

CHANNELSTO TRIP

2/steamgenerator

2

HINIHUHCHANNELSOPERABLE

3/steamgenerator

3

APPLICABLEMODES ACTION

1,2,3 16

l,2,3

1, 2, 3

13*, 14

12

1,2,3,4 . 12

1,2,3

1, 2, 3

17

12

1, 2, 3(c) 13", 14

Il

I l4 t

I

i

CONTAINMENT SYSTEMS

CONTAINMENT LEAKAGE

LIMITING CONOITION FOR OPERATION

3.6.1.2 Containment leakage rates shall be limited to:a. An overall integrated leakage rate of:

1. Less than or equal to L , 0.50 percent by weight of thea'ontainmentair per 24 hours at P , 41.8 psig, oras

b.

C.

2. Less than or equal to Lt, 0.35 percent by weight of thecontainment air per 24 hours at a reduced'ressure of Pt,20e9 psig. I

A combined leakage rate of less than or equal to 0.60 L for allapenetratfons and valves subject to Type B and C tests whenpressurized to P . DZMT'E. 59A combined bypass leakage rate of less of than or equal 0. 12 L for

aall penetrations identified'n Table 3.6-1 as secondary containmentbypass leakage paths when pressurized to P .

a'PPLICABILITY:HOOES 1, 2, 3, and 4.

ACTION:

With either (a) the measured overall integrated oontainment leakage rataexceeding 0.75 L or 0.75 Lt, as applicable, or (b) with the measured combinedleakage rate for all penetrations and valves subject to Types 8 and C test"exceeding 0.60 L , or (c) with the combined bypass leakage rate exceeding

a'.L2

L , restore the overall integrated leakage rate to less than or equalto 0.75 L or less than or equal to 0.75 Lt, as applicable, and the combinedleakage rate for all penetrations and valves subject to Type B and C tests toless than or equal to 0.60 L , and the bypass leakage rate to less than orequal to O.i2 L, prior to increasing the Reactor Coolant System temperatureabove 2004F.


4.6. 1.2 The containment leakage rates shall be demonstrated at the followingtest schedule and shall be determined in conformance with the criteriaspecified in Appendix J of 10 CFR 50:

a. Three Type A tests (Overall Integrated Containment Leakage Rate)shall be conducted at 40 + 10 month intervals during

ST. LUCIE - UNIT 2 3/4 6-2 Amendment No. 3g, 37

CONTAINMENT SYSTEMS

3/4.6.5 VACUUM RELIEF VALVES


3.6.5 The primary containme vess 1

OPERABLE with an actuatio set point0.35 inches water gauge.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

to annulus vacuum relief valves shall bef less than or equal to 9.85 +

oBL~76, A~Q C.c pM (GDo I fib 5 c.tpbIA+

With one primary containment vessel to annulus vacuum relief 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 'SHUTDOWN 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

Ice

cc

~ j

ELECTR ICAL

SURVEILLANCE RE UIRENENTS Continued

7,

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 injectionactuation signal.

Verifying the diesel generator operates for at least 24 hours.*+**Our ing the first 2 hours of this test, the diesel [email protected] be loaded within a load band of 3800 to 3985 kW~ andduring the remaining 22 hours of this test, the dieselgenerator shall be loaded within a load band of 3450 to3685 kW<. The generator voltage and frequency shall be4160 + 420 volts and 60 + 1.2 Hz within 10 seonds afterthe start signal; the steady-state generator voltage andfrequency shall be maintained within these limits duringthis test. Within 5 minutes after completing this 24 hourtest, perform Surveillance Requirement 4.8.1.1.2e.4.b)

8.

9.

Verifying that the auto-connected loads to each dieselgenerator do not exceed the 2000-hour rating of 3935 kW.

Verifying that the diesel generator's capability to:

a) Synchronize with the offsite power source while thegenerator is loaded with its emergency loads upon asimulated restoration of offsite power.

b) Transfer its loads to the offsite power source, and

DLLC ) G.

10.

c) He restored to its standby status.r

Verifying that with the diesel generator operating in a testmode (connected to its bus), a simulated safety injectionsignal overrides the test mode by (1) returning the dieselgenerator to standby operation and (2) automatically energizesthe emergency loads with offsite power.

Verifying that the fuel transfer pump transfers fuel from eachfuel storage tank to the engine-mounted tanks of each dieselvia the installed cross connection lines.

PThis 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 shallnot invalidate this test.

****Thistest may be conducted in accordance with. tFie manufacturer"srecommendations concerning engine pr elu5e period,

ST. LUCIE- UNIT 2 3/4 8-7 Amendment No. 39

R FUR IHQ OP( TIOHS

LOW MlTER LEV

LIHITINQ CONOITION FOR OPERATION

3. 9.8. 2 Two- fndependent shutdown coolfng loops shall be OPERABLE and at leastone shutdown coo1fng loop shall be fn operatfon.

APPLICABILITY: HOOE 6 when the water level above the top of the reactorpressure vessel flange fs less than 23 feet.

ACTEON:

a.

b.

fifth less than the requfred shutdown coolfng loops OPERABLE,wfthfn 1 hour inftfate correctfve actfon to return the requfrad loopsto OPERABLE status, or to establfsh greater than or equal to 23 feetof water above the reactor pressure vessel flange, as soon aspossfble.

fifth no shutdown coolfng loop fn operatfon, suspend all operatfonsfnvolvfng a reduction fn boron concentratfon of the Reactor CoolantSystii and wfthfn 1 hour fnftfate correctfve actfon to return therequfred shutdown coolfng loop to operatfon. Close all contafnmentpenetratfons provfdfng dfrect access froa the contafnment atmosphere'o the outsfde atmosphere wfthfn 4 hours.

SURVE LANCE R UIRENEHT

4.9.8.2 At least one shutdown coolfng loop shall be verfffed to be fnoper atfon and cfrculatfng reactor coolant at a flow rate of greater than orequal to 3000 gpi at least once per 12 hours,

ST. LUCIE - UNIT 2 3/4 9 9 Amendment No.4,

1

H

akf

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4» i4 ' 'p

V

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~ i ~

TABLE 4. 11-2 Continued

TABLE NOTATION

b."

C.

d.

e.

Sampling and analysis shall also be performed following shutdown, startup,or a THERMAL POWER change exceeding 15K of RATED THERMAL POWER within1 hour unless (1) analysis shows that the DOSE EQUIVALENT I-131concentration in the primary coolant has not increased more than a factorof 3; and (2) the noble gas activity monitor shows that effluent activityhas not increased by more than a factor of 3.

Samples shal 1 be changed at least 4 times a month and analyses shal 1 becompleted within 48 hours after changing (or after removal from sampler).Sampling shall also be performed at least once per 24 hours for at least7 days following each shutdown, startup or THERMAL POWER change exceeding15X of RATED THERMAL POWER in 1 hour and analyses shall be completed within48 hours of changing if (1) analysis shows that the DOSE EQUIVALENT I-131concentration in the primary coolant has increased more than a factor of 3;and (2) the noble gas activity monitor shows that effluent activity hasincreased by more than a factor of 3. When samples collected for 24 hoursare analyzed, the corresponding LLDs may be increased by a factor of 10.

'he

ratio of the sample flow rate to the sampled stream flow rate shallbe known for the time period covered by each dose or dose rate calculatiogmade in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

The principal gamna emitters for which the LLD specification appliesexclusively are the following radionuclides: Kr-87, Kr 88, Xe-133,Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59,Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 forparticulate emissions. This list does not mean that onl the nucl idesare to be detected and reported. Other peaks which ar measureable ndidentifiable, together with the above nuclides, shall a so e identifiedand reported.

QE~Tg QNQ kERAQ~«H "woes,ut gg4"

ST. LUCIE = UNIT 2 3/4 11-10 Amendment No. 25

TABLE 4. 12-1 Continued)

TABLE NOTATION

it should be recognized that the LLD is defined as an a priori (beforethe fact) limit representing the capability of a measurement system andnot as an a gosterior i (after the fact) limit for a par ticular measure-ment. Analyses shall be performed in such a manner that the stated LLOswill be achieved under routine conditions'ccasionally backgroundfluctuations, unavoidable small sample sizes, the presence of interferingnuclides, or other uncontrollable circumstances may render these LLOsunachievable. In such cases, the contributing factors shall be identifiedand described in the Annu adiological Environmental Operatin Repursuant to Specificatio 6.9.1.11. ~+~< ~<+ ~p~ ~qyH i> gq ~ g

dLLD for drinking water samples. If no drinking water pathway exists, theLLO of gamma isotopic analysis may be used.

An equilibrium mixture of the parent and daughter isotopes whichcorresponds to 15 pCi/R of the parent isotope.

ST. LUCIE " UNIT 2 3/4 12" 10

4N

$ % tlk+ 'I@0 V

.I

ADMINIST ATIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM PCP

6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the Semiannual RadioactiveEffluent Release Report for the period in which the chan (s) wmade. This submittal shall contain: P g Qhlg ~t3Kl=

tA~ ~ 5MPggf1a. Sufficiently detailed information to totall supor the rat ona e

for .the change without benefit of additional r pplementalinformation;

b. A determination that the change did not reduce the overallconformance of the dewatered bead resin to existing criteriafor radioactive wastes; and

c. Documentation of the fact that the change has been reviewed andfound acceptable by the FRG.

2. Shall become effective upon review and acceptance by the FRG.

6. 14 OFFSITE DOSE CALCULATION MANUAL ODCM

6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCM:.

Shall be submitted to the Commission in the Semiannual RadioactiveEffluent Release Report for the period in which the change(s) wasmade effective. This submittal shall contain:

a. Sufficiently detailed information to totally support therationale for the change withdut benefit of additional or supple-mental information. Information submitted should consist of apackage of those pages of the ODCM to be clianged with each pagenumbered and provided with an approval and date box, togetherwith appropriate analyses or evaluations justifying the change(s);

b.

Co

A determination that the change will not reduce the accuracy orreliability of dose calculations or setpoint determinations;and

Documentation of the fact that the change has been reviewed andfound acceptable by the FRG.

-2 Shall become effective. upon review and acceptance by the FRG.

ST. LUCIE - UNIT2'-25

ATTACHMENT 3

SAFETY ANALYSIS

INTRODUCTION

These amendments are submitted to make administrative changes tothe St. Lucie Unit 1 and Unit 2 Technical Specifications andachieve consistency throughout the Technical Specifications byremoving outdated material, making minor text changes, andcorrecting errors.DISCUSSION

The proposed amendments represent an administrative update to theSt. Lucie Units 1 and 2 Technical Specifications. Each of theproposed changes are discussed below:

St. Lucie Unit 1

1) Item: Page IIIAction:

In Section 3/4.1.2 on page III of the, Index an "s" should beadded after Charging Pump, the corrected revision should readCharging Pumps-Shutdown instead of Charging Pump-Shutdown.

Discussion:

Adding an "s" corrects the index reference and providesaccuracy and consistency with the rest of the listings in.Section 3/4.1.2'.

2) Item: Page 3/4 0-3

Action:

Correct Surveillance Requirement 4.0.5 part d. by inserting aperiod (".") at the end of the statement.

Discussion:

Adding a period at the end of the statement corrects thesentence grammatically.

3) Item Page 3/4 1-12

Action:

On Page 3/4 1-12 add an "s" after the word "pump" at the topof the page.

l lP

C),

r

Discussion:

This revision will correct a typographical error.Item: Page 3/4 3-41

Action:

Correct Limiting Condition For Operation Section 3.3.3.8 byinserting a period (".") at the end of the statement.

Discussion:

Adding a period at the end of the statement corrects thesentence grammatically.

Item: Page 3/4 4-l,bAction:

Correct Limiting Condition For Operation Section 3.4.1.3 part(b) by changing the period (".") at the end of the statementto a comma (",").Discussion:

Changing the period to a comma at, the end of the statement inpart b provides a clearer transformation from part b to partc. By placing a comma after parts a,b and c, and a periodafter part d a more consistent format is established.

Item: Page 3/4 4-8

Action:

Change Surveillance Requirement 4.4.5.5 a to read, "Within 15days following the completion of each inservice inspection ofsteam generator tubes, the number of tubes plugged in eachsteam generator shall be reported to the Commission in aSpecial Report pursuant to Specification 6.9.2. ChangeSurveillance Requirement 4.4.5.5 b to read, "The completeresults of the steam generator tube inservice inspection shallbe submitted to the Commission in a Special Report pursuant toSpecification 6.9.2 within 12 months following completion ofthe inspection.Discussion:.

This change will allow the Unit 1 Technical Specj.fications tobe consistent with the Unit 2 Technical Specifications.Item Page 3/4-14

a

/

tp

l~

F,c

p

Action:,

Add the words "Surveillance Requirements" above the doubleline to accurately label Surveillance Requirement 4.4.6.2.This revision will accurately allow the reader to identifythat below the double line is the appropriate SurveillanceRequirement.

Discussion:

This addition willmake the format consistent with the rest ofthe Surveillance Requirements in the Unit 1 TechnicalSpecifications.

8) Item Pages 3/4 6-53/4 6-203/4 6-21

Action:

Delete the new valve tag numbers, and any references to them,as requested in Florida Power & Light's (FPL) Proposed LicenseAmendment letter L-87-123 and replace them with the existingvalve tag numbers.

Discussion:

FPL is returning to the Technical Specifications that existedprior to the proposed modification described in FPL letter L-87-123. The modification was planned as an enhancement to theBreathing Air System to save time and provide operationalflexibility during a refueling outage. The penetration isisolated during normal operation. It has been determined thatthis modification is economically unjustified. No detrimentaleffects to the Breathing Air System will result and the systemwill remain as originally designed. Minor valve numberdiscrepancies are corrected.

9) Item Page 3/4 8-6b

Action:

Correct misspelling of the word "accumulated."

Discussion:

None.

10) Item Page 3/4 11-2

Action:

Delete the dash ("-") in the page number.

Discussion:

This revision corrects a typographical error.Item Page 3/4 11-10

Action:

Correct misspelling of the word "measurable".

Discussion:

None.

Item Page 3/4 12-10

Action:

Change the referenced Technical Specification from "6.9.1.11"to "6.9.1.8". This change-.vill accurately identify the AnnualRadiological Environmental Operating Report in Chapter 6, theAdministrative Controls Section, of the Unit 1 TechnicalSpecifications.Discussion:

In Chapter 6 of the Unit 1 Technical Specifications, Section6.9.1.11 presently does not exist. Specification 6.9.1.8 isthe correct Specification.Item Page B 3/4 1-4

Action:

-Correct statement (1) in Bases Section 3/4.1.3 by adding an"r" below the "T" in the symbol F g which represents thesteady-state radial peak.

Discussion:

This revision will amend a typographical error and accuratelyrepresent the steady-state radial peak. symbol as shown.

Item Page B 3/4 7-5

Action:

In Bases Section 3/4.7.7 the last sentence should referenceGeneral Design Criteria 19 of Appendix A, 10 CFR 50 instead ofGeneral Design Criterion 10 of Appendix "A", 10 CFR 50.

Discussion:

Pli,, i )

4

P 1» s f t l twW

Criterion 19 of Appendix A, 10 CFR 50 pertains to the ControlRoom and specifically states that, "Adequate radiationprotection shall be provided to permit access and occupancy ofthe control room under accident conditions without personnelreceiving radiation exposures 'in excess of 5 REMs whole body,or its equivalent to any part of the body, for the duration ofthe accident." This correction will accurately reference theproper Design Criteria.

15) Item Page B 3/4 9-3

Action:

Change the word "exceed" to "excess".Change the word "insure" to "ensure".

Discussion:

These changes will correct the appropriate sentencegrammatically.

16) Item Page 6-24

Action:

In Administrative Controls Section 6.15.1 part (c) change theperiod (".") at the end of the statement and replace it witha semi-colon (" ") .

Discussion:

By making this revision Sections 6.15.1 parts (a) through (h)will read grammatically correct. The only statement in thissection that should end with a period is the last statement,

~ "Section 6.15.1 part (h).

ST. LUCIE UNIT 2

1) Item Pages 2-42-5

Action:

Add the referenced note "(1)" after the word "pressure" inTable 2.2-1 part 4., Thermal Margin/Low Pressure. Add thereferenced note "(1)" after the word "Low" in Table 2.2-1part 14., Reactor Coolant Flow-Low.

Discussion:

Note (1), in Table 2.2-1, is the reference note for Zero PowerMode Bypass. Note (1) states, "Trip may be manually bypassedbelow 0.54 of RATED THERMAL POWER during testing pursuant to

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Special Test Exception 3.10.3; bypass shall be automaticallyremoved when the THERMAL POWER is greater than or equal to0.5> of RATED THERMAL POWER." The Zero Power Mode Bypassallows manual bypassing of Variable Powel Level-High, ThermalMargin/Low Pressure (including ASGT), Steam Generator PressureDifference — High and the Reactor Coolant System (RCS) LowFlow. The addition of these two references will correct Table2.2-1, to include all of the above trips.

2) Item Page 3/4 2-9

Action:.

In Surveillance Requirement 4.2.3.2 part (a) change the period(".") at the end of the statement to a comma (",").Discussion:

This revision will correct the statement grammatically.

3) Item Pages 3/4 3-2

Action:

Add the referenced note "(a)" after the "2" and before the"(d)" in Table 3.3-1 part 4., Thermal Margin/Low Pressureunder the column of Channels to Trip. Add the referenced note"(a)" after'the "2/SG" and before the "(d)" in Table 3.3-1part 14., Reactor Coolant Flow-Low under the column ofChannels to Trip.Discussion:

Note (a), in Table 3.3-1, is the reference note for Zero PowerMode. Bypass. ,Note (a) states, "Trip may be manually bypassedbelow 0.5> of RATED THERMAL POWER in conjunction with (d)below; bypass shall be automatically removed when THERMAL,POWER is greater than or equal to 0.54 of RATED THERMALPOWER." Note (d), as referenced above, states, "Trip may bebypassed during testing pursuant to Special Test Exception3.10.3." The Zero Power Mode Bypass allows manual bypassingof Variable Power Level-High, Thermal Margin/Low Pressure(including ASGT), Steam Generator Pressure Difference-High andthe Reactor Coolant System (RCS) Low Flow. The addition ofthe these two references will correct Table 3.3-1, to includeall of the above trips.

4) Item Page 3/4 3-13

Action:

Add a close parenthesis (")") after the word "buttons" inTable 3.3-3 Part 4 (a).

Discussion:

This revision willgrammatically correct the reference to TripButtons.


Action:

Delete the word "of" from the sentence and add the word "to"to the sentence.

Discussion:

This revision will correct the sentence grammatically.

6) Item Page 3/4 6-26

Action:

Change the words "set point" to "setpoint".Discussion:

The word "setpoint" should be one word instead of two.


Action:

Delete the close parenthesis (")") at the end of paragraph 7.

Discussion:

This revision will correct the sentence grammatically.


Action:

Delete the comma (",") after Amendment No. 48 at the bottom ofthe page.

Discussion:

This revision will eliminate a typographical error.9) Item Page 3/4 11-10

Action:

Correct misspelling of the word "measurable".

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Discussion:

None

10) Item Page 3/4 12-10

Action:

Change the referenced Technical Specification from "6.9.1.11"to "6.9.1.8". This change will accur'ately identify the AnnualEnvironmental Operating Report, in Chapter 6 (theAdministrative Controls Section) of the Unit 2 TechnicalSpecifications.Discussion:

Specification 6.9.1.11 presently does not exist in Chapter 6of the Unit 2 Technical Specification. Specification 6.9.1.8is the correct Specification and this correction needs to b'

«»'-made:to accurately reflect the proper reference to the AnnualEnvironmental Operating Report.

11) Item Page 3/4 6-25

Action:

Correct misspelling of the word "support".Discussion:

None

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ATTACHMENT 4

DETERMINATION OF NO SIGNIFICANT HAZARDS

The standards used to arrive at a determination that a requestfor amendment involves no significant hazards considerationare included in 10 CFR 50.92 which states that no significanthazards considerations are involved if the operation of thefacility in accordance with the proposed amendment would not(1) involve a significant increase in the probability orconsequences of an accident previously evaluated; or (2)create tne possibility of a new or different kind of accidentfrom any accident previously evaluated; or (3) involve asignificant reduction in a margin of safety. Each standard isdiscussed as follows:

(1) Operation of the facility in accordance with the proposed'",— amendment would not involve a significant increase in the

probability or consequences of an accident previouslyevaluated.

Administrative changes to the Technical Specifications donot affect assumptions in plant safety analysis, nor dothey affect Technical Specifications that preserve safetyanalysis assumptions. The corrections and clarificationsreferenced in this proposed license amendment do notaffect the probability or consequences of accidentspreviously analyzed.

(2) Operation of the facility in accordance with thisproposed license amendment would not create thepossibility of a new or different kind of accidentpreviously evaluated.

A new or different kind of accident is not created sinceTechnical Specification LimitingConditions for Operation,(LCO) and ACTION statement requirements remain unchanged.

The administrative changes to the Breathing Air Systembeing proposed by FPL willnot lead to material procedurechanges or to physical modifications to the St. LuciePlant. The Breathing Air System proposed modificationwas not completed due to budgetary considerations. TheTechnical Specifications are being returned to those thatexisted prior to the proposed modification.'inor valvenumbering corrections are being made. Therefore, theproposed changes do not create the possibility of a newor different kind of accident.

(3) Operation of the facility in accordance with the proposedamendment would not involve a significant reduction in amargin of safety.

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The changes being proposed do not relate to or modify thesafety margins defined in and maintained by the TechnicalSpecifications. Therefore, the proposed changes wouldnot involve any reductions in a margin of safety.

Based on the above, we have determined that -the amendmentrequest does not (1) involve a significant increase in theprobability or consequences of an accident previouslyevaluated; or (2) create the possibility of a new or differentkind of accident from any accident previously evaluated; or(3) involve a significant reduction in a margin of safety; andtherefor does ,not involve a significant hazardsconsideration.

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attachment st. unit technical specifications · marked-up st. lucie unit 1 technical specifications...

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