Forwards comments on proof & review version of Tech Specs ...

ACOKLILRATED DISTRIBUTION DEMONSTRATlON SYSTEM

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8811100265 DOC.DATE: 88/11/02 NOTARIZED: NOFACIL:50-250 Turkey Point Plant, Unit 3, Florida Power and Light C

50-251 Turkey Point Plant, Unit 4, Florida Power and Light CAUTH.NAME AUTHOR AFFILIATION

CONWAY,W.F. Florida Power & Light Co. ~6~RECIP.NAME RECIPIENT AFFILIATION ~e"o1V

Document Control Branch (Document Control Desk)

SUBJECT: Forwards comments on proof & review version of Tech Specsissued to util on 880609,in form of marked-up pages.

DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR J ENCL J SIZE:TITLE: OR Submittal: General DistributionNOTES

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RECIPIENT'D

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INTERNAL: ARM/DAF/LFMBNRR/DEST/CEB 8HNRR/DEST/MTB 9HNRR/DOEA/TSB 11NUDOCS-ABSTRACT

01

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LTTR ENCL1 01 1y'

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NOPE M ALL "RIDS" RECIPIENIS:

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P. O. 4000, JUNO BEACH, FL 33408-0420

4i X+

NOVEMBER 2 1988

L-88-478

U. S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, D. C. 20555

Gentlemen:

Re: Turkey Point Units 3 and 4Docket Nos. 50-250 and 50-251Revised Technical SpecificationsProof and Review Comments

The purpose of this letter is to transmit Florida Power 6 Light Company (FPL)comments on the Proof and Review version of the revised Technical Specificationsfor Turkey Point Units 3 and 4. This Proof and Review version was issued to FPLwith an NRC letter dated June 9, 1988.

The FPL comments are provided in a marked up form in the attachment to thisletter. Justification sheets for non-editorial comments are provided, which areindexed to the numbers adjacent to the individual comments.

Further comments, as a result of issues that were identified late in the reviewprocess, are expected to be provided during the comment resolution period.These comments are expected to involve control room emergency ventilation andintake cooling water strainers.

The numerical values for items such as pump flows, tank levels, and setpointsare being reviewed through a parallel effort as we have previously discussed.Should any of these values change as a result of our review, we will advise theTechnical Specification Branch and the Project Manager.

As previously discussed with the staff, we have not provided comments on Section3/4.8 - Electrical Power Systems and this section has been deleted from oursuomittal. A separate license amendment on this subject is being prepared whichwe will be providing later this month. Upon approval of this amendment, it willreplace our previous submittal of Section 3/4.8.

ADOCg »0~<'OSOOOZgo

po~»Ilo<

(((an FPL Group company

U. S. Nuclear Regulatory CommissionL-88-478Page two

IAt our last meeting with the staff a milestone schedule for the remainder of thisproject was discussed. The period between November 4, 1988 and January 6, 1989was provided for resolution of comments. Due to the extent and complexity ofour comments, we would like to work with the staff to develop a working scheduleto detail the process that will be necessary to reach this next milestone. Weare prepared to provide the support necessary to meet this milestone. Thedetails of the schedule to meet the January 6, 1989 date can be discussed withthe FPL manager for this project, Mr. J. Arias, at (305) 246-6007.

Very truly yours,

W. . CoSenior Vice resident - Nuclear

WFC/PLP/gp

Attachment

cc: Malcolm L. Ernst, Acting Regional Administrator, Region II, USNPCSenior Resident Inspector, USNRC, Turkey Point Plant

SECTION 1.0

DEFINITIONS

TURKEY POINT - UNITS 3 8 4 1-0

-PQ

~Ij s. I .

pc~ ., ~ > ge 77ov al/( //~

0

DEFINITIONS

SECTION

1 . 0 DEFINITIONS..;................................................~ 1 ACTIONe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~1

1.2 ACTUATION LOGIC TEST.............................,............1.3 ANALOG CHANNEL OPERATIONAL TEST...............................1.4 AXIAL FLUX DIFFERENCE.........................................1. 5 CHANNEL CALIBRATION...........................................1.6 CHANNEL CHECK......................... ~ .......................1.71.8

CONTAINMENT INTEGRITY...............CORE ALTERATION.....................

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.9l. 10

l. 11

l. 12

DOSE EQUIVALENT I-131..............E-AVERAGE DISINTEGRATION ENERGY

FREQUENCY NOTATION.................GAS DECAY TANK SYSTEM..............

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ pt ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1. 13

1. 14

IDENTIFIED LEAKAGE...........................................MASTER RELAY TEST............................................

1. 15

1. 16

MEMBER(S) OF THE PUBLIC.;..........OFFSITE DOSE CALCULATION MANUAL....~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~

1. 17

1. 18

OPERABLE - OPERABILITY.............OPERATIONAL MODE - MODE............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~

l. 19

1. 20

1. 21

1. 22

1. 23

1. 24

PHYSICS TESTS................................................

PROCESS CONTROL PROGRAM......................................P URGE PURGINGo ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~

QUADRANT:,ONER TILT, RATIO..........RATED T L. P56................

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

PRESSURE BOUNDARY LEAKAGE....................................

1.261. 27

EPO RTNK EVElfFo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~R

HUTDOO MARGINo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o' ~ ~ ~ ~ ~ ~'o

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~S

1. 28 SITE BOUNDARY.....'...................................;.......1.29 SLAVE RELAY TEST..

c4uu e.a t mis To E

Fird|OL t VC-A ~< @ S /56CS,TURKEY POINT - UNITS 3 4 4

INOEX

OEF INITIONS

SECTION

1. 30

l. 31

1. 32

1. 33

1. 34

1. 35

1. 36

1. 37

1. 38

TABLE

TABLE

SOLIDIFICATION.........................SOURCE CHECK...........................STAGGERED TEST BASIS.........................................

HERMAL POWER.................T

TRIP ACTUATING DEVICE OPERATIONAL TEST.......................

UNRESTRICTED AREA..............VENTILATION EXHAUST TREATMENT SYSTEM.........................VENTING ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.1 FRE(UENCY NOTATION................1. 2 OPERATIONAL MODES..........,......

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

UNIDENTIFIED LEAKAGE.........................................

1"6

1-6

1-6

1-61-61-61-6

1-7

1-81-9

C'URKEY

POINT - UNITS 3 Ea 4JUK 09 19<a

INDEX

SAFETY LIMITS ANO LIMITING SAFETY SYSTEH SETTINGS

SECTION

2.1 SAFETY LIMITS

2. 1. 1 REACTOR CORE..............,.....,.........,,.....,,.........2. 1.2 REACTOR COOLANT SYSTEM PRESSURE.............................FIGURE 2.1"1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION..

2.2 LIHITING SAFETY SYSTEM SETTINGS

2.2.1 REACTOR TRIP SYSTEM INSTRUHENTATION SETPOINTS...............

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUHENTATION TRIP SETPOINTS....

PAGE

2-1

2-1

2-2

2-3

2-4

BASES

SECTION PAGE

2. 1 SAFETY LIMITS

2. 1.1 REACTOR CORE................................................2.1.2 REACTOR COOLANT SYSTEH PRESSURE.............................

B 2-1

B 2-2

2.2 LIMITING SAFETY SYSTEM SETTINGS

2.2.1 REACTOR TRIP SYSTEM. INSTRUMENTATION SETPOINTS.............. B 2-3

~ ~

~ I

~ ~

'TURKEY POINT - UNITS 3 4 4gUg i0 N8

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS

SECTION

3/4. 0 APPLICABILITY...............................................PAGE

3/4 0-1

Flow Path - Shutdown................................Fl ow Paths - Operating.......... ~ . ~... - .. ~ " - " . ~ = ~ ~

Charging Pumps - Operating..........................Borated Water Source - Shutdown......,..............

~ ~ ~ ~ 4

~ ~ ~ ~ ~

Borated Water Sources - Operating........................Heat Tracsng..........................-.-....-.

3/4. 1.3 MOVABLE CONTROL ASSEMBLIES

roup Hei ghta ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ a ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~G

TABLE 3.1-1 ACCIDENT ANALYSES RE(UIRING REEVALUATION IN THEEVENT OF AN INOPERABLE FULL-LENGTH ROD..............--

Position Indication Systems - Operating.............TABLE 4. 1-1 ROD POSITION INDICATOR SURVEILLANCE RE/UIREMENTS.

Position Indication System - Shutdown...............i

Rod Orop Time.......................................Shutdown Rod Insertion Limit............. " . """"Control Rod Insertion Liaits.....-...---......-... "

FIGURE 3.1-2 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER

THREE-LOOP OPERATION........-...............-...

3/4. 1 REACTIVITY CONTROL SYSTEMS

3/4. 1. 1 BORATION CONTROL ~ /I ~d dr~p<)r~Shutdown Nargin -g Greater Than 2 OoF................

FIGURE 3.1"1 REQUIRED SHUTDOWN MARGIN VERSUS REACTOR COOLANTBORON CONCENTRATION.....................,.............

Shutdown Margin - + , Less Than or Equal to g004F.......Moderator Temperature Coet'tie)ent.4 ......................Minimum Temperature for Criticality......................

3/4. 1. 2 BORATION SYSTEMS

3/4 1-1

3/4 .1" 33/4 1-4 )f3/4 1-5

3/4 1-7

3/4 1-8

3/4 1-9

3/4 1-11

3/4 1-12

3/4 1"14

3/4 1-16

3/4 1-17

3/4 1-19

3/4 1-20

3/4 1-22

3/4 1-23

3/4 1-24

3/4 1-25

3/4 1-26

3/4 1-27

TURKEY POINT - UNITS 3 Ec 4 IVJUN OS l98S

INDEX

LIMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE RE UIREMENTS

SECTION'I

'/4. 2 POWER DISTRIBUTION LIMITS

3/4.2.1 AXIAL FLUX DIFFERENCE..........................FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION

RATED THERMAL POWER......................:.3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR...................

OF

~ ~ ~ ~ ~ ~ ~ ~ ~ ~

FIGURE 3.2-2 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT.

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.................3/4.2.4 QUADRANT POWER TILT RATIO................................3/4.2.5 ONB PARAMETERS...;.......................................TABLE 3. 2-1 DNB PARAMETERS........................................

PAGE

3/4 2-1

3/4 2-3

3/4 2"4

3/4 2-5

3/4 2-8

3/4 2-11

3/4 2-13

3/4 2-14

3/4. 3 INSTRUMENTATION

3/4.3. 1 REACTOR TRIP SYSTEM INSTRUMENTATION'.;....................TABLE 3. 3-1 REACTOR TRIP SYSTEH INSTRUMENTATION...................TABLE 4. 3-1 REACTOR TRIP SYSTEH INSTRUMENTATION SURVEILLANCE

REQUIREMENTS...................'.......................~ ~

~

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEHINSTRUMENTATION..........................................

TABLE 3. 3-2 ENGINEEREO SAFETY FEATURES, ACTUATION SYSTEMINSTRUMENTATION..........................................

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEHINSTRUMENTATION TRIP SETPOINTS...........................

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4. 3, 3 MONITORING INSTRUMENTATION

Radfatfon, Honftorfng For Plant Operatfons......:.........

3/4 3-g3/4 3-2

3/4,3-8

3/4 3-13

3/4 3-14

3/4 3-22

3/4 3-28

3/4 3-34

TURKEY POINT - UNITS 3 S 4 V .-4UN i~ ee

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEIL'LANCE REUIREMENTS'ECTION

TABLE 3-3-4 RADIATION MONITORING INSTRUMENTATIONFOR PLANT OPERATIONS........................,.....

TABLE 4. 3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANTOPERATIONS SURVEILLANCE RE(UIREMENTS.....................Movable Incore Detectors.................................Accident Honitoring Instrumentation......................

TABLE 3. 3-5 ACCIDENT MONITORING INSTRUMENTATION...................TABLE 4.3-4 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE

REQUIREMENTS.............................................Fire Detection Instrumentation.............;.............

TABLE 3. 3-6 FIRE DETECTION INSTRUMENTS............................Radioactive Liquid Effluent Honitoring Instrumentation...

TABLE 3.3-7 RADIOACTIVE LI(UID EFFLUENT MONITORING INSTRUHENTATION

TABLE 4.3-5 RADIOACTIVE LIQUID EFFLUENT MONITORINGINSTRUMENTATION SURVEILLANCE REQUIREMENTS................Radioactive Gaseous Effluent Monitoring Instrumentation..

TABLE 3.3-8 RADIOACTIVE GASEOUS EFFLUENT MONITORINGINSTRUMENTATION......................................-...

TABLE 4.3-6 RADIOACTIVE GASEOUS EFFLUENT HONITORINGINSTRUMENTATION SURVEILLANCE RE(UIREHENTS................

E

3/4 3-35

3/4 3-38

3/4 3-39

3/4 3-40

3/4 3-41

3/4 3-45

, 3/4 3-48

3/4 3-49

3/4 3-51

3/4 3-52

3/4 3-54

3/4 3-55

3/4 3-56

3/4 3-59

TURKEY POINT - UNITS 3 8s 4 VISUN ie tyg

INOEX

LIHITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREHENTS

SECTION

3/4.4 REACTOR COOLANT SYSTEH

3/4.4. 1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

PAGE

3/4.4. 2

Startup and Power Operation..............................ot Standby........................................H

ot Shutdown ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~H

Cold Shutdown - Loops Filled.............................Cold Shutdown - Loops Not Filled.........................SAFETY VALVES

3/4 4-1

3/4 4-2

3/4 4-3

3/4 4-5

3/4 4-6

hutdown ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~S

perating.............................................0

344.3/4 4.34 ~ 4o 3 P ESS RIZERo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~

LOC K3 /4. 4. 4 VALVES~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.4. 5 STEAM GENERATORS...................;........;............TABLE 4.4-1 MINIHUH NUMBER OF STEAM GENERATORS TO BE INSPECTED

DURING INSERVICE INSPECTION.............................TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................3/4. 4. 6 REACTOR COOLANT SYSTEM LEAKAGE

Leakage Detection Systems.............'...................Operational Leakage......................................

TABLE 3.4-1 REACTOR COOLANT SYSTEH PRESSURE ISOLATION VALVES......3/4o 4o 7 CHEMISTRYo ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

TABLE 3.4-2 REACTOR COOLANT SYSTEH CHEMISTRY LIMITS...............TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE

RE(UIREHENTSo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.4. 8 SPEC@/'CTIVHY'........................................FIGURE 3.4-1 . E(VIVALENT I-131 REACTOR COOLANT SPECIFIC

A LIHI'F VERSUS PERCENT OF RATED THERHAL POWER

WI . REACTOR COOLANT SPECIFIC ACTiVITY> AC)/grasDOSE EQUIVALENT I-3.31..............,.......,...............

TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAHPLE AND ANALYSISP ROG RAMo ~ ~ ~ ~ ~ ~ ~

'~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o

3/4 4-7

3/4 4-8

3/4 4-9

3/4 4-1O j ~3/4 4-1X

3/4 4-16

3/4 4-17

3/4 4-18

3/4 4-19

3/4 4-21

3/4 4-22

3/4 4-23

3/4 4-24

3/4 4-25,

3/1 4-26

3/4 4-27

TURKEY POINT - UNITS 3 4 4 VII JUN 00 )gas

INOEX


Pressurizer.Overpressure>

3/4.4. 10 STRUCTURAL

~ J' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

I &in Systems................. ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.4. 11 REACTOR COOLANT SYSTEM VENTS.......................-.....

SECTION

3/4.4. 9 PRESSURE/TEHPERATURE LIMITS

Reactor Coolant System...............................FIGURE 3.4-2 UNIT 3 REACTOR COOLANT SYSTEM HEATUP LIHITATIONS-

APPLICABLE UP TO 10 EFPY.................................FIGURE 3.4-3 UNIT 3 REACTOR COOLANT SYSTEM COOLDOWN LIHITATIONS-

APPLICABLE UP TO 10 EFPY.................................FIGURE 3.4-4 UNIT 4 REACTOR COOLANT SYSTEH HEATUP LIHITATIONS-

APPLICABLE UP TO 10 EFPY.................................FIGURE 3.4-5 UNIT 4 REACTOR COOLANT SYSTEM COOLDOWN LIHITATIONS-

APPLICABLE UP TO 10 EFPY.................................TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM

WITHDRAWAL SCHEDULE.............................

3/4 4"29

3/4 4-30

3/4 4-31

3/4 4-32

3/4 4-33

3/4 4-34

3/4 4-35

V4 4-33) 8.3/4 4-38

3/4 4"40

3/4. 5 EMERGENCY CORE COOLING SYSTEHS

3/4.5.1 ACCUMULATORS 3/4 5-1

3/4.5. 2

3/4.5. 3

Rcs ANWA6E cc E34 A~ ~KP~~e-FECCS SUBSYSTEHS - ~ GREATER THAN OR EQUAL TO 3SO~F....

e,cs Awe~~ Mo~ur W&v p~~~a-E,ECCS SUBSYSTEHS - ~ LESS'HAN 35O~F...................

3/4 5-3

3/4 5-7

3/4.5.4 REFUELING WATER STORAGE TANK..........."""" 3/4 5-8

TURKEY POINT - UNITS 3 4 4 VIIIJUN 00 1988

PR

INOEX


SECTION

3/4.6 CONTAINMENT SYSTEMS ATMOSPHERIC TYPE CONTAINMENT

PAGE

Air Temperature..........................Containment Structural Integrity.........Containment Ventilation System...........

~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ )0 ~ ~

3/4. 6. 2 DEPRESSURIZATION AND COOLING SYSTEMS

Containment Spray System.................Containment Cooling System.....'..........

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4. 6. 3 EMERGENCY CONTAINMENT FILTERING SYSTEM ..................

'3/4. 6. 1 PRIMARY CONTAINMENT

Containment Integrity.....................,...............Containment,Leakage......................................Containment Air Locks.....................,..............Internal Pressure........................................

3/4 6-1

3/4 6-2

3/4 6-4

3/4 6"6

3/4 6-7

3/4 6-8

3/4 6-11

3/4 6-12

3/4 6-fO

3/4 6-15

~ ~

~

3/4.6.4 CONTAINMENT ISOLATION VALVES.......'.....TABLE 3.6-1 CONTAINMENT ISOLATION VALVES.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ 1 ~

3/4 6-17

3/4 6-19

3/4.6.5r

=Hydrogen Monitor ........................................ 3/4 6-22

3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM.................... 3/4 6-23

TURKEY POINT - UNITS 3 4 4 JUN Ok 1988

PR

INDEX


3/4.7.23/4.7.33/4.7.43/4. 7. 5

3/4.7. 6

3/4.7. 7

COMPONENT COOLING WATER SYSTEM..........................-INTAKE COOLING WATER SYSTEM.........................-....ULTIMATE HEAT SINK......................-....-...........CONTROL ROOM VENTILATION SYSTEM................ -...- - -...NUBBERS.................................................S

SEALED SOURCE CONTAMINATION..............................

SECTION'/4.

7 PLANT SYSTEMS

3/4.7. 1 TURBINE CYCLE

afety Valves............................................S

TABLE 3.7"1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGHSETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURINGFOUR LOOP OPERATION......................'................

TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP.....................Auxiliary Feedwater System...........................

TABLE 3.7"3 AUXILIARYFEEDWATER SYSTEM OPERABILITY............Condensate Storage Tank..............................SPecific Activity................................... ~

TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLEAND ANALYSIS PROGRAM.............;...................Main Steam Line Isolation Valves.....................Standby Feedwater System.............................

3/4 7-1

3/4 7-2

3/4 7"2

3/4 7-3

3/4 7-5

3/4 7-6

3/4 7-8

3/4 7-9

3/4 7-103/4 7-11

3/4 7-12

3/4 7-14

3/4 7-15

3/4 7-16

3/4 7-18

3/4 7-22

~

'URKEY

POINT - UNITS 3 4 4 SUN OQ 1gse

INDEX

LIMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE RE UIREMENTS

SECTION

3/4.7.8~ ~ FIRE SUPPRESSION SYSTEMS

PAGE

Yard Fire Hydrants and Hydrant Hose Houses....TABLE 3.7-5 FIRE HYDRANTS.......'.......................

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4. 7.9 FIRE RATED ASSEMBLIES....................................

Fire Suppression Water System.................Spray and/or Sprinkler Systems................Fire Hose Stations.............,..............

TABLE 3.7-4 FIRE HOSE STATIONS....................................

3/4 7-24

3/4 7-27

3/4 7-29

3/4 7-30

3/4 7-31

3/4 7-32

3/4 7-33

3/4.8 ELECTRICAL POWER SYSTEMS

/4.8.1 .C. SOURCES

0 p ating......................;........... .............TABLE 4.8-1 DI L GENERATOR TEST SCHEDULE.....

TABLE 4.8-2 DIESEL ENERATOR TEST FREgUENC

~ ~

~

~ ~

Shutdowno ~ t ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ao ~ ~ ~ ~ ~ 0 ~

3/4.8.2 D.C. SOURCES

perating............0

TABLE 3. 8-1 BATTERY CHARGE LLOWABL UT-OF-SERVICE TIMES........, TABLE 4.8-3 BATTERY SU ILLANCE REQUIR S.....................

hutdown1 ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ 0 ~ ~ ~S

3/4.8. 3 ONSITE WER DISTRIBUTION

per tinge ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~0

TABLE 3.8- NTOR'CONTROL. CENTER ALLOWABLE OUT-OF-SERVICE TIS h ~ ~ ~ ~ ~ e ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4 B-X.',

3/4 8-8

3/4 8-9

3/4 8-10

3/4 8-11

3/4 8-12

3/4 8-14

3/4 8-15

3/4 8-16

3/4 8-19

3/4 8-20

(A~t

TURKEY POINT - UNITS 3 4 4 XIJUN 0 S 1988

INDEX


SECTION

3/4. 9 REFUELING OPERATIONS

3/4. 9. 1

3/4. 9. 2

BORON CONCENTRATION......................................

INSTRUMENTATION..........................................

3/4 9-1

3/4 9-2

3/4.9.3 DECAY TIME............................................... 3/4 9-3

3/4. 9.4 CONTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4

3/4. 9. 5 COMMUNICATIONS...........................................

3/4. 9. 6 MANIPULATOR CRANE...............;........................

3/4 9-5

3/4 9-6

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS........'.......... 3/4 9-7

~ ~3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION

H'gh Water Level......................................... 3/4 9"8

Low Water Level...................-...-....-..---- .. " . 3/4 9-9

'TURKEY POINT - UNITS 3 4 4 XII JUN 09 1988

INOEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREHENTS

~ ~

SECTION

3/4. 11.2 GASEOUS EFFLUENTS

PAGE

0 ose Rate.............................................-.. 3/4 11 '7

Dose - Iodine-131, Iodine-133, TritiMaterial in Particulate Form........Gaseous Radwaste Treatment System...Explosive Gas Mixture...............

um, and Radioactive

as Decay Tanks..........................................G

TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSISROGRAMo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~P

Dose - Noble Gases.......................................3/4 11-8

3/4 11-12

3/4 11-13

3/4 11"14

3/4 11"15

3/4 11-16

3/4.11. 3 SOLID RADIOACTIVE WASTES...........;.................... 3/4 11-17I

3/4. 11. 4 TOTAL DOSE............................................... 3/4 11-18

3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING

3/4. 12. 1 HONITORING PROGRAM............................... -.......TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM........TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITYCONCENTRATIONS

IN ENVIRONMENTAL SAHPLES.....'............... -............TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE

ANALYSISo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.12.2 LAND USE CENSUS...............'............ "...........3/4.12.3 INTERLABORATORY COMPARISON PROGRAM........ " . "..... "-..

3/4 12-1

3/4 12-3

3/4 12-7

3/4 12-8

3/4 12-11

3/4 12-13

TURKEY POINT - UNITS 3 da 4 XIV gag a0 1888

INOEX


SECTION

3/4. 9. 9 CONTAINMENT VENTILATION ISOLATION SYSTEM.................

PAGE

1

3/4 9-10

3/4.9.10 WATER LEVEL - REACTOR VESSEL............................. 3/4'9-11

3/4.9. 11 WATER LEVEL - STORAGE POOL .............................. - 3/4 9-12

3/4.9. 12 HANDLING OF SPENT FUEL CASK..............................

3/4. 9. 13 RADIATION MONITORING.....................................

3/4.9. 14 SPENT FUEL STORAGE......................................,

3/4 9-13

3/4 9-14

3/4 9-15

TABLE F 9. 1 SPENT FUEL BURNUP REQUIREMENTS. FOR STORAGE IN

REGION II OF THE SPENT FUEL PIT..........................., 3/4 9-15

3/4. 10 SPECIAL TEST EXCEPTIONS

3/4.10.1 SHUTDOWN MARGIN..........................;............... 3/4 10-1

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2

3/4. 10. 3 PHYSICS TESTS............................................ 3/4 10-3

3/4 10-4

3/4. 10. OSITION INDICATION SYSTEM - SHUTDOWN.................... 3/4 10-5

3/4. 11 RADIO

3/4.11.1 LIQ

Co

Vf EFFLUENTS

LUENTS

400llor ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-1

TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND'ANALYSISROGRAMo ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~P

Oseo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~D

Liquid Radwaste Treatment Systea..................;......

3/4 11-2~ 3/4 11-5

3/4 11-6

'TURKEY POINT - UNITS 3 Sc 4 XIII. JUN Sr >as'

BASES

INDEX

PAGE

3/4. 0 APPLICABILITY............................................... B 3/4 0-1

3/4.1 REACTIVITY CONTROL SYSTEMS

3/4. 1. 1 BORATION CONTROL..........................................3/4. 1.2 BORATION SYSTEMS..........................................3/4. 1.3 MOVABLE CONTROL ASSEMBLIES................................

B 3/4 1-1

B 3/4 1-2

B 3/4 1-4

3/4.2 POWER DISTRIBUTION LIMITS..............................,....3/4.2.1 AXIAL FLUX DIFFERENCE.............

FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL'LUXDIFFERENCE VERSUSTHERMAL POWERo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.........

3/4.2.4 QUADRANT POWER TILT RATIO.................................3/4.2.5 ONB PARAMETERS............................................

B 3/4 2-1

B 3/4 2-X

B 3/4 2-0

B 3/4 2-4

B 3/4 2-5

B 3/4 2-6

3/4. 3 INSTRUMENTATION

3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETYFEATURES ACTUATION SYSTEM INSTRUMENTATIOM.......,... B 3/4 3-1

3/4.3.3 MONITORING INSTRUMENTATION.............................. B 3/4 3-3

TURKEY POINT - UNITS 3 8L 4 XV JUN C$ t9<:.

BASES

INDEX

SECTION

3/4.4 REACTOR COOLANT SYSTEM

3/4.4.1 REACTOR COOLANT LOOPS ANO COOLANT CIRCULATION............. B 3/4 4-1

3/4.4.8 'PECIFIC ACTIVITY.........................3/4.4. 9 PRESSURE/TEMPERATURE LIMITS...............TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS - UNIT 3.

TABLE B 3/4.4-2 REACTOR VESSEL TOUGHNESS - UNIT 4.

3/4.4. 10 STRUCTURAL INTEGRITY.....................3/4.4. 11 REACTOR COOLANT SYSTEM VENTS.............3/4. 5 EMERGENCY CORE COOLING SYSTEMS

~ ~ ~ ~ ~ l ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.5. 1 ACCUMULATORS...................-.....-...........3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS...................- ".----....3/4.5.4 REFUELING MATER STORAGE TANK..................

3/4.4.2 SAFETY VALVES................

3/4. 4. 3 PRESSURIZER...................................... -........3/4.4. 4 RELIEF VALVES.............................................3/4.4.5 STEAM GENERATORS..........................................3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................3/4.4.7 CHEMISTRY.................................................

B 3/4 4-2

B 3/4 4-2

B 3/4 4"3

B 3/4 4-3

B 3/4 4-4

B 3/4 4-5

B 3/4 4-5

8 3/4 4-7

B 3/4 4-9

B 3/4 4-10

B 3/4 4-16

B 3/4 4-16

B.3/4 5-1

B 3/4 5-1

B 3/4 5-2

TURKEY POINT - UNITS 3 Ei 4 XVIJUN 09 1SSS

BASES

INOEX

SECTION

3/4.6 CONTAINMENT SYSTEHS

3/4. 6. 1 PRIMARY CONTAINMENT.......................................3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEHS...................

PAGE

8 3/4 6-1

B 3/4 6-3

3/4. 6. 3 EMERGENCY CONTAINMENT FILTERING SYSTEM.................... B 3/4 6-3

3/4.6.4 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-3

3/4.6. 5 COMBUSTIBLE GAS CONTROL................................... B 3/4 6-4

3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM..................... B 3/4 6-4

TURKEY POINT - UNITS 3 4 4 XVII'UNot Ms'

BASES

INDEX

SECTION,

3/4. 7 PLANT SYSTEHS

3/4. 7. 1 TURBINE CYCLE............... "............'............... B 3/4 7-1

3/4.7.2 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-5

3/4.7.3 INTAKE COOLING WATER SYSTEM............................... B 3/4 7-5

3/4.7.4 ULTIMATE HEAT SINK........................................3/4.7.5 CONTROL ROOM VENTILATION SYSTEM...........................3/4. 7. 6 NUBBERS......; ...........................................S

3/4.8 ELECTRICAL POWER SYSTEMS

3/4.7.7 SEALED SOURCE CONTAMINATION.........................-'.....3/4.7. 8 FIRE SUPPRESSION SYSTEMS..................................3/4.7.9 FIRE RATED ASSEHBLIES................................ - -...

B 3/4 7-5

,B 3/4 7-6

B 3/4 7-6

B 3/4 7-7

B 3/4 7-8

03/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and

ONSITE POWER DISTRIBUTION................................. B 3/4 8-1

TURKEY POINT - UNITS 3 8L 4 XVIII JUN 0 ~ 1988

BASES

INDEK

PAGE


3/4. 9. 1 BORON CONCENTRATION..........................,............

3/4. 9. 2 INSTRUMENTATION....................,......................

3/4o9o3 DECAY TIMEo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~

8 3/4 9-1

8 3/4 9-1

8 3/4 9-1

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................: 8 3/4 9-1

3/4. 9. 5 COMMUNICATIONS..................................;.........

3/4. 9o 6 MANIPULATOR CRANE.........................................3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS...................3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.............

3/4.9.9 CONTAINMENT VENTILATION ISOLATION-SYSTEM..................

8 3/4 9-1

8 3/4 9-2

8 3/4 9-2

8 3/4 9-4

8 3/4 9-2

3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSELSTORAGE POOL..............................

AND

8 3/4 9-3

3/4.9. 12 HANDLING OF SPENT FUEL CASK............................... 8 3/4 9-3

3/4. 9.13 RADIATION MONITORING......................................3/4 9 14 SPENT + FUEL STORAGE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

8 3/4 9-3

8 3/4 9-3


3/4. 10. 1 SHUT% MARGINS...........................................3/4.10.2 GRO $GST, INSERTION, AND POWER DISTRIBUTION LIMITS....

3 /4o 10o 3 PHY TESTS'e ~ ~ e ~ ~ ~ ~ ~ ~ ~ 'e ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.10.4 (Thfs spec$ ffcat$ on nueber $ s not used)...................

8 3/4 10-1

8 3/4 10-1

8 3/4 10-1

8 3/4 10-1

3/4.10.5 POSITION INDICATION SYSTEM - SHUTDNM..................... 8 3'/4 10-1

TURKEY POINT - UNITS 3 4 4 XIX',

BASES

INDEX

SECTION

3/4. 11 RADIOACTIVE EFFLUENTS

3/4. 11. 1 LI(UID EFFLUENTS....................3/4.11.2 GASEOUS EFFLUENTS................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.11. 3 SOLID RADIOACTIVE WASTES................................3/4. 11.4 TOTAL DOSE..............................................3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING

B 3/4 11-1

B 3/4 11-3

B 3/4 11-6

B 3/4 11-6

3/4. 12. 1 MONITORING PROGRAM..................3/4.12. 2 LAND USE CENSUS.....................3/4. 12.3 INTERLABORATORY COMPARISON PROGRAM.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

B 3/4 12-1

B 3/4 12-1

B 3/4 12-2

TURKEY POINT - UNITS 3 4 4 XX JL/N 0 ~ 1S88

INDEX

DESIGN FEATURES

SECTION0 '.1. 1 EXCLUSION AREA,..................................,.....,...,

PAGE

5-15. 1. 2 LOW POPULATION ZONE......................................... 5-15.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY

RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS..........FOR~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-1

5. 2 CONTAINMENT

5.2.1 CONFIGURATION...............................................5.2.2 DESIGN PRESSURE AND TEMPERATURE.............................FIGURE 5.1-1 SITE AREA MAP........................................

5-15-1

5-2

5. 3 REACTOR CORE

5.3. 1 FUEL ASSEMBLIES.. ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~

5.3.2 CONTROL ROD ASSEHBLIES......................................5-$5-3

5.4 REACTOR COOLANT SYSTEM

5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-3.4.2 VOLUME......................................................5 5-3

5.5 HETEOROLOGICAL TOWER LOCATION................................. 5-3

5.6 FUEL STORAGE

5. 6. 1 CRITICALITY.;............................................... 5-45 ~ 6e 2 DRAINAGEe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-5

5 .6.3 CAPACI ~ I ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-5

5. 7 COMPONE CVCt.IC OR TRANSIENT LIMIT........................... 5-5

TABLE 5.7-1 CONTINENT CYCLIC OR TRANSIENT LINITS...,............... 5-6

TURKEY POINT - UNITS 3 Sc 4 XXI " JUN 0 0 ]egg

INOEX

ADMINISTRATIVE CONTROLS

SECTION

6. 1 RESPONSIBILITY.......,,........,.......,...,...,,...,.......PAGE

6-1"

6. 2 ORGANIZATION,...~........,..............,...,,...........,... 6-1

6. 2. 1 ONSITE AND OFFSITE ORGANIZATION........................... 6-16.2.2 UNIT STAFF.............. 6-2TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION...................... 6-46.2.3 SHIFT TECHNICAL ADVISOR................................... 6"5

6. 3 FACILITY STAFF UALIFICATIONS............................... 6-5

6 . 4 TRAINING.................................................... 6-5

6o5 REVIEW AND AUDITo ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-5

6.5. 1 'LANT NUCLEAR SAFETY COMMITTEE

unct) Ono ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~F 6-5C

~ ~Ompas 1 t) On ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

lternates.........................--......-".-...""".A

6-6

6-6M teetsng Frequency......................................... 6-6uorum.................................................... 6-6

Respanslb I llties................. ". - ~ " ~ ~ " - ~ -""~ - ~ - ~ ~ ~

6-6'cardSo

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~R 6-8

~of

''TURKEY POINT - UNITS 3 5 4 XXII JUt4 CS 19gg

INDEX


~ ~

SECTION

6.5..2 COMPANY NUCLEAR REVIEW BOARD

unctiono ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~F 6-8

Composition... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8

lternates..................................,.....A ~ ~ ~ ~ ~ ~ ~ ~ ~ < ~ 6-8

onsultants........C 6-9

M t.eetlng Frequency......................................... 6-9

uorum. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-9

Revsew...............A d tUdlts ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6-9

6-10

ecords ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~R 6-11

6.6 REPORTABLE EVENT ACTION...........!....'..................... 6-11

6. 7 SAFETY LIMIT VIOLATION...................................... 6-11

6.8 PROCEDURES AND PROGRAMS..................................... 6-12

6. 9 REPORTING RE UIREMENTS...................................... 6-14

6.9. 1 ROUTINE REPORTS......................................;.... 6-14I ESStartup Report............................................ 6-14

Annual Reports...........................Semiannual Radioactive Effluent Release

Monthly Operating Report................Radial Peek)ng Factor Limit Report......

6.9.2 SPECIAREPORTS.........................

Reporto ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6-15

6-17

6-18

6-18

6-19

6. IO RECORD RETENTION............................................ 6-19

TURKEY POINT - UNITS 3 L 4 XXIII


INOEX

SECTION

6. 11 RADIATION PROTECTION PROGRAM............................... 6-21'

6.12 HIGH RADIATION AREA........................................ 6-21

6.13 PROCESS CONTROL PROGRAM PCP .............................. 6-22

6. 14 OFFSITE DOSE CALCULATION MANUAL ODCM ...................... 6-22

6. 15 MAJOR CHANGES TO LI UID GASEOUS AND SOLID

RADWASTE TREATMENT SYSTEMS................................. 6-23

TURKEY POINT - UNITS 3 4 4 XXIV JUN 09 198~

p~ g~~ ~c ~OLD g 6 /<<USE~l. 0 DEFINITIONS g~q~ gj)Q;Qs.M &~6S~-

The defined terms of this section appear in capitalized type and are applicablethroughout these Technical Specifications.

ACTION

l. 1 ACTION shall be that part of a Technical Specification which prescribesremedial measures required under designated conditions.

ACTUATION LOGIC TEST

1.2 An ACTUATION LOGIC TEST shall be the application of various simulatedinput combinations in conjunction with each possible interlock logic state andveri ficati on of the required 1 ogi c output. The ACTUATION LOGIC TEST shal 1

include a continuity check, as a minimum, of output devices.

ANALOG CHANNEL OPERATIONAL TEST

1.3 An ANALOG CHANNEL OPERATIONAL TEST. shall be the injection of a simulatedsignal into the channel as close to the sensor as practicable to verifyOPERABILITY of alarm, inter lock and/or trip functions. The ANALOG CHANNELOPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-

'ockand/or Trip Setpoints such that the setpoints are within the requiredrange and accuracy,

AXIAL FLUX OIFFERENCE ( t- )1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signalsbetween the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION

1.5 A CHANNEL.CALIBRATION shall be the adjustment, as necessary, of thechannel such that it responds within the required range and accuracy to knownvalues of input. The CHANNEL CALIBRATION shall encompass the entire channelincluding the sensors and alarm, interlock and/or trip functions and may beperformed by any series of sequential, overlapping, or total channel stepssuch that the entire channel is calibrated.

CHANNEL CHECK'~';"

1.6 A CHANNEKCHECK shall be the qualitative assessment of channel behaviorduring operation by observation. This determination shall include, wherepossible, comparison of the channel indication and/or status with otherindications and/or statuq derived from independent instrument channelsmeasuring the same paramiter.

JUN 09 >'.;i

TURKEY POINT - UNITS 3 4 4

CsuTRC ILES /gk'~GQi '74 ~I'7P~u.& 0'LCSeC~ s/ J/ /~ 4~v ~~./ ~Per r'~~~ ~

/'l e.re ~Per zeal», ('~ ge../s.DEFINITI NS

CONTAINMENT INTEGRITY

1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditionsare either: t

1) Capable of being closed by an OPERABLE containment automaticisolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automaticvalves secured in their closed positions, except as provided inTable 3.6-1 of Specification 3.6.4.

b. The equipment hatch is closed and sealed,

C.

d.

e.

Each air lock is in compliance with the requirements of Specification3.6.1.3,

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

The sealing mechanism associated with each penetration (e. g., welds,bellows, or 0-rings) is OPERABLE.

CORE ALTERATIONS

1.8 CORE ALTERATIONS shall be the movement or manipulation of any componentwithin the reactor pressure vessel with the vessel head removed and fuel inthe vessel. Suspension of CORE ALTERATIONS shall not preclude completion ofmovement of a component to a safe conservative position.

DOSE E UIVALENT I-131

1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microCurie/gram)which alone would produce the same thyroid dose as the quantity and isotopicmixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroiddose conversion factors used for this calculation shall be those listed inTable III of TI0-14844, "Calculation of Distance Factors for Power and TestReactor Sites".-'.or Table E-7 of NRC Regulatory Guide 1.109, Revision 1,October 1977; "

f - AVERAGE DISINTEGRATION ENERGY

1.10 Z shall be the average (weighted in proportion to the concentration of each .

radionuclide in the reactor coolant at the time of sampling) of the s~ of theaverage beta and gamma energies per disintegration (HeV/d) for the radionuclidesin the sample isotopes, other than iodines, with half lines- greater than 15 minutes,making up at least 95 percent of the total non-iodine activity in the coolant.

TURKEY POINT - UNITS 3 8a 4 1-2 JUN OQ 188t:

DEF INITIONS

FREIR E CY TATION.

l. 11 The FRE(UENCY NOTATION specified for the performance of SurveillanceRequirements shall'orrespond to the intervals defined in Table 1.1.

GAS DECAY TANK SYSTEM

l. 12 A GAS DECAY TANK SYSTEM shall be any system designed and installed toreduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup forthe purpose of reducing the total radioactivity prior to release to theenvironment.

IDENTIFIED LEAKAGE

1.13 IDENTIFIED LEAKAGE shallbe:( (~ccp1 COV1ROLi'Eb LEAK866~~

a. Leakag to closed systems, such as pump seal or valve packing leaks Jthat are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are bothspecifically located and known either. not to interfere with theoperation of Leakage Detection Systems or not to be PRESSURE

BOUNDARY'EAKAGE,or

c. Reactor Coolant System leakage through a steam generator to theSecondary Coolant System.

'of OPE

e

MEMBER S OF THE PUBLIC

1.15 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant.. This category does not include employeesof the licensee, its contractors, vendors or members of the Armed Forces usingproperty locat~within the SITE, BOUNDARY. Also excluded from this categoryare persons wh4Ventar the site to service equipment or to make deliveries.This category ~ include persons who use portions of the site for- recre-ational, occupaffonal, or other purposes not associated with the plant..

OFFSITE DOSE CALCULATION MANUAL

1.1S The OFFSITE 00SE CALCULATION MANUAL (ODCM) shall contain the methodologyand parameters used in the calculation of offsite doses due to radioactivegaseous and liquid effluents, in the calculation of gaseous and liquideffluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

TURKEY POINT - UNITS 3 8E 4 1-3

DEFINITIONS

OPERABLE - OPERABILITY

1.17 A system, subsystem, train, component or device shall be OPERABLE orhave OPERABILITY when it is capable of performing its specified function(s),and when all necessary attendant instrumentation, controls, electrical power,cooling or seal water, lubrication or other auxiliary equipment that arerequired for the system, subsystem, train, component, or device to perform itsfunction(s) are also capable of performing their related support'unction(s).

OPERATIONAL MODE - MODE

1.18 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusivecombination of core reactivity condition, power level, and average reactorcoolant te"verature specified in Table 1.2.

PHYSICS TESTS

1. 19 PHYSICS TESTS shall be those tests performed to measure the fundamentalnuclear characteristics of the reactor core. and related instrumentation:(1) described in Chapter 13.5 of the FSAR, (2) authorized under theprovisions of 10 CFR 50.59, or (3) otherwise approved by the Comnission.

PRESSURE BOUNDARY LEAKAGE

1.20 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tubeleakage) through a nonisolable fault in a Reactor Coolant System componentbody, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM

1.21 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas,sampling, analyses, 'tests, and determinations to be made to ensure thatprocessing and packaging of solid radioactive wastes based on demonstrated

'rocessing of actual or simulated wet solid wastes will be accomplished insuch a way as to assure compliance with 10 CFR Parts 20, 61, and 71 andFederal and State regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.

PURGE - PURGI II

1.22 PURGE ~PURGING shall be any controlled process of discharging air or gasfrom a confinement to maintain temperature, pressure, humidity, concentrationor other operating condition, in such a manner that'eplacement air or gas isrequired to purify the confinement.

TURKEY POINT - UNITS 3 4 4 1-4 JUh' 5 195j

DE F INITION 5

UADRANT POWER TILT RATIO

1.23 QUADRANT POWER TILTdetector calibrated out u

RATIO shall be the ratio of the maximum upper excorep t to the average of the upper excore detector cali-

brated outputs, or the ratio of the maximum lower excore detector calibratedoutput to the average of the lower excore detector calibrated outputs, whicheveris greater. With one excore detector inoperable, the remaining three detectorsshall be used for computing the average.

RATED THERMAL POWER

1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate tothe reactor coolant of 2200 MWt.

ENCE POSITION

REPORTABLE EVENT

1.26 A REPORTABLE EVENT shall be any of those conditions specified inSection- 50.73 of 10 CFR Part 50.

SHUTDOWN MARGIN

1.27 SHUTDOWN MARGIN shall be the instantaneous amount of'reactivity by whichthe reactor is subcritical or would be subcritical from its present conditionassuming all fo)l-length rod cluster assemblies (shutdown and control) arefully insertiexcept for the single rod cluster assembly of highest reactivity.worth which $~smed to be fully withdrawn.

SITE BOUNDARY'

''.28

The SITE BOUNDARY:shall be that line beyond'which the land is neitherowned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST

1.29 A SLAVE RELAY TEST shall be the energization of each slave relay andverification of OPERABILITY of each relay. The SLAVE RELAY TEST shall includea continuity check, as a minimum,: of associated tes~le actuation devices.

1. 25 Analo d Position Indication System REFERENCE POSITIO defined as:

a. For all Shu Banks and Control Banks A B; the group demandcounter indicate ition between 0 0 steps withdrawn inclusiviand between 200 and 2 s wi wn inclusive.

b. For Control Banks C a ; the gr emand counter indicated positionbetween 0 and 30 ps withdrawn inclus d between 150 and 228

~ steps withdr inclusive. For the withdrawa e of 31 to 149steps i sive the REFERENCE POSITION shall be the idual rodcal ation curve noting indicated analog rod position vers ndicated

oup demand counter position.

TURKEY POINT - UNITS 3 4 4 1-5 JUN 0 9 19»:

DEFINITIONS

SOLIDIFICATION

1. 30 SOLIDIFICATION shallmeets shipping and

SOURCE CHECK4$ c "c-<c'<

1.31 A SOURCE CHECK shallwhen the channel sensor is

be the conversion of wet wasteth inty a form thatlk aPPbr~48 /zgeHSingr~ ~sr'>+~a O r~Co~ »ice.

dbe the qualitative assessment of channel responseexposed to a source of increased radioactivity.

STAGGERED TEST BASIS

1.32 A STAGGERED TEST BASIS shall consist of:

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

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

THERMAL POWER

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

TRIP ACTUATING DEVICE OPERATIONAL TEST

1.34 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating theTrip Actuating Device and verifying OPERABILITY of alarm, interlock and/ortrip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall includeadjustment, as necessary, of the Trip Actuating Device such that it actuatesat the required setpoint within the required accuracy

UNIDENTIFIED LEAKAGE

1.35 UNI ENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIEDLEAKAGE Or COPQOLCEO ~P,~UNRESTRICTED.

1.36 An UNREINECTN AREA shall be any area at or beyond the SITE BOUNDARYaccess to which is not controlled by the licensee. for purposes of protection ofindividuals frois exposure to radiation and radioactive aaterials, or any areawithin the SITE BOUNDARY used for residential quarters. or for industrial,-commercial, institutional,, and/or recreational purposes.-

VENTILATION EXHAUST TREATMENT SYSTEM

1.37 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any systea designed and"installed to reduce gaseous radioiodine 'or radioactive material in particulateform in effluents by passing ventilation or vent exteust gases through charcoal

TURKEY POINT - UNITS 3 4 4 1-6.pcc<; t', 9 )QQrc

DEFINITIONS

VENTILATION EXHAUST TREATMENT SYSTEM (Continued)

adsorbers and/or HEPA filters for the purpose of removing iodines or particulatesfrom h ex u stream~rice to the release to the environment. Such

, ysteq is not onsldered to have any ef ct on noble gas effluents. Engineeredafetysneaturas tmospheric peanupgystemd are not considered to be VENTILATION )Q

EXHAUST TREATME SYSTEM compon nts.

VENTING

1.38 VENTING shall be the controlled process of discharging air or gas from aconfinement to maintain temperature, pressure, humidity, concentration, or otheroperating condition, in such a manner that replacement air or gas is not pro-vided or required during VFNTING. Vent, used in system names, does not implya VENTING process.

sa~ a+

TW

TURKEY POINT - UNITS 3 Ec 4 1-7 JUN 00 1988

TABLE 1.1

FRE UENCY NOTATION

NOTATION FRE UENCY

At least once per 12 hours.


At least once per 7 days.

,no-grea

SA

S/U

N.A.




At least once per refuelingnA'rior

to each reactor startup.

Not applicable.

Completed prior to each release.


TABLE 1.2

OPERATIONAL MODES

MODE

1. POWER OPERATION

2. STARTUP

3. HOT STANDBY

4. HOT SHUTDOWN

5. COLD SHUTDOWN

6. REFUELING"*

REACTIVITYtt Kt

> 0.99

> 0.99

< 0.99

< 0.99

< 0.99

QR5

X RATEDTHERMAL POWER"

> 5X

< 5X

AVERAGE COOLANTTEMPERATURE

> 350'F

> 350'F

> 350 F

350 F > T> 200 F

< 200 F

<140F

"Excluding decay heat.""Fuel in the reactor vessel with the vessel head closure bolts less than fully

tensioned or with the head removed.

. TURKEY POINT - UNITS 3 8I 4 1-9.' 0S 1S88

/Vl 5 n /V~p

/~ 5!<.k,

a1

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 SAFETY LIMITS

~

~

REACTOR CORE

2. l. 1 'The combination of THERMAL POWER, pressurizer pressure, and the highestoperating loop coolant temperature (T ) shall not exceed the limits shown inavgFigure 2. 1-1, for 3 loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loopaverage temperature and THERMAL POWER has exceeded the appropriate pressurizerpressure line, be in HOT STANDBY within 1 hour, and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE

2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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

ACTION:

MODES 1 and 2:

'Whenever the Reactor Coolant System pressure has exceeded 2735 psig, bein HOT STANDBY with the Reactor Coolant System pressure within its limitwithin 1 hour, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig,reduce the Reactor Coolant System pressure to within its limit within5 minutes, and coeply with the requirements of Specification 6.7.1.

TURKEY POINT - UNITS 3 8c 4 2-1JL.. 0 S 1988

tte

ZcgpP$ ]'y

e;SO>$

< ScS

ScS

o tl~~sa .

coopP$ f~

g AHVCO

8. .s .~ .s. .a .v .I.POvEI f free i'teoaini t l- ~-

FIGURE'.1-1

REACTOR CORE SAFETY LINIT- THREE LOOPS IN OPERATION

TURKEY POINT - NITS 3 4 4 2-2 JUY 0> 1868

SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS

REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

2.2.1 .The Reacto~ Trip System Instrumentation and Interlock Setpoints shallbe set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

Wi a Reac or Trip stem Ins umentation r Interl ck Set ointss conse vative t an the va e shown in he Trip etpoi col n

ut more onserva ve than e value sho in th liow e V uecolumn o Table .2-1, adj t the setpo nt cons tent th t e Tripset oin value.

p. Nlth the Reactor Trip System Inetrumentetlon or Interlock Setpointless conservative than the value shown in~i P y.h-l, P *1 h h 1 i p hh P pphy happlicable ACTION statement requirement of Specification 3;3.1 untilthe channel is restored to OPERABLE status with its Setpoint adjustedconsistent with the Trip Setpoint value.

TURKEY POINT - UNITS 3 4 4 2-3J„;i C < 19~i

TABLE 2.2-1REACTOR TRIP SYSTEH INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT

1. Hanual Reactor TripI

C 2. Power Range, Neuta. High Setpoinl .

b. Low Setpoint\

3. Interaediate Range,Neutron Flux

4. Source Range, Neutron Flux

5. Overteaperature hT

6. Overpower hT

7. Pressurizer Pressure-Low

8. Pressurizer Prcssure-High

9. Pressurizer Mater Level-High

10. Reactor Coolant Flow-Low

TRIP SETPOINT

N.A.

<1QQ of RTP*~

<25K of RTP*"

<25K of RTP"*

<10s cps

See Note 1

See Not 2-

>1835 psig

<2385 psig

<92K of instreaent span

>90X of looplesign flow"

AL WABLE ALUE

.A.

of Pa*

eX RTP*<

<3 of RT *

<1.3 x Os cp

See ote 2

Se Note .

> 825 sig

<23 psi

of nstr

>8% f loopdesi n flow

t span

11. Stem Generator MaterLevel Low-Low

>15K of narrow rangeinstrument span

>14% of rrow rang instr entspan /

Loop design flow = 89,500 gpa"*RTP = RATEO THERHAL POWER

ib

TABLE 2. 2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

g FUNCTIONAL UNIT

12. Steam/Feedwater FlowHiseatch

Q Coincident Mith

Steam Generat MaterQe Level-L

TRIP SETPOINT

Feed Flow<0.64 x 10 lb/hrbelow steam flow

5X of narrowange instrument span F

ALLOWABL VANE

Feed ow<0.7 x 106 lb/bel steam fl<1 of nar ow range i trumentspan

13. Undervoltage - 4.16 kVBusses A and B

14. Underfrequency - Trip of ReactorCoolant Pump Breaker(s)'pen

15. Turbine Trip

a. Auto Stop Oil Pressure

b. Turbine Stop ValveClosure

16. Safety In)ection Inputfroa ESF

17. Reactor Trip SysteaInterlocks

a. Interaediate RangeNeutron Flux, P-6

>2496 volts-each bus

>56.1 Hz

>45 psig

Fully Closed ¹

N. A.

>1x10 ao am

>2456 ol ts-eac us

.0 Hz

>40 ig

Ful y Close

.A.

x 10-'i mp"

Liant switch is set when Turbine Stop Valves are fully closed.

TABLE 2. 2-1 Continued

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT

b. Low Power Reactor TripsBlock, P-7

1) P-104npu0-,. ~~et2) Tugine Firyt Stage

Pressure

c. Power Rangy HeyttonFlux'-8

d. P.ower Range NeutronFlux, P-10

I

18. Reactor Coolant PuepBrqaker Popjfion Trip

19. Reactor $I'ip greaQrs

20. Autowatic Trip and InterlockLogfq

TRIP SETPOINT

<gg of RTP**

<lOX Turbine Power

<45K of RTP*"

>10X of RTP*+

N.A.

N.A.

N.A.

ALLO LE V UE

<1 of R *

T ine P Ger

<46 of RTP

of P*"

N.A.

N.

**RTATED THERMAL POWER

NOTE 1: OVERTNPERATURE hTP

AT(—) <AT (K-I 1+x) oI 1

C

1Qo ~+Xf

I

Kg

1+v S+ Y

~ Le K

bvf ''T.'f".:f~

T

.TABLE 2. 2-1 Continued

TABLE NOTATIONS

- K ~ (T ( ) — T'] + K (P - P') — f (A())2 (1 + t3S) 1 + xgS 3

Heasured hT by RTD Instrumentation;

Lag compensator on measured hT;

Time constants utilized in the lag compensator for AT, xg = 2 s;

Indicated 4T at RATED THERMAL POWER;

ms, ). ops0.0107/OF;

The function generated by the lead-lag compensator for Tdynaeic compensation; avg

f„

eo constants utilized in the lead-lag compensato~ for T „, x2 =25s,'/=

35'rage

temperature, OF;

Lag ciipensator on measured T

avg'ime

constant utilized in the measured T lag compensator, t< = 2 s.avg

574.2'F (Nominal T at RATED THERMAL POWER);

0.000453/psig;

Pressurizer pressure, psig;K y, +lb ~)e,gl


TABLE NOTATIONS Continued

NOTE 1: (Continued)CIM

IC

Qo

D

pl 2235 psig (Nominal RCS operating pressure);

S : ,, —. Laplace transform operator, s-~;

/,aand f '(4I) is a function of the indicated difference between top and bottom detectors of thepower-range neutron ion chambers; with gains to be selected based on measured instrumentresponse during plant startup tests such that:

(1) For qt -qb bebeen -. 14K and + 10', f (AI) = 0, where qt and qb are percent RATED THERMAL

POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL

POWER in percent of RATED THERMAL POWER;

ICO

(2) For each percent that the magnitude of qt -qb exceeds - 14K, the hT Trip Setpoint shall

be automatically reduced by 2.0X of its value at RATED THERMAL POWER; and

(3) For each percent that the magnitude of qt -qb exceeds + 10K, the AT Trip Setpoint shall

be autoaatically reduced by 3.5X of its value at RATED THERMAL POWER.

NOTE 2: The nq 'axiam Tr t oint shall t ts co ted Setpoi by mo h

0TABLE 2. 2-1 Continued


NOTE : OVERPOWER dT

AT ( 1 .)(N N

(~tS ) ( 1 ) T ( 1 )(1+ ts$ )o e s (1+ sss) (1+ ses) e (1+ ass)

s ~~ ~ s

~e

%here: . hT = As defined in Note 1,

As defined in Note 1,1

4T

+ ss~sS

As defined in Note 1,


1.09,

0.02/4F for increasing average teaperature and 0 for decreasing averagetemperature, s

The function generated by the rate-lag coapensator for T dynamiccoipensation, avg

Tiae constants utilized in. the rate-lag compensator for T x = 10avg's

def)ned in Note 1,




NOT : (Continued)

Kg .

', js)~.1

T .',, ~"

OO gNQo

f (aI)

0.00068/ F for T > T" and Kz = 0 for T < T",


Indicated T „ at RATED THERHAL POWER (Calibration temperature for bT

instruaentation, < 574.2 F),

As defined in Note 1, and

As defined in Note 1.

NOTEW

he channel's i Trip Setpoi all not exce ts computed' point b're an

2. CL

BASES

FOR

SECTION 2.0

SAFETY LIMITS

AND

LIMITING SAFETY SYSTEM SETTINGS

NOTE

The BASES contained in succeeding pages summarizethe reasons For the Specifications in Section 2.0,but in accordance with 10 CFR 50.36 are not partof these Technical Specifications.

TURKEY POINT - UNITS 3 81 4 B 2-0

0

I5 'pA.ri8 In.78nlidvlz(l~/ g

4p~

2. 1 SAFETY LIMITS

BASES

2.1. 1 REACTOR CORE

The restrictions of this Safety Limit prevent overheating of the fueland possible cladding perforation which would result in the release of fissionproducts to the reactor coolant. Overheating of the fuel cladding is preventedby restricting fuel operation to within the nucleate boiling regime where theheat transfer coefficient is'arge and the cladding surface temperature isslightly 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 (DNB). and the resultant sharp reduction in heat transfercoefficient. DNB is not a directly measurable parameter during operation andtherefore THERMAL POWER and reactor coolant temperature and pressure have beenrelated to DNB. This relationship has been developed to predict the ONB fluxand the location of DNB for axially uniform and nonuniform heat flux distribu-tions. The local ONB heat flux ratio (ONBR)'is defined as the ratio of theheat flux that would cause DNB at a particular cort location to the local heat:t.flux and is indicative of the margin to ONB.

'

The ONB design basis fs as follows: there must,bc at least a 95 percentprobability with 95 percent confidence that the minimus ONBR of the limitingrod during Condition I and II events is greater than or equal to the ONBRlimit of the DNB correlation being used. The correlation ONBR limit isestablished based on the entire applicable experimental data set such thatthere is a 95 percent probability with 95 percent confidence that ONB willnot occur when the minimum DNBR is at the ONBR limit.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER,Reactor Coolant System pressure and average temperature for which the minimumONBR is no less than the design ONBR value, or the average enthalpy at thevessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F~, of 1.62 andN

and a reference cosine with a pegk of 1.55 for axial power shape. An allowanceis included for ah increase in F~ at reduced power based on the expression:

F~ < 1. Q+ 0.3 (I-P)3

1&ere P iFthe-fraction of RATED THERMAL PNER.

These limiting heat flux conditions are higher than those calculated forthe range of all control:rods fully withdrawn to th» maxim'llowable controlrod insertion limit assuming the axial power imbalance is within the l.imits ofthe f (dI) function of the Overtemperature trip. %hen the axial power imbalanceis not within the tolerance, the axial power imbalance effect on the Overtemperature hT trips will reduce the setpo$ nts. to provide protection consistentwith core Safety Limits.

TURKEY POINT - UNITS 3 4 4 B 2-1 JUN 35 )9>:

SAFETY LIMITS

BASES

2. 1. 1 REACTOR CORE (Continued)

Fuel rod bowing reduces the values of DNB ratio (DNBR). .The amount ofthe DNBR reduction is 4.7X for LOPAR fuel with the j.-grid ONB correlation and5.5X for the OFA,fuel with the WRB-1 DNB correlation. The penalties arecalculated pursuant to "Fuel Rod Bow Evaluation," WCAP-8691-P-A Revision 1(Proprietary) and WCAP-8692 Revision 1 (Non-Proprietary). The restrictionsof the Core Thermal Hydraulic Safety Limits assure that an amount of DNBRmargin greater than or equal to the above penalties is retained to offset therod bow DNBR penalty.

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE

The restriction of this Safety Limit protects the integrity of the ReactorCoolant System (RCS) from overpressuriiation and thereby prevents the release iof radionuclides contained in the reactor coolant froa reaching the containmentatmosphere. J!rggf< 4t;Cl f~

0, est~

The reactor vessel ressurizeelv mstd7he RCS p!p!ng, valves and flttlngsare es gne to w c e s a maximum transient ressure 20Kof desi n pressure of 2485 ps! Sect!on III of odc for Nuclear

ower Plan s permfte a max>mum trans!ent pressure of 110K (2735 pslg) of design gressure. The afety Limit of 2735 psig is therefore more conservative thane esxgn cri eri nd consisten iated"Co

~

~

rements.'84< < 4<gL ~ i%5» ASIDE

e entire ydrotested at 125K (3107 psig) of design pressure, todemonstrate integrity prior to initial operation.

TURKEY POINT - UNITS 3 Cc 4 B 2-2 JUtf'7 fi l9»~

2. 2 LIMITING SAFETY SYSTEM SETTINGS

BASES

PR~gg~)o4a gns$ru~C cl riA

opcraYiun+ ge.sly~ +E ~~ t~,„$ ,g nw-n ~lay

4- pr1C!M COYlSI2YV~ 1VC, IM'4JAAA

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINT mqpoiry a.ll~~

The Reactor Trip Setpoint Limits specified in Table . - are the ~~values at which the Reactor trips are set for each functional unit. The TriSetpoints have been selected to ensure that the core and Reactor Coolant i~i iSystem are prevented from exceeding their safety limits during normal operaand design basis anticipated operational occurrences and to assist the Engi«neered Safety Features Actuation System in mitigating the consequences ofaccidents.

o acco odate e instr ent drift assume to occur etween ope tional< tes and t accur cy to w ch setp nts can e measure and calib ted,~ Al owable alues r the actor T p Setpo ts have b n specifi in

T ble 2. -1. 0 ration th a tr p set le s conserv ive than i Tripetpoi but w'in it specifi Allowa e Value acceptabl on the b is

that e diff rence b tween e h Trip tpoint an the Allow e Value 'sequ to or ess th the dr t allow ce f tri s inc ding tho tri sas med in he tra ient s ety ana ses.

Cf'llN 4M ~

The methodology .to derive the Trip Setpo n s lin the channels. Inherent the determination of the

Trip Setpoints are the magnitudes of these channel uncertainties. Sensors andother instrumentation utilized in these channels are expected to be capable of

ig ill t 11 f 1 t y gt d

The various Reactor trip circuits automatically open. the Reactor tripbreakers whenever a condition monitored by the Reactor Trip System reaches apreset or calculated level. In addition to redundant channels and trains, thedesign approach provides a Reactor Trip System which monitors numerous systemvariables, therefore providing Trip System functional diversity. The functionalcapability at Qa specified trip setting is. required for those. anticipatory ordiverse Reac trips for which no direct credit was assed in the safetyanalysis to o the overall reliability of the Reactor Trip System. TheReactor Trip fnitiataL a Turbine trip signal whenever Reactor trip isinitiated. Thfi prevents the reactivity instrtion that would otheneise result .:from excessive Reactor. Coolant System cooldown and thus avoids unnecessaryactuation of the Engiaeered Safety Features hctuatfoo Syatea

Manual Reactor Tri

The Reactor Trip Systea includes 'manual Reactaw.tHp capability

TURKEY POINT - UNITS 3 4 4 1 2-3 Jub '. i 1b~t


BASES

Power Ran e Neutron Flux

. eIn each of the Power Range Neutron Flux channels there are two independent

bistables, each with its own trip setting used for a High and Low Range tripsetting. The Low Setpoint trip provides protection during subcritical and lowpower operations to mitigate the consequences of a power excursion beginningfrom low power, and the High Setpoint trip provides protection during poweroperations for all power levels to mitigate the consequences of a reactivityexcursion which may be too rapid for the temperature and pressure protectivetrips.

The Low Setpoint trip may be manually blocked above P-10 (a power levelof approximately 10K of RATED THERMAL POWER) and fs automatically reinstatedbelow the P-10 Setpoint.

Intermediate and Source Ran e NeutronFlux'he

Intermediate and Source Range, Neutron Flux trips provide coreprotection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal froa a subcrfticalcondition. These trips provide redundant. protection to the Low Setpoint tripof the Power Range, Neutron Flux channels. The Source Range channels wi11initiate a Reactor trip at about 10s counts per second unless manually blockedwhen P-6 becomes active. The Intermediate Range. channels will initiate aReactor trip at a current level equivalent to approximately 25K of RATED

H L POWER unless manually blocked when P-10 becomes active, No credit~~~ taken for operation of the trips associated with either the Intermediateource Range Channels fn the accident analyses; however, their functi'onal

capability at the specified trip settings is required by this specificationto enhance the overall reliability of the Reactor Protectfon System.

Qvertem erature hT

JUH 00 )g..B 2-4TURKEY POINT - UNITS 3 4 4

I

The Overteayerature ALT. trip provides core protection to prevent GNS forall combfnatfoo0t,of pressure, power, coolant temperature,'nd axial powerdistribution; jftovfdsd that the transient fs slow wfth respect to pfpfnytransit delays'.fisc the core to the taaperature detectors (about 4 seconds),and pressure fs within the rangI between the Pressurfzer. High and Low Pressuretrips. The setpofnt is automatically varfed with.". (1) coolant temperature tocorrect for temperature induced changes. fn density and heat capacity of waterand includes dynamic compensation for piping delays from the core to the looptemperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power dfstrfbutfon, this Reactor trip lfaft fs a1waysbelow the core Safety Limit as shown. in Figure 2.1-1. If axial peaks aregreater than design, as, indicated by the dtrferenoe between top and bottompower range nuclear detectors, the Reactor trip is automatically reducedaccording to the notations in Table 2.2-1.

se pin


BASES

The Overpower 4T trip prevents power density anywhere in the core fromexceeding 118K of the design power density. This provides assurance of fuelintegrity (e. g., no fuel pellet melting and less than IX cladding strain)under all possible overpower conditions, limits the required range for Over-temperature hT trip, and provides a backup to the High Neutron Flux trip. Thesetpoint is automatically varied with: (1) coolant temperature to correct fortemperature induced changes in density and heat capacity of water, (2) rate ofchange of temperature for dynamic compensation for piping delays from the coreto the loop temperature detectors, and (3) axial power distribution, to ensurethat the allowable heat generation rate (kW/ft) is not exceeded.

Pressurizer Pressure

In each of the pressurizer pressur'e channels, there are two independentbistables, each with its own trip setting to provide for a High and Low Pressuretrip thus limiting the pressure range in which reactor operation is permitted.'.The Low Setpoint trip protects against low pressure which could lead to DNB bye .

tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7(a power level of approximately 10K of RATED THERMAL POWER with turbine firststage .pressure at approximately lOX of full power equivalent); and on increasingpower, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizersafety valves to protect the Reactor Coolant System against system 4- Q

rpressure.

Pressurizer Water Level

The Pressurizer Mater Level-High trip is provided to prevent water reliefthrough the pressurizer safety valves. On decreasing power the PressurizerHigh Mater LI 'trip is automatically blocked by P-7 (a peer lovel ofapproximatel .of RATED THERMAL POMER with a turbine first stage pressureat approxima 1l% of full power equivalent); and on increasing power, auto-matically re d by P-7.

Reactor Coolant Flow

The Reactor Coolant Flow-Low trip provi e core protection to prevent DNb

by mitigating the consequences of a loss of ,resulting fmo-the loss of-one or more reactor coolant pumps..

On increasing power above P-7 (a powIr level of approxiaately 10K'ofRATED THERMAL POMER or a turbine first stage pressure at approximately 10K

TURKEY POINT - UNITS 3 dc 4 B 2-5

LIMITING SAFETY SYSTE> SETTINGS

BASES

Reactor Coolant Flow (Continued)'% ~ 1

of full power equivalent), an automatic Reactor trip will occur if the flow inmore than one loop drops below 90K of nominal full loop flow. Above P-8 (apower level of approximately 45X of RATED THERMAL POWER) an automatic Reactortrip will occur if the flow in any single loop drops below 90K of nominalfull loop flow. Conversely, on decreasing power between P-8 and the P-7 anautomatic Reactor trip will occur on low reactor coolant flow in more thanone loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level-

The Steam Generator Mater Level Low-Low trip protects the reactor fromloss of heat sink in the event of a sustained steam/feedwater flow mismatchresulting from loss of normal feedwater.'he specified setpoint providesallowances for star ting delays of the Auxiliary Feedwater System.

Steam/Feedwater Flow Mismatch and Low Steam Generator Mater Level

The Steam/Feedwater Flow Mismatch in coincidence with a Steam GeneratorWater Level-Lo trip is not used in the transient and accident analyses )~but is included n Table 2.2-1 to ensure the functional capability of thespecified trip settings and thereby enhance the overall reliability of theReactor Trip System. This trip is redundant to the Steam Generator MaterLevel Low-Low trip. The Steam/Feedwater Flow Mismatch portion of this trip isactivated when the steam flow exceeds the feedwater flow by greater an orequal to 0.64 x 10e lbs/hours The Steam Generator Mater Level-Loportion of the trip is activated when the water level drops belo , asindicated by the narrow range instrument. These trip values inclu e su c enallowance in excess of normal operating values to preclude spurious trips butwill initiate a Reactor trip before the steam generators are dry. Therefore,the required capacity and starting tiaa requirements of the auxiliary feedwaterpumps are reduced and the resulting thiraal transient on e Reactor CoolantSystem and s~ generators is minimized.

I,Ac. '$'e p(n

Undervolta e - 4.16 kV Bug A and S Tri s

The 4.16 kV Bus A and B Undervoltage trips provide core protection againstDNB as a result of complete loss of forced coolant flow. The specifiedsetpoint assures a Reactor trip signal is generated before the Low Flow TripSetpoint is reached. Time delays are incorporated.. in the Undervoltage tripsto prevent spurious Reactor trips from momentary electrical power transients.The delay is set so that the time required. for a signa) to reach the Reactortrip breakers following the trip of at least one undervoltage relay in both ofthe associated Units 4.16 kV busses shall not exceed 1.3 seconds. On decreasing

'URKEY POINT - UNITS 3 4 4 B 2-6 'UN ti9 )~


BASES

Undervolta e and - 4. 16 kV Bus A and B Tri s (Continued)

power the Undervoltage Bus trips are automatically blocked by P-7 (a powe~level of approximately 10K 'of RATED THERMAL POWER with a turbine first stagepressure at approximately 10K of full power equivalent}; and on iacreasingpower, reinstated automatically by P-7.

~pgaWnr 4v'ipA Turbine trip initiates a Reactor trip. On decreasing powe Turbine

trip is automatically blocked by P-7 (a power level of approximately 10K ofRATED THERMAL POWER with a 4urbine first stage pressure at approximatelylOX of full power equivalent); and on increasing power, reinstated automatic-ally by P-7.

Sa fet In 'ecti on In ut from ESF

If a Reactor trip has not already been generated by the Reactor TripSystem instrumentation, the ESF automatic actuation logic channels willinitiate a Reactor trip upon any signal which initiates a Safety In)ection.The ESF instrumentation channels which initiate a Safety Injection signal areshown in Table 3.3-3.

Reactor Coolant Pum Breaker Position Tri

The Reactor Coolant Pump Breaker Position Trips are anticipatory tripswhich provide reactor core protection against, DNB'. The open/close positiontrips assure a reactor trip signal is generated before the low flow tripsetpoint is reached. No credit was taken in the accident analyses for operationof these trips Their functional capability at the open/close position settingsis required to enhance the overall reliability of the Reactor Protection System.Above P-7 (a power level of approximately 10K of RATED THERMAL POWER or aturbine first stage pressure at approximately 10% of full power equivalent) anautomatic reactor trip will occur if more than one reactor coolant pump breakeris opened. Above P-8 (a power level of approxiaately 45% of RAT THERMAL Q EPOWER) an automatic reactor trip will occur if one reactor coolant p rea eris opened. On."jftcreasing power between P-8 and P-7, an automatic reactor tripwill occur if mme than one reactor coolant pump breaker is opened and belowP-7 the trip f&ation is automatically blocked.

Underfrequency sensors are also installed on the 4.16 kV busses to detectunderfrequency and initiate breaker trip on underfrequency. The underfrequencytrip setpoints preserve the coast down energy of the reactor coolant pumps, incase of a grid frequency decrease so DNB does not occur.

TURKEY POINT - UNITS 3 SL 4 B 2-7 JuN <> |S>


BASES

Reactor Tri S stem Interlocks

The Reactor Trip System interlocks perform the following functions:'-6

On increasing power, P-6 allows the manual block of the Source Rangetrip (i.e., prevents premature block of Source Range trip) anddeenergizes the high voltage to the detectors. On decreasing power,Source Range Level trips are automatically reactivated and highvoltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on lowflow in more than-one reactor coolant loop, more than one reactorcoolant pump breaker open, reactor coolant pump bus undervoltage andunderfrequency, Turbine trip, pressurizer low pressure and pressurizerhigh level. On decreasing power, the above listed trips are auto-matically blocked.

P-8 On increasing power, P-8 automaticafly enables Reactor trips on low-:.flow in one or more reactor coolant loops, and one or more reactor

„'oolantpump breakers open. On decreasing power, the P-8 interlockautomatically blocks the trip on low flow in one coolant loop orone coolant pump breaker open.

P-10 On increasing power, P-10 allows the manual block of the IntermediateRange trip and the Low Setpoint Power Range trip; and automaticallyblocks the Source Range trip and deenergizes the Source Range highvoltage power. On decreasing power, the Intermediate Range trip andthe Low Setpoint Power Range trip are automatically reactivated.P-10 also provides input to P-7..

~ a

TURKEY POINT - UNITS 3 81 4 B 2-8

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS

3/4. 0 APPLICABILITY

LIMITING CONDITIONS FOR OPERATION

3.0. 1 Compliance with the Limiting Conditions for Operation contained inthe succeeding specifications is required during the OPERATIONAL NOES orother conditions specified therein; except that upon failure to meet theLimiting 'Conditions for Operation, the associated ACTION requirements shallbe met.

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

3.0.3 When a Limiting Condition for Operation is not met, except as providedin the associated ACTION requirements,'within 1 hour action shall'be initiated.to place it, as applicab1e, in:

a. At least HOT STANDBY within the next 6hours'.

At least HOT SHUTDOWN within the following 6 hours, and

c. At least COLD SHUTDOWN within the subsequent 24 hours.

Where corrective measures are completed that permit operation under the ACTIONrequirements, the action may be taken in accordance with the specified timelimits as measured from the time of failure to meet the Limiting Condition forOperation. Exceptions to these requirements are stated in the individualspecifications.:

This specification is not applicable in MODES 5 or 6.

3.0.4 Entry into an OPERATIONAL NODE or other specified condition shall notbe made when the conditions for the Limiting Conditions for Operation are notmet and the associated ACTION requires a shutdown if they are not met withina specified ti~ interval. Entry into an OPERATIONAL NDE or specifiedcondition may Q made in accordance with ACTION requirements when conformanceto them permiti: continued operation of the facility for an unlimited periodof time. This provision shall not prevent passage through or to OPERATIONALMODES as required to comply with ACTION requirements. Exceptions to theserequirements are stated in the individual specifications.

+'Excel'! ~/en 6e r,cTloi) apg-,. "7g gr!! gr 75~ 5<~v!~~nc'>~15((""l

d,e~ koln'„6 ~!di( 4,,„g< lpga~] //07 ~7~ND8~:I( /Ill'8 /2 I"Qci f (.

. TURKEY POINT - UNITS 3 5 4 3/4 0-1 JUNik '-'

APPLICABILITY

LIMITING CONDITIONS FOR OPERATION Continued

3.0.5 Limiting Conditions for Operation including the associated ACTIONrequirements shall apply to each unit individually unless otherwise indicatedas follows:

a. Whenever the Limiting Conditions for Operation refers to systems orcomponents which are shared by both units, the ACTION requirementswill apply to both units simultaneously.

b. Whenever the Limiting Conditions for Operation applies to only oneunit, this will be identified in the APPLICABILITY section of thespecification; an/

c. Whenever certain portions of a specification contain operatingparameters, Setpoints, etc., which are different for each unit, thiswill be identified in parentheses, footnotes or body of therequirement.

5.o,g .

When a system, subsystem, train, component or device is determined tobe inoperable soley because its emergency power source is inoperable, or solelybecause its normal power source is inoperable, it may be considered OPERABLEfor the purpose of satisfying the requirements of its applicable LimitingCondition for Operation, provided: (1) its corresponding normal or emergencypower source is OPERABLE; and (2) all of its redundant system(s), subsystem(s),train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the+ requirements of this specification. This specification is not applicable in NODES

~

~ ~

~ ~ ~

TURKEY POINT - UNITS 3 4 4 3/4 0-2 - 'i 0 0 1Sct

APPLICABILITY

SURVEILLANCE RE UIREMENTS

4.0. 1 Surveillance Requirements shall be met during the OPERATIONAL MODESor other conditions specified for individual Limiting Conditions forOperation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specifiedtime interval with:

a. A maximum allowable extension not to exceed 25X of the surveillanceinterval, but

b. The combined time interval for any three consecutive surveillanceintervals shall not exceed 3.25 times the specified surveillanceinterval.

4.0.3 Failure to perform a Surveillance Requirement within the allowedsurveillance interval, defined by Specification 4.0.2, shall constitutenoncompliance with the OPERABILITY requirements for a Limiting Conditionfor Operation. The time limits of the ACTION requirements are applicable atthe time it is identified that a Surveillance Requirement has not beenperformed. The ACTION requirements may be delayed for up to 24 hours topermit the completion of the surveillance when the allowable outage timelimits of the ACTION requirements are less than 24 hours. SurveillanceRequirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAI MODE or other specified condition shall notbe made unless the Surveillance Requirement(s) associated with a LimitingCondition for Operation has been performed within the stated surveillanceinterval or as'otherwise specified. This provision shall not prevent, passagethrough or to OPERATIONAL MODES as required to comply with ACTION requirements.

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

Inservice inspection of ASME Code Class 1, 2, and 3 components andinsoiHce testing of ASME Code Class 1, 2, and 3 pumps and valvesshaQ be performed in accordance with Section XI of the ASME Boilerand P&ssur» Vessel Code and applicable Addenda as required by10 CFR 50, Section 50.55a(g), except where specific written reliefhas been granted by the Commission pursuant to 10 CFR 50, Section50.55a(g)(6)(i,).

'URKEY POINT - UNITS 3 Sc 4 3/4 0-3 r4 CS ';nP

APPLICABILITY

SURVEILLANCE RE UIREHENTS CONTINUEO

b. Surveillance intervals specified in Section XI of the ASIDE Boilerand Pressure Vessel Code and applicable Addenda shall be applicableas follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies forCode and applicable Addenda performing inserviceterminology for inservice inspection and testingins ection and testin activities activities

sleekly At east once per 7 daysMonthly At least once per 31 d~

quarterly or every 3 months At least once per 92 da.Semiannually or every 6 months At least once per 184 days

Every 9 months At least once per 276 daysYearly or annually At least once per 366 days

c. The provisions of Specification 4;0.2 are applicable to the aboverequired frequencies for performing .inservice inspection and testing..activities.

d. Performance of the above inservice inspection and testing activitiesshall be in addition to other specified Surveillance Requirements.

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

4.0.6 Surveillance Requirements shall apply to each unit individually unlessotherwise indicated as stated in Specification 3.0.5 for individualspecifications or whenever certain portions of a specification containsurveillance parameters different for each unit, which will be identified inparentheses, footnotes or body of the requirement.

'TURKEY POINT - UNITS 3 4 4'4

3/4 0-4 -t4 C S '.go;


3/4. 1. 1 BORATION CONTROLRt" 5 /~err'<~<>'p "=~"

SHUTDOWN MARGIN ~M GREATER THAN 200 F

LIMITING CONDITION FOR OPERATION

3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to the applicablevalue shown in Figure 3.1-1.

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

ACTION:

With the SHUTDOWN MARGIN less than the applicable value shown in Figure 3.1-1,immediately initiate and continue boration at greater than or equal to 10gpm of a solution containing greater than or equal to 20,000 ppm boron orequivalent until the required SHUTDOWN MARGIN is restored.


4.1.1.1.1 The SHUTDOWN MARGIN shall be determfned to be greater than or equah-to the applicable value shown in Figure 3;1-1:

a. Within 1 hour after detection of an inoperable control rod(s) andat least once per 12 hours thereafter while the rod(s) is inoperable.If the inoperable control rod is immovable or untrippable, the aboverequired SHUTDOWN MARGIN shall be verified acceptable with an increasedallowance for the withdrawn worth of the immovable or untrippablecontrol rod(s);

b. When in MODE 1 or MODE 2 with K ff greater than or equal to 1 atleast once per 12 hour's by verifying that control bank withdrawal iswithin the limits of Specification 3.1.3.6;

c. When in MODE 2 with K ff less than 1, within 4 hours prior to achievingreactor criticality bg verifying that the predicted critical controlrod position is within the limits of Specification 3.1.3.6;

d. Prfto initial operation above 5X RATED THERMAL POWER after eachfu~loading, by consideration of the factors of Specification4.3@5 I.le. below, with the control banks at the maximum insertionlimit of Specificatfon 3.1.3.6; and

"See Special Test Exceptions Specification 3.10.1.

TURKEY POINT - UNITS 3 4 4 3/4 1-1 JUti 09 1='-'

REACTIVITY CONTROL SYSTEMS

SURVEILLANCE RE UIREMENTS Continued

e. When in MODE 3 or 4, at least once per 24 hours by consideration ofthe following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

0

4.1.1.1.2 When in MODE 1 or 2, the overall core reactivity balance shall becompared to predicted values to demonstrate agreement within 4 3X hk/k at leastonce per 31 Effective Full Power Days (EFPD). This comparison shall consider..-at least those factors stated in Specification 4.1.1. l.le, above. Thepredicted reactivity values ~~ be ad)usted. (normalized) to correspond to 43the actual core conditions prio to exceeding a fuel burnup of 60 EFPD aftereach fuel loading.

TURKEY POINT - UNITS 3 4 4 3/4 1-2JUN 09 1888

),77% h k/ k

CI

X

750 ~~<

lS .

fThoeeea4e3ECS. RÃ%N «MSNIRaTICR (HN)

Figure 3.1-1Requ)red Shutdown Hargfn vs Reactor Coolant

Boron Concentration

TURKEY POINT - UNITS 3 4 4 3/4 1-3


pc> uututptA 7tumpeiakpoSHUTDOWN MARGIN -p~ LESS fHAN'R EQUAL TO 200 F


3. 1.1.2 . The SHUTDOWN MARGIN shall be greater than or equal to IX hk/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than 1X hk/k, immediately initiate and continueboration at greater than or equal to 1Q gpm of a solution containing greater

20, t p t'~ u t t ttMARGIN is restored. ~,~ ~~ go~ c p,c,V„- None, OpeeeeLC

t:IttttOiuIpump tttiN r«uaan4o u utttoF e~SLG @~M


r4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equalto 1X hk/k:

t

a. Within 1 hour after detection of an inoperable control rod(s) and atleast once per 12 hours thereafter while the rod(s) is inoperable.If the inoperable control rod is iaeovable or untrippable, theSHUTDOWN MARGIN shall be verified acceptable with an increasedallowance for the withdrawn worth of the iaaevable or untrippablecontrol rod(s); and

b. At least once per 24 hours by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant Systea average temperature,

4) Fuei burnup based on gross thermal energy generation,

5) .Icon concentration, and..I

6) Saaarim concentration.

at

~ ~ u tj

TURKEY POINT" UNITS 3 4 4 3/4 1-4 JUN I 9 198|I


MODERATOR TEMPERATURE COEFFICIENT


0

3.1.1.3 The moderator temperature coefficient (MTC) shall be:

Less positive than or equal to 5.0 x 10-s hk/k/ F for all rodswithdrawn, beginning of the cycle life (BOL), hot zero'THERMAL POWER(HZP) conditions; and

b. Less positive than or equal to .5.0 x 10- hk/k/ F from HZP to 70KRATE THERMAL POWER condition; and

c. Less positive thW or equal to 5.0 x 10- hk/k/ F from 70X RATEDTHERMAL POWER decreasing linearly to less positive than or equal.to0 hk/k/ F at 100K RATED THERMAL POWER conditions; and

d. Less negative than -3.5 x 10 ~ hk/k/oF for the all rods withdrawn,end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3a, b and c. - MODES 1 and 2~ only*".Specification 3.1.1.3d. - MODES 1, 2, and 3 only**.

ACTION:

With the MTC more positive than the limit of Specification 3.1.1.3a,b or c above, operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintainedsufficient to restore the MTC to less positive or equal tolimits described in 3.1.1.3a, b and c above within 24 hours orbe in HOT STANDBY within the next 6 hours. These withdrawallimits shall be in addition to the insertion limits ofSpecification 3.1.3.6;

2. The control rods are maintained within the withdrawal limitsestablished above until a subsequent calculation verifies thatthe NTC has been restored to within fts limit for the all rods%6Chdrawn condition; and 30

3. 4 Special Report fs prepared and submitted to the Coaefssfon,pursuant to Specification 6.9.2, within %0 days, describing thevalue of the measured MTC, the fnterfm control rod withdrawallimits, and the predicted average core burnup necessary forrestoring the positive MTC to within fts lfaft,for the all rodswithdrawn condition.


'With K ff greater than or equal to 1.effe *"See Special Test Exceptions Specification 3.10.3.,

QQN 0 9 1968


LIMITING CONDITION .FOR OPERATION

ACTION: (Continue'd)

b. With the MTC more negative than the limit of Specification 3.1.1.3d.above, be in HOT SHUTDOWN within 12 hours.


4.1.1.3cycle as

aO

b.

, C.

The MTC shall be determined to be within its limits during each fuelfol 1 ows:

The MTC shall be measured and compared to the BOL limit of Specifi-cation 3.1.1.3a., above, prior to initial operation above SX ofRATED THERMAL POWER, after each fuel loading; and

The MTC shall be measured at any THERMAL POWER and compared to-3.0 x 10-4 hk/k/DF (all rods withdrawn, RATED THERMAL POWERcondition) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates theMTC is more negative than -3.0 x 10-i hk/k/oF, the MTC shall beremeasured, and compared to the EOL MTC limit of Specification3.1.1.3d., at least once per 14 EFPD during the remainder of thefuel cycle.

Perform design calculation to verify conformance to Specifications3.1.1.3b and c.

'URKEY POINT - UNITS 3 8L 4 3/4 1-6 dUN Os ~s'.8


MINIMUM TEMPERATURE FOR CRITICALITY


3. 1. 1.4 The Reactor Coolant System lowest operating loop temperature (T )shall be greater than or equal to 541 F. avg

APPLICABILITY: MODES 1 and 2" "".

ACTION:

With a Reactor Coolant System operating loop temperature (T ) less thanavg541'F, restore T „ to 'within its limit within 15 minutes or be in HOT

STANDBY within the next 15 minutes.


04. 1.1.4 The Reactor Coolant System temperature (Tav ) shall be determined tobe greater than or equal to 541 F:

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

b. At least once per 30 minutes when the reactor is critical and theReactor Coolant System T is less than 5474F with the T -T

avg refDeviation Alarm not reset.

"With K <f g~yter than or equal to 1.

""See Special Test Exceptions Specification 3.10.3.:

TURKEY POINT - UNITS 3 81 4 3/4 1-7 JUN 09 l988


3/4. 1. 2 BORATION SYSTEMS

FLOW PATH - SHUTDOWN


3.1.2.1 As a minimum, one of the following boron injection flow paths shallbe OPERABLE and capab1e of being powered from an OPERABLE emergency powersource:

a. A flow path from the boric acid storage tanks via a boric acidtransfer pump and a charging pump to the Reactor Coolant System ifthe boric acid storage tank in Specification 3.1.2.5a. isOPERABLE, or

b. The f1ow path from the refueling water storage tank via a chargingpump to the Reactor Coolant System if the refueling water storagetank in Specification 3.1.2.5b. fs OPERABLE;


ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an ~OPERABLE emergency power source, suspend all operations involving COREALTERATIONS or positive reactivity changes.


4.1.2.1 At least one of the above 'required flow paths shall be demonstratedOPERABLE:

",7da 5 >

a. At least once per'. by verifying that the temperature of the /heat. traced portion of the flow path is greater than or equal to14 ' a flow path froa the boric acid inks is up/ and

~ g~c ~Echeb. A - 'nce. per 31 days by verifying that eactr>vive- nual,

p pirated, or autoeatic) in the flow. path tQt is not locked,sealed; or otherwise secured in position; is in its correctposition.

TURKEY POINT - UNITS 3 4 4 3/4 1-8 JUN 09 1II


FLOW PATHS - OPERATING


3. 1.2.2 The~f ing boron injection f1ow paths sha11 be OPERABLE:, ~urce-

a. The+~path from a boric acid storage tank via a boric acidtransfer pump to the charg n ump suction*, and

~ source.b. At least one of the two~+~a&s from the refueling water storage

tank to the charging pump suction; and,

c. , The flow path from the charging pump discharge to the ReactorCoolant System vie the regenerative heat exchanger.

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

ACTION:

a. With no boration source path from a boric acid storage tank OPERABLE,

1. Demonstrate the OPERABILITY of the second flow path from therefueling water storage tank to the charging pump suction byverifying the flow path valve alignment; and

2. Restore the boration source path from a boric acid storage tankto OPERABLE status within 72 hours or be in at least HOT STANDBYand borated to a SHUTDOWN MARGIN equivalent to at least lX h k/kat 200'F within the next 6 hours; r'estore the boration sourcepath from a boric acid storage tank to OPERABLE status withinthe next 72 hours or be in COLD SHUTDOWN within the next 30hours.

b. With only one boration source path OPERABLE or the regenerative heatexchanger flow path to the RCS inoperable, restore the required flowpaths to OPERABLE status within 72 hours or be in at least HOTSTANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 3X6 k/k at 2004F within the next 6 hours; restore at least twoboration source paths to OPERABLE status within the next 72 hours orbe irt COlD SHUTDOWN within the next 30 hours.

c. With the boratfon source path froa a boric acid storage tank and thecharging pump discharge path via the, regenerative heat exchangerinoperable, within one hour initiate boration to a SHUTDOWN MARGINequivalent to 1% h k/k at 2004F and go to COLD SHUTDOWN as soon aspossible within the limitations of the boration and pressurizerlevel control functions of the CVCS.

"The flow required in Specification 3.1.2.2.a above, shall be isolated fromthe other unit.

TURKEY POINT - UNITS 3 5 4 3/4 1-9JUN 05 t9oo


SURVEILLANCE RE UIREHENTS

4.1.2.2

a ~

b.

C.

The above required f s shall be demonstrated OPERABLE:

7Jag>At least once p ~~ b verifying that the temperature of theheat traced porti ow path from the boric acid tanks isgreater than or equal to 1454F when it is a re~uired wa~ source;

< acce.es i'd@+~At least once per 31 days by verifying that eacRPalve manual,power-operated, or automatic) in the flow path that is not locked,sealed, or otherwise secured in position, is in its correct position;

At least once per 18 months by verifying that the flow path requiredby Specification '3. 1.2.2a. and c. delivers at least 10 gpm to the RCS.



CHARGING PUMPS - OPERATING


3.1.2.3 At least two charging pumps with independent power supplies shall be

~y age'" 'ni.mEPgs )i~i»~ +a. With two charging pumps OPERABI.E and powered from a common power

supply, restore at least two charging pumps from independentpower supplies to OPERABLE status within 7 days or be in at leastHOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to atleast 1% 6 k/k at 200'f within the next 6 hours; restore at leasttwo charging pumps from independent power supplie to OPERABLEstatus within 22 hours or be in TDOWN wit in the next 30hours. ~ L'ttln 'sv4e~~e ~~<

b. With only one charging pump OPERABLE, restore any two chargingpumps to OPERABLE status 72 hours or be in at least HOTSTANDBY and borated to a SHUTDOWN MARGIN 'equivalent to at least1X h k/k at 200'F within 6 hours.„ restore any two charging pumpsto OPERABLE status 72 hours o be in COLD SH~GWN withinthe next 30 hours. Ou in6euuewd

c. The provisions of Speci cation 3.0.4 are not ap licable toACTION a, provided the 7 day limit of ACTION a s notexceede .


4. 1.2.3.1 The required charging pumps shall be demonstrated OPERABL bytesting pursuant to Specification 4.0e5 The provisions of Specific ion 4.04are not applicable for entry into MODES .3 and 4. h,

TURKEY POINT - UNITS 3 Ec 4 3/4 1-11~ s


BORATEO WATER SOURCE - SHUTDOWN

LIMITING CONOITION FOR OPERATION

3. 1.2.4 As a minimum, one of the following borated water sources shall beOPERABLE:

a. A Boric Acid Storage System with:

1) A minimum indicated borated water volume of 500 gallons,

2) A boron concentration between 20,000 ppm and 22,500 ppm, and

3) A minimum solution temperature of 145 F.

b. The refueling water storage tank ( ST wAtt.

1) A minimum indicated borate'd. er volum f 20,000 gallons,

2) A minimum boron concentrat on ~of 1 0 ppm and

3) A minimum solution tempera re '-of 39o .


ACTION:

With no borated water source OPERABLE, suspend all operations involving COREALTERATIONS or positive reactivity changes.

I&


4.1.2.4 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per ? days by:

1) 4irifyinO the boron'concentration of the mater,Io)iwfcA

2 Ver inO the< rated water vol~, anda)

'tterifying the boric acid storagetank solution temperature whe it is the source of borated water.

TURKEY POINT - UNITS 3 8s 4 3/4 1-12JUN 09 1988

REACTIVITY CONTROL SYSTEHS-


By verifying the RWST temperature is above its iimit whenever theoutside air temperature is less than 394F at the followingfrequencies:

1) Within one hour when the outside temperature is below. 394F for23 consecutive hours, and

2) At least once per 24 hours when the outside temperature is below39 F.

TURKEY POINT - UNITS 3 5 4 , 3/4 1-13

JUN 09 )98&


BORATED WATER SOURCES - OPERATING


3.1.2.5 The following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

1) A minimum indicated borated water volume of 3080 gallons,

2) A boron concentration between 20,000 ppm and 22,500 ppm, and

3) A minimum solution temperature of 145 F.

b. The refueling water storage tank (RWST) with:

1) A minimum indicated borated water volume of 320,000 gallons,

2) A minimum boron concentration of 1950 ppm,

3) A minimum solution temperature of 394F, and

4) A maximum solution temperature of 1004F.


ACTION:

a. With the required Boric Acid Storage System inoperable verify thatthe RWST is OPERABLE; restore the system to OPERABLE status within72 hours or be in at least HOT STANDBY within the next 6 hours andborated to a SHUTDOWN MARGIN equivalent to at least lX dk/k at 200'F;restore the Boric Acid Storage System to OPERABLE status within thenext 72 hours or be in COLD SHUTDOWN within the next 30 hours.

b. With the RWST inoperable, restore the tank to OPERABLE statuswithin 1 hour or be in at least HOT STANDBY within the next6 hours and in COLD SHUTDOWN within the following 30 hours.


REACTIVITY CONTROL SYSTEHS


4. 1.2.5 Each borated water source shal.l be demonstrated OPERABLE:

a. At least once per 7 days by:

b.

1) Verifying the bor ion in the water,LYl CI 1 CN

2) Verifying the ~Haed~bor ted water volume of the watersource, and

At least once per by verifying the Boric Acid Storage Systemsolution temperature when it is the source of borated water.

c. By verifying the RWST temperature is within limits whenever theoutside air temperature is less than 39~F or greater than 100'F atthe following frequencies:

1) Within one hour upon the outside .temperature excee'ding its limitfor 23 consecutive hours, and

2) At least once per 24 hours while the outside temperature exceedsits limits.


JUN OQ ]pap


HEAT TRACING


3.1.2.6 At least two independent channels of heat tracing shall be OPERABLEfor the boric acid storage tank and for the heat traced portions of theassociated flow paths required by Specification 3.1.2.2.

APPLICABILITY: MODES 1, 2, 3 and 4MODES 5 and 6 (when the boric acid storage tank is the boratedwater source per Specification 3.1.2.4)

ACTION:

MODES 1 2 3 and 4

With only one channel of heat tracing on either the boric acid storage tank oron the heat traced portion of an associated flow path OPERABLE, operation maycontinue for up to 30 days provided the tank and flow path temperatures areverified to be greater than or equal to 145 F at least once per 8 hours;otherwise, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWNwithin the following 30 hours.

MODES 5 and 6

With only one channel of heat tracing on either the boric acid storage tank oron the heat traced portion of an associated flow path OPERABLE, operationsinvolving CORE ALTERATIONS or positive reactivity additions may continue forup to 30 days provided the tank and flow path temperatures are verified to begreater than or equal to 1454F at least once per 8 hours; otherwise, suspendall activities involving CORE ALTERATIONS or positive reactivity changes.


4.1.2.6 Each heat tracing channel for the boric acid storage tank andassociated flow path required by Specification 3.1.2.2 shall be demonstratedOPERABLE:

a. At least once per 31 days by energizing each heat tracing channel,and

b. At least once per y verifying the tank and flow pathtemperatures to be greater than or equal to 1454F. The tanktemperature shall be determined by measurement. The flow pathtemperature shall be determined by either measurement orrecirculation flow until establishment of equilibrim temperatureswithin the tank.

4TURKEY POINT - UNITS 3 4 4

k

3/4 1-16JUN 09 @8<


3/4.1.3 MOVABLE CONTROL ASSEMBLIES

GROUP HEIGHT


/Pod(f'to33733clicttttotr,

'

3. 1.3. 1 All full length (shutdown and c ntrol) rods shall be OPERABLE andpositioned within + 12 steps (' of the

group step counter demand position within one hour afterrod motion.

APPLICABILITY: MODES 1" and 2*

ACTION:

With one or more full length rods inoperable due to being immovableas a result of excessive friction or mechanical interference or knownto be untrippable, determine that the SHUTDOWN MARGIN requirement ofSpecification 3. 1. 1.1 is satisfied within 1 hour and be in HOTSTANDBY within 6 hours.

b.

C.

With more than one full length rod inoperable or misaligned from thegroup step counter demand position by more than t 1lp steps,gieA'mated-;.~po~~),'e in HOT STANDBY within 6 hours. /~An loj Pg 'Pootto~, r

133 3 clgf'pe: 33.With one full length rod inoperable dua to causes otherthan-addressed by ACTION a, above, or misaligned from its group stepcounter demand position by more than k 12 steps ( „),POWER OPERATION may continue provided that within on hogr ejthez:

1. The rod is restored to OPERABLE status within th bqv% aligkiWYi<.,„.requirements, or

2. The remainder of the rods in.the bank with the inoperable rod arealigned to within f 12 steps of the inoperable rod whilemaintaining the rod sequence and insertion limits ofFigure 3.1-2; the THERMAL POWER level shall be restrictedpursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGINrequirewtent of Specification 3.1.1.1 is satisfied. POWER OPERATIONmay then continue provided that:

~dp * 3 * 3 1 113.1 3 3.13.3.

TURKEY POINT - UNITS 3 8L 4 3/4 1-17

JUN 0 9 lanai


LIMITING CONDITION FOR OPERATION Continued

g m) A reevaluatfon of each accident analysfs of Tab'le 3.1-1 ls performedwithin 5 days; this reevaluation shall confirm that the previouslyanalyzed results of these accidents remain valid for the duration ofoperation under these conditions,

b) The SHUTDOWN MARGIN requirement of S citric tion 3.1.1.1 isdetermined at least once per 12 hou s ar4

c) A power distribution map is obtained from the movable incoredetectors and,F~(Z) and F are verified to be within their limits within( p i g72 hours>stroll /~

q*P) The THERMAL POWER level is reduced to less than or equal to 75% ofRATED THERMAL POWER within one hour and within the next 4 hours thepower range neutron flux high trip.setpoint is reduced to less thanor equal to 85'f RATED THERMAL POWER. THERMAL POWER shall bemaintained less than or equal to ~f RATED THENQ~tIWER untilcompliance w TONS 3.1.3.1.ci3.m' d 3.1.3.1.4.3.d'I aredemonstrat a~&. ( c. l g beo~u

SURVIELLANCE RE UIREMENTS

p „afo~ g +eLt om X ice'f crt ')

4.1.3.1.1 The p n of eac full length rod shall be determined to bewithin + 12 step ( ) of the ml4e &e-group step demand position at least once per 12 hours (allowing for onehour thermal soak fter rod motion) except during time invervals when the RodPosition Deviation Monitor is inoperable, then verify the group positions atleast once per 4 hours.

4.1.3.1.2 Each full length rod not fully inserted in the core shall bedetermined to be OPERABLE by movement of at least 10 steps.in any onedirection at least once ptr 31 days.

+%2.> ~

0TURKEY POINT - UNITS 3 Cs 4 3/4 1-18

JUN 01 ice~

TABLE 3.1-1

ACCIDENT ANALYSES RE UIRING REEVALUATION

IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD

Rod Cluster Control Assembly Insertion Characteristics

Rod Cluster Control Assembly Misalignment

Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks inLarge Pipes Which Actuates the Emergency Core Cooling System

Single Rod Cluster Control Assembly Withdrawal at Full Power

Major Reactor Coolant System Pipe Ruptures (Loss-of-CoolantAccident)

Major Secondary Coolant System Pipe Rupture

Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster ControlAssembly Ejection)


JUN Q Q 1888


POSITION INDICATION SYSTEMS - OPERATING

LIMITING'ONDITIONFOR OPERATION

mg o<aJcg rob Po>Lt(a~ Lwhtc44Lo~ +g4c'~ ~ f ~ dl~A,~ pcsLfLc~L~ I ~ 5VCle U3.1.3.2 n shall be

OPERABLE and capable of determining the actual and demanded rod positions<asfollows:

a. Analog rod position indicators, within one hour after rod motion(allowance for thermal soak);

All Shutdown Banks: + 12 steps of the group demand counters forwithdrawal ranges of 0-30 steps and 200-228 steps.

Control Bank A and B: k 12 steps of the group demand counters forwithdrawal ranges of 0-30 steps and 200-228 steps.

Control Banks C and 0: + 12 steps of the group demand counters for.~228 p .

b. . Group demand counters; k 2 steps.


ACTION:

With a maximum of one analog rod position indicator per bank inoperableeither:

1. Determine the position of the non-indicating rod(s) indirectlyby the movable incore detectors at least once per 8 hours andwithin one hour after any motion of the non-indicating rod whichexceeds 24 steps in one diriction since the last determinationof the rod's position, or

UFO2. . Reduce THERMAL POWER to less than~ of RATED THERMAL POWER

. within 8 hours.

b. With a maximum of one demand position indicator per bank inoperableeither:

l. Verify that all analog rod position indicators for the affectedbank are OPERABLE and that the most withdrawn rod and the leastwithdrawn rod of thi bank are within a maximum of 12 steps ofeach other at least once per 8 hours, or

vsr2. Reduce THERMAL POWER to less than 59K'f RATED THERMAL POWER

within 8 hours.

JUN I9 1988




4, 1.3.2. 1 Each'analog rod position indicator shall be determined to be OPERABLEby verifying that the demand position indication system and the rod positionindication system agree within 12 steps(allowing for one hour thermal soak after rod motion) at least once per 12 hoursexcept during time intervals when the Rod Position Deviation Monitor is inoper-able, then compare the demand position indication system and the rod positionindication system at least once per 4 hours.

4. 1. 3. 2, 2 Each of the above required rod position indicator(s) shall bedetermined to be OPERABLE by performance of a CHANNEL CHECK, CHANNEL CALIBRA-TION and ANALOG CHANNEL OPERATIONAL TEST performed in accordance withTable 4.1-1.

vtmsp

JUN 09 gals

TURKEY POINT - UNITS 3 8c 41

3/4 1-21

TABLE 4.1-1

ROD POSITION INDICATOR SURVEILLANCE RE UIREHENTS

Functional Unit

Individual Rod Position

Demand Position

Check C lib l ~CT T

N/A


VuN O9 >sea

PR


POSITION INDICATION SYSTEM - SHUTDOWN


3. 1.3.3 The group step demand position indicator shall be OPERABLE and capableof determining within i 2 steps the demand position for each shutdown and controlrod not fully inserted.

APPLICABILITY: MODES 3"8, 4*8, and '5*8

ACTION:

With less than the above required group step demand position indicator(s)OPERABLE, open the reactor <rip system breakers.


4, 1.3.3. 1 Each of the above required group step demand position indicator(s)~ ~ ~ ~

~

~shall be determined to be OPERABLE by movement of the associated control rodat least 10 steps in any one direction at least once per 31 days.

4. 1.3,3.2 A CHANNEL CHECK CALIBRATION AND ANALOG CHANNEL OPERATIONAL TEST shallbe performed per Table 4. 1-1.

"With the Reactor Trip System breakers in the closed position.SSee Special Test Exceptions Specification 3.10. .+

TURKEY POINT - UNITS 3 4 4 3/4 1-23 JUN 09 lg><


ROD DROP TIME

LIMITING CONDITION FOR OPERATIONe

3. 1.3.4 The individual full"length (shutdown and control) rod drop time fromthe fully withdrawn position shall be less than or equal to 2.4 seconds frombeginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T greater than or equal to 541DF, andavg

b. All reactor coolant pumps operating.


ACTION:

With the drop time of any full-length rod determined to exceed the abovelimit, restore the rod drop time to within the above limit prior to proceedingto MODE 1 or 2.


4. 1.3.4 The rod drop time of full-length rods shall be demonstrated throughmeasurement prior to reactor criticality:

b.

C.

For all rods following each removal of the reactor vessel head,

For specifically affected individual rods following any maintenanceon or modification to the Control Rod Drive System which couldaffect the drop time of those specific rods, and

At least once pe ~oaths.refocliop

L

rook to e~ Z4 OLOnl4 ~

TURKEY POINT - UNITS 3 SL 4 3/4 1-24


SHUTDOWN ROD INSERTION LIMIT


3. 1.3.5 All shutdown rods shall be fully withdrawn.

APPLICABILITY: MODES 1" and 2"8

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillancetesting pursuant to Specification 4.1.3.1.2, within 1 hour either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification3. 1.3. l.


4. 1 3 5 Ea h shutdo rod sh ll be determi'ne to b 11 withdrawn;~~

a. Within 15 minutes prior to withdrawal of any rods in controlbanks A, 8, C, or D during an approach to reactor criticality, and

b. At least once per 12 hours thereafter.

*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.IIWith K ff greater than or equal to 1 . 0

TURKEY POINT - UNITS 3 8 4 3l'4 1-254UN 0 9 1888


CONTROL ROD INSERTION LIMITS

LIMITING CONDITION FOR OPERATION0

3. l. 3. 6 The control banks shall be limited in physical insertion as shown inFigures 3. 1-2.

APPLICABILITY: MODES 1* and 2"8

ACTION:

With the control banks inserted beyond the above insertion limits, except forsurveillance testing pursuant to Specification 4.1.3.1.2 either:

l d

a. Restore the control banks to within the limits within 2 hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to thatfraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the above figures, or

c. Be in at least HOT STANDBY within 6 hours.


4.1.3.6 The position of each control bank shall be determined to be withinthe insertion limits at least once per 12 hours

d ig l i 1 dg ddt l ill gis inoperable, then verify the individual rod positions at least once per4 hours.

"See Special Test Exceptions Specifications 3.10.2 and 3.10.3.4Wth K ff greater than or equal to 1.0eff

0J:,'t 09 1988

TURKEY POINT - UNITS 3 8c 4 3/4 1-26

215

192

(lS, Sn SnPS)

ace c

120

(OS, 1O7 S>ESS)

72

24

0'ltS. 0 SmrS)

40 '0~OWN LKNLo tQCNT NT

FIGURE 3.1-2

ROD BANK INSERTION LIMITS VERSUS THERMAL POWER

THREE LOOP OPERATION

TURKEY POINT - UNITS 3 5 4 3/4 NVJVN ( S 5--

/ ~tS VGA +- i w,~ v V <Qwo.~ai w

'I .t

3/4. 2 POWER DISTRIBUTION LIMITS

3/4.2.1 AXIAL FLUX DIFFERENCE


3.2. 1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained withina t 5X target band (flux difference units) about the target flux difference.

The indicated AFD may deviate outside the above required target band at greaterthan or equal to 50K but less than 90K of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour during the previous 24 hours.

The indicated AFD may deviate outside the above required target band at greaterthan 15K but less than 50K of RATED THERMAL POWER provided the cumulativepenalty deviation time does not exceed 1 hour during the previous 24 hours.

APPLICABILITY: MODE 1, above lSX of RATED THERMAL POWER."

ACTION:

'a ~

b.

With the indicated AFD outside of the above required target band andwith THERMAL POWER greater than or equal to 90'f RATED THERMALPOWER, within 15 minutes either:

1. Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL'OWER to less than 90K of RATED THERMAL POWER.

With the indicated AFD outside of the above required target band formore than 1 hour of cumulative penalty deviation time during theprevious 24 hours or outside the Acceptable Operation Limits ofFigure 3.2-1 and with THERMAL POWER less than 90X but equal to orgreater than 50K of RATED THERMAL POWER:

1. Reduce THERMAL POWER to less than 50K of RATED THERMAL POWER

within 30,minutes, and

2. Reduce the Power Range Neutron Flux~ "" - High Trip Setpoints toless than or equal to 55K of RATED THERMAL POWER within the next4 hours.

"See Special Test Exceptions Specification 3.10.2.Surveillance testing of the Power Range Neutron Flux Channels may be performedpursuant to Specification 4.3.1.1 provided the indicated AFD is maintained withinthe Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours operationmay be accumulated with the AFD outside of the above required target band duringtesting without penalty deviation.

, TURKEY POINT - UNITS'3 4 4 3/4 2-1J:.', ~:988

AJd Coue gg W <Z.l.(.a. Z)

*Qv )Qr ~a~(.Q,PqI'~i«TVos@ qI ri.,

c ~e4 to ~0 Ac We <VA m g I e.Sa &SviTu+e2 Qn +Lie Wgc i~mgv t.

POWER DISTR!BUTION LIMITS


ACTION Continued

c. With the indicated AFD outside of the above required target band formore than 1 hour of cumulative penalty deviation time during theprevious 24 hours and with THERMAL POWER less than 50X but greaterthan 15'f RATED THERMAL POWER, the THERMAL POWER shall not be

'ncreasedequal to or greater than 50K of RATED THERMAL POWER untilthe indicated AFD is within the above required target band.


4.2. l. 1 The indicated AFO shall be determined to be within its limits duringPOWER OPERATION above 15X of RATED THERMAL POWER by:

a. Monitoring the indicated AFO for each OPERABLE excore channel:

1) At least once per 7 days when the alarm used to monitor the AFQ:is OPERABLE, and

r

2) At least once per hour for the first ~ hours~after ~toringthe alarm used to monitor the AFD to OPERABLE statu

b. Monitoring and logging the indicated AFD for each OPERABLE excorechannel at least once per hour for the first 24 hours and at leastonce per 30 minutes thereafter, when the alarm used to monitor theAFO is inoperable. The logged values of the indicated AFO shall beassumed to exist during the interval preceding each logging.

4.2. 1.2 The indicated AFD shall be considered outside of its target band whentwo or more OPERABLE excore channels are indicating the AFD to be outside thetarget band. Penalty deviation outside of the above required target band shallbe accumulated on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER OPERATIONoutside of the target band at THERMAL POWER levels equal to or above5~of RATED THERMAL POWER, and

b. One~if minute penalty deviation for each 1 minute of POWER OPERATIONoutside of the target band at THERMAL POWER levels between -15K and50K of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be..determined by measurement at least once per 92 Effective Full Power Days.The provisions of Specification 4.0.4 are not applicable.

4.2.,1.4 The target flux difference shall be updated at least once per31 Effective Full Power Days by either determining the target flux differencepursuant to Specification 4.2. 1.3 above or by linear interpolation between themost recently measured value and the predicted value at the end of the cyclelife. The provisions of Specification 4.0.4 are not applicable.


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PR

POWER OISTRIBUTION LIMITS

3/4. 2. 2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)


g ALA~~>v+ FrL~>A.t» Alp

3. 2 F (Z) shall be limited by the following relationships: 0(Z) < 2.32 t;K(Z}] for P > 0.5

P

F~(Z) < .64 [K(Z)3 for P < 0.50.

where:

Thermal Po r , andRated hermal er

K(2) is the function obtaine from Figure 3.2-2 or a givencore height location.

APPLICABILITY: MOOE 1

ACTION:

With F~(Z} exceeding its limit:a. Reduce THERMAL POWER at least for ch IX F~(Z) exceeds the limit

within 15 minutes and similar reduce. t e Power Range NeutronFlux - High Trip Setpoints w hin the nex 4 hours; POWER OPERATIONmay proceed for up to a tot of 72 hours; ubsequent POWER OPERATIONmay proceed provided the erpower 4T Trip S oints

have. been reduced at le st 1X for each 3X F~(Z) xceeds thelimit; and

b. Identify and colrec the cause of the out-of-limit ndition priorto increasing. THE L POWER above the reduced power sit required byACTION a., above ERMAL PlNER may then be increased rovided F~(Z)is dhonstrated hrough incore mapping to be within its limit.

TURKEY POINT - UNITS 3 8 4 3/4 2"4Jl" < 9 l'.".'9

,Iel

~ )[

II~

~ ~J

':li~eee

(I.O,

~ I

;. r j[!.

SAI)

)le

!1

II~it!~ 1

~ / f ~

eef'SOA

~ 0.0

g I.i

I

ac Li

ItxNIIRNHTFtl~

FIGURE 3.2-2K(Z) NORMALIZED Fq(Z) AS A FUNCTION OF CORE HEIGHT

TURKEY POINT - UNITS 3 8 4 3/4 2-5 I ~ e

POWER DISTRIBUTION LIMITS


R E.pv~c.C un H.

P PU mLQ14$

4.2.~ The provisions of Specification 4.0.4 are not applicable.4.2.2.2 shall be evaluated to determinetif F~(Z) is within its ~lrmit by:

a.. Using the movable incore detectors to obtain a power*distr'ibution mapat any THERMAL POWER greater than 5X of RATED THERMAL POWER,

b. Increasing the measured Fx component of the power di tribution map

by 3X to account for manufacturing tolerances and rther increasingthe value by 5X to account for measurement uncer inties,

c. Comparing t Fx computed (F ) obtained in ecification 4.2.2.2b.,above to:

d.

1) The F limit for RATED THERMAL P ER (F ) for thexy xyappropriate mea ured core plans ven in Specification4.2.2.2.e. and fi below, and

2) The relationship:

F = F [1+0.2 1-],'yxy

. Where Fx is the li t for actional THERMAL POWER operation

expressed as a fu ction of F nd P is the fraction of RATED

THERMAL POWER which F was sured.xy.Remeasuring Fx a ording to the follow g schedule:

C) .When Fx s greater than the F limit for the appropriateRTP

measur core plane but less than the F relationship,addi onal power distribution maps shall b taken and Fxyc ared to Fx and „ either:RTP E.

Within 24 hours after exceeding by 20% of TED THERMAL

POWER or greater, the THERMAL POWER AT whic Fx was lastC

determined, or xy

b) At least once per 31 Effective Full Powe~ Days EFPD),whichever occurs first.

TURKEY POINT - UNITS 3 4 4 3/4 2-6 o~ ii I 0 a)



gypped~ LDl CA

~L ~QDi ~4

e.

2 ) When the F is less than or equal to the F limit foe ther

appropriate measured core plane, additional power distributionmaps shall be taken and Fx compared to Fx and F„ at leastC RTP ~ L

once per 31 EFPO.

e F limits for RATEO THERMAL POWER (Fx ) shalVbe provided forRTP

xy xyall core planes containing Bank "0" control rods,and all unroddedcore lanes in a Radial Peaking Factor Limit Rep'ort perSpeci ication 6.9.1.6;

The F imits of Specification 4.2.2.2e., above, are not applicablexy /in the fo owing core planes regions as measured in percent of coreheight fro the bottom of the fuel:

1) Lower co e region from 0 to 15X inclusive,

2) Upper core egion from 85 to OOX, inclusive,

3) Grid plane re ions at l7. i 2X;"32.1 4 2X, 46.4 k2X,'0.6x 2X, an 74.9 t, inclusive, and

4) Core plane regioabout the bank de d

With Fx exceeding FC

1) The F~(Z) li it shallexceeds F , and (forusing AP S)

hin t 2X of core height t:4 2.88 inches3position of the Bank "0" control rods.

e reduced at least 2X for each 3X FxC

p nts with F~(Z) less than 2.32 and

2) The e ects of Fx on F~(Z) hall be evaluated to determine ifF~( is within its limits.

4.2.2.3 When F Z) is, measured for other than f determinations, an overallmeasured F~( shall be obtained from a power di tribution map and increased

by 3X to a ount for manufacturing tolerances and urther increased by SX toaccount f r meisurement uncertainty.

TURKEY POINT - UNITS 3 4 4 3(4 2-7~ ~

~ ~

) rl gm /80<-

feR- h/„k


3/4 2.2 HEAT FLUX HOT CHANNEL FACTOR - F Z


UsE,

PLM08-b lh3$

3.2.2 Fq(Z) shall be limited by the following relationships:L

F~(Z) < LF~] X [K(Z)j for P > 0.5

P

Fq(2) < [Fqj X [K(Z)l for P < O.S

~5where: [Fq] = 2. 32 1 imitL

P = Thermal PowerRated Thermal Power

M

[Fqj = The Measured Value,

and K(Z) is the function obtained from Figure 3.2-2 for a givencore height location.

APPLICABILITY: MODE 1

ACTION:

thWith the measured value of Fq(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% F~(Z) exceedsM

F (2) within 15 minutes and similarly reduce the Power Range NeutrontF ux - High Trip Setpoints within the next 4 hours; POWER OPERATION

may proceed for up to a total of 72 hours; subsequent POWER OPERATION

may proceed provided the Overpower Delta-T Trip Setpoints (value of

K4) have been reduced at least 1% for each 1% F~(Z) exceeds theM

F~(Z); and

b. Identify and correct the cause of the out-of-limit condition priorto increasing THERMAL POWER above the reduced power limit required by

ACTION a., above; THERMAL POWER may then be increased provided F~(Z)M

is demonstrated through incore mapping to be within its limit.

3/4 2-4

/ vtAvifidunlit

).e 8/. 6

0



vsE 7H]sFPLMQP-Di dQ

4.2.2.1 If [Fn] ~s predicted by approved physics calculations is greaterthan pFqj and P is greater than PT+as defined in 4.2.2.2, F~(Z)

shall be evaluated by MIDS (Specification 4.2.2.2), BASE LOAD

(Specification 4.2.2.3) or RADIAL BURNDOWN (Specification 4.2.2.4) todetermine if F~ is within its limit ([Fqj = Predicted Fq).

If [F~] , is less than [Fq]" or P is less than

PT, F~(Z) shall be evaluated to dermine if F~(Z) is within itslimit as follows:

a. Using the movable incore detectors to obtain power

distribution map at any THERMAL POWER greater than 5X of RATED

THERMAL POWER.

b. Increasing the measured Fq(Z) component of the power

distribution map by 3$ to account for manufacturing tolerances

and further increasing the value by 5X to account formeasurement uncertainties. Verifying that the requirements

of Specification 3.2.2 are'satisfied.

c. F<(Z) < F<(Z)

Where Fq(Z) is the measured Fq(Z) increased by the allowanceM

for manufacturing tolerances and measurment uncertainty and

Fq(Z) is the F~ limit defined in 3.2;2.L

g p = Reactor sower 'leva) z> mkic~ peed~'~~ed. Fq '»o'~;!T I

8'~CC~ J j <5 l)m< t.

3/4 2-5

0

/A /yl, gw l 4-6

),p- g/~.


SURVEILLANCE RE IREMENTS Continued

ÃE 7ffcsPP1

43oR.D(8$

d. Measuring Fq(Z) according to the following schedule:M

1. Prior to exceeding 75% of RATED THERMAL POWER*, afterrefueling,

2. At least once per 31 Effective Full Power Days.

e. With the relationship specified in Specification 4.2.2.1.c above

not being satisfied:

1) Calculate the percent Fq(Z) exceeds its limit by theM

fol 1 owing expression:

F (Z)

[F<j" X V(Z)/P

- 1 X 100 for P > 0.5

F (Z)

[Fqj X K(Z)/0.5

- 1 X 100 for P ( 0.5

* During power escalation at the beginning of each cycle, power level may beincreased until a power level for extended oper'ation has been achieved andpower distribution map obtained.

3/4 2-6

~ y~ fp pi t(Q H6~f~~ )



UsC 7H u'

~os>]Wp

2) The following action shall be taken:

a) Comply with the requirements of Specification 3.2.2 for

FQ(2) exceeding its limit by the percent calculatedM

above.

4.2.2.2 MIDS

Operation is permitted at power above PT where PT equals the ratio ofLF ]L divided by [FQ] if the following Augmented Surveillance

Q

(Movable Incore Detection System, MIDS) requirements are satisfied:

a. The axial power distribution shall be measured by MIDS when

required such that the limit of LFQ] /P times Figure 3.2.2 isnot exceeded. F>(Z) is the normalized axial power distributionfrom thimble j at core elevation (Z).

I. If F>(Z) exceeds [F>(Z)judas defined in the bases by

< 4%, iomediately reduce thermal power one percent forevery percent by which [F>(Z)]s is exceeded.

2. If F>(Z) exceeds LF>(Z)]s by > 4X imnediately reduce

thermal power below PT. Corrective action to reduce F>(Z)

below the limit will permit return to thermal power not toexceed cvrrent Ifj4 defined in the bases.

+ j'F:(z)J (g ),< o,or,,i i=:'~io Nil)5J

~ 'P„ is raixc4r )Peri cl i.o~-idr exjare~e3. od ci frocf'r~l (

of '( fine'f 's !!5ro( (o coic:,.oii [('(~)j

3/4 2-7

I/ V~ pl~qE'w I r. ~.) /Pic~ l(

kA l/..e

POWER DISTRI BUTION LIMITS ~6PPL


b. F ~ (Z) shall be determined to be within limits by using MIDS tomonitor the thimbles required per specification 4.2.2.2.c at thefollowing frequencies.

1. At least once every 24 hours, and

2. Immediately following and as a minimum at 2, 4 and 8 hours

following the events listed below and every 24 hours there-after.

1) Raising the thermal power above PT, or

2) Movement of control-bank 0 more than an accumulated

total of 15 steps in any one direction.

c. MIDS shall be operable when the thermal power exceeds PT with:

1. At least two thimbles available for which 0> and>

as

defined in the bases have been determined.

2. At least two movable detectors available for mapping F>(Z).

3. The continued accuracy and representativeness of the

selected thimbles shall be verified by using the most

recent flux map to update the 0 for each selected thimble.The flux map must be updated at least once per 31 effectivefull power days.

I

:, her< ~

)/ 'I f—+0[P.l p+op'/lg,gr'<pf f( g~ Q "j(> l ", la '~ i': >.

1

eed. to t„bo. ial peak:~" ~ ~~ <w H~, 4t~, c

44. 'l4v~e!~ !o'coki08 0 (Ec4) fgpp<>@nit

3/4 2-p 7A

C

le& -~'-k ,


SURVEILLANCE RE U IREMENTS Continued

0567+a'p

L~ o P-> /'AQ

4.2.2.3 Base Load

Base Load operation is permitted at powers above PT if the followingrequirements are satisfied:

a. Either of the following preconditions for Base Load operationmust be satisfied.

1. For entering Base Load operation with power less than PT,

a) Maintain THERMAL POWER between PT/1.05 and PT for atleast 24 hours,

b) Maintain the AFD (Delta-I) to within a + 2X or + 3X

target band for at least 23 hours per 24 hour period.

c) After 24 hours have elapsed, take a full core flux map

to determine F~(Z) unless a valid full core flux mapM

was taken within the time period specified in 4.2.2.1d.

d) Calculate PBL per 4.2.2.3b.

2. For entering Base Load operation with power greater than

a) Maintain THERMAL POWER between PT and the power

limit determined in 4.2.2.2 for at least 24 hours, and

maintain Augmented Surveillance requirements of 4.2.2.2during this period.

b) Maintain the AFD {Delta-I) to within a + 2X or + 3X

target band for at least 23 hours per 24 hour period,

7..;~c,c,e;,,'':-, /-,: //g/w ~".~.4



Vs& VH«

La~A>'~Cj

c) After 24 hours have elapsed, take a full core flux map

to determine FQ(Z) unless a valid full core flux map

was taken within the time period specified in 4.2.2.1d.

d) Calculate PBL per 4.2.2.3b.

b. Base Load operation is permitted provided:

1. THERMAL POWER is maintained between PT and PBL or between

PT and 100% (whichever is most limiting).

2. AFD (Delta-I) is maintained within a + 2% or + 3X targetband.

3. Full core flux maps are taken at least once per 31

effective Full Power Days.

PBL and PT are defined as:

PBL [FQ] X K(Z)

Fg~z~ X M~Z~ BLX 1'09

PT = [FQ] /[FQ]

where: FQ(Z) is the measured FQ(Z) with no allowance forM

manufacturing tolerances or measurement uncertainty. For the

purpose of this Specification [FQ(Z)] shall be obtained betweenM

elev~tions bounded by 10% and 90% of the active core height.[Fn] is the FQ limit. K(Z) is given infigure 3.2-2. W(Z)qL is the cycle dependent function thataccounts for limited'ower distribution transients encounteredduring base load operation.

3/4 2-Sl



The function is given in the Peaking Factor Limit Report as per

Specification 6.9.1.6. The 9% uncertainty factor accounts formanufacturing tolerance, measurement error, rod bow and any

burnup and power dependent peaking factor increases.

c. During Base Load operation, if the THERMAL POWER is decreased

below PT, then the conditions of 4.2.2.3.a shall be satisfiedbefore re-entering Base Load operation.

d. If any of the conditions of 4.2.2.3b are not maintained, reduce

THERMAL POWER to less than or equal to PT, or, within 15 minutes

initiate the Augmented Surveillance (MIDS) requirements of4.2.2.2.

4.2.2.4 RADIAL BURNDOWN

I

Operation is permitted at powers above PT if the following RadialBurndown conditions are satisfied:

a. Radial Burndown operation is restricted to use at powers between

PT and PRB or PT and 1.00 (whichever is most limiting).The maximum relative power permitted under Radial Burndown

operation, PRB, is equal to the minimum value of the ratio

of [F~(Z)j/[F~(Z)]RB Meas. where:L

[Fq(Z)]RB Meas. = [Fxy(Z)lMap Meas. x Fz(Z) x 1.09 and

[F~(Z)j is equal to [F~j x K(Z).L L

b. A full core flux map to determine [Fxy(Z)]Map Meas. shall be

taken within the time period specified in Section 4.2.2.ld.2.for the PurPose of the sPecification, [Fxy(Z)]Map Meas. shall

be obtained between the elevations bounded by 10% and 90$ of the

active core height.

3/4 2-EC

~/~ C 8 'i < (. «'V ~6v"i/~ I //

k+ 2/<,.p

tl lttlttilNhtNttttttII tt tl!NNIIIIHIIS

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fill I.'tiff fit lilt iltflfftlflflfl ltlflftfi llllflitlltllllftlllll !!Iftt tfftfltltI IIIlfftIIIII<

tttltttlltt tlttfH ttlftlllftfttltttlttltltttttlIIIIL'll

g5 j~Q, /+



7H cJ

~PLWOt i»AG,

c ~ The function Fz(Z), provided in the Peaking Factor Limit Report

(6.9.1.6), is determined analytically and accounts for the most

perturbed axial power shapes which can occur under axial power

distribution control. The uncertainty factor of 9% accounts formanufacturing tolerances, measurement error, rod bow, and any

burnup dependent peaking factor increases.

d. Radial Burndown operation may be utilized at powers between

PT and PRB, or, PT and 1.00 (whichever is most limiting)provided that the AFD (Delta-I) is within + 5% of the targetaxial offset.

e. If the requirements of Section 4.2.2.4d are not maintained, then

the power shall be reduced to less than or equal to PT, or

within 15 minutes Augmented Surveillance of hot channel factors

shall be initiated if the power is above PT.

4.2.2.5 When F~(Z) is measured for reasons other than meeting the

requirements of specification 4.2.2.1, 4.2.2.2, 4.2.2.3 or

4.2.2.4 an overall measured F~(Z) shall be obtained from

a power distribution map and increased by 3% to account formanufacturing tolerances and further increased by 5% to account

for measurement uncertainty.

3/4 2-$R "7P


3/4.2;3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR


3.2,3 FH

shall be limited by the following relationship:

AH< 1.62 l1.0 + 0.3 (1-P)],N

Whe~e:

THERMAL POWER

RATED THERMAL POWER

APPLICABILITY: MODE l.ACTION:

With F<H exceeding its limit:

a. Within 2 hours either:

2.

Restore F~ to within the above limit, or

Reduce THERMAL POWER to less than 50K of RATEO THERMAL POWER

and reduce the Power Range Neutron Flux - High Trip Setpoint toless than or equal to 55X of RATED THERMAL POWER within thenext 4 hours.

b. Within 24 hours of initially being outside the above limit, verifythrough incore flux mapping that F~ has been restored to within the

above limit, or reduce THERMAL POWER to less than 5X of RATEDTHERMAL POWER within the next 2 hours.

C. Identify and correct the cause of the out-of-limit condition priorto increasing THERMAL POWER above the reduced THERMAL POWER limitrequired by ACTION a.2. and/or b., above; subsequent POWER OPERATION

may proceed provided that F is demonstrated, through fncore fluxN

mapping, to be within the limit of acceptable operation prior toexceeding the following THERMAL POWER levels:

l. A nominal 50K of RATEO THERMAL POWER,.

2. A nominal 75K of RATEO THERMAL POWER, and

3. Within 24 hours of attaining greater than or equal to 95K ofRATED THERMAL POWER.


ICf'l~ p '~7~ 'i ~' '+1<' ll~


SVRYEILLANCE RE UIREMENTS

4. 2. 3. 1 The provisions of Specification 4. Q.4 are not applicable.

4. 2. 3. 2 When a measurement of F<H is taken, the measured FH

shall beN N

nHincreased by 4X to account for measurement error..

4.2.3.3 This corrected F<H shall be determined to be within its limit throughincore flux mapping:

a. Prior to operation above 75K of RATED THERMAL POWER after each fuelloading, and

b. At least once per 31 Effective Full Power Days.

TURKEY POINT - UNITS 3 & 4 3/4 2-9


3/4. 2.4 UAORANT POWER TILT RATIO


3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50K of RATED THERMAL POWER".

ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 butless than or equal to 1.09:

Calculate the QUADRANT POWER TILT RATIO at least once per houruntil either:

a) The QUADRANT POWER TILT RATIO is reduced to withinits limit, or

b) THERMAL POWER is reduced 'to less than 50K of RATED THERMALPOWER.

2. Within 2 hours either:

a) Reduce the QUADRANT POWER TILT RATIO to within itslimit, or

b) Reduce THERMAL POWER at least 3X from RATED THERMAL POWER

for each 3X of indicated QUADRANT POWER TILT RATIO inexcess of 1 and similar ly reduce the Power Range NeutronFlux-High Trip Setpoints within the next 4 hours.

3. Verify that the QUADRANT POWER TILT RATIO is within its limitwithin 24 hours after exceeding the limit or reduce THERMALPOWER to less than 50K of RATED THERMAL POWER within the next2 hours and reduce the Power Range Neutron Flux-High TripSetpoints to less than or equal to 55K of RATED THERMAL POWER

within the next 4 hours; and

Identify and correct the cause of the out-of-limit conditionprior to increasing THERMAL POWER; subsequent POWER OPERATIONabove 50% of RATED THERMAL POWER say'roceed provided that theQUADRANT POWER TILT RATIO is verified within its limit at leastonce per.hour for 12 hours or until verified acceptable at 95Kor greater RATED THERMAL POWER.

*See Special Test Exceptions Specification 3. 10.2.

. TURKEY POINT - UNITS 3 4 4 3/4 2-10



ACTION Continued

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due tomisalignment of either a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hourunti 1 ei ther:


b) THERMAL POWER is reduced to less than SO% of RATED THERMALPOWER.

2. Reduce THERMAL POWER at least 3X from RATED THERMAL POWER foreach lX of indicated QUADRANT POWER TILT RATIO in excess of1, within 30 minutes;

3. Verify that the QUADRANT POWER TILT RATIO is within its limitwithin 2 hours after exceeding the limit or reduce THERMALPOWER to less than 50K of RATED THERMAL POWER within the next2 hours and reduce the Power Range Neutron Flux-High TripSetpoints to less than or equal to 55'f RATED THERMAL POWER

within the next 4 hours; and

4. Identify and correct the cause of the out-of-limit conditionprio~ to increasing THERMAL POWER; subsequent POWER OPERATIONabove 50K of RATED THERMAL,POWER may proceed provided that theQUADRANT POWER TILT RATIO is verified within its limit at leastonce per hour for 12 hours or until verified acceptable at 95Kor greater RATED THERMAL POWER.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due tocauses other than the misalignment of either a shutdown or controlrod:

1. Calculate the QUADRANT POWER TILT RATIO at least, once per houruntil either:


b) THERMAL POWER is ~educed to less than 50K of RATED THERMALPOWER.

TURKEY POINT - UNITS 3 5 4 3/4 2-11~ l ~

l~L~A ~ ~



ACTION Continued

2. Reduce THERMAL POWER to less than 50K of RATED THERMAL POWERwithin 2 hours and reduce the Power Range Neutron Flux-HighTrip Setpoints to less than or equal to 55K of RATED THERMALPOWER within the next 4 hours; and

3. Identify and correct the cause of the out-of-limit conditionprior to increasing THERMAL POWER; subsequent POWER OPERATIONabove 50K of RATED THERMAL POWER may proceed provided that theQUADRANT POWER TILT RATIO is verified within its limit at leastonce per hour for 12 hours or until verified at 95K or greaterRATED THERMAL POWER.

d. The provisions of Specification 3.0.4 are not applicable.


4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within thelimit above SOX of RATED THERMAL POWER by:

4

a. Calculating the ratio at least once per 7 days when the Power RangeUpper Detector High Flux Deviation and Power Range Lower DetectorHigh Flux Deviation Alarms are OPERABLE, and

b. Calculating the ratio at least once per 12 hours during steady-stateoperation when either alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within thelimit when above 75K of RATED THERMAL POWER with one Power Range channelinoperable by using the movable incore detectors to confirm that the normalizedsymmetric power distribution, obtained either from two sets of four symmetricthimble locations or fu11-core flux map> is nsistent with the indicatedgUAD WER-TAT-RATIO yt'least.onc pe hours.

or by Ixcor-c. jkcf'mockup(< ~<p4.2.W 4f the QUADRANT POWER 'fILT RATIO is not within its limit within 24hours and the POWER DISTRIBUTION LIMITS of 3.2.2 and 3.2.3 are within theirlimits, a Special Report in accordance with 6.9.2 shall be submitted within 30days including an evaluation of the cause of the discrepancy.



3/4.2.5 DNB PARAMETERS


>3.2.5 The following DNB-related parametersI„);C lesIlimits ah

a. Reactor Coolant System T

b. Pressurizer Pressure, and

c. Reactor Coolant System Flow


ACTION:

shall be maintained within the fo'lto+IECiL>m tt4 ZT(.3'Fo gag.g.pdicb 'st

) Zaq,BOO qP~

With any of the above parameters exceeding its limit, restore the parameter towithin its limit within 2 hours or reduce THERMAL POWER to less than 5X ofRATED THERMAL POWER within the next 4 hours.


.p gg 6 3 g.,5.13 o,kdvC.211 I «2 «E~ 1112 11; I 2

its limits at least once per 12 hours.

4.2.5.2 The RCS flow rate indic e ubjected to CHANNELAAIBAAI I 2

«Tasm sake'tE'4 5.3 The RCS flow rate shall be demonstrated by measureme once per ~ ~ J

n~$ go ~cgAJI Z4'lbdJ4

~ nbdd 5/d,'er Enc dEA er Ep Tkeimnpo~


TABLE 3.2-1

DNB PARAMETERS

LIMITS

PARAMETER

Indicated Reactor Coolant System Tavg

Indicated Pressurizer Pressure

Indicated Reactor Coolant Flaw

< 576. F

) 2 7 psia"

?7,900 gpaP

I3

iL)Jl ~

I

3/4 2-14TURKEY POINT - UNITS 3 5 4

"Li it not applicable during either a THERMAL POWER ramp in excess of SX ofR TEO THERMAL POWER per minute or a THERMAL POWER step in excess of 10K of

TEO THERMAL POWER.

3/4.3 INSTRUMENTATION

3/4.3. 1 REACTOR TRIP SYSTEM INSTRUMENTATION


3.3. 1 As a minimum, the Reactor Trip System instrumentation channels andinterlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.


4.3. 1.1 Each Reactor Trip System instrumentation channel and interlock andthe automatic trip logic shall be demonstrated OPERABLE by the performance ofthe Reactor Trip System Instrumentation Surveillance Requirements specified inTable 4.3"1.


TABLE 3 ~ 3-1

REACTOR TRIP SYSTEH INSTRUHENTATION

FUNCTIONAL UNITI

1. Hanual Reactor Trip

2. Power Range, Neutron Fluxa a. High Setpoint4h b. Low Setpoint

3. Interaediate Range, Neutron Flux

TOTAL NO.OF CHANNELS

CHANNELSTg TRIP

HINIHUHCHANNELSOPERABLE

33

APPLICABLEMODES

1, 23* 4* 5A

1, 2ltII, 2

leak', 2

ACTION

19

Rag

IFO

4. Source Range, Neutron Fluxa. Startupb. Shutdownc. Shutdown

5. Overteaperature hT

6. Overpower 4T

7. Pressurizer Pressure —Low

Above P-7)8. ressurizer Pressure —High

9. Pressurizer Ha er Level —HighAt)ovc v-7

10. Reactor Coolao Flow-Lowa. Single Loop (Above P-8)

b. Two Loops (Above P-7 andbelow P-8)

3/loop

3/loop

2v

2

2/loop

2/loop

2

2/loop

2/loop

1, 2

1. 2

1, 2

0 )Th. (

,TABLE 3. 3-1 Continued

REACTOR TRIP SYSTEM INSTRUMENTATION

C)

FUNCTIONAL UNITI

ll. Steam Generator MaterLevel-"Low-Low

12. Steam Generator Mater Level-Low Coincident With Steam/Feedwater Flow Hismatch

13. Undervoltage —4.16 KV BussesA and 8 +$~pt ~7)

14. Underfrequency —Trip of ReactorCoolant Pump'.Breaker(s) Open<Above 'P=/)

15. Turbine Tripa. 'Autostop Oil Pressureb. Turbine Stop Valve

(Abaa< 1 -7)


3/stm. gen.

2 stm. g en.level and2 stm./feed-water flowmismatch ineach stm. gen.

2/bus

2/bus

1 stm. gen.level coin-cident with1 stm./feed-water flowmismatch insam'e stm.gen.

1 stm. gen. 1, 2level and2 stm./feed-water'lowmismatch insame stm. gen.or 2 stm. gen.level and 1stm./feedwaterflow mismatchin same stm.gen.

1/bus on 2/busboth busses

1 to trfp gg/bus '//1 1

RCPs*~* .

MINIMUMCHANNELS CHANNELS APPLICABLETO TRIP OPERABLE MOOES

2/stm. gen. 2/stm. gen. 1, 2

7Cm

FUNCTIONAL UNITI

Q 16. Safety Injection Inputfrom ESF


HINIHUHCKANNELS CMANNELSTO TRIP OPERABLE


REACTOR TRIP SYSTEH INSTRUHENTATION

APPLICABLEHODES

1, 2

ACTION

47. Reactor Trip Systea Interlocksa. Interaediate R n e / I - ~~e/

Neutron Fl P-6 (6'CC EP&lgf 2

Tr sP- 0 &<pat'><t 4

orTurbine First 2Stage Pressure

p ~)+II

3 1

1/breaker1/breaker

19. Reactor Trip Breakers

20. Automatic Trip and InterlockLogic

c. Power Ran Neutron(increagi~ pyre) ~

d. Paar Ran Neutronl ( J8 C<076 <+

f'OrJC/'8.

Reactor Coolant Puep BreakerPosition Tripa..Above P-8b. Above P-7 and below P-8

yg 4/~

12

1/breaker1/breaker

1s 2

1, 2 8, 103A 4* 5* 9

1, 2. 123A'* 5* g

TABLE 3.3-1 Continued

TABLE NOTATIONS

"When the Reactor Trip System breakers are in the closed position and theControl Rod Drive System is capable of rod withdrawal.

*"When the Reactor Trip System breakers are in the open position, one or bothof the backup NIS instrumentation channels may be used to satisfy thisrequirement. For backup NIS testing requirements, see Specification 3/4.3.3.3,ACCIDENT MONITORING.

"*"Reactor Coolant Pump breaker A is tripped by underfrequency sensor UF-3A1(UF-4A1) or UF-3Bl(UF-4Bl). Reactor Coolant Pump breakers B and C aretripped by underfrequency sensor UF-3A2(UF-4A2) or UF-3B2(UF-4B2).

SBelow the P-6 (Interme4iate Range Neutron Flux Interlock) Setpoint.

PPBelow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

eACTION 1-

ACTION 2-

ACTION STATEMENTS

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, restore the inoperable channelto OPERABLE status within 48 hours or be in HOT STANDBY withinthe next 6 hours.

With the number of OPERABLE channels one less than the TotalNumber of Channels, STARTUP and/or POWER OPERATION may proceedprovided the following conditions are satisfied:a'. The inoperable channel is placed in the tripped condition

within 1 hour,

b. The Minimum Channels OPERABLE requirement is met; however,the inoperable channel may be bypassed for up to 2 hoursfor surveillance testing of other channels per Specification4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equalto 75K of RATED THERMAL POWER and the Power Range NeutronFlux Trip Setpoint is reduced to less than or equal to85K of RATED THERMAL POWER within 4 hours; or, the QUADRANTPOWER TILT RATIO is monitored per Specification 4.2.4.2.


~ ~


ACTION STATEMENTS Continued

ACTION 3 - With the number of channels OPERABLE one less than the MinimumChannels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)Setpoint, restore the inoperable channel to OPERABLEstatus prior to increasing THERMAL POWER above. the P-6Setpoint, and

ACTION 4-

ACTION 5-

ACTION 6-

ACTION 7-

ACTION 8-

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)Setpoint but below lOX of RATED THERMAL POWER, restore theinoperable channel to OPERABLE status prior to increasingTHERMAL POWER above 10X of RATED THERMAL POWER.

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, suspend all operations involvingpositive reactivity changes.

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, suspend all operations involvingpositive reactivity changes and verify compliance with the SHUTDOWNMARGIN requirements of Specification 3.1. 1.1 or 3. 1.1.2, as appli-cable, within 1 hour and at least once per 12 hours thereafter.

With the number of OPERABLE channels one less than the TotalNumber of Channels, STARTUP and/or POWER OPERATION may proceeduntil performance of the next required ANALOG CHANNEL OPERATIONALTEST provided the inoperable channel is placed in the trippedcondition within 1 hour.

With less than the Minimum Number of Channels OPERABLE, within1. hour determine by observation of the associated permissiveannunciator window(s) that the interlock is in its required statefor the existing plant condition, or apply Specification 3.0.3.

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, be in at least HOT STANDBYwithin 6 hours; however, one channel say, be bypassed for up to2 hours for surveillance testing per Specification 4.3.1. 1,provided the other channel is OPERABLE.

ACTION 9 - Nth the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, restore the inoperable channelto OPERABLE status within 48 hours or open the Reactor TripSystem breakers within the next hour.

ACTION 1Q - With one of the diverse trip features (undervoltage or shunttrip attachment) inoperable, restore it to OPERABLE statuswithin 48 hours or declare the breaker inoperable and applyACTION 8. The breaker shall not be bypassed while one of the

TURKEY POINT - UNITS 3 4 4 3/4 3-6 ~ ~ ~



diverse trip features is inoperable except for the time requiredfor performing maintenance to restore the breaker to OPERABLEstatus.

ACTION ll - With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, be in at least HOT STANOBY within6 hours.

ACTION 12 - With the number of OPERABLE channels one less than the Minimumhannels OPERABLE requirement, be in at least HOT STANDBY within

6 hours; ho~eve~, one channel may be bypassed for up to 4 hours forsurveillance testing per Specification 4.3. 1. 1, provided the otherchannel is OPERABLE. Upon determination of an inoperable relay,the associated channel may be out of service for up to 8 hours, forrepair and testing.

I

ACTION 13 - With the number of OPERABLE channels one less than the TotalNumber of Channels, STARTUP and/or POWER OPERATION may proceeduntil per formance of the next„required ACTUATION LOGIC TESTprovided the inoperable channel ip placed in the tipped conditionwithin 1 hour. P„//>».rr $~<~ ~~„j~/<«-,

~ro rrorrrr r f~r'.err r.'j n

?f'URKEY

POINT " UNITS 3 4 4 3/4 3-7

TABLE 4.3-1

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS

Cl

I

FUNCTIONAL UNITCHANNELCHECK

TRIPANALOG ACTUATING MODES FORCHANNEL DEVICE WHICH

CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCECALIBRATION TEST TEST ~E"

MC/l 1. Hanual Reactor Trip N.A. N.A. N.A. R(13) N.A. 3A 4* 5*

'2. Power Range, Neutron Fluxa. High Setpoint

b. Low Setpoint S

3. Interwediate Range, Scu Neutron Flux

Source Range, Neutron Flux S

5. Overteaperature AT

6. Overpower AT

7. Pressurizer Pressure —Low S

8. Pressurizer Pressure —High S

9. Pressurizer Water Level — S

High

10. Reactor Coolant Flow—Low S

ll. Steam Generator Water Level — SLow-Low

D(2, 4),H(3, 4),Q(4, 6),R(4)R(4)

R(4)

R(4)

R(12)

S/U(1),H

S/U(1),N(9)

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

N.A,

N.A.

N.A.

1, 2

]*Ah'***

2"", 3, 4, 5

1, 2

1, 2

1, 2

1, 2


REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMEHTS

I

FUNCTIONAL UNITCHANNELCHECK

ANALOGCHANNEL

CHANNEL OPERATIONALCALIBRATION TEST

TRIPACTUATING MOOES FOROEVICE WHICHOPERATIONAL ACTUATION SURVEILLANCETEST L l«E

12. Steam Generator Mater Level— S

Low Coincident with SteamlFeedwater Flow Hissatch

13. Undervoltage - 4.16 kV Busses N.A.A and B

N.A.

H.A.

N. A'.

N.A. 1, 2

N.A.

14. Underfrequency - Trip ofReactor Coolant PulpBreaker(s) Open

15. Turbine Tripa. Autostop Oil Pressureb. Turbine Stop Valve

Closure

N.A.

N.A.N.A.

N.A.

N.A.N;A.

N.A. H.A.

. S/U(1, 10) H.A.S/U(1, 10) N.A.

16. Safety Injection Input from N.A.ESF

1T. Reactor Trip Systemr Interlocksa. . Interaediate Range

Neutron. Flux, P"6 N.A.

N.A.

R(4)

N.A.

N.A.

H.A.

N.A.

1 2

b.

C.

Low Power ReactorTrips Block, P-(includes P-10and Turbine FirsStage Pressure)

Power Range NeutronFlux, P-8

H.A.

N.A.

R(4)

R(4)

M(8)

M(8)

N.A.

N.A.

N.A.

N.A.

~ ~~ ~


REACTOR TRIP SYSTEH INSTRUHENTATION SURVEILLANCE

ANALOGCHANNEL

CHANNEL CHANNEL OPERATIONALCHECK CALIBRATION TEST

R(4) H(8)

18. Reactor Coolant Puep Breaker N.A.Position Trip

N.A. N.A.

CA3

ICl

19. Reactor Trip Breaker

20. Autoaatic Trip and InterlockLogic

N.A.

N.A.

N.A.

N.A.

N.A.

N.A.

I

FUNCTIONAL UNIT

v 17. Reactor Tr ip System Interlocks (Continued)

Qe d. Pmer 'RangeD Neutron Flux, P-10 N.A.

RE UIREHENTS

TRIPACTUATINGDEVICEOPERATIONALTEST

N.A. N.A.

N.A.

1, 2

l9H(7,~ g0') N.A.

1'fN.A. H(7,i+)

1, 2, 3*, 4*, 5)91, 2, , (

* *

HODES FORWHICH

ACTUATION SURVE ILLANCEl E«Ettl


TABLE NOTATIONS

"When the Reactor Trip System breakers are closed and the Control Rod DriveSystem is capable of rod withdrawal.

""Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

""*Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

If not performed in previous 7 days.

(2)

(3)

(4)

(5)

(6)

Comparison of calorimetric to excore power indication above 15K of RATEDTHERMAL POWER. Adjust excore channel gains consistent with calorimetricpower if absolute difference is greater than 2X. The provisions ofSpecification 4. 0.4 are not applicable to entry into MODE 2 or l.Single point comparison of incore to excore AXIAL FLUX DIFFERENCEabove 15K of RATED THERMAL POWER. Recalibrate if the absolutedifference is greater than or equal to 3X. The provisions ofSpecification 4.0.4 are not applicable for entry into MODE 2 or l.Neutron detectors may be excluded from CHANNEL CALIBRATION.

This table Notation number is not used.

Incore-Excore Calibration, above 75K of RATH) THERMAL POWER (RTP). Ifthe quarterly surveillance requirement coincides with sustained operationbetween 30K and 75K of RTP, calibration shall be performed at this lowerpower level.

(7)

4e-add+The provisions of Specification 4.0.4

are not applicable for entry into MODE 2 or l.Each train shall be tested at least every 62 days on a STAGGEREO

TEST BASIS.

With power greater than or equal to the Interlock Setpoint the requiredANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that, theinterlock is in the required state by observing the permissive annun-ciator window.

Monthly surveillance in MODES 3", 4", and 5" shall also includeverification that permissives P-6 and P-10 are in their requiredstate for existing plant conditions by observation of the permissiveannunciator window. Monthly surveillance shall include verificationof the High Flux at Shutdown Alarm Setpoint of 1/2 decade above theexisting count rate.

TURKEY POINT - UNITS 3 4, 4 3/4 3-11



(1Q) Setpoint verifica is t applicable.r

<uel'11)At least once p r a following maintenance or adjustment of theReactor trip bre P ACTUATING DEVICE OPERATIONAL TEST shallinclude independent verification of the Undervoltage and Shunt trips.

(12) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(13) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verifythe OPERABILITY of the undervoltage and shunt trip circuits for the ManualReactor Trip Function. The test shall also verify the OPERABILITY. of theBypass Breaker trip circuit(s).

lg~) Interlock Logic Test shall consist of verifying that the interlock is inits required state by observing the permissive annunciator window.

0

, TURKEY POINT - UNITS 3 4 4 3/4 3-12~ ~

~ ~ I ~

INSTRUMENTATION

3/4.3.2 ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION


3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation~ ~

~

~

~

channels and interlocks shown in Table 3.3-2 shall be OPERABLE with their TripSetpoints set consistent with the values shown in the Trip Setpoint column ofTable 3.3-3.


ACTION:

a. With an ESF Instrum ntation or terlock rip Setpoi tr'p lessco ervati e than th value show in the T ip Setpoi col mn butm re cons rvative t an the valu shown i the Allow le V lue c umnf Table 3.3-3, ad 't the Set oint con stent wi the rip S point

~value.

P. With an SSFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in able3.3-3, declare the channel inoperabl'e and apply the applicable ACTIONstatement requirements of Table 3.3-2 until the channel is restoredto OPERABLE status with its Setpoint adjusted consistent with theTrip Setpoint value.

With an ESFAS instrumentation channel or interlock inoperable, takethe ACTION shown in Table 3.3-2.


4.3.2. 1 Each ESFAS instrumentation channel and interlock and the automaticactuation logic and relays shall be demonstrated OPERABLE by performance ofthe ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

TURKEY POINT - UNITS 3 ic 4 3/4 3-13

TABLE 3.3.2

ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION

CI'I

FUNCTIONAL UNIT

1. Safety In'ect (Reactorrsp, ee ater Isolation,

Control Room Isolation, StartDiesel Generators, ContainmentCooling Fans, and EssentialService Mater).


CHANNELSTO TRIP


APPLICABLEHODES ACTION

OiD00I

a. Hanual Initiation

b. Automatic ActuationLogic and ActuationRelays

c.. ContainwentPressure-Hig

d. PressurizerPressure —Low

e. High DifferentialPressure Betweenthe Steam LineHeader and anySteam Line

f. Hi Steam Line Flow-Co ncident with:

ee Steee ~ACjenera@Iressure- r

orhssowT„

'0

3/steam line

2/steam line

1/steam Qgg

1/loop

2/steam linein any steamline

geIIerci&or

1/steam linein any twosteam lines1/steam Qaoin any twosteam lines1/loop in anytwo loops

2

2/steam line

1/steam linein any twosteam lines1/steam Beein any twosteam lines1/loop in anytwo loops

1, 2, 3, 4

1, 2, 3, 4

1, 2, 3

1, 2, 3f

1, 2, 3"

1, 2, 3*

1, 2, 3*

1, 2, 3*

17

14

1SB )g

15

15

15

15

15

yTABLE 3.3-2 Continued

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

Cl

FUNCTIONAL UNIT

2. Containment Spray

a. Automatic ActuationLogig and ActuationRelays

TOTAL NO. CHANNELSOF CHANNELS TO TRIP

MINIMUMCHANNELSOPERABLE

APPLICABLEMODES

1, 2, 3, 4

ACTION

14

00

ColI

Ql

b. Containment Pressure—High-HighCoincident with:Containment Pressure—High

3. Containient Isolation

a. Phase "A" Isolation1) Nanual Initiation2) Autoeatic Actuation

Logic and ActuationRelays

3) Safety Injection

1, 2, 3

1,2,3

1, 2, 3, 41, 2, 3, 4

See Item 1. above for all Safety Injection initiating functions andrequirements. (Manual S.I. initiation sill not initiate Phase AIsolation)

15

15

1714

b. Phase "8" Isolation. I) Hanual Initiation

2) Automatic ActuationLogic and ActuationRelays

2 (Both buttons 2must be pushedsimultaneouslyto actuate)

1, 2, 3, 4

1, 2, 3, 4

17


ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUNENTATION

FUNCTIONAL UNIT

3. Containment Isolation (Continued)


CHANNELSTO TRIP


APPLICABLENODES ACTION

3) Containment 3Pressure —High-HighCoincident with:Containment Pressure—High 3 2

1, 2, 3

1,2,3

15

15

c. Containment Vent)lationIsolation1) Containment Isolation

Manual Phase A orPhase B


3) Safety Injection

4) High Containmentoactivity-

4. Steam Line Isolation

See Items 3.a.l and 3.b.l above for all Hanual ContainmentVentilation functions and requirements.

1, 2, 3, 4

See Item 1. above for all Safety Injection initiating functions andrequirements.

1, 2, 3, 4

16

X6 )g

a. Nanual Initiation(individual)

b. Automatic ActuationLogic and ActuationRelays

1/steam line

1/steam line

1/steam line

1, 2, 3

1, 2, 3

21

20

0


ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION

FUNCTIONAL UNIT

4. Steam Line Isolation (Continued)


CHANNELSTO TRIP

HINIHUNCHANNElSOPERABLE

APPLICABLENODES ACTION

c. Containment Pressure—High-HighCoincident with:Containment Pressure—High

d. Hi Steam Line Flow-Co>ncident with:

2/steam line

Lo Steam 44ee-Qc> ei~gzt- 1/steamPressure

2ev ei'e$ nf"

1/steam inein any twosteam lines1/steam +ice/in any twosteam lines

2

1/steam linein any twosteam lines1/steam kinein any twosteam lines

1 2 3

1,2,31, 2,

15

15

@I x5

Low T„

5. Feedwater isolation

1/loop 1/loop in. anytwo loops

'/loop in anytwo loops

1, 2, 3

a. Automatic ActuationLogic and Actuation .

Relaysb. Safety-injection

6. Auxiliary FeedwaterÃf

1, 2

See Item 1. above for all Safety Injection initiating functions andrequirements.

22

a. Automatic Actuation Logic 2and Actuation Relays

1, 2, 3 20

rg

~ e

C:

m



C)

I

C

Qo

TOTAL NO.FUNCTIONAL UNIT OF CHANNELS

6. Auxiliary FeedwaterfN (Continued)

b. Ste. Gen. Water Level — 3/steamLos-Low generator

CHANNELSTO TRIP

2/steamgeneratorin any

MINIMUMCHANNELSOPERABLE

2/steamgenerator

APPLICABLEMODES

1,2,3

ACTION

~ 15

Ca>I

CD

c. Safety Injection

d. Bus Stripping 1/bus

e. Trip of All NainFeedwater Pueps Breakers 1/Breaker

1, 2, 3

(1/Breaker)/operatingpuep

(1/Breaker)/operatingpulap

1, 2

steam ~eg

See Item 1. above for all Safety Injection initiating functionsand 'requirements.

I/bus 1/bus 23 )P23

7. Loss of Paara. 4.16 kV Busses A and B

(Loss of Voltage)-b. '480 V Load Centers

3A, 3B, 3C, 3D and4A, 4B, 4C, 4D(2 instantaneous relaysper load center)Degraded Voltage

2/bus

2 per loadcenter

2/bus

2 on anyload center

2/bus

2 per loadcenter

1, 2, 3, 4

1, 2, 3, 4

18

18

Coincident with:Safety Injection See Item 1. above for all Safety Injection initiating functions and

requirements



C)

FUNCTIONAL UNITI

7. Loss of Power

Qo

c. 480 V Load Centers3A, 3B, 3C, 3D and4A, 48, 4C 4D(2 inverse timerelays per loadcenter) DegradedVoltage

b. Lee T

9. Control Room Isolation

a. Automatic ActuationLogic and.ActuationRelays

8. Engineered Safety FeaturesActuation System Interlocks

IW a. Pressurizer Pressure

2 per loadcenter

2 on anyload center

TOTAL NO. CHANNELSOF CHANNELS TO TRIP

HINIHUMCHANNELSOPERABLE

2 per loadcenter

APPLICABLEMODES

1, 2, 3, 4

1 2 3

1, 2, 3

2 3 4~*

ACTION

18

19

19 ) 9S'~

b. Safety Injection

c. Hi ContainwentRadioactivity-

See Item 1. above for all Safety Injection initiating functionsand requireaients.

1 2 3 4""" 16

d. Containment Isolation „2Hanual Phase A or Phase 8

1, 2, 3, 4 17


TABLE NOTATIONS

STrip function may be blocked in this MODE below the Pressurizer PressureInterlock Setpoint of 2000 psig.

¹¹Channels are R11 for gaseous radioactivity and R12'for articulatradioactivity.

a

¹¹¹Auxiliary feedwater manual initiation is included in Specification 3.7. 1.2.

*Trip function may be blocked in this MODE below the o T „ -InterlockSetpoint. avg

"" Applicableirradiated

ACTION 14-

ACTION 15-

ACTION 16-

ACTION 17-

)~during MODES 1, 2, 3, 4 or during CORE ALTERATIONS or movement offuel within the containment o> spent fuel pool.

ACTION STATEMENTS

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, be in at least HOT STANDBYwithin 6 hours and in COLD SHUTDOWN within the following30 hours; however, one channel may be bypassed for up to 2 hoursfor surveillance testing per Speci ca o 4.3.2. rovthe other channel is OPERABLE.

With the number of OPERABLE channels one s an eNumber of Channels, operation may procee until performance ofthe next required ANALOG CHANNEL OPENATIONA TEST provided the ) Vrginoperable channel is placed in the tripped conditioq withinI hour. iavlliacI Tvoau a d pfaci~g pc pere cap iae oooo ih place

. AC C4aeneI i n TriP LOWO jADN.,With less than the'inimum Channels OPERABLE requirement, complywith the ACTION statement requirements of Specification 3.3.3. 1Item 1a of Table 3.3-4.

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, restore. the inoperable channelto OPERABLE status within 48 hours or be in at least HOT STANDBYwithin the next 6 hours and in COLD SHUTDOWN within the following30 hours.

),c LIe cIovcrag gocjcts I>E.JS l f ov

Qfgg QQpfATcggs dd vvadaeephM p 'iv'rcto'~~lac lflcaa ~ CdutIduc n Tp ~~ .0

, TURKEY POINT - UNITS 3 8L 4 3/4 3-20

TABLE 3 '-2 Continued


ACTION 18 - With the number of OPERABLE channels one less than the TotalNumber of Channels, STARTUP and/or POWER OPERATION may proceedprovided the inoperable channel is placed in the trippedcondition within 1 hour.

ACTION 19-

ACTION 20-

ACTION 21-

ACTION 22-

ACTION 23-

With less than the Minimum Number of Channels OPERABLE, within1 hour determine by observation of the associated permissiveannunciator window(s) that the interlock is in its requiredstate for the existing plant condition, or apply Specification3. 0. 3.

With the number of OPERABLE channels one less than the HinimumChannels OPERABLE requirement, be in at least HOT STANDBYwithin 6 hours and in at least HOT SHUTDOWN within the following6 hours; however, one channel may be bypassed for up to 2 hoursfor surveillance testing per Specification 4.3.2.1 provided theother channel is OPERABLE.

With the number of OPERABLE channels one less than the TotalNumber of Channels, restore the inoperable channel to OPERABLEstatus within 48 hours or declare the associated valve inoperableand take the ACTION required by Specification 3.7.1.5.

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, be in at least HOT STANDBY within6 hours; however, one channel may be bypassed for up to 2 hoursfor surveillance testing per Specification 4.3.2.1 provided theother channel is OPERABLE.

With the number of OPERABLE channels one less than the MinimumChannels OPERABLE requirement, comply with Specification 3. 0. 3.

I


TABLE 3.3-3

ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT

Safet In'ectian (Reactor Trip,eedwa er so ation, Control

Room Isolation, Start DieselGenerators, Containment CoolingFans, and Essential Service Water)

TRIPSETPOINT ALLOW VALUE

a ~

b.

C.

d.

e.

Manual Initiati on

Automatic Actuation Logic

Containment Pressure —High

Pressurizer Pressure--Low

High Differential PressureBetween the Steam LineHeader and any Steam Line.

N.A.

N.A.

.$l50 psi

.A.

N.

$4.5 psi

?171 ps ig1

51 2 psi

2.

H>gh teamne Flow

c,~ ~0of-C 'ncident with.

o Steam 'essure-Low T „

Containment Spray

Automatic Actuation Logicand Actuation Relays

$A function definedas follows: A hpcorresponding to0.64 x 10e lbs/hrat OX load increas-ing linearly to ahp corresponding to3 84 x 10e lbs/hrat full load

?600 psig

2525 F

S3g

N.A.

$A funct ndefine as 11

A hp rrespo ito .76 x 10 lbs/hrat load i creasinglin arly to a pcorr ndin to3.96 x 1 lb /hr atfull loa .

?58 sig

? 3IF ~

N.A.

b. Containment Pressure-High-HighCoincident with:Containment Pressure-High

530. 0's ig

sP'.0 psig)+ 76

$32.0 p g

54. 5 ig

, TURKEY POINT - UNITS 3 4 4 3/4 3-22

~ ~ ~ p )V ~


ENGINEERED SAFETY FEATURES ACTUATION SYSTEMINSTRUMENTATION TR P SETPOINTS

FUNCTIONAL UNITTRIP

SETPOINT AL ABLE VALUE

3. Containment Isolation

a. Phase "A" Isolation

1} Manual Initiation N.A. N.A.

b.

2) Automatic Actuation Logicand Actuation Relays

3) Safety Injection

Phase "B" Isolation

1) Manual Initiation


N.A. N.

'N.A.

N.A.

See Item 1 above for all SafetyInjection Trip Setpointsiaa4

N

N.A.

3) Containment Pressure-High-HighCoincident with:Containment Pressure —High

530.0 psig

5)0 psig

$32. s ig

$4.5 sig

C. Containment Ventilation Isolation

1) Containment Isolation N.A.Manual Phase A or Phase B

N.A.


N.A. N.A.J

3) Safety Injection

4) High ontainmentioactivity-(1)

See Item 1. above for all Safety InjectionTrip Setpointsi

Particulate (R-ll) (S s ri56.1 x 10s CPH Se pointGaseous (R-12)See (2)

t

TURKEY POINT - UNITS 3 Ec 4 3/4 3-23

~ Q r~ ~ JI


ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT TRIP SETPOINT ALLO BLE VAL

4.

5.

Steam Line Isolation

a. Hanual Initiation

b. Automatic Actuation Logicand Actuation Re1ays

C.

d.

Containment Pressure--High-HighCoincident with:Containment Pressure--High

High team Line Flow-

@~errorCoincident with

ow Steam kkee- ressure-0

avg~Feedwater Iso ation

a. Automatic Actuation LogicActuation Relays

N.A.

N.A.

530.0 psig&

$$.0 psig 8

$A function defineas follows: A hpcorresponding to0.64 x 10e lbs/hrat GX load increas

~ ing linearly to ahp corresponding t3.&4 x 1Qe lbs/hrat full load.

2600 psig

N.A.

I N.A.

„'.A.

$32. psig

54.5

$A functiondefined asfollows: A hpcorres nding to0.76 10'bs/hat OX loadincrea ng linea r'ly

?580 sig

2531

N.A.

to a hp rrespon~to 3a96 x 10s lbat full oad.

6.

b. Safety Injection

Auxiliary Feedwater (3)

See Item 1. above for all SafetyInjection Trip Setpointa, ene

)

b.

Automatic Actuation Logicand Actuation Relays

Steam Generator MaterLevel —Low-Low

N.A.

215K of narrowrange instru-ment span.

214 of narrowran strument jspan.

c. Safety Injection See Item 1. above for all SafetyInjection Trip Setpoints, and

'URKEY POINT - UNITS 3 4 4 3/4 3-24

TABLE 3.3"3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

~

~FUNCTIONAL UNIT

6. Auxiliary Feedwater (Continued)

d. Bus Stripping

e. Trip of All Hain FeedwaterPump Breakers.

TRIP SETPOINT

N.A.

See Item 7. below for all BusStripping Setpoints,

4eiees-.

N.A.

+)!

7. Loss of Power

4.16 kV Busses A and B

(Loss of Voltage)N.A. N.A.

b. 480V Load Centers(Instantaneous Relays)Degraded Voltage

Load enter

3A

3B

3C

30

4B

4C

40

Coincident With:Safety In5ection

~g,y'(8 la436VPBI-sec delay)

~6-Iy(4@'16VQesec delay);

=rll~,'2417VQQ sec delay);

6

Within + 5volts'of

S point

--/&io42f~sec delay)al big

415fR8 sec delay)mv'(hl0

414 ~ sec delay)~svk/0

401V sec delay)h+mfbe

403V secdelay)'ee

Item 1. above for all SafetyInjection Tain Setnoints aad 'j~ )

7


TABLE 3.3-3 (Continued)


FUNCTIONAL UNIT TRIP SETPOINT ALLOWAB VALUE

7. Loss of Power (Continued)

C. 480V Load Centers(Inverse Time Relays)Degraded Voltage

Load Center

3A

3B

3C

30

4B

4C

40

419$ (60 sec 130sec delay)

~eY426V<(60 sec + 30sec delay)

~c/427V(KO sec 430sec Belay)

~ e ~

"436V'(65 sec + 30sec Belay)-

427llg(( eec + 30sec delay)

424V„ sec + 30sec delay)

+m/413V(60 sec + 30sec delay)

+~/412Vf60 sec + 30sec delay)

Wit n t5 voltsof etpoi nt

8. Engineered Safety FeaturesActuation System Interlocks

a. Pressurizer Pressure $2000 psig

b. Low T „~ RID 0set'F

Control Room 4ajoc44en-C sar~ r W

a. Automatic Actuation N.A.Logic and Actuation Relays

$2010 p ig

R531 $535 F

N.

b. Safety Injection See Item 1. above for all SafetyIn)ection Trip Setpoints, eel




FUNCTIONAL UNIT. +~/~pl FVl

c ')F-ioacti vi ty-(1)

d. Containment IsolationManual Phase A or Phase B

TRIP SETPOINT

Particulate (R-Il)S6.1 x 10s CPMGaseous (R-12)See (2)

N.A.

ALLOWABLE VALUE

(Sa as TripSetpoi t)

N.A.

TABLE NOTATIONS

(1) Either the particulate or gaseous channel in the OPERABLE status willsatisfy this LCO.

r3,2 x 10~>(2) Containment Gaseous Monitor Setpoint= .'

CPM,

Where FActual Purge Fl ow

esign Purge Flow 35,000 CFM

Setpoint may vary according to current plant conditions provided thatthe release rate. does not exceed allowable limits provided inSpecification 3.11.2.1.

(3) Auxiliary feedwater manual initiation is included in Specification3. 7. 1. 2.:


CHANNELFUNCTIONAL UNIT

Safet Injection (Reactorrip, ee a er-Isolation,

Control Rooa Isolation,Start Diesel Generators,Containment Cooling Fans,and Essential ServiceMater)

3TABLE 4.3"2

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIONSURVEILLANCE RE UIREMENTS

TRIPANALOG ACTUATINGCHANNEL DEVICE

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATIONCHECK CALIBRATION TEST TEST . LOGIC TEST

HOOES

MASTER SLA i FOR WHICHRELAY REL Y. SURVEILLANCE~E

a. Manual Initiation N.A.

b. Autoaatic Actuation N.A.Logic and ActuationRelays

N.A.

N.A.

N.A.

N.A. N.A.

N.A.

M(1)

N.A. N.A.I 1,2,3,4

c. Contaiment Pressure- N.A.High

d. Pressurizer Pressure- SLow

e. High OifferentialPressure Between theStew Line Header andany Steaa Line

f. High Steaw Line Flow- S Ro ncident with:ow SteaNHee ressure- E. R

or

N.A.

N.A.

N.A.

N.A.

N.A.

M(1)

N.A.

N.A.

N.A.

N.A.

N.A. N.A

N.A N.A.

N.. N.A.

.A. N.A.

.A. N.A.

1, 2, 3

ow Tavg M

iN.A. N.A. - N.A. N.A.

m

CI

I


c 2. Containment Spray

a. Automatic ActuationLogic and ActuationRelays

N.A. N.A. N.A. N.A. H(1)


ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATIONSURVEILLANCE RE UIREHENTS


CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATIONCHECK CALIBRATION TEST TEST LOGIC TEST

MASTER S AVRELAY R LATEST ST

MODES

FOR WHICHSURVEILLANCEIS RE UIREO

1, 2, 3, 4

Ca7

LD

b. Containment Pressure-High-HighCoincident with:Containment Pressure—High

3. Containeent Isolation-

a. Phase "A" Isolation

N.A.

N.A.

N.A.

N.A.

H(l)

H(1)

N.A.

N.A

N.A.( 1, 2, 3

N.A.)

1, 2, 3

1) Hanual Initiation N.A. N.A. N.A. N.A. .A. N.A 1, 2, 3, 4

2) Automatic'Actua-tion Logic andActuation Relays

~N.A. N.A. N.A. N.A. H(1) R 1, 2, 3, 4

3) Safety Injection

b. Phase "8" Isolation

1) Hanual Initiation

2) Automatic Actua-tion Logic andActuation Relays

N.A.

N.A.

N.A.

N.A.

N.A.

N. A. N.A.

N.A.

H(l)

N. A.

See Item l. above for all Safety Injection Surveillance Requirements.

.A,

1, 2, 3, 4

1, 2, 3, 4

R7PCfll

KBCI


ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATIONSURVEILLANCE RE UIREMENTS

TRIPANALOG ACTUATINGCHANNEL OEVICE

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATIONCHECK CALIBRATION TEST . TEST LOGIC TEST

MODES

FOR WHICHSURVEILLANCEIS RE VIREO

MASTER SL E

RELAY R AYTEST T ST

E= CHANNELFUNCTIONAL UNIT

o'. Containment Iso

3) . Conta

1 ation (Continued)

inaent N.h.Pressure-High-High Coincidentwith: ContainwentPressure-High N.A.

c. Containeent Venti-lation Isolation1) Containeent N.A.

IsolationManual Phase Aor Phase B

2) Autoaatic Actua- N.A.tion Logic andActuation Relays

3) Safety Injectio See Item

4). igh Containaent SM

jl g.g. N(1)N.A. N.A. 1,2,3.A.

N.A. N.A. 1,2,3N.A.CAB

00I

ClN.A. N.A. ~ A. N.A.N.A. 1, 2, 3, 4

N.A. N.A. N.A. 1, 2, 3, 4

1. above for all Safety Injection Surveillance Requirements.

R N N.A. N.A. N.A. N... 1, 2, 3, 4

4,

Ra oactivity-Steaa LIne IeolatIona. Hanual Initiationb. Autoaatic Actuation N.A.


N.A.

N.A

N.A.

N.A N.A.

N.A.

R R s ~) a."pN.A. N.A. 1, 2, 3


g


ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATIONSURVEILLANCE RE UIREHENTS

TRIPANALOG ACTUATING HOOESCHANNEL DEVICE HASTER S VE fOR WHICH

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY R LAt SURVEILLANCECHECK CALIBRATION TEST TEST

Steaa Line Isolation (Continued)

c. Containeent Pressure- N.A.High-HighCoincident with:Containment Pressure- N.A.High

d. H teaa Line Sow ~ncident With:

Le Steee &tee(oenord orssures Sor

avgFeedwater o ationa. Autoaatic Actuation N.A.


b. Safety InjectionAuxiliary Feedwater (2)a. Automatic Actuation


b. Steas Generator S

Water Level-Low-Low

P g.A. R(1)

J( /J $ ~ R(1)

N.A. .A.'- 1, 2, 3N.A.

N.A. 'N.A. N.A. 1,2,3

1. 2. 34%N.A. N.A N.A.

N.A. N.A. N. N.A. 1, 2, 33K.V-

1,2,~&N.A. N.A. N A. N.A.

N.A. N.A. N.A. 1, 2

N.A N.A. R gR q 1,2,3

A NA 1,2,3

N.A.

N.A.. N.A.

See Itea 1. above for all Safety Injection Surveillance Requirements.

N A

ClM


ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATIONSURVEILLANCE RE UIREHENTS

TRIP .

ANALOG ACTUATING

~6. AuQo

c. Safety Injection See Itea l. above for all Safety Injection Surveillance Requirements.

N.A. R N.A. R N.A. N

.A

N.A.Bus Strippingd.

N.A.N.A.N.A.N.A.N.A.e. Trip of All HainFeedwater PumpBreakers.

Loss of Power

D04

tQ

I CHANNEL DEVICE HASTER AVECHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY

M~ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TE TEST.

xiliary Feedwatel (Continued)

HOOES

FOR WHICHSURVEILLANCEIS RE UIREO

1, 2, 3

1, 2

a. 4.16 kV Busses Aand B (Loss ofVoltage)

N.A. R N.A. N.A. N.A N.A. 1, 2, 3, 4

b. 4BD V Load Centers3A,38,3C,3D and4A,48,4C,4D(Instantaneousrelays) Degraded

'Voltage

R. N.A. N.A. .A. N.A. 1, 2, 3, 4

C.

Coincident With:Safety Injection

480 V Load Centers3A,38,3C,30 and4A,4B,4C,4D(Inverse Time Relays)Degraded Voltage

See Item l. above for all Safety Injection Surveillance Requirements.

S R N.A. H(1) N.A.<N..A 1, 2, 3, 4


ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATIONPCmSURVEILLANCE RE UIREHENTS

CD

I


CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATIONCHECK CALIBRATION TEST TEST LOGIC TEST

HODESHASTER SL V FOR WHICHRELAY R LA}'URVEILLANCEc= CHANNEL

~ FUNCTIONAL UNIT

cu 8.Qo

Engineered SafetyFeatures ActuationSystem Interlocks

1, 2, 3~a. Pressurizer Pressure N.A. N.A.

N.A.

N.A. N.A. N.A.

I, 2, 3IE+)gN.A. N.A. N.A.b. Low T

Control Room Isolation

N.A.CAR

4 gI + +roa. Automatic Actuation


N.A. N.A. N.A. N.A.

See Item 1. above for all Safety Injection Surveillance Requirements. ~S+~ R H N.A. N.A. N.A.

b. Safety Injection

c. High'ontainmentoactivity

N.A. N.A.d. ContainmentIsolationHanual Phase'or Phase B

.A. N.A.N.A. N.A. 1, 2, 3, 4

TABLE NOTATION

)~)~g

) v

(1) t least every 62 days on a STAGGERED TEST BASIS.(2) ux ary feedwater manual initiation is included in Specification 3.7.1.2.

Applicable in HODES 1, 2, 3, 4 or during CORE ALTERATIONS or movement ofirradiated fuel within the containment or in the spent fuel pool.'++ SN veil/o„,Cg P>o7 r

enquire/fo 5nfi r HQDE. '3.

'"+++ ~~I«~4CL C.'„"~P'.~ ".=:f.F . Sc ueCZ Ceres

INSTRUMENTATION

3/4. 3. 3 MONITORING INSTRUMENTATION

RADIATION MONITORING FOR PLANT OPERATIONS


3.3.3. 1 The radiation monitoring instrumentation channels for plant operationsshown in Table 3.3-4 shall be OPERABLE with their Alarm/Trip Setpoints withinthe specified limits.


ACTION:

a.

b.

C.

With a radiation monitoring channel Alarm/Trip Setpoint for plantoperations exceeding the value shown in Table 3.3-4, adjust the

'etpointto within the limit within 4 hours or declare the channelinoperable.

With one or more radiation monitoring channels for plant operationsinoperable, take the ACTION shown in Table 3.3-4.

The provisions of Specifications 3.0.3 are not applicable.


4.3.3. 1 Each radiation monitoring instrumentation channel for plant operationsshall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNELCALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at thefrequencies shown in Table 4.3-3.

TURKEY POINT - UNITS 3 8s 4 3/4 3-34

TABLE 3.3-4

RADIATION HONITORING INSTRUHENTATION FOR PLANT OPERATIONS

FUNCTIONAL UNIT

1. Containtaent

a. Containaent AbnosphereRadioactivity-High(Par ticulated~h-or Gaseous 4.~(See Note 1.)

ACTION

All" Particulate 26 for HOOES 1, 2, 3, 4$ 6.lxlOsCPH orGaseousSee Rote 2. 27 HODES 5 AND 6

HINIHUHCHANNELS CHANNELS APPLICABLE ALARH/TRIPTO TRIP/ALARH OPERABLE HOOES SETPOINT

b. RCS Leakage DetectionParticulate Radio-activity ~ orGaseous Radioactivity

N.A. 1, 2, 3, 4 N.A. 26

2. Spent Fuel Storage Pool Areas

a. Unit 3 Radioactivity-High Gaseous

<5.5xl0 2 pCi 28CC

b. Unit 4 Radioactivity-High

Gaseous'i

0 va8)

<2.8x10 2 ~Ci 28CC

~~".~ (~PI~C)or

<1. Oxl06CPH

(PR I15)'


TABLE NOTATIONS

During CORE ALTERATIONS or movement of irradiated fuel'within thecontainment comply with Specification 3/4.9. 13.

With irradiated fuel in the spent fuel pits.

Unit 4 Spent Fuel Pool Area is monitored by Plant Vent radioactivityinstrumentation.

Note 1 Either the particulate or gaseous channel in the OPERABLE statuswill satisfy this LCO.

(3.2 x 10')Note 2 Containment Gaseous Monitor Setpoint=

( F

Actual Pur ge F 1 owWhe~e F =

Design Purge Flow (35,000 CFH)

CPM,

Setpoint may vary according to current plant conditions providedthat the release rate does not exceed allowable limits provided inSpecification 3.11.2.1.

ACTION STATEMENTS

ACTION 26- In MODES 1 thru 4: With both the Particulate and GaseousRadioactivity Honitoring Systems inoperable, operation maycontinue for up to 7 days p~ovided:

per 5 ec,<TicaT(ow 5A,b.i 01) A Containment sump level'monitoring system<is OP L ,

2) Appropriate grab samples are obtained and analyzed at leastonce per 24 hours,

3) A Reactor Coolant System water inventory balance isperformed at least once per & hours during steady stateoperation except when operating in shutdown cooling mode,and

4) Containment Purge, Exhaust and Instrument Air Bleed Valvesare maintained closed.

Otherwise, be in at least NOT STANDBY within the next 6 hoursand in COLD SHUTDOWN ~ithin the following 30 hours (ACTION 27

'ppliesin HODES 5 and 6).

TURKEY POINT - UNITS 3 CL 4 3/4 3-36


ACTION STATEMENTS (Continued)

ACTION 27- In MODES 5 or 6 (except during CORE ALTERATION or movement ofirradiated fuel within the containment or spent fuel pool): Withthe number of OPERABLE Channels less than the Minimum ChannelsOPERABLE requirement perform the following:

1) Obtain and analyze appropriate grab samples at least onceper 24 hours, and

2) Monitor containment atmosphere with area radiation monitors.

Otherwise, isolate all penetrations that provide direct accessfrom the containment atmosphere to the outside atmosphere.

During CORE ALTERATION or movement of irradiated fuel withinthe containment or spent fuel pool: With the number ofOPERABLE Channels less than the Minimum Channels OPERABLErequirements, comply with ACTION statement requirements ofSpecification 3.9.9 and 3.9. l3.

ACTION 28 '- With the number of OPERABLE channels less than the MinimumChannels OPERABLE requirement, immediately suspend operationsin the Spent Fuel Pool area involving spent fuel manipulations.


TABLE 4.3"3

RADIATION MONITORING INSTRUMENTATION FOR PLANTOPERATIONS SURVEILLANCE RE UIREMENTS

I

FUNCTIONAL UNIT

cn 1. Containment

o a. Contaiment AtoosphereRadioactivity-Highg~k

ANALOGCHANNEL MODES FOR WHICH

CHANNEL CHANNEL OPERATIONAL SURVEILLANCECHECK CALIBRATION TEST Ettt

Al 1 )~cb. RCS Leakage Detection

CAB

CABI

CO2)

Particulate Radio-activityfRWQ-

Gaseous Radioactivity

1, 2, 3, 4

1, 2, 3, 4

2. Spent Fuel Pool Areas

a. Unit 3 Radioactivity-High Gaseous )</

b, Unit 4 Radioactivity-High Gaseous' H

~ad-63&MhanneIM~-And-B-IB)~

TABLE NOTATIONS(Plawk V~) ) g

* With irradiated fuel in the-fuel storage pool areas.f Unit 4 Spent Fuel Pool Area is Itonitored by Plant Vent radioactivity instrumentation.

~C/.'8/4 /i 4tH/'P+Cl( JPct<i' 'I:. -' 'if~ ~('./// Q/p-

~ ~

)>-/

INSTRUMENTATION

MOVABLE INCORE DETECTORS


3.3.3,2 'he Movable Incore Detection System shall be OPERABLE with:

'a ~ At least 16 detector thimbles when used for recalibration andcheck of the Excore Neutron Flux Detection System and monitoringthe QUADRANT POWER TILT RATIO", and at least 38 detector thimbles

when used for monitoring F~ F~(Z) and F (Z).xyA minimum of two detector thimbles per core quadrant, and'. Sufficient movable detectors, drive, and readout equipment to mapthe above required thimbles.

APPLICABILITY: When the Movable Incore Detection System is used for:

a ~

b.

C.

ACTION:

Recalibration and check of the Excore Neutron Flux DetectionSystem, or

Monitoring the QUADRANT POWER TILT RATIO*, or

Measurement of F>H F~(Z) and Fx (Z).N

With the Movable Incore Detection System inoperable, do not use the system forthe above applicable monitoring or calibration functions. The provisions ofSpecifications 3.0.3 are not apglicable.


4. 3. 3.2 The Movable Incore Detection System shall be demonstrated OPERABLE atleast once per 24 hours by normalizing each. detector output when required for:

a. Recalibration and check of the Excore Neutron Flux Detection System,or

b. Monitoring the QUADRANT POWER TILT RATIO,* or

c. Measurement of F~ F~(Z) and F„ (Z).N

«Exception to the 16 detector thimble requirement of monitoring the QUADRANT

POWER TILT RATIO is acceptable when performing Specification 4.2.4.2 usingtwo sets of four symmetric thimbles.

TURKEY POINT - UNITS 3 & 4 3/4 3-39~ ( I'.'

~

INSTRUMENTATION

ACCIDENT MONITORING INSTRUMENTATION


3.3.3.3 The accident monitoring instrumentation channels shown in Table 3.3-5shall be OPERABLE.


ACTION:

a. As shown in Table 3.3-5

b. The provisions of. Specification 3.0.4. are not applicable,


~c c

4.3.3.3 Each accident monitoring instrumentation channel shall be demonstratedOPERABLE by performance of the CHANNEL CHECK, ANALOG CHANNEL OPERATIONAL TEST tand CHANNEL CALIBRATION at the frequencies shown in Table 4.3-4.


TABLE 3.3"5

ACCIDENT MONITORING INSTRUHENTATION

INSTRUMENTI

Q 1. Containment Pressure (Wide Range)

2. Containment Pressure (Narrow Range)

'Qo Reactor Coolant Outlet TemperatureTT (Wide Range)

Reactor Coolant Inlet Temperature-TCOLD (Mid R g )

TOTALNO. OF

CHANNELS

2

HINIMUHCHANNELSOPERABLE

1

Ch

/ &anne/-2-Oetec-hn~er

APPLI-CABLEMODES ACTIONS

1, 2, 3

1, 2, 3

1, 2, 3

36

31, 32

31, 32

1, 2, 3 31, 32

5

6.

7.

Reactor Coolant Pressure - Mide Range

Pressurizer Mater Level

Auxiliary Feedwater Flow Rate 2/steamgenerator

1/steamgenerator

1 2 3

1, 2, 3

1, 2, 3

31, 32

31, 32

31, 32

8.

9.

10.

12.

13.

Reactor Coolant System Subcooling HarginHonitor

. PDRV Position Indicator (Primary Detector)

PORV Block Valve Position Indicator

-Safety Valve Position Indicator (PrimaryDetector)

Containment Mater Level (Narrow Range)

Containment Mater Level (Wide Range)

2(2)

1/valve

1/valve

1/valve

1(2) 1, 2, 3 31, 32

1/valve

1/valve

1, 2, 3 33

I, 2, 3 32

1, 2, 3 36

1, 2, 3 31, 32

1/valve 1, 2, 3 33

2(1)


ACCIDENT HONITORING INSTRUHENTATION

TOTALNO. OF

INSTRUMENT CHANNELSI

14. In Core Theraocouples (Core Exit Thermo- 4/corecouples) quadrant

c 15. Containment High Range Area Radiation 2Qo

16. Reactor Vessel Level HonitoringSystem

ACTIONS

31, 32

1, 2, 3 34

1, 2, 3 37, 38

HINIHUM APPLI-CHANNELS CABLEOPERABLE MORES

2/core 1, 2, 3

17.

o /g&.

Neutron Flux, Backup NIS (Mide Range)

High Range-Noble Gas Effluent Honitors

1, 2, 3 31, 32

~~--35-~' RO g

lfPf.

Pt tt ttt t~cannel-9)—

b. Unit 3-Spent Fuel Pit Exhaustltttttllttt~

c. Condenser Air Ejectors ad-64~d. Nain Steaa Lines

RMST Mater Level

ALL

ALL

1, 2, 3

1, 2, 3

1, 2, 3

34

34

34

34

31, 32

TABLE NOTATIONS

1. A channel is eight sensors in a probe. A channel is OPERABLE if a minimum of four sensors are OPERABLE.

2. Inputs to this instr~nt are from instrument items 3, 4, 5 and 14 of this Table.

ACTION 31

ACTION 32


ACTION STATEMENTS

With the number of OPERABLE accident monitoring instrumentationchannel(s) less than the Total Number of Channels either restorethe inoperable channel(s) to OPERABLE status within 7 days, or bein at least HOT STANDBY within the next 6 hours and in at least HOTSHUTDOWN within the following 6 hours.

With the number of OPERABLE accident. monitoring instrumentationchannels less than the Minimum Channels OPERABLE, either restorethe inoperable channel(s) to OPERABLE status within 48 hours, or bein at least HOT STANDBY within the next 6 hours and in at least HOTSHUTDOWN within the following 6 hours.

ACTION 33 Close the associated block valve and open its circuit breaker.

ACTION 34

ACTION 35

ACTION 36

ACTION 37

With the number of OPERABLE Channels less than required by theMinimum Channels OPERABLE requirements, initiate the preplannedalternate method of monitoring the appropriate parameters(s),within 72 hours, and:

1) Either restore the inoperable channel(s) to OPERABLE statuswithin 7 days of the event, op.

$ 02) Prepare and submit a Special Rep to the Commission pursuant

to Specification 6.9.2 within ays following the eventoutlining the action taken, the cause of the inoperability, andthe plans and schedule for restoring the system to OPERABLEstatus.

With one or both hydrogen monitor(s) inoperable, comply with ActionRequirements of Specification 3.6.5.

With the number of OPERABLE accident monitoring instrumentationchannels less than the Minimum Channel OPERABLE, either restore theinoperable channel to OPERABLE status within 30 days, or be in atleast HOT STANDBY within the next 6 hours and in at least HOTSHUTDOWN within the following 6 hours.

With the number of OPERABLE channels one less. than the Total Numberof Channels, restore the system to OPERABLE status within 7 days.If repairs are not feasible without shutting down, prepare andsubmit a Special Report to the Commission pursuant to Specification6.9.2 within 30 days following the event outlining the actiontaken, the cause of the inoperability and the plans and schedulefor restoring the system to OPERABLE status;



ACCIPENT MONITORING INSTRUMENTATION

ACTION BB Nith the number of OPERABLE channels less than the Minimum Channels ~OPERABLE requirements, restore the inoperable channel(s) toOPERABLE status within 48 hours. If repairs are not feasiblewithout shutting down:

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

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

3. Restore at least one channel to OPERABLE status at the nextscheduled refueling.

O.


ycm

TABLE 4.3-4

ACCIDENT HONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS

INSTRUMENTC

~ 1. Containment Pressure (Mide Range)

R" 2. Containeent Pressure (Narrow Range)

4 3 Reactor Coolant Outlet Teeperature -THOT (Wide Range)

CHANNELCHECK

HP)

H(fJ

CHANNELCALIBRATION

ANALOGCHANNEL'PERATIONAL

TEST

N.A.

N.A.

N.A.

5.

4 6.QlI

7 0

Reactor Coolant Inlet Temperature -TCOLD (Wide Range)

Reactor Coolant Pressure - Wide Range

Pressurizer Mater Level

Auxiliary Feedwater Flow Rate

N.A.

N.A.

N.A.

8. Reactor Coolant Systea Subcooling Hargin Honitor

9. PORV Position Indicator (Priltary Detector)

10. PORV Block Valve Position Indicator

11. Safety Valve Position Indicator (Prieary Detector)

12. Contaitaent Mater Level (Narrow Range)

13. Containoent Mater Level (Mide Range)

14. In Core Theraocouples (Core Exit Thermocouples)

15. Contaieient - High Range Area Radiation Honitor

16.'eactor Vessel Level Honitoring System

H(g)

HP)

HQ

H(P)

Sgl5

H(P

R

R(g)

R

N.A.

N.A.

N.A.

N.A.

N.A.

H(3)

N.A.


ACCIOENT HONITORING INSTRUMENTATION SURVEILLANCE RE UIREHENTS

4l ~

4 ~IE

High Range - Noble Gas Effluent Honitors

a. Plant Vent Exhaustb. Unit 3 - Spent Fuel Pit Exhaustc. Condenser Air E)ectorsd. Ni St Lf

cu PK'I RMST Mater Level

Cl

INSTRUHENT

~ 17. Neutron Flux, Backup NIS (Mide Range)

gen

CHANNEL CHANNELCHECK CALIBRATION

H R(g)

~ 8['3) R

R~ g(~) R

5~ q(~) R

HP$ R

ANALOG CHANNELOPERATIONAL

TESTI

R(6)

)wy


TABLE NOTATIONS

5

T ese re no ana g cha els. he req remen is s isfi if ehan 1 tes inc des t ver'catio and n essar adj tme of

cha el su th t the hanne outp esu. in pm er dic onsand or al m f nction .

Acceptable criteria for calibration are provided in Table II.F.1-3 ofNUREG-0737.

3. n a ition o req remen in NOD 1, 2 d 3, e sur > lian forhe chan ls sh 1 be rformed ithin ne sur eilla e i rval iorto heatu bove 04F.

Q 4. By observation of the acoustic monitor power light "ON" and test of theassociated alarm.,

5. ~secv meters---5oukc.S c'.P&cy

6. Neutron detectors may be excluded from CHANNEL CALIBRATION.


's i~9 ~y.-

INSTRUMENTATION

FIRE DETECTION INSTRUMENTATION


3.3.3.4 As a minimum, the fire detection instrumentation for each fire detectionzone shown in Table 3.3-6 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the fire detection instrumentis required to be OPERABLE.

ACTION:

a. With any, but not more than one-half the total in any fire zone,Function A fire detection instruments shown in Table 3.3-6 inoper-able, restore the inoperable instrument(s) to OPERABLE status within14 days or within the next 1 hour establish a fire watch patrol toinspect the zone(s) with the inoperable instrument(s) at least onceper hour, unless the instrument(s) is located inside /he containment,then inspect that containment zone at least once pe hours (or moni- ~Ztor the containment air temperature at least once per hour at the 'i

locations listed in Specification 4.6.1.5). ~bc >Ia p

b. With more than one-half of the Function A fire detection instrumentsin any fire zone shown in Table 3.3-6 inoperable, or with any Func-tion B fire detection instruments shown in Table 3.3-6 inoperable, orwith any two or more adjacent fire detection instruments shown inTable 3.3-11 inoperable, within 1 hour establish a fire watch patrolto inspect the zone(s) with the inoperable instrument(s) at leastonce per hour, unless the instrument(s) is located inside~/he. contain-ment, then inspect that containment zone at least once peP'8'hours j>p(or monito~ the containment air temperature at least once per hourat the locations listed in Specification 4.6.1.5).

fire match patrol not established at the 1 ~et-otthe turbine area, ro within one hour, orpr~m n~55it a Special Report to e

The provisions of Specification 3.0.3 are not applicable.


4.3.3.4.1 Each of the above required fire detection instruments which areaccessible during plant operation shall be demonstrated OPERABLE at least onceper 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Firedetectors which are not accessible during plant operation shall be demonstratedOPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST duringeach COLD SHUTOOWN exceeding 24 hours unless performed in the previous 6 months.

4.3.3.4.2 The NFPA Standard 72D supervised circuits supervision associatedwith the detector alarms of each of the above required fire detection instrumentsshall be demonstrated OPERABLE at least once per 6 months.

TURKEY POINT - UNITS 3 5 4 3/4 3-48 .', ", i~ 9 >gyp

INSTRUMENT LOCATION

FIRE ZONE AREA

TABLE 3.3-6FIRE DETECTION INSTRUMENTS

FOR ESSENTIAL E UIPMENT

HEA~xy)"

FLAME

(x/y)SMOKE

~xy

TOTAL NUMBEROF INSTRUMENTS

4 - Aux. Bldg. Corridor El. 10'- Chem. Ora'ry/Shower Tank Room

9 - Laundr emi al Drain Tank Room10 - Pipewll - Unit RHR ea xchanger Room12 - RHR mp 3A Roo13 - RHR ump 3B Room14 - Unit ~R, I]eat Exchanger Room15 - RHR Pump 4A Room16 " RHR Pump 4B Room19 - Unit 3 W Elect Penet Room20 - Unit 3 S Elect Penet Room21 - Instrument Shop22 - Radioactive Laboratory26 - Unit 4 N Elect Penet Room27 - Unit 4 W Elect Penet Room30 - Unit 4 Piping and Valve Room40 - Unit 3 Piping and Valve Room45 - Unit 4 Charging Pump Room

47 - Unit 4 Component Cooling Water Area54 - Unit 3 Component Cooling. Water Area55 - Unit 3 Charging Pump Room58 - Aux Bldg Corridor, El.

18'9- Unit 4 Containment ElectricalPenet. Area"*

60 - Unit 3 Containment ElectricalPenet. Area""

61 - Reactor Control Rod Eqpmt Room - Unit 462 - Computer Room63 - Reactor Control Rod Eqmt Room - Unit 367 - 4160V Switchgear 4B68 - 4160V Switchgear 4A70 - 41 itchgear 3B71 - 4 OY w tchgear 3A72 - E y Diesel Gen B73 - Emergency Diesel Gen A74 - Emergency Oay Tank Room B75 - Emergency Day Tank Room A76 - Unit 4 Turbine Lube Oil Reservoir79A- North-South Breezeway81 - Unit 4 Main Transformer82 - Unit 4 Aux Transformer Area84 - Unit 3 and 4 Aux Feedwater Pump Area

(DC Enclosure Bldg)

(0/4)

(0/4)(0/4)(0/4)

(0/3)(0/3)(1/1){1/1)-(1/0)(0/6) .

(1/0)(1/0)

(5/2)AAA(4/2) M*

(1/0)(1/0)

(2/0)(2/0)(1/0)(11/0)

(2/P)kak(2/0))4k'5/P)AAA

(2/P)aaa(2/0)*AN(5/P)A*A'll/0)

(2/0)(2/0)(8/0)(6/0)(4/0)~~(4/P)AA*(3/0)

(3/0)(18/0)(10/0)

(16/0)

(4/0)(11/0)(4/0)(10/0)(6/0)(10/0)(6/0)(1/0)(1/0)

(4/0)

(3/0)

TURKEY POINT - UNITS 3 8E 4 3/4 3-49

TABL 3.3' ntinued)

FIRE OETEC ON INSTRUHENTATIONFOR ESSENTIAL E UIPHENT

INSTRUMENT LOCATION

FIRE ZONE AREA

TOTAL NUMBEROF INSTRUHENTS

HEAT FLAME SMOKE

~x y)" (x/y) ~x7yy

87-93 "94-95-96-97-98-

101-102-103-104-106-108A-108B-109-110-113-116-119-120-132-

Unit 3 Aux Transformer Area480V Load Center 4A and 4B480V Load Center 4C and 40480V Load Center 3A and 3B480V Load Center 3C and 30Mechanical Equipment RoomCable Spreading RoomRPI Inverter and MG SetsBattery Rack 4BBattery Rack 3ARPI Inverter and MG SetsControl RoomTrain A InvertersTrain B InvertersBattery Rack 4ABattery Rack 3BUnit 4 Feedwater PlatformUnit 3 Feedwater PlatformUnit 4 Intake Cooling Water Pump AreaUnit 3 Intake Cooling Water Pump AreaControl Room Electrical Chase

I

(1/0)

(1/0)(1/0)

(1/0)

(1/0)(1/0)

(2/0)***(2/0)*A*(4/0)AAA(4/0)"aa

(1/0)(2/0)(1/0)(2/0)(1/0)(16/15)(1/0)

(2/0)(17/0)(3/4)(4/4)

(1/0)

TABLE NOTATIONS

* (x/y): x is number of Function A (early warning fire detection and notificationonly) instruments.

y is number of Function B (actuation of Fire Suppression Systems andearly warning fire detection and notification) instruments.

The fire detection instruments located within the containment are notrequired to be operable during the performance of Type A ContainmentLeakage Rate Test.

Installed to meet the requirements of 10 CFR,Part 50, Appendix R,Section III.G.

0r——-A-.fire-watch-patrol-eh&i-be-estabNshe~4espeh4 ~~I

TURKEY POINT - UNITS 3 L 4 3/4 3-50~ I

~ ~ ~

INSTRUMENTATION

RAQIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION

LIMITING CONQITION FOR OPERATION

3.3.3.5 The radioactive liquid effluent monitoring instrumentation channelsshown in Table 3. 3-7 shall be OPERABLE with their Alarm/Trip Setpoints set toensure that the limits of Specification 3.11. 1. 1 are not exceeded. The Alarm/Trip Setpoints of these channe1s shall be determined and adjusted in accordancewith the methodology and parameters in the OFFSITE QOSE CALCULATION MANUAL(OQCM).

APPLIOA8ILITY: At all times mYcctst as A~cdJ sa Tccgim 8.3-7. g 3ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channelAlarm/Trip Setpoint less conservative than required by the abovespecification, immediately suspend the release of radioactive liquideffluents monitored by the affected channel, declare the channelinoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of:radioactive liquid effluentmonitoring instrumentation channels OPERABLE, take the ACTION shownin Table 3.3-7. Restore the inoperable instrumentation to OPERABLEstatus within 30 days and, if unsuccessful, explain in the nextSemiannual Radioactive Effluent Release Report pursuant to Specifi-cation 6. 9. 1.4 why this inoperability was not corrected in a timelymanner.

c. The provisions of Specification 3. 0.3 are not applicable.


4.3.3.5 Each radioactive liquid effluent monitoring instrumentation channelshall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE

CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at thefrequencies shown in Table 4.3-5.


C

m

TABLE 3.3-7

RADIOACTIVE LI UID EFFLUENT HONITORING INSTRUHENTATION

C)

I

C

lA

Qo

INSTRIRIENT

1. Gross Radioactivity Honitors Providing Alarm andAutoaatic Termination of Release

a. Liquid Radwaste Effluent Line+0-:~b. Steae Generator Blowdown Effluent Lin~~

2. Flow Rate Heasureaent Devices

HINIHUMCHANNELSOPERABLE ACTION

35

36

CARI

a. Liquid Radwaste Effluent Line

b. Steaa Generator Slowdown Effluent Lineg

37

lpgaW+ 37

CIAO!f'A0 V

TABLE 3 3-7 Continued

ACTION STATEMENTS

ACTION 35- With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, effluent releases viathis pathway may continue provided that prior to initiatin arelease:

sng a

ACTION 36-

ACTION 37-

a. At least two independent samples are analyzed in accordancewith Specification 4. 11. 1.1. 1, and

b. At least two technically qualified members of the facilitystaff independently verify the release rate calculationsand discharge line valving.

Otherwise, suspend release of radioactive effluents via thispathway.

With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, effluent releases viathis pathway may continue provided grab samples are analyzedfor gross (beta or gamma) radioactivity at a lower limit of detec-tion of no more than 10-~ microCurie/ml or analyzed isotopically(Gamma t a limit o detection of at least 5 x 10 ~ microcurie/ml:

over /p',a. A a t once 12 hours when the specific activity of

the secondary coolant is greater than 0.01 microCurie/gramDOSE E(UIVALENT I-131, or

b. At least once per 24 hours when the specific activity ofthe secondary coolant is less than or equal to0.01 microCur ie/gram DOSE E)UIVALENT I-131.

With the number of channels OPERABLE less than required by theNinimum Channels OPERABLE requirement, effluent releases via--this pathway may continue provided the flow rate is estimatedat least once per 4 hours during actual releases. Pump perfor-mance curves may be used to estimate flow.


TABLE 4.3"5

RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS

CHANNELCHECK

D(3)

Q INSTRUMENT

1. Gross Radioactivity Honitors ProvidingAlarm and Automatic Terminationof

Release'.

Liquid RaGuaate Effluent Lie~~ ~1 BGi) pal)b. Stean Generator Blouuown Effluent Line+lMQ 0(g)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line D(3)

b. Steam Generator Blowdown Effluent Line.

C CK CALIBRATION

R(2)*

R(2)

R*

SOU E CHANNEL

ANALOGCHANNEL

OPERATIONALTEST

(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway andcontrol room alarm annunciation occur if the instrument indicates measured levels above the Alarm/TripSetpoint.

~Channel calibration frequency shall be at least once per 18 months.

TABLE NOTATIONS

( t

(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certifiedby the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers thatparticipate in measurement assurance activities with NBS. These standards shall permit calibrating thesystem over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION,sources that have been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECKshall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are

~ ~ -~-

INSTRUHENTATION

RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUHENTATION

LIHITING CONDITION FOR OPERATION

3.3.3,6 The radioactive gaseous effluent monitoring instrumentation channelsshown in Table 3.3-8 shall be OPERABLE with their Alarm/Trip Setpoints set toensure that the limits of Specifications 3. 11.2. 1 and 3. 11.2.5 are not exceeded.The Alarm/Trip Setpoints of these channels meetin'g Specification 3. 11.2. 1shall be determined and adjusted in accordance with the methodology andparameters in the ODCH.

APPLICABILITY: As shown in Table 3.3-8

ACTION:

a.

b.

With a radioactive gaseous effluent monitoring instrumentationchannel Alarm/Trip Setpoint less conservative than required by theabove specification, immediately suspend the release of radioactivegaseous effluents monitored by the affected channel, declare thechannel inoperable or change the setpoint so it is acceptablyconservative.

With less than the minimum number of radioactive gaseous effluentmonitoring instrumentation channels OPERABLE, take the ACTION shownin Table 3,3-8. Restore the inoperable instrumentation to OPERABLEstatus within 30 days and, if unsuccessful explain in the next Semi-ann'ual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.4 why this inoperability was not corrected in a timelymanner.

c. The provisions of Specification 3. 0.3 are not applicable.


4.3.3.6 Each radioactive gaseous effluent monitoring instrumentation channelshall be demonstrated OPERABLE by performance of the CHANNEL„.CHECK, SOURCE

CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at thefrequencies shown in Table 4.3-7.

TURKEY POINT - UNITS 3 AND 4 3/4 3-55 '. v~ l."

Ã77Cm

TABLE 3.3-8

RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION

IC

Qe

INSTRUHENT

1. GAS DECAY TANK SYSTEH

a. Noble Gas Activity Honitor-Providing Alarm and AuteaaticTeraination of Release(Plant Vent Honitor,-i&44)

HINIHUM CHANNELSOPERABLE APPLICABILITY ACTION

45

b. Effluent Systea Flow RateHeasuring Device~arMSN

Wl

a 2. GAS DECAY TANK SYSTEH (Explosive Gasc, Honitoring Systee)

a. Hydrogen and Oxygen Honitors

3. Condenser Air Ejector Vent System

a. Noble Gas Activity Ho itor~S-or-(5PAJ& or PÃg5

c.'articulate-Saap1 -NN

gf. Effluent Systea Flow Rate Heasuring Device 1 N

Sampler Flow Rate Heasuring Device

46

49

47

.48—~4a

46

C:

7c


RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION

CIHINIHUH CHANNELS

OPFRABLEINSTRUHENTI

Q 4. Plant Vent Systea (Includes Unit 4'sSpent Fuel Pool)

a. Noble Gas Activity Honitor~4-er - 1RacH5 S Pi'lJ6 os- PP+jg

APPLICABILITY ACT ION

48

Effluent System Flm Rate HeasuringDevice.

-Channe~+-

Saapler Flm Rate Heasuring Device~d-65. Unit 3 Spent'Fuel Pit Building Vent

a. Noble Gas Activity Nonitor-6hanne~

-h—.-XocÃ

46

47

-* ———.- -- 48 ss

Soapier Flow Rate Neasurieg Oevice~Md

46

)


TABLE NOTATIONS

At all times."" During GAS DECAY TANK SYSTEM operation.Applies during MODE 1, 2, 3 and 4.

H Applies during MODE 1, 2, 3 and 4 when primary to secondary leakage isdetected as indicated by condenser air ejector noble gas activity monitor.

ACTION STATEMENTS

ACTION 45 - With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, the contents of thetank(s) may be released to the environment provided that priorto initiating the release:

a. At least two independent samples of the tank's contentsare analy'zed, and

b. At least two technically qualified members of the facilitystaff independently verify the release rate calculationsand discharge valve lineup.

Otherwise, suspend release of radioactive effluents via thispathway.

ACTION 46 - With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, effluent releases viathis pathway may continue provided the flow rate is estimated atleast once per 4 hours.

ACTION 47-

ACTIO 48 "

With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, effluent releases viath'is pathway may continue provided grab samples are taken at,least once per 12 hours and these samples are analyzed forradioactivity within 24 hours.

With the n er of annels PERAB less than quire y theMin'mum Ch nnels 0 RABLE r quir ent, e fluen relea es viath affec ed pat ay may c ntin provi ed s les e contu usly c lected ith aux lia saapl g eq pment s requ red

n Tab 4.11- and anal zed least weekly.

'gl

ACTION 49 - With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, operation of the GASDECAY TANK SYSTEM may continue rovided thatgrab samples are collected and analyzed for hydrogen and oxygenconcentration at least a) once per hour s during degassingoperations, and b) once per day dur ng other operations.

, TURKEY POINT - UNITS 3 4 4 3/4 3-58 ~,g P~ ~ ~, ~~ IOCi

TABLE 4.3-6

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS

RR

CO

C

Cia

INSTRUMENT

1. GAS DECAY TANK SYSTEM

CHANNELCHECK

URCE

CHECCHANNEL

CALIBRATION

ANALOG

CHANNELOPERATIONAL

TEST

MODES FOR WHICHSURVEILLANCE

IS RE UIRED

a. Noble Gas Activity Monitor-Providing Alara and AutoeaticTernination of Release p(g) P(y)- (Plant Vent Monitor,~+)

b. Effluent Systea Flow Rate<ss

g 2. GAS DECAY TANK SYSTEM (ExplosiveGas Monitoring Systea)

a. Hydrogen and Oxygen Monitors

N.A.

N.

R(3) '(1)

N.A.

q(4,5) i*3. Condenser Air Ejector Vent System

a. Noble Gas Activity Nonitor~~ D(p)~~~Co'(3) 9(2)

Effluent System Flow RateMeasuring Device

N.A. N.A.

C:R77C


RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUHENTATION SURVEILLANCE RE UIREHENTS

RgCI

CHANNELCHECK

'I

INSTRUHENT

R 3. Condenser Air E)ector Vent SystemR (Continued)Qe

d G p'. Saepler Fine Rate Heaauring ~ D

Device eHi4~WrQndicato~

S RC

CH K

.A.

CHANNELCALIBRATION

ANALOGCHANNEL

OPERATIONALTEST

N.A.

HODES FOR WHICHSURVEILLANCEEll

RAA

CARI

OlC)

4. Plant Vent Systea (Includes Unit4's Spent Fuel Pool)

a. Noble. Gas Activity Honitor < g(7) (3,6) 0(2)

4;A—.— *

'.A

—." —H:A:—

P'-

Effluent Systea Fle» Rate., Heasur ing Oevice+hd~4-

nnel-~-

~ p, Sampler Flow Rate HeasuringII I -Ed daddd~

N.

N

(6)

(6)

N.A.

N.A.


RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS

Cl

Cib

INSTRUMENT

5. Unit 3 Spent Fuel Pit Building Vent

a. Noble Gaa Activity Hooitor

CHANNELCHECK

G(7)

RCE CHANNELCHEC CALIBRATION

- R(3)

OPERATIONALTEST

SURVEILLANCEIS RE UIRED

0(2)

ANALOG

CHANNEL MODES FOR WHICH

CayI

CJl

Sampler Flow Rate Measuring 0D

N.A.

TABLE NOTATIONS

N.A.

* At all times.-"* During GAS DECAY TANK SYSTEH operation.

Applies during NDES 1, 2, 3 and 4Applies during NDES 1, 2, 3 and 4 when primary to secondary leakage is detected as indicated bycondenser air ejector noble gas activity monitor

(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway andcontrol room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/TripSetpoint.

o~ Co~pAe,t- >co~(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room~alarm annunciation ~&nf-

ter-roea+ occurs if the instrument indicatesmeasured levels above the Alarm Setpoint.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified bythe National Bureau of Standards (NBS) or using standards that have been obtained from suppliers thatparticipate in measurement assurance activities with NBS. These standards shall permit ca1ibrating thesystem over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sourcesthat have been related to the initial calibration shall be used.

TABLE 4, 3" 6 Continued


(4) The CHANNEL CALIBRATION shall include the use of standard gas samplescontaining a-nominal:

a. One volume percent hydrogen, balance nitrogen, and

b. Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall include the use of standard gas samplescontaining a nominal:

a. One volume percent oxygen, balance nitrogen, and

b. Four volume percent oxygen, balance nitrogen.

(6) CHANNEL CALIBRATION frequency shall be at least once per 18 months.

Pf) c8AANSL ~au™.p crnsisk a4 o. 5'oUR.cG c H-fc,K ~ ) 0 z


3/4. 4 REACTOR COOLANT SYSTEM

3/4.4. 1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

STARTUP AND POWER .OPERATION


3.4.1.1 All reacto~ coolant loops shall be in operation.


ACTION:

With less than the above required reactor coolant loops in operation, be inat least HOT STANDBY within 6 hours.


4.4. 1. 1 The above required reactor coolant loops shall be verified inoperation and circulating reactor coolant at-least once per X2 hours.

TURKEY POINT - UNITS 3 4 4 3i4 4-1~ ij C

A

REACTOR COOLANT SYSTEM

HOT STANDBY


3.4. 1.2 All of the reactor coolant loops listed below shall be OPERABLE withall reactor coolant loops in operation when the Reactor Trip Sys4am- breakersare closed and two reactor coolant loops listed below shall be OPERABLE with atleast one reactor~ooiant loop in operation when the Reactor Trip k~r.breakers are open .

> Q ~f 7a. Reactor Coolant

reacto~ coolantb, Reactor Coolant

reactor coolantc. Reactor Coolant

reactor coolantAPPLICABILITY: MODE 3.

Loop A and its associated steam generator andpump,

Loop B and its associated steam generator andpump, and

Loop C and its associated steam generator andpump.

ACTION:

b.

C.

With less than the above required reactor coolant loops OPERABLE,restore the required loops to OPERABLE status within 72 hours or bein HOT SHUTDOWN within the next 12 hours.

With less than three reactor coolant loop in operation and the ReactorTrip i~wbreakers in the closed position, within 1 hour open theReactor Tr Ip C~e breakers.With no reactor coolant loop in operation, suspend all operationsinvolving a reduction in boron concentration of the Reactor CoolantSystem and immediately initiate corrective action to return therequired reactor coolant loops to operation.


4.4. 1.2. 1 At least the above required reactor coolant pumps, if not inoperation, shall be determined OPERABLE once per 7 days by verifyingcorrect breaker alignments and indicated power availability.

4.4. 1.2.2 The required st'earn generators shall be determined OPERABLE by verifyingsecondary side water level to be greater than or equal to 1 at least onceper 12 hours. C~rroQ i-a nq44.4.1.2.3 The required reactor coolant loops shall be verified in operationand circulating reactor coolant at least once per 12 hours.

"All reactor coolant pumps may be deenergized for up to 1 hour provided:(1) no operations are permitted that would cause dilution of the ReactorCoolant System boron concentration, and (2) core outlet temperature ismaintained at least 104F below saturation temperature.~ As ~~ a~~«~"-"<~ '4 '5c renvoi trip

5C. md c~+~.t sos m y bc PI ~ ~ ~A. b aK S i~No@, ~)4, ca. gg„+g~ p~k, Qe+wJ


0


HOT SHUTDOWN


3.4.1.3 At least two of the loops listed below shall beone of these loops shall be in operation:"

OPERABLE and at least

a. Reactor Coolant Loop A and its associated steam generator andreactor coolant pump,""

b.

C.

d.

e.

Reactor Coolant Loop B and its associated steamreactor coolant pump,"*

Reactor Coolant Loop C and its associated steamreactor coolant pump,""

RHR Loop A, d

RNR Loop ~4~

generator and

generator and


ACTION:

a. With less than the above required loops OPERABLE, immediatelyinitiate corrective action to return the required loops to OPERABLEstatus as soon as possible; if the remaining OPERABLE loop is an RHR

, loop, be in COLD SHUTDOWN within 24 hours.

b. With no loop in operation, suspend all operations involving a reduc-tion in boron concentration of the Reactor Coolant System andimmediately initiate corrective action to return the required loop(5)to operation..

"All reactor coolant pumps and RHR pumps may be deenergized for up to 1 hourprovided: (1} no operations are permitted that would cause dilution of theReactor Coolant System boron concentration, and (2) core outlet temperatureis maintained at least 10 F below saturation temperature.

*~A reactor coolant pump shall not be started with one or more of the Reacto~Coolant System cold leg temperatures less than or equal to 275oF unlessthe secondary water temperature of each steam generator is less than 504Fabove each of the Reactor Coolant System cold leg temperatures.

A> IAcyli )(o. ggP t Oo[ fA~g Q AO. OP&griWl+n P'HR lPO)

y» o Wo z L,~Tg Q» yurt.;i)~.c~ feofi~g. e~)~t(is( o)«~4lQ).



HOT SHUTOOWN

SURVEILLANCE RE UIREHENTS 04.4.1.3.1 The required reactor coolant pump(s), if not in operation, shall bedetermined OPERABLE once per 7 days by verifying correct breaker alignments andindicated power availability.

4. 4. 1. 3. 2 The required steam generator(s) shall be determined OPERABLE byverifying secondary side water level to be greater than or equal to lOX atleast once per 12 hours. anooo «~g) ~g4.4.1.3.3 At least one reactor coolant or RHR loop shall be verifie noperation and circulating reactor coolant at least once per 12 hours.

TURKEY POINT - UNITS 3 8E 4 3l4 4-4 l1'o~ ~ Iv) l


COLO SHUTOOWN - LOOPS FILLED


3.4. 1.4.1 At least one residual heat removal (RHR) loop shall be OPERABL&andin operation", and either:

a. One additional RHR loop shall be OPERAB P , or

b. The secondary side water level of at least two steam generatorsshall be greater than 10K.

APPLICABILITY: MOOE 5 with reactor coolant loops filled""".ACTION:

a.

b.

With one of the RHR loops inoperable or with less than the requiredsteam generator water level, immediately initiate corrective actionto return the inoperable RHR loop to OPERABLE status or restore therequired steam generator water level as soon as possible.

With no RHR loop in operation, suspend all operations involving areduction in boron concentration of-.the Reactor Coolant System andimmediately initiate corrective action to return the required RHR

loop to operation.


4.4. 1.4. 1. 1 The secondary side water level of at least two steam generatorswhen required shall be determined to be within limits at least once per12 hours,

4.4. 1.4. 1. 2 At least one RHR loop shall be determined to be in operation andcirculating reactor coolant at least once per 12 hours.

"The RHR pump may be deenergized for up to 1 hour provided: (1) no operationsare permitted that would cause dilution of the Reactor Coolant System boronconcentration, and (2) core outlet temperature is maintained at least 104Fbelow saturation temperature.

""One RHR loop may be inoperable for up to 2 hours for surveillance testingprovided the other RHR loop is OPERABLE.

"""A reactor coolant pump shall not be started. with one or more of the ReactorCoolant System cold leg temperatures less than or equal to 275oF unlessthe secondary water temperature of each steam generator is less than 504Fabove each of the Reactor Coolant System cold leg temperatures.



COLD SHUTDOWN " LOOPS NOT FILLED


3.4. 1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE" and atleast one RHR loop shall be in operation.""

APPLICABILITY: MODE 5 with reactor coolant loops not filled.ACTION:

b.

With less than the above required RHR loops OPERABLE, immediatelyinitiate corrective action to return the required RHR loops toOPERABLE status as soon as possible.

With no RHR loop in operation, suspend all operations involving areduction in boron concentration of the Reactor Coolant System andimmediately initiate corrective action to return the required RHR

loop to operation.


4.4. 1.4.2 At least one RHR l.oop shall be determined to be in operation andcirculating reactor coolant at least once per, 12 hour s.

"One RHR loop may be inoperable for up to 2 hours for surveillance testingprovided the other RHR loop is OPERABLE.

""The RHR pump may be deenergized for up to 1 hour provided: (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant Systemboron concentration, and (2) core outlet temperature is maintained at least10'F below saturation temperature.



."j'd. I" 5e peg] a.l le&

'y OPERATING


.13.4.2.Z A'll pressurizer Code safety valves shall be OPERABLE with a lift setting) Qof 2485 psig a lX."

APPLICABILITY: MOOES 1, 2, and 3.

ACTION:

With one pressurizer Code safety valve inoperable, either restore theinoperable valve to OPERABI E status within 15 minutes or be in at least HOTSTANDBY within 6 hours and in at least HOT SHUTOOWN within the following6 hours.


14.4.2.Z No additional requirements other than those required bySpecification 4.0.5.

"The lift setting pressure shall correspond to ambient conditions of the valveat nominal operating temperature and pressure.



~evts- SAFETY VALVES

SHUTDOWN


3.4.2M A minimum of one pressurizer Code safetya lift etting of 2485 psig a lX."

QQD~ H . ("QAPPLICABILITY: MOD@ 4 and~5~

~Ac'~I"~~I ed,MACTION:

valve shall be OPERARLE with ) C

1r ~ure l'.-: J

With no pressurizer Code safety valve OPERABLE, immediately suspend alloperations involving positive reactivity changes-loop into-operation i'~ ~n-cool-i-ng-mode:.exbep. ~eT as OPERA Lie /gal Iaop >4ss~~ ssv. plat Ysg lave

p+ „,4u. sI flan ooI~ Mp


4.4.2.Z No additional requirements other than those required bySpeci 'tion 4.0.5.

"The lift setting pressure shall correspond to ambient condit4ons of the valveat nominal operating temperature and pressure.~ RCS PPOS~ure BOus 9 F"y ~ ~~4~~Y~A LFFile~e W PC~ig

+e' ~Ysvs~n+nTh~vssv~ a Vvs~sn V slperu pTURKEY POINT - UNITS 3 4 4 3/4 ~

C~ g'


3/4.4.3 PRESSURIZER


3.4.3 The pressurizer shall be OPERABLE with a water volume of less than orequal to 92K of indicated level, and at least two groups. of pressurizer heaterseach having a capacity of at least 125 kW and capable of being supplied byemergency power.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at leasttwo groups to OPERABLE status within 72 hours or be in at least HOTSTANDBY within the next 6 hours and in HOT SHUTDOWN within thefollowing 6 hours.

With the pressurizer otherwise inoperable, be in at least HOT STANDBYwith the Reactor Trip System breakers open within 6 hours and in HOTSHUTDOWN within the following 6 hours.


4.4.3. 1 The pressurizer water volume shall be determined to be within itslimit at least once per 12 hours.

4.4. 3. 2 The capacity of each of the above required groups of pressurizerheaters shall be verified by energizing the heaters and measuring circuitcurrent at least once per 92 days.

4.4.3.3 The emergency power supply fog the urizel heaters shall bedemonstrated OPERABLE at least once p6r by demonstrating thecapability to power the heaters from the'mergency power.

IN,I.vC,tin) ho4 ~+ <<<~2 Cj msn44



3 .4.4 REi;.IEF VALVES

LIHITI CONDITION FOR OPERATION

INsazT F8-g ~g.l~

3.4.4 All p er"operated relief valvesvalves shall b OPERABLE.

APPLICABILITY: H ES 1, 2, and 3.

ACTION:

iPORVs) end their dssocieted block

a r

b,

With one or m e PORV(s) inoperable, ithin 1 hour either restore thePORV(s) to OPE BLE status or clos the associated block valve(s) andremove power fro the block valv s); otherwise, be in at least HOTSTANDBY within th next 6 hour and in COLD SHUTDOWN within the fol-lowing 30 hours.

With one or more block va) e(s) inoperable, within 1 hour either restothe block valve(s) to 0 'RABLE status, or close the block valve(s) andremove power from the o k valve(s); otherwise, be in at least HOTSTANDBY within the n t 6 urs and COLD SHUTDOWN within thefollowing 30 hours.


4.4.4. 1 In addition the requirements of ecification 4.0.5, each PORVshall be demonstrate OPERABLE at least once p r 18 months by:

a. Perfor nce of a CHANNEL CALIBRATION, a d

b. Oper ting the valve through one complete cle of full travel withth motive force supplied by the normal Ins rument Air System ant e backup Nitrogen Gas System.

4.4.4.2 ach block valve shall be demonstrated OPERABLE t least once per92 day by operating the valve through one complete cycle f full travelunles the block valve is closed with power removed in orde to meet the r quire-

n of Specification 3.4.4 or is closed due to PORV leakag

.TURKEY POINT - UNITS 3 5 4 3/4 4-10


3/4.4.4 PORV BLOCK VALVES

LIMITING COND'ITION FOR OPERATION

Cfg( S

/pe. RR+Rx74

3.4.4 Each Power Operated Relief Valve (PORV) Block valve shall be OPERABLE.

APPLICABILITY: MODES I, 2 and 3.

ACTION:

a.

b.

With one or more block valve(s) inoperable, within 1 hour eitherrestore the block valve(s) to OPERABLE status or close the blockvalve(s) and remove power from the block valve(s); otherwise, be inat least HOT STANDBY within the next 6 hours and in SHUTDOWNwithin the following Pf hours.

The provisions of Specification 3.0 .4 are not applicable.


4.4.4 Each block valve shall be demonstrated OPERABLE at least once per92 days by operating the valve through one complete cycle of full travelunless the block valve is closed with power removed in order to gee theqi fAkl . 'p if' 3.4.4. '

f ~

pro v J ~ ( so)

f' l'

Ie

f \

l1


3/4.4.5 STEAM GENERATORS


3.4.5 'ach steam generator shall be OPERABLE.


ACTION: h~~~With one or more steam generators inoperable, restore the inoperable generator(s)to OPERABLE statu 'v CIt 4L E< QfHOT AU'D ()(~4 ~ Co kc(or~ >Q (~ COLD <HU~L,(g (O(qg ~(~

SlfkE1LLANCE RE REBATE

4. 4. 5, 0 Each steam generator shall be demonstrated OPERABLE by performance ofthe following augmented inservice inspection-proSpecification 4.0.5. -~ g ~, „ $ p ~~ z4. rn~Qs

4.4.5. 1 Steam Generator Sam le S 1 tion and In ection - Each steam generatorshall be determined OPERABLE duri g by select ng and inspecting atleast the minimum number of steam gen rator< specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sam le Selection and Ins ection - The steamgenerator tube minimum sample size, inspect>on result classification, and thecorresponding action required shall be as specified in Table 4.4-2. Theinservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall beverified acceptable per the acceptance criteria of Specification 4.4.5.4. Thetubes selected'for each inservice inspection shall include at least 3X of thetotal number of tubes in all steam generators; the tubes selected for theseinspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistryindicates critical areas to be inspected, then at least 50K of thetubes inspected shall be from these critical areas;

b. The first sample of tubes selected for each inservice inspection(subsequent to the preservice inspection) of each steam generatorshall include:

TURKEY POINT " UNITS 3 4 4 3/4 4-11


STEAM GENERATORS


1) All nonplugged tubes that previously had detectable wallpenetrations (greater than 20X),

2) Tubes in those areas where experience has indicated potentialproblems, and

3) A tube inspection (pursuant to Specification 4.4.5.4ar8) shallbe performed on each selected tube. If any selected tube doesnot permit the passage of the eddy cur rent probe for a tubeinspection, this shall be recorded and an adjacent tube shallbe selected and subjected to a tube inspection.

c. The tubes selected as the second and third samples in the inserviceinspection may be less than a full tube inspection by concentrating(selecting at least 50X of the tubes to be inspected) the inspectionon those areas of the tube sheet 'array and on those portions of thetubes where tubes with imperfections. were previously found.

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

~Cate or Ins ection Results

C-1 Less than 5X of the total tubes inspected aredegraded tubes and none of the inspected tubesare defective.

C"2

C-3

One or more tubes, but not more than IX of thetotal tubes inspected are defective, or between5X and 10X of. the total tubes inspected aredegraded tubes.

More than 10X of the total tubes inspected aredegraded tubes or more than 3X of the inspectedtubes are defective.

Note: In all inspections, previously degraded tubes must exhibitsignificant (greater than 10X) further wall penetrationsto be included in the above percentage calculations.

TURKEY POINT - UNITS 3 L 4 3/4 4-12


STEAM GENERATORS


4. 4. 5. 3 Ins ection Fr e uencies - The above required inservice inspections ofsteam generator tubes shall be performed at the following frequencies:

a ~ The first inservice inspection shall be performed after 6 EffectiveFull Power Months but within 24 calendar months following replacementof steam generators. Subsequent inservice inspections shall be per-formed at intervals of not less than 12 nor more than 24 calendarmonths after the previous inspection. If two consecutive inspectionsfollowing service under AVT conditions, not including the preserviceinspection, result in all inspection results falling into the C-1category or if two consecutive inspections demonstrate that previouslyobserved degradation has not continued and no additional degradationhas occurred, the inspection interval may be extended to a maximum ofonce per 40 months;

b. If the inservice inspection of a steam generator conducted in accord-ance with Table 4.4-2 requires a third sample inspection whose resultsfall in Category C-3, the inspection frequency shall be increased toat least once per 20 months. The increase in inspection frequencyshall apply until a subsequent inspection demonstrates that a thirdsample inspection is not required.

C. Additional, unscheduled inservice inspections shall be performed oneach steam generator in accordance with the first sample inspectionspecified i'n Table 4.4-2 during the shutdown subsequent to any ofthe following conditions:

1) Primary-to secondary tubes leak (not including leaks originatingfrom tube-to-tube sheet welds) in excess of the limits of

'pecification 3.4.6.2, or

2) A seisaic occurrence greater than the Operating Basis Earthquake, or

3) A loss-of-coolant accident resulting in rapid depressurizationof the primary system, or

4) A main steam line or feedwater line break resulting in rapiddepressurization of the affected steam generator.

TURKEY POINT - UNITS 3 8 4 3/4 4-13 JUN ( ~


STEAM GENERATOR


4.4. 5.4 Acce tance Criteria

As

2)

3)

4)

5)

6)

7)

8)

used in this specification:

~T" * p hcontour of a tube from that required by fabrication drawings orspecifications. Eddy-current testing indications below 20K ofthe nominal tube wall thickness, if detectable, may beconsidered as imperfections;

d d 1, g,general corrosion occurring on either inside or outside of atube;

~ddp h 1 1gl 1 1 gthan or equal to 20X of the nominal wall thickness caused bydegradation;

g~g" T h p g 1 h h 11 P 1affected or removed by degradation;

gefect means an imperfection of such severity that it exceedsthe plugging iimit. A tube containing a defect is defective.Any tube which does not permit the passage of the eddy currentinspection probe shall be deemed a defective tube;

~PT 1 11 lp 1 1 dphthe tube shall be removed from service because it may becomeunserviceable prior to the next inspection and is equal to 4(Cof the nominal tube wall thickness;

Unserviceable describes the condition of a tube if it leaks orcontains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break asspecified in Specification 4.4.5.3c., above;

Tube Ins ection means an inspection of the steaa generator tubefrom the po nt of entry (hot leg side) completely around theU-bend to the top support of the cold leg, or from the point ofentry (cold leg side) completely around the U-bend and to thebottom of the hot leg; and

0TURKEY POINT - UNITS 3 & 4 3/4 4-14 1 + de t ~ ~


STEAM GENERATOR

SURYEILLANCE RE UIREMENTS Continued

b.

9) Preservice Ins ection means an inspection of the full length ofeach tube sn each steam generator performed by eddy currenttechniques prior to service to establish a baseline conditionof the tubing.

The steam generator shall be determined OPERABLE after completingthe corresponding actions (plug all tubes exceeding the plugginglimit and all tubes containing through"wall cracks) required byTable 4.4-2.

4.4. 5. 5

a.

b.

C.

~Re orts

Withi days following the completion of each inservice inspectionof steam generator tubes, the number of tubes plugged in each steamgenerator shall be reported to the Commission in a Special Reportpursuant to Specification 6.9.2;

'he

complete results of the steam generator tube inservice inspectionshall be submitted to the Commission in a Special Report pursuant toSpecification 6.9. 2 within 12 months following the completion of theinspection. This Special Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for eachindication of an imperfection, and

3) Identification of tubes plugged.

Results of steam generator tube inspections which fall into CategoryC-3 shall be reported to the Commission pursuant to 10 CFR Part 50.72and prior to resumption of plant operation. This report shall providea description of investigations conducted to determine cause of thetube degradation and corrective measures taken to prevent recurrence.

T'URKEY POINT - UNITS 3 5 4 3/4 4-15

~ ~I ~

1

ClM

IC

Qo

ININIMUMNUNIIIER OF STEhAI GENEAATOAS TD SE

IMStECTEtl OURING IIJSERVICE INSPECTION

Ne. el Slewa GcacrHols per Ustil

Flrcl I~lee IalpcLCtea

Second 0 Sehceiwal laccttae lnritccltons

Yet

Two Theet

One Two

O IO~I

crtTchie NoIHlea:

0. The htretelee leryecllea awt be fitnired lo one cretin ycacteror oa o rolHby ccltcthere eneoettrttr&sy lM 4 ol lhe reelect~ N le Ihe tnrtnbec ol i!coen NetrctHotc hr lhe phnr) il lhe ltlellc ol lite Iiear ot litcieiatr intiiecriont intricate thar ~

H Heter QNNlHotc ere gletletlnleN in ~ lke tncne»t. l4OH I+ ltndet lotne cltetitnrrtticer, It» or<titlingconkllont in~ae oc ~ clem oeetHotc ~ be hrttnd lo he arete cctete rI»a lhote ia osl»r sicken yenceHtttc. Under tttclt circtttnNaeei lhe searle serlaeace ANhe aredtitcd Io hapccl lhe rail ccvcre condtlionL

" f.. The ochcc Ilcwa jsetHet aol hrrpccled keiey Ihe Ilail hratvice hrrpcctioa lheli be inritccietl. The lhltd entI atitMtittcnl.hrrlteclieae Aerthl hrwew I» inrintHlonc ritter%ted Ia I ebote.

TABLE 4.4-1 HININJH NlNBER OF STEAM GENERATORS TO BE INSPECTEODURING INSERVICE INSPECTION

STEhil GEtKRATOR %VS'NSPECTIIOH

lst SAMPLE INSPECTION 2nd SAMPLE INSPECTION 3rd SAMPLE INSPECTION

ltasaft hction ~iced Result Action Required

Aminie~ofST~yerSQ

C-I

C4 'ug defects eblis «ndImpact additional 25tubes In this S.G.

hsyect aN ahes In this5A yhg*factlve tubesand Inspect 2$ hgea lneach other 54,

Notification so INCswat to Par«Ipaph3tM%4Mof IO CFR $0

N/A

C-I

C-2

C-3

NlotherS MareC-I

Saae S.~C-2 but noadditionalS.Ga areC-3

AdditionalSG IsC-3

N/A

Nane

Phg*fectbe tubes andlmpect additional 25 tubestubes in this S.C.

Perform action for C 3result of first sample

Perforce action for C-2result of second sample

Inspect all tubes in eachS.G. and plug defectbetubes. Notification toNRC pursuant toPar~ape 3O.nSX2)of IOCFR%

N/A

C-I

C-2

C-3

N/h

N/A

Plug defectl~ tubes

Perform action fw C-3result of first sample

N/A

N/A

++ IlIs the +anbar of steam cenerators In the unit> and n Is the numler of steam generators ieyacted Arlng an Inspection

TABLE 4.4"2 STEAH GENERATOR TUBE INSPECTIOH


3/4. 4. 6 REACTOR COOLANT SYSTEM LEAKAGE

LEAKAGE DETECTION SYSTEMS


3.4.6. 1 'The following Reactor Coolant System Leakage Detection Systems shallbe OPERABLE:

a. The Containment Atmosphere Gaseous or Particulate RadioactivityMonitoring System, and

b. A Containment Sump Level Monitot ing System.

APPI ICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With both the Particulate and Gaseous Radioactivity Monitoring Systemsinoperable, operation may continue for up to 7 days provided:

1) A Containment Sump Level Monitoring System is OPERABLE;

2) Appropriate grab samples are obtained and analyzed at leastonce per 24 hours;

3) A Reactor Coolant System water inventory balance is performedat least once per 8 hours during steady state operation exceptwhen operating in shutdown cooling mode; and

4) Containment Purge, Exhaust and Instrument Air Bleed valves aremaintained closed.

b.

Otherwise, be in at least HOT-STANDBY within the-next 6 >ours and inCOLD SHUTDOWN within the folloqing 30~hours.

o+Q Covetous'nme3 Cvw.n LevJ YvnPori eggWith no ontainment Sump Level Monitoring System op~erable,.restoreat least System to OPERABLE status within 7 days, or be in atleast HOT STANOBY within 6 hours and in COLO SHUTOONNP(ithin thefollowing 30 hours.


, b. Containment Sump Level Monitoring System-performanceof CHANNEL CALIBRATION at least once per 48-moaths

f-eaucjinq n ~ tZ + mes3/4 4-18TURKEY POINT - UNITS 3 4 4

4.4.6. 1 The Leakage Detection Systems shall be deeonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNELOPERATIONAL TEST at the frequencies specified in Table 4.3-3,


PERATIONAL LEAKAGE

|.. 30~@ Co4rro<~H I EevnG-E. w%ea~~~r Coo lc,~c Cycle~ pv.ess~re.

<>35 +2.0 prig~ @no{


0

3. 4. 6. 2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,prlWarq

c. 1 GPM total +ea~Zo-secondary leakage through all steamgenerators and500 gallons per day through any one steam generator.

d. 10 GPM IDENTIFIED LEAKAGE fr e R tor Coolant System,~s s,pecifted 'in Table 3A-I

+g. ~4'- 4akage„ up to a maximum. of5 GPM at a Reactor Coolant System pressure of 2235 k 20 psig from anyReactor Coolant System Pressure Isolation Valve specified inTab1e 3.4-1."


ACTION:

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

With any Reactor Coolant System leakage greater than any one of theabove limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage fromReactor Coolant System Pressure Isolation Valves, reduce the leakagerate to within limits within 4 hours or be in at least HOT STANDBYwithin the next 6 hours and in COLD SHUTDOWN within the following30 hours.

C. With any Reactor Coolant System Pressure Isolation Valve leakagegreater than a'l1owed by 3.4.6..g bove operation may continue ) Fprovided:

Within 4 hours verify that at least two valves in each highpressure line having a non-functional valve are in, and remainin that mode corresponding to the isolated condition, i.e.,manual valves shall be locked in the closed position; aotoroperated valves shall be placed in the closed position and .

power supplies d gized. Follow applicable ACTIONstatement for t e >if cted system, and

Test pressures loss than 2235 psig are allowed. Minimuti differential testpressure shall not be less than 150 psid. Observed leakage shall be adjustedfor the actual test pressure up to 2235 psig assuming the leakage to bedirectly proportional to pressure differential to the one-half power.



OPERATIONAL LEAKAGE


d.

2. The 'leak om the remaining isolation valves in each highpressure line having a valve not meeting the criteria ofTable 3.4-1, as listed in Table 3.4-1, shall be determined andrecorded daily. The positions of the other valves located inthe high pressure line having the leaking valve shall berecorded daily unless they are manual valves located insidecontainment,

Otherwise be in at least HOT STANDBY within the next inCOLO SHUTOOWN within the following 30 hours.

With any Reactor Coolant System Pressure Isolation alve ea agegreater than 5 gpm, reduce leakage to below 5 gp or be in at leastHOT STANDBY within the next 6 hours and in COLD UTDOWN within thefollowing 30 hours.


4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within'achof the above limits by:

a. Monitoring the containment atmosphere gaseous or particulateradioactivity monitor at least once per 12 hours;

~ ~g(Kb. Monitoring the containment sump level at least once per 12 hours;

)dy. Performance of a Reactor Coolant System water invent lancewithin 12 hours after achieving steady-state operati *ea d at least

) gonce per 24 hours thereafter during steady-state opera son, exceptthat not more than 48 hours shall elapse between any two successiveinventory balances; and

e+. Monitoring the Reactor Head Flange Leakoff System at least once per ) E .24 hours.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Va e s cified inTable 3.4-1 shall be demonstrated OPERABLE by verifying leak gPt be within t3its limit:

a. At least once per refueling sbutdovsv no) fo CX.-- - I z.4. meiW

)E.b. Prior to entering MODE'2 whenever the plant has been in COLO

SHUTDOWN for 72 hours or sere and if leakage testing has not beenperformed in the previous 9 months, and

c. Prior to returning the valve to service following maintenancerepair or replacement work on the valve~ Nomic.h cold pote~tsaliZ ~~~oh W(oI rt lac kAiOL4- ~

The provisions of 'Specification 4.0.4 are not applicable for. entry into NODE 3or 4.

d W~/CVCT ECS average coolant temperature being changed by less than Sop/hour. ) j

TURKEY POINT - UNITS 3 dc 4 3/4 4-20

Meosurem~ a4 'ttie, D ~<oLi-~ LEAK<ce6 frccnaf gy.Qyy MO(4-~+ Pq~ ZeC ~S. Q he.~ 74 Re&ceV

~(a ~t 5~ s%~pressure is 22.'ZS ~ 2o psfq

a~ leasW Crnes par 3 l days Wa dna«is ieras

cipeaiPc'a~4<'~ 4.O.P o.re nat ~p( c&4.Jr'krq'(do Mole 3 ov 4

7:s. SA. +,C.a

5o S~b g AlALA- u i carciecds ) le~4 )e. yriagiviecueureat indi really ( as Pram Aa pe.r+ca~aau

f'~eu

tc. InJ I alLfo i s) tg 4 acsrrapliSLcl ln Ac c ar JocccC

wiK ~piaVa.cl ti&Waice ~ s pulo aCcvcl 3ld + in+~ i ~ Ll

.4 ~a,d~iig V lVe ~p'Li'~ ~'i4 g4 (. k pCYl ~Yi W

l ~seas ~ 'T t.5. ~~4..

F. lieu> i'c(ue o.c.p oki irci clue t. caAi ra~Pic

Qr fncknvcJ d.ctC ccn DY 'P( oui tKrdv)kvo-lvc:

h. ldiSin 24 ho«+ Ig Vevcfg in> yalVe

C t450W)

P ria v 4a e.wfe.riagVe.i i a) leaked r~t,

COlt 'A'$8 Mlitt WHO'D

4

I

r,

4k

'

I 'S

is

4 t s

4

iss

ii

s

'4

TABLE 3.4-1

REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES

VALVE NUMBER

Unit 3 Unit 4

FUNCTION

High-Head Safety Injection Check Valves

3-874A3-875A3-873A

3-87483"8?5B3-873B

3-8?5C3"873C

4-874A4-875A4-873A

4-8?4B4-87584-8738

4-875C4-873C

Loop A, hot legcold legcold leg

Loop B, hot legcold legcold leg

Loop C, cold legcold leg

Residual Heat Removal Line CheckValves

3-876A 4"876A4-876E

Loop A, cold leg

3-876B3"8760

4-876B4-876O

Loop B, cold leg

3"876C 4-876C3-876EMOV3-750 MOV4-750MOV3-751 MOV4-751

Loop C, cold leg

Loop A, hot leg to RHR

Loop A, hot leg to RHR

2.

3.

ACCEPTABLE LEAKAGE LIMITS

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

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm areconsidered acceptable provided that the latest measured rate has notexceeded the rate determined by the previous test by an amount thatreduces the margin betwee w asured leakage rate and the maximumpermissible rate of 5.0 gpm b reater.

pmi~uglgLeakage rates greater than 1. gpm u ess than or equal to 5.0 gpm areconsidered unacceptable if the latest measured rate exceeded the ratedetermined by the previous test by an amount that reduces the marginbetwee wmeasured leakage rate and the maximum permfsstble rate of 5.0 gpm

) Eby 50'r

p WplVMSIQ,Leakage rates grea e than 5.0 gpm are considered unacceptable.

TURKEY POINT " UNITS 3 Ec 4 3/4 4"21 tJl'V 0 ~ ""


3/4.4.7 CHEMISTRY


3.4.7 The Reactor Coolant System chemistry shall be maintained within thelimits specified in Table 3.4-2.

APPLICABILITY: At al 1 times.

ACTION:

MODES 1, 2, 3, and 4:

a.

b.

With any one or more chemistry parameter in excess of,its Steady"State Limit but within its Transient Limit, restore the parameter towithin its Steady-State Limit within 24 hours or be in at least HOTSTANDBY within the next 6 hours and in COLD SHUTDOWN within thefollowing 30 hours

With any one or more chemistry parameter in excess of its TransientLimit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWNwithin the following 30 hour s.

At All Other Times:

With the concentration of either chloride ot fluoride in the ReactorCoolant System in excess of its Steady-State Limit for more than 24 hoursor in excess of its Transient Limit, reduce the pressurizer pressure toless than or equal to 500 psig, if applicable, and perform an engineeringevaluation to determine the effects of the out-of-limit condition on thestructural integrity of the Reactor Coolant System; determine that theReactor Coolant System remains acceptable for continued operation priorto increasing the pressurizer pressure above 500 psig or prior toproceeding to MODE 4.


4.4.7 The Reactor Coolant System chemistry shall be determined to be withinthe limits by analysis of those parameters at the frequencies specified inTable 4.4-3.

TURKEY POINT - UNITS 3 8c 4 3/4 4-22~ ~'j

TABLE 3.4-2


CHEMISTRY LIMITS

PARAMETER

Dissolved Oxygen"

Chl ori de

Fluoride ~~

STEADY-STATELIMIT

< 0.10 ppm

< 0.15 ppm

< 0.15 ppm

TRANSIENTLIMIT

< 1.00 ppm

< 1.50 ppm

< 1.50 ppm

V4

gpa.ua. < )ak4-.

Limit not applicable with average reactor coolant temperature less than orequal to 250 F.

TURKEY POINT - UNITS 3 4 4 3/4 4-23 v('. i~ 1:

TABLE 4.4-3


CHEMISTRY LIMITS SURVEILLANCE RE UIREMENTS

PARAMETER

Oissolved Oxygen*

Chloride

Fluoride

SAMPLE ANDANALYSIS FRE UENCY

At least 5 times per week not toexceed 72 hours between samples



gglc IU ~ 0 c5 QGKcL

clrc4l Plagal g Unada.l l 4 ~4

Not required with average reactor coolant temperature less than or equal to2500F



3/4. 4. 8 SPECIFIC ACTIVITY


3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT I-131,and

b. Less than or equal to 100/E microCuries per gram of gross radioactivity.

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

ACTION:

HODES 1, 2 and 3":

a. With the specific activity of the reactor coolant greater than1 microCurie per gram DOSE EQUIVALENT I-131 for more than 48 hoursduring one continuous time interval or exceeding the limit line shownon Figure 3.4-1, be in at least HOT STANDBY with average reactorcoolant temperature less than 500 F within 6 hours; and

b. With the specific activity of the reactor coolant greater than100/E microCuries per gram, be in at least HOT STANDBY with averagereactor coolant temperature less than 500 F within 6 hours.

h~J. inMODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than1 microCurie per gram DOSE E(UIVALENT I-131 or greater than 100/E micro-Curies per gram, perform the sampling and analysis requirements of Item 6.a)of Table 4.4-4 until the specific activity of the reactor coolant isrestored to within its limits.


~,

4.4.8 The specific activity of the reactor coolant shall be determined to bewithin the limits by performance of the sampling and analysis program ofTable 4.4-4.

Jlty.Jim)3/4 4-25

"With the average reactor coolant temperature greater than or equal to 5004F.


P ~

~ ~

~ ~ ~ ~

~ ~ 0 oo

~ ~ ~

~ ~

~ ) ~ o + +~ ~ o. ~

Oo

t'~t ~

mo

~o ~

~ ~

0 ~ ~

~ . I ~ ~

~ ~ o ~

o p ~ ~

~ ~ ~ '~ ~ ~

~ % ~

~ ~ ~

~ o I f.o o4 ~

~ f~ ~ o ~

4 ~ ~

o o

;jul~ oo

I

P}o foo< ~ Ip ~o.oo f ) ~

~ omfqr

f

'Q+1

r~~o

200~ ~

~ ~

~ P\ ~ ~

6 ~r ~

150

~ ~ ~

~ ~ ~

~ ~ ~ ~

~ ~ ~

o ~ ~ ~ ~

'

~ ~

~ ~

P ~ I

~ ~

~ ot ~

~ ~ ~ ~

~ ~ ~

f f ~ ~

'~~ ~ ~

~ 4 ~ P

~ ~

) f f lUNACCEPTABLEt I

f P TlOMI,.f.i .I.fo),o~ ) ~ o }t ~ Qo O

~ ~

f af~.$ »~~e ~ ~I ~ f

~ or f + ~ ~ ~ ~ oo ~~ ~ ~ ~

r «+»P fao f-«I

~oopsI ~~ 4~~ y j't'op

P f ~

$ foo

~ ~

~ ~ ~

~ ~

~ ~

~ ~r ~ p~ ~

~ ~ ~

~ ~ P'I ~

~ ~

~ ~ ~f oooo«.f o

~ f ~

~o oo ~ ~

~ ~ ~

~ f ~ ~ ~ ~

OQP~ o ~

~ o l.~ I

~ ~ ~

fef ~

~ ~ o ~ ~

P ~

~ ~ ~

~ ~ ~

o ~ ~II

P ~

I ~

)Wol P

~ i o ~

~ ~ ~

~ M ~~ ~ P ~~ ~ ~ ~

~ ~

~ ~ ~ ~

~ ~ P~ P ~ ~

~ ~ % ~

LPP

~ 4 ~ ePo ~ ~

~ ~

o p ~

~ ~ ~

~ o ~

~ ~

ACCEPTABLE�

'PERATlON

~ ~

~ P

~ ~ ~ ~ ~

~ ~ ~

~ ~

~ ~ ~ ~

P ~ ~

0 oopI.

of ~

f o ~

~ ~ ~

~ ~ ~

~ ~

~ ~ ~

~ ~ ~

~ ~

o ~

~ ~

~ ~

~ ~ ~

~ ~ ~

~ P ~ ~

~ ~ ~ ~

~ ~~ ~

o ~

~ ~ ~ ~

~ o~ P ~ ~

~ ~ ~

~ ~ ~

~ ~

oo a so a e a so coo'ECKMTIfGATED %ERAL tONER

FIGURE 3.4-1

00SE E/UIVALENT I-.131 REACTOR COOLANT SPECIFIC ACTIVITYLIMITVERSUSPERCENT OF RATED THERMAL RNER %TH THE REACTOR COOLANT SPECIFICACTIVITY>1 pC)/gras DOSE E/UIVALENT I-131

TURKEY POINT - UNITS 3 4 i 3/4 4-26J.', (g

.7Cm

TABLE 4.4-4REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE

AND ANALYSIS 'PROGRAM

C/l

TYPE OF MEASUREMENT

AND ANALYSIS

1. Gross RadioactivityDetermination

2. Tritium Activity Determination

3. Isotopic Analysis for OOSE EQUIVA-LENT I-131 Concentration

4. Radiochemical IsotopicDetermination Including GaseousActivity

5. Radiochemical for E Determination

6. Isotopic Analysis for IodineIncluding I-131, I-133, and I-135

SAMPLE AND ANALYSISFRE UENCY


1 per 7 days.

1 per 14 days.

Monthly

1 per 6 months"

a) Once per 4 hours,whenever the specificactivity exceeds 1pCi/gram 00SEEQUIVALENT I-131or 100/E pCi/gram ofgross radioactivity, and

b) One sample between 2and 6 hours followinga THERMAL POWER changeexceeding 15Kof the RATED THERMALPOWER within a 1-hourperiod.

MODES IN WHICH SAMPLEAND ANALYSIS RE UIRED

1, 2, 3, 4

1, 2, 3, 4

1, 2, 3, 4

1¹, 2¹, 3¹, 4¹, 5¹

1, 2, 3

~ Qt

~ ~


TABLE NOTATIONS

* Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATIONhave elapsed since reactor was last subcritical for 48 hours or longer.

8 U'ntil the specific activity of the Reactor Coolant System is restoredwithin its limits.

0

'. TURKEY POINT - UNITS 3 5 4 3/4 4-28l. ~ ~


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 Figures3.4-2 and 3.4-3 for Unit 3 and Figures 3.4-4 and 3.4-5 for Unit 4 during heatup,cooldown, criticality, and inservice leak and hydrostatic testing with:

a, A maximum heatup of 100 F in any 1-hour period,

b. A maximum cooldown of 1004F in any 1-hour period, and

c. A maximum temperature change of less than or equal to 54F in any1-hour period during inservice hydrostatic and leak testing operationsabove the heatup and cooldown limit curves.

'I

APPLICABILITY: At al l times.

ACTION:> Asz4~

any of the above limits exceeded,,restore the temperature and/or pressuto wi 'he limit within 30 minutes;Iperform an engineering e atfon todetermine t cts of the out-of-limit condition o ructural integritof the Reactor Coolan determine t eactor Coolant System remainacceptable for continued e in at least HOT STANDBY within thenext 6 hours ce the,RCS-T and

avg e to less than 2004F and5 , respectively, within the following 30 hours.


4.4.9.1. 1 The Reactor Coolant System temperature and pressure shall bedetermined to be within the limits at least once per 30 minutes during systemheatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimensshall be removed and examined, to determine changes in material properties,as required by 10 CFR Part 50, Appendix H, in accordance with the schedulein Table 4.4-5. The results of these examinations shall be used to updateFigures 3.4-2 to 3.4-5.


1

eaI ~

l g,%sr

f0N

g CO

I4N8

NTKRl fNPKNY SA$1$

%N %fM. Ol 0.31%lNNN. Nfl. 8 F

N l0 KFSCtltC M.L, Nea TENSNF N 1/l TNINNKSS a m F

abet At 8/4 TNCIKS$ a 1Q P

%hTlF lATE.0 TO 100'/Nt

LKAK TKST Llll1

RlTlCALlTYLllllT

1N 150 200 25Q XO

T~ flSMO150 5 550

FIGURE 3.4-2TURKEY POINT UNIT 3

REACTOR COOLANT SYSTEH HEATUP LIHITAT!ONS - APPLICABLE UP TO 10 EFPY

ld it L any of thc 4 Ve ( inni4 cyceec(ec(

.'estoret'e iona pcratu re W /ar pressure

te iuigin ttic (iini t Lui"Lin 3o iviinutes cvJ)

Ii-u" the ..a.. 'p(etc a.n erig iiiecring ella.(uatl sN go

de4e ',.~ the elf~ c$ .5e ~f-.f- li'~hLaVLe(kevin

m 5e St»ctur ( inVegv-ttg

g.C5 cu c( defe on Ine, tko.t t(ie

RCS y'erne in'' mes'~t'a.44,

IQgiri) I nl/c.d epev"a I I Irrl

)oy-

Lo ~ Ba in HoT'7R4bC'I W redueiL 4auera.cp, coo(cu+ CcMpei atul c.

(iressvrc 'ta (e.ss +a.n. ~~ F

f4 g(la~in~

au r5 iunc( urn

p�(etc th c. ~ i na~ < ng

I)eva.(ua.hm in Z.a- 'coasts pr'

ri~ «ge i- Zoo'F ai Qoo tos i y.

Tp I8 pp 4w I~~t nQALQ

8'LAA

C

CA

4l

ICab

~ ~ ~ '

ERI PROPERTY SASISSNNOR0Cfee

1'N

p ss

gI NTKS

e L

SN

I p ioo

RN %TR, CO i 0.31$NlfQLNag + 3 F

gf 10 KHKCflTKSllL NKR %NSat~ at lie mrCenS ~ SCi C

keg AT Sf'NNIESS 15 F

flCNQPt5 gverage l~e~vi~ l'0 .

FIGURE 3.4-3. TURKEY POINT UNIT 3

REACTOR COOLANT SYSTEH COOLDOMN LIHITATIONS - APPLICASLE UP TO 10 EFPY

S8NI

isofiN

I metea

IIII. aa

I~

TEkl PROPGNY SA$1$

%$ %N. CO ~ 0.ÃSQlTN. NTgg ~ 0 F

fiEFFKCOS ML KRR %NSNgg AT fl4 ONCllESS ~ 3C fNFL AT 3I4 TNNIESS ~ 030 F

LEAK TEST LllllT

%AM RATElt TO fOO F/l

ISO m m ae 350 4

7~> puree IarCee meeeeE (%

FIGURE 3-4 4TURKEY POINT UHIT 4

REACTOR COOLAHT SYSTEH HEATUP LIHITATIOHS " APPLICABLE UP TO 10 EFPY

0

t0Ri%8NNOfae

IQTERIAL PROPERTY SASIS

%$ CITAL CQ ~ 0.38lalalL arar ~ O'

N 10 EFFKCTBK FOR KNN mN5e~ Ar 1/4maaaSS i Se'%Fag Af 0/i lHICNKSS ~ t$0

tea.l010N

OLNNN NATES

~y F/N "

eo LNO

l 150 2N tO 3N

es u..).i.-q, J . ~- mens eeeti50 500 $%

FIGURE 3.I-STURKEY POINT UNIT I

REACTOR COOLANT SYSTEN COOUNNN LIIIITATIONS- APPLICABLE UP TO 10 EFPY

TA8LE 4.4-5

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE

CAPSULENUMBER

U

CAPSULENUMBER

VESSELLOCATION

so4o

ICcZ30

VESSELLOCATION

300Nc

gG

l5~'c'NIT

3

LEADFACTOR

o,3'4o Z4e Qo.M

UNIT 4

LEADFACTOR

p, A.q:oi 1 f

g+

o Po 34

WITHDRAWAL TIME EFPY

Stan

33 yeal'sStandbyStandby

WITHDRAWAL TIME EFPY

Standby24 yeatsStandbyStandbyStandbyStandby



1 D 6]&7idb)OVERPRESSURE SYSTEMS

3.4.9.3 The high pressure safety injection capability to the Reactor CoolantSystem (RCS) shall be isolated, and below an RCS average coolant temperatureof 275'F at least one of the fol'lowing Overpressur 'ystems shallbe OPERABLE:

a) Two power-operated relief valves (PORVs) with a'lift setting of 415a 15 psig, or

b) The RCS depressurized with a RCS vent of greater than or equal to2.20 square inches.

APPLICABILITY: NODE 3", 4, 5 and 6 with the reactor vessel heed err no~ re~russ j ~

ACTION:

g, With the high pressure safety injection capability to the RCS unisolated, takeimmediate action to isolate this capability.

, g, In HOOE 4 with RCS average coolant temperature less than or equal to 275'F, andin MOOE 5 or in MODE 6 with the reactor vessel 4oac&~no+ r~z~~J, ~

)))f. With one PORV inoperable, perform at least one of the followingwithin the next 7 days:

Restore the inoperable PORV to OPERABLE status, or

Depressurize and vent the RCS through at least a 2.20 squareinch, or

(:.jf. Depressurite and maintain a RCS vent through at least one openPORV and open associated block valve.

W ~

~

4

~~!9» p ~ol vroklnl 8'o( r!OJ Qofg)f I' c'Tlo(,

co(octal

~If/ lrf I '(d ~ '~':u.f~ ~>ol~Y~(oR~r rcgu'co)A"-5 c,o3 l..rl "i" ddt ,yr +C) ~ BED r.<4 IO i«

TURKEY POINT - UNITS 3 4 4 3/4~4-36 .J s)1s I

po f cmckiwm ~~Qof-s


PRESSURIZER


3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 1004F in any 1-hour'eriod,

b. A maximum cooldown of 200'F in any 1-hour period, andgZo .P

c. A maximum spray water temperature differential of ~-'..APPLICABILITY: At all times.

ACTION:

res rszer temperature limits in excess of any of t arestore t e e to within the limits w nutes; perform anengineering evaluation to ects of the out-of-limit conditioon the structural in ' the pre~ri-zap~ determine that the pressurizeremains acce for continued operation or be fr("at-.least HOT STANDBY with nthe nex ours and reduce the pressurizer pressure to less Chan-50Q gsig

the following 30 hours.

SURVEILLANCE RE UIRB1ENTS

R~cA~ l %5 ~~~

4.4.9.2 The pressurizer temperatures shall be determined to be within thelimits at least once per 30 minutes during system heatup or cooldown. Thespray water temperature differential shal,l be determined to be within thelimit at least once per 12 hour s during auxiliary spray operation.

'TURKEY POINT - UNITS 3 8I 4 3/4 4-35

R i 4 ~a. pr&ssU r it~ fc.w pa.va-t<~ ( 1~TS

&XcesS' ore l nni S

I . RCSiO~ f4 pP~+Vrf+At fO ~(Q<< ~ (~~>ff

lu~4i ~ gc nninOges>

z. 4i6'n ho~.s ~(ft,~:tII*

a. ~ ~pl@:fw cw, aeg:t.~c i>g e~al<~>'~ fe

WL1AC C C o C ~ o (M i

~ J Ci~ n the sf~ ~Pung ('~+~ r, g

QC. p~~ Vly~ ~ ~QCVllhl hC. ga.f f4.pre.ssv v i y~

mrna.inc

ca.cc epona.k 4.I~( i su~ ape~~t <m

~

Bc u HnT SWAN 8'/ cu-~ ~<4 ca T4.pvessuv,aav p~ssv~ ge less (ha~ ~opsig

Q llO~ing j( gee r(~ ega u

~)~vs ~v< n~ t ex~eAi~)p5 lp ~

+gas pp gC t pfPi'.~i~&A~'f


LIMITING CONOITION.FOR OPERATION Continued

Q)p. 'ith both PORYs inoperable, depressor ize and vent the RCS through atleast a 2.20 s

' t within 24 hours.Ur p)onv

In the even eiNer ~ RVs or a 2.20 square inch vent is used tomitigate an RCS pressure transient, a Special Report shall be preparedand submitted to the Commission pursuant to Specification 6.9.2 within30 days. The report shall describe the circumstances initiating thetransient, the effect of the PORVs or RCS vent(s) on the transient,and any corrective action necessary to prevent recurrence.


4.4.9.3. 1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuationchannel, but excluding valve 'operation, within 31 days prior to enteringa condition in which the PORV is required OPERABLE and at least onceper 31 days thereafter when the PORV is required OPERABLE.

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channelat least once each refueling; and .

ncg o «x~ 24- Alc. Verifying the PORV block valve is eas once per 72 hours

when the PORV is being used for overpressure protection.

d. While the PORVs are required to be OPERABLE, the backup air supplyshall be verified OPERABLE at least once per 24 hours.

4;4.9.3.2 The:2 '0 square inch vent shall be verified to be open at least onceper 12 hours" when the vent(s) is being used for overpressure protection.

4.4. 9. 3. 3 Verify the high pressure infection capability to the RCS is isolatedat least once per 24 hours.

Except when the vent pathway is provided with a valve which is locked, sealed,or "otherwise secured i.n the open position, then verify these valves open atleast once per 31 days.



3/4.4. 10 STRUCTURAL INTEGRITY


3.4. 10 The structural integrity of ASHE Code Class 1, 2, and 3 components shallbe maintained in accordance with Specification 4.4. 10.

APPLICABILITY: Al 1 MODES.

ACTION:

a ~

b.

C.

With the structural integrity of any ASME Code Class 1 component(s)not conforming to the above requirements, restore the structuralintegrity of the affected component(s) to within its limit or isolatethe affected component(s) prior to increasing the Reactor CoolantSystem temperature more than 50 F above the minimum temperature .

required by NOT considerations.

With the structural integrity of any ASME Code Class 2 component(s)not conforming to the above requirements, restore the structuralintegrity of the affected component(s) to within its limit or isolatethe affected component(s) prior..to increasing the Reactor CoolantSystem temperature above 2004F.

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


4.4.10 Requirements of Specification 4.0.5 shall be met including the followingspecific requirements.

at Reactor Coolant System integrity shall be demonstrated as followsafter the system is closed following normal opening, modification orrepair.

1) When the Reactor Coolant System is closed, the system will beleak tested at not less than 2335 psig while meeting NDTTrequirements for temperature.

2) When Reactor Coolant System modifications or repairs have beenmade which involved new strength welds on components greater )Q Zthan 4-fn. diameter, the new welds will receive both a surfaceand 100K volumetric examination.

0 lh

. TURKEY POINT - UNITS 3 5 4 3/4 4-38

REACTOR COOLANT SYSTEH


3) Mhen Reactor Coolant System modificati'ons or repairs have beenmade which involve new strength welds on components ~ diameter ~g

)@gal gk~ 4

TURKEY POINT - UNITS 3 4 4 3(4 4-39


3/4.4. 11 REACTOR 'COOLANT SYSTEM VENTS


3.4. 11 At least one Reactor Coolant System vent path consisting of at leasttwo vent valves in series and powered from emergency busses shall be OPERABLEand closed at each of the following locations:

a, Reactor vessel head, and

b. Pressurizer steam space


ACTION:

a ~

b.

With one of the above Reactor Coolant System vent paths inoperable,STARTUP and/or POWER OPERATION may continue provided the inoperablevent path is maintained closed with power removed from the valveactuator of all the vent valves in the inoperable vent path; restorethe inoperable vent path to OPERABLE status within 30 days, or, be inHOT STANDBY within 6 hours and in COLD SHUTDOWN within the following30 hours.

With both Reactor Coolant System vent paths inoperable; maintain theinoperable vent path closed with power removed from the valve actuatorsof all the vent valves in the inoperable vent paths, and restore atleast one of the vent paths to OPERABLE status within 72 hours or bein HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following30 hours.


4.4.11 Each React t S stem vent path shall be demonstrated OPERABLEat least once pe %8-a~Ra b :

efveli~, no< fo t'-Aaccl 4 M 44.

a. Verifyi m ation va ves n each vent path are locked inthe open posit on,

b. Cycling each vent valve through at least, one complete cycle of fulltravel from the control room, and

c. Verifying flow through the Reactor Coolant System vent paths duringventing.

TURKEY POINT - UNITS 3 SI 4 3/4 4-40

JI ~0

A ~ ~ ~

EMERGENCY CORE COOLING SYSTEMS

3/4. 5. 1 ACCUMULATORS


3.5;1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:II

a. The isolation valve open and its circuit breaker open,

b, A contained borated water volume of between 6545 and 6665 gallons,

c. A boron concentration of between 1950 and 2350 ppm,

d. A nitrogen cover-pressure of between 600 and 675 psig, and

e, A water level andpressure channel OPERABLE.

APPLICABILITY: MODES 1, 2, and 3".

ACTION:

a ~

b.

With one accumulator inoperable, except as a result of a closedisolation valve, restore the inoperable accumulator to OPERABLEstatus within 4 hours or be in at least HOT STANDBY within the next6 hours and reduce pressurizer pressure to less than 1000 psigwithin the following 6 hours.

With one accumulator inoperable due to the isolation valve beingclosed, either immediately open the isolation valve or be in atleast HOT STANDBY within 6 hours and reduce pressurizer pressure toless than 1000 psig within the following 6 hours.


4.5. 1. 1 Each accumulator shall be demonstrated OPERABLE:

a ~ At least once per 12 hours by:

1) Verifying the contained borated water volume and nitrogencover-pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open bycontrol room indication (power may be restored to the valveoperator to perform this survei)lance if redundant indicator isinoperable).

"Pressurizer pressure above 1000 psig.




b. At least once per 31 days and within 6 hours after each solutionvolume increase of greater than or equal to IX of tank volume byverifying the boron concentration of the solution in thewater-filled accumulator;

c. At least once per 31 days:

d.

er to the isolation valve operato~ ispen breaker when the RCS pressure is

1) By verifying thadisconnected byabove 1000 psig, n

2) Each accumulator water level and pressure channel shall bedemonstrated OPERABLE by the performance of an ANALOG CHANNELOPERATIONAL TEST, and

At least ante per ref eul nlg~rtof"4 elk~ Z+ rnuk~k

1) Each accumulator water level and pressure channel shall bedemonstrated OPERABLE by the performance of a CHANNELCALIBRATION, and

,kka 1 *k k 1 kkkk k kfor operability..

TURKEY POINT - UNITS 3 4 4 3/4 5-2s ~ ~ i s ~ ~


Q5 Puc'no~ c+~e4 Ten pere,'furl3/4.5.2 ECCS SUBSYSTEMS -~ GREATER THAN OR E(UAL TO 35O~F


3.5.2 The following Emergency Core Cooling System (ECCS) equipment and flowpaths shall be OPERABLE:

a. Four OPELABLE Safety Injection (SI) pumps with discharge aligned tothe RCS cold legs,

b. Two OPERABLE RHR heat exchangers,

c. Two OPERABLE RHR pumps with discharge aligned to the RCS cold legs,

d. An OPERABLE flow path capable of taking suction from the refuelingwater storage tank as defined in Specification 3.5.4, and

e. Two OPERABLE flow paths capable of taking suction from thecontainment sump.

APPLICABILITY: MODES 1, 2, and 3".

ACTION:

a.

/gcf

With any one of the required ECCS compo ntso flow pat iyoperable,except for inoperable Safety Injection , restorb' einoperable component or flow path to OPERABLE status within 72 hoursor be in at least HOT STANDBY within the next 6 hours and in HOTSHUTDOWN within the following 6 hours.

In the event the ECCS is actuated and injects wa orCoo stem, a Special Report sh epared and submitted tothe Commiss cation 6.9.2 within 90 daysdescri c rcums e actuation and the totala ulated actuation cycles to date.

With one Safety Injection pump inoperable, restore the pump toOPERABLE status within 30 days or be in at least HOT STANDBY withinthe nex hours and in HOT pHUTOOHN within the folloying g hours.

/Cll4I upp/Ieh +p CIn> 5 innulianepu=~~/.i h two afety Xrl)ection Pumps inoperahle, restore ohe of tha two

inoperable pumps to OPERABLE status within.72 hours or be in at'eastHOT STANDBY within the next 6 hours and. in HOT SHUTDOWN within

the following'ours.gas rfC77pAn'rnplirs /c hc/< union'Irnultuncc~'<lu l+0/pL

IDurin hcofup hi g I"<«J Qnjalg in''ccfio~ ce mal /Py Q )"-'"'': 't-/

u

~ t

only 8 IcsforCgn r rene','c OS@'p 5$ 8ri rccP~cvt '9 5:P,Q. i4 CLED/'g~i; .e,TURKE( POINT - UNITS 3 8t 4 C 3/4 5-3

nit'n gC5 CO/„:I /"~.CCf/ "«~uf> + ~"S'O'F.



4.5:2 Each ECCS component and flow path shall be demonstrated OPERABLE:0

a. At least once per 12 hours by verifying by control room indicationthat the following valves are in the indicated positions withpower to the v.lve operators removed:

Valve Number Valve Function Valve Position

864A and B

862A and B

863A and B

866A and BHCV-758"

Supply from RWST to ECCSRWST Supply to RHR pumpsRHR RecirculationH. H. S. I. to Hot LegsRHR HX Outlet

OpenOpenClosed

+pen- C, l ~sec(Open

To permit temporary operation of these valves for surveillance ormaintenance purposes, power may be restored to these valves for aperiod not to exceed 24 hours.

b. At least once per 31 days by:

1) Verifying that the ECCS piping is full of water by venting theECCS pump casings and accessible discharge piping,

2) Verifying that each valve (manual, power-operated, or automatic)in the flow path that is not locked, sealed, or otherwisesecured in position, is in its correct position, and

3) Verifying that each RHR Pump develops the'ndicated differentialpressure applicable to the operating conditions when testedpursuant to Specification 4.0.5:

RHR Pump > 131 psid at a metered flowrate > 150 gpm(recirculation. mode), or

> 231 ft at metered flowrate > 3600 gpm(normal cooldown mode).

c. At least once per 92 days by:

1) Verifying that each SI puap develops the indicated differentialpressure applicable to the operating con'ditions when testedpursuant .to Specification 4.0.5:

SI pump > 1126 psid at a metered flowrate > 300 gps (normalalignment and Unit 4 SI pumps aligned to Unit 3 RWST), or

> 1156 psid at a metered flowrate > 280 gpmYUnit 3 SI pumps aligned to Unit 4 RWST).

Air Supply to HCV-758 shall be verified shut off once per 31 days.TURKEY POINT - UNITS 3 5 4 3/4 5-4



d. By a visual inspection which verifies that no loose debris (rags,trash, clothing, etc.) is present in the containment which could betransported to the containment sump and cause restriction of thepump suctions during LOCA conditions. This visual inspection shallbe performed:

For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and

e.

2)

At

2)

Of the areas affected within containment at the completion ofeach containment entry when CONTAINMENT INTEGRITY isestablished.-

ra4 (in(, n t4 eycW 24 ~~+~least once per ~walks 5y:

Verifying automatic isolation and interlock action of the RHRsystem from the Reactor Coolant System by ensuring that with asimulated or actual Reactor Coolant System pressure signalgreater than or equal to 525 psig the interlocks cause thevalves to automatically close and prevent the valves from beingo d, and

V~ g correct interlock action to ensure that the EST isisolated from the RHR System during RHR System operation and toensure that the RHR System cannot be pressurized from theReactor Coolant System unless the above EST Isolation Valvesare closed.

3) A visual inspection of the containment sump and verifying thatthe-sob.~e suction inlets are not restricted by debris andthat the ~component (trash racks, screens, etc.} show noevidence of structural osion.

a PAe In e7 o g~e ~,~~,At least once per

~(ig~ ~o+ fo e)zap~ 24 >m~1) Verifying that e automhic valve in the flow path actuates

to its correct position on Safety Injection actuation testsignal, and

2} Verifying that each of the follo~ing pumps start automaticallyupon receipt of a Safety Injection actuation test signal:

a) Safety Injection pump, and

b) RHR pump.

TURKEY POINT - UNITS 3 8( 4 3/4 5-59 )@


SURVEILLANCE RE UI'REMENTS

g. By verifying the correct position of each electrical and/ormechanical position stop for the following ECCS throttle valves:

1) Within 4 hours following completion of each valve strokingoperation or maintenance on the valve when the ECCS~ required to be OPERABLE, and g$ U f

PtNetm

2} At least once per 1& months.

S stemVal ve umber

-"-758MOV-*-872

Fcv-*-cp05

TURKEY POINT - UNITS 3 5 4 3/4 5-6rit )jttj.


Qc5 Avrcra5lc Cool a T ) dvugo< roti) fro3/4. 5. 3 ECCS SUBSYSTEMS -~ LESS TEAN 350 F


3.5.3 .As a minimum, the following ECCS components and flow path shall beOPERABLE:

a. One OPERABLE RHR heat exchanger,

b. One OPERABLE RHR pump, and(i)

c. An OPERABLE flow path capable of<taking suction from therefueling water storage tank upon being manually realigned and(Q )transferring suction to the containment sump during therecirculation phase of operation.

APPLICABI LITY: MODE 4.

ACTION:

b.

With no OPERABLE ECCS flow path from'-the refueling water storagetank, restore at least one ECCS flow path to OPERABLE status within '

hour or be in COLD SHUTDOWN within the next 20 hours.

With either the residual heat removal heat exchanger or RHR pumpinoperable, restore the components to OPERABLE status or maintainthe Reactor Coolant System ~ less than 3504F by use of alternateheat removal methods.

the ECC'S ~s ac ~ated"'anal rn eactoCoolant System, p e prepared and submitted tothe Commission purs pecific s ithin 90 daysdescribin rcumstances of the actuation andac ed actuation cycles to date.


4.5.3. The ECCS components shall be demonstrated OPERABLE per the applicablerequirements of Specification 4.5.2.

, TURKEY POINT - UNITS 3 8 4 3/4 5-7r

Lit ~ tr 5vs

fP1EZCEJJCV CjdC'odLiAI6 gPg7g~g

3T4-.5.4 "REFUELING. WATER STORAGE TAN%~


3.5.4„For single Unit operation, one refueling water storge tank (RWST) shallbe OPERABLE or for dual Unit operation two RWSTs shall be OPERABLE with:

I.Jic~i'~la. A minimum ~~ borated water volume of 320,000 gallons per RWST,

b. A minimum boron concentration of 1950 ppm of boron,

c. A minimum solution temperature of 394F, and

d. A maximum solution temperature of 1004F.


ACTION:

With less than the required number of RWST(s) OPERABLE, restore the tank(s) toOPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours andin COLD SHUTDOWN within the following 30 hours.

'


4.5'.4 The required RWST(s) shall be demonstrated OPERABLE:

a. At least once per 7 days by: I

1) Verifying the ~%eeet borated water volume in the tank, and

2) Verifying the boron concentration of the water.

b. By verifying the RWST temperature is within limits whenever theoutside air temperature is less than 394F or greater than 1004F atthe following frequencies:

1) Within one hour upon the outside temperature exceeding itslimit for consecutive 23 hours, and

2) At least once per 24 hours while the.outside temperatureexceeds its limit.

* TURKEY POINT - UNITS 3 4 4 3/4 5-8 >U4 4; 1g"Cia

3/4 6 coNTAINMENT sYsTEMs ++ ~Yacc- CIoo TYICLQ 44.+a )eh, vs\~3/4.6.1 PRIMARY CONTAINMEN Pa er ~~ << QAkrolc fV Cpl?PLLOMof'eW~in IA tVe~ ~ ~'irloc j:.~ ~CONTAINMENT INTEGRITY

yf es'Cin re O i rarrI&S .LIMITING CONDITION FOR OPERATION

3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintaine .


ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within1 hour or be in at least HOT STANDBY within the next 6 hours and in COLDSHUTDOWN within the following 30 hours.

) 4g


4.6.1. 1 CONTAINMENT INTEGRITY shall be demonstrated:

At least once per 31 days by verifying that all penetrations" notcapable of being closed by OPERABLE containment automatic isolationvalves and required to be closed during accident conditions areclosed by valves, blind flanges, or deactivated automatic valvessecured in their positi xcept as provided in Table 3.6-1 ofSpecIfLcatIon 3.6.4.1;

After each c1osfng of each penetratfon sob)act to Type 6 testin~except the containment air lock), if opened following a Type A or B /est, by leak rate testing the Seal with gas at a pressure not less

than 50 psig, and verifying that when the measure'd leakage ratefor these seals is added to the leakage rates determined pursuant toSpecification 4.6.1.2d. for all other Type B and C penetrations,the combined leakage rate is less than 0.60 L .

*Except valves, blind flanges, and deactivated automatic valves which are .

located inside the containment and are locked, sealed or otherwise securedin the closed position. These penetrations shall be verified closed duringeach COLD SHUTDOWN except that such verification need not be performed moreoften than once per 92 days.


CONTAINMENT SYSTEMS

CONTAINMENT LEAKAGE


3.6. 1. 2 Containment leakage rates shall be limited to:

a ~ An overall integrated leakage rate

i1) Less than or equal to L , 0.25K by weight of the containment

a'ir

per 24 hours at~ 50 psig, or~ de <~ww~

+e2 ~h

b. A combined leakage rate of less than 0.60 L for. all penetrationsaand valves subject to Type B and C tests, when pressurized to50 psig.


ACTION:

A,. With either the measured overall integrated containment leakage rate exceeding0.75 L or the measured combined leakage rate forall penetrations and valves subject to Types B and C tests exceeding 0.60 L ,a'

21 kd 1 2 2121a

, and the combined leakage rate for all penetrationssubject to Type B and C tests to less than 0.60 L prior to increasing theReactor Coolant System temperature above 200 F.

jihudert a4alej Fkoqd7SURVEILLANCE RE UIREMEAS

.2.1.2.11 1 1 k 11 2t determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSIN45; 4-1972:

Three Type A tests (Overall Integrated Containaent Leakage Rate)shall be conducted at 40 k 10 month intervals durin shutdown aa pressure not less than~~ 50 psig,during each 10-year service period. The third test of each setshall be conducted during the shutdown for the 10-year plantinservice inspection;

- TURKEY POINT - UNITS 3 4 4 3/4 6-2JL'4 ng tg<~

c.ow o i "ir..r-) I eccl'a.q ~ I"a.f' GfG

)I Ga4I 4a~ Qr" eQUO.C gC) Q Co.O Qcr )Or

a.j I pe~ic3 i cd.o>~~ a;.) uglier ~u'c>I ace tc fuge(r t,/

4 l

I3 .'>;c' ~, +~ LIAerr p'' I '".":g, '0 g-:.'g

I /1:.. -= lc c~e../ale ~c -'- ='.': ". d'.CC lc.

g~'ifnr~ ) />c'"~ or 6g in HOT 57/1)P/3$/

/nJ7 C~i~ a ~i+ 4Y,Fr 8 C ~le

CONTAINMENT SYSTEMS

SURVEILLANCE RE UIREHENTS Continued

If any periodic Type A test fails to meet either 0.7 L .75 L,athe tes for subsequent Type crt% shall be reviewed andapproved by the Commss o consecutive Type A tests fail tomeet either 0.75 L . 5 Lt, a test shall be performed atleast ev months until two consecutive Typ s meet either0 or 0.75 Lt at which time the above test schedul.e m ~~umed

The accuracy of each Type A test shall be verified by a supplemental.test which:

1) nfirms the accuracy of the test by verifying that the supple-men 1 test result, L , is in accordance witP the appropriatefollow equation:

c am+ ) 3 < 0. 25 La or c Ltm +

o ~ — '

where L or Lt is th me red Type A test leakage and Ltm ois the superimposed lea;

2) Has a duration ficient to es ablish accurately the change inleakage rat etween the Type A tesetnd the supp lame'ntal test;and

3) Requ's that the rate at which gas is injecte nto the conta n-me or bled from the containment during the supp ntal tes

between 0.75 L and 1.25 L or'.7 and 1.25 L .

Type B nd C tests shalq/~./2.Qthan, 50 psig, d i g e'achno ca, at inter s hereatinvolving:.

ducted with gas at a pressure not lessactor shutdown for refueling, but inthan 24 months except for tests

)C

a g) Air locks,

L 'R) Purge supply and exhaust isolation valves, and

C 8) Equipment access opening which shall be tested at least onceevery 12 months and after each use.

Air locks shall be tested and demonstrated OPERABLE by the require- )gments of Specification 4.6;1.3;

4.(.12.5Purge supply and exhaust isolation valves seals shall be tested and. )>demonstrated OPERABLE by the requirements of Specification 4.6.1.7.2,as applicable;

qc.S.Z.SThe provisions of Specification 4.0.2 are not applicable.

TURKEY POINT - UNITS 3 8( 4 3/4 6-3 "": "-9i&aF

CONTAINMENT SYSTEMS

CONTAINMENT AIR LOCKS

LIMITING CONDITION. FOR OPERATION

3.6. 1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normaltransit entry and exit through the containment, or during theperformance of containment air lock surveillance and/or testingrequirements, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to ~WL at50 psig. O.Q,

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

ACTION:

With one containment air lock door'noperable:

1. Maintain at least the OPERABLE air lock door closed and eitherrestore the inoperable air lock door to OPERABLE status within24 hours or lock the BP8%H&air lock door closed;

~ov5~r i2. Operation may then continue until performance of the next

required overall= air lock leakage test provided that thefollowing applicable action is performed at least once per31 days:

a) If the outer air lock door is inoperable, verify bycontrol room indication or local observation, that theinner airlock door is closed and verify the outer air lockdoor is locked closed, or

3.

verify that theb) If the inner air lock door is peraouter air lock door is locked closed>

Otherwise, be in at least HOT STANDS thi he next 6 hoursand in COLD SHUTDOWN within the following 30 hours; and

b. With. the containment air lock inoperable, except as the result of aninoperable air lock door, maintain at leait one air lock door closed;restore the inoperable air lock to OPERABLE status within 24 hoursor be in at least HOT STANDBY within the next 6 hours and in COLDSHUTDOWN within the following 30 hours.

', TURKEY POINT - UNITS 3 8L 4 3/4 6-4

CONTAINMENT SYSTEMS


4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

b.

C.

Within 72 hours following each closing, except when the air lock isbeing used for multiple entries, then at least once per 72 hours, byverifying that the seals have not been damaged and have seatedproperly by vacuum testing the volume between the door seals inaccordance with approved plant procedures.

By conducting overall air lock leakage tests at not less than50 psig, and verifying the overall air lock leakage rate is withinits limit at least once per 6 months."

At least once per 6 months by verifying that only one door in eachair lock can be opened at a time.

"The provisions of Specification 4.0.2 are not applicable.

TURKEY POINT - UNITS 3 4 4 3/4 6-5Jll> 0~ lo~

CONTAINMENT SYSTEMS

INTERNAL PRESSURE


3.6. 1.4 Primary containment internal pressure shall be maintained between -2and +3 psig.


ACTION:

With the containment internal pressure outside of the limits above, restorethe internal pressure to within the limits within 1 hour or be in at least HOTSTANDBY within the next 6 hours and in COLO SHUTDOWN within the following 30hours.


4.6. 1.4 The primary containment internal pressure shall be determined to bewithin the limits at least once per 12 hours.

.TURKEY POINT - UNITS 3 8L 4 3/4 6-600

1988

CONTAINMENT SYSTEMS

AIR TEMPERATURE


3.6. 1.5 Primary containment average air temperature shall not exceed 120'F;


ACTION:

With t conta>nment erage air temp ture greater han 120' reduce heaver e air temper re to within th limit withi hours, be in at eastHOT TANDBY within the next 6 hours and in COLD S UTDOWN wi hin the f owing

0 ours

/DE toS~T 5


4.6.1.5 The primary containment average air temperature shall be the arith-metical average of the temperatures at the following locations and shall bedetermined at least once per 24 hours:

Location

a. ~RN— 04 Azimuth - 58 feet elevation

b. ~TOT- 1204 Azimuth - 58 feet elevation

$4II'M IQtg~pera,'4rc- lncLcc&dN NLRB

Mithgaify of the abov~<empp:/ ~ i', II

go/tou)in > <>d i~tor ken'ho v>5-a. North Mall

b. ~49& 'est Mall

c. 7&4AQQ- C-Filter

AUD IHS~ 4

58 feet elevation

eh ~a Pe.ne r.esP af flan

- 58 feet elevation

58 feet elevation

58 feet elevation

)Wi/

TURKEY POINT " UNITS 3 4 4 3/4 6-7JN 0g

CONTAINMENT SYSTEMS

CONTAINMENT STRUCTURAL INTEGRITY


3.6. 1.6 The structural integrity of the containment shall be maintained at alevel consistent with the acceptance criteria in Specification 4.6.1.6.


ACTION:

a ~ With more than one tendon with an observed lift-offforce betweenthe predicted lower limit and 90K of the predicted lower limit orwith one tendon below 90K of the predicted lower limit, restore thetendon(s) to the required level of integrity within 15 days andperform an engineering evaluation of the containment and provide .aSpecial Report to the Commission within 30 days in accordance withSpecification 6.9.2 or be in at least HOT STANDBY within the next6 hours and in COLD SHUTDOWN within the following 30 hours.


(n4.6.1.).1shall beintegrity

Containment Tendons. The containment tendons'tructural integritydemonstrated every fth year from the date of the initial structuraltest. The tendons'tructural inte rity shall be demonstrated by:

l2Determinin that a random but rep esentative sample of at least Mtendons Q'.dome, 8'ertical, and hoop) each have an observedlift-offforce within predicted limits for each. For each subsequentinspection one tendon from each group may be kept unchanged todevelop a history and to correlate the observed data. If the observedlift-offforce of any one tendon in the original sample populationlies between the predicted lower limit and 90K of the predicted lowerlimit, two tendons, one on each side of thia tendon should'e checkedfor their lift-offforces. If both of these adjacent tendons arefound to be within their predicted limits, all three tendons shouldbe restored to the required level of integrity. This single deficiencymay be considered unique and accceptable.

av

b. With any abnormal degradation of the structural integrity other thanq („/.Q.j , ACTION 6, at a 'level below the acceptance criteria of Specification

mv,d 4.6. 1.6~mrestore the containment to the required level of integrity e~

~ ~

~

~ ~

~ ~

~within 72 hours and perform an engineering evaluation of the contain-ment and provide a Specia'l Report to the Commission within~days inaccordance with Specification 6.9.2 or be in at least HOT STANDBYwithin the next 6 hours and in COLD SHUTDOWN within the f '

30 hours. QOg

: TURKEY POINT - UNITS 3 4 4 3/4 6-8 ' "8 1988

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CONTAINMENT SYSTEMS


b. Performing tendon detensioning, inspections, and material tests on apreviously stressed tendon from each group (dome, vertical, and hoop).A randomly selected tendon from each group shall be completelydetensioned in order to identify broken or damaged wires and deter-mining that over the entire length of the removed wire or strandthat:

1) The tendon wires or strands are free of corrosion, cracks, anddamage,

2) There are no changes in the presence or physical appearance ofthe sheathing filler-grease, and

I')

A minimum tensile strength of 24D,OOO psi (guaranteed ultimat'estrength of the tendon material) for at least three wire orstrand samples (one from each end and one at mid-length) cutfrom each removed wire or strand. Failure of any one of thewire or strand samples to meet the minimum tensile strengthtest is evidence of abnormal degradation of the containmentstructure.

C.

.O~o~6~

LK54h"+(

Performing tendon retensioning of those tendons detensioned forinspection to their observed lift-offforce with a tolerance limitof CNo. goring retensioning of these tendons, the changes in load /&and elongation should be measured simultaneously at a minimum ofthree approximately equally spaced levels of force between zero andthe seating force. If the elongation corresponding to a specificload differs by more than 5X from. that recorded during installation,an investigation should be made to ensure that the difference is notrelated to wire failures or slip of wires in anchorages;

uring the observed lift-offstresses exceed the ave ge mznimumdes> value given below, which are adjusted to uat-tbr elasticlosses; a

Oom~VerticalMoop

133 ksiSi

133 k

e. Verifying the OPERABILITY of the sheathing filler grease by:

1) Minimum grease coverage exists for the different parts of theanchorage .system, and

2) The chemical properties of the filler material are within thetolerance limits as specified by the manufacturer.

TURKEY POINT - UNITS 3 & 4 3/4 6-9'ii''

8„-g

CONTAINMENT SYSTEHS


in<8'<7 gEnd Anchora es and Ad acen Concrete Surfaces. The str ural

integrity of t e orages of all tendons 1n uant to Specification 4. 6. 1. 7. 1 and the adga aces shall be demonstrated bydetermining through inspect a no ap changes have occurred in thevisual appearance end anchorage or the conc ack patterns adjacento the end ages. Inspections-of-the-cona etc-sh ed-dUF4AQ

-the-T-

Conta'>ament Sur aces. The structural integrity of theaccessible sn e error surfaces of the conta, c uding theliner plate, shall be determine th own for each Type A containmentleakage rate test (reference Spe >on . . . a visual snspectson ofthese surfaces. This i on shall be performed pr ~o. he Type A contain-ment leakage r to verify no apparent changes in appearanc therabnorm radation. This first, inspection performed will form the base '~e „f uture surveillances.

4.6.1.6.3 o tainment urfaces

In accordance with 10 CFR 50'ppendix J. Section V. Aia visual inspection of the accessible interior andexterior surfaces of the containment, including the linerplate, shall be performed during the shutdown for (butprior to) each Type A containment leakage rate test(Technical Specification 4.6.1e2 F 1) ~ The purpose of thisinspection shall be to identify any evidence ofstructural deterioration which may affect containmentstructural, integrity or leaktightness. The visualinspection shall be general in nature; its intent shallbe to detect gross areas of widespread cracking,spalling, gouging, rust, weld degradation, or greaseleakage. The visual examination may include theutilization of binoculars or other optical devices.Co'rrective actions taken, and recording of structuraldeterioration and corrective actions, shall be inaccordance with 10 CPR 50, Appendix J, Section V. A.Records of previous, inspections shall be reviewed toverify no apparent changes in appearance. The firstinspection performed will form the baseline for futuresurveillances.


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4.6.1.6.2 End Anchora es and Ad'acent Concrete Surfaces

The structural integrity of the end anchorages of all'endonsinspected pursuant to Specification 4.6.1.6.1 and

the adjacent concrete surfaces shall be demonstrated bydetermining through visual inspection that nounacceptable levels of corrosion exist on the endanchorages and no unacceptable cracking exists in theconcrete adjacent to the end anchorages. Determinationof acceptance levels shall be by engineering evaluationof the areas in question. If unacceptable conditions arefound, the tendons inspected during the previoussurveillance shall be examined to determine whether thecorrosion levels or concrete cracking have increasedsince the previous surveillance. Inspection of adjacentconcrete surfaces shall be performed concurrently withthe containment tendon surveillance (TechnicalSpecification 4.6.1.6.1).

~/ /~ p~p~ y~ 7~~i p/8 l~&lfpf

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CONTAINMENT SYSTEMS

CONTAINMENT VENTILATION SYSTEM

NQ IT ION OP

gg>LACK IAII&IIles~

p g pp y / exhaust >so tion valve openor not scale clued, close and seal that valve isolate thepenetration(s) within 4 hours, otherwise be in least HOT STANDBYwithin next 6 h rs and in COLD SHUTDOWN with'he following 30 hourWith the containm t purge supply and/or ex ust isolation valve(s)open for more than 50 hours during a cale ar year, or for more than200 hours during a c'endar year while i DES 1 or 2, close thevalve(s) or isolate t penetration(s} thin 4 hours, otherwise bein at least HOT STANDS within the ne 12 hours and in COLO SHUTDOWNwithin the following 30 urs. This CTION shall apply to both unitssimultaneously wi.th the its appl able to the sum of cumulativetimes for both units.With a containment purge supp nd/or exhaust isolation valve(s)having a measured leakage rat n excess of the limits of Specifica-tion 4.6. 1.8.3, restore the op rable valve(s) to OPERABLE statuswithin 24 hours, otherwise in t least HOT STANDBY within the next6 hours, and in COLD SHUT N with the following 30 hours.

b.

C.

3.6. .7 Each containment purge supply (48-inch) and exhaust isolation(54-'alvehall be OPERABLE and shall be closed and sealed closed. Operation th

the pu e supp1y and/or exhaust iso1ation valves open shall be limited t 'ssthan or ual to 250 hours during a calendar year and to less than or e al to200 hours er calendar year while in MODES 1 or 2; The purge supplyexhaust iso tion valves shall not be opened wider than 33 or 3Q. deg es,respective1y 0 degrees is fully open).APPLICABILITY: ODES 1, 2, 3, AND 4.

ACTION:

a. With a con inment ur e su 1 and or


4.6. 1.7. 1 Each containment urge supply and exhaus isolation valve shall beverified to be sealed clos at least once per 31 da

4. 6. l. 7. 2 The cumulativ times that Unit 3 and 4 pu supply and/or exhaustvalves have been open ring the calendar year shall b etermined at leastonce per 7 days.

4.6.1.7.3 At leas once per refueling interval, not to ex ed 24 months, eachcontainment purge upply and exhaust isolation valve shall b demonstratedOPERABLE by ver ying that the measured leakage rate is less an or equal to0.20 L when p ssurized to P .

4.6.1.?.4 F each containment purge supply or exhaust isolati valve thatis not sea .

'd closed, at least once per 24 hours verify that. the alve is openo more an 33 or 30 degrees, respectively.


CONTAINHENT SYSTEHS

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEHS

CONTAINMENT SPRAY SYSTEM


3.6.2. 1 Two independent Containment Spray Systems shall be OPERABLE with eachSpray System capable of taking suction from the RWST.


ACTION:

a ~ With one Containment Spray System inoperable restore the inoperable SpraySystem to OPERABLE status within 72 hours or be in at least HOTSTANDBY within the next 6 hours and in COLD SHUTDOWN within thefollowing 30 hours.

b. With two Containment Spray Systems inoperable restore at least oneSpray System to OPERABLE status within 1 hour or be in at least HOTSTANDBY within the next 6 hours and in COLO SHUTDOWN within thefollowing 30 hours. Restore both Spray Systems to OPERABLE statuswithin 72 hours of initial loss or be in at least HOT STANDBY withinthe next 6 hours and in COLD SHUTDOWN within the following 30 hours.


4.6.2. 1 Each Containment Spray System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual,power operated, or automatic) in the flow path that is not locked,sealed, or otherwise secured in position, is in its correct positionand that power is available to flow path components that requirepower for operation;

b. By verifying that on recirculation flow, each pump develops theindicated differential pressure, when tested pursuant toSpecification 4.0.5:

Containment Spray Pump R268 psid at a metered flowrate 2400 gpm(Unit 3), or

h269 psid whil'e aligned in recirculation mode(Unit 4).

TURKEY POINT - UNITS 3 4 4 3/4 6-12 Al( A ~

CONTAINMENT SYSTEMS



3.6.1.7 Each containment purge supply (48 inch) and exhaust (54 inch)isolation valves shall be OPERABLE» ~~ s4dl not' ape~ek e'Idev't ~~

33 or So )~re g, r~~p 4~valyAPPLICADILITT: MODES I, 2, 3 ., AND 4 (CI A;s q„gACTION:

2 ~

3 ~

With the purge supply and/or exhaust isolation valve(s) open wider than 33or 30 degrees, respectiveTy, within $ houe close the valve(s) to less thanor equal to 33 or 30 degrees, respectively; or close the: valve(s), or bein HOT STANDBY within next 6 hours and in COLD SHUTDOWN within thesubsequent 30 ho

sl c, (uri,i's 3~4) "ir fAoJ s I WZWith the purge supp y an or ex aust isolation valve(s) open for more than200 hours per erH+ per calendar year, close the open valve(s) within +hoursor be in at least HOT STANDBY within the next 6 hours and in COLDSHUTDOWN within the subsequent 30 hours.

With the purge supply and/or exhaust isolation valve(s) having a measuredleakage rate in excess of the limits of Specification 4.6.1.7.2 and thecontainment combined leakage for Type B and C penetrations is less than0.60 L , restore the valve leakage within its limit during the next COLD

SHUTDOWN condition of 72 hours or longer.

'Ih Necks I cwJ 2 Per zi4 (IJiih

The.* 4aeh~containmen purge supply and exhaust isolation valve (s) totalopening tim uring a calendar year shall not exceed 200 hours unless arelief has been requested and received from the Coranission.

TH «PA$ 6 tnTC&m ~au J

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CONTAINMENT SYSTEMS



Un< 3caw'.6.1=.7,3The cumulative time that thegurge supply and exhaust valves have

been open in MODES 1 and 2 during the calendar year shall be determined atleast once per 7 days.

j nor™lc C)C @em 24 Ata7l

4,6.1.7.2 At least once per refueling interval each containment purge supplyand exhaust isolation valve shall be demonstrated OPERABLE by verifying thatthe measured leakage rate is less than or equal to 0.20 La when pressurizedto Pa.

4.6.1.7-3 [ea~k on~ pe~ re$ ve,li~~.id~rva(, net ta c»c.easel

Nml4, var fy Act 1hc. tNcc.~~vied sf.ps fepu~ „M e~4ausW is

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CONTAINMENT SYSTEMS

SURVEILLANCE RE UIREHENTS Continu

C.

d.

„.g f. ~~M x< m,~WAt refueling by:

1) Verifying that each automatic valve in the flow path actuatesto its correct position on a containment spray actuation testsignal, and

2) Verifying that each spray pump starts automatical'ly on acontainment spray actuation test signal. The manual isolationvalves in the spray lines at the containment shall be lockedclosed for the performance of these tests.

At least once per 5 years by performing an air or smok ~ estthrough each spray header and verifying each spray nozz e isunobstructed.


CONTAINMENT SYSTEMS

p(<d'>' CONTAINMENT COOLING SYSTEMl~


3.6.2.2 Three emergency containment cooling units shall be OPERABLE.


ACTION:

a 0

b.

~t ~rqen.C gWith one of the above required ontainment cooling units inoperablerestore the inoperable cooling 'unit to OPERABLE status within 72 hoursor be in at least HOT STANDBY within the next 6 hours and in COLDSHUTDOWN within the following 30 hours.

~ e,vft C. rq e v c gWith two or more yf the above required>contain8ent Pooling unitsinoperable, restore at least two coolin to OPERABLE statuswithin 1 hour or be in at least HOT STA Y within the next 6 hoursand in COLD SHUTDOWN within the followi g 30 hours. Restore all ofthe above required cooling units to OP BLE status within 72 hoursof initial loss or be in at 1'east HOT TANDBY within the next 6 hoursand in COLD SHUTDOWN within the follow ng 30 hours.

SURVEILLANCE RE UIREMENTS uni@

2.4.6.2.8 Each emergency containment cooling unit shall be demonstrated OPERABLE: )

a. At least once per 31 days byo

Ql 1 tarting each cooler ueii4p,from the control room and verifyingthat eac 'un4e motor reaches the nominal operating current forthe test conditions and operates for at least 15 minutes, mnd-st-

2) Verifying a coo ing water flow rate of greater t an or enu o2000 gpm to each cool e

b. At least once per refuelin by: Verifying that each unit startsautomatically on a safety in)eftion (SI) test signal> oA<

2.0 IVL

'URKEY POINT - UNITS 3 4 4 3/4 6-14

47i ; .

CONTAINMENT SYSTEMS

3/4.6.3 EMERGENCY CONTAINHENT FILTERING SYSTEM

LIHITING CONOITION FOR OPERATION

3.6.3 Three emergency containment filtering units shall be OPERABLE.


ACTION:

Mith one emergency containment filtering unit inoperable, restore the inoperablefilter to OPERABLE status within 7 days or be in at least HOT STANOBY withinthe next 6 hours and in COLO SHUTDOWN within the following 30 hours.


4.6.3 Each emergency containment filtering unit shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED- TEST BASIS by initiating,from the control room, flow through the HEPA filters and charcoaladsorbers and verifyin that the system operates for at least15 minutes;

C.'KCC.CA 2.4'l~b. At least once pe or (1 a er any ru ura ma ntenance

on the HEPA filter oal adsorber housings, or (2) followingoperational exposure of filters to effluents from painting, fire, orchemical release or (3) after every 720 hours of system operation by:

1) Performance of a visual inspection for foreign material andgasket deterioration, and verifying that the filtering unitsatisfies the in-place penetration and bypass leakage testingacceptance criteria of greater than or equal to 99K removal ofOOP and halogenated hydrocarbons at the system flow rate of37,500 cfm ilOX;

2) Verifying within 31 days after removal, that a laboratory analy-sis of a representative carbon sample obtained in accordancewith applicable portions of Regulatory Position C.S.b of Regula-tory Guide 1.52, Revision 2, March 1978, and performed inaccordance with ANSI N-510-1975, aeets the acceptance criteriaof greater than 99.SX remeval of elemental iodine; and that anycharcoal failing to meet this criteria be. replaced with charcoalthat meets or exceeds the criteria of position C.6a of RegulatoryGuide 1.52; Rev. 2; and

3) Verifying a system flow rate of 37,500 cfm ilOX and a pressuredrop across the HEPA and charcoal filters of less than 6 incheswater gauge during system operation when tested in accordancewith ANSI N510-1975;


CONTAINMENT SYSTEMS


c.'fter maintenance affecting flow distribution, by performance of avisual inspection and an air distribut' t t at a system flow rateof 37,500 cfm ale.

~,f„g(~, n.t t. e) c,e4 Za ~M@d. At least once pe by:

1) Verifying that the system starts on a Safety In)ection testsignal and;

2) Verifying that the filter cooling solenoid valves can be openedby operator action and are opened automatically on a loss offlow signal.

e. After each complete or partial replacement of a HEPA filter bank, byperformance of a visual inspection for foreign material and gasketdeterioration and by verifying that the filtering unit satisfies thein-place penetration and bypass leakage testing acceptance criteriaof greater than 9% removal ot Dgp test aeroso'I while operating the j gsystem at a f„ 37,500 cfm ilOX; and"ore Va o

f. After each complete o partial replacement of a charcoal adsorbed bank,by performance'f a visual inspection for foreign material and gasketdeterioration and by verifying that the filtering unit satisfies thein-place penetration and bypass leakage testing acceptance criteriaof greater than or equal to 99K removal of halogenated hydrocarbon*

while operating the system at a flow rate of 37,500 cfm ilOX.

TURKEY POINT - UNITS 3 Ec 4 3/4 6-16I

s ~

~ ~

CONTAINMENT SYSTEMS

3/4. 6. 4 CONTAINMENT ISOLATION VALVES


3.6.4 The containment isolation valves specified in Table 3.6-1 shall beOPERABLE


ACTION:

*With one or more of the isolation valve(s) specified in Table 3.6"1 inoperable,maintain at least one isolation valve OPERABLE in each affected penetrationthat is open and:

Restore the inoperable valve(s) to OPERABLE status within 4 hours,or

Isolate each affected penetration within 4 hours by use of at leastone deactivated automatic containment isolation valve secured in theisolation position, or

C.

d.

Isolate each affected penetration within 4 hours by use of at leastone closed manual/valve or ind flange, or

C0+ a~nvne~ l~c ionBe in at lea t HOT'STAMBYwithin e next 6 hours and in COLDSHUTDOWN within the following 30 hours.


4.6.4.1 The isolation valves specified in Table 3.6-1 shall be demonstratedOPERABLE prior to returning the valve to service after maintenance, repair ofreplacement work is performed on the valve or its associated actuator, controlor power circuit y performance of a cycling test, )NI.~

CM g +gQ~ ~gC'j I~~

"CAUTION: The inoperable isolation valve(s) may be part of a system(s).Isolating the affected penetration(s) may affect the use of the system(s).Consider the technical specification requirements on the affected system(s)and. act accordingly.

TURKEY POINT - UNITS 3 Ec 4 3/4 6-17 "'~'g

CONTAINMENT SYSTEMS

SURVEILLANCE RE UIREMENTS Continued e'jcAseoI

4 .6.4.2 Each isolation valve specified in Table 3.6-1 rOPERABLE during the COLD SHUTDOWN or REFUELING NODE t least once per refueling

'a ~

b.

C.

4. 6.4. 3

Verifying that on a Phase "A" Isolation test signal, each Phase "A"isolation valve actuates to its isolation position;

Verifying. that on a Phase "B" Isolation test signal, each Phase "B"isolation valve actuates to its isolation position; and

Verifying that on a Containee Vs+nil thon Isolation test signai,each purge, exhaust and instr ment ble d valve actuates to itsisolation position. a,i~

TURKEY POINT - UNITS 3 4 4 3/4 6-18~ ~

) P

FUNCTIONVALVE NUMBER

A. Phase "A" Isolation

TABLE 3.6-1

CONTAINMENT ISOLATION VALVES

MAXIMUISO L IOTI SE ONDS NOTE

1.2,3.4,5.6 ~

7.

15.16.

CV-"-200ACV-*-200BCV-"-200CCV""-204MOV-"-381CV-"-516CV-"™519ACV-A-855CV-*-956ACV-"-956BCV-"-9560MOV-"-1417MOV-"-1418

MOV-"-1426*

Letdown LineLetdown LineLetdown LineLetdown LineRCP Seal Water Leakoff BypassPRT Gas Analyzer LinePRT Makeup Primary Water SupplyNz Supply to AccumulatorsPressurizer Steam Space SamplePressurizer Liquid Space SampleAccumulator Sample LinesCCW to Normal CTMT CoolersCCW Return from Normal CTMT

CoolersG C 8 owdown amp e

S/G B owlowdown Sam le

10101

010

'0

6010101

/AN/A

/N/

33

1 31111,1)11 31 3

,4 7

$ 40

"-Applicable to Unit 3 and Unit 4


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CONTAINMENT ISOLATION VA1 VES

~

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VALVE NUMBER FUNCTION

A. Phase "A" Isolation Continued

MAXIMUISOL IONTI SECONDS OTE

33.34.35.36.

SV-""6428

CV-"-6CV-"-6275CCV-"-6275A-1CV""-627

- 275C-1

CV"""2821CV""-2822SV"""2911SV-"™2912SV-"-2913CV-"-4658ACV-"™4658BCV-*-4659A

gg K CV-"-4659B

CV-"™4668ACV-"-4668BCV-""6165SV-"-6385MOV-""6386

Containment Sump DischargeContainment Sump DischargeContainment Air SampleContainment Air SampleContainment Air SampleRC Drain Tank VentRC Drain Tank VentRC Drain Tank Line to H2

AnalyzerRC Drain Tank Line to H2

AnalyzerRC Drain Tank Pump DischargeRC Drain Tank Pump DischargeBreathing AirPRT Gas Analyzer LineExcess Letdown and RCP Seal

Mater Return to CVCSRCS Sam le

owdownBlowdown

S/G C

S/G A B ypassBlowdown Bypass

S/G C Blowdown B ass

101

01010101010

1010.1010

-18'/A

/A

33

1 31 31 31 3

1,1 j

1,1 $

11 313

l41~4)~4),4,


TURKEY POINT " UNITS 3 8( 4 3/4 6-20~ I, ''

~ i'',

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VALVE NUHBER

B. Phase "B" Isolation

FUNCTION


CONTAINHENT ISOLATION VALVES

HAXIHUMISO LAT N

TIHE SECONDS'OTES

1.2.3.4.

<OY~*-626HOV"'™716AHOV-"-716BHOV-*-730

CC Return from RCPCC Supply to RCPCC Supply to RCPCC Return from RCP Thermal

Bar rier Cool ers

3,4,73,4,7 i

3,4,7 "

3,4,7 t

C. Containment Ventilation Isolation

1.2.

"3.

5.6.

POV-"™2600POV-"-2601POV-*-2602POV-"-26Q3CV-""2819CV""-2826

Containment Purge SupplyContainment Purge SupplyContainment Purge ExhaustContainment Purge ExhaustInstrument Air BleedInstrument Air Bleed

1.313131,31 $ 31i3


TURKEY POINT - UNITS 3 5 4 3/4 6-21't ~v ~ ~

~ f4% ~V

CONTAINMENT SYSTEMS

3/4.6.5

HYDROGEN MONITORS


3.6.5 Two independent containment hydrogen monitors shall be OPERABLE.


ACTION:

a. With one hydrogen monitor inoperable, restore the inoperable monitor toOPERABLE status within- 30 days or be in at least HOT STANDBY within thenext 6 hours.

b. With both hydrogen monitors inoperable, restore at least one monitor toOPERABLE status within 72 hours or be in at least HOT STANDBY within thenext 6 hours.


4.6.5. 1 Each hydrogen monitor shall be demonstrated OPERABLE by the performanceof a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONALTEST at least once per 31 days, and at least once per 92 days on a STAGGEREDTEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:

a. One volume percent hydrogen, balance nitrogen, and

b. Four volume percent hydrogen, balance nitrogen.

4.6.5.2 The flow path to each hydrogen monitor shall be demonstrated OPERABLEat least once per 31 days by a system walkdown to verify that each accessiblemanual, power operated, or automatic valve is in its correct position and thatpower is available to those components related to the operability of theflowpath.

TURKEY POINT - UNITS 3 4 4 3/4 6"22

CONTAINMENT SYSTEMS

3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM


3.6.6 A Post Accident Containment Vent System shall be OPERABLE.


ACTION:

With the Post Accident Containment Vent System inoperable, restore the PostAccident Containment Vent System to OPERABLE status within 7 days or be in atleast HOT STANDBY within 6 hours.


4.6.6 The Post Accident Containment Vent System shall be demonstrated OPERABLE:

0 b.

At least once per 31 days by demonstrating system flow path operabilityvia a system walkdown to verify that each accessible manual valve isin its correct position.-

,nA 4o ~xM 24. w~

At least once per or (1) after any structural maintenanceof the HEPA filter or charcoal adsorber housings, or (2) followingoperational exposure of filters to effluents from painting, fire, orchemical release in any ventilation zone communicating with thesystem, or (3) after 720 hours of system operation or (4) afterreplacement of a filter by:

1) A visual inspection of the system for foreign materials andgasket deterioration and verifying that the filter system satisfiesthe penetration and bypass leakage testing acceptance criteriaof less than 1X for DOP and halogenated hydrocarbon testsconducted at a design flow rate of 55 cfm ilOX;

2) Verifying, within 31 days after removal, that a laboratory analysisof a representative carbon sample

performed in accordance with ANSIN510-1975, meets the methyl iodide removal criteria of greaterthan or equal to St% and that any charcoal failing to Neet thecriteria be replaced with charcoal that meets or exceeds thecriteria of Position C.6.a of Regulatory Guide 1.52, Revision 2.

TURKEY POINT - UNITS 3 5 4 3/4 6-23 J(ft'& >:;;.

CONTAINMENT SYSTEMS


r~k,g;.~ ~.4 t; ~x~ z4 n~c. At least once per Rby

1) Verifying that the pressure drop across the combined HEPA filterand charcoal adsorber is less than 6 inches Water Gauge at aflow rate of 55 cfm k 10K,

2) Visual inspection of the system and operation of all valves.

0

TURKEY POINT - UNITS 3 4 4 3/4 6-24":<' 9 H>a

3/4.7 PLANT SYSTEMS

3/4.7.1 TURBINE CYCLE

SAFETY VALVES

f v ves associated with each steamhall be OPERABLE with lift3.7.1.1 Al mai s ea i C

generatorsettings as specs se s a e 3. - .


ACTION:

With (3) reactor coolant loops and associated steam generators in operation andwith one or more main steam line Code safety valves inoperable, operation inMODES 1, 2, and 3 may proceed provided, that within 4 hours, either the in-operable valve is restored to OPERABLE status or the Power Range Neutron FluxHigh Trip Setpoint is reduced per Table 3.7;1; otherwise, be in at least HOTSTANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30hours. A

4.7. 1. 1 No additional requirements other than those required by Specification4. 0. 5.

TURKEY POINT - UNITS 3 4 4 3/4 7-1 ) [lgJ'l ~ a ~

TABLE 3.7-1

MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITHINOPERABL STEAM L E SA E Y VALVES

MAXIMUM NUMBER OF INOPERABLESAFETY VALVES ON ANY

OPERATING STEAM GENERATOR

MAXIMUM ALLOWABLE POWER RANGENEUTRON FLUX HIGH SETPOINT

PERCENT OF RATED THERMAL POWER

82

54

27

TABLE 3.7-2

STEAM LINE SAFETY VALVES PER LOOP

VALVE NUMBER

~Loo A ~Loo 8 ~Loo C

1. RV1400 RV1405 RV1410

2. RV1401 RV1406 RV1411

3. RV1402 RY1407 RV1412

4. RV1403 RV1408 RV1413

LIFT SETTING %IX "

1085 psig

1100 psig

1115 psig

1130 psig

ORIFICE SIZES UARE INCHES

16

16

16

16

*The lift setting pressure shall correspond to ambient conditions of the valveat nominal operating temperature and pressure.


PLANT SYSTEMS

AUXILIARYFEEDWATER SYSTEM


0

3.7.1.2 Two independent auxiliary feedwater trains including 3 pumps asspecified in Table 3.7-3 and associated flowpaths shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3

ACTION:

1) With one of the two required independent auxiliary feedwater trainsinoperable, either restore the inoperable train to an OPERABLEstatus within 72 hours, or place fected unit(s) in at leastHOT STANDBY within the next 6 ho r n in HOT SHUTDOWN within thefollowing 6 hours.

2) With both required auxiliary feedwater trains inoperable, within 2hours either restore both trains to an OPERABLE status, or restoreone train to an OPERABLE status and follow ACTION statement 1 abovefor the other train. If neither train can be restored to anOPERABLE status within 2 hours, verif the availability of bothstandby feedwater pumps and plac tPe ffected unit(s) in at leastHOT STANDBY within the next 6 ho s%n in HOT SHUTDOWN within thefollowing 6 hours. Otherwise, in i corrective action to restoreat least one auxiliary feedwater train to an OPERABLE status as soonas possible and follow ACTION statement 1 above for the other train.

3) With a single auxiliary feedwater pump inoperable, within 4 hours,verify OPERABILITY of two independent auxiliary feedwater trains, orfollow ACTION statements 1 or 2 above as applicable. Upon verifica-tion of the OPERABILITY of two independent auxiliary feedwater trains,restore the inoperable auxiliary feedwater pump to an OPERABLE statuswithin 30 days, or p e e operating unit(s) in at least HOT

STANDBY within 6 ho Pan in HOT SHUTDOWN Within the following6 hours. The provis f Specification 3.0.4 are not applicableduring the 30 day period for the inoperable auxiliary feedwater pump.


4.7.1.2.1 The required independent auxiliary feedwater trains shall bedemonstrated OPERABLE:

a. At least once 'per 31 days on a STAGGERED TEST BASIS by:

1) Verifying by control panel indication and visual observation ofequipment that each steam turbine-driven pump operates for 15minutes or greater and develops a flow of greater than or

/j > gg7/y(J p-.l.,- f~ kc X/i c nfl 5i'~ ;j7'a Pcs:.I/i> .g )'fl@W~~~I~AO'OBQ,> '14m <4c, new/ ]g Rod~~ an) (n IKON 5h'045CdR) u>'tlirt lk

f~l'TURKEY

POINT - UNITS 3 8 4 3 4 7 3 "OU/5

PLANT SYSTEMS

AUXILIARY FEEDMATER SYSTEM


b.

equal to 373 gpm to the entrance of the steam generators. Theprovisions of Specification 4.0.4 are not applicable for entryinto MODES 2 and 3;

2) Verifying by control panel indication and visual observation ofequipment that the auxiliary feedwater discharge valves and thesteam supply and turbine pressure valves operate as required todeliver the required flow during the pump performance testabove;

3) Verifying that each non-automatic valve in the flow path thatis not locked, sealed, or otherwise secured in position is inits correct position; and

4) Verifying that er is 'available to those corn onents whichrequire power or path operability.

ref'vcr ~k t ex~ z.4 ~At least once pe by:

1) Verifying that each automatic valve in the flow path actuatesto its correct position upon receipt of each AuxiliaryFeedwater Actuation test signal, and

2) Verifying that each auxiliary feedwater pump receives a startsignal as designed automatically upon receipt of each AuxiliaryFeedwater Actuation test signal.

4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall bedemonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 daysprior to entering MODE 1 by verifying normal flow to each steam generator.

TURKEY POINT - UNITS 3 Ec 4 3/4 7-4~ ~

0

TABLE 3.7-3

AUXILIARYFEEDWATER SYSTEM OPERABILITY

UNIT TRAIN STEAM SUPPLY FLOWPATH PUMP OISCHARGE MATER FLOWPATH

3 '1 SG 3C via MOV-3-1405<)

A or C SG 3A via CV-3-2816(2)or SG 3B via MOV-3-1404 (1)

SG 3B via CV-3-2817SG 3C via CV-3-2818

SG 3A via MOV-3-14031

B or C

or SG 3B via MOV-3-1404( ) SG 3A via CV-3-2831SG 3B via CV-3-2832SG 3C via CV-3-2833

SG 4C via MOV-4-1405<>

A or C(2)

or SG 4B via HOV-4-1404 (1) SG 4A via CV-4-2816SG 4B via CV-4-2817SG 4C via CV-4-2818

SG 4A via MOV-4-1403>

B or C

or SG 4B via HOV-4-1404SG 4A via CV-4-2831SG 4B via CV-4-2832SG 4C via CV-4-2833

ei NOTES:

Steam admission valves MOV-3-1404 and MOV-4-1404 can be aligned to eithertrain (but not both) to restore OPERABILITY in the event MOV-3-1403 or

* MOV-3-1405, or MOV-4-1403 or MOV-4-1405 are inoperable.

During single and two unit operation, one pump shall be OPERABLE in eachtrain and the third auxiliary feedwater pump shall be OPERABLE and capableof being ed from, and supplying water to either train, except as notedin ACTIO P>o Technical Specification 3.7.1.2. The third auxiliaryfeedwate (normally the "C" pump) can be aligned to either train torestore OPERABILITY in the event one,of the required pumps is inoperable.

If any local manual realignment of valves is required when operating theauxiliary feedwater pumps, a dedicated individual, who is in communicationwith the control room, shall be stationed at the auxilia area. Uponinstructions from the control room, this operator would real gn the valvesin the AFW system train to its normal operational alignment.


Sii N',,

PLANT SYSTEHS

CONDENSATE STORAGE TANK


3.7.1.3 The Condensate Storage Tanks shall be OPERABLE with a contained watervolume of at least 185,000 gallons of water as

follows:'/Ag/QMb"i+ 7~iot ~ 4$ca(~i~I i~~Ops Q~ ~ 8

a) ONE water supply from either Condensate Storage Tank ihcludingflowpath piping and valves.

Qe~~g Lfng+ +I@V TO &Cata l(w ItIl'g /kgOQQ P

~aWa) ONE water supply from 4he- unit's corresponding Condensate Storage

Tank including flpwpath piping and valves.

APPLICABILITY: HODES 1, 2 and 3.

ACTION:It ll t At o AkovCP&+ +

1) With one water supply from a Condensate Storage Tank inoperable, within 4hours, either realign the other Condensate Storage Tank containing therequired water volume to the suction of the Auxiliary Feedwater pumps orrestore the inoperable water supply to OPERABLE status or be in at least l

HOT STANDBY in the next 6 hours and in HOT SHUTDOWN within the following 6hours.

~ 2) th water supplies from the Co le,;rl~8f Y within 4 er supply from either Condensate StorageTank to status st HOT STANDBY within the next 6

and in HOT SHUTDOWN within the fo owWoW M~r*At e~A4~~ HC>E3

2))pcgp7

With one water supply from a Condensate Storage Tank inoperable, restorethe inoperable water supply to OPERABLE status within 4 hours or place oneunit in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWNwithin tht following 6 hours. Refer to Single Unit Operation ACTION forsingle unit at or above HODE 3.

ith both water supplies from the Condensate Storagewi hour restore one water supply from sate Storage Tank toOPERABLE s or place one unit in st HOT STANDBY within the next6 'hours and in H he following 6 hours. If unable torestore at least one supply ensate Storage Tank toOPERABLE statu in 4 hours from initial dec of inoperability,the seco t shall be placed in at least HOT STANDBY w e next 6h and in HOT SHUTDOWN within the. following 6 hours.

TURKEY POINT - UNITS 3 SL 4 3/4 7-6 t ~

-f,s >! lW

/

/ w ~ICQ5ieraue/ ~

7A vi&51

ld ~IG I'8 ,'." np

stoJ~Mewf

~g

fQC J0l ~Q~

CQf'l ~~6 ) JC

to t~ jOI .8

g//~ Ac7ioLi

ct" Q I ta vI[4 a

ted ~

ca<4 gee Co~de ...c i'e,"'ocaae(i

G~ 9 6 '4.8

Q.NDVP- TG f

0

I ~W~ion3 r ivg L~ 5M~K

PLANT SYSTEHS


4.7.1.3 The Condensate Storage Tanks shall be demonstrated OPERABLE at leastonce per 12 hours by verifying the contained water volume is within its limitwhen the tank is the supply source for the auxiliary feedwater pumps.


PLANT SYSTEHS

SPECIFIC ACTIVITY


3.7. 1.4 The specific activity of the Secondary Coolant System shall be lessthan or equal to 0. 10 microCurie/gram DOSE E(UIVALENT I-131.

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

ACTION:

With the specific activity of the Secondary Coolant System greater than 0.10microCurie/gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6hours and in COLO SHUTDOWN within the following 30 hours.


4.7.1.4 The specific activity of the Secondary Coolant System shall bedetermined to be within the limit by performance of the sampling and analysisprogram of Table 4.7-1.

'TURKEY POINT - UNITS 3 4 4 3/4 7-8

TABLE 4.7"1

SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY

SAMPLE AND ANALYSIS PROGRAM

TYPE OF MEASUREMENTAND ANALYSIS

1. Gross RadioactivityDetermination

2. Isotopic Analysis for DOSEE(UIVALENT I-131 Concentration

SAMPLE AND ANALYSISFRE UENCY


a) Once per 31 days, when-ever the gross radio-activity determinationindicates concentrationsgreater than 10X of theallowable limit forradioiodines.

b) Once per 6 months, when-ever the gross radio-activity determinationindicates concentrationsless than or equal to lOXof the allowable limitfor radioiodines.

"A gross radioactivi analy s shall sist of the antitati measu ementof th total spec c act ity of t secondary co ant exc for r

io-'uc

des with h f-live less th IO minutes. termina on of tco tributors the g ss speci c activity sh be ba d upon ose ergypeaks identi iable w' a 95K onfidence level


PLANT SYSTEMS

HAIN STEAM LINE ISOLATION VALVES


3.7.1.5 Each main steam line isolation valve (HSIV) shall be OPERABLE.


ACTION:

MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue~

~rovided the inoperablg valve is restored to OPERABLE status within24'ours; otherwise be i OT STANDBY within the next 6 hours and

sn HOT SHUTDOWN within the following 6 hours.

MODES 2 and 3: a (e.4With one HSIV inoperable, subsequent operation in MODE 2 or 3 ma roceedprovided the isolation valve Is malntalned closed. Othervlse, be n OT 'gSTANDBY within the next 6 hours and in HOT SHUTDOWN within the follow ng6 hours.


4.7.1.5 'ach MSIV shall be demonstrated OPERABLE by verifying full closurewithin 5 seconds when tested pursuant to Specification 4.0.5. The provisionsof Specification 4.0.4 are not applicable for entry into HODE 3. P2

TURKEY POINT - UNITS 3 8t 4 3/4 7-10

PLANT SYSTEMS

STANDBY FEEDWATER SYSTEM


3.7. 1.6 Two standby feedwater pumps shall be OPERABLE" and at least 60,000gallons of water (available volume), shall be in the Demineralized WaterStorage Tank"".

APPLICABILITY: MODES 1, 2 and 3

ACTION:

b.

With one standby feedwater pump inoperable, restore the inoperable pumpsto available status within 30 days or submit a SPECIAL REPORT per 3.7. 1.6d.

With both standby feedwater pumps inoperable: g~~yQ'ICa(1. Within 24 hours, notify the NRC and provide cause for

and plans to restore pump(s).to status and,

2. Submit a SPECIAL REPORT per 3.7.1.6d." Q~g.//LANc. With less than 60,000 gallons of water in the Demineralized Water Storage

Tank restore the available volume to at least 60 00 lions within 24hours or submit a SPECIAL REPORT per 3.7.1.6d.

~n~eV4td. If a SPECIAL REPORT is required per the above sp cs ca ons submit a

report describing the cause of the , action taken and aschedule for restoration within 30 da s in accordance with 6.9.2.

E'. The Isrovisirns eF + >4,'< 1'iirn s.o.Z are hot apphavuk4 .SURVEIL U EMENTS

4.7. 1.6. 1 The Demineralized Water Storage tank water volume shall be deter-mined to be within limits at least once per 24 hours.

4.7.1.6.2 At least monthly verify the standby feedwater pumps are OPERABLE- bytesting in recirculation on a STAGGERED TEST BASIS.

no+to ~X~ 24- +~ ~4.7.1.6.3 During each refueling<eu4age; verify operability of the respectivestandby feedwater pump by powering from the non-safety grade diesel generatorsand providing feedwater to the steam generators.

These pumps are not safety related equipment and do not require plant safetyrelated emergency power sources for operability

"*The Demineralized Water 'Storage Tank is non-safet grade.

'.IIV'4flounged~Is lnotlYlvllg l Die~~

3/4 7-11

PLANT SYSTEMS

3/4.7.2 COMPONENT COOLING WATER SYSTEM


3.7.2 The Component Cooling Water System (CCW) shall be OPERABLE with:

Three CCW pumps,%LAO tn ServlCc, 7'~ a.~

CCW heat exchangers+ s C

gQ

EI

O

b.capable of removing design basis

heat loads~~

restore the knoperahle CCW pump to OPERABLE status w$ th$ n ays or )4 gbe in HOT STANDBY within the next 6'hours and in COLD SHUTOOWNwithin the following 30 hours. The provisions of Specification3.0.4 are not applicable.

73.With only one CCW pump OPERABLE or with two CCW pumps OPERABLE butnot from independent power supplies, restore two pumps fromindependent power supplies to OPERABLE status within hours or bein HOT STANDBY within the next 6 hours and in COLD S DOWN withinthe following 30 hours.

z "o

+1a

C,W M

b.

C.ol be in HOT STANOBY + ~within th t 6 hours and in COLD SHUTOOWN within the following

30 hours <+ +8c

lB

APPLICABILITY S 1,insevvrpee

oP rernovine riescj n A»see heo:P (e»efs restore ~e»»tii ~ iyhin i beer Qo

ae on y two C pump n ependent power supplies OP LE,

URVEILLANCE RE UIREMENTS

<CaJ hespf egcJl4.n~rS'~5ayl/I I kj vAc 77~ p l%

~g8f','

.7.2 'he Component Cooling Water System (CCW) shall be eaonstrated OPERABLE:

a. At least once per 12 hours, by v ing thatone p ~j4 c pable o removing design

basis heat loads.0

7 / + ~~4,kly~lS Jg gong tlat s ~l~ Ml Pu~P'"~g,p



b.

C.

At least once per 31 days by: (1) verifying that each valve (manual,power-operated, or automatic) servicing safety-related equipment thatis not locked, sealed, or otherwise secured in position is in itscorrect position, and (2) verifying by a performance test the heatexchanger surveillan v

,ylok- ~ ~~ ~ IYIMRAt least once pe < , by verifying that:

1) Each automatic valve servicing safety-related equipment actuatesto its correct position on a SI test signal, and

2) Each Component Cooling Water System pump starts automaticallyon a SI test signal.

3) Interlocks required for CCW operability are OPERABLE.

'TURKEY POINT - UNITS 3 8L 4 3/4 7-13

PLANT SYSTEMS

3/4.7.3 INTAKE COOLING WATER SYSTEM


3.7.3 .The Intake Cooling Water System (ICW) shall be OPERABLE with.

ufhree ICW pumps.4(4/+ 4Z

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

L gQ~) gfr~) nevy L$7Egf.ACTION:

(-")

) b.

With only two ICW.pumps with independent power supplies OPERABLE,restore the inoperable ICW pump to OPERABLE status within 7 days orbe in HOT STANDBY within the next 6 hou LQ

wit in fol lowin 30 hours. ~d pro v'lsi~ o Specs i~~f~ ' 2.Z. 0.4 tl 4dt r ~

~th only one I W pump PERABLE or with two ICW pumps OPERABLE butfrom independent power supplies, restore o pumps from independent ) P.power supplies to OPERABLE status within hours or be in HOTSTANOBY within the next S hours and in CO SNUTOOWN within the )+3following 30 hours.


4.7.3 The Intake Cooling Water System (ICW) shall be demonstrated OPERABLE:

At least once per Sl days by verifying that each valve (manual,power-operated, or automatic) servicing safety-related equipment thatis not locked, sealed, or otherwise secured in position is in itscorrect position; and

re)gg ) I Vl>

V)dp~ dAt least once pe wn, by verifying that: )

'.1) Each automatic valve servicing safety-related equipment actuates

to its correct position on a SI test signal, and

2) Each Intake Cooling Water System pump starts automatically ona SI test signal.

3) Interlocks required for system operability are OPERABLE.

TURKEY POINT - UNITS 3 5 4 3)'4 7-14J( ~ )'

PLANT SYSTEMS

3/4. 7.4 ULTIMATE HEAT SIN

LIM ING ONDITION FOR OPERATION

vexw ~sSPS 4

~ ~

~3.7.4 The ul imate heat sink shall be OPERABLE with an average upply watertemperature to the Intake Cooling Water System less than or e l to 954F.

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

ith the requirement ',of the above specification not s isfied, be in at leastHOT STANDBY within 12 ours and in COLD SHUTDOWN wit the following 30 hours.This action shall be a licable to both units simul neously.


4.7.4 The ultimate heat sink shall e determined OPERABLE at least once per24 hours by verifying the average u ly water temperature" to the IntakeCooling Water System to be withi its imit.

Porta e monitors may be used to measure the temperature.


PLANT SYSTEMS

R lVICL6BQC'f3/4.7.5 CONTROL ROOM VENTILATION SYSTEM


2 Wdo Yg~Q3.7. 5 The Control R Ventilation Syste sla11 be OPERABLE.

APPLICABILITY: All MODES.

ACTION:

MODES 1, 2, 3 and 4: graf.Grn~~gNth the Control Room~Ventilatio System inoperable, suspend allmovement of fuel in the sp uel pool and restore the inoperable systemto OPERABLE status within ours 'or be in at least HOT STANDBY withinthe next 6 hours and in COL SHUTDOWN within the following 30 hours.This ACTION shall apply to both units simultaneously.g

MODES 5 and 6:&~%

With the Control RoomAVentllatlon System'inoperable, suspend all operations ) E,involving CORE ALTERATIONS, movement of fuel in the spent fuel pool, orpositive reactivity changes. This action shall apply to both units tsimultaneously.


4. 7. 5 -Ea~ ControlRoom~Ventilation System shall be demonstrated OPERABLE:

a. At least once per 12 hours by verifying that the control room airtemperature is less than or equal to 120 F;

b. At least once per 31 days by initiating, from the control room, flowthrough the HEPA filters and charcoal adsorbers and verifying thatthe system operates for at least 15 minutes;

c. At least once per 18 months or (1) after 720 hours of systemopera-'ion,

or (2) after any structural maintenance on the HEPA filter orcharcoal adsorber housings, or (3) following operational exposure ofthe filters to effluents from painting, fire, or chemical release inany ventilation zone comaunicating with the system, or (4) aftercomplete or partial replacement of a filter bank by:

~ -J r g =. AICT!od .pp: rS 7o

W"''g/'~'yt~/:/'''vc l - '

ISJi>~«de y'!ba i.~] GO rtOOroTURKEY POINT - UNITS 3 4 4 3/4 7-16

PLANT SYSTEMS


d.

e.

y(<Verifying that th eanup system satisfies the in-place pene-tration an'd bypass leakage testing acceptance criteria of greaterthan or equal to 99K 00P and halogenated hydrocarbon removal ata system flow rate of 1000 cfm %10K.

2) Verifying, within 31 days after removal, that a laboratory analysisof a representative carbon sample obtained in accordance withRegulatory Position C. 6. b of Regulatory Guide 1.52, Revision 2,March 1978, and analyzed per ANSI N510-1975, meets the criteriafor methyl iodine removal efficiency of greater than or equal to90 o the charcoal be replaced with charcoal that meets orexce s the criteria of position C.6.a. of Regulatory Guide 1.52(Revision 2) and

3) Verifying by a visual inspection the absence of foreignmaterials and gasket deterioratioy in We. HEPT Pte~~ ~~Mc~<~f %4tSO~~

At least once per 12 months by verifying that the pressure dropacross the combined HEPA filters and charcoal adsorber banks is lessthan 6 inches Water Gauge while operating the system at a flow rateof 1000 cfm tlOX;

At least once per 18 months by verifying that on a Containment Phase"A" Isolation test signal the, system automatically switches into therecirculation mode of operation.


PLANT SYSTEMS

3/4.7.6 SNUBBERS


03.7.6 All snubbers shall be OPERABLE. The only snubbers excluded from therequirements are those installed on nonsafety-related systems and then only-if their failure of failure of the system on which they are installed wouldhave no adverse effect on any safety-related system.

APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located onsystems required OPERABLE in those MODES.

ACTION:

With one or more snubbers inoperable on any system, within 72 hours replace or re-store the inoperable snubb to OPERABLE status and perform an engineering eval-uation per Specification .7.6 on the attached component or declare the attachedsystem inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE RE UIREMENTS I. 7 (cpf

No. of Inoperable Snubbers of Each Type Subsequent Visualon an s stem er Ins ection Period In ection Period" ""

11 p i- „$+o 12 months 4 25%2 Negate. i>gi" 6 months g 25%

3,4 eMcaad Z4 ~~ ". 124 days k 25%

5,6,7 62 days 4 25~ 8 or more 31 days 4 25K

"The inspection interval for each type of snubber (on a given system) shall notbe lengthened more than one step at a time unless a generic problem has beenidentified and corrected; in that event the inspection interval may be length-ened one step the first time and two steps thereafter if no inoperable snubbersof that type are found (on that system).

*"The provisions of Specification 4.0.2 are not applicable.

~ . TURKEY POINT - UNITS 3 4 4 3/4 7-18

4.7.6 Each snubber shall be demonstrated OPERABLE by performance of thefollowing augmented inservice inspection program in addition to the require-ments of Specification 4.0.5.~ltt l,:

As used in this specification, type of snubber shall mean snubbersof the same design and manufacturer, irrespective of capacity.

b. Visual Ins ectionsSnubbers are categorized as inaccessible or c s ible during reactoroperation. Each of these groups (inaccessib accessible)'may gZbe inspected independently according to the sc edule below. Thefirst inservice visual inspection of each type of snubber shall beperformed after 4 months but within 10 months of. commencing POWER

OPERATION and shall include all snubbers. If all snubbers of eachtype (on any system) are found OPERABLE during the first inservicevisual inspection, the second inservice visual inspection (of thatsystem) shall be performed at the first refueling outage. Otherwise,subsequent visual inspections of a given system shall be performed inaccordance with the following schedule:

PLANT SYSTEMS


C. Visual Ins ection Acce tance Criteria

per Specification +1+-condition and determine

Visual inspections shall verify that: (1) there are no visible indi-cations of damage or impaired OPERABILITY, (2) attachments to thefoundation or supporting structure are secure, and (3) fasteners forattachment of the snubber to the component and to the snubber anchorageare secure. Snubbers which appear inoperable as a result of visualinspections may be determined OPERABLE for the purpose of establishingthe next visual inspection interval, provided that: (1) the causeof the rejection is clearly established and remedied for that partic-ular snubber and for other snubbers th'at may be generically susceptible;and (2) the affected snubber is functionally tested in the as-found

d OPERABLE

4

q.q. (gtd. Functional Tests

For each unit during refueling , a representative sample ofsnubbers shall be tested using he following sample plan:

1) At least 10K of the total number of safety related snubbers for.the respective unit identified by site records shall be func-tionally tested either in-place or in a bench test. For eachsnubber of a type that does not meet the functional test accep-tance criteria of Specification,4.7.6e, an additional 10K ofthat type of snubber shall be functionally tested until no morefailures are found or until all snubbers of that type have beenfunctionally tested;

2) The representative sample selected for functional testing shallinclude the various configurations, operating environments andthe range of size and capacity of snubbers. At least 25X ofthe snubbers in the representative sample shall include snubbersfrom the following categories;

A. Snubbers within 5 feet of heavy equipment (ex. valves,pumps, turbines, motors, etc.)

B. Snubbers within 10 feet of the discharge from a safetyrelief valve.'0

( ~ 3) Snubbers identified by site records as "Especially Oifficult tove" or in "High Radiation Zones DuHng Shutdown" shall also

be i uded in the representative sample."ol

"Perma nt other emptions from functional testing for individual snubbersin th se cate s may be granted by the Commission only if a justifiablebasis mption is presented and/or snubber life destructive testing wasperformed to qualify snubber OPERABILITY for all dysign conditions at eithe~the completion of their fabrication or at a subsequent date.

TURKEY POINT - UNITS 3 8L 4 3/4 7"19

PLANT SYSTEHS


In addition to the regular sample, snubbers which failed theprevious functional test shall be retested during the next testperiod. If a spare snubber has been installed in place of afailed snubber, then both the failed snubber (if it is repairedand installed in another position) and the spare snubber shallbe retested. Test r'esults of these snubbers may not be includedfor the re-sampling.

e. Mechanical Snubbers Functional Test Acce tance Criteria

('hesnubber functional test shall verify that:

1) Activation (restraining action) is achieved with the specifiedrange of velocity or acceleration in both tension andcompression;

2) Snubber release rate, where required, is within the specifiedrange in tension and compression,

3) The force required to initiate or maintain motion of thesnubber is within the specified range in both directions oftravel.

f. Functional Test Failure Anal sis

An engineering evaluation shall be made of each failure to meet thefunctional test acceptance criteria to determine the cause of thefailure. The results of this evaluation shall be used, if applicable,in selecting snubbers to be tested in an effort to determine theOPERABILITY of other snubbers irrespective of type which may besubject to the same failure mode.

If any snubber selected for functional testing either fails toactivate or fails to move, i.e., frozen-in-place, the cause will beevaluated under the provisions of 10 CFR Part 21.

Should the results of the evaluation indicate that the failure wascaused by either manufacturer or design deficiency, further actionshall be taken, if needed, based on manufacturer or engineeringrecoaeendations.

For the snubber(s) found inoperable, an evaluation shall be pe~formedon the components to which thi inoperable snubbers are attached.The purpose of this evaluation shall be to determine if the componentsto which the inoperable snubber(s) are attached were adversely affectedby the inoperability of the snubber(s) in order to ensure that thecomponent remains capable of meeting the designed service.


PLANT SYSTEMS


g. Snubber Service Life Monitorin Pro ram

A record of the service life of each snubber, the date at which thedesignated service life commences and the installation andmaintenance records on which the designated service life is basedshall be maintained as required by Specification 6.10.3m.

Concurrent with the first inservice visual inspection and duringrefueling shutdown thereafter, the installation and maintenancerecords for each safety related snubber as identified by site recordsshall be reviewed to verify that the indicated service life has notbeen exceeded or will not be exceeded prior to the next scheduledsnubber service life review, the snubber service life shall bereevaluated or the snubber shall be replaced or reconditioned so asto extend its service life beyond the date of the next scheduledservice life review. This re-evaluation, replacement or reconditioningshall be indicated in the records.

.TURKEY POINT - UNITS 3 8E 4 3/4 7-21

PLANT SYSTEMS

3/4.7.7 SEALED SOURCE CONTAMINATION


3.7.7 Each sealed source containing radioactive material either in excess of100 microCuries of beta and/or gamma emitting material or 5 microCuries of alphaemitting material shall be free of greater than or equal to 0.005 microCurieof removable contamination.

APPLICABILITY: At all times.

ACTION:

aO With a sealed source having removable contamination in excess of theabove limits, immhdiately withdraw the sealed source from use andeither:

b.

l. Decontaminate and repair the sealed source, or

2. Dispose of the sealed source in accordance with CommissionRegulations.

The provisions of Specification 3. 0. 3 are not applicable.


4.7.7.1 Test Requirements - Each sealed source shall be tested for leakageand/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or anAgreement State.

The test method shall have a detection sensitivity of at least 0.005microCurie per test sample.

4.7.7.2 Test Frequencies - Each category of sealed sources (excludingstartup sources and fission detectors previously sub)ected to core flux) shallbe tested at the frequency described below.

a. Sources in use - At least once per 6 months for all sealed sourcescontaining radioactive materials:

1) With a half-life greater than 30 days (excluding Hydrogen 3),and

2) , In any form other than gas.


PLANT SYSTEMS


b. Stored sources not in use - Each sealed source and fission detectorshall be tested prior to use or transfer to another licensee unlesstested within the previous 6 months. Sealed sources and fissiondetectors transferred without a certificate indicating the last testdate shall be tested prior to being placed into use; and

C. Startup sources and fission detectors - Each sealed startup sourceand fission detector shall be tested within 31 days prior to beingsubjected to core flux or installed in the core and following repairor maintenance to the source.

4.7.7.3 Reports - A report. shall be prepared and submitted to the Commissionon an annual basis if sealed source or fission detector leakage tests revealthe presence of greater than or equal to 0.005 microCurie of removablecontamination.

4.7.7.4 A complete inventory of licensed radioactive materials in possessionshall be maintained current at all times.


PLANT SVSTEHS

3/4.7. 8 FIRE SUPPRESSION SYSTEHS

FIRE WATER SUPPLY AND DISTRIBUTION SYSTEH


3.7.8.1 The Fire Mater Supply and Distribution System shall be OPERABLE with:

a. At least two fire suppression pumps, one electric dr iven ~~d u Idl

-ca , with their discharge aligned to the firesuppression header,

b. Separate water supplies, each with a minimum contained volume of300,000 gallons, and'

. An OPERABLE flow path capable of taking suction from the, Raw WaterTank I and Raw Mater Tank II and transferring the water through dis-tribution piping with OPERABLE sectionalizing control or isolationvalves to the yard hydrant curb valves, the last valve .ahead of thewater flow alarm device on each sprinkler or hose standpipe, and thelast valve ahead of the deluge valve on each Deluge or Spray Systemrequired to be OPERABLE per Specifications 3.7.8.2, 3.7.8.g, and3.7.8.$ . 3'-


ACTION:

a. With one pump and/or one water supply inoperable, restore the inoper-able equipment to OPERABLE status within 7 days or provide an alter-nate backup pump or supply. The provisions of Specification 3.0.3are,not applicable. This action applies to both units simultaneously.

b. With the Fire Mater Supply and Distribution System otherwise inoper-able, establish a backup fire water capability within 24 hours. Thisaction applies to both units simultaneously.


PLANT SYSTEMS


4.7.8. 1. 1 The Fire Mater Supply and Distribution System shall be demonstratedOPERASLE:

a. At least once per 7 days by verifying the contained water supplyvolume,

b. At least once per 31 days by starting the electric motor-driven pumpand operating it for at least 15 minutes on recirculation flow,

c. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position,

d. At least once per. 12 months by performance of a system flush,

e. At least once per 12 months by cycling each testable valve in the~ flow path through at least one complete cycle of full travel,

f. At least once per 18 months by performing a system functional testwhich includes simulated automatic actuation of the system throughoutits operating sequence, and:

1) Verifying that each automatic valve in the flow path actuatesto its correct position,

2)

3)

Verifying that the electric-driv p mp evelops at least 1880gpm at a system pressure of 130. sQ a d the diesel-dr en p pdevelops at least 2350 gpm at a ste p essure of 130.

ps'yclingeach valve in the flow path that is not testableplant operation through at, least one complete cycle of fulltravel, and

4) Verifying that each fire pump starts sequentiallyto maintain the Fire Mater Supply and Distribution Systempressure greater than or equal to 125 psig.

g. At least once per 3 years by performing a flow test of the system inaccordance with Chapter 5, Section 11 of the Fire Protection Handbook,14th Edition, published by the National fire Protection Association.

TURKEY POINT - UNITS 3 8E 4 3/4 7"25

PLANT SYSTEHS


4 7.8.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying:

1) The fuel storage tank contains at least 375 gallons of fuel,and

2) The diesel starts from ambient conditions and operates for atleast 30 minutes on recirculation flow.

b. At least once per 92 days by verifying that a sample of diesel fuelfrom the fuel storage tank, obtained in accordance with ASTH-0270-1975is within the acceptable limits specified in Table 1 of ASTH 0975-1977when checked for viscosity and water and sediment; and

c. At least once per 18 months by subjecting the diesel to an inspec-tion in accordance with procedures prepared in con)unction with itsmanufacturer's recommendations for the class of service.

4.7.8.1.3 The fire pump diesel starting 24-volt battery bank and chargershall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The electrolyte level of each battery is above the plates, and

2) The overall battery voltage is greater than or equal to 24 volts.

b. At least once per 92 days by verifying that the specific gravity isappropriate for continued service of the battery, and

C. At least once per 18 months by verifying that:

2)

The batteries, cell plates, and battery racks show no visualindication of physical damage or abnormal deterioration, and

The battery-to-battery and terminal connections are clean,tight, free of corrosion, and coated with &ticorrosion material.

TURKEY POINT - UNITS 3 4 4 3/4 7-26 dU> .:

PLANT SYSTEMS

SPRAY AND/OR SPRINKLER SYSTEMS


3.7.8.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:&vtLsg L{ Li gag Q C~p0etisnt COO" Q ~e 4rCcg

a ~

F'm ~et 95<~4 5'5 —Gl a~i~g V~~bb WOO~Xb.~ Fi~C 20+LL. '7RA - DO~-SO~K E~aema~C.d. FLM Bsbt& V2., 73~ '7Q ciinol '7p — E~e~en~ DieSe( Genres" And WQ l4st4

RA)&S'PPLICABILITY:Whenever equipment protected by the Spray/Sprinkler System isrequired to be OPERABLE.

ACTION:

a ~ With one or more of the above required Spray and/or Sprinkler Systemsinoperable, within 1 hour establish a continuous fire watch withbackup fire suppression equipment.&A+c +his action applies to both units

'imultaneously.

b.C



4.7.8.2 Each of the above required Spray and/or Sprinkler Systems shall bedemonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position,

b.

C.

At least once per 12 months by cycling each testable valve in theflow path through at least one complete cycle of full travel,

At least once per 18 months:

1) By performing a system functional test which includes simulatedautomatic actuation of the system, and:

a) Verifying that the automatic valves in theactuate to their correct positions on asignal, and

athtest ) g

of ful


b) Cycling each valve in the flow path that is not testableduring plant operation through at least one complete cycle

1 travel.

PLANT SYSTEMS


2) By'a visual inspection of the dry pipe spray and sprinklerheaders to verify their integrity; and

3) By a visual inspection of each nozzle's spray area to verify thespray pattern is not obstructed.

d. At.1east once per 3 years by performing an a$ r or water eeasr test ) fthrough each open head spray/sprinkler header and verifying eachopen head spray/sprinkler nozzle is unobstructed.

..TURKEY POINT - UNITS 3 4 4 3/4 7-28~ ~s ~

s ~

FIRE HOSE STATIONS


3.7.8.3 The fire. hose stations given in Table 3.7-4 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the fire hose

ACTION:

a ~

b.

With one or more of the fire hose stations given in Table 3.7-4inoperable, provide an equivalent capacity fire hose from the nearestequivalent OPERABLE water source. The fire hose shall be of a lengthof hose sufficient to provide coverage for the area left unprotectedby the inoperable hose station, and shall be stored in a roll at theoutlet of the OPERABLE water supply. The above ACTION requirementshall be accomplished within 1 hour if the inoperable fire hose isthe primary means of fire suppression; otherwise route the additionalhose within 24 hours. This action applies to both units'imultaneously

~ ~ ~ lwThe provisions of Specification 3.0.3 are not applicable.


4.7.8.3 Each of the fire hose stations given in Table 3.7-4 shall bedemonstrated OPERABLE:

a. At least once per 31 days, by a visual inspection of the fire hosestations accessible during plant operations to assure all requiredequipment is at the station.

b. At least once per 12 months, by:

1) Visual inspection of the stations not accessible during plantoperations to assure all required equipment is at the station,

2) Removing the hose for inspection and re-racking, and

3) Inspecting all gaskets and replacing any degraded gaskets inthe couplings.

C. At least once per 3 years, by:

1) Partially opening each hose(station valve to verify valveOPERABILITY and no flow blockage, and

2) Conducting a hose hydrostatic test at a pressure of 150 psig orat least 50 psig above maximum fire main operating, pressure,whichever is greater.


TABLE 3.7-4

FIRE HOSE STATIONS

IDENTIFICATION

HS-03-0HS-03-0HS-03-03HS-03-04HS-03-05HS-03-06HS-03-07HS-03-0

HS-04-0HS-04"0HS-04-03HS-04-04HS-04-0HS-04-06

HS-04-07

HS-04-08

HS-AB-0HS-AB-0HS-AB-0

HS-AB-0HS-AB"0

HS-RW-01HS-RW-02

LOCATION

EL. 18' East of 4160V SWGR Room on ColumnEL. 18' West of 3A Condensate Pump on PedestalEL. 18' Passageway South of SG Feed Pump RoomEL. 30' East of 480V Load Center on ColumnEL. 30' South End of Mezzanine DeckEL. 42' NW End of Turbine DeckEL. 42' North of 6A HPFW HeaterEL. 42' NW Corner of Entrance to Elevator

EL. 18' South of 4160V SWGR Room on ColumnEL. 18' Passageway South of SG Feed Pump RoomEL. 30' East of 480V Load Center at StairwayEL. 30' South End of Mezzanine DeckEL. 42' West End of Turbine DeckEL. 42' East Side of Turbine Deck and North of6A FW HeaterEL. 42' East Side of Turbine Deck and North of6B FW HeaterEL. 42' Southwest Corner of Turbine Deck

EL. 18' East-West Passageway at West EndEL. 18' East-West Passageway at East End ,

EL. 18' North-South Passageway Outside Unit 3Charging Pump RoomEL. 50' Roof of Unit 3 New Fuel Storage AreaEL. 50' Roof of Unit 4 New Fuel Storage Area

EL. 18' Radwaste Bldg - Main HallwayEL. 38' W of Control Room

8783

105105117117

79

8278

105105117117

117

117

585858

118118

126126


NT SYST HS

VAPO fIR HYDRANTS AND HYDRANT HOSE HOUSES

NG CONDITION FOR OPERATION

3 .7.8.4 The fire hydrants and associated hydrant hose houses givTable 3.7-5 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected b the ~f$ rehydrants is required to be OPERABLE.

ACTION:

a ~ With one or more of the fire hydrants or associated hydrant hosehouses given in Table 3.7-5 inoperable, within 1 hour have sufficientadditional lengths of 2 1/2 inch diameter hose located in an adjacentOPERABLE hydrant hose house to provide service to the unprotectedarea(s) if the inoperable fire hydrant or associated hydrant hosehouse is the primary means of fire suppression; otherwise, providethe additional hose within 24 hours.

b. The provisions of Specification 3.0.3 and are not applicable.

SURVEILLANCE RE UIRE NT

4.7.8.4 Each of the yerbs% hydrants and associated hydrant hose housesgiven in Table 3. shall b demonstrated OPERABLE:

a ~

b.

C.

At least once per 31 days, by visual inspection of the hydrant hosehouse to assure all required equipment is at the hose house,

At least once per 6 months by visually inspecting each fire hydrantand verifying that the hydrant is not damaged, and

At least once per 12 months by:

2)

Conducting a hose hydrostatic test at a pressure of 150 psig orat least 50 psig above maximum fire main operating pressure,whichever is greater,

Inspecting all the gaskets and replacing any degraded gasketsin the couplings, and

3) Performing a flow check of each hydrant to verify itsOPERABILITY.

TURKEY POINT - UNITS 3 84 4 3/4 7-31 JU" " '+

TABLE 3.7-5

FIRE HYDRANTS

FH-06

FH-07

FH-08

NA

81

IDENTIFICATION FIRE ZONE

FH-01 124

LOCATION

NE Corner of Unit 3 near Vehicle Gathinto RCA

W of Nuclear Maintenance Building

Unit 3 Transformer Area

Unit 4 Transformer Area

NUMBER OFHYDRANTS

1

FH-09

FH-10FH-11

FH"12

FH-13

FH-17

FH"16

76

77

NA

Unit 4 Turbine-Generator Area

Unit 4 Condensate Storage Tank Area

Unit 4 New Fuel Storage Area

Refueling Water Storage Area

Nuclear Ory Storage Area

Steam Generator Storage Area

TOTAL

2


A

PLANT SYSTEMS

3/4.7.9 FIRE RATED ASSEMBLIES


Qfn Affrayfer3.7.9 All fire rated assemblies (walls, floor/ceilings, wsal&e penetration se~ls, jE

d 8 I ) p tg f y- 1 dflseparating portions of redundant systems important to safe shutdown within afire area and all sealing devices in fire rated assembly penetrations (firedoors, fire Ptindows, fire dampers, cable, piping, and ventilation duct pene-tration sea~a)shall be OPERABLE.


ACTION:

as With one or more yf the above required fire rated assemblies and/orsealing devices inoperable, within 1 hour either establish acontinuous fire watch on at least one side of the affected assembly,or verify the OPERABILITY of fire detectors on at least one side ofthe inoperable assembly and establish an hourly fire watch patrol.

b. The provisions of Specification 3.0.3 are not applicable.


4.7.9.1 At least once per 18 months the above required fire rated assembliesand penetration sealing devices shall be verified OPERABLE by performing avisual inspection of:

a. The exposed surfaces of each fire rated assembly,

b. Each fire window/fire damper and associated hardware, and

Cs At least 10X of each type of sealed penetration. If apparentchanges in appearance or abnormal degradations are found, a visualinspection of an additional 10K of each type of sealed penetrationshall be made. This inspection process shall continue until a 10Ksample with no apparent changes in appearance or abnormal degradationis found. Samples shall be selected such that each penetration willbe inspected every 15 years.


PLANT SYSTEMS

SURVEILLANCE RE UIRENENTS Continued

4.7.9.2 Each of the above required fire doors shall be verified OPERABLE byinspecting the release and closing mechanism and latches at least once per6 months, and by verifying:

a; The OPERABILITY of the fire door supervision system for eachelectrically supervised fire door by performing a TRIP ACTUATINGDEVICE OPERATIONAL TEST at least once per 31 days,

b. That each locked closed fire door is closed at least once per7 days,

c. That door s with automatic hold-open and release mechanisms are freeof obstructions at least once per 24 hours, and a functional testis performed at least once per 18 months, and

d. That each unlocked fire door without electrical supervisionis closed at least once per 24 hours.


3/vg A~

oW v 4e Wi~ie> &~4

4~e ~stwa ~ +le F'.~p ~< ie ~~i o~

7~ <,i/~e rusw.dd/~4pc~A Pg) >~

~ W~~~~~8ce~re A~ >~~~~ ~ <4e

cccv<m4 ~~( ~ +pCC ~


3/4.9. 1 BORON CONCENTRATION

0e'IMITING

CONDITION FOR OPERATION

3.9.1 The boron concentration of all filled portions of the Reactor CoolantSystem and the refueling canal shall be maintained uniform and sufficient toensure that the more restrictive of. the following reactivity conditions is met;

cs5a. A Keff of JA6 or less, or

b. A boron concentration of greater than or equal to 1950 ppm.

APPLICABILITY: MODE 6."

ACTION:

With the requirements of the above specification not satisfied, immediatelysuspend all operations involving CORE ALTERATIONS or positive reactivitychanges and initiate and continue boration at greater than or equal to 10 gpmof a solution containing greater than or equal to 20,000 ppm boron or ~

K111 d d 1 1 1 1t dt-dtt tt 109Kconcentration is restored to reate~ n or equal to 1950 m whichever is

the more restrictive. gp fo e cuP ~p«i y aFone Op~g 8~In PIP&P +lg gS~i 'ow b'd~ o rt Q+Pg gag @ps>T

SURVEILLANCE RE UIREMEN

4.9. l. 1 The more restrictive of the above two reactivity conditions shall bedetermined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length control rod in excess of 3 feet fromits fully inserted position within the reactor vessel.

4.9. 1.2 The boron concentration of the Reactor Coolant System and the refuelingcanal shall be determined by chemical analysis at least once per 72 hours.

4.9. 1.3 Valves isolating unborated water sources*" shall be verified closedand secured in position by mechanical stops or by removal of air or electricalpower at least once per 31 days.

4.9.1.4 The spent fuel pit boron concentration shall be determined'at leastonce per 31 days.

adms n>strative controls

TURKEY POINT - UNITS 3 8 4!!;a~" r "i C3/4 9-1

"The reactor shall be maintained in MODE 6 whenever fuel is in the reactorvessel with the vessel head closure bolts less than fully tensioned or withthe head removed.

""The primary water supply to the boric acid blender may be opened underfor makeup.

REFUELING OPERATIONS

3/4. 9. 2 INSTRUMENTATION


3.9.2 As a minimum, one primary Source Range Neutron Flux Monitor with con-tinuous visual indication in the control room and audible indication in thecontainment and control room, and one of the remaining three Source RangeNeutron Flux Monitors (one primary or one of the two backup monitors) withcontinuous visual indication in the control room shall be OPERABLE.


ACTLON:

at

b.

With one of the above required monitors inoperable or not operating,immediately suspend all operations involving CORE ALTERATIONS orpositive reactivity changes.

With both of the above required monitors inoperable or not operating,determine the boron concentration of the Reactor Coolant System atleast once per 12 hours.


4. 9.2 Each required Source Range Neutron Flux Monitor shall be demonstratedOPERABLE by performance of:

a. A CHANNEL CHECK at least once per 12 hours,

b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initialstart of CORE ALTERATIONS, and

c. An ANALOG CHANNEL OPERATIONAL TEST at least once-per 7 days.

0TURKEY POINT - UNITS 3 8L 4 3/4 9-2


3/4. 9. 3 DECAY TIME


~ ~3.9.3 The reactor shall be subcritical for at least 100 hours.

APPLICABILITY: During movement of irradiated fuel in the reactor vessel.

ACTION:

With the reactor subcritical for less than 100 hours, suspend all operationsinvolving movement of irradiated fuel in the reactor vessel.


4.9.3 The reactor shall be determined to have been subcritical for at least100 hours by verification of the date and time of subcriticality prior tomovement of irradiated fuel in the reactor vessel.

e J



3/4 '9. 4 CONTAINMENT BUILDING PENETRATIONS

LIMITING CONDITION FOR OPERATION 03.9.4 The containment building penetrations shall be in the following status:

a. The equipment door closed and held in place by a minimum of fourbolts,

b. A minimum of one door in each airlock is closed, and

c. Each penetration providing direct access from the containmentatmosphere to the outside atmosphere shall be either:"

1) Closed by an isolation valve, blind flange, or manual valve, or

2) Be capable of being closed by an OPERABLE automatic containmentventilation isolation valve.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within

~

~

~

~ ~

~

ACTION:

With the requirements of the above specification not satisfied, immediatelysuspend all operations involving CORE ALTERATIONS or movement of irradiatedfuel in the containment building.


4.9.4 Each of the above required containment building penetrations shall bedetermined to be either in its closed/isolated condition or capable of beingclosed by an OPERABLE automatic containment ventilation isolation valve within100 hours prior to the start of and at least once per 7 days during COREALTERATIONS or .movement of irradiated fuel in the containment building by:

'a 4 Verifying the penetrations are in their closed/isolated condition,or

b. Testing the containment ventilation isolation valves per the appli-cable portions of Specification 4.6.4.2.

"Exception may be taken under Administrative Controls for opening of certainvalves and airlocks necessary to perform. surveillance or testing requirements.

'l

:TURKEY POINT - UNITS 3 8 4 3/4 9-4


3/4.9. 5 COMMUNICATIONS


~ ~

~3.9.5 Direct communications shall be maintained between the control room andpersonnel at the refueling station.

APPLICABILITY: During CORE ALTERATIONS.

ACTION:

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


4.9.5 Direct communications between the control room and personnel at therefueling station shall be demonstrated within 1 hour prior to the start ofand at least once per 12 hours during CORE ALTERATIONS.

TURKEY POINT - UNITS 3 8s 4 3/4 9-5


3/4. 9. 6 MANIPULATOR CRANE


3.9.6 The manipulato~ crane and auxiliary hoist shall be used for movement ofdrive rods or fuel assemblies and shall be OPERABLE with:

a. The manipulator crane used for movement of fuel assemblies having:

1) A minimum capacity of 2750 pounds, and

2) An overload cutoff limit less than or equal to 2700 pounds.

b. The auxiliary hoist used for latching and unlatching drive rodshaving:

1) A minimum capacity of 610 pounds, and

2) A load indicator which shall be used to prevent lifting loadsin excess of 600 pounds.

APPLICABILITY: During movement of drive rods or fuel assemblies withinthe reactor vessel.

With the requirements for crane and/or hoist OPERABILITY not satisfsed, suspenduse of any inoperable manipulator crane and/or auxiliary hoist from operationsinvolving the movement of drive rods and fuel assemblies within the reactorvessel.


4.9.6.1 At least once each refueling, each manipulator crane used for movementof fuel assemblies within the reactor vessel shall be demonstrated OPERABLEwithin 100 hours prior to the start of such operations by performing a loadtest of at least 2750 pounds and demonstrating an automatic load cutoff whenthe crane load exceeds 2700 pounds.

4.9.6.2 At least once each refueling, each auxiliary hoist and associated'oadindicator used for movement of drive rods within the reactor vessel shall bedemonstrated OPERABLE within 100 hours prior to the start of such operations byperforming a load test of at least 610 pounds.



3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS

3.9.7 Loads in excess of 2000 pounds shall be prohibited from travel overfuel assemblies in the storage pool."

APPLICABILITY: With fuel assemblies in the storage pool.

ACTION:

a. With the requirements of the above specification not satisfied, placethe crane load in a safe condition.



4.9.7 Prior to crane operation over fuel assemblies in the spent fuel storagepool, verify that each load is 2000 pounds or less.

*Exception may be taken for the temporary construction crane. to be used forthe re-rack operation which may be carried over irradiated fuel to facilitateinstallation of the crane. Lift rigs which meet the design and operationalrequirements of NUREG-0612 "Control of Heavy Loads at Nuclear Power Plants"will be used while performing this installation.



3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION

HIGH WATER LEVEL

LIMITING CONDITION FOR OPERATION0

3.9.8. 1 At least one residual heat removal (RHR) loop shall be OPERABLE andin operation."

APPLICABILITY: MODE 6, when the water level above the top of the reactor2 1 2 2 1 2 1 t.

ACTION:

With no RHR loop OPERABLE and in operation, suspend all operations involving'n increase in the reactor decay heat load or a reduction in boron concentrationof the Reactor Coolant System and immediately initiate corrective action toreturn the required RHR loop to OPERABLE and operating status as soon aspossible. Pose all contalnment penetratlons provfdlng direct access fromthe con ainment atmos here to the outside atmosphere within 4 hours. 4l


4.9.8. 1. 1 At least one RHR loop shall be verified in operation and circulatingreactor coolant at a flow rate of greater than or equal to 3000 gpm at leastonce per 12 hours.

4.9.8. 1.2 The RHR flow indicator shall be subjected to a CHANNEL CALIBRATIONat least once per refuelingi hog f~ <<+~+ Z4. ~Qs

"The RHR loop may be removed from operation for up to 1 hour per 8-hour periodduring the performance of CORE ALTERATIONS in the vicinity of the reactor vesselhot legs.



LOW WATER LEVEL


3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE,and at least one RHR loop shall be in operation.

APPLICABILITY: MODE 6, when the water level above the top of the reactorvessel flange is less than 23 feet.

ACTION:

a ~ With less than the required RHR loops OPERABLE, immediately initiatecorrective action to return the required RHR loops to OPERABLEstatus, or to establish greater than or equal to 23 feet of waterabove the reactor vessel flange, as soon as possible.

b. With no RHR loop in operation, suspend all operations involving areduction in boron concentration of the Reactor Coolant System andimmediately initiat corrective action to return the required RHR

loop to operation. goose all containment penetrations providingdirect access from the containment atmosphere to the outsideatmosphere within 4 s.

+a

SURVEILLANCE RE UIREH

4.9.8.2 At least one RHR loop shall be verified in operation and circulatingreactor coolant at a flow rate of greater than or equal to 3000 gpm at leastonce per 12 hours.

TURKEY POINT - UNITS 3 8L 4 3/4 9"9

)eggs+


3/4. 9. 9 CONTAINMENT VENTILATION ISOLATION SYSTEM


3.9.9 The Containment Ventilation Isolation System shall be OPERABLE.


ACTION:

'a ~

b.

With the Containment Ventilation Isolation System inoperable, closeeach of the containment ventilation penetrations providing directaccess from the containment atmosphere to the outside atmosphere.



(~bio~4.9.9 The Containment Ventil tiongystem s be demonstrated OPERABLE within100 hours prior to the start o and t 1 once per 7 days during COREALTERATIONS by. verifying that C n nt Ventilation Isolation occurs on aHigh g gi i « ig i i tghght t i gi i i i ginstrumentation channel/.

+c rs~ug ~ Oi EMfSL



3/4.9.10 WATER LEVEL - REACTOR VESSEL


3.9.10 't least 23 feet of water shall be maintained over the top of thereactor vessel flange.

APPLICABILITY: Ouring movement of fuel assemblies or control rods within theconta>nment when either the fuel assemblies being moved or the fuel assembliesseated within the reactor vessel are irradiated while in MODE 6.

ACTION:

With the requirements of the above specification not satisfied, suspend alloperations involving movement of fuel assemblies or control rods within thereactor vessel.

e SURVEILLANCE RE UIREHENTS

4.9. 10 The water level shall be determined to be at least its minimum requireddepth within 2 hours prior to the start of and at least once per 24 hoursthereafter during movement of fuel assemblies or control rods.

TURKEY POINT - UNITS 3 4 4 3/4 9-11 II ~

~ I pe


3/4. 9. 11 WATER LEVEL - STORAGE POOL


s

3.9.11 The water level shall be maintainp4 ' " in Plthe sPent fuel stonege Pool.*" gr~Tf, gJ;„o„~j ro e/s~„+o

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

as With the requirements of the above specification not satisfied,suspend all movement of fuel assemblies and crane operations withloads in the fuel. storage areas and restore the water level to withinits limit within 4 hours.



4.9. 11 The water level in the storage pool shall be determined to be at leastits minimum required depth at least once per 7 days when irradiated fuelassemblies are in the fuel storage pool.

"During spent fuel rerack operation, the water level shall be maintained .atleast at 49'0" elevatio h i no movement of fuel assemblies with

) 1 t 4 ill *k p i . +i'iO'f"*The requirements of t s speci ica ion may be suspended for more than 4 hours

hours to perform maintenance provided a safety evaluation is prepared prior tosuspension of the above requirement and all movement of fuel assemblies andcrane operation with loads in the fuel storage areas are suspended. If thelevel is not restored within 7 days, the NRC shall be notified within the next24 hours.

's

TURKEY POINT - UNITS 3 4 4 3/4 9-12tsstf

r p,


3I4.9.12 HANOLING OF SPENT FUEL CASK


~ ~~ ~

3.9.12 The handling of spent fuel cask shall be limited to the followingconditions:

1) The spent fuel cask shall not be moved into the spent fuel pit untilall the spent fuel in the pit has decayed for a minimum of onethousand five hundred twenty-five (1,525) hours.*

2) Only a single element cask may be moved into the spent fuel pit.') A fuel assembly shall not be removed from the spent fuel pit in a

shipping cask until it has decayed for a minimum of one hundredtwenty (120) days.

APPLICABILITY: Ouring movement of spent fuel cask in the spent fuel storagearea.

ACTION:

Mith the requirement of the above specification not satisfied, suspend allmovement of the spent fuel cask within the spent fuel storage area.

SURVEIlLANCE RE UIREMENTS

4.9. 12.1 The following required decay times of the spent fuel assembliesshall be determined prior to the movement of a spent fuel cask by verificationof date and time th ~ mott ~@+pl diSC,go~eclgi e.( ~g e T'iiel iT ~ la=T'ce',4cai'.

1525 hours of decay o a spen ue assem s snpit for movement of a spent fuel cask into the s t fuel pit.

I @LE lSb. 120 days of decay of the spent fuel assemb > e spent fuel cask

prior to removal of the spent fuel cask from the spent fuel pit.

4.9.12.2 Prior to any operations involving spent fuel cask movement into thespent fuel pit, verify only a single element cask will be moved into the spentfuel pit.4.8. 12.3 The spent fuel cask crane interlock shall bewithin 7 days of crane operation and at least once pertime between tests; specification 4.0.2 does not applybeing used to maneuver the spent fuel cask.

demonstrated OPERABLE7 days (7 days is maximumhere) when the crane is

"The spent fuel cask can be moved into the Unit 4 spent fuel pit after aminimum decay of 1000 hours until the new two-region high density spent fuelracks are installed.

e

TURKEY POINT - UNITS 3 4 4 3/4 9-13 ) 1

, 4 ~(


3/4. 9 e 13 RADIATION MONITORING

LIMITING CONDITION„FOR OPERATION

Qha.3.9. 13 ~ C ntainment Radiation monitor which initiatencontainment andcontro oom ventilation isolation shall be OPERABLE.


ACTION:

a)

b)

AOIh ~r i i < p b1, i" Ytinue provided the containment ventilat solation valves are maintained closed> ~~

Dfc. Pvi( c:~QO

With radiation monitors erabl, within 1 hour isolatthe Contro m t 1 tion System and n ate operation of theControl Room Ventilation System in the recirculation mode.

Einmr

SURVEILLANC E UIREMENTS

Om ietre 'OL 5 CCI3LCC4'LOw< g 1 ~ ~

4.9. 13 EsafyLContainment Radiation monitor e monetra e LE by ) +the performance of the CHANNEL CHECK, CHAN~tlEL CALIBRATION and ANALOG CHANNEL~OPERATIONAL TEST at the frequencies shown in Table 4.3-3.



3/4.9.14 SPENT FUEL STORAGE

~ ~


3.9.14 The following conditions shall apply to spent fuel storage:

a. Fuel assemblies containing more than 4. 1 weight percent of U-235shall not be placed in the single region spent fuel storage racks.After installation of the two-region high density spent fuel racks,the maximum enrichment loading for the fuel assemblies in the spentfuel racks shall be 4.5 weight percent of U-235.

b. The minimum boron concentration inI the Spent Fuel Pit shall be1950 ppm.

c.* Storage in Region II of the Spent Fuel Pit shall be further restrictedby burnup and enrichment limits specified in Table 3.9-1.

d.* Ouring the re-racking operation only, fuel that does not meet theburnup requirement for normal. storage in Region II may be stored inRegion II in a checkerboard arrangement (i.e., no fuel stored inadjacent spaces).

APPLICABILITY: At all times when fuel is stored in the Spent Fuel Pit.1

ACTION:

a. With any of conditions a, c or d not satisfied, suspend movement ofadditional fuel assemblies into the Spent Fuel Pit and restore thespent fuel storage configuration to within the specified conditions.

b. With bo~on concentration in the Spent Fuel Pit less than 1950 ppm,suspend movement of spent fuel in the Spent Fuel Pit and initiateaction to restore boron concentration to 1950 ppm or greater.


4.9.14 The boron concentration of the Spent Fuel Pit shall be verified to be1950 ppm or greater at least once per month.

"These requirements are applicable only after installation of the new two-' ~region high density spent fuel racks.


TABLE 3.9-1

SPENT FUEL BURNUP RE UIREHENTS FOR STORAGEIN GION II 0 THE S N UL

Initialw/o

1.5

1. 75

2.0-

2.2

2.4

2.6

2.8

3.0

3.2

3.4

3.6

3.8

4.0

4.2

4.5

Oischarge BurnupGWD/HT

0.

5.0

9.0

12. 0

14. 8

17. 6

20. 1

22. 6

25. 0

27. 4

29. 6

. 31.8

34. 0

36. 1

39. 0

Linear interpolation between twoconsecutive points will yield

conservative results.



3/4. 10. 1 SHUTDOWN MARGIN

LIMITING CONDITI'ON FOR OPERATION


ACTION:

With any full-length control rod not fully inserted and with lessthan the above reactivity equivalent available for trip insertion,immediately initiate and continue boration at greater than or equalto 10 gpm of a solution containing greater than or equal to20,000 ppm boron until the GNUTOONN MARGIN requiredby Specification 3. 1. 1. 1 is restored.

With all full-length control rods fully inserted and the reactorsubcritical by less than the above'reactivity equivalent,

immedi-'telyinitiate and continue boration at greater than or equal to10 gpm of a solution containing greater than or equal to 20,000 ppmboron or its equivalent until the SHUTDOWN MARGIN re uired bSpecification 3.1.1.1 'st d.

As(.~ c,n ~"< «. ~ '~'> ~ K'g "pu+ uiW wl H

6L,& ~'T

b.


3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may besuspended for measurement of control rod worth and SHUTDOWN MARGIN providedreactivity equivalent to at least the highest estimated control rod worth isavailable for trip insertion from OPERABLE control rod(s}.

4.10.1.1 The position of each full-length control rod either partially orfully withdrawn shall be determined at least once per 2 hours.

4. 10. 1.2 Each full-length control rod not fully inserted shall be demonstratedcapable of full insertion when tripped from at least the 50K withdrawn positionwithin 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits ofSpecification 3.1.1.1.

r


SPECIAL TEST EXCEPTIONS

3/4.10. 2 GROUP HEIGHT INSERTION AND POWER DISTRIBUTION LIMITS


3. 10.2 The group height, insertion, and power distribution limits ofSpecifications 3. 1.3. 1, 3.1.3.5, 3. 1.3.6, 3.2. 1, and 3.2.4 may be suspendedduring the performance of PHYSICS TESTS provided:,

a. The THERMAL POWER is maintained less than or equal to BSX of RATEDTHERMAL POWER, and

b. The limits of Specifications 3.2.2 and 3.2.3 are maintainedand determined at the frequencies specified in Specification4.10.2.2 below.

APPLICABILITY: MODE l.ACTION:

With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded whilethe requirements of Specifications 3. 1. 3. 1; 3. 1. 3.5, 3.1.3. 6, 3.2.1, and 3. 2.4are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirementsof Specifications 3.2.2 and 3.2.3, or

b. Be in HOT STANDBY within 6 hours.


4. 10.2. 1 The THERMAL POWER shall be determined to be less than or equal to85X of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 The requirements of the below listed specifications shall be performedat least once per 12 hours during PHYSICS TESTS:

a. Specifications 4.2.2.1 and

b. Spec<1<catkon 4.2.3P $

TURKEY POINT -.UNITS 3 4 4 3/4 10-2


3/4. 10. 3 PHYSICS TESTS


3. 10.3 The limitations of Specifications 3. l. 1.3, 3. l. 1.4, 3. 1.3. 1, 3. 1.3.5,and 3. 1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5X of RATED THERMAL POWER,

b.

C.

The Reactor Trip Setpoints on the OPERABLE Intermediate and PowerRange channels are set at less than or equal to 25X of RATEDTHERMAL POWER, and g~The Reactor Coolant System lowe trloperating loop temperature (T )is greater than or equal to 531 F. avg

ACTION:



r -1-'-'SR'/L +war R Lelosd S Po~>+ ~A4cca. With the T R L POWER greater than 5X of RATED THERMAL POWER,

immediately open the Reactor tri br kers.avero e.

b. With a Reactor Coolant System~ p r ng loop temperature ~-+less than 5314F, restore T—to within its limit within15 minutes or be in a least HOT STANDBY within the next15 minutes.

RGS aueca~e Pe pero~ure

)~

1

4. 10.3.1 The THERMAL POWER shall be determined to be. less than or equal to 5Xof RATED THERMAL POWER at least once per hour du~ing PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range channel shall be subjected to anANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICSTESTS.

aaen ae-4. 10.3.3 The Reactor Coolant System temp rature gT—+ shall be determined tobe greater than or equal to 5314F at least once per 30 minutes during PHYSICSTESTS.

TURKEY POINT - UNITS 3 4 4 3/4 10-3 l C,.5


3/4. 10: is s e 'rati number 'ot us



3/4. 10. 5 POSITION INDICATION SYSTEM - SHUTDOWN


3.10+ The limitations of Specification 3.1.3.3 may be suspended during the I Eperformance of individual full-length shutdown and control rod drop timemeasurements pr vi d;

tuZ) 4 3'8

a. Onl ene.sh tdown or control b kale,wi hdrawn fram the fu]]y inserted idf=ipos tion a a time, and

b. The rod position indicator is OPERABLE during the withdrawal of therods.

APPLICABILITY: HOOES 3, 4, and 5 during performance of rod drop time measurements.

ACT!OM:

TiesWith the Position Indication Systems inoperable or with more tha eve.rods withdrawn, immediately open the Reactor trip breakers.


4. 10.5 The above required Position Indication Systems shall be determinedto be OPERABLE within 24 hours prior to the start of and at least once per24 hours thereafter during rod drop time measurements by verifying the DemandPosition Indication System and the Digital Rod Position Indication Systemagree;

a. Within 12 steps when the rods are stationary, and

b. Within 24 steps during rod motion.

3/4 10-5

/r1/4~/rc'n el~


3/4.11.1 LI UID EFFLUENTS

CONCENTRATION


3.11.1.1 The concentration of radioactive material released in liquid effluentsto UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the concentrationsspecified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclidesother than dissolved or entrained noble gases. For dissolved or. entrainednoble gases, the concentration shall be limited to 2 x 10-~ microCurie/mltotal activity.


ACTION:

With the concentration of radioactive material released in liquid effluents toUNRESTRICTED AREAS exceeding the above limits, immediately restore the concen-tration to within the above limits.


4. 11.1.1.1 Radioactive liquid wastes shall be sampled and,analyzed accordingto the sampling and analysis program of Table 4.11-1.

4. 11.1.1.2 The results of the radioactivity analyses shall be used in accordancewith the methodology and parameters in the ODCH to assure that the concentrationsat the point of release are maintained within the limits of Specification3. 11. 1. 1.

I


TABLE 4.11-1

RADIOACTIVE LI UID WASTE SAMPLING AND ANALYSIS PROGRAM

LIQUID RELEASETYPE

1. Batch WasteRelease

Tanks

SAMPLING'REQUENCY

P

Each Batch

MINIMUM

ANALYSISFREQUENCY

P

Each Batch

TYPE OF ACTIVITYANALYSIS

Principal Gaiiima

EmittersI-131

LOWER LIMITOF DETECTIO

(LLD)(1)(pCi/ml)

Sxl0-7

lxl0-s

P

One Batch/M

P

Each BatchM

Composite

Dissolved andEntrained Gases(Gamma Emitters)

H-3

Gross Alpha

lxl0-s

lx10-s

lx10-7

2. Continuous

Releases

a. SteamGener atop)Blowdown

('.~'4 eDrain

P

Each Batch

M(8)

W(8)

W(8)

Q 4Composite(")

M(8)

(')Composite

Q(8)Composite

Sr-89, Sr-90

Fe-55

Principal Gamma

Emitters

I-131

Dissolved andEntrained Gases(Gamma Emitters)

Gross Alpha

.Sr-89, Sr-90

Fe-55

Principal Gaaea

Emitters

I-131

5x10-s

Ix10-s

5x10-7

lx10-s

lxl0-s

lx10-s

lx10-7

5x10 s

lx10-s

5xlO-~

1x10-s

jp

- TURKEY POINT - UNITS 3 Ec 4 3/4 11-2

TABLE 4.11" 1 continued

TABLE NOTATIONS

(1) The LLD is the smallest concentration of radioactive material in a sample0

that will be detected with 95K probability with only SX probability offalsely concluding that a blank observation represents a "real" signal.

for a particular measurement system, which may include radiochemicalseparation:

4.66LLD =

~ V ~ .2 x 1 ~ ~ exp -UtWhere:

LLD' the "a priori" lower limit of detection as defined above

for a blank sample (microCurie per unit mass or volume),

sb = the standard deviation of the background counting rateor of the counting rate of a blank sample as appropriate(counts per minute),

.E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or volume),

2.22 x 10' the number of disintegrations per minute per microCurie,

Y = the fractional radiochemical yield, when applicable,

the radioactive decay constant for the particularradionuclide, and

Lt = the elapsed time between the midpoint of sample collectionand the time of counting (for plant effluents, notenvironmental samples).

The value of sb used in the calculation of the LLD for a detection system

shall be based on the actual observed variance of the background countingrate or of the counting rate of the blank samples (as appropriate) ratherthan on an unverified theoretically predicted variance. Typical values ofE, V, Y, and ht should be used in the calculation.

A batch release is the discharge of liquid wastes of a discrete volume.(2)Prior to sampling for analyses, each batch shall be isolated, and thenthoroughly mixed by a method described in the ODCM to assure representa-tive sampling.




The principal gamma emitters for which the LLO specification exclusivelyapplies are the following radionuclides: Hn-54, Fe-59, Co-58, Co-60,Zn-65, Ho-99,'s-134, Cs-137, Ce-141, and Ce-144. This list does not meanthat only these nuclides are to be considered. Other gamma peaks that areidentifiable, together with those of the above nuclides, shall also beanalyzed and reported in the Semiannual Radioactive Effluent ReleaseReport pursuant to Specification 6.9.1.4.

A composite sample is one in which the quantity of liquid sampled isproportional to the quantity of liquid waste discharged and in which themethod of sampling employed results in a specimen that is representativeof the liquids released.

A continuous release is the discharge of liquid wastes of a nondiscretevolume, e.g., from a volume of a system that has an input flow during thecontinuous release.

C'rior to analyses, all samples taken for the composite shall be thoroughly(6)mixed in order for the composite sample to be representative of theeffluent release.

Sampling and analysis of steam generator blowdown is not required duringMode 5 or 6.

Sampling and analysis of steam generator blowdown on the applicable unitis only necessary for these species when primary t con y leakage isoccurring as indicated by the condenser air e$ ec r nile s activitymonitor. (See Specification 3.3.3.7 in Table 3. -8, Stem 3 ).


RADIOACTIVE EFFLUENTS

DOSE

LIMITING CONDITION 'FOR OPERATION

3. 11. 1.2 The dose or dose commitment to a HEHBER OF THE PUBLIC from radioactive~ ~~

~

~materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS(see Figure 5. 1-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems tothe whole body and to less than or equal to 5 mrems to any organ,and

b. During any calendar year to less than or equal to 3 mrems to thewhole body and to less than or equal to 10 mrems to any organ.


. ACTION:

With the calculated dose from the release of radioactive materialsin liquid effluents exceeding any. of the above limits, prepareand submit to the Commission within 30 days, pursuant to Specification6.9.2, a Special Report that identifies the cause(s) for exceedingthe limit(s) and defines the corrective actions that have been takento reduce the releases and the proposed corrective actions to betaken to assure that subsequent releases will be in compliance withthe above limits.



4.11.1.2 Cumulative dose contributions from liquid effluents for the currentcalendar quarter and the current calendar year shall be determined in accordancewith the methodology and parameters in the ODCM at least once per 31 days.

TURKEY POINT - UNITS 3 8( 4 3/4 11-5


LI UID RAOWASTE TREATMENT SYSTEM


3.11. 1.3 The Liquid Radwaste Treatment System shall be OPERABLE and appropriateportions of the system shall be used to reduce releases of radioactivity whenthe projected doses due to the liquid effluent, from each unit, to UNRESTRICTEDAREAS (see Figure 5. 1-1) would exceed 0.06 mrem to the whole body or 0.2 mremto any organ in a 31-day period.

I


ACTION:

a. With radioactive liquid waste being discharged without treatment and .

in excess of the above limits and any portion of the Liquid RadwasteTreatment System not in operation, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6. 9.2, a Special Reportthat includes the following information:

l. Explanation of why liquid radwaste was being discharged withouttreatment, identification of any inoperable equipment orsubsystems, and the reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLEstatus, and

3. Summary description of action(s) taken to prevent a recurrence.



4. 11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREASshall be projected at least once per 31 days in accordance with the methodologyand parameters in the ODCM when Liquid Radwaste Treatment Systems are not beingfully utilized.

4. 11.1.3.2 The installed Liquid Radwaste Treatment System shall beconsidered OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2.

TURKEY POINT - UNITS 3 8( 4 3/4 11-6



DOSE RATE


~ ~ ~3. 11.2. 1 The dose rate due to radioactive materials released in gaseouseffluents from the site to areas at and beyond the SITE BOUNDARY (see Figure5. 1-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrems/yr to the wholebody and less than or equal to 3000 mrems/yr to the skin, and

b. For Iodine-131, for Iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days:Less than or equal to 1500 mrems/yr to any organ.

APPLICABILITY:,At all times.

ACTION:

With the dose rate(s) exceeding the above limits, immediately restore therelease rate to within the above limit(s).


~ ~ ~~ ~

~

4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall bedetermined to be within the above limits in accordance with the methodologyand parameters in the ODCM.

4.11.2.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and allradionuclides in particulate form with half-lives greater than 8 days ingaseous effluents shall be determined to be within the above limits inaccordance with the methodology and parameters in the ODCM by obtainingrepresentative samples and performing analyses in accordance with the samplingand analysis program specified in Table 4.11-2.

3/4 11-7

TABL

RADIOACTIVE GASEOUS AS SAMPLI AND ANALYSIS PROGRAM

WI

CO

~ ~g O

o~ CQA

8~ ~Cgg

J(b

~tP

GASEOUS RELEASE TYPE

1. Gas DecayTank (Batch)

2. Containment PurgeBatch)

3. Condenser AirEjectors

lant Vent (Tp nt vent opor e nit 4Sp 1 Pit

ilding qt.)5. Unit 3 Spent Fuel

Pit BuildingVent

6. All Release Typesas listed in 3.,4., and 5. above

SAMPLINGFREqUENCY

P

Each TankGrab SaN le

P6

Grab Salple

HGrab Saiiple

1Grab S leM(4),(5)Grab Saaple.

HGrab Sa le

Grab S leContinuous

Continuous

Continuous

Continuous

Cont'ious

MIN M

ANAL SISFREQU CY

P

Each Tank

P

Each PURGE

Gas Sample

M

Gas Sa le

HGas S le

CharcoalS le

ParticulateS le

M

Coiiiposite Par-ticulate S 1

Coliposite Par-ticulate Sam 1

Noble GasMonitor

E OFACTIVI ANALYSIS

ncipal Gawaa Emiitters

Princi al Gaeaa EoittersH-3

Principal Gaaea Eilitters

H-3

Principal Gaaaa Eiaitters

H-3

Principyl aiaaa Eeitters

H-3

I-131

Principal Gama EIiitters

Gross Alpha

Sr-89, Sr-90

le Gass Beta or Gamma

LOWER LIMI 0DETECTION (LLD)

(pCi/cc)

1x10-

1x10-~

1xlO-6

x10 B

1 iO-~

1x10-6

1x10-i

1x10-~~

1x]O-»

1x10-»

1x10-6n


TABLE NOTATIONS

(1}The LLD is the smallest concentration of radioactive material in a samplethat will be detected with 95X probability with only 5X probability offalsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemicalseparation:

LLD4.66 sb

E ' (2.22 x 10 ) ~ Y . [exp (-Mt)]Where:

LLD = the "a priori" lower limit of detection as defined aboveas a blank sample (microCurie per unit mass or volume),

sb = the standard deviation of the background counting rate orof the counting rate of a blank sample as appropriate(counts per minute},

E = the counting efficiency (counts per disintegration)

V = the sample size (units of mass or volume),

2.22 x 10' the number of disintegrations per minute per microCurie,

Y = the fractional radiochemical yield, when applicable,

the radioactive decay constant for the particularradionuclide, and

ht = the elapsed time between the midpoint of sample collectionand the time of counting (for plant effluents, notenvironmental samples)

The value of sb used in the calculation of the LLD for a detection system

shall be based on the actual observed variance of the background countingrate or of the counting rate of the blank samples (as. appropriate) ratherthan on an unverified theoretically predicted variance. Typical valuesof E, V, Y and ht shall be used in the calculation.




The principal gamma emitters for which the LLD specification will apply(2)are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133,Xe-133m, Xe-135, and Xe-138 in noble gas emissions and Mn-54, Fe-59,Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 infor particulate emissions. This list does not mean that only thesenuclides are to be detected and reported. Other ganna peaks that aremeasurable and identifiable, together with the above nuclides, shall alsobe identified and reported pursuant to Specification 6.9.1.4.

Nuclides which are below the LLD for the analyses should not be reportedas being present at the LLD for that nuclide. When a radionuclide'scalculated LLD is greater than its listed LLD limit, the calculated LLDshould be assigned as the activity of the radionuclide; or, the activityof the radionuclide should be calculated using measured ratios with thoseradionuclides which are routinely identified and measured.

The ratio of the sample flow rate to the sampled stream flow rate shall beknown for the time period covered by each dose or dose rate calculationmade in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.

When a Unit's refueling canal is flooded Tritium grab samples shall betaken on that Unit only from the following respective area(s) at leastonce per. 24 hours:

For Unit 3 sample the plant vent and the Unit 3 spent fuel pool areaventilation exhaust.

For Unit 4 sample the plant vent only.

When spent f 1 is in Unit's spent fuel pool, tritium grab samples shallbe taken o at~only rom the following respective area at least once per7 days: pviif

For Unit 3, s the Unit 3 spe 1 pool area ventilation exhaust

For Unit 4, sample the p lant v'nt.mnli/,C.

Sampling and analysis shall also performed following shutdown, startup,or a THERMAL POWER change exceeding 15K of RATED THERMAL POWER within a1-hour period if (1) analysis shows that the DOSE E)UIVALENT I-131 con-centration in the primary coolant has increased by moro than a factor of3; (2) the noble gas activity monitor shows that effluent activity hasi creased by more than a factor of 3. 4Q

Ay)g ~




(7) Sample collection media on the applicable Unit shall be changed at leastonce per 7 days and analyses shall be completed within 48 hours afterchanging, or after removal from sampler. Sample collection media on theapplicable Unit shall also be changed at least once per 24 hours for atleast 7 days following each shutdown, startup, or THERMAL POWER changeexceeding 15X of RATED THERMAL POWER within a I-hour period and analysesshal be completed within 48 hours of changinaP

if: (1) analysis shows that the DOSE EglJIVALENT I-131 concentrationin the primary coolant has ~increased more than a factor of 3; and (2)the noble gas monitor shows that effluent activity has ~increased morethan a factor of 3. When samples collected for 24 hours are analyzed, thecorresponding LLQs may be increased by a factor of 10.



DOSE - NOBLE GASES


3.11.2.2 The air dose due to noble gases released in gaseous effluents, fromeach unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shallbe limited to the following:

a. During any calendar quarter: Less than or equal to 5" mrads forgamma radiation and less than or equal to 10 mrads for beta radiation,and

b. During any calendar year: Less than or equal to 10 mrads for gammaradiation and less than or equal to 20 mrads for beta radiation.


ACTION

With the calculated air dose from radioactive noble gases in gaseouseffluents exceeding any of the above limits, prepare and submit tothe Commission within 30 days, pursuant to Specification 6.9.2, aSpecial Report that identifies the cause(s) for exceeding the limit(s)and defines the corrective actions that have been taken to reducethe releases and the proposed corrective actions to be taken toassure that subsequent releases will be in compliance with the abovelimits.

b. The pr ov i s ions of Speci ficati on 3.0. 3 are not app1 icab 1 e.


4. 11.2.2 Cumulative dose contributions for the current calendar quarter andcurrent calendar year for noble gases shall be determined in accordance withthe methodology and parameters in the ODCM at least once per 31 days.



DOSE - IODINE-131 IODINE-133 TRITIUM AND RADIOACTIVE MATERIAL INPARTICULATE FORM


3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133,tritium, and all radionuclides in particulate form with half-lives greaterthan 8 days in gaseous effluents released, from each unit, to areas at andbeyond the SITE BOUNDARY (see Figure 5. 1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to anyorgan and,

b. During any calendar year: Less than or equal to 15 mrems to anyorgan.


ACTION:

aO With the calculated dose from the release of Iodine-131, Iodine-133,tritium, and radionuclides in particulate form with half-livesgreater than 8 days, in gaseous effluents exceeding any of'he abovelimits, prepare and submit to the Commission within 30 days, pursuantto Specification 6.9.2, a Special Report that identifies the cause(s)for exceeding the limit(s) and defines the corrective actions that havebeen taken to reduce the releases and the proposed corrective actionsto be taken to assure that subsequent releases will be in compliancewith the above limits.



4.11.2.3 Cumulative dose contributions for the current calendar quarter andcurrent calendar year for Iodine-131, Iodine-133, tritium and radionuclidesin particulate form with half-lives greater than 8 days shall be determinedin accordance with the methodology and parameters in the ODCM at least onceper 31 days.



GASEOUS RADWASTE TREATMENT SYSTEM


3. 11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and the GAS DECAY TANKSYSTEM shall be OPERABLE and appropriate portions of these systems shall beused to reduce releases of radioactivity when the projected doses in 31 daysdue to gaseous effluent releases, from each unit, to areas at and beyond theSITE BOUNDARY (see Figure 5. 1-1) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.


ACTION:

With radioactive gaseous waste being discharged without treatmentand in excess of the above limits, prepare and submit to theCommission within 30 days, pursuant to Specification 6.9.2, aSpecial Report that includes the following information:

l. Identification of any inoperable equipment or subsystems, andthe reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLEstatus, and

3. Summary description of action(s) taken to prevent a recurrence.



4.11.2.4.1 Doses due to gaseous releases from each unit to areas at andbeyond the SITE BOUNDARY shall be projected at least once per 31 days inaccordance with the methodology and parameters in the ODCM when GaseousRadwaste Treatment Systems are not being fully utilized.

4.11.2.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and GAS DECAYTANK SYSTEM shall be considered OPERABLE by meeting Specifications 3. 11.2.1 and.3.11.2.2 or 3.11.2.3.

,'TURKEY POINT - UNITS 3 4 4 3/4 11-14


EXPLOSIVE GAS MIXTURE


3. 11. 2. 5 The concpntration of oxygen in the GAS DECAY TANK SYSTEM (as measuredin the inservice /fang'ecay fank) shall be limited to less than or equal to 2%

) gby volume whenever the hydrogen concentration exceeds 4X by volume.


ACTION:

a ~

b.

C.

With the concentration of oxygen in the inservicepAS CAY JANKgreater than 2X by volume but less than or equal to 4 by volume,reduce the oxygen concentration to the above limits within 48 hours.

With the concentration of oxygen in the inservicepAS JECAYQANK.greater than 4X by volume and the hydrogen concentration greaterthan 4X by volume, immediately suspend all additions of waste gasesto the gas J[ecay inks and reduce the concentration of oxygen toless than or equal to 4S by volume, then take ACTION a., above.

The provisions of Specification 3.0:3 are not applicable.


4.11.2.5 The concentrations of hydrogen and oxygen in the inservicegAS/ECAYfANKS shall be determined to be within the above limits by continuouslyW ')mZmonitoring the waste gases in the PP)P]kgb/ g+ SYSTEM with the hydrogen )db )and oxygen monitors required OPE HLE by'able 3.3-8 of Specification3.3.3.7. fg-5cI AGK

~lsd I,8 is sw orner ~~~gCaw[ llnudug Ivio>~'"J p

T $g ~ ~ g ~ll~s We Usc of

TURKEY POINT - UNITS 3 8[ 4 3/4 11-15


GAS DECAY TANKS


3. 11.2.6 The quantity of radioactivity contained in each gas decay tankshall be limited to less than or equal to 70,000 Curies of noble gases (con-sidered as Xe-133 equivalent).


ACTION:

With the quantity of radioactive material in any gas decay tankexceeding the above limit, immediately suspend all additions ofradioactive material to the tank, within 48 hours reduce the tankconte'nts to within the limit, and describe the events leading to thiscondition in the next Semiannual Radioactive Effluent Release Report,pursuant to Specification 6.9.1.4.



~ ~ ~4. 11.2.6 The quantity of radioactive material contained in each gas decaytank shall be determined to be within the above limit at least once per 24hours when radioactive materials are being added to the tank and the ReactorCoolant System total activity exceeds the limit of Specification 3.4.8.

TURKEY POINT - UNITS 3 Sc 4 3/4 11-16


3/4. 11. 3 SOLID RADIOACTIVE WASTES


l ~g'I''+/~+7"r/ 1 / I

4. 11. 3. 1with the PCP.

4.11.3e2 SOLIDIFICATION (excluding dewatering) of at least one representativetest specimen from at least every tenth batch of each type of wet radioactivewastes (e.g., filter'ludges, spent resins, evaporator bottoms, boric acidsolutions, and sodium sulfate solutions) shall be verified in accordance withthe PROCESS CONTROL PROGRAM:

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATIONof the batch under test shall be suspended until such time as additionaltest specimens can be obtained, alternative SOLIDIFICATION parameterscan be determined in accordance with the PROCESS CONTROL PROGRAM,and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION ofthe batch may then be resumed using the alternative SOLIDIFICATIONparameters determined by the PROCESS CONTROL PROGRAM;

b. If the initial test specimen from a batch of waste fails to verifySOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for thecollection and testing of representative test specimens from eachconsecutive batch of the same type of wet waste until at least threeconsecutive initial test specimens demonstrate SOLIDIFICATION.The PROCESS CONTROL PROGRAM shall be modified as required, as providedin Specification 6.13, to assure SOLIDIFICATION of. subsequent batchesof waste; and

c. With the installed equipment incapable of meeting Specificati'on3. 11.3 or declared inoperable, restore the equipment to OPERABLEstatus or provide for contract capability to process wastes asnecessary to.satisfy all applicable transportation and disposalrequirements.

TURKEY POINT - UNITS 3 8 4 3/4 11-17 /I ~~

1

I F.r3. 11.3 gadioactive wastes shall be in accordance withthe PROCESS CONTROL PROGRAM to meet shipping and transportation requirements

,..p ~ qi ~h I h1$ G op/cool llccr~5~i~ ~B %e conoicnec' ifyam )coflrA+<.

APPLICABILITY: At e11 'times'.

ACTION:o I,coo m Ilcmll5 nid reef <?)f ori..e, ~ GT

e. With SOLIDIFICATION or re iiot me et'i ngf' 'en~ roe..oirrG

shipping and transportation requirements, suspend shipment of theinadequately processed wastes and correct the PROCESS CONTROL PROGRAM,the procedures, and/or the Solid Waste System as necessary to preventrecurrence.

b. With SOLIDIFICATION or dewatering not performed in accordance withthe PROCESS CONTROL PROGRAM, test the improperly processed waste ineach container to ensure that it meets and shippingrequirements a aQ~pp dmi ietrmtisrs-ection-ta.~r euen~recurrence. re

aphonica linen i'e~uireeerF'm oF Fge coro)see >

c. The provisions of peC are not-ayplkcabl e-.

SURVEILLANCE RE UIREMENTS r


3/4.11.4 TOTAL DOSE


3. 11.4 The annual (calendar year) dose or dose commitment to any HEHBER OF

THE PUBLIC due to releases of radioactivity and to radiation from uranium fuelcycle sources shall be limited to less than or equal to 25 mrems to the wholebody or any organ, except the thyroid, which shall be limited to less than orequal to 75 mrems.

APPLICABILITY: At all times.ACTION:

0

aO

b.

With the calculated doses from the release of radioactive materialsin liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or3. 11.2.3b., calculations shall be made including direct radiationcontributions from the units to determine whether the above limitsof Specification 3.11.4 have been exceeded. If such is the case,prepare and submit to the Commission within 30 days, pursuant toSpecification 6. 9. 2, a Special Report that defines the correctiveaction to be taken to reduce subsequent releases to prevent recur-rence of exceeding the above limits and includes the schedule forachieving conformance with the above limits. This Special Report,as defined in 10 CFR 20.405(c), shall include an analysis thatestimates the radiation exposure (dose) to a HEHBER OF THE PUBLICfrom uranium fuel cycle sources, including all effluent. pathwaysand direct radiation, for the calendar year that includes therelease(s) covered by this report. It shall also describe levelsof radiation and concentrations of radioactive material involved, andthe cause of the exposure levels or concentrations. If the estimateddose(s) exceeds the above limits, and if the release condition result-ing in violation of 40 CFR Part 190 has not already been corrected,the'Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190. Submittal of the reportis considered a timely request, and a variance is granted until staffaction on the request is complete.



4.11.4.1 Cumulative dose contributions froa liquid and gaseous effluentsshall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and4.11.2.3, and in accordance with the methodology and parajaeters in the ODCH.

4.11.4.2 Cumulative dose contributions from direct radiation from the unitsand the methodology used shall be indicated in the Semiannual RadioactiveEffluent Release Report. This requirement is applicable only under conditionsset forth in ACTION a. of Specification 3.11.4.

TURKEY POINT - UhITS 3 4 4 3/4 11-18 ~~


3/4. 12. 1 MONITORING PROGRAM


3. 12. 1 The Radiological Environmental Monitoring Program shall be conducted~ ~~

~ ~

~as specified in Table 3.12-1.


ACTION:

b.

With the Radiological Environmental Monitoring Program not beingconducted as specified in Table 3.12-1, prepare and submit tothe Commission, in the Annual Radiological Environmental OperatingReport required by Specification 6.9.1.3, a description of the reasonsfor not conducting the program as required and the plans for preventinga recurrence.

With the level of confirmed"" radioactivity as the result of planteffluents in. an environmental sampling medium at a specified locationexceeding the reporting levels of-,Table 3.12-2 when averaged over anycalendar quarter, prepare and submit to the Commission within 3O days,pursuant to Specification 6.9.2,, a Special Report that identifies thecause(s) for exceeding the limit(s) and defines the correctiveactions to be taken to reduce radioactive effluents so that thepotential annual dose" to a MEHBER OF THE PUBLIC is less than thecalendar year limits of Specifications 3.11.1.2, 3.11.2.2, or3. 11.2.3. When more than one of the radionuclides in Table 3. 12-2are detected in the sampling medium; this report shall be submittedif:

concentration 1 + concentration 2reporting leve 1 report ng eve 2

When radionuclides other than those in Table 3.12-2 are detected andare the result of plant effluents, this report shall be submitted ifthe potential annual dose" to a MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits ofSpecification 3.11.1.2, 3.11.2.2; or 3.11.2.3. This report is notrequired if the measured level of radioactivity 'Was not the resultof plant effluents; however, in such an event, the condition shallbe reported and described in the Annual Radiological EnvironmentalOperating Report required by Specification 6.9.1.3.

"The methodology and parameters used to estimate the potential annual dose to '

MEMBER OF THE PUBLIC shall be indicated in this report.""A confirmatory reanalysis of the original, a duplicate, or.a new sample may

be desirable, as appropriate. The results of the confirmatory analysis shallbe completed at the earliest time consistent with the analysis, but in anycase within 30 days.

TURKEY POINT - UNITS 3 8L 4 3/4 12-1JUN,' o

RADIOLOGICAL ENVIRONMENTAL MONITORING


ACTION Continued

C. With milk or fresh leafy vegetation samples unavailable from one ormore of the sample locations required by Table 3.12-1, identifyspecific locations for obtaining replacement samples and add themwithin 30 days to the Radiological Environmental Monitoring Programgiven in the ODCM. The specific locations from which samples wereunavailable may then be deleted from the monitoring program. Pursuantto Specification 6.14, submit in the next Semiannual RadioactiveEffluent Release Report documentation for a change in the ODCMincluding a revised figure(s) and table for the ODCM reflecting thenew location(s) 'with supporting information identifying the cause ofthe unavailability of samples and justifying the selection of thenew location(s) for obtaining samples.

d. The provisions of Specification 3.0.3 are not applicable.


4.12. 1 The radiological environmental monitor ing samples shall be collectedpursuant to Table 3.12-1 from the specific locations given in the table andfigure(s) in the ODCH, and shali be analyzed pursuant to the requirements ofTable 3.12"1 and the detection capabilities required by Table 4.12-1.


TABLE 3.12"1

RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM

EXPOSURE PATHWAY

AND/OR SAMPLE

1. Direct Radiation

2. Airborne

Radioiodine andParticulates

3. Waterborne~

a. Surface~ ~

NUHBER OF

REPRESENTATIVESAMPLES AND

SAMPLE LOCATIONS

21 monitoring locations

Five locations

'Three locations

SAHPLING AND«EEE Ffl I

Continuous monitoringwith sample collectionquarterly

Continuous sampler oper-ation with sample collec-tion weekly, or morefrequently if requiredby dust loading.

Monthly

TYPE AND FRED((ICYOF ANALYSIS

Gamma exposure ratequarterly

Radioiodine FilterI-131 analyst s weekly.

Par ticulate FilterGross beta radioactivityanalysis > 24 hoursfollowing filter change;

Gamma isotopic analysisof composite (bylocation) quarterly.

Gama isotopicand tritium analyses monthly.

I~ *g ~

CI

EXPOSURE PATWAYAND/OR SAMPLE

C

3. waterborne (Continued)

c b. SedimentQo from

Shoreline

4. Ingestion

a. Fish and- Inverte-

brates

Three locations Semiannually.

TABLE 3.12-1

RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM

NUMBER OF

REPRESENTATIVESAHPLES AND

2 3SAMPLING AND

NNIPLE LOINT (ONE( (( ) OLI Tl I FNEILUENOT( (TYPE ARD FREQ(g)CY

OF ANALYSIS

N

Gama isotopic analysissemiannually.

1. Crustacea

2. Fish

Two locations

Two locations

Semiannually

Semiannually

Gaaea isotopic analysissemiannually

Gaama isotopic analysissemiannually

'

b. FoodPro

leafy1.„~ TN Iegetatson

Honthly when available Gaaea isotopic and I-131 panalyses monthly.


TABLE NOTATIONS

(2)

(3)

Deviations are permitted from the required sampling schedule if specimensare unobtainable due to circumstances such as hazardous conditions, sea-sonal unavailability, and malfunction of automatic sampling equipment orother legitimate reasons. If specimens are unobtainable due to samplingequipment malfunction, corrective action shall be taken prior to the endof the next sampling period. All deviations from the sampling scheduleshall be documented in the Annual Radiological Environmental OperatingReport pursuant to Specification 6. 9. 1. 3.

Specific parameters of distance and direction sector from the centerlineof the plant vent stack and additional description where pertinent, shallbe provided for each and every sample location in Table 3.12-1 in a tableand figure(s) in the OOCM.

At times, it may not be possible or practicable to continue to obtainsamples of the media of choice at the most desired location or time. Inthese instances suitable alternative media and locations may be chosen forthe particular pathway in question and. appropriate substitutions madewithin 30 days in the Radiological Environmental Monitoring Program givenin the OOCM.

(4)

(5)

The following definition of frequencies shall apply to Table 3.12-1 only:

W~eekl - not less than once per calendar week. A maximum Interval of 11days ss allowed between the collection of any two consecutive samples.

Semi-Monthl - Not less than 2 times per calendar month with an intervalof not ess than 7 days between sample collections. A maximum intervalof 24 days is allowed between collection of any two consecutive samples.

Monthl - Not less than once per calendar month with an interval of notess than'0 days between collection of any two consecutive samples.

~uarterl - Not less than once per calendar quarter.

Semiannuall - One sample each between calendar dates (January 1 - June 30)and u y 1 - Oecember 31). An interval of not less than 30 days will beprovided between sample collections.

The frequency of analyses is to be consistent with the sample collectionfrequency.

One or more instruments, such as a pressurized ion chamber, for measuringand recording dose. rate continuously may be used in place of, or in addi-tion to, integrating dosimeters. For the purposes of this table, a thermo-luminescent dosimeter (TLD) is considered to be one phosphor; two or more

phosphors in a packet are considered as two or. more dosimeters.

TURKEY POINT - UNITS 3 8 4 3/4 12-5 ~'p



(6) Refers to normal collection frequency. More frequent sample collectionis permitted when conditions warrant it.

(7) Airborne particulate sample filters are analyzed for gross beta radio-activity 24 hours ore l r n and thorondau hter deca

I ddi'ment for a gamma isotopic on a composite sample, a gamma isotopic

is also required for each sample having a gross beta radioactivity whichis > 1.0 pCi/m3 and which is also > 10 times that of the most recentcontrol sample.

(8) Gamma isotopic analysis means the identification and quantification ofgamma-emitting radionuclides that may be attributable to the effluentsfrom the facility.

(9) Off-shore grab samples.

(10) Discharges from the Turkey Point Plant do not influence drinking water orground water

re S le(ll) Samples of vegetation grown nearest each of two different off-site locations o h ghest predicted annual average ground level D/g, andone sample of similar egetation at an available location15-30 km distant in the leas prevalent wind direction based upon his-torical data in the ODCM.

(~el, l eZg

~ TURKEY POINT - UNITS 3 8L 4 3/4 12-6

04T

giO+f

.TABLE 3. 12-2

REPORTING LEVELS FOR RADIOACTIVITYCONCENTRATIONS IN ENVIRONMENTAL SAMPLES

REPORTING LEVELS

ANALYSIS

H-3

Hn-54

Fe-59

MATER

(pCi/1)

30,000*

1,000

400

Co-58 1,000

Co-60 300

Zn-65 300I

Z~Nb-95*** 400

AIRBORNE PARTICULATEOR GASES (pCi/e~)

FISH(pCi/kg, wet)

30,000

10,000

30,000

10,000

20,000

MILK(pCi/1)

FOOD PRODUCTS

(pCi/kg, wet)

I-131

Cs-134

Cs-137

Ba La 140***

2*0

30

50

200

0.9

10

20

1,000

2,000

60

70

300

100

1,000

2,000

Qnce no drinking water pathway exists, a value of 30,000 pCi/1 is used. For drinking water samples, a valueof 20,MO pCi/l is used. This is 40 CFR Part 141 value.

**Applies to drinking water*""An equilibriua aixture of the parent and daughter isotopes which corresponds to the reporting value of the

parent isotope.

TABLE 4.12-1

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS

LOWER LIHIT OF DETECTION LLD

WATER

ANALYSIS (pCi/1 )AIRBORNE PARTICULATE

OR GASES (pCi/m3)FISH HILK

(pCi/kg, wet) (pCi/1)FOOD PRODUCTS SEDIMENT(pCi/kg, wet) (pCi/kg, dry)

c Gross Beta

a H-3D

Hn-54

Fe-59

Co-58,60

Zn-65

Zr-Nb-95

I"131

Cs-134

Cs-137

Ba-La-140

3000*

15

30

15

30

15(5)

1(4)

15

18

15(')

0. 01

0. 07

0. 05

0.06-

130

260

130

260

130

150

15

18

15(

60

60

80

150

180

*Since no drinking water pathway exists, a value of 3,000 pCi/1 is used. For drinking water samples, a valueof 2;000 yCi/1 is, used.


TABLE NOTATIONS

(l)This list does not mean that only these nuclides are to be considered.Other peaks that are identifiable, together with those of the abovenuclides, shall also be analyzed and reported in the Annual RadiologicalEnvironmental Operating Report pursuant to Specification 6. 9. l. 3.

(2)Required detection capabilities for thermoluminescent dosimeters usedfor environmental measurements are given in Regulatory Guide 4. 13.

(3)The LLQ is defined, for purposes of these specifications, as the smallestconcentration of radioactive material in a sample that will yield a netcount, above system background, that will be detected with 95K probabilitywith only 5X probability of falsely concluding that a blank observationrepresents a "real" signal.

For a particular measurement system, which may include radiochemicalseparation:

4.et (sb)LLD

E ~ Y ~ 2. 22 ~ Y iexp(-Alt)]

Where:

LLQ = the "a priori" lower limit of detection as defined above aspicoCuries per unit mass or volume,

sb = the standard deviation of the background counting rate or of thecounting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

Y = the sample size (units of mass or volume),

2.22 = the number of disintegrations per minute per picoCurie,

Y = the fractional radiochemical , plicable,

E, V, ndhts u in the culation.

he radioactive decay constant for the par >cula radionuclide, and/ i8dt~for environmental samples ~the elapsed time betwe samecollection, or end of the sample collection period, nd time ofcounting

ical valu

TURKEY POINT - UNITS 3 8( 4 3(4 12-9



It should be recognized that the LLD is defined as an a priori (%afore thefact) limit representing the capability of a measurement system and not asan a osteriori (after the fact) limit for a particular measurement.AnaTyses shal be performed in such a manner that the stated LLDs will beachieved under routine conditions. Occasionally background fluctuations,unavoidable small sample sires, the presence of interfering nuclides, orother uncontrollable circumstances may render these LLDs unachievable.In such cases, the contributing factors shall be identified and describedin the Annual Radiological Environmental Operating Report pursuant toSpecification 6. 9. 1. 3.

(4)LLD for drinking water samples. If no drinking water pathway exists, theLLD of gamma isotopic analysis may be used.

(5)An equilibrium mixture .of the parent and daughter isotopes which correspondsto 15 pCi/1 of the parent isotope.

. ~

TURKEY POINT - UNITS 3 81 4 3/4 12-10C

(Ill


3/4. 12. 2 LAND USE CENSUS


3.12.2 A Land Use Census shall be conducted and shall identify within adistance of 8 km (5 miles) the location in each of the 16 meteorologicalsectors of the nearest milk animal, the nearest resid nce the nearestgarden" of greater than 50 m2 (500 ft ) producin egetation.


ACTION:

res en

a.

b.

C.

With a Land Use Census identifying a location(s) that yields acalculated dose or dose commitment greater than the values currentlybeing calculated in Specification 4. 11.2.3, pursuant to Specifica-tion 6.9. 1.4, identify'he new location(s) in the next SemiannualRadioactive Effluent Release Report.

With a Land Use Census identifying a location(s) that yields acalculated dose or dose commitment (via the same exposure pathway)20K greater than at a location from which samples are currentlybeing obtained in accordance with Specification 3.12.1, add the newlocation(s) within 30 days to the Radiological Environmental Moni-toring Program given in the ODCH. The sampling location(s), exclud-ing the control station location, haying the lowest calculated doseor dose commitment(s), via the same exposure pathway, may be deletedfrom this monitoring program after October 31 of the year in whichthis Land Use Census was conducted. Pursuant to Specification 6.14,submit in the next Semiannual Radioactive Effluent Release Reportdocumentation for a change in the ODCH including a revised figure(s)and table(s) for the ODCM reflecting the new location(s) with informa-tion: supporting the change in sampling locations.

The provisions of Specification 3.0;3 are not applicable.

pvegetation sampling may be performed at the SITE BOUNDARY in eacho two different direction sectors with the highest predicted D/gs in lieu ofthe garden census. Specifications for egetation sampling inTable 3. 12-1, Part 4.b., shall be followed, nclu n anal sis of controlsamples.


)~i'L (Nd)



4. 12.2 The Land Use Census shall be conducted during the growing season atleast once per 12 months using that information that will provide the bestresults, such as'by a door-to-door survey, aerial survey, or by consultinglocal agriculture authorities. The results of the Land Use Census shall beincluded in the Annual Radiological Environmental Operating Report pursuant toSpecification 6.9. 1.3.



3/4.12.3 INTERLABORATORY COMPARISON PROGRAM


3.12.3 Analyses shall be performed on all radioactive materia s, supplied aspart of an erlaboratory Comparison Program that has been approved by theCommission " <


ACTION:

a ~ Mith analyses not being performed as required above, report thecorrective actions taken to prevent a recurrence to the Commissionin the Annual Radiological Environmental Operating Report pursuantto Specification 6.9. 1.3.



4.12.3 A summary of the results obtained as part of the above requiredInterlaboratory Comparison Program shall be included in the Annual RadiologicalEnvironmental Operating Report pursuant to Specification 6.9.1.3.

0

Th'i S spe i<i crim Zh~1~ applyvnaAev'|A th~4 eIpr see* +e se.mp ed ~e4<~ ~pro| toom n.ne)QSes o 4 is+pcs iiJ~cl

l c 44es 3.lZ.-) ~J Q.E~-2.-

~ ~

F" his condition is satisfied by participation in the Environmental. Radio-activity Laboratory Intercomparison Studies Program conducted by theEnvironmental Protection Agency (EPA).


g g Q )+i jg<t» //~a,.k

BASES FOR

SECTIONS 3.0 AND 4.0

LIMITING CONDITIONS FOR OPERATION

AND

SURVEILLANCE REQUIREMENTS

NOTE

The BASES contained in succeeding pages summarizethe reasons for the Specifications in Sections 3. 0and 4.0, but in accordance with 10 CFR 50.36 arenot part of these Technical Specifications.

TURKEY POINT - UNITS 3 AND 4 B 3/4 0"0

'ipQ.Pp Q '. )VL"( P 'IK~CONCi j4

eA

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS


BASES

eTURKEY POINT - UNITS 3 and 4 B 3/4 0-1

S ecification 3.0. 1 throu h 3.0.4 establish the general requirementsapplicable to Limiting Condst ons for Operation. These requirements are basedon the requirements for Limiting Conditions for Operation stated in the Codeof Federal Regulations, 10 CFR 50.36(c)(2):

"Limiting conditions for operation are the lowest functional capability orperformance levels of equipment required for safe operation of the facility.When a limiting condition for operation of a nuclear reactor is not met, thelicensee shall shut down the reactor or follow any remedial action permitted bythe technical specificatioq, until the condition can be met."

.I Il Appll y tll y I ~ I hl h'ndividual specsfscation as the requirement for when (i.e., in whipchOPERATIONAL MODES or other specified conditions) conformance to the LimitingConditions for Operation is required for safe operation of the facility. TheACTION requirements establish those remedial measures that must be takenwithin specified time limits when the requirements of a Limiting Condition forOperation are not met.

There are two basic types of ACTION requirements. The first specifies theremedial measures that permit continued operation of the facility which is notfurther restricted by the time limits of the ACTION requirements. In thiscase, conformance to the ACTION requirements provides an acceptable level ofsafety for unlimited continued operation as long as the ACTION requirementscontinue to be met. The second type of ACTION requirement specifies a timelimit in which conformance to the conditions of the Limiting Condition forOperation must be met. This time limit is the allowable outage time torestore an inoperable system or component to OPERABLE status or for ~estoringparameters within specified limits. If these actions are not completed withinthe allowable outage time limits, a shutdown is required to place the facilityin a MODE or condition in which the specification no longer applies. It isnot intended that the shutdown ACTION requirements be used as an operationalconvenience which permits (routine) voluntary 'removal of a system(s) orcomponent(s) from service in lieu of other alternatives that would not resultin redundant systems or components being inoperable.

The specified time limits of the ACTION requirements are applicable from thepoint in time it is identified that a Limiting Condition for Operation is notmet. The tiae limits of the ACTION requirements are also applicable when asystem or component is removed from service for surveil,lance testing or in-vestigation of operational problems. Individual specifications aey include aspecified time limit for the completion of a Surveillance Requirement whenequipment is removed from service. In this case, the allowable outage timelimits of the ACTION requirements are applicable when this limit expires ifthe surveillance has not been completed. When a shutdown is required to complywith ACTION requirements, the plant may have entered a MODE in which a newspecification becomes applicable. In this case, the time limits of the ACTION


BASES

requirements would apply from the point in time that the new specificationbecomes applicable if the requirements of the Limiting Condition for Operationare not met.

~if' .0.2 bit h pit I h p I!!exists when the requirements of the Limiting Condition for Operation are notmet and the associated ACTION requirements have not been implemented withinthe specified time interval. The purpose of this specification is to clarifythat (1) implementation of the ACTION requirements within the specified timeinterval constitutes compliance with a specification and (2) completion ofthe remedial measures of the ACTION requirements is not required whencompliance with a LimitingLondition of Operation is restored within the timeinterval specified in the associated ACTION requirements.

S ecification 3.0.3 establishes the shutdown ACTION requirements that must bemplemented when a Limiting Condition for Operation is not met and the

condition is not specifically addressed by the associated ACTION requirements..The purpose of this specification is to delineate the time limits for placingthe unit in a safe shutdown MODE when plant operation cannot be maintainedwithin the limits for safe operation defined by the Limiting Conditions forOperation and its ACTION requirements. It is not intended to be used as anoperational convenience which permits (routine) voluntary removal of redundantsystems or components from service in lieu of other alternatives that wouldnot result in redundant systems or components being inoperable. One hour isallowed to prepare for an orderly shutdown before initiating a change inplant operation. This time permits the operator to coordinate the reductionin electrical generation with the load dispatcher to ensure the stability andavailability of the electrical grid. The time limits specified to reachlower MODES of operation permit the shutdown to proceed in a controlled andorderly manner that is well within the specified maximum cooldown rate andwithin the cooldown capabilities of the facility assuming only the minimumrequired equipment is OPERABLE. This reduces thermal stresses on componentsof the primary coolant system and the potential for a plant upset that couldchallenge safety systems under conditions for which this specificationapplies.

If remedial measures permitting liaited continued operatidn of the facilityunder the provisions of the ACTION requirements are completed, the shutdown maybe terminated: The time limits of the ACTION requirements are applicable fromthe point in time there was a failure to eeet a Lioiting Condition forOperation. Therefore, the shutdown may be terminated. if the ACTION require-ments have been met or the time limits of the ACTION requirements have notexpired, thus providing an allowance for the completion of the requiredactions.

~,

The time limits of Specification 3.0.3 allow 37 hours for the plant to be inthe COLD SHUTDOWN MODE when a shutdown is required during the POWER MODE ofoperation. If the plant is in a lower HODE of operation when a shutdown isrequired, the time limit for reaching the next lower HODE of operation 0TURKEY POINT - UNITS 3 and 4 8 3/4 0-2 ~ al

' ~ g


BASES

applies. However, if a lower MODE of operation is reached in less time thanallowed, the total allowable time to reach COLD SHUTDOWN, or other'applicable.MODE, is not reduced. For example, if HOT STANDBY is reached in 2 hours;~theWtime allowed to reach HOT SHUTDOWN is the next 11 hours because the total timeto reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours.Therefore, if remedial measures are completed that would permit a return toPOWER operation, a penalty is not incurred by having to reach a lower HODE ofoperation in less than the total time allowed.

The same principle applies with regard to the allowable outage time limitsof the ACTION requirements, if compliance with the ACTION requirements forone specification results ip entry into a MODE or condition of operation foranother specification in which the requirements of the Limiting Condition forOperation are not met. If the new specification becomes applicable in lesstime than specified, the difference may be added to the allowable outagetime limits of the second specification. However, the allowable outage timelimits of ACTION requirements for a higher HODE of operation may not be usedto extend the allowable outage time that is applicable when a LimitingCondition for Operation is not met in a lower HODE of operation.

The shutdown requirements of Specification 3.0.3 do not apply in HODES 5 and6, because the ACTION requirements of individual specifications define theremedial measures to be taken.

.tl.4 blah lit tl INDE h g 1 LMigCondstlon for Operation is not met. It precludes placing the facility in ahigher MODE of operation when the requirements for a Limiting ion forOperation are not met and continued noncompliance to these nditio wouldresult in e shutdown to comply with the ACTION requ$ remen if<chen $ n NOOESwere permitted. The purpose of this specification is to e ure th facilityoperation is not initiated or that higher HODES of operati not enteredwhen corrective action is being taken to obtain compliance with a specificationby restoring equipment to OPERABLE status or parameters to specified limits.Compliance with ACTION requirements that permit continued operation of thefacility for an unlimited period 0f time provides an acceptable level of safetyfor continued operation without regard to the status of the plant before orafter a HODE change. Therefore, in this case, entry into 'an OPERATIONAL MODEor other specified condition may be made in accordance with the provisions ofthe ACTION requirements. The provisions of this specification should not,however, be interpreted as endorsing the failure to exercise good practice inrestoring systems or components to OPERABLE status before plant startup.

When a shutdown is required to comply with ACTION requirements, theprovisions of..Specification 3.0.4 do not apply because they would delayplacing the facility in a. lower MODE of"operation.~AC7/W) fi~rs all~a) Sc'r a~ aVrr/g gd d~ fQ yfuPJ~~~

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TURKE'g POINT - UNITS 3'and 4'

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BASES

S ecification 3.0.5 delineates the applicability of each specification toUni nd Unit peration.

3 +S ec ication 4. . 1 throu h 4.0.5 establish the general requirements applicableto Survei lance equirements. hese requirements are based on the SurveillanceRequirements stated in the Code of Federal Regulations, 10 CFR 50.36(c)(3):

"Surveillance requirements are requirements relating to test, calibrationor inspection to ensure that the necessary quality of systems and components ismaintained, that facility operation will be within safety limits, and that thelimiting conditions of operation will be met."

S ecification 4.0.1 establishes the requirement that surveillances must be. performed dur ng the OPERATIONAL MODES or other conditions for which the

requirements of the Limiting Conditions for Operation apply unless otherwisestated in an individual Surveillance Requirement. The purpose of thisspecification is to ensure that surveillances are performed to verify theoperational status of systems and components and that parameters are withinspecified limits to ensure safe operation of the facility when the plant isin a MODE or other specified condition for which the associated LimitingConditions for Operation are applicable. Surveillance Requirements do nothave to be performed when the facility is in an OPERATIONAL MODE for whichthe requirements of the associated Limiting Condition for operation do notapply unless otherwise specified. The Surveillance Requirements associatedwith a Special Test Exception are only applicable when the Special TestException is used as an allowable exception to the requirements of aspecification.

S ecification 4. 0.2 establishes the conditions under which the specified timeinterval for Surve llance Requirements may be extended. Item a. permits anallowable extension of the normal surveillance interval to facilitatesurveillance scheduling and consideration of plant operating conditions thatmay not be suitable for conducting the surveillance; e.g., transientconditions or other ongoing surveillance or maintenance activities. Item b.limits the use of the provisions of item a. to ensure that it is not usedrepeatedly to extend the surveillance interval beyond that specified. Thelimits of Specification 4.0.2 are based on engineering 5udgment and therecognition that the most probable result of any particular surveillancebeing performed is the verification of conformance with the SurveillanceRequirements. These provisions are sufficient to ensure that the reliabilityensured through surveillance activities is not significantly degraded beyondthat obtained from the specified surveillance interval.

TURKEY POINT - UNITS 3 and 4 B 3/4 0-4I I ~

v

>3.0 This soecification delineates what additiona! conditions must be satisiied'o

permit operation to continue, consistent with the ACTION statementsfor power sources, when a normal or emergency power source is notOpERABLE. It specifically prohibits operation when one division isinoperable because its normal or emergency power source is inoperableand a system, subsystem, train, component or device in another division isinoperable for another reason. isis a su~ca gkar o e, ivl/diwsro Qe „)geC Zb tucr crCccC a~sprcy pPLv~ 'cs,asiptglg ~l/IW Optg+gg. pThe provisions of this specification permit the ACTION statementsassociated with individual systems, subsystems, trains, components ordevices to be consistent with the ACTION statements of the associatedelectrical power sources It allows operation to be governed by the timelimits of the ACTION statement associated with the Limiting Conditionfor Operation for the normal or emergency power source, not theindividual ACTION statements for each system, subsystem, train,component or device that is determined to be inoperable solely because ofthe inoperability of its normal or emergency power source.z8',rFor example, Specification ~l requires in part that two emergencydiesel generators be OPERABLE. The ACTION statement provides for anout-of-service time when one emergency diesel generator is notOPERABLE. If the definition of OPERABLE were applied withoutconsideration of Specification 3.0P',~ all systems, subsystems, trains,components and devices supplied by the inoperable emergency powersource would also be inoperable. This would dictate invoking theapplicable ACTION statements for each of the applicable LimitingConditions for Operation. However, the provisions of Specification 3.0.0permit the time limits for continued ooeration to be consistent with theACTION statement for the inoperable emergency diesel generatorinstead, provided the other specified conditions are satisfied. In this case,this would mean that the corresponding normal power source must beOPERABLE, and all redundant systems, subsystems, trains, componentsand devices must be OPERABLE, or otherwise satisfy Specification 3.0.lf(i.e., be capable of performing their design function and have at least oneaetna~ one emergency power source OPERABLB. If they are notsatisfied, shutdown is

required.'obes

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S ecification 4.0.3 establishes the failure to perform a SurveillanceRequirement withsn the allowed surveillance interval, defined by theprovisions of Specification 4.0.2, as a condition that constitutes a failureto meet the OPERABILITY requirements for a Limiting Condition for Operation.Under the provisions of this specification, systems and components areassumed to be OPERABLE when Surveillance Requirements have been satisfactorilyperformed within the specified time interval. However, nothing in thisprovision is to be construed as implying that systems or components areOPERABLE when they are found or known to be inoperable although still meetingthe Surveillance Requirements. This specification also clarifies that theACTION requirements are applicable when Surveillance Requirements have not beencompleted within the allowed surveillance interval and that the time limits ofthe ACTION requirements apply from the point in time it is identified that asurveillance has not been performed and not at the time that the allowedsurveillance interval was exceeded. , Completion of the Surveillance Requirementwithin the allowable outage time limits of the ACTION requirements restorescompliance with the requirements of Specification 4.0.3. However, this doesnot negate the fact that the failure to have performed the surveillance withinthe allowed surveillance interval, defined by. the provisions of Specification4.0.2, was a violation of the OPERABILITY requirements of a Limiting Conditionfor Operation that is subject to enforcement action. Further, the failure toperform a surveillance within the provisions of Specification 4.0.2 is aviolation of a Technical Specification requirement and is, therefore, a re-portable event under the requirements of 10 CFR 50.73(a)(2)(i)(B) because itis a con pion yrohiPited by .the plant's Tec al Specifications.

Cn@en QC. C <OtJ Sg~'Ve~eIf thea wa e uae > >ms so ON requirements are less than24 hours or a shutdown is required to comply with ACTION requirements, e.g.,Specfficatfon 3.0.3, a 24-hour allowance ia provided to permit a delay 1n

This provides an adequate time limit tocomplete Surveillance Requirements that have not been performed. The purposeof this allowance is to permit the completion of a surveillance before ashutdown is required to comply with ACTION requirements or before otherremedial measures would be required that may preclude completion of asurveillance. The basis for this allowance includes consideration for plantconditions, adequate planning, availability of personnel, the time required toperform the surveillance, and the safety significance of the delay incompleting the required surveillance. The provision also provides a timelimit for the completion of Surveillance Requirements that become applicableas a consequence of HODE changes imposed by ACTION requirements and forcompleting Surveillance Requirements that are applicable when an exception tothe requirements of Specification 4.0.4 is allowed. If a surveillance is notcompleted within the 24-hour allowance, the time limits of the ACTIONrequirements are applicable at that time. %hen a surveillance is performedwithin the 24-hour allowance and the Surveillance Requirements are not met,the time limits of the ACTION requirements are applicable at'he time that thesurveillance is terminated.

TURKEY POINT - UNITS 3 and 4 B 3/4 0-5


BASES

Surveillance Requirements do not have to be performed on inoperable equipmentbecause the ACTION requirements define the remedial measures that apply.However, the Surveillance Requirements have to be met to demonstrate thatinoperable equipment has been restored to OPERABLE status.

S ecification 4.0.4 establishes the requirement that all applicablesurveys llances must be met before entry into an OPERATIONAL MODE or othercondition of operation specified in the Applicability statement. The purposeof this specification is to ensure that system and component OPERABILITYrequirements or parameter limits are met before entry into a MODE orcondition for which these systems and components ensure safe operation of thefacility. This provision applies to changes in OPERATIONAL MODES or otherspecified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable SurveillanceRequirements must be performed within the specified surveillance interval toensure that the Limiting Conditions for Operation are met during initial plantstartup or following a plant outage.

When a shutdown is required to comply with ACTION requirements, the provisionsof Specification 4.0.4 do not apply. because this would delay placing thefacility in a lower MODE of operation.

S ecification 4.0.5 establishes the requirement that inservice inspection ofASHE Code Class 1, 2, and 3 components and inservice testing of ASIDE Code Class1, 2, and 3 pumps and valves shall be performed in accordance with aperiodically updated version of Section XI'of the ASME Boiler and PressureVessel Code and Addenda as required by 10 CFR 50.55a. These requirementsapply, except when relief has been provided in writing by the Coaeission.

This specification includes a clarification of the frequencies for performingthe inservice inspection and testing activities required by Section XI'of theASME Boiler and Pressure Vessel Code and applicable Addenda. Thisclarification is provided to ensure consistency in surveillance intervalsthroughout the Technical Specifications and to remove any ambiguities relativeto the frequencies for performing the required inservice inspection andtesting activities.

Under the texas of this specification, the sere restrictive requirements ofthe Technical Specifications take precedence over the ASIDE Boiler and PressureVessel Code and applicable Addenda. The requirements of Specification 4.0.4to perform surveillance activities before entry into an OPERATIONAL MODE orother specified condition takes precedence over the ASHE Boiler and Pressure



BASES

Vessel Code provision which allows pumps and valves to be tested up to oneweek after return to normal operation. The Technical Specification definitionof OPERABLE does not allow a grace period before a component, that is notcapable of performing its specified function, is declared inoperable and takesprecedence over the ASME Boiler and Pressure Vessel Code provision whichallows a valve to be incapable of performing its specified function for up to24 hours before being declared inoperable.

S ecification 4.0.6 delineates the applicability of the surveillance activitiesto Un>t and Un~tg operations.

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3/4.1. 1 BORATION CONTROL

3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN

A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be madesubcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptablelimits, and (3) the reactor will be maintained sufficiently subcritical topreclude inadvertent criticality in the shutdown condition.

SHUTDOWN, MARGIN requirements vary throughout core life as a function offuel depletion, RCS boron concentration, and RCS Tavg The most restrictivecondition occurs at EOL. with Tav at no load operating tmperature. and isassociated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. Figure 3.1-1 shows the SHUTDOWN MARGIN equivalent to1.77'k/k at the end-of-core-life with respect to an uncontrolled cooldown.Accordingly, the SHUTDOWN MARGIN requirement is based upon this limitingcondition and is consistent with FSAR safety a'nalysis assump5ions. '6th T—-~ ~„jq« ~

less than 200oF, the reactivity transients resulting f~romha1d M I d e uk.NIIIDML ~ )rl

aeequa<e pro<ac<<on., r g+ gCg op y~ ino3eerfeef dilufio~ o$ PQQ nero'~5T

3/4.1.1. 3 MODERATOR TEMPERATURE COEFFICIENT

The limitations on moderator temperature coefficient (MTC) are providedto ensure that the value of this coefficient remains within the limiting.condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set ofplant conditions; accordingly, verification of MTC values at conditions otherthan those explicitly stated will require extrapolation to those conditions inorder to permit an accurate comparison.

The most negative MTC, value equivalent to the most Positive moderatordensity coefficient (MDC), was obtained by incrementally correcting the MDC

used in the FSAR analyses to nominal operating conditions. These corrections

TURKEY POINT - UNITS 3 AND 4 B 3/4 1-1


BASES

involved subtracting the incremental change in the MDC associated with a corecondit'ton of all rods inserted (most positive MDC) to an all rods withdrawncondition and, a conversion for the rate of change of moderator-density withtemperature at RATED THERMAL POWER conditions. This value of the HDC was thentransformed into the limiting MTC value -3.5 x 10-~ hk/k/4F. The HTC valueof -3. 0 x 10-~ hk/k/4F represents a conservative value (with corrections forburnup and soluble boron) at a core condition of 300 ppm equilibrium boronconcentration and is obtained by making these corrections to the limiting HTCvalue of -3.5 x 10-~ hk/k/4F.

The Surveillance Requirements for measurement of the MTC at the beginningand near the end of the fuel cycle are adequate to confirm that the MTC remainswithin its limits since this coefficient changes slowly due principally to thereduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUMTEMPERATURE FOR CRITICALITY

This specification ensures that the reactor will not be made criticalwith the Reactor Coolant System average temperature less than 5414F. Thislimitation is required to,ensure: (1) the moderator temperature coefficientis within it analyzed temperature range, (2) the trip instrumentation is withinits normal operating range, (3) the pressurizer is capable of being in anOPERABLE status with a steam bubble, and (4) the reactor vessel is above itsminimum RTNDT temperature.

3/4.1. 2 BORATIOX SYSTEMS

The Boron In)ection System ensures that negative reactivity control isavailabla 4uring each mode of facility operation.'he components required tope~form this function include: (1) borated water sources, (2) charging pumps,(3) separate flow paths, (4) boric acid transfer pumps, (5) associated HeatTracing Systems, and (6) an emergency power supply froe OPERABLE dieselgenerators.

With the RCS average temperature above 2004F, a minima of two boroninjection flow paths are required to ensure single'functional capability inthe event an assumed failure renders one of the flow paths inoperable. Oneflow path from the charging pump discharge is acceptable since the flow pathcomponents sub)ect to an active failure are upstream of the charging pumps.



BASES

BORATION SYSTEMS (Continued)

The boration flow path spe icati allows the RWST and the boric acidstorage tank to be the boron so e5. Du to the lower boron concentration inthe RWST, borating the RCS from this s ce is less effective than Loratingfrom the boric acid tank and a time may be required to achieve thedesired SHUTDOWN MARGIN required by ACTION statement restrictions.

The ACTION statement restrictions for the boration flow paths allowcontinued operation in mode 1 for a limited time period with either borationsource flow path or the normal flow path to the RCS (via the regenerative heatexchanger) inoperable. In this case, the plant capability to borate andcharge into the RCS is limited and the potential operational impact of thislimitation on mode 1 operation must be addressed. Nth both the flow pathfrom the boric acid tanks and the regenerative heat exchanger flow path

~ inoperable, immediate initiation of action to go to COLD SHUTDOWN is requiredbut no time is specified for the mode reduction due to the reduced plantcapability with these flow paths inoperable.

Two charging pumps with independent power supplies are required to beOPERABLE to ensure single functional capability in the event an assumedfailure renders one of the pumps or power supplies inoperable. However, theACTION statement 'restrictions allow 7 days to restore an inoperable pumpprovided that two charging pumps are available. This restriction isacceptable based on the low probability of losing the power source common toboth charging pumps. The bus supplying the pumps can be fed from either theEmergency Diesel Generator or the offsite grid through the startuptransformer.

The boration capability of either flow path is sufficient to provide therequired SHUTDOWN MARGIN in accordance with Figure 3.1-1 from expectedoperating conditions after xenon decay and cooldown to 2004F. The maximumexpected boration capability requirement occurs at EOL froa full power.equilibrium xenon conditions and requires 3080 gallons of 20,000 PPM boratedwater from the boric acid storage tanks or 320,000 gallons of 1950 PPM boratedwater from the refueling water storage tank (EST).

Nth the RCS temperature below 2004F, one boron in$edtion source flowpath is acceptable without single failure consideration on the basis of thestable reactivity condition of the reactor and the additional restrictionsprohibiting CORE ALTERATIONS and positive reactivity changes in the event thesingle boron infection systea source flow path becowes inoperable.

r

The boron capability required below 2004F is sufficient to provide aSHUTDOWN MARGIN of 1X hk/k after xenon decay and cooldown.froa 2004F to1404F. This condition requires either 500 gallons of 20,000 ppa borated waterfrom the boric acid storage tanks or 20,000 gallons of 1950 .ppe borated waterfrom the RWST.

TURKEY POINT - UNITS 3 AND 4 B 3/4 1-3 i '\


BASES

BORATION SYSTEMS (Continued)

The charging pumps are demonstrated to be OPERABLE by testing as requiredby Section XI of the ASME code or by specific surveillance requirements in thespecification. These requirements are adequate to determine OPERABILITYbecause no safety analysis assumption relating to the charging pumpperformance is more restrictive than these acceptance criteria for the pumps.

The limits on contained water volume and boron concentration of the RWST

also ensure a pH valu'e of between 8.5 and 11.0 for the solution recirculatedwithin containment after a LOCA. This pH band minimizes the evolution ofiodine and minimizes the effect of chloride and caustic stress corrosion onmechanical systems and components. The.,cr ~era~ore. f'eq J <re y e~f5 -.<r ',.eRld'5W p rP knOP'p,'' ~ 'r, ' ~: «, "r ..; <eOI-r4 ling ~~~/'/~~~ 055+~Df«t ~

The OPERABILITY of one Boron In)ection System derring REFUELIHG ensures ~that this system is available for reactivity control while in %DE 6.

The OPERABILITY of the redundant heat tracing channels associated with theboric acid tank system ensures that the solubility of the boron solution willbe maintained above the solubility limit of 135 F at 22,500 ppa boron.

One channel of heat tracing is'ufficient to maintain the specifiedtemperature limit. Since one channel of heat tracing is sufficient to maintainthe specified temperature, operation with one channel out-of-service ispermitted for a period of 30 days provided additional temperature surveillanceis performed.

3/4.1. 3 MOVABLE CONTROL ASSEMBLIES

. —;

between 0 andY:f)':~'" withdrawn inclusive. 8l ~Hog-rM-position-vs- i

1

The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and(3) the potential effects of rod misalignment. on associated accident analyses arelimited. OPERABILITY of the control rod position indicators is required todetermine control rod positions and thereby ensure compliance with the controlrod alignment and insertion limits continue. OPERABLE condition for theanalog rod position indicators is difined as heing cspahl ~ of indicatin odposition to within 412 steps of the . or e utdownBanks and Control Banks h and B, the R is defined as thegroup demand counter indicated position between 0 and ps rawninclusive, and between 200 and 228 steps withdrawn inclusive. This permits >'~,'the operator to verify that the control rods in these'binks are either fully~" '-

'ithdrawnor fully inserted, the normal operating modes for these banks.Knowledge of these bank positions in these two areas satisfies all accident

i analysis assonptions concerning thoir position. For Control Banks C and D, thes defined as the group demand counter indicated position

228 steps

TURKEY POINT - UNITS 3 AND 4 B'3/4 1-4


BASES

MOVABLE CONTROL ASSEMBLIES (Continued)~ ~

0

;- Comparison of the group demand counters to the bankinsertion limits with verification of rod position with the analog rodposition indicators (after thermal soak after rod motion) is sufficientverification that the control rods are above the insertion limits.

Rod position indication is provided by two methods: a digital count ofactuating pulses which shows demand position of the banks and a linearposition indicator Linear Variable Differential Transformer which indicatesthe actual rod position. The relative accuracy of the linear positionindicator Linear Variable Differential Transformer is such that, with the mostadverse error, an alarm will be actuated if any two rods within a bank deviateby more than 24 steps for rods in motion and 12 steps for rods at rest.Complete rod misalignment (12 feet out of alignment with its bank) does notresult in exceeding core limits in steady-state operation at RATED THERMALPOWER.

The ACTION statements which permit limited variations from the basicrequirements are accompanied by additional restrictions which ensure that theoriginal design criteria are met. Misalignment of a rod requires measurementof peaking factors and a restriction in THERMAL POWER. These restrictions pro-vide assurance of fuel rod integrity during continued operation. In addition,those safety analyses affected by a misaligned rod are reevaluated to confirmthat the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed roddrop time used in the safety analyses. Measurement with T greater than orequal to 5414F and with all reactor coolant pumps operating ensures that themeasured drop times will be representative of insertion times experiencedduring a Reactor trip at operating conditions.

. Control rod positions and OPERABILITY of the rod position indicators arerequired to be verified on a nominal basis of once per 12'hours with more fre-quent verifications required if an automatic monitoring channel is inoperable.These verification frequencies are adequate for assuring that the applicableLCOs are satisfied.

TURKEY POINT - UNITS 3 AND 4 B 3/4 1-5 'l ~ le g(

I P l ~i f'8'<'Ci /j<

3/4. 2 POWER DISTRIBUTION LIMITS

BASES

The specifications of this section provide assurance of fuel integrityduring Condition I (Normal Operation) and II (Incidents of Moderate Frequency)events by: {1) m'aintaining the minimum DNBR in the core greater than or equalto 1.30 during normal operation and in short-term transients, and (2) limitingthe fission gas release, fuel pellet temperature, and cladding mechanicalproperties to within assumed design criteria. In addition, limiting the peaklinear power density during Condition I events provides assurance that theinitial conditions assumed for the LOCA analyses are met and the- ECCS acceptancecriteria limit of 22004F is not exceeded.

The definitions of certain hot channel and peaking factors as used inthese specifications are as follows:

F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heatflux on the surface of a fuel rod at core elevation Z divided by theaverage fuel rod heat flux, allowing for manufacturing tolerances onfuel pellets and rods;

N Nuclear Enthalpy Rise Hot Channel- Factor, is defined as the ratio ofthe integral of linear power along the rod with the highest integratedpower to the average rod power; and

F (Z) Radial Peaking Factor, is defined as the ratio of peak power densityto average power density in the horizontal plane at core elevation Z.

4.2.1 AXIAL FLUX DIFFERENCE gc"Act '3/v'.0.l R LThe 'ts on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) uppe

bound envelope o 32 times the normalized axial peaking factor is not exceededduring either normal ration or in the event of xenon redistribution followingpower changes.

Target flux difference is de ined at equilibrium xenon conditions.The full-length rods may be positione thin the core in accordance withtheir respective insertion limits and shou e inserted near their normalposition for steady-state operation at high po levels. The value of thetar get flux difference obtained under these condit diVided by the fractionof RATED THERMAL POWER is the target flux difference a TED THERMAL POWER

for the associated core burnup conditions. Target flux dif ces for otherTHERMAL POWER levels are obtained by multiplying the RATED THE POWER valueby the appropriate fractional THERMAL POWER level. The periodic up ing 'ofthe target flux difference value is necessary to reflect core burnupconsiderations.



+j9+ ~mCOi Wg

AXIAL FLUX FFERENCE (Continued)

Although s is intended that the plant will be operated with the FDwithin the tar get bynd required by Specification 3.2.1 about the ta et fluxdifference, during r'hpid plant THERMAL POWER reductions, control d motionwill cause the AFD to 6eviate outside of the target band at red ed THERMALPOWER levels. This deviation will not affect the xenon redi ibution suffi-ciently to change the envel'bye of peaking factors which m be reached on asubsequent return to RATED THERMAL POWER (with the AFD hin the target band)provided the time duration of th+deviation is limit . Accordingly, a 1-hourpenalty deviation limit cumulative>guring the pre us 24 hours is provided foroperation outside of the target banMut withi he limits of Figure 3.2-1while at THERMAL POWER levels between 5 a 90K of RATED THERMAL POWER. ForTHERMAL POWER levels between 15K and 50K RATED THERMAL POWER, deviations ofthe AFD outside of the target band ar ess nificant. The penalty of2 hours actual time reflects this uced signi ance.

Provisions for monitori the AFD on an automati asis are derived fromthrough the alarm used pit p FD.

The de4eeokoac Phe~ni4i= -. . the

OPERABLE excore de or outputs and provides"an alarm messag immediately ifthe AFD for two more OPERABLE excore channels are outside th target band,

. Duringoperatio t THERMAL POWER levels between 50K and 9'nd between and 50KRATED ERMAL POWER, the computer outputs an alarm message when the naltyde ation accumulates beyond the limits of 1 hour and 2 hours, respect vely.

Figure B 3/4 2-1 shows a typical monthly target band.



ASES

THIS SIOURCOO NOT

ILLIIaTRATION ONlYSOR OS% RATION

40% 'I0% 0 +10%, +00%

IHDICATlDAXIAlSlVX DISS(haSSC'm

FIGURE B 3/4 2-1

TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER

8 3/4 2-3


I ES

MoM/3/4.2. nd 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY~NISEHOT CHAN L FACTOR

The lim> s on heat flux hot channel factor and nuclear e alpy rise hotchannel factor nsure that: (1) the design limits on peak cal power densityand minimum DNBR re not exceeded and (2) in the event of LOCA the peak fuelclad temperature w 1 not exceed the 2200 F ECCS accep nce criteria limit.

Each of these is asurable but will normall only be determinedperiodically as specific in Specifications 4. and 4.2.3. This periodicsurveillance is sufficient o ensure that th imits are maintained provided:

a. Control rods in < si le grou ove together with no individual rodinsertion differing by ore an a 12 steps, indicated, from thegroup demand position;

b. Control rod groups ar sequenin Specification 3 .3.6;

1

with overlapping groups as described

c. The control ro insertion limits of"S ifications 3.1.3.5 and3. 1.3.6 are intained; and

d. The axi power distribution, expressed in te of AXIAL FLUXDIFFE CE, is maintained within the limits.

F<H wil be maintained within its limits provided Condition a. through d.N

above are aintained. The relaxation of F~ as a function of THE POWER

allows anges in the radial power shape for all permissible rod insert nlimit .


POWER DISTRIBUTION LIHITS

BASES

HEA CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

Co ued

Fue rod bowing reduces the value of DNB ratio. Credit is av'ai-3able tooffset th reduction in the generic margin. The generic margins, totaljh'.

1X DNBR c pletely offset any rod bow penalties. This margin inclu thefollowing.

a. Desi limit DNBR of j1.30 vs 1.283,b. Grid Sp cing (K ) of [0.046 vs 0.0597,

c. Thermal D fusion Coefficient of f0.038 vs 0.05

d. ONBR Noltlp iee of [0.86 vs 0.88], eod

e. Pitch reductio

The applicable values of ro bow penalties are ferenced in the FSAR.

When an F~ measurement is ken, an al owance for both experimental errorand manufacturing tolerance must mad An allowance of 5X is appropriatefor a full-core map taken with the re Detector Flux Mapping System, and a3X allowance is appropriate for manu uring tolerance.

The Radial Peaking Factor, (Z), i easured periodically to provide

assurance that the Hot Charm Factor, F~(Z), emains within its limit. The

RTP

F limit for RATED TH L POWER (Fx ) as provi 'n the Radial Peakingxy

, Factor Limit Report er Specification 6.9.1.6 was dete ed from expectedpower control man ers over the full range of burnup con ions in the core.

When RCS ow rate and F~ are measured, no additional a wances areN

necessary p or to comparison with the limits of Figures 3. 2-3 an .2-4.

Heasurem t errors of 2.1X for RCS total flow rate and 4X for F~ have eenN

allowe for in tion of the design DNBR value.

3/4. 2. 4 UADRANT POWER TILT RATIO

The (VAGRANT POWER TILT RATIO limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis.Radial power distribution measurements are made during STARTUP testing andperiodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB

and linear heat generation rate protection with x-y plane power tilts. A

limit of 1.02 was selected to provide an allowance for the uncertainty asso-ciated with the indicated power tilt.

TURKEY POINT - UNITS 3 AND 4 B 3/4 2-5 ~ > j!<


BASES

UADRANT POWER TILT RATIO (Continued)

The 2-hour time allowance for operation with a tilt condition greaterthan 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such action actiondoes not correct the tilt, the margin for uncertainty on F (Z) is reinstatedby reducing the maximum allowed power by 3X for each percent of tilt in excessof 1. ~or iecore iYierr ocovple r'ia

For purposes of monitoring QUADRANT POWER TILT TIO when one excoredetector is inoperable, the movable incore detectors are used to confirm thatthe normalized symmetric power distribution is consistent with the QUADRANTPOWER TILT RATIO. The incore detector monitoring is done with a full incoreflux map or two sets of four symmetric thimbles. The two sets of four sym-metric thimbles is a unique set of eight detector locations. These locationsare C-8, E-5, E-ll, H-3, H-13, L-5, L-ll, N-8.

3/4.2. 5 DNB PARAMETERS

The limits on the DNB-related parameters assure that each of the param-eters are maintained within the normal steady-state envelope of operationassumed in the transient and accident analyses. The limits are consistentwith the initial FSAR assumptions and have been analytically demonstrated ade-quate to maintain a minimum DNBR above the applicable design limits throughouteach analyzed transient. The 'ted T value of 576.3 F and the indicatedpressurizer ssure value of sig correspond to analytical limits of QP.578.2 F an sig respective y, with allowance for measurement uncertainty.

22'he

indicated RCS flow value of 277,900 gpm corresponds to an analyticallimit of 268,500 gpm which is assumed to have a 3.5X measurement uncertainty.The above measurement uncertainty estimates assume that these instrumentchannel outputs are averaged to minimize the uncertainty.

The 12-hour periodic surveillance offIi ht 1 p t t d tthI hglimits following load changes and other expected transient operatio

mug sine( ~pnmouriacr prso~iure™" fQ~~e ~4e,/i~i-~ave /l~ c.e R~ ~CG ~ib~ g~

TURKEY POINT - UNITS 3 AND 4 B 3/4 2-6t988

3/4.2 POWER DISTRIBUTION LIMITS

BASES

os 6 7HigFPL Ac COIAQ

3/4 .2 .I AXIAL FLUX DIFFERENCE

The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper boundenvelope of 2.32 times the nomaalized axial peaking factor )s not exceededduring either normal operation or in the event of xen'on redistributionfollowing power changes.

Target flux difference is determined at equilibrium xenon conditions. Thefull-length rods may be positioned within the core in accordance with theirrespective insertion limits and should be inserted near their normal positionfor steady-state operation at high power levels. The value of the target fluxdifference obtained under these conditions divided by the fraction of RATEDTHERMAL POWER is the target flux difference at RATED THERMAL POWER for theassociated core burnup conditions. Target flux differences for other THERMALPOWER levels are obtained by multiplying the RATED THERMAL POWER value by theappropriate fractional THERMAL POWER level. The periodic updating of thetarget flux difference value is necessary to reflect core burnupconsiderations.

Strict control of the flux difference (and rod position) is not as necessaryduring operation at less than 90% power. ,This is because xenon distributioncontrol is not as significant as the control at full power and allowance hasbeen made in predicting the heat flux peaking factors for less strict controlat less than 90% power. Although it is intended that the Plant will beoperated with the AFD within the target band required by Specification 3.2.1about the target flux difference, during rapid plant THERMAL POWER reductions,control rod motion will cause the AFD to deviate outside of the target band atreduced THERMAL POWER levels. This deviation will not affect the xenonredistribution sufficiently to change the envelope of peaking factors whichmay be reached on a subsequent return to RATED THERMAL POWER (with the AFDwithin the target band) provided the time duration of the deviation islimited . Accordingly, a I-hour penalty deviation limit cumulative during theprevious 24 hours is provided for operation outside of the target band butwithin the limits of Figure 3.2-1 while at THERMAL POWER levels between 50$and 905 of RATED THERMAL POWER. If the flux difference exceeds the limit fora cumulative period of one hour in any 24 hours, then xenon distributions maybe significantly changed and hence operation at less than 501 power isrequired to protect against potentially more severe consequences of some

0 T ERREL 0 ER l T ~f TER TRERRRL 0 ER,deviations of the AFD outside of th target band are less significant. Thepenalty of 2 hours actual time ref cts this reduced significance.

4e'toseen IZffo Eoe 50 fyoStrict control of flux differences is not possible during certain physicstests or during the required periodic excore calibrations. Therefore, thisspecification is not applicable during physics testing or excore calibrationprovided the duration of the deviation is limited. This is acceptable due tothe extremely low probability of a significant accident during these

'17lE fi»E& Ou fgirvf0 th'e, targe7 4»dz f5 li»EE tea

2B3/4 2-g


BASES

USE 7lh is@PL, Rl~i4A

AXIAL FLUX DIFFERENCE Continued

Provisions for monitoring the AFD on an automatic basis are derived from theplant process computer through the AFD Monitor Alarm. The computer 4eteriim~r oviform

the OPERABLE excore detector outputs andprovides an alarm message iomediately if the AFD for two or more OPERABLEexcore channels are outside the target band

g P i" EMAL 0

between 505 and 90% and hei~,505 RATED THERMAL, the computer outputs an alarmmessage when the penalt deviation accumulates eyond the limits of 1 hour and2 hours, respectively. <i~+<M l5 ro an

) D~~F<Figure B 3/4 2-1 shows a typsca month y target band.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISEHOT CHANNEL FACTOR

The limits on heat flux hot channel factor and nuclear enthalpy rise hotchannel factor ensure that: (1) the design limits on peak local powerdensity and minimum DNBR are not exceeded and (2) in the event of a LOCA thepeak fuel clad temperature will not exceed the 2200 degrees-F ECCS acceptancecriteria limit. The LOCA peak fuel clad temperature limit may be sensitive tothe, number of steam generator tubes plugged. The current limit is valid fortube plugging levels up to 5X.

Fq(Z), Heat Flux Hot Channel Factor, is defined as the maximum local heat

flux on the surface of a fuel rod at core elevation Z divided by the averagefuel rod heat flux.

F<H, Nuclear Enthal Rise Hot Channel Factor, is defined as the ratio ofN

the integral of linear power along the rod with the highest integrated power

to the average rod power.

Each of these is measurable but will normally only be determined periodicallyas specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance issufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rodinsertion differing by more than + 12 steps, indicated, from the groupdemand position;

b. Control rod groups are sequenced with overlapping groups as described inSpecification 3.1.3.6.

0B 3/4 2-g


BASES

TNIE FIGUhE FDN ILLUEThATIONONLYDO NOT USE FOh OfthATION

40% K% I0% 0 +10% +00% +X%

INDICATED AXIALFLVX DIFFEhENCE

FIGURE B 3/4 2-1

TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER

B 3/4 2-

/ > g pg p (~ppzi>7>cy'C'.~

POWER Q I 5 TR I BU T I ON L IMITS

use 78.«A'<Wea 0 iP4

8ASES

HEAT FL'Jg 1OT CHANNEL FACTOR ANO NUCLEAR ENTHALPY R ISE HOT CHANNEL FACTORonto'ueo

c. Thc control rod fnsertfon limits of Spccfffcat1ons 3.1.3.S and 3.1.3.6 aremaf ntd intd ~ and

d. Thc axial power distribution, expressed fn terms of AXIAL FLUX DIFFERENCE,is maintained within the limits.

'%hen an F~ measurement 1s taken, both experimental error and manufacturingtolerance must be allowed for. F1ve percent fs the appropriate allowance fora full core map taken wfth the movable incore detector flux mapp1ng system anathree percent 1s thc appropr1ate allowance for manufactur 1ng tolerance. Theseuncertainties only apply ff the map fs taken for purposes other than thedetermination of P8L and PR8.

F<< will be mafnta1ned wfthfn fts limits provided Conditions a. throughN

d. above are maintained.

In the.spcc1ffed limit of FaN, there fs an 8 percent allowance forN

uncertainties which means Chat normal operation of the core fs expected to

rcsulc fn F<H < l,62/1.08. The logic behind the larger uncertainty in thisN

case is that (a} normal perturbatfons fn the radial po~er shape (e.g., rod

misa11gnment} affect F<H, fn most cases without necessarily affecting F<,N

(b) although the operator has a dfrect influence on Fn through movementof rods, and can lfmft ft to the desired value, he haE no d1rcct control over

F<H and (c) an error 1n the prediction for radial power shape, which may be

detected during startup physics tests can be ccmpensatcd for in F~ by tighteraxial control, but compensation i'r F<H 1s less readily available. Qhcn a

mcasur cmcnt of F>H fs taken, experimental error must be allowed for and 4% isN

the appropriate allowance for a full core map taken with the movable incoredetector flux mapping systea.

83/4 2-4A

I

If s//

U

POWER OISTRIBUTION LIMITS

BAS S

Use 70 (

Mok5 i~(

HEAT Ft UX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNE FACTOR

ont nued

O dMl

j~crtW ~~~ ~ g~go/J

g.~Cd'glcPCÃi

The analyt1cally detera1ned [F~]P 1s formulated to gener ate 11aft1ng shapes

for all load follow maneuvers cons1stent w1th control to a a 5% band aboutthe target flux 41fference. For Base Load operat1on the sever1ty of theshapes that need to be cons1dered 1s s1gn1f1cantly reduced relat1ve toload follow operatfon.

WM 4 Xo(/ocJ<uy

u4~ fo gt~r~«e p&i~6a~es:I dpi- ~q( r eamrekw.I.OU ~ K(a)p~

cJ/8/ e: d(w)> Jalap~/. g P dt's

Fqg~= B~ lMpe~"iq Facfoi.

~ The sever1ty of poss1ble shapes 1s small due to the restrfct1ons 1mposed

by Sect1ons 4.2.2.3. To quant1fy the effect of"the 11mtt1ng trans1ents: wh1ch could occur dur1ng Base Load operat1on, the funct1on Q(Z)BL 1s

calculated f~ the follow1ng relat1onsh1p:

'(Z)BL Wax F Z Base Load Case s 150 HN,'T ,F Z Base Case s 5X EOL BU

Fq Z) AROo 150 $8/T Fq AR , 8 L .]

Ii,,

(~> 'P

~l

I

J

) gl' pp". l ' YCi

I I

UD+ WH-(4FPL mao~

Bur. ere.~ — 4; wep~od'c4~ /re ollocuI-r.e > .a~ O e err-~tvtg ia I~ aCdl'0".

Fq(~)p~ = V (z)„„„g x F~C>) ) I.gq

u/ere '8) ac~ou~fs fw v~cerfai~l~~(k) uecoo~tS Fd1 ~i@/Po emglca/<y

C2') nneayuMd(= <aFid pf fE'ale j7ucdct4 uvero e ~a~ Je~slgf

P el ed~(aF~ ~

= PdJ Bur. «Pedip FaZ>~

For Radial Surndown operation the full spectral of possible shapes consistentw1th control to a e 5% Delta-l band needs to be considered fn determiningpower capability. Accordingly, to quantify the effect of the liait1ng

i transients which could occur during Rad1al Surndown operation, the functionFz(Z) 1s calculated from the following relationship:

:. Ez(Z) ~ tF<(Z)1 FAC Analysislt:Fxy(Z)] ARO

The essence of the procedure fs to maintain the xenon distribution in thecore as close to the equilibria full power condition as poss1ble. Thiscan be accompl1shed by usfng the boron system to position the full lengthcontrol rods to produce the requ1red 1ndicated flux difference.

Above the power level of PT, additional flux shape monitor1ng fs required.

?n order to assure that the total power peaking factor, F~, is maintained at

or belo~ the limit1ng value, the movable incore instr~ntation will beut11ized. Thimbles are selected init1ally during startup physics tests sothat the measurements are representat1ve of the peak core power density.Sy limiting the core average axial power distribut1on, the total power

peaking factor Fq can be limited since all other components remain relatively"'ixed. The reaaining part of the total power peak.ing factor can be derivedI

froi incore measurements, i.e., an effective radial peaking factor 0, can be

.determined as the ratio of the total peak1ng factor resulting from a full coreflux map and the axial peaking factor in a selected thimble.

B3/i 2-$ QC.

/ „. r, a

g la nr'C

POQER OISTRIBUTION.LIMITS

BASES

vs,e

Fpi~CD I Rg

HEAT FLUX iOT .-HANNEL FACTOR ANO NUCLEAR ENTHALPY RISE HOT CHANNEL FACTORCont>nued

The limiting value of (Fj (E)]s is derived as fo'llows:

[Fj (E)]s I F " x K E

PPj (1+ 5'j) (1 03)(1 07)%here:

')

F, (E) is the normalized axial power distribution from thimble j atelevation E.

b) PL is reactor thermal power expressed as a fraction of 1.

c) K (E) is the reduction in the Fg limit as a function of core elevation(E) as determined from Figure 3.2-2.

d) {:Fj (l)3sis the alarm setpo1nt for NIGS.

e) Jt,, for thimble j, is determined from n~6 incore flux maps covering thefill configuration of permiss1ble rod patterns at the thermal powerlimit of PT.

s ia1 g Rij

Rij ~ Fqi meas ~

max

and Fij (E) 1s the normalized axial distribution at elevation E fromthimble j in map 1 which has a measured peaking factor withoutunce~tainties og densification allo~ance of F~i meas.

f) (5 j is the standard deviation, expressed as a fraction or percentage

of IIj, and is der1ved from n flux maps and the relationship belo~, or0.02 (2X), whichever 1s greeter .

n1 E (Rij - Rj)

n-t 1~1

kj

8 3/4 2/ qg

1/2

I/'~@DE~

)67C'~iK '0'~"'l


BASES

(35~ 7+rSPPL ~Los~)

HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTORContinued

g) The factor 1.03 reduction in the kw/ft limit is the engineeringuncertainty factor.

h) The factors (1+ $ j) and 1.07 represent the margin between (F'(Z)]Ljlimit and the MIDS alarm setpoint [F.(Z)3s. Since (1+ 8'.) is bounded

by a lower limit of 1.02, there is at least a 9$ reduction of the alarmsetpoint. Operations are permitted in excess of the operational limit< 4X, while making power adjustment on a percent for percent basis.

/i i -'P-I

/VIYCr 'i''."i

lf

3/4.3 INSTRUMENTATION

BASES

g q~~O~~ggAS e. In el ~VIC-e ] ~g ja I y r ui I sa/ j,yl/5 peal Inall<g .. I 4? r a I.Y'( ~ COr!5E't uhl<.->~

~!e, -etpLi T.cillacooo'g 7ac') Je e.Z-S

dc ~tze7pg,~Pliw)/

3/4. 3. 1 and 3/4. 3. 2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES

A UATION SYS EM INS RUM N A ION

The OPERABILITY of the Reactor Trip System and the Engineered SafetyFeatures Actuation System instrumentation and interlocks ensures that: (1) theassociated ACTION and/or Reactor tr3p will be initiated when the parametermonitored by each channel or combination thereof reaches its Setpoint (2) thespecified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance,and (4) sufficient system functional capability is available from diverseparameters.

The OPERABILITY of these systems is required to provide the overallreliability, redundancy, and diversity assumed available in the facilitydesign for the protection and mitigation of accident and transient conditions.The integrated operation of each of these systems is consistent with theassumptions used in the safety analyses. The Surveillance Requirements speci-fied for these systems ensure that the overall system functional capability ismaintained comparable to the original design standards. The periodic surveil-lance tests performed at the minimum frequencies are suff c t to demonstratethis capability.

l<+'he

Engineered Sa eatures Actuation System Ins r ation TripSetpoints specified i Table 3.3-3 are the values at which the tripsare set for. each functiona unit. A Setpoint is considered to be ad]ustedconsistent with the value when the "as measured" Setpoint is withinthe band allowed for calibration accuracy.

modate the instrument drift assumed to occur betw ionalthe acy to which Setpoints can be an calibrated,Values for een specified in Table 3.3-3. Opera-Setpo ess conserv Trip Setpoint but within the

ue is acceptable since an allowance de in thesafety analysis to accommodate this error.

ia u e a oiuaace far in ev Once

0tests andAllowabletion withAllowabl

ent

The Engineered Safety Features Actuation System senses selected plantparameters and determines whether or not predetermined limits are. being exceeded.If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents events, and transients. Once therequired logic combination is completed, the system sends actuation signals to

pggII md'.5

The methodology to derive the Trip e nn the channels. Inherent to the determination of the

Trip Setpoints are the magnitudes of these channel uncertainties. Sensor andrack instrumentation utilized in these channels are expected to be capable fo crating within the allowances of these uncertainty magnitudes.

e rack has notmet its allowa a stical chance that thiswill happen, an i s v is expected. Rack or sensor drift,in exce e allowance that is more h onal', may be indicative of

serious problems and should warrant further inve on.


INSTRUMENTATION

BASES

REACTOR TRIP SYSTEM and ENGINEEREO SAFETY FEATURES ACTUATION SYSTEMINSTRUM NTATION Continued

those Engineered Safety Features components whose aggregate function bestserves the requirements of the condition. As an example, the following actionsmay be initiated by the Engineered Safety Features Actuation System to mitigatethe consequences of a steam line break or loss-of-coolant accident: (1) SafetyInjection pumps start and automatic valves position, (2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containmentspray pumps start and automatic valves position (6) containment isolation,(7) steam line iso1ation, (8) turbine trip, (9) auxiliary feedwater pumpsstart and automatic valves position, (10) containment cooling fans start andautomatic valves position, (11 water pumps start and auto-matic valves position, and (1 R ol ~t nd Ventilation Systemsstart. llipiQ ~~]q +pge,r ot ng gg+t ~g 8~7

The Engineered Safety Features Actuation ystem interlocks perform thefollowing functions:

HIGH STEAM FLOW SAFETY INJECTION BLOCK -. This permissive is used to blockthe safety injection (SI)- signal generated by High Steam Line Flow coincidentwith Low Steam Line Pressure or Low T . The permissive is generated when

avg'wo

out of three Low Tav channels drop below their setpoints and the manual

SI Block/Unblock switch is momentarily placed in the block position. Thisswitch is a spring return to the normal position type. The permissive willautomatically be defeated if two out of three Low T channels rise above

avgtheir setpoints. The permissive may be manually defeated when two out ofthree Low T channels are below their setpoints and the manual SI Block/

avgUnblock switch is momentarily placed in the unblock position.

LOW PRESSURIZER PRESSURE SAFETY -INJECTION BLOCK - This permissive is usedto block the safety injection signals generated by Low Pressurizer Pressureand High Oifferential Pressure between the Steam Line Header and any SteamLine. The permissive is generated when two out of three pressurizer pressurepermissive channels drop below their setpoints and the manual SI Block/Unblockswitch is momentarily placed in the block position. This is the same switchthat is used to manually block the High Steam Flow Safety Injection signalsmentioned above. This permissive will automatically be defeated if two out ofthree pressurizer pressure perwissive channels rise above their setpoints.The permissive channels rise above their setpoints.. The permissive may bemanually defeated when two out of three pressurizer pressure permissivechannels are below their setpoints and the manual SI Block/Unblock switchmomentarily placed in the Unblock position.

0

/4: 7-.-1-.

, TURKEY POINT - UNITS 3 and 4 B 3/4 3-2

INSTRUMENTATION

BASES

3/4. 3. 3 MONITORING INSTRUMENTATION

3/4. 3. 3. 1 RADIATION MONITORING FOR PLANT OPERATIONS

The OPERABILITY of the radiation monitoring instrumentation for plantoperations ensures that conditions indicative of potential uncontrolledradioactive releases are monitored and that appropriate actions will beautomatically or manually initiated when the radiation level monitored by eachchannel reaches its alarm or trip setpoint.

3/4.3.3.2 MOVABLE INCORE DETECTORS

The OPERABILITY of the movable incore detectors with the specified minimumcomplement of equipment ensures that the measurements obtained from use ofthis system accurately represent the spatial neutron flux distribution of thecore. The OPERABILITY of this system is demonstrated by irradiating eachdetector used and determining the acceptability of its voltage curve.

For the purpose of measuring F~(Z) or F~ a full incore flux map is used.N

A

quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used inrecalibration of the Excore Neutron Flux Detection System, and full incoreflux maps or symmetric incore thimbles may be used for monitoring the QUADRANTPOWER TILT RATIO when one Power Range channel is inoperable.

3/4. 3. 3. 3 ACCIDENT MONITORING INSTRUMENTATION

The OPERABILITY.of the accident monitoring instrumentation ensures thatsufficient information is available on selected plant parameters to monitorand assess these variables following an accident. This capability is consis-tent with the recommendations of Regulatory Guide 1.97, Revision 3, "Instrumen-tation for Light-Water-Cooled Nuclear Power Plants to Assess Plant ConditionsDuring and Following an Accident," May 1983 and NUREG-0737, "Clarification ofTMI Action Plan Requirements," November 1980.

3/4. 3. 3. 4 FIRE DETECTION INSTRUMENTATION

The OPERABILITY of the fire detection instrumentation ensures that bothadequate warning capability is available for prompt detection of fires and thatFire Suppression Systems, that are actuated by fire detectors, will dischargeextinguishing.agents in a timely manner. Prompt detection and suppression offires will reduce the potential for damage to safety-related equipment and isan integral element in the overall facility Fire Protection Program.

Fire detectors that are used to actuate Fire Suppression Systems representa more critically important component of a plant Fire Protection Programthan detectors that are installed solely for early fire warning and notifica-tion. Consequently, the minimum number of OPERABLE fire detectors must be

'reater.

The loss of detection capability for Fire Suppression Systems, actuatedby fire detectors, represents a significant degradation of fire protection for


INSTRUMENTATION

BASES

FIRE DETECTION INSTRUMENTATION (Continued)

any'area. As a result, the establishment of a fire watch patrol must be ini-tiated at an earlier stage than would be warranted for the loss of detectorsthat provide only early fire warning. The establishment of frequent firepatrols in the affected areas is required to provide detection capabilityuntil the inoperable instrumentation is restored to OPERABILITY.

3/4. 3. 3. 5 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION

The radioactive liquid effluent instrumentation is provided to monitorand control, as applicable, the, releases of radioactive materials in liquideffluents during actual or potential releases of liquid effluents. TheAlarm/Trip Setpoints for these instruments shall be ca3culated and ad)usted inaccordance with the methodology and parameters in the ODCH to ensure that thealarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. TheOPERABILITY and use of this instrumentation is consistent with the requirementsof General Design Criteria 60, 63, and. 64 of Appendix A to 10 CFR Part 50.

3/4. 3. 3. 6 RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION

The radioactive gaseous effluent instrumentation is provided to monitor andcontrol, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential releases of gaseous effluents. The Alarm/TripSetpoints for these instruments shall be calculated and adjusted in accordancewith the methodology and parameters in the ODCM to 'ensure that the alarm/tripwill occur prior to exceeding the limits of 10 CFR Part 20. This instrumenta-tion also includes provisions for monitoring (and controlling) the concentrationsof potentially explosive gas mixtures in the GAS DECAY TANK SYSTEM. The OPERA-BILITY and use of this instrumentation is consistent with the requirements ofGeneral Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. Thesensitivity of any noble gas activity monitors used to show compliance with thegaseous effluent release requirements of Specification 3.11.2.2 shall be suchthat concentrations as low as 1 x 10-'~ pCi/ml're measurable.

TURKEY POINT - UNITS 3 and 4 8 3/4 3-4l

4f~r

3/4.4 REACTOR COOLANT SYSTEH

BASES

3/4.4. 1 REACTOR CDOLANT LOOPS AND COOLANT CIRCULATION

The plant is designed to operate with all reactor coolant loops inoperation and maintain DNBR above the applicable design limit during all normaloperations and anticipated transients. In HODES 1 and 2 with one reactorcoolant loop not in operation this specification requires that the plant be inat least HOT STANDBY within S hours.

In HODE 3, three reactor coolant loops provide sufficient heat removalcapability for removing core decay heat in the event of a bank withdrawalaccident; however, a single reactor coolant loop provides sufficient heatremoval capacity if a bank withdrawal accident can be prevented, i.e., byopening the Reactor Trip Nyo4em breakers or-by placing the Rod Control Systemin the Bank Select Hode with a Shutdown Bank Selected.

In HODE 4, and in HODE 5 with reactor coolant loops filled, a singlereacto~ coolant loop or RHR loop provides sufficient heat removal capabilityfor removing decay heat~

In HODE 5 with reactor coolant loops not filled, a single RHR loop providessufficient heat removal capability for removing decay heat; but single failureconsiderations, and the unavailability of the steam generators as a heatremoving component, require that at least two RHR loops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RHR pump providesadequate flow to ensure mixing, prevent stratification and produce gradualreactivity changes during boron concentration reductions in the Reactor CoolantSystem. The reactivity change rate associated with boron reduction will,therefore, be within the capability of operator recognition and control.

The restrictions on starting an RCP with one or more RCS cold legs lessthan or equal to 2754F are provided to prevent RCS pressure transients, causedby energy additions from the Secondary Coolant Systeo, which could exceed thelimits of Appendix G to 10 CFR Part 50. The RCS will be, protected againstoverpressure transients and will not exceed the limits of Appendix G by either:(1) restricting the water volume in the pressurizer and thereby providing avolume for the reactor coolant to expand into, or (2) by restricting starting ofthe RCPs to when the secondary water temperature of ea nefa&r4s lt+~0 F ag of the RCS cold leg temperatures Q~ ~'F hpvif inc e5gjh Ctumewl el"IO/,

T~eehnical~Specifications for Hot Shutdown and Cold Shutdown allow aninoperable RHR pump to be the operating RHR pump for up to 2 hours for sur-veillance testing to establish operability. This is required because of thepiping arrangement when the RHR system is being used for Decay Heat Removal.

'URKEY

POINT - UNITS 3 AND 4 B 3/4 4-1J 111 ~

t'.e y 9~ Jv


BASES

C:

3/4.4.2 SAFETY VALVES gpss, 3~g)

The pressurize Code safety valves operate to prevent the RCS from beingpressurized above its Safety Limit of 2735 psig. Each safety val e is designedto relieve 480-4'I89- lbs per hour of saturated steam at the valv etpoint. The J E:relief capacity of a single safety valve $ s adequate to relieve aug overpressurecondition which could occur'uring shutdown. In the event that no safetyva1ves are OPERABLE, an RCS vent opening of at least 2.20 squareinches willprovide overpressure relief capability and will prevent RCS overpressurization.r, 0 ~ y pprotection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safefy val s must be OPERABLE toprevent the RCS from being pressurized above its Safety Limit of 2735 psig.The combined relief capacity of all of these valves is greater than the maximumsurge rate resulting from a complete loss-of-load assuming no Reactor tripuntil the first Reactor Trip System Trip Setpoint is reached (i.e., nocredit is taken for a direct Reactor trip on the loss-of-load) and also assumingno operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves'ift settings will occur only duringshutdown and will be performed in accordance with the provisions of Section XIof the ASIDE Boiler and Pressure Code.

3/4.4. 3 PRESSURIZER

~ g

The limit on the maximum water volume in the pressurize'r assures that theparameter is maintained within the normal steady-state envelope of operationassumed in the, FSAR. , The limit is consistent with the initial FSAR assumptions.The 12-hour periodic surveillance is sufficient to ensure that the parameteris restored to within its limit following expected transient operation. Themaximum water volume cubic feet) also ensures that a steam bubble isformed and thus the RC is not a hydraulically solid system. The requirementthat both backup pre urizer heater groups be OPERABLE enhances the capabilityof the plant to con ol Reactor Coolant System pressure and establish naturalcirculation.

( //93~ L

Jot-]

'URKEY POINT - UNITS 3 AND 4 B 3/4 4-2


BASES

pokd BLe3/4. 4, 4 l~~ VALVES

T4o~w~es

gn rans en s up

The opening of the PORVs fulfills no safety-related function and no credit istaken for their operation in the safety analysis for MODE 1, 2 or 3. Each PORVhas a remotely operated block valve to provide a positive shutoff capabilityshould a relief valve become inoperable.

3/4.4.5 STEAM GENERATORS

The Surveillance Requirements for inspection of the steam generator tubesensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes isbased on a modificati'on of Regulatory Guide 1.83, Revision 1. Inserviceinspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence ofmechanical damage or progressive degradation due to design, manufacturingerrors, or inservice conditions that lead to corrosion. Inservice inspectionof steam generator tubing also provides a means of characterizing the natureand cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondarycoolant will be maintained within those chemistry limits found to result innegligible corrosion of the steam generator tubes. If the secondary coolantchemistry is not maintained within these limits, localized cor rosion maylikely result in stress corrosion cracking. The extent of cracking duringplant operation would be limited by the limitation of steam generator tubeleakage between the Reactor Coolant System and the Seconda Coolant Syste(reactor-to-secondary leakage = 500 gallons per day peP earn genera or . ~ ~

~) ECracks having a reactor to-secondary leakage less than this limit during ~"f o'

i

operation will have an adequate margin of safety to withstand the loads im s'il'uringnormal operation and by postulated accidents. Operating plants have

demonstrated that reactor-to-secondary leakage of 500 gallons per day persteam generator can readily be detected by radiation monitors of steam generatorblowdown. Leakage in excess of this limitvill require plant shutdown and anunscheduled inspection, during ~hich the leaking tubes will be located andplugged.

wastage-type defects are unlikely with the all volatile treatment (AVT)of the secondary coolant. However, even if a defect should develop in service,it will be found during s'cheduled inservice steam generator tube examinations.Plugging will be required. for all tubes with imperfections exceeding theplugging limit of 40K of the tube nominal wal1 thickness. Steam generatortube inspections of operating plants have demonstrated the capability toreliably detect degradation that has penetrated 20K of the original tube wallthickness.



BASES

STEAM GENERATORS (Continued)

Whenever the results of any steam generator tubing inservice inspectionfall into Category C-3, these results will be promptly reported to the Commissionin a Special Report pursuant to Specfffcation 6.9.2 within 30 days and prior toresumption of plant operation. Such cases will be considered by„the Commissionon a case-by-case basis and may result in a requirement for analysis, laboratoryexaminations, tests, additional eddy-current inspection, and revision of theTechnical Specifications, if necessary.

3/4.4. 6 REACTOR COOLANT SYSTEM LEAKAGE

3 4'4'6'I LEAKAGE UETEGTIo sYGTEMs A0 lect<+ end 'c~cL~~

The RCS Leakage Detection stems required by this specification areprovided to monitor and detec leakage from the reactor coolant pressureboundary to the containment. The containment sump level system is the normalsump level instrumentation. ~/he Post Accident Containment Wat er Level Monitor - 0Narrow range instrumentatfon also satisfies the requirement for a sump levelmonitoring system. In addition, gross leakage will be detected by changes inmakeup water require ents, visual inspection, and audible detection. Leakageto" other systems will detected by activity changes (e.g., within the tcomponent cooling system or water inventory changes (e.g ,, tank levels).3/4.4.6.2 OPERATIONAL LEAKA 'per loIple. ~ ~ -5"

)

PRESSURE BOUNDARY LEAKAGE of any magnitude fs unacceptable since it maybe indicative of an impending gross failure of the pressure boundary. Therefore,the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptlyplaced in COLD SHUTDOWN.

Industry experience has shown that while a limited amount of leakage isexpected from the RCS, the unidentified portion of this leakage can be reducedto a threshold value of less than 1 gpm. This threshold value fs sufficientlylow to ensure early detection of'dditional leakage.

The total steam generator tube leakage limit of 1 gpm for all steamgenerators ensures that the dosage contribution from the tube leakage will belimited to a small fraction of 10 CFR Part 100 dose guideline values in theevent of either a steam generator tube rupture or steam line break. The 1 gpmlimit is consistent with the assumptions used fn the analysis of these acci-dents. The 500 gpd leakage limit per steam generator ensures that steamgenerator tube integrity is maintained fn the event of a main steam linerupture or under LOCA conditions.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limitedamount of leakage from known sources whose presence will not interfere withthe detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

g<~Q /~WAG-E. f3 e 3+4 tc"IS P ~Jr I K+ ClpeMOn ~<4egp4sc, o,- c~ .Ic„;. p>+p se~ls pseefs 30SP~ 87

W Z nO,„»i„t gC.S PreSS~mTURKEY POINT - UNITS 3 ANO 4 6 3/4 4-4 +o4'235 .si~.


BASES

OPERATIONAL LEAKAGE (Continued)

The leakage from any RCS pressure isolation valve is sufficiently low toensur early detection of possible in-series valve failure. It is apparentthat when pressure isolation is provided by two in"series valves and whenfai lure of one valve in the pair can go undetected for a substantial lengthof time, verification of valve integrity is required. Since these valves areimportant in preventing overpressurization and rupture of the ECCS low pressurepiping which could result in a LOCA, these valves should be tested periodicallyto ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provideadded assurance of valve integrity thereby reducing the probability of grossvalve failure and consequent intersystem LOCA.

. Q~ )w+G ibe<4(led. Q~iM 4c5c. gus-dclll~qc.e.5 0Am// lioi6g n»~i ~l < <~ ZLW77 F'IC D 4.CA'KA&E polio~ +he iepuIP i~ ac+pc..l~a5~ jw~ fm~ the QC5 |n >t ic4. case +he 1'~ >h<~I

bi'/4.4.7 CHEMISTRY ~ggT|F tgpThe limitations on Reacto~ Coolant System chemistry ensure that corrosion

of the Reactor Coolant System is minimized and reduces the potential forReactor Coolant System leakage or failure due to stress corrosion. Maintainingthe chemistry within the Steady-State Limits provides adequate corrosionprotection to ensure the structural integrity of the Reactor Coolant Systemover the life of the plant. The associated effects of exceeding the oxygen,chloride, and fluoride limits are time and temperature dependent. Corrosionstudies show that operation may be continued with contaminant concentrationlevels in excess of the Steady-State Limits, up to the Transient Limits, forthe specified limited time intervals without having a significant effect onthe structural integrity of the Reactor Coolant System. The time intervalpermitting continued operation within the restrictions of the Transient Limitsprovides time for taking corrective actions to restore the contaminant concen-trations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentrationsin excess of 'the limits will be detected in sufficient time to take correctiveaction.

3/4.4. 8 SPECIFIC ACTIVITY

The limitations on the specific activity of the reactor coolant ensurethat the resulting 2-hour doses at the SITE BOUNDARY will not exceed an



BASES

SPECIFIC ACTIVITY (Continued)

appropriately small fraction of 10 CFR Part 100 dose guideline values followinga steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm.. The valuesfor the limits on specific activity represent limits based upon a parametricevaluation by the NRC of typical site locations. These values are conservativein that specific site parameters of the Turkey Point site, Units 3 and 4 site,such as SITE BOUNDARY location and meteorological conditions, were not con-sidered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limitedtime periods with the reactor coolant's specific activity greater than1 microCurie/gram DOSE EQUIVALENT I-131, but within the allowable limitshown on Figure 3.4-1, accommodates possible iodine spiking phenomenon whichmay occur following changes in THERMAL POWER.

The sample analysts for determining th o s specific activity and E canexclude the radioiodines because of the 1 reactor coolant limit of 1 microCurie/"gram DOSE EQUIVALENT I-131, and because, f the limit is exceeded, theradioiodine level is to be determined e ery 4 hours. If the gross specificactivity level and radioiodine level i the reactor coolant were at theirlimits, the radioiodine contribution uld be approximately 3X. In a releaseof reactor coolant with a typical mix ure of radioactivity, the actual radio-iodine contribution would probably be about 20K.. The exclusion of radio-nuclides with half-lfves less than minutes from these determinations has )

™~

been made for several reasons. The first consideration is the difficulty toidentify short-lived radionuclides in a sample that requires a significanttime to collect, transport, and analyze., The second consideration is thepredictable delay time between the postulated release of radioactivity fromthe reactor coolant to its release to the environment and transport to theSITE BOUNDARY, which is relatable to at least 30 minutes decay time. Thechoice o Pf minutes for the ha'It-life cutoff was made because of the nuclearcharacteristics of the typical reactor coolant radioactivity. s

n ewe

the.t-to'%e YLT~OU

TURKEY POINT " UNITS 3 AND. 4 B 3/4 4-6 J(q... g


BASES

Based upon the above considerations for excluding certain radionuclidesfrom the sample analysis, the allowable time of 2 hours between sample takingand completing the initial analysis is based upon a typical time necessary toperform the sampling, transport the sample, and perform the analysis of about90 minutes. After 90 minutes, the gross count should be made in a reproduciblegeometry of sample and counter having reproducible beta or gamma self-shieldingproperties. The counter should be reset to a reproducible efficiency versusenergy. It is not necessary to identify specific nuclides. The radiochemicaldetermination of nuclides should be based on multiple counting of the samplewithin typical counting basis following sampling of less than 1 hour, about2 hours, about 1 d about 1 week, a d out 1 month.

a,<~r~ ce l~Reducing gto s an 'revents the release of activity should p

a steam generator ube rupture since the saturation pressure of the reactorcoolant is below the liftpressure of the atmospheric steam relief valves.The Surveillance Requirements provide adequate assurance that excessive specificactivity levels in the reactor coolant will be detected in sufficient time totake corrective action. A reduction in frequency of isotopic analyses followingpower changes may be permissible if justified by the data obtained.

3/4.4. 9 PRESSURE/TEMPERATURE LIMITS~ ~

All components in the RCS are designed to withstand the effects of cyclicloads due to system temperature and pressure changes. Ouring RCS heatup andcooldown, the temperature and pressure changes must be limited to beconsistent with design assumptions and to satisfy stress limits for brittlefracture.

The temperature and pressure changes during heatup and cooldown arelimited to be consistent with the requirements given in the ASME Boiler andPressure Vessel Code, Section III, Appendix G:

1. The reactor coolant temperature and pressure and system heatup and cooldownrates (with the exception of the pressurizer) shall be limited in accordancewith Figures 3.4-2 to 3.4-5 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specifictemperature change rates are below and to the right of the limitlines shown. Limit lines for cooldown rates between those presentedmay be obtained by interpolation; and

b. Figures 3.4-2 to 3.4-5 define limits to assure prevention ofnon-ductile failure only. For normal operation, other inherent plantcharacteristics, e.g., pump heat addition and pressurizer heatercapacity, may limit the heatup and cooldown rates that can beachieved over certain pressure-temperature ranges.



BASES

C'RESSURE/TEMPERATURE

LIMITS (Continued)

2. These limit 1ines shall be calculated periodically using methods providedbelow,

3. The secondary side of the steam generator must not be pressurized above200 psig if the temperature of the steam generator is below 70oF,

The pressurizer heatup and cooldown rates shall not exceed 1004F/h and200 F/h, respectively. The spray shall not be used if the temperaturedifference between the pressurizer and the spray fluid is greater than3004F, and

System preservice hydrotests and inservice leak and hydrotests shall beperformed at pressures in accordance with the requirements of ASME Boilerand Pressure Vessel Code, Section XI.

The fracture toughness properties of the ferritic materials in the reactorvessel are determined in accordance with the NRC Standard Review Plan, ASTHE185-73, and in accordance with additional reactor vessel requirements. Theseproperties are then evaluated in accordance with Appendix G of the 1976 SummerAddenda to Section III of the ASHE Boiler and Pressure Vessel Code and thecalculation methods described in CAP-7924-A, "Basis for Heatup and CooldownLimit Curves," April 1975.

Heatup and cooldown limit curves are calculated using the most limitingvalue of the nil-ductility reference temperature, RTNOT, at the end of10 effective full power years (EFPY) of service life. The 10 EFPY servicelife period is chosen such that the limiting RTNOT at the 1/4T location inthe core region is greater than the RTN>T of the limiting unirradiated material.The selection of such a limiting RTN>T assures that all components in theReactor Coolant System will be operated conservatively in accordance withapplicable Code requirements.

The reactor vessel materials have been tested to determine their initialRTN>T., the results of these tests are shown in Tables B 3'.4-1 and B 3/4.4-2.Reactor operation and resultant fast neutron (E greater than 1 HeV) irradiationcan cause an increase in the RTN>T. Therefore, an adjusted reference tempera

ture, based upon the fluence and chemistry content of .the material has beenpredicted using Regulatory Guide 1.99, Effects of Residual Elements on Pre-dicted Radiation Damage.to Reactor Vessel Materials." The heatup and cooldownlimit curves of Figures 3.4-2 to 3.4-5 include predicted adjustments for thisshift in RTNOT at the end of 10 EFPY as well as adjustments for possible errorsin the pressure and temperature sensing instruments.

TURKEY POINT - UNITS 3 ANO 4 B 3/4 4-8

L

TABLE 8 3/4.4-1REACTOR VESSEL TOUGHNESS OATA

TURKEY POINT - UNIT 3

Component

Mater ialType

Cu P NDTT

(<) (>) ('F)

50 ft lb/35 milsLateral Expansion

Tem F

Long Tr ansNDT

('F)

MinimumUpper Shelf

ft lbLong Trans

Cl. Hd. Dome A302 Gr. B

Cl. Hd, Flange A508 Cl. 2

Yes. Sh. Flange A508 Cl. 2

Inlet Nozzle A508 Cl. 2



Outlet Nozzle A508 Cl. 2


Outlet Nozzle

Upper Shell

A508 Cl. 2

A508 Cl. 2

0

44

23(a)

60(')

60(')

60(')

27(a)

7(a)

42(a)

50

Inter. Shell

Lower'hel 1

Trans. Ring

A508 C1. 2 0.079 0.010

A508 Cl. 2

30

60(a)

Bot. Hd. Dome A302 Gr. B

Inter. to Lower SAW

Shell Girth Weld0. 31 0. 011

-10

0('a)

A508 Cl. 2 0.058 0.010 40

„(a)

-41(')p

NA

9(a)

22(a)

23(')

44(a)

25(')

2(a)

58(')

NA

-23

60

60

60

27

50

40

30

60

30(')

> 70 > 45.5

«18 76.5(')

>120 > 78( )

NA — NA

NA NA

NA NA

>110 >71.5( )

111 72(')

140 91(')

>129 >83.5( )

122 >79(a)

163 106

>109 >70.5(')

HAZ HAZ 0(a) 168

(a) Estimated values based on NUREG-0800, Branch Technical Position - MTEB 52

ASLE 8 3/4.4-2REACTOR VESSEL TOUGHNESS DATA

TURKEY POINT - UNIT 4

ClM

I Component

Haterial

Type

Cu P HDTT

(>) (X) ('F)

50 ft lb/35 milsLateral Expansion

Tem 4F

Long TransHDT

('F)

NinimumUpper Shelf

ft lbLong Trans

C

l/l00

C7

IC)

Cl. Hd. Dome A302 Gr. 8

Cl. Hd. Flange A508 Cl. 2

Ves. Sh. Flange A508 Cl. 2


Inlet Hozzle A508 Cl. 2

Inlet Hozzle A508 Cl. 2

Outlet Nozzle A508 Cl..2



"20

4(a)

1(a)

eo')

eo(')

16(')

7(a)

38(')

eo(')

Upper Shell.

Inter. Shell

Lower Shell

. Trans. Ring

A508 Cl. 2 40

%08 Cl. 2 0.054 0.010 50

A508 Cl. 2 0.056 0.010

A508 Cl. 2

40

eo(')

Bot. Hd. Dome A302 Gr. 8

Inter. to Lower SAW

Shell Girth iield

HAZ

0. 31 0. 011

10

0('a)

a) Estimated values t 4 on NUREG-0800, Bra

HA

27(a)

-11(a)

NA

HA

13(a)

25(a)

16(a)

42(a)

32(')

90(a)

38(a)

30(a)

30(a)

63

nical Position - HTEB

'0

60

60

16

38

60

40

50

40

60

10

NA . NA

199 129

176 114

NA NA

HA NA

162 105

165 107(')

160 104( )

143 93

156 101( )

149 97(')

HA NA

NA NA

HA 63

NA 140


BASES

0

PRESSURE/TEHPERATURE LIMITS (Continued)

Values of IRTNOT determined in this manner may be used until the resultsfrom the material surveillance program, evaluated according to ASTH E185, areavailable. Capsules will be removed in accordance with the requirements ofASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsuleand the inner wall of the reactor vessel. Therefore, the results obtainedfrom the surveillance specimens can be used to predict future radiation damageto the reactor vessel material by using the lead factor and the withdrawaltime of the capsule. The heatup and cooldown curves must be recalculated whenthe hRTN>T determined from the surveillance capsule exceeds the calculated

BRTNOT for the equi val ent capsul e radiation exposure. The survei 1 1 ance capsul e"T" results from Unit 3 (WCAP 8631) and Unit 4 (SWRI 02-4221) were used togenerate the heatup and cooldown curves in Figure 3.4-2 through 3.4-5.

Allowable pressure-temperature relationships for various heatup andcooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASHE Boiler and Pressure Vessel Code as required by Appendix G

to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A.

The general method for calculating heatup and cooldown limit curves isbased upon the principles of the linear elastic fracture mechanics (LEFH)technology. In the calculation procedures a semielliptical surface defectwith a depth of one-quarter of the wall thickness, T, and a length of 3/2Tis assumed to exist at the inside of the vessel wall as well as at theoutside of the vessel wall. The dimensions of this postulated crack,referred to in'Appendix G of ASME Section III as the reference flaw, amplyexceed the current capabilities of inservice inspection techniques.Therefore, the reactor operation limit curves developed for this referencecrack are conservative and provide sufficient safety margins for protectionagainst nonductile failure. To assure that the radiation embrittlementeffects are accounted for in the calculation of the limit .curves, the mostlimiting value of the nil-ductility reference temperature, RTNpT is used

and this includes the radiation-induced shift, hRTNDT, corresponding tothe end of the period for which heatup and cooldown curves are generated.

The ASHE approach for calculating the allowable limit curves for variousheatup and cooldown rates specifies that the total stress intensity factor,KI, for the combined thermal and pressure stresses at any time during heatup

or cooldown cannot be greater than the reference stress intensity factor, KIR,for the metal temperature at that time. KIR is obtained from the reference

fracture toughness curve, defined in Appendix G to the ASHE Code. The KIR'urveis given by the equation:

TURKEY POINT - UNITS 3 AND 4 B 3/4 4-11 LI~ ~ ~


BASES

PRESSURE/TEMPERATURE LIMITS (Continued)

KIR 26.78 + 1.223 exp [0.0145(T-RTNDT 160)] o)Where: KIR is the reference stress intensity factor as a function of the metaltemperature T and the metal nil-ductility reference temperature RTNDT., Thus,the governing equation for the heatup-cooldown analysis is defined in Appendix Gof the ASME Code as follows:

C KIM + KIt c KIR (2)

Where: KIM= the stress intensity factor caused by membrane (pressure) stress,

KIt = the stress intensity factor caused by the thermal gradients,

KIR= constant provided by the Code as a function of temperature

relative to the RTNDT of the material,

C = 2.0 for level A and B service limits, and

C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, KIR is determined bythe metal temperature at the tip of the postulated flaw, the appropriate valuefor RTNDT, and the reference fracture toughness curve. The thermal stressesresulting from temperature gradients through the vessel wall are calculatedand then the corresponding thermal stress intensity factor, KIT, for thereference flaw is computed. From Equation (2) the pressure stress intensityfactors are obtained and, from these, the allowable pressures are calculated.

COOLDOWN

For the calculation of the allowable pressure versus coolant temperatureduring cooldown, the Code reference flaw is assumed to exist at the inside ofthe vessel wall. During cooldown, the controlling location of the flaw isalways at the inside of the wall because the thermal gradients produce tensilestresses at the inside, which increase with increasing cooldown rates. Allowablepressure-temperature relations are generated for both steady-state and finitecooldown rate situations. From these relations, composite limit curves areconstructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessarybecause control of the cooldown procedure is based on measurement of reactorcoolant temperature, whereas the limiting pressure is'actually dependent on thematerial temperature at the tip of the assumed flaw.'uring cooldown, the1/4T vessel location is at a higher temperature than the fluid ad)acent to thevessel ID. This condition, of course, is not true for the steady-state situa-tion. It follows that at any given reactor coolant temperature, the hTdeveloped during cooldown results in a higher value of K at the 1/4T locationIRTURKEY POINT - UNITS 3 AND 4 B 3/4 4-12'


BASES

PRESSURE/TEMPERATURE LIMITS (Continued}

for finite coo1down rates than for steady-state operation. Furthermore, ifconditions exist such that the increase in KIR exceeds KIt, the calculatedallowable pressure during cooldown will be greater than the steady-statevalue.

The above procedures are needed because there. is no direct control ontemperature at the 1/4T location; therefore, allowable pressures may unknowinglybe vio1ated if the rate of cooling is decreased at various intervals along acooldown ramp. The use of She composite curve eliminates this problem andassures conservative operation of the system for the entire cooldown period.

HEATUP~illa'hreeseparate calculations are required to determine the limit curves

for finite heatup rates. As is done in the cooldown analysis, allowablepressure-temperature relationships are developed for steady-state conditionsas well as finite heatup rate conditions assuming the presence of a 1/4Tdefect at the inside of the vessel wall. The thermal gradients during heatupproduce compressive stresses at the inside of the wall that alleviate thetensile stresses produced by internal pressure. The metal temperature at thecrack tip lags the coolant temperature; therefore, the KIR for .the 1/4T crackduring heatup is lower than the KIR for the 1/4T crack during steady-stateconditions at the same coolant temperature. During heatup, especially at theend of the transient, conditions may exist such that the effects of compressivethermal stresses and different KIR's for steady-state and finite heatup ratesdo not offset each other and the pressure-temperature curve based on steady-stateconditions no longer represents a lower bound of all similar curves for finiteheatup rates when the 1/4T flaw is considered. Therefore, both cases have tobe analyzed in or der to assure that at any coolant temperature the lower valueof the allowable pressure calculated for steady-state and finite heatup ratesis obtained.

The second portion of the heatup analysis concerns the calculation ofpressure-temperature limitations for the case in which a 1/4T deep outsidesurface flaw is assumed. Unlike the situation at the vessel inside surface,the thermal gradients established at the outside surface during heatup producestresses which are tensile in nature and thus tend to reinforce any pressurestresses present. These thermal stresses, of 'course, are dependent on boththe rate of heatup and the time (or coolant temperature) along the heatupramp. Furthermore, since the thermal stresses at the outside are tensile andincrease with increasing heatup rate, a lower bound curve cannot be defined.Rather, each heatup rate of interest must be analyzed on an individual basis.

TURKEY POINT - UNITS 3 AND 4 B 3/4 4-13 j~g ' ~ ~

~~

REACTOR COOLANT SYSTEH

BASES

g-)

PRESSURE/TEMPERATURE LIMITS (Continued)Following the generation of pressure-temperature curves for both the

steady-state and finite heatup rate situations, the final limit curves areproduced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At anygiven temperature, the allowable pressure is taken to be the lesser of thethree values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatuplimitations because it is possible for conditions to exist such that over thecourse of the heatup ramp the controlling condition switches from the insideto the outside and the pressure limit must at all times be based on analysisof the most critical criterion.

Finally, the new 10 CFR 50 Appendix G rule which addresses the metaltemperature of the closure head flange and vessel flange regions isconsidered. The rule states that the minimum metal temperature for the flangeregions should be at least 120 F higher than the limiting RTNOT for theseregions when the pressure exceeds 20 percen't of the preservice hydrostatictest pressure (621 psig). Since the limiting RTNpT for the flange regions forTurkey Point Units 3 and 4 is 44 F, the minimum temperature required forpressure of 621 psig and greater based on the Appendix G rule is 164 F. Theheatup and cooldown curves as shown in Figures 3.4-2 to 3.4"5 clearly satisfythe above requirement by ample margins.

Finally, the composite curves for the heatup rate data and the cooldownrate data are adjusted for possible errors in'he pressure and temperaturesensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those forwhich there is reason for concern of nonductile failure, operating limitsare provided to assure compatibility of operation with the fatigue analysisperformed in accordance with the ASHE Code requirements.

RVEIIPREEEURE RVRREERR PRIV AAVl IE, VVE''VV VlThe Technical Specifications provide requirements to isolate High

Pressure Safety Injection from the RCS

an4 to prevent thestart of an idle RCP if secondary temperature is more than 50 F above the RCScold leg temperatures. These requirements are designed to ensure that massand heat input transients more severe than those assumed in the lowtemperature overpressurization protection analysis cannot occur.

The OPERABILITY of two PORVs or an RCS vent opening of at least 2.20square inches ensures that the RCS will be protected from pressure transientswhich could exceed the limits of Appendix G to 10 CFR Part 50 when one or moreof the RCS cold legs are less than or equal to 275OF. Either PORV hasadequate relieving capability to protect the RCS from overpressurization whenthe transient is limited to either: (1) the start of an idle RCP with thesecondary water temperature of the steam generator less than or equal to 50'Fabove the RCS cold leg temperatures including margin for instrument error, or(2) the start of a HPSI pump and its injection''nto a water solid RCS.

TURKEY POINT - UNITS 3 ANP 4 B 3/4 4-14~ ~ 'V

E

~ 'aV


BASES

PRESSURE/TEMPERATURE LIMITS (Continued)

REACTOR MATERIAL SURVEILLANCE PROGRAM

Each Type I capsule contains 28 V-notch specimens, ten Charpy specimensmachined from each of the two shell forgings. The remaining eight Charpyspecimens are machined from correlated monitor material. In addition, eachType I capsule contains four tensile specimens (two specimens from each of thetwo shell forgings). Dosimeters of copper, nickel, aluminum-cobalt, andcadmium"shielded aluminum-cobalt wire are secured in holes drilled in spacersat the top, middle and bottom of each Type I capsule.

Each Type II capsule contains 32 Charpy V-notch specimens: eightspecimens machined from one of the shell forgings, eight specimens of weld'etal and eight specimens of HAZ metal, the remaining eight specimens arecorrelation monitors. In addition, each Type II capsule contains four tensilespecimens and four WOL forgings and the weld metal. Each Type II capsulecontains a dosimeter block at the center of the capsule. Two cadmium-oxide-shielded capsules, containing the two isotopes uranium-238'and neptunium-237,are contained in the dosimeter block. The double containment afforded by thedosimeter assembly prevents loss and contamination by the neptunium-237 anduranium-238 and their activation products. Each dosimeter block containsapproximately 20 milligrams of neptunium-237 and 13 milligrams of uranium238contained in a 3/8-inch OD sealed brass tube. Each tube is placed in a 1/2-inchdiameter hole in the dosimeter block (one neptunium-237 and one uranium-238tube per block), and the space around the tube is filled with cadmium oxide.After placement of this material, each hole is blocked with two 1/16-inchaluminum spacer discs and an outer 1/8-inch steel cover disc, which is weldedin place. Dosimeters of copper, nickel, aluminum-cobalt and cadmium-shieldedaluminum-cobalt are also secured in holes drilled in spacers located at thetop, middle and bottom of each Type II capsule.

Ca sule T e Ca sule Identification

IIIIII

IIIII

S

VT0XW

Y— Z



BASES

3/4.4. 10 STRUCTURAL INTEGRITY

The inservice inspection and testing programs for ASME Code Class 1, 2,and 3 components ensure that the structural integrity and operational readinessof these components will be maintained at an acceptable level throughout thelife of the plant. These programs are in accordance with Section XI of theASME Boiler and Pressure Vessel Code and applicable Addenda as required by10 CFR 50.55a(g) except where specific written relief has been granted bythe Commission pursuant to 10 CFR 50.55a(g)(6)(i .

Components of the Reactor Coolant System were designed to provide accessinservice inspections in accordance with Section XI of the ASME

Boiler and Pressure Vessel Code, 4%&-Edition and Addenda through winterM&h- Ll1Q 1'le

The surveillance requirements for post=RCS opening, modifications andrepairs ensure that RCS integrity is demonstrated following conditions thatmay ve affected system integrity.

For ormal opening, the integrity of th em, n terms of strength, isunch nged. If the system does not leak a 35 psig (o crating pressure + 100psi 100 ps is normal system pressure f ctuatlon)P it ill be leak tightduri g o al operation.

s or u 'tosli moanF repairs or components greater than< diameter, the thorough

nondestructive testing gives a very high degree of confidence in the integrityof the system, and will detect any significant defects in and near the newwelds.

/gal& fLan 0-ln445Repairs on componentsp4~4 in diameter or smaller are relatively minor

in comparison and the surface examination assures a similar standard ofintegrity. In all cases, the leak test will ensure leak tightness duringnormal operation.

3/4.4.11 REACTOR COOLANT SYSTEM VENTS

Reactor Coolant System vents are provided to exhaust ooncondensible gasesand/or steam from the Reactor Coolant System that could inhibit naturalcirculation core cooling. The OPERABILITY of least one Reactor Coolant Systemvent path froe the reactor vessel head and the pressurizer steam space ensuresthat the capability exists to perform this function,

The valve redundancy of the Reactor Coolant System vent paths serves tominimize the probability of inadvertent or irreversible actuation while ensuringthat a single failure of a vent valve, power supply, or control system does notprevent isolation of the vent path. The performances of the specifiedsurveillances will verify'the operability of the system.

The function, capabilities, and testing requirements of the Reactor CoolantSystem vents are consistent with the requirements of Item II.B.1 of NUREG-0737,"Clarification of TMI Action Plant Requirements," November 1980.

TURKEY POINT - UNITS 3 AND 4 8 3/4 4-16 -"-'4'9)ggg

0TURKEY POINT - UNITS 3 AND 4

3/4 5 EHERGENCY CORE COOLING SYSTEHS

BASES

3/4. 5. 1 ACCUMULATORS

The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensuresthat a sufficient volume of borated water will be immediately forced into thereactor core through each of the cold legs in the event the RCS pressure fallsbelow the pressure of the accumulators. This initial surge of water into thecore provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensurethat the assumptions used for accumulator injection in the safety analysis aremet.

e

The accumulator isolation valves fail to meet single failure criteria,therefore, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reasonexcept an isolation valve closed minimizes the time exposure of the plant to aLOCA event occurring concurrent with failure of an additional accumulator whichmay result in unacceptable peak cladding temperatures. If a closed isolationvalve cannot be immediately opened, the full capability of one accumulator isnot available and prompt action is required to place the reactor in a modewhere this capability is not required.

3/4. 5. 2 and 3/4. 5. 3 ECCS SUBSYSTEHS

The OPERABILITY of ECCS components and flowpaths required in Modes 1, 2and 3 ensures that sufficient emergency core cooling capability will be avail-able in the event of a LOCA assuming any single active failure consideration.Two SI pumps and one RHR pump operating in conjunction with two accumulatorsare capable of supplying sufficient core 'cooling to limit the peak claddingtemperatures within acceptable limits for all pipe break sizes up to andincluding the maximum hypothetical accident of a circumferential rupture of areactor coolant loop. In addition, the RHR subsystem provides long-term corecooling capability in the recirculation mode during the accident recoveryperiod

With the RCS temperature below 350 F, operation with less than fullredundant equipment is acceptable without single failure consideration on thebasis of the stable reactivity condition of the reactor and the limited corecooling requirements.

A/7 /gQ fjene5 olfocd /ol on overly'e don'ha/5llu'toom'P

boll Unii4 u)4o< Pne wpodrrn@liP'f a compo'~e~~(<)

o6ck= boisei u. 6 ~ ill eo„'/.e) r:f/,..

TURKEY POINT " UNITS 3 5 4 B 3/4 5-1


BASES

ECCS SUBSYSTEMS (Continued)

The Surveillance Requirements provided for each component ensures thatECCS OPERABILITY is maintained and verified periodically. Surveillance Require-ments for throttle valve position stops prevent total pump flow from exceedingrunout conditions when the system is in its minimum resistance configuration.

Pump performance requirements are obtained from accident analysis assumptions.Varying flowrates are provided to acco odate testing during modes and alignments.In the case of the 3600 gpm (normal o own mode) RHR test, differential headis specified in "feet". This criter a will allow for compensation of test datawith water density due to varying temp rature.

3/4.5.4 REFUELING WATER STORAGE TANKCOCi <

The OPERABILITY of the refueling water storage tank (RWST) as part of theECCS ensures that a sufficient supply of borated water is available for injec-tion by the ECCS in the event of a LOCA. The 'limits on RWST minimum volume andboron concentration ensure that: (1) sufficient water is available withincontainment to permit recirculation cooling flow to the core, and (2) thereactor will remain subcritical in the cold condition following mixing of theRWST and the RCS water volumes with all control rods inserted except for themost reactive control assembly. These assumptions are consistent with the LOCA

analyses.|.J c~ e

The 'ater volume limit includes an allowance f orwater notusable because of tank discharge line location or other physical characteristics.

The temperature limits on the RWST solution ensure that: 1) the solubilityof the borated water will be maintained, and 2) the temperature of the RWST

solution is consistent with the LOCA analysis. RQ@ie. in~tron e~tc4'ion. wo/Udc] 40 monitor't4'RQ57 icmmerclt" r8 ~

gl

TURKEY POINT - UNITS 3 4 4 8 3/4 5-2 ~UN us

3/4.6 CONTAINMENT SYSTE

BASES

ge c~h~'i ~ kgi~ Mss~ uf 5gp~,)~,g ~e } e if Ae. >4~~2~<<~ ~P~~ ~ r ajar L.ocA was a.s n uD

PMS ~~ g p„,. y< g 1~i'nmM (s Jsoui'Aw~ <be. kvhel Y+cii>vl ofp.s sg

~ ~3/4. 6. 1 PRIMARY CONTAINMENT

3/4. 6; l. 1 CONTAINMENT INTEGRITY

Primary CONTAINMENT INTEGRITY ensures that the release of radioactivematerials from the containment atmosphere will be restricted to those leakagepaths and associated leak rates assumed in the safety analyses. Thisrestriction, in conjunction with the leakage rate limitation, will limitthe SITE BOUNDARY radiation doses to within the dose guideline values of10 CFR Part 100 during accident conditions.

3/4. 6.1. 2 CONTAINMENT LEAKAGE

The limitations on containment leakage rates ensure that the totalcontainment leakage volume will not exceed the value assumed in the safetyanalyses at the peak accident pressure, P . As an added conservatism, themeasured overall integrated leakage rate is further limited to less than orequal to 0.75 L , during performance of the periodictest to account for possible degradation of the containment leakage barriersbetween leakage tests.

The surveillance testing for measuring leakage rates is consistent withthe requirements of Appendix J of 10 CFR Part 50.

3/4. 6.1. 3 CONTAINMENT AIR LOCKS

The limitations on closure and leak rate for the containment air locksare required to meet the restrictions on CONTAINMENT INTEGRITY and containmentleak rate. Surveillance testing of the air lock seals provides assurance thatthe overall air lock leakage will not become excessive due to seal damageduring the intervals between air lock leakage tests.

3/4.6.1.4 INTERNAL PRESSURE

T limi i s on cont en int ma pre sure ensure th t: (1ntai ent str ure i prevented om exceedjpg its dsign neyftive pressu e

diffe ential 2.5 p g with res ct to the odtside osphe 4, and 2') thecon inment ak pre ure does exceed t desi pressu of 59 sigdu ng LOC condit ns.

'h

maximu peak pres re expecte to be tained rom a L A ev t isk5 psi . For e limitin initial po itive ntainme pressu of psithe t tal pr sure is ca culated to e 52.9 sig. T s value as eenadjusted to 5 psig as he nominal tructural design pressure, wh ch i

ss than esi n essure an is consi tent with the safe analys

TURKEY POINT-UNITS 3 and 4 8 3/4 6-1 """ "" 1988

CONTAINMENT SYSTEMS

BASES

3/4.6.1.5 AIR TEMPERATURE

The limitations on containment average air temperature ensure that the aver-all containment average air temperature does not exceed the initial temperature

dl I dl 5 f y ly I f Lggh~g, )shall be made at all listed locations, whether by fixed or portable instruments,prfor to determining the average afr temperature.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY

This limitation ensures that, the structural integrity of the containmentwill be maintained comparable to the orfgfnal design standards for the life ofthe facility. Structural integrity is required to ensure that the containment

h d h t ll "" fgtgdiiif th I LAAA~JEThe measurement of containment tendon lift-offforce, the tensile tests of thetendon wires or strands, the visual examination of tendons, anchorages andexposed interior and exterior surfaces of the containment, and the Type Aleakage test are sufficient to demonstrate this capability.

The required Special Reports from any engineering evaluation of contain-ment abnormalities shall include a description of the tendon condition, thecondition of the concrete (especfally at tendon anchorages), the inspection5 d,th I Aid. 5 lt MW%leMldgevaluation, and the corrective actions take

( SJI J,„se~ 0~> mau

3~4 6 I 7 coNTAINMENT YENTILATIQN sYsTEM '

g,~ ~ALQ iusi)cc fo> ~ali crefgfe

person el p.qrThe containment purge supply and exhau t-fsbht8o ce-requtrud to

h I dd lg Lgdh~gf f1 5 h I Id I dd lgplant operation ensures that excessive quantltles of radloactlve aaterlals willnot be released via the Containment Purge System.

The total time the containment purge (vent) system isolation valves maybe open during HOOES 1, 2, 3, and 4 fn a calendar year fs a function ofanticipated need and operating experience.

Leakage integrity tests with a maximum allowable leakage rate for contain-ment purge supply and exhaust supply valves will provide early indication ofresilient material seal degradation and will allow opportunity for repair beforegross leakage failures could develop. The Oe60 L .leakage limit of Specifica-tion 3.6.1.2b. shall not be exceeded when the leakage rates determined by theleakage integrity tests of these valves are added to the previously determinedtotal for all valves and penetrations subject to Type B and C tests.

TURKEY POINT-UNITS 3 and 4 B 3/4 6-21 s ~

CONTAINMENT SYSTEMS

BASES

3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS

3/4.6.2. 1 CONTAINMENT SPRAY SYSTEM

The OPERABILITY of the Containment Spray System ensures that containmentdepressurization capability will be available in the event of a LOCA. Thepressure reduction and resultant lower containment leakage rate are consistentwith the assumptions used in the safety analyses.

The allowable out-of-service time requirements for the Containment SpraySystem have been maintained consistent with that assigned other inoperable ESFequipment and do not reflect the additional redundancy in cooling capabilityprovided by the Containment Cooling System. Pump performance requirements areobtained from the accidents-analysis assumptions.

CK<g~~cp'/4.

6. 2. 2 CONTAINMENT COOLING SYSTEM )cThe OPERABILITY of the Containment Cooling System ensures that adequate

heat removal capacity is available during past-LOCA conditions. The emergency=containment coolers are a full capacity system and are redundant to the spraysystem in terms of heat removal function for design basis accident.

The allowable out-of-service time requirements for the Containment CoolingSystem have been maintained consistent with that assigned other inoperable ESFequipment and do not reflect the additional redundancy in cooling capabilityprovided by the Containment Spray System.

3/4.6.3 EHERGENCY CONTAINMENT FILTERING SYSTEH

0

The OPERABILITY of the Emergency Containment Filtering System ensures thatsufficient iodine removal capability will be'vailable in the event of a LOCA.The reduction in containment iodine inventory reduces the resulting SITEBOUNOARY radiation doses associated with containment leakage. The operationof this system and resultant iodine removal capacity are consistent with theassumptions used in the LOCA analyses. System components are not subject torapid deterioration. Visual inspection and operating/performance tests aftermaintenance, prolonged operation, and at the required frequencies provideassurances of system reliability and will prevent system failure. Filterperformance tests are conducted in accordance'ith the methodology and intentof ANSI N510- 1975.

A Filter Efficiency of 90K is applicable to Turkey Point Safety Analysis.The HEPA filter efficiency combined with greater than 99K elemental iodineremoval and 30K organic iodine efficiency for the activated carbon filter willgive the applicable overall efficency.

3/4. 6. 4 CONTAINMENT ISOLATION VALVES

The OPERABILITY of the containment isolation valves ensures that the con-tainment atmosphere will be isolated from the outside environment in the event

TURKEY POINT-UNITS 3 and 4 B 3/4 6-3

CONTAINMENT SYSTEMS

BASES

CONTAINMENT ISOLATION VALVES (Continued)

of a release of radioactive material to the containment atmosphere or pressuri-zation of the containment.

3/4.6. 5 HYDRo&~v ~0~(7 op g )aThe OPERABILITY of the Hydrogen Monitors ensures the detection of hydrogen

buildup within containment following a LOCA to allow operator action to reducethe hydrogen concentration below its flammable limit.3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM

The OPERABILITY of the Post Accident Containment Vent System ensures thecapability for emergency venting of containment following a LOCh to reducethe hydrogen concentration to below its flaaeable limit.

k

PACVS systems components are not subject to rapid detirioration, havinglifetimes of many years, even under continuous flow conditions. Visualinspection and operating tests provide assurance of system reliability andwill ensure early detection of conditions which could cause the system tofail or operate improperly. The performance tests prove that filters havebeen properly installed, that no deterioration or damage has occurred, andthat all components and subsystems operate properly. The tests are performedin accordance with the methodology and intent of ANSI N510-1975 and provideassurance that filter performance has not deteriorated below requiredspecification values due to aging, contamination or other effects.

A filter efficiency of 99K is applicable to Turkey Point Safety Analysis.The HEPA filter efficiency combined with greater than 90K methyliodide removalefficiency for the activated carbon filter will give the applicable overallefficiency.

PTURKEY POINT-UNITS 3 and 4 B 3/4 6-4

':0>c;

3/4.7 PLANT SYSTEMS

BASES

3/4.7. 1.1 SAFETY VALVES

The OPERABILITY of the main steam line Code safety valves ensures that theSecondary System pressure will be limited to within 110K (1193.5 psig) of itsdesign pressure of 1085 psig during the most severe anticipated system opera-tional transient. The maximum ~elieving capacity is associated with a Turbinetrip from lOOX RATED THERMAL POWER coincident with an assumed loss of condenserheat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordancewith the requirements of Section VIII of the ASME Boiler and Pressure Code,1971 Edition. The total relieving capacity for all valves on all of the steamlines is 10,670,000 lbs/h which is 11'f the total secondary steam flow of9,600,000 lbs/h at 100K RATED THERMAL POWER. A minimum of two OPERABLE safetyvalves per steam generator ensures that sufficient relieving capacity isavailable for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperablewithin the limitations of the ACTION requirements on the basis of the reductionin Secondary Coolant System steam flow and THERMAL POWER required by thereduced Reactor trip settings of the Power Range Neutron Flux channels. TheReactor Trip Setpoint reductions are derived on the following bases:

Where:

SP

109

SP —X

x (109)

Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER,

Maximum number of inoperable safety valves per steam line,

Power Range Neutron Flux-High Trip Setpoint,

Total relieving capacity of all safety valves per steamline in 1bs/hour, and

Maximum relieving capacity of any one safety valve in lbs/hour

3/4.7.1. 2 AUXILIARYFEEDWATER SYSTEM

The OPERABILITY of the Auxiliary Feedwater System ensures that the ReactorCoolant System can be cooled down to less than 3504F from normal operatingconditions in the event of a total loss-of-offsite power. Steam can be

'TURKEY POINT - UNITS 3 4 4 - B 3/4 7-1

PLANT SYSTEMS

BASES.

supplied to the pump turbines from either or both units through redundantsteam headers. Two D. C. motor operated valves and one A. C. motor operatedvalve on each unit isolate the three main steam lines from these headers.Both the D.C..and A.C. motor operated valves are powered from safety-relatedsources. Auxiliary feedwater can be supplied through redundant lines to thesafety-related portions of the main feedwater lines to each of the steamgenerators. Air operated fail closed flow control valves are provided tomodulate the flow to each steam generator. Each steam driven auxiliary feed-water pump has sufficient capacity for single and two unit operation to ensurethat adequate feedwater flow is available to remove decay heat and reduce theReactor Coolant System temperature to less than 350 F when the Residual HeatRemoval System may be placed into operation.

ACTION statement 2 describes the actions to be taken when both auxiliaryfeedwater trains are inoperable. The requirement to verify the availabilityof both standby feedwater pumps is to be accomplished by verifying that bothpumps have successfully passed their monthly surveillance tests within thelast surveillance interval. The requirement..to complete this action beforebeginning a unit shutdown is to ensure that an alternate feedwater train isavailable before putting the affected unit through a transient. If no alter-nate feedwater trains are available, the affected unit is to stay at the samecondition until an auxiliary feedwater train is returned to service, and theninvoke ACTION statement 1 for the other train. If both standby feedwaterpumps are made available before one auxiliary feedwater train is returned toan OPERABLE status, then the affected unit(s) shall be placed i ANDBYwithin 6 hour&'and HOT SHUTDOWN within the following 6 hours.

ACTION statement 3 describes the actions to be taken when a singleauxiliary feedwater pump is inoperable. The requirement to verify that twoindependent auxiliary feedwater trains are OPERABLE is to be accomplished by

. verifying that the requirements for Table 3.7-3 have been successfully met foreach train within the last surveillance interval.

+he provisions of Specification 3.0.4 are )6 .

not applicable to the third auxiliary feedwater pump provided it has not beeninoperable for longer than 30 days. This means that a unit(s) can changeOPERATIONAL MODES during a unit(s) heatup with a single auxiliary feedwaterpump inoperable as long as the requirements of ACTION statement 3 are satisfied.

The monthly testing of the auxiliary feedwater pumps will verify theiroperability. Proper functioning of the turbine admission valve and the opera-tion of the pumps will demonstrate the integrity of the system. Verificationof correct operation will be made both from instr~ntation within the controlroom and direct visual observation of the pumps.

3/4.7.1.3 CONDENSATE STORAGE TANK

There are two (2) seismically designed 250,000 gallons condensate storagetanks. A minimum of 185,000 gallons is maintained in each tank. The OPERABIL-ITY of the condensate storage tank with the minimum water volume ensures that"' ~'<<'PPI" 'fo hot u-«f~dsi~v~td 'eoudlu 4, nt oj jq«f'.'. ~ ITURKEY POINT - UNITS 3 5 4,,B 3/4 7-2 ( /

ag7- gag, ugey ~ftl»" .,'..~ ',e;—,'.ago-~ a V ~ ~~AllAtua fV'Q ~Oi J>M ~Cf D /'~' ~'

PLANT SYSTEMS

BASES

CONDENSATE STORAGE TANK (Continued)

sufficient water is available to maintain the Reactor Coolant System at HOTSTANDBY conditions for approximately 23 hours or maintain the Reactor CoolantSystem at HOT STANDBY conditions for 15 hours and then cool'down the ReactorCoolant System to below 350'F at which point the Residual Heat Removal Systemmay be placed in operation.

3/4. 7. 1. 4 SPECIFIC ACTIVITY

The limit on secondary coolant specific activity is based on a postulated.release of secondary coolant equivalent to the contents of three steam genera-tors to the atmosphere due to a net load re)ection. The limiting dose for thiscase would result from radioactive iodine in the secondary coolant. One tenthof the iodine in the secondary coolant is assumed to reach the site boundarymaking allowance for plate-out and retention in water droplets. The inhalationthyroid dose at the site boundary is then;

Dose (Rem) = C V c B OFC ~ X/Q + 0.1

Where: C = secondary coolant dose equivalent I-131 specificactivity

0.2 curies/m3 (pCi/cc) or O.l Ci/m3, each unit

V = equivalent secondary coolant volume released = 214 m3

B = breathing rate = 3.47 x 10-~ m3/sec.

X/Q = 'tmospheric dispersion parameter = 1.54 x 10-~ sec/m3

0.1 = equivalent fraction of activity released

DCF = dose conversion factor, Rem/Ci

The resultant thyroid dose is less than 1.5 Rem.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES

The OPERABILITY of the main steam line isolation valves ensures that nomore than one steam generator +11 blow down in the event of a steam line .

rupture. This restriction is required to: (1} minfmiza the positive reac-tivity effects of the Reactor Coolant System cooldown associated with theblowdown, and (2) limit the pressure rise within containment .in the event thesteam line rupture occurs within containment. The OPERABILITY of the mainsteam isolation valves within the closure times of'he Surveillance Require-ments are consistent with the assumptions used in the safety analyses.

TURKEY POINT - UNITS 3 4 4 B 3/4 7-3

Au..t~~ CF'CT 44/mpeVmoiig 4~S

PLANT SYSTEMYviol'4iD 6'o d.s jo posifivcl~ stvo'icJ ~rg

ii'vp|i'cdisv si- cosil ag~tiovi ~P Qc Qc ink

Fee/~~ ggleann is, dcgi~ocJ LDY'cdiriBASES

S 4 g g CVYi ~isirCSV3/4.7.1.6 STANDBY FEEDWATER SY EN Sc-i rivi« I +~44 Chvi <srini

gus),FI c &isrs p.fa~ rack. esndvy csigThe. purpose of this speci ficati IinAPe s6/@rpngls ll n Ul

ments is to provide for administrative controls which will assure<and performance of the non-safety grade Standby Feedwater System. he Stan yFeedwater System consists of commercial grade components designed nd construct dto industry and FPL standards of this class of equipment located in the outdoo~plant environment typical of FPL facilities system wide. The system is expectedto perform with high reliability, i.e., comparable to that typically achievedwith this class of equipment. FPL intends to maintain the system in goodoperating condition with regard to appearance, structures, supports, componentmaintenance, calibrations, etc.

The function of the Standby Feedwater System for OPERABILITY determinationsis that it can be used as a backup to the Auxiliary Feedwater (AFW) System inthe event the AFW System does not function properly. The system would be ymanually starte and controlled by the operator when needed. In the event of a

g loss of offsite power the pumps can be powered via the non-safety grade diesel)generators. connected to the aon=ii4eeh 416D voig us.,

nm-+e-A supply of 60,000 gallons from the em r zed ater Storage Tank for

the Standby Feedwater Pumps is sufficient water to remove decay heat from thereactor for six (6) hours for a single unit or two (2) hours for two units.This was the basis used for requiring 60,000 gallons of water in the non-safetygrade Demineralized Water Storage Tank and is fudged to provide sufficient timefor restoring the AFW System or establishing make-up to the Demineralized WaterStorage Tank.

The motor. driven Standby Feedwater Pumps are not designed to NRC require-ments applicable to Auxiliary Feedwater Systems and not required to satisfydesign basis events requirements. These pumps may be out of service for up to24 hours before initiating formal notification because of the extremely lowprobability of a demand for their operation.

The guidelines for NRC notification in case of both pumps being out ofserv fo longer than 24 hours are provided in applicabl'e plant procedures,as p ~4-hour notification. E;

Adequate demineralized water for the standby feedwater system will beverified once per 24 hours. The Demineralized Water Storage Tank provides asource of water to several systems and therefore, requi es daily verification.

oP P-N,LEThe standby feedwater pumps will be verifie . onthly on a

STAGGERED TEST BASIS by starting and operating them n the recirculation modetypically from their normal power supply. Also, during each unit's refuelingoutage, the respective standby feedwater pump will be powered from the unit'sC b'us utilizing Units 1 and 2 non-safety grade diesel generators and flow

TURKEY POINT - UNITS 3 8L 4 B 3/4 7-4SU)t:'' "

~ ~ qg g~ ~

PlANT SYSTEMS

BASES

I ~ ~

aliis)l,

.J

STANDBY FEEDWATER. SYSTEM (Continued)

,tested to the nuclear unit's steam generators. Prior to this test, therefueling unit's C bus will be de-energized and the necessar loads will be

'ransferred to the other unit's C bus.

This surveillance regimen will thus demonstrate f theentire flow path, backup non-safety grade power supp y and pump associatedwith a unit at least each refueling outage. The pump, motor driver, andnormal power; supply availability would typically be dern s r b o erationof the pumps in the recirculation mode month on a a gered test basis.

3/4.7.2 COMPONENT COOLING WATER SYSTEM

EO $47%

The OPERABILITY of the Component Cooling Mater System ensures that suf-ficient cooling capacity is available for continued operation of safety-relatedquipment during normal and accident conditions. The redundant cooling

capac , assuming a singl failure, is consistent with theassumptions used in the safety analyses.

AD~ A tt A~~ l us% LV 4C +C

3/4.7.3, INTAKE COOLING WATER SYSTEM 8CL e ~Ci'OM

The OPERABILITY of the Intake Cooling Water System ensures that sufficientcooling capacity is available for continued operation of safety-relate equip-ment during normal and accident conditions. The o this system,assuming a single failure, 4e consistent with t e assum t e in the safetyanalyses. 'tom cnsupcs cooling co.po.~%~/4.. ULTIMATE HEAT SINK

im~ tions on the ultimate heat sink temperature ensure that-Suf ientcooling capa vailable either: (1) to provide norm e5ldown of thefacility or (2) to m te the effects of accident cond'5 within accepta lelimits.

The limitations on minimum wa er and maximum temperature are basedon providing a 30-day cooling-wat p~ ety-related equipment withouexceeding its design bpsis rature and fs .cons t with the recommend-ations of Regulator-G 1.27, "Ultimate Heat Sfqk for ear Plants,"March 1974.

TURKEY POINT - UNITS 3 8a 4 B 3/4 7-5 JuviQQ8

PLANT SYSTEMS

BASES

Evm~~3/4.7.5 CONTROL ROOM VENTILATION SYSTEM E~ev~~

The OPERABILITY of the Control Room>Ventllatfon System ensures that:(1) the ambient air temperature does not exceed the allowable temperature forcontinuous-duty rating for the equipment and instrumentation cooled by thissystem, and (2) the control room will remain habitable for operations personnelduring and following all credible accident conditions. The OPERABILITY of this,system in conjunction with control room design provisions is based on limitingthe 'radiation exposure to personnel occupying the control room to 5 rems orless whole body, or its equivalent. This limitation is consistent with therequirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50.

System components are not subject to rapid deterioration, having lifetimesof many years, even under continuous flow conditions. Visual inspection andoperating tests provide assurance of system reliability and will ensure earlydetection of conditions which could cause the system to fail. or operateimproperly. The filters performance tests prove that filters have been properlyinstalled,'that no deterioration or damage has occurred, and that all componentsand subsystems operate properly. The tests are performed in accordance withthe methodology and intent of ANSI N510 (1975) and provide assurance thatfilter performance has not deteriorated below returned specification values dueto aging, contamination, or other effects.

3/4.7.6 SNUBBERS

All snubbers are required OPERABLE to ensure that the structural integrityof the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads.

The visual inspection frequency is based upon maintaining a constant levelof snubber protection to each safety-related system during an earthquake orsevere transient. Therefore, the required, inspection interval varies inverselywith the observed snubber failures and is determined by the number of inoperablesnubbers found during an inspection. Inspections performed before that intervalhas elapsed may be used as a new reference point to dete~ine the next inspection.However, the results of such early inspections performed before the originalrequired time:interval has elapsed (nominal time less 25K) may not be used tolengthen the required inspection interval. Any inspection whose resultsrequire a shorter inspection interval will override'he previous schedule.

When the cause of the rejection of a snubber is visual inspection isclearly established and'emedied for the snubber and for any other snubbersthat may be generically susceptible, and verified operable by inservice func-tional testing, that snubber may be exempted from being counted as inoperablefor the purposes of establishing the next visual inspection interval.Generically susceptible snubbers are those which are of a specific make ormodel and have the same design features directly related to rejection of thesnubber by visual inspection, or are similarly located'or exposed to the sameenvironmental conditions such as temperature, radiation, and vibration.


Tha plant design inc)udes three CCM heat exchangars. However,ana)ysis results have shown that one pump and the combinedperformance of two heat exchangers will meet the coolingrequirements assumed in the accident analysis. A surveillanceprogram monitors the system capability to meet these requirements.The 51 day surveillance establishes the heat exchanger performance

'characteristics. The 12 hour surveillance monitors intake coolingwater inlet temperature and correlates it with the heat exchangerperformance characteristics and other system parameters to verifyheat removal capability. To provide additional assurance of heatexchanger availability~ all three heat «xchangers will be includedin this monitoring prograe.

/-.> 7< g~ge~7id~a.!!

~/'

PLANT SYSTEMS

BASES

When a snubber is found inoperable, an evaluation is performed, in addi-tion to the determination of the snubber mode of failure, in order to determineif any Safety Related System or component has been adversely affected by theinoperability of the snubber. The evaluation shall determine whether or notthe snubber mode of failure has imparted a significant effect or degradation onthe supported component or system.

To provide assurance of snubber functional reliability, a representativesample of the installed snubbers will be functionally tested during plantrefueling SHUTDOWNS. Observed failure of these sample snubbers shall requirefunctional testing of additional units.

In cases where the cause of the functional failure has been identifiedadditional testing shall be based on manufacturer's or engineering recommenda-tions. As applicable, this additional testing increases the probability oflocating possible inoperable snubbers without .testing 100K of the safety-related snubbers.

The service life of a snubber is established via manufacturer input andinformation through consideration of the snubber service conditions and asso-ciated installation and maintenance records (newly installed snubbers, sealreplaced, spring replaced, in high radiation area, in high temperature area,etc.). The requirement to monitor the snubber service life is included toensure that the snubbers periodically undergo a performance evaluation in viewof their age and operating conditions. These records will provide statisticalbases for future consideration of snubber service life. The requirements forthe maintenance of records and the snubber service life review are not intendedto affect plant operation.

3/4.7.7 SEALEO SOURCE CONTAMINATION

The limitations on removable contamination for sources requiring leaktesting, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits forplutonium. This limitation will ensure that leakage from Byproduct, Source,and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use,with Surveillance Requirements coaeensurate with the probability of damage to asource in that group. Those sources which are frequently handled are requiredto be tested more often, than those which are not. Sealed sources which arecontinuously enclosed within a shielded mechanism (i.e., sealed sources withinradiation monitoring or boron measuring devices) are considered to be storedand need not be tested unless they are removed from the shielded mechanism.

TURKEY POINT - UNITS 3 8a 4 B 3/4 7-7~Univ' g

PLANT SYSTEHS

BASES

3/4.7.8 FIRE SUPPRESSION SYSTEHS

("

The OPERABILITY of the Fire Suppression Systems ensures that adequatefire suppression capability is available to confine and extinguish firesoccurring in any portion of the facility where safety-related equipment islocated. The Fire Suppression System consists of the water system, spray,/ 8 k1,~fl h f . dy ~ tf hyd . )NLThe collective capability of the Fire Suppression Systems is adequate tominimize potential damage to safety-related equipment and is a major elementin the facility Fire Protection Program.

In the event that portions of the Fire Suppression Systems are inoperable,alternate backup fire-fighting equipment is required to be made available inthe affected areas until the inoperable equipment is restored to service.Mhen the inoperable fire-fighting equipment is intended for use as a backupmeans of fire suppression, a longer period of time is allowed to provide analternate means of fire fighting than if the inoperable equipment is theprimary means of fire suppression.

The Surveillance Requirements provide assurance that the minimum OPERABILITYrequirements of the Fire Suppression Systems are met.

In the-event the Fire Suppression Mater System becomes inoperable,immediate corrective measures must be taken since this system provides themajor fire suppression capability of the plant.

3/4. 7. 9 FIRE RATED ASSEHBLIES

The functional integrity of the fire rated assemblies and barrier penetrationsensures that fires will be confined or adequately retarded from spreading to

adjacer'ortionsof the facility. These design features minimize the possibility of a singfire rapidly involving several areas of the facility prior. to detection and extin-guishing of the fire. The fire barrier penetrations are 8 passive element in thefacility Fire Protection Program and are subject to periodic inspections.

Fire barrier penetrations, including cable penetration barriers, fire doorsand dampers are considered function 1 when the visually .observed condition is thes as the as-desi ne con tion

During periods of time when a barrier is not functional, either: (1) a contin-uous fire watch is required to be maintained in the vicinity of the affected bar~ieor (2) the fire detectors on at least one side of the affected barrier must beverified OPERABLE and an hourly fire watch patrol established until the barrieris restored to functional status.

TURKEY POINT - UNITS 3 4 4 B 3/4 7-8~ ~

3


BASES

3/4.9. 1 BORON CONCENTRATION

'he limitations on reactivity conditions during REFUELIN ensure that:(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) auniform boron concentration is maintained for reactivity control, in the watervolume having direct access to the reactor vessel. These limitations areconsistent with the initial conditions assumed for the boron dilution incidenin the safety analyses. he required valve dur ngrefueling operations-pea'&eden the possibility of uncontrolled bor n dilutionof the filled portion of the RC . This action pr vents flow to the RCS ofunborated water by closing flow paths f ource of unborated water.

~ ~3/4.9.2 INSTRUMENTATION

lS grec vThe OPERABILITY of the our e Range Neutron Flux Monitors ensures that

redundant monitoring capability is available to detect changes in the reactivitycondition of the core. There are four source range neutron flux channels, twoprimary and two backup. All four channels have visual and alarm indication inthe control room and interface with the containment evacuation alarm system.The primary source range neutron flux channels can also generate reactor tripsignals and provide audible indication of the count rate in the control roomand containment. At least one primary source range neutron flux channel toprovide the required audible indication, in addition to its other functions,and one of the three remaining source range channels shall be OPERABLE tosatisfy the LCO.

3/4. 9. 3 DECAY TIME

The minimum requirement for reactor subcriticality prior to movement ofirradiated fuel assemblies in the reactor vessel ensures that sufficient timehas elapsed to'allow the radioactive decay of the short-lived fission products.This decay time is consistent with the assumptions used in the safety analyses.

3/4. 9. 4 CONTAINMENT BUILDING PENETRATIONS

The requirements on containment building penetration closure and OPERABILITYensure that a release of radioactive material within containment will berestricted from leakage to the environment. The OPERABILITY and closurerestrictions are sufficient to restrict radioactive material release from afuel element rupture based upon the lack of containment pressurization potentialwhile in the REFUELING MODE.

3/4.9. 5 COMMUNICATIONS .

The requirement for communications capability ensures that refuelingstation personnel can be promptly informed of significant changes in thefacility status or core reactivity conditions during CORE ALTERATIONS.

TURKEY POINT - UNITS 3 8E 4 B 3/4 9-1


BASES

3/4.9.6 MANIPULATOR CRANE

The OPERABILITY requirements for the manipulator cranes ensure that:(1) manipulator cranes will be used for movement of drive rods and fuel assem-blies, (2) each crane has sufficient load capacity to lift a drive rod or fuelassembly, and (3) the core internals and reactor vessel are protected fromexcessive lifting force in the event they dna vertently engaged duringlifting operations.

(The requirement that t e Puxiliary jloist goad gndi a r be used to proven

'I fting excessive loads wII squire a manual action .T e juxiliaryPoist goadnNcator does not include a y aut tic h r e edtrical interlocks

that prevent lifting loads in cess 600 pounds.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS

The restriction on movement of loads in excess of the nominal weight of afuel and control rod assembly and associated handling tool over other fuelassemblies in the storage pool ensures that in the event this load is dropped:(1) the activity release will be limited to that contained in a single fuelassembly, and (2) any possible distortion of fuel in the storage racks will notresult in a critical array. This assumption is consistent with the activity

~ ~

release assumed in the safety analyses.

3/4.9.8 RESIOUAL HEAT REMOVAL ANO COOLANT CIRCULATION

The requirement that at least one residual heat removal (RHR) loop be inoperation ensures that: (1) sufficient cooling capacity is available to removedecay heat and maintain the water in the reactor vessel below 140 F as requiredduring the REFUELING MOOE, and (2) sufficient coolant circulation is maintainedthrough the core to minimize the effect of a boron dilution incident and preventboron stratification.

The requirement to have two RHR loops OPERABLE when there is less than23 feet of water above the reactor vessel flange ensures that a single failureof the operating RHR loop will not result in a complete loss of residual heatremoval capability. lith the reactor vessel head removed and at least 23 feetof water above the reactor pressure vessel flange, a large'eat sink is avail-able for core cooling. Thus, in the event of a failure of the operatingRHR loop, adequate time is provided to initiate emergency procedures to coolthe core.

0

3/4. 9. 9 CONTAINMENT VENTILATION ISOLATION SYSTEM

The OPERABILITY of this system ensures that the containment ventilationpenetrations will be automatically isolated upon detection of high radiationlevels within the containment. The OPERABILITY of this system is required torestrict the release of radioactive material from the containment atmosphere tothe environment.

TURKEY POINT - UNITS 3 & 4 B 3/4 9-2 J]'QNgy


BASES

3/4.9. 10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL~ ~

The restrictions on minimum water level ensure that sufficient waterdepth's available to remove 99K of the assumed 1(C iodine gap activityreleased from the rupture of an irradiated fuel assembly. The minimum waterdepth is consistent with the assumptions of the safety analysis.-

3/4.9.12 HANDLING OF SPERM FUEL CASK

Limiting spent fuel decay time to a minimum of 1,525 hours prior tomoving a spent fuel cask into the spent fuel pit will ensure that potentialoffsite doses are a fraction of 10 CFR Part 100 limits should a dropped caskstrike the stored fuel assemblies.

The restrjction to allow only a single element cask to be moved into thespent fuel pit will ensure the maintenance of water inventory in the unlikelyevent of an uncoqtgolled cask descent. Use of a single element cask whichnominally weig about twenty-five tons will also increase crane safetymargins by about a facto~ of four.

Requiring that spent fuel decay time be at least 120 days prior to movinga fuel assembly outside the fuel storage pit in a shipping cask will ensurethat potential offsite doses are a fraction of 10 CFR 100 limits should adropped cask and ruptured fuel assembly release activity directly to theatmosphere.

3/4. 9. 13 RADIATION MONITORING

PThe OPERABILITY,of ~containment radiation monitonC censures continuous

monitoring of radiation evels to provide immediate ind~ca ion of an unsafecondition.

)w]

3/4.9. 14 SPENT FUEL STORAGE

The spent fuel storage racks provide safe subcritical storage of fuelassemblies by providing sufficient center-to-center spacin'g or combinationof spacing and poison to assure k ff is equal to or less tha . for normaleff o.95operations and postulated accidents.

The spent fuel racks are divided into two regions. Region I racks havea 10.6 inch center-to-center spacing and Region II racks have a 9.0 inchcenter-to-center spacing. Because of the larger center-to-center spacing andpoison (B~o) concentration of Region I cells, the only restriction for place-ment of fuel is that the initial fuel assembly enrichment is. equal to or lessthan 4.5 weight percent of U-235. The limiting value of U-235 enrichmentis based upon the assumptions in the spent fuel safety analyses and assuresthat the limiting criteria for criticality is not exceeded. Prior to placement

)~f

TURKEY POINT - UNITS 3 8L 4 B 3/4 9-3


BASES

SPENT FUEL STORAGE (Continued)

in Region II cell locations, strict controls are employed to evaluate burnup ofthe spent fuel assembly, .Upon determination that the fuel assembly meets theburnup requirements of Table 3.9-1, placement in a Region II cell is authorized.These positive controls assure the fuel enrichment limits assumed in the

safety'nalyseswill not be exceeded. "

"This Technical Specification is applicable upon installation of the new two-region high density spent fuel racks.

TURKEY POINT - UNITS 3 8L 4 B 3/4 9-4~ o ~i'


BASES

3/4. 10. 1 SHUTDOWN MARGIN~ ~

This special test exception provides that a minimum amount of control rodworth is immediately available for reactivity contr'ol when tests are performedfor control rod worth measurement. This special test exception is required topermit the periodic verification of the actual versus predicted core reactivitycondition occurring as a result of fuel burnup or fuel cycling operations.

3/4. 10. 2 GROUP HEIGHT INSERTION AND POWER DISTRIBUTION LIMITS

This special test exception permits individual control rods to be positionedoutside of their normal group heights and insertion limits during the performanceof such PHYSICS TESTS as those required to measure control rod worth.

3/4. 10. 3 PHYSICS TESTS

This special test exception permits PHYSICS TESTS to be performed at lessthan or equal to SN of RATED THERMAL POWER with the RCS 7@ alkghtly lower

than normally allowed so that the fundamental nuclear characteristics of thecore and related instrumentation can be verified. In order for various charac-teristics to be accurately measured, it is at times necessary to operateoutside the normal restrictions of these Technical Specifications. For instance,to measure the moderator temperature coefficient at BOL, it is necessary toposition the various control rods at heights which may not normally be allowedby Specification 3. 1.3.6 which in turn may cause the RCS ~ to fall slightlybelow the minimum temperature of Specification 3. 1. 1.4.

3/4. 10.4 This s ecification number is not used.

Slq. 10. 5 POSITION INDICATION SYSTEM - SHUTDOWN

This special test exception permits the Position Indica son Systems to beinoperable during rod drop time measurements. The exception is required sincethe data necessary to determine the rod drop time are derived from the inducedvoltage in the position indicator coils as the rod is dropped. This inducedvoltage is small compared to the normal voltage and, therefore, cannot beobserved if the Position Indication Systems remain OPERABLE.

TURKEY POINT - UNITS 3 8E 4 B 3/4 10-1

/ y5 ~ g /p77+pz p(/~Wc l~4/ 6


BASES

3/4.11.1 LI UID EFFLUENTS

3/4 ~ 11. 1. 1 CONCENTRATION

This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents to UNRESTRICTED AREAS willbe less than the concentration levels specified in 10 CFR Part 20, Appendix B,Table II, Column 2. This limitation provides additional assurance that thelevels of radioactive materials in bodies of water in UNRESTRICTED AREAS willresult in exposures within: (1) the objectives of Appendix I, 10 CFR Part 50,to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to thepopulation. The concentration limit for dissolved or entrained noble gases isbased upon the assumption that Xe-135 is the controlling radioisotope and itsMPC in air (submersion) was .converted toian equivalent concentration in water,'sing the methods described in International Commission on Radiological Protec-

t tion (ICRP) Publication 2.

This specification applies to the release of radioactive materials inliquid effluents from all units at the site;The required detection capabilities for radioactive materials in liquid

waste samples are tabulated in terms of the lower limits of detection (LLDs).Detailed discussion of the LLD, and other detection limits can be found inCurrie, L. A., "Lower Limit of Detection: Definition and Elaboration of aProposed Position for Radiological Effluent and Environmental Measurements,"NUREG/CR-4077 (September 1984'), in HASL Procedures Manual, HASL-300 (revisedannually) and in Hartwell, J. K., "Detection Limits for Radioaana ytTcalCounting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215(June 1975).

3/4. 11. l. 2 DOSE

This specification is provided to implement the requirements of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guidesset forth in Appendix I. The ACTION statements provide the required operat-e

~

~~

~~ ~~ ~ing flexibilityand at the same time implement the guides set forth inppendix I to assure that the releases of radioactive material in liquidfluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achiev-

able." The dose calculation methodology and parameters in the ODCM implementthe requirements in Appendix I that conformance with the guides of Appendix Ibe shown by calculational procedures based on models and data, such that theactual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is .

unlikely to be substantially underestimated. The equations specified in theODCM for calculating the doses due to the actual release rates of radioactivematerials in liquid effluents are consistent with the methodology provided inRegulatory Guide 1.109, "Calculation of Annual Doses to Man from RoutineReleases of Reactor Effluents for the Purpose of Evaluating Compliance with10 CFR Part 50, Appendix I," March, 1976 and Regulatory Guide 1.113, "Estimat-ing Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releasesfor the Purpose of Implementing Appendix I," April 1977.

TURKEY POINT - UNITS 3 4 4 B 3/4 ll-lvv4 ' )95p


BASES

DOSE (Continued)

This specification applies to the release of radioactive materials in liquideffluents from each unit at the site. For units with shared Radwaste Systems,the liquid effluents from the shared system are to be proportional among theunits sharing that system.

3/4.11. 1.3 LI UID RADWASTE TREATMENT SYSTEM

The OPERABILITY of the Liquid Radwaste Treatment System ensures that thissystem will be available for use whenever liquid effluents require treatmentprior to release to the environment. The requirement that the appropriate portionsof this system be used when specified provides assurance that the releases ofradioactive materials in liquid effluents will be kept "as low as is reasonablyachievable." This specification implements the requirements of 10 CFR 50.36a,General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the ob)ectivesgiven in Section II.D of Appendix I to 10 CFR Part 50. The specified limitsgoverning the use of appropriate portions of the Liquid Radwaste TreatmentSystem were specified as a suitable fraction of the dose ob$ ectives setforth in Appendix I, 10 CFR Part 50 for liquid effluents.

This specification applies to the release of radioactive materials inliquid effluents from each unit at the site. For units with shared Radwaste System~the liquid effluents from the shared system are to be proportioned among theunits sharing that system.

TURKEY POINT - UNITS 3 5 4 B 3/4 11-2JUh' f l~g

I


BASES


3/4. 11. 2. 1 DOSE RATE

This specification is provided to ensure that the dose at any time at andbeyond the SITE BOUNDARY from gaseous effluents from all units on the sitewill be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS.The annual dose limits are the doses associated with the concentrations of10 CFR Part 20, Appendix 8, Table II, Column I. These limits pr'ovide reasonableassurance that radioactive material discharged in gaseous effluents will notresult in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA,either within or outside the SITE BOUNDARY, to annual average concentrationsexceeding the limits specified in Appendix B, Table II of 10 CFR Part 20(10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be withinthe SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually besufficiently low to compensate for any increase in the atmospheric diffusionfactor above that for the SITE BOUNDARY. Examples of calculations for suchMEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be givenin the ODCM. The specified release rate limits restrict, at all times, thecorresponding gamma and beta dose rates above background to a MEMBER OF THEPUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/yearto the whole body or to less than or equal to 3000 mrems/year to the skin.These release rate limits also restrict, at all times, the correspondingthyroid dose rate above background to a child via the inhalation pathway toless than or equal to 1500 mrems/year.

This specification applies to the release of radioactive materials ingaseous effluents from all units at the site.

The required detection capabilities for radioactive material in gaseouswaste samples are tabulated in terms of the lower limits of detection (LLDs).Detailed discussion of the LLD, and other detection limits can be found inCurrie, L. A , "Lower Limit of Detection: Definition and Elaboration of aProposed Position for Radiological Effluent and Environmental Measure tNUREG/ - 077 (September 1984), in HASL Procedures Manual, HASL-300

nd in Hartwell, J.K., "Oetection Limits for Radiosina ayt ca oon sngec nsques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4. 11.2. 2 DOSE - NOBLE GASES

This specification is provided to implement the requirements of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guidesset forth in Appendix I. The ACTION statements provide the required operat-ing flexibilityand at the same time implement the guides set forth in Appendix Ito assure that the releases of radioactive material in gaseous effluents toUNRESTRICTED AREAS will be kept "as low as is reasonably achievable." TheSurveillance Requirements implement the requirements in Appendix I thatconformance with the guides of Appendix I be shown by calculational procedures'ased on models and data such that the actual exposure of a MEMBER. OF THEPUBLIC thr ough appropriate pathways is unlikely to be substantially under-estimated. The dose calculation methodology and parameters established

TURKEY POINT - UNITS 3 4 4 8 3/4 11-3


BASES

DOSE-NOBLE GASES (Continued)

in the ODCM for calculating the doses due to the actual release rates ofradioactive noble gases in gaseous effluents are consistent with the methodologyprovided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man fromRoutine Releases of Reactor Effluents for the Purpose of Evaluating Compliancewith 10 CFR Part 50, Appendix I," Harch 1976, and Regulatory Guide 1.111,"Methods for Estimating Atmospheric Transport and Dispersion of GaseousEffluents in Routine Releases from Light-Water Cooled Reactors," Revision 1,July 1977. The ODCM equations provided for determining the air doses at andbeyond the SITE BOUNDARY are based upon the historical average atmosphericconditions.

This specification applies to the release of radioactive materials ingaseous effluents from each. unit at the site. For units with shared radwastetreatment systems, the gaseous effluents from the shared system are proportionedamong the units sharing that system.

3/4. 11.2. 3 DOSE - IODINE-131 IODINE-133 TRITIUM AND RADIOACTIVE MATERIALN UL FORH

This specification is provided to implement the requirements of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides setforth in Appendix I. The ACTION statements provide the required operatingflexibilityand at the same time implement the guides set forth in Appendix Ito assure that the releases of radioactive materials in gaseous effluents toUNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM

calculational methods specified in the Surveillance Requirements implement therequirements in Appendix I that conformance with the guides of Appendix I beshown by calculational procedures based on models and data such that the actualexposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely tobe substantially underestimated. The ODCM calculational methodology andparameters for calculating the doses due to the actual release rates of thesubject materials are consistent with the methodology provided in RegulatoryGuide 1. 109, "Calculation of Annual Doses to Man from Routine Releases ofReactor Effluents f he Purpose of Evaluating Compliance with 10 CPR Part 60,.Appendix 1," Rare 96 , and Regulatory Guide 1.111, "Rethods for EstimatingAtmospher spor and Dispersion of Gaseous Effluents in Routine Releases

ght-Water Cooled Reactors," Revision 1, July 1977. ~ These equations alsoprovide for determining the actual doses based upon the historical averageatmospheric conditions. The release rate specifications for Iodine-131Iodine-133, tritium, and radionuclides in particulate form with half-livesgreater than 8 days are dependent upon the existing radionuclide pathways toman in the areas at and beyond the SITE BOUNDARY.- The pathways that wereexamined in the development of the calculations wero: (1) individual inhala-tion of airborne radionuclides, (2) deposition of radionuclides onto greenleafy vegetation with subsequent consumption by man, (3) deposition onto grassyareas where milk animals and meat producing animals graze with consumption ofthe milk and meat by man, and (4) deposition on the ground with subsequentexposure of man.

TURKEY POINT - UNITS 3 Si 4 B 3/4 11-4~ > '": E'eak


BASES

DOSE " IODINE-133 TRITIUM ANO RADIOACTIVE MATERIAL IN PARTICULATE FORM

Continued

This specification applies to the release of radioactive materials ingaseous effluents from each unit at the site. For units with shared radwastetreatment systems, the gaseous eff1uents from the shared system are proportionedamong the units sharing that system.

3/4. 11.2.4 GASEOUS RADWASTE TREATHENT SYSTEM

The OPERABILITY of the GAS DECAY TANK SYSTEM and the VENTILATION EXHAUSTTREATHENT SYSTEM ensures that the systems will be available for use whenevergaseous effluents require treatment prior to release to the environment. Therequirement that the appropriate portions of these systems be used, when specified,provides reasonable assurance that the releases of radioactive materials ingaseous effluents will be kept "as low as is reasonably achievable." Thisspecification implements the requirements of 10.CFR 50.36a, General DesignCriterion 60 of Appendix A to 10 CFR Part 50 and the ob)ectives given inSection II.D of Appendix I to 10 CFR Part 50. The specified limits governingthe use of appropriate portions of the systems were specified as a suitablefraction of the dose objectives set forth in. Appendix I, 10 CFR Part 50, forgaseous effluents.

This specification applies to the release of radioactive materials ingaseous effluents from each unit at the site. For units with shared radwastetreatment systems, the gaseous effluents from the shared system are proportionedamong the units sharing that system.

TURKEY POINT - UNITS 3 5 4 B 3/4 11-5 tl I ~ ~)ge


BASES

3/4. 11. 2. 5 EXPLOSIVE GAS MIXTURE

This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the GAS DECAY TANK SYSTEM (asmeasured in the inservice gas decay tank) is maintained below the flammabilitylimits of hydrogen and oxygen. Maintaining the concentration of hydrogen andoxygen below their flammability limits provides assurance that. the releases ofradioactive materials will be controlled in conformance with the requirementsof General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4 11 ~ 2.6 GAS DECAY TANKS

The tanks included in this specification are those tanks for which thequantity of radioactivity contained is not limited directly or indirectly byanother Technical Specification. Restricting the quantity of radioactivitycontained in each Gas Decay Tank provides assurance that in the event of anuncontrolled release of the tank's contents, the resulting whole body exposureto a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem.

3/4. 11. 3 SOLID RADIOACTIVE WASTES

This specification implements the requirements of 10 CFR 50.36a andGeneral Design Criterion 60 of Appendix A to 10 CFR Part 50. The processparameters included in establishing the PROCESS CONTROL PROGRAM may include,but are not limited to, waste type, waste pH, waste/liquid/SOLIDIFICATIONagent/catalyst ratios, waste oil content, waste principal chemical constituents,and mixing and curing times.

3/4. 11. 4 TOTAL DOSE

This specification is provided to meet the dose limitations of 10 CFRPart 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. Thespecification requires the preparation and submittal of a Special Report when-ever the calculated doses due to releases of radioactivity and to radiation fromuranium fuel cycle sources exceed 25 mrems to the whole body or any organ, exceptthe thyroid, which shall be limited to less than or equal to 75 mrems. Forsites containing up to four reactors, it is highly unlikely that the resultantdose to a MEMBER OF THE PUBLIC All exceed the dose limits of 40 CFR Part 190if the individual reactors remain within twice the dose design objectives ofAppendix I, and if direct radiation doses from the units > ~

~



BASES

TOTAL DOSE (Continued)

are kept small. The Special Report will describe a courseof action that should result in the limitation of the annual dose to a MEMBER

OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of theSpecial Report, it may be assumed that the dose commitment to the MEMBER ofthe PUBLIC from other uranium fuel cycle sources is negligible, with theexception that dose contributions from other nuclear fuel cycle facilities atthe same site or within a radius of 8 km must be considered. If the dose toany MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFRPart 190, the Special Report with a request for a variance (provided therelease conditions resulting in violation of 40 CFR Part 190 have not alreadybeen corrected), in accordance with the provisions of 40 CFR 190. 11 and 10 CFR20.405c, is considered to be a timely request and fulfills the requirements of40 CFR Part 190 until NRC staff action is completed. The variance only relatesto the limits of 40 CFR Part 190, and does not apply in any way to the otherrequirements for dose limitation of 10 CFR Part 20, as addressed in Specifi-cations 3. 11. 1. 1 and 3.11.2.1. An individual is not considered a MEMBER OF THEPUBLIC during any period in which he/she is engaged in carrying out any operationthat is part of the nuclear fuel cycle.

TURKEY POINT - UNITS 3 8( 4 B 3/4 11-7

~- @DC/P- In7en(idnal/.~I

/


BASES

3/4. 12. 1 MONITORING PROGRAM

The Radiological Environmental Monitoring Program required by thisspecification provides representative measurements of radiation and of radio-active materials in those exposure pathways and for those radionuclides thatlead to the highest potential radiation exposure of MEMBERS OF THE PUBLICresulting from the plant operation. This monitoring program implementsSection IV.B. 2 of Appendix I to 10 CFR Part 50 and thereby supplements theRadiological Effluent Monitoring Program by verifying that the measurableconcentrations of radioactive materials and levels of radiation are not higherthan expected on the basis of the effluent measurements and the modeling ofthe environmental exposure pathways.

The required detection capabilities for environmental sample analyses aretabulated in terms of the lower limits of detection (LLDs). The LLDs requiredby Table 4. 12-1 are considered optimum for routine environmental measurementsin industrial laboratories. It should be recognized that the LLD is definedas an a griori (before the fact) limit representing the capability of a measure-

s d ~ i l(ft tt f )ill f pmeasurement.

Detailed discussion of the LLD, and other detection limits, can be foundin Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of aProposed Position for Radiological Effluent and Environmental Measurem tNUREG/ -400? (September 1984), in HASL Procedures Manual, HASL-300 ~itch

and Hartwell, J. K. "Oetection Limits for RadioanaTyytsca Coun snechnsques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4. 12.2 LAND USE CENSUS

This specification is provided to ensure that changes in the use of areasat and beyond the SITE BOUNDARY are identified and that modifications to theRadiological Environmental Monitoring Program are made if required by theresults of this census. The best information from the door-to-door survey,from aerial survey or from consulting with local agricultural authoritiesshall be used. This census satisfies the requirements of Section IV.B.3 ofAppendix I to 10 CFR Part 50. Restricting the census to gardens of greaterthan 50 m~ provides assurance that significant exposure pathways via leafyvegetables will be identified and monitored since a garden of this size is theminimum required to produce the quantity (26 kg/year) of leafy vegetablesassumed in Regulatory Guide 1.109 for consumption by a child. To determinethis minimum garden size, the following assumptions were made: (1) 20X of thegarden was used for growing broad leaf vegetation (i.e., similar to lettuce andcabbage), and (2) a vegetation yield of 2 kg/m2.

TURKEY POINT - UNITS 3 8( 4 B 3/4 12-1


BASES

3/4. 12.3 INTERLABORATORY COMPARISON PROGRAM

The requirement for participation in an approved Interlaboratory ComparisonProgram .is provided to ensure that independent checks on the precision andaccuracy of the measurements of radioactive materials in environmental samplematrices are performed as part of the quality assurance program for environ-mental monitoring in order to demonstrate that the results are valid for thepurposes of Section IV. B. 2 of Appendix I to 10 CFR Part 50. This condition issatisfied by participation in the Environmental Radioactivity LaboratoryIntercomparison Studies Program conducted by the Environmental ProtectionAgency (EPA).

0

,TURKEY POINT " UNITS 3 5 4 B 3/4 12-2

SECTION 5.0

DESIGN FEATURES

~ t

@CD)M [ W+CV:9 l p YC<

~~

7" Q C(ln, Q

T. S. 3 4.3.2

Justifications:1 ~

2.

The CTS does not address "allowable" values. The 1986version of RTS adopted the STS 2-column format containingnominal setpoints and allowables. The RTS allowableswere best estimate numbers determined by Westinghouseusing generic tolerances. However, FPL would requireadditional analysis to adequately support these values.

This change clarifies an oversight. It has been revisedto be consistent with Functional Unit 4.d of Table 3.3-2 ~

3 ~

4 ~

Use of jumpers; fuses to place channel in trip conditionwas agreed to by NRC technical reviewer in FPL/NRCworking meetings.

For some channels, (e.g., containment press hi), ananalog channel operational test is not applicable.

5. There is no testing of relays that can be accomplishedbeyond that done by ACTUATION LOGIC TEST.

6. A monthly TRIP ACTUATING DEVICE OPERATIONAL TEST cannotbe performed without initiating an actual high pressurecondition. This is performed during refuelingcalibration.

7.

8.

With deletion of the allowable value column, the tripsetpoints are revised to reflect the CTS values.

The automatic actuation logic and actuation relays arerequired surveillances in MODES 3 and 4; however, theyare not testable until MODE 3. This problem exists withitems 1.b, 1.d through l.f, 4.b, 4.d, 8.a and 8.b.

9. Several instruments do not have independent channelsmeasuring the same parameter. To perform a CHANNEL CHECKon these instruments, a known source is needed tomanually check each instrument with an independentchannel.

10. Functional Unit 9.a is not testable for ACTUATION LOGICTEST of radiation monitors.

High Ste'am Flow-Coincident with Steam Generator Pressure-Low or Tzpg Low can only be blocked for safetyinjection. Steam Line Isolation cannot be blocked.

12. With deletion of the allowable value column, the tripsetpoints are revised to reflect the CTS values.

13. BASES — The deleted wording provides no necessaryinformation.

14. There are no control room isolation functions associatedwith high-containment radioactivity monitors during fuelmovement in the spent fuel pool.

Justifications:1 ~ The monitor numbers were removed to prevent making a

License Amendment if instrument make/model changes aremade in the future.

2. Several instruments do not have independent channelsmeasuring the same parameter. To perform a CHANNEL CHECKon these instruments, a known source is needed tomanually check each instrument with an independentchannel.

T. S. 3 4.3.3.3

Justifications:1 ~ Two different specifications exist for the

same'equirement3.3.3.3 and 3.6.5. Specification 3.0.4 isnot applicable to Table 3.3-5, but it is applicable toSpecification 3.6.5. This is a source of confusion thatneeds to be eliminated.

2 ~

3 ~

The accident monitoring instruments do not providetrip/active functions; therefore, there is no need toconduct the ANALOG CHANNEL OPERATIONAL TEST whichrequires simulated input signals to test interlock/alarm/trip functions.

The refueling surveillance is consistent with the STS.

4 ~ The note describes the only possible check that can beperformed because there is no visible indication duringnormal conditions.

5.

6.

This note allows the neutron detector to be excluded fromchannel calibration. This calibration is not practical.Comparing wide range instruments during normal operationis not a meaningful CHANNEL CHECK. A source must be usedto obtain a response. This change is consistent withSTS.

7. This surveillance is inconsistent with the LCO's whichrequire these channels only in Modes 1, 2 and 3.

8. The monitor numbers were removed to prevent making aLicense Amendment if instrument make/model changes aremade in the future.

9. For consistency of reporting requirements, 30 days isallowed for submittal of the report.

Justifications:

0 1 ~ Failure to comply with the ACTION Statement is reportablevia 10 CFR 50.73. This was a CTS requirement, not an STSrequirement.

2.

3

Containment entry to inspect the fire zone is consistentwith the containment temperature surveillance frequency.(4.6.1.5) and is justified based on ALARA considerationsand the likelihood of recognizing fire concerns throughother instrumentation/equipment feedback.

Fire watch patrol has been replaced by additionalinstrumentation as part of Appendix R upgrades.

Justifications:1. The monitor numbers were removed to prevent making a

License Amendment if instrument. make/model changes aremade in the future.

2 ~ Several instruments do not have independent channelsmeasuring the same parameter. To perform a CHANNEL CHECKon these instruments, a known source is needed tomanually check each instrument with an independentchannel.

3 ~ Applicability should be limited to those situations whenreleases are made via these pathways. This is consistentwith CTS (Amendment 103/97).

T. S. 3 4.3.3.6

Justifications:The iodine and particulate samplers are counted filters,not online monitors. Their function is covered in Table4.11-2.

2. Several instruments do not have independent channelsmeasuring the same parameters. To perform a CHANNELCHECK on these instruments, a known source is needed tomanually check each instrument with an independentchannel.

3.

4.

This is a continuous release pathway and should bemonitored daily as well as prior to any release.

Four (4) hours would require upgrades to equipment.Eight (8) hours is a reasonable amount of time to performthis test.

5. The monitor numbers were removed to prevent making aLicense Amendment if instrument make/model changes aremade in the future.

Justifications:1. BASES — This clarification is consistent with the CTS.

Justifications:1. Refer to attached Westinghouse letters.2. The single active failure wording is not applicable to

Turkey Point due to electrical system design.

Nestfnghouse Water ReactorBectrlc Corporstlcn Dlvlslons

Nr, Steve CraigFlorida Power and Light CompanyP.O, Box 013100Miami, FL 33}52

Mucker fuel Divigeii

kx Sl12N1thlglPIN4ylrloll15250 $ 912

84F'P*-9-}28October 8, 1984Keywords: FPL Rod-NthdrawalReference: 1 FPL-84-729> 8/15/84

2 NS-EPR-2935) 7/9/843 FP-FP-697

82FP~-G-011

Dear Nr, Craig;

FLORIDA POWER AND LIGHT COMPANY,TURKEY POINT UNITS 3 AND 4

PREVENTION OF ROD MIT DRAWL FROM SUBCRITICAL DURIN MODE 3

The purpose of this memo is to provide additional requested informationregarding an incons1stency between the safety analysis and the Tech Specswith respect to the number of operating reactor coolant pumps when in Node 3(or the equivalent) as defined 1n the Tech Specs. Reference ) provided thelatest update on th1s issue. The purpose of this letter is to provideconfirmation of the safety analys1s assumptions and FPL's proposedadministrative procedure to address this issue, as discussed with Nr. RickMande, Reactor Kngineer. at Turkey Point.

As discussed in Reference 2, a latter from ifestinghouse to the NRC, the onlyevent impacted by this issue is the RCCA bank withdrawal from subcritical.This accident is defined as an uncontrolled addition of reactivity to thereactor core caused by withdrawal of RCChs resulting in a power excur s1on.Such a transient could be caused by a malfunction oT the automat1c rodcontrol system or by operator e~~or. The analysis assumes a react1vityinsert1on rate of 75 pcm/sec, greater than the maximum resulting from thesimultaneous withdrawal of the combinatior of two sequent1al control bankshaving the maximum combined worth at maximum speed.

ha most recent RCCA bank withdrawal from subcr'itical analysis for TurkeyPoint Units 3 hard 4 was performed as par'. of the positive HTC study in1981 (Reference 3),.and assumes that all three reactor coolant pumps are

,operating. However, the Turkey Point Tech Specs state that "In Hothutdown at les.', '.wo Reactor Coolant L."~ps shall ba operable and at least

one Rear-or C~c.'.!~: Loon shall be in cra~at1on." (T.S. 3.4. l.d, pg 3.4-2A).Thus, the Tech 5;a.s are inconsistent w5xh the analysis assumption. Asnoted in Refers«e P, it is not a realms«c requirement to have all threereactor cools~'.: ~, ~ ooerating whar t~~ plant is cooling down prior to goingto Cold Shutdciv''. ln aoait1on, administrative procedures are preferable to

. phys1cal prevention of rod withdrawal for Turkey Point> s1nce the plant

5. 0 DESIGN FEATURES

5. 1 SITE

EXCLUSION AREA

5. 1. 1 The Exclusion Area shall be as shown in Figure 5. . 1

LOW POPULATION ZONE

5. 1. 2 The Low Population Zone shall be as shown in Figure 5.1-1.

MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS ANDLI UID EFFLUENTS

5.1.3 Information regarding radioactive gaseous and liquid effluents, whichwill allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible toMEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-1.

5. 2 CONTAINMENT

CONFIGURATION

5.2.1 The containment building is a steel-lined, reinforced concrete buildingof cylindrical shape, with a dome roof and having the following designfeatures:

a. Nominal inside diameter = 116 feet.

b. Nominal inside height = 170.5 feet

c. Minimum thickness of concrete walls = 3.75 feet.

d. Minimum thickness of concrete roof = 3.25 feet.

e. Minimum thickness of concrete floor pad = 10.5 feet.

f. Nominal thickness of steel liner = 0.25 inches.

g. Net free volume = 1 550 000 cubic feet.

DESIGN PRESSURE AND TEMPERATURE

5.2.2 The containment building is designed and shall be maintained for amaximum internal pressure of 59 psig and a temperature of 2834F. The con-tainment building is also structurally designed to withstand an internalvacuum of 2.5 psig.

I

TURKEY POINT - UNITS 3 8 4 5-1 JUl' &;:,a;

FIGURE 5.1-1 SITE AREA HAP


DESIGN FEATURES

5. 3 REACTOR CORE

FUEL ASSEMBLIES

speci ic relcrrg o r . i q-1

.',ejlP~r".if.'n',

/age gal I bing] 'f ~pi Si <~~KVL'l Q") ig, . s, ' Qg ) ])gp.corbi ~ m; cicu. AoQ <Teei. o( h/ '!'r i rccY'hJ s) <" t 'y'rnf'Q [p, gggI

Omit /PIC-sy-. 'eJ

5.3.1 The core shall contain 157 fuel assam ies with each fuel assembly nomnoll]containing 204 fuel rods clad with Zircaloy-4 The reactor core containsapproximately 71 metric tons of uranium in t form of natural or slightlyenriched uranium dioxide pellets. Each fuel rod shall have a nominal activefuel length of 144 inches. 4

CONTROL ROD ASSEMBLIES

5.3.2 The core shall contain 45 full-length control rod assemblies. Thefull-length control rod assemblies shall contain a nominal 142 inches ofabsorber material. The absorber material shall be silver, in ium, andcadmium. All control rods shall be ad withstainless steel tubing.

5.4 REACTOR COOLANT SYSTEMI ~

DESIGN PRESSURE AND TEMPERATURE

q p ection 4. 1 ofthe FSAR, with allowance for normal degradation pursuant to theapplicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650 F, except for the pressurizer which is6800F.

5.4. 1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code re uirements s ecified in S

VOLUME

5.4.2 The nominal water and steam volume of the Reactor Coolant System is9343 cubic feet at a nominal T „ of 574.2 F.

5. 5 METEOROLOGICAL TOWER LOCATION

5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1.

!~!/ ~

i

wr'

URKEY POINT UNITS 3 g 4 5 3L

DESIGN FEATURES

5.6 FUEL STORAGE

5. 6. 1 CRITICALITY.

5.6.1.1 The spent fuel storage racks are designe to provide safestorage of fuel assemblies by providing suffic t center-to-centera combination of spacing and poison and shall maintained with:

subcriticalspacing or

a. A k ff equivalent to less than or equal to 0.95 when flooded witheffunborated water, which includes a conservative allowance of 2.55Xhk/k for uncertainties for single region spent fuel storage racks.

b. A k f equivalent to less than or equal to 0.95 when flooded withunb3rfted water, which includes a conservative allowance in region 1of 2.66K hk/k ance in region 2 of 2.94K 4k/k for uncertainties for tworegion fuel storage racks.

c. A nominal 13.7 inch center-to-center distance between fuel assem-blies placed in the single-region storage racks. A nominal 10.6inch center-to center distance for Region 1 and 9.0 inch center-to-center distance for Region 2 for two region fuel storage racks.

d. Fuel assemblies stored in the single-region spent fuel storage racksshall contain no more than 4.1 weight percent of U-235.

e. After installation, of the two-region high density spent fuel storageracks, the maximum enrichment loading for fuel assemblies is 4.5weight percent of U-235.

5.6.1.2 The racks for new fuel storage are designed to store fuel in a safesubcritical array and shall be maintained with:

a. A nominal 21 inch center-to-center spacing to assure k ff equal toor less than 0.98 for optimum moderation conditions and equal to orless than 0.95 for fully flooded conditions.

b. Fuel assemblies placed in the New Fuel Storage brea shall contain nomore than 4.5 weight percent of U-235.

TURKEY POINT - UNITS 3 & 4

DESIGN FEATURES

5.6.1.3 Credit for burnup is taken in determining placement locations forspent fuel in the two-region spent fuel racks." Administrative controls areemployed to evaluate the burnup of each spent fuel assembly stored in areaswhere credit for burnup is taken. The burnup of spent fuel is ascertained bycareful analysis of burnup history, prior to placement into the storage loc-cations. Procedures shall require an independent check of the analysis ofsuitability for storage. A complete record of such analysis is kept for thetime period that the spent fuel assembly remains in storage onsite.

DRAINAGE

5.6.2 The spent fuel storage pit is designed and shall be maintained toprevent inadvertent draining of the pool below a level of 6 feet above thefuel assemblies in the storage racks.

CAPACITY

5.6.3 The spent fuel storage pool is designed and shall be maintained with astorage capacity limited to no more than 621"" fuel assemblies in one regionstorage racks or 1404 in two region storage. racks

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT

5.7.1 The components identified in Table 5.7-1 are designed and shall bemaintained within the cyclic or transient limits of Table 5.7-1.

During rack installation, it will be necessary to .temporarily store Region Ifuel in the Region II spent fuel racks. Administrative controls will beutilized to maintain a checkerboard storage configuration, i.e., alternate celloccupation, in the Region II racks.

""The fuel assembly storage capacity for Unit 4 single region storage racks is614.

TURKEY POINT - UNITS 3 4 4 5"5

~ l~rTABLE 5.7-1

COMPONENT CYCLIC OR TRANSIENT LIMITS

ClMCOMPONENT

I

Reactor Coolant System

CYCLIC OR

TRANSIENT LIMIT

200 heatup cycles at < 1004F/hand 200 cooldown cycles at< 100 F/h.

200 pressurizer cooldown cyclesat < 2004F/h.

80 loss of load cycles, withoutimmediate Turbine or Reactor trip.

40 cycles of loss-of-offsite .

A.C. electrical. power.

80 cycles of loss of flow in onereactor coolant loop.

400 Reactor trip cycles.

f50AO'eak tests.

5 hydrostatic pressure tests.

DESIGN CYCLEOR TRANSIENT

Heatup cycle - T from < 200 F

to > 550 F

CooTdown cycle - T from> 550'F to < 200 F.

Pressurizer cooldown cycletemperatures from > 6504F to< 200 F.

> 15X of RATED THERMAL POWER toOX of RATED THERMAL POWER.

Loss-of-offsite A.C. electrical.ESF Electrical System.

- Loss of only one reactorcoolant pump.

100X to OX of RATED THERMAL POWER.

Pressurized to > 2435 psig.

Pressurized to > 3100 psig.

Secondary Coolant System 6 loss of secondary pressure

5 leak tests3S~ hydr ostatic pressure tests.

Loss of Secondary pressure

Pressurized to ?; 1085 psig

Pressurized to g 1356 psig.

- ~


6. 1 RESPONSIBILITY

6. 1. 1 The Plant Manager - Nuclear shail be responsibie for or roll"unit opera- ) QEtion and shall delegate in writing the succession to this responsibility duringhis absence.

6. 1.2 The Plant Supervisor - Nuclear (or during his absence from the controlroom, a designated individual) shall be responsible for the control room com-mand function. A management directive to this effect, signed by the Site VicePresident shall be reissued to ail station personnel on anannual basis.

6. 2 ORGANIZATION

ONSITE AND OFFSITE ORGANIZATION

6.2.1 An onsite and an offsite organization shall be established for facilityoperation and corporate management. The onsite and offsite organization shallinclude the positions for activities affecting the safety of the nuclear powerpl ant.

a ~

b.

C.

d.

e.

Lines of authority, responsibility and communication shall beestablished and defined from the highest management levels throughintermediate levels to an including all operating organizationpositions. Those relationships shall be documented and updated, asappropriate, in the form of organizational charts. These organiza-tional charts will be documented in the Topical guality AssuranceReport and updated in accordance with 10 CFR 50.54(a)(3).

The Senior Vice President-Nuclear shall be responsible for overallplant nuclear safety, and shall take any measures needed to ensureacceptable performance of the staff in operating, maintaining, andproviding technical support to the plant to ensure nuclear safety.

The Plant Manager-Nuclear shall be responsible for overall unitsafe operation and shall have control over those onsite activitiesnecessary for safe operation and maintenance of the plant.

Although the individuals who train the operating staff and those whocarry out the quality assurance functions may report to the appro-priate manager onsite, they shall have sufficient organizationalfreedom to be independent from operating pressures.

Although health physics individuals may report to any appropriatemanager onsite, for matters relating to radiological health andsafety of employees and the public, the health physics manager shallhave direct access to that onsite individual having responsibilityfor overall unit management. Health physics personnel shall havethe authority to cease any work activity when worker safety isjeopardized or in the event of unnecessary personnel radiationexposures.

TURKEY POINT - UNITS 3 8E 4 6"1t s g fin's

i% K


FhciLiTy~ET- STAFF

3~46.2.2 The ~ rganization shall be subject to the following:

a. ach on-duty shift shall be composed of at least the minimum shiftcrew composition shown in Table 6.2-1;

b. At least one licensed Operator shall be in the control room whenfuel is in the reactor.

c. At least two licensed Operators shall be present in the control roomduring reactor startup, scheduled reactor shutdown and duringrecovery from reactor trips. In addition, while the unit is inMODE 1; 2, 3, or 4, at least one licensed Senior Operator shall bein the control room;

d. A Health Physics Technician* shall be on site when fuel is in thereactor;

e. All CORE ALTERATIONS shall be observed and directly supervised byeither a licensed Senior Operator or. licensed Senior Operator Limitedto Fuel Handling who has no other concurrent responsibilities duringthis operation;

9.

A site Fire Brigade of at least five members" shall be maintained onsite at all times. The Fire Brigade shall not include the ShiftSupervisor and the two other members of the minimum shift crewnecessary for safe shutdown of the unit and any personnel requiredfor other essential functions d a fire emergency; and

Cs)Administrative procedures shall developed and implemented tolimit the working hours of unit staff who perform safety-related ) 'CEEIfunctions (e.g., licensed Senior Operators, licensed Operators,health physicists, auxiliary operators, and key maintenance personnel).

or

Adequate shift coverage shall be maintained without routine heavyuse of overtime. The objective shall be to have. operating personnelwork a normal 8-hour day, 40-hour week while the unit is operating.However, in the event that unforeseen problems require substantialamounts of overtime to be used, or dur ing extended periods of shut-down for refueling, major maintenance, or major plant modification,on a temporary basis the following guidelines shall be followed:

The Health Physics Technician and Fire Brigade composition may be less thanthe minimum requirements for a period of time not to exceed 2 hours, in orderto accommodate unexpected absence, provided immediate action is taken to fillthe required positions.

TURKEY POINT - UNITS 3 8E 4 6-2 ~ ~

~ w se'


UNIT STAFF Continued

l. An individual should not be permitted to work more than 16 hoursstraight, excluding shift turnover time.

2. An individual should not be permitted to work more than 16 hoursin any 24-hour period, nor more than 24 hours in any 48-hourperiod, nor more than 72 hours in any 7-day period, all excludingshift turnover time.

3. A break of at least 8 hours should be allowed between workperiods, including shift turnover time.

4. Except during extended shutdown periods, the use of overtimeshould be considered on an individual basis and not for theentire staff- on a shift.

des agnesAny deviation from the above guidelsnes all be authorized by thePlant Manager - Nuclear or his, or higher levels of manage-ment, in accordance with established procedures and with documenta-tion of the basis for granting the deviation. Controls shall beincluded in the procedures such that individual overtime shall bereviewed monthly by the Plant Manager - Nuclear or his designee toassure that excessive hours have not been assigned. Routine devia-tion from the above guidelines is not authorized.

h. The Operations Supervisor shall hold a Senior Reactor OperatorLicense.

i. The Operations Superintendent shall

.;e, g.ll ~ 4-.- h-l~ - s'-"g~,g, Q; th Tul<gP ~ Pl

ov i@ave. he[I a Gni r Rc~~<~ Q~v~4<gt~i(ay. (~~+ < IC>

p~gLV I I't~l lk)OM ~~~ )

TURKEY POINT - UNITS 3 8( 4 6-3

TABLE 6.2-1

MINIMUM SHIFT CREW COMPOSITION

POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION

BOTH UNITS INMODE 1; 2, 3,

or 4

BOTH UNITS INMODE 5 or 6OR DEFUELED

ONE UNIT IN MODE 1, 2, 3, or 4AND

ONE UNIT IN MODE 5 or 6 or DEFUELED

PSN

SRO none""

RO

AO

3'4

3*

STA none

PSN - Plant Supervisor Nuclear with a Senior Operator licenseSRO - Individual with a Senior Operator license

RO - Individual with an Operator licenseAO - Auxiliary Operator

STA - Shift Technical Advisor

The shift crew composition may be one less than the minimum requirements ofTable 6.2-1 for a period of time not to exceed 2 hours in order to accommodateunexpected absence of on-duty shi 't crew members provided immediate action istaken to restore the shift crew composition to within the minimum requirementsof Table 6.2-1. This provision does not permit any shift crew position to beunmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Plant Supervisor Nuclear from the control room whilea unit is in MODE 1, 2, 3, or 4, an individual (other than the Shift TechnicalAdvisor) with a valid Senior Operator license shall be designated to assumethe control room command function. During any absence of the Plant SupervisorNuclear from the control room while both units are in MODE 5 or 6, an indi-vidual with a valid Senior Operator license or Operator license shall bedesignated to assume the control room command function.

At east one of the required individuals must be assigned to the designatedposition for each unit.

""At least one licensed Senior Operator or licensed Senior Operator Limitedto Fuel Handling must be present during CORE ALTERATIONS on either unit,who has no other concurrent responsibilities.

*""The STA position shall be manned in MODES 1, 2, 3, and 4 unless the PlantSupervisor Nuclear or the individual with a Senior Operator license meetsthe qualifications for the STA as required by the NRC.

TURKEY POINT - UNITS 3 5 4 6-4 ,ftlgI CQQ


6. 2. 3 SHIFT TECHNICAL ADVISOR

6.2.3.1 The Shift Technical Advisor shall provide advisory technical supportto the Plant Supervisor Nuclear in the areas of thermal hydraulics, reactoren ineering, and plant analysis with regard to the safe operation of theun' The Shift Technical Advisor shall have a bachelor's degree or equiva-len in a scientific or engineering discipline and shall have receivedspecific trainin in the response and analysis of the uni for transients andaccidents, and in un'esign and layout, including the capabilities ofinstrumen 'o and cont ~ in the control room.

S'.

3 FAC STAFF UAL CATIONS

6 .3.1 Each member of the facility staff shall meet or exceed the minimumqualifications of ANSI N18.1-1971 for comparable positions, except for theHealth Physics Supervisor Who shall meet or exceed the qualifications of Regu-latory Guide 1.8, September 1975 The licensed Operators and Senior Operatorsshall also meet or exceed the minimum qualifications of the supplementalifi d'1

i ccFR 55 gnoi <NSX 3, I l98 t.6.3.2 When e ea Physics Superv sor does not meet the above require-ments, compensatory action shall be taken which the Plant Nuclear SafetyCommittee determines and the NRC office of Nuclear Reactor Regulation concursthat the action meets the 'ent o S ecification 6.3

)o.n Qe, Oper~~<~ Svporingeq ~k whoa@ re~oir<rrieP6.4 TRAINING Par A Senior- gea.8c r Qper~tor t i<cnre. is a.s "st~f~JCpLc,i/i cgdlayl 4 2 Z A ~

6.4.1 A retraining em n ra>ning pro ram or the facility staffshall be maintained under the direction of the Training Superintendent andshall meet or exceed the requirements and recommendations of Section 5.5 ofANSI N18.1-1971 an Amen4s-

, and shall include familiarization with relevant ylindu t r

io cue 55 ~~el RASE Z. I, tq8 t.'.

4. 2 A ra n ng program o e fire brigade shall be maintained under thedi 1 f h Fi P 1 "~ d I 11 d hrequirements of 10 CFR 50.48 and 0 CFR 50 Appendix R.

IQe6.5 REVIEW AND AUDIT Super viSor

6.5. 1 PLANT NUCLEAR SAFETY COMMITTEE PNSC

FUNCTION

6.5. l. 1 The PNSC shall function to advise the Plant Manager '- Nuclear on allmatters related to nuclear safety.

TURKEY POINT - UNITS 3 8( 4 6-5


COMPOSITION

6.5.1e2 The PNSC .shall be composed of the:

Member:Member:Henber:Hehber:Henber:Henber:Hehber:Heaber:Mehber:Henber:Menber:Hectber:

Plant Hana9er Ntac (~rOperat1ons Super1ntendentOperations Superv1sorMaintenance Super intendantlnstrugient C Control SupervisorReactor Superv1sorHealth Physics SupervisorTechn1cal SupervisorChenistry SupervisorQuality Control SupervisorAssistant Plant Supt. ElectricalAss1stant Plant Supt. Mechanical

Pa

an:Vice ChaiMember:Member:Member:Member:

r:Tllf'hiis'ada gism (]

Plant Manager - NuclearOperations Superintend uclear

al Depar SupervisorHainte rintendent - Nuclear

rument and Con u ervisorReactor Supervisor

Ph s e or~4.'pypg~ < sf fp ++cQ dpgggsjja//

r/,Q @dan) sfmgvy( IPf Airfv'is)i

~/6.5.1.3 All alternate members shall be appointed in writing by the PNSC

Chairman to serve on a temporary basis; however, no more than two alternatesshall participate as members in PNSC activities at any one time.

HEETING FRE UENCY

QUORUM

~ ~ ~6. 5. 1. 5 The quorum of the PNSC necessary for the performance of the PNSC

responsibility and authority provisions of these Technical Specificationsshall consist of the. Chairman ' and four members including ) oh (alternates.

6.6. 1.4 The PMSC shall meet at least once per calendar month and as convened )dh /by the PNSC Chairman.

RESPONSIBILITIES

6.5.1.6 The PNSC shall be res s b for:„J<,„,> $'sic +HJ e<~psfCy npd'~4t~Q

a. Review of: (1) all proposegprocedures

and ~ anyother proposed procedures or changes thereto as determinfd bythe Plant Manager - Nuclear;

b. Review of all proposed tests and experiments that affect nuclearsafety;

C.

d.

Review of all proposed changes to Appendix "A" TechnicalSpecifications;

Review of all proposed changes or modifications to unit systems orequipment that affect nuclear safety;

s

TURKEY POINT - UNITS 3 8L 4'-6

AOMINISTRATIVE CONTROLS

RESPONSIBILITIES Continued)

Investigation of all violations of the Technical Specifications,including the preparation and forwarding of reports covering evalua-tion and recommendat'ons t rev t recurre c to the Senior VicePresident-Nuclear and to theChairman of the Compan uc ear evsew oard;

f. Review of all REPORTABLE EVENTS;

g. Review of facility operations to detect potential hazards tonuclear safety;

h. Performance of special reviews, investigations, or analyses andreports thereon as requested by the Plant Manager - Nuclear orthe Chairman of the Company Nuclear Review Board;

). j(.

Review of the Emergency Plan and implementing procedures andsubmittal of recommended changes to the Chairman of the CompanyNuclear Review Board;

SeniorReview of any accidental, unplanned, or uncontrolled radioac verelease including the preparation of reports- covering evaluation,recommendations, and disposition of the corrective action t re-vent recurrent, and the rding of these reports to theVice President Auclear and to the Chairman of the CompanyNuclear Revie/Board; an

o th

)Q~

) dPI

6.5.1.7 The PNSC shall:

) ih)

Render determinations in writing with regard to whether or noteach item considered under Specification 6.5.1.6a. through Q d. i Qr-

constitutes an unreviewed safety question; andlaA'anacea- L)urea

Provide written notification within 24 hours to the Sen or ice C +lPresident-Nuclear and the Company Nuclear Review Board of disagree-ment between the PNSC and the Plant Manager Nuclear; however, thePlant Manager - Nuclear shall have responsibility for resolutionof such disagreements pursuant to Specification 6.1.1.

TURKEY POINT - UNITS 3 5 4 6"7


RECORDS

6.5.1.8 The PNSC shall maintain written minutes of each PNSC meeting that, ata minimum, document the results of all PNSC activities performed under theresponsibility provisions of these Technical Specifications. Copies shall be

provided to the Senior Vice President-Nuclear and the Company Nuclear Review

)+ (

A>~ Board.[AlSENT ~~

OA6 5 9 COMPANY NUCLEAR REVIEW BOARO CNRB )QE

(p 4;BA) FUNCTION

6.5. P I The CNRB shall function to provide independent review and audit of ) 06

designated activities in the areas of:

a ~

b.C.d.e.f.gh.

Nuclear power plant operations,Nuclear engineering,Chemistry and radiochemistry,Metallurgy,Instrumentation and control,Radiological safety,Mechanical and electrical engineering, andguality assurance practices.

The CNRB shall report to and advise the Executive Vice President on thoseareas of responsibility specified in Specifications 6.5. 2.7 and 6. 5. 2. 8. )„,0COMPOSITION

I

gwS6'W ... CNRB shall be composed he:

Chairman: ior Vice Pre 'dent-NuclearOIL

I t ~~ .. Vic reside ucl ear EnergyMember.Member:Member:Member:Member:Member:MembeMe r:

mber:

Vice P i nt-Engineering Project Management and ConstructionChief er Power Plant EngineeringDir r-qua y Assurance

'ctor-Nuclea icensinganager-Power Pla Engineering

Manager-Nuclear Ener ServicesManager-Nuclear FuelGroup Vice President

ALTERNATES

6.5. p3 All alternate members shall be appointed in writing by the CNRB Chairman ) QF

to serve on a temporary basis; however, no more than two alternates shallparticipate as voting members in CNRB activities at any one time.

, TURKEY POINT - UNITS 3 5 4 6-8 sI] sg

6.5.2 TECHNICAI. REYIEN AND CONTROL ACTIVITIES~Teelb».eaT l3eTaAb»e» Sfytpey'bfbeer

6. 5. 2. 1 The shall assure that eachprocedure and program required by Specification 61 8 and other procedures whichaffect nuclear safety, and changes thereto, is, prepared by a qualified indivi-dual/organization. Each such procedure, and changes thereto, shall be reviewedby an individual/group other than the individual/group which prepared the proce-dure, or changes thereto, but who may be from the same organization as the indi-vidual/group which prepared the procedure, or changes thereto.

6.5.2. 2. Individuals responsible for reviews performed in accordance with 6.5.2. 1

b b" f««f 1»»»'visory staff, prev>ouslg degignc&ct ~Q +4e Plw~+ Wa~~zer - Nuc(eaY- <Ope"&<~ such reY'le~5. Each such review shall include a determination of whetheror not additional, cross-disciplinary, review is necessary. If deemed necessary,

111 f Pby pp pf ~ 1 1

AGvl6.5.2.Z The station security ~~ and implementing procedures shall bereviewed. Recommended changes shall be approved by the ~F4~ Site Servicesor designated alternate and transmitted to the;cH As

A

OPERATiOHS ~APE.Rir4T'E.t4 aOCt4T~6. 5.2.~ Theance of a review bsite release of radioactive m

iew y a qua ified individual/organization. of every unplanned o-t 1 e material to the environs including the preparationne on-

an orwarding of reports covering the evaluation, recommendations and disposi-tion of the corrective action to prevent recurrence.

Of'GkhYE+ADA>PE<> A~<lOG<T

6.5.2.5 The ,'hall assure the perfor-

mance of a review by a qualified individual/organization of changes to the PRO-

CESS CONTROL PROGRAM, OFFSITE OOSE CALCULATION MANUAL

Tumed.V Poidr —0>n-s 34 I

j j

r ~~, .j z P <~l fP Il(f s ~ a, i M/

members to th and sha 1 designate from this membership a Chairman and atleast one Vice Chairman. The membership shall collectively possess experienceand competence to provide independent review and audit in the areas listed inSection 6.5.2. 1 The Chairman and Vice Chairman shall have nuclear backgroundin engineering or operations and shall be capable of determining when to callin experts to assist the ~6-review of complex 'problems. All members shallhave at least a bachelo .s degree in engineering or related sciences TheChairman shall have a least eats of professional level managemen eri-ence in the power fi d and each f the other members shall have at least

years of cumulativ professional level experience in one or more of thefields listed in Se tion 6.5,2. 1. Or yiVa 8» eX Cries aS @<I

c.<RQ . QraQ C4t»<m lg&,~uo~ 9. ~ ).

TQ < 4 PA44 l OJ7 %~i~gp~y


CONSULTANTS

~ ~~

6.5+4 Consultants shall be utilized as determined by the CNRB Director ) OF-

to provide expert advice to the CNRB.

MEETING FRE UENCY

3/6.5 g.5 The CNRB shall meet at least once per 6 months and as convened by ) QE.the RB chairman or his designated alternate.

UORUM

6.5 g.6 The quorum of the CNRB necessary for the performance of the CNRB9f

review and audit functions of these Technical Specifications shall consistof the Chairman or his designated alternate and at least four CNRB membersincluding alternates. No more than a minority of the quorum shall haveline responsibility for operation of the facility.REVIEW

3'.5.$.7 The CNRB shall be responsible for the review of:

The safety evaluations for: (I) changes to procedures, equipment,or systems; and (2) tests or experiments completed under the provi-sion of 10 CFR 50.59, to verify that such actions did not con-stitute an unreviewed safety question;

)O~

b.

C.

e.

Proposed changes to procedures, equipment, or systems which involvean unreviewed safety question as defined in 10 CFR 50.59;

Proposed tests or experiments which involve an unreviewed safetyquestion as defined in 10 CFR 50.59;

Proposed changes to Technical Specifications or this OperatingLicense; .

Violations of Codes, regulations, orders, Technical Specifications,license requireme'nts, or of internal procedures or instructions.having nuclear safety significance;

Significant operating abnormalities or deviations from normal andexpected performance of unit equipment that affect nuclear safety;

g. All REPORTABLE EVENTS;

h. All recognized indications of an unanticipated deficiency in some

aspect of design or operation of structures, systems, or componentsthat could affect nuclear safety; and

Reports and meeting minutes of the PNSC.



AUDITS

~ ~

~

~ ~36.5 $ .8 Audits of unit activities shall be per forced under the cognizance ) Qdof the CNRB. These audits shall encompass:

as

b.

C.

d.

e.

9.

h.

k.

The conformance of facility operation to provisions containedwithin the Technical Specifications and applicable license condi-tions at least once per 12 months;

The performance, training, and qualifications of the entirefacility staff at least once per 12 months;

The results of actions taken to correct deficiencies occurring infacility equipment, structures, systems, or method of operationthat affect nuclear safety, at least once per 6 months;

The performance of activities required by the guality AssuranceProgram to meet the criteria of Appendix B, 10 CFR Part 50, atleast once per 24 months;

The fire protection programmatic controls including the implement-ing procedures at least once per 24 months by qualified licenseegA personnel;

The fire protection equipment and program implementation at leastonce per 12 months utilizing either a qualified offsite licenseefire protection engineer or an outside independent fire protectionconsultant. An outside independent fire protection consultantshall be used at least every third year;

The Radiological Environmental Monitoring Program and the resultsthereof at least once per 12 months;

The OFFSITE DOSE CALCULATION MANUAL'and implementing procedures atleast once per 24 months;

The PROCESS CONTROL PROGRAM and implementing procedures for proc-essing and packaging of radioactive wastes at least once per24 months;

The performance of activities required by the guality AssuranceProgram for effluent and environmental monitoring at least onceper 12 months; and

Th er enc Plans and implementing procedures at least once per12 Kin

The Security Plans and implementing procedures at least once per@5

Any other area o fac ity operation considered appropriate by theCNRB or the Executive Vice President.

TURKEY POINT - UNITS 3 & 4 6-10 ~l((4'Vo


RECORDS

~~

~ ~~ ~

~ ~ ~9

6.5.$ .9 Records of CNRB activities shall be prepared, approved, aod dis- ) Q@

tributed as indicated below:

a. Minutes of each CNRB meeting shall be prepared, approved, andforwarded to the Executive Vice President within 14 days follow-ing each meeting;

b. Reports of reviews encompassed by Specification 6.5.2.7 shall beprepared, approved, and forwarded to the Executive Vice Presidentwithin 14 days following completion of the review; and

C. Audit reports encompassed by Specification 6.5.2e8 shall beforwarded to the Executive Vice President and to the managementpositions respons'ible for the areas audited within 30 days aftercompletion of the audit by the auditing organization.

0

6.6 REPORTABLE EVENT ACTION

6.6. 1 The following actions shall be taken for REPORTABLE. EVENTS:

a. The Commission shall be notified and a report submitted pursuant tothe requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the PNSC, and the resultsof this review shall be submitted to the CNRB,

and the Senior Vice President-Nuclear. GE

6.7 SAFETY LIMIT VIOLATION

6.7.1 The following actions shall be taken in the event a Safety Limit isviolated:

a. In accordance with 10 CFR 50.72, the NRC Operations'enter, shallbe notified by telephone as soon as practical and in all caseswithin one hour after the violation has been determined. TheSenior Vice President-Nuclear, and the CNRB shall be notifiedwithin 24 hours.

b. A Licensee Event Report shall be prepared in accordance with10 CFR 50.73.

c. The License Event Report shall be submitted to the Commission inaccordance with 10 CFR 50.73, and to the CNRB, and the Senio~ VicePresident-Nuclear within 30 days after discovery of the event.

d. Critical operation of the unit shall not be resumed until authorizedby the Nuclear Regulatory Commission.

TURKEY POINT " UNITS 3 8( 4 6"11


6.8 PROCEDURES AND PROGRAMS

6.8. 1 Written procedures shall be established, implemented, and maintainedcovering the activities referenced below:

a ~

b.

The applicable procedures recommended in Appendix A of RegulatoryGuide 1. 33, Revision 2, February 1978, Sections 5. 1 and 5. 3 of ANSIN18.7-1972; and the Facility Fire Protection Program;

The emergency operating procedures required to implement therequirements of NUREG-0737 and Supplement I to NUREG-0737 as statedin Generic Letter No. 82-33;

c. Security Plan implementation;

d. Emer gency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation;

f. OFFSITE SE CA CULATION MANUAL implementation; andBeen .l P~r~

g. guality or effluent monitoring using the guidance inRegulatory Guide 1.21, Revision 1, June 1974; and

(.) h. guality Control Program for environmental monitoring using theguidance in Regulatory Guide 4. 1, Revision 1, April 1975.

6.8.2 Each procedure of Specification 6.8.1 (a through g), and changesthereto, shall be reviewed b

pi i i,i . iforth in administrative pro dure inarcogar.ciZ turk pcciFird,r~ r,,5: I'n4 (.5.2. as appliCa6lc6.&.3 Temporary changes to procedures o Me iftc'Stion-&.& 6(a tlirough g) ) 9 Zmay be made provided:

a. The intent of the original procedure is not altered;

b. The change is approved by two members of the plant managementstaff, at least one of whom holds a Senior Operator license on theunit affected; and

c. The change is documented, reviewed , and approved &y-tissuithin 14 days of implementation~

;, in iirrrrgncr aria r 5l a+5 5:2.,, pd appl.(eagle

The g sty Con Program(6;8. .h re r vs wed by t d proce res for nviro ntal oni ing

tate orid .

TURKEY POINT - UNITS 3 4 4, 6-12

ADHINISTRATIVE CONTROLS

PROCEOURES ANO PROGRAMS Continuedt 6.8.4 The following programs shall be established, implemented, andmaintained:

a. Primar Coolant Sources Outside Containment

A program to reduce leakage from those portions of systems outsidecontainment that could contain highly radioactive fluids during aserious transient or accident to as low as practical levels. Mhe-

The program shallinclude the following:

(1) Preventive maintenance and periodic visual inspectionrequirements, and

(2) Integrated leak test requirements for each system at refuelingcycle intervals or less.

b. In-Plant Radiation Monitorin

C.

A program which will ensure the capability to accurately determinethe airborne iodine concentration in vital areas under accidentconditions. This program shall include the following:

(1) Training of personnel,

(2) Procedures for monitoring, and

(3) Provisions for maintenance of sampling and analysis, equipment.

Secondar Water Chemistr

A program for monitoring of secondary water chemistry to inhibitsteam generator tube degradation. This program shall include:

(1) Identification of a sampling schedule for the criticalvariables and control points for these variables,

(2) Identification of the procedures used to measure the valuesof the critical variables,

(3) Identification of process sampling points, which shall includemonitoring the discharge of the condensate pumps for evidenceof condenser in-leakage,

I

(4) Procedures'for the recording and management of data,

TURKEY POINT - UNITS 3 8L 4 6-13


PROCEDURES AND PROGRAMS Continued

(5) Procedures defining corrective actions for all off-controlpoint chemistry conditions, and

(6) A procedure identifying: (a) the authority responsible forthe interpretation of the data, and (b) the sequence and tim-ing of administrative events required to initiate correctiveaction.

d. Post-Accident Sam linA program which will ensure the capability to obtain and analyzereactor coolant, radioactive iodines and particulates in plantgaseous effluents, and containment atmosphere samples under acci-dent conditions. - The program shall include the following:

(1) Training of personnel,

(2) Procedures for sampling and analysis, and

(3) Provisions for maintenance of sampling and analysis equipment.

6.9 REPORTING RE UIREMENTS

) ROUTINE REPORTS

6.9.1 In addition to the applicable reporting requirements of Title 10,Code of Federal Regulations, the following reports shall be submitted to theU.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC

pursuant to 10 CFR 50.4.

STARTUP REPORT

p,E~ch~e unH p,~p,c,ling W5%7TURKEY POINT - UNITS 3 4 4 6-14 Jl]is r

6.9.1.1 A summary report of plant startup and power escalation testing shallbe submitted following: (1) receipt of an Operating License, (2) amendmentto the license involving a planned increase in power level, (3) installationof fuel that has a different design or has been manufactured by a differentfuel supplier, and (4) modifications that may have significantly altered thenuclear, theraal, or hydraulic performance of the unit.

T .4niQ -. Startup eport squall addr s each of the startup testsiden fied i Chapter of the inal Saf y Analys Report nd sh 1

inc de a d cription f the m sured va es of t operati con ions orc racteri ics obta ed duri the te program nd a co ariso of these

ives wi design redictio and sp ificatio . Any rrecti e actionthat wer require to obtai satisfa ory oper ion sha also e descri ed.Any ad tional sp cific de ils requ red in li en ns based oncommitments shal be inclu ed in this re ort Subsequent Star up eportssha a ress startup tests that are necessary to demonstrate the accept-ability of changes and/or modifications. 0

6.9

fch procedure~and admin'ative pol' of 6e8.1 a "ve, and

changes the@ho, excep the Qualit Control Fr gram forenvironmen I monitori, shall be viewed by t PNSC andapproved the Plant anager - Nu ear prior to i plementationand per'ically as ovided by pr dure.

6.8.3 Te orary chases to proced es of 6.8.1 Jove may be madepr ided:

IThe intent of the ori 'lprocedure is not altered.

b. The change is a proved by twa members of the plant

~Operators Lic se on the unit Pouted.

anagement st f, at least one of whom holg a Senior

<c. The chan is documente,'eviewed by, the PNSC andapproved y the Plant M ager - Nuclea within fourteendays of plementation.

REPOR G RE UIRE NTS

the FSAR and shall in general include a description of themeasured values of the operating conditions of characteristicsobtained during the test program and a comparison of thesevalues with design predictions and specifications. Anycorrective actions that were required to obtain satisfactoryoperation shall also be described. Any additional specificdetails required in license conditions based on othercommitments shall be included in this re ort.

l h35%AT

g.9.l. I

g In addition t the applicable reporting requirements of Title 10, Code oXl.

Federal Re lations, the foll ding identified reports shall be submittePothe U. Nuclear Regul ory Commission ~ Document Control sk,Washin 'n DC. pursuant t 10 CFR 50.0.

6.9.1 ROUTINE RE RTS

a. 5tartu e ort - A summaryreport of plant star p and powerescalap on testing shall be<'submitted followin (1) receipt ofan op6rating license, (2) amendment to the 'nse involving aplanned increase in power level, (3) insta tion of fuel:thatha/a different design or has been manuf tured by a differentfidel supplier and (0) modifications that ay have significantlyaltered he nuclear, thermal'or h drau ic'rformance of the

la The repor s a a ress eac o t e tests > en i ie >n

u~E, YH«A~

Startup eports shall be submitted w hin (I') 90 da followingcomp tion of. the artup test pro am, (2) 90 d s following.res ption or co mencement o ommercial p er operation ~'o (3) 9 mon s following i ial criticali, whichever~if .

arliest. If t Startup Rep does not co r all three e ~ts(i.e., initial riticality, co etion of star p test progr andresumpti or 'orn cement of commerci poweroperati, supplement y reports sh be submit d at leastever y three month until all ree events have beencomp eted.

and ~~

~ '5 pi Cj P ) v V 8' V gr, i~", Ifj)

a'>nr r(


STARTUP REPORT (Continued)

Startup Reports shall be submitted within: (1) 90 days followingcompletion of the Star tup Test Program, (2) 90 days following resumption orcommencement of commercial power operation, or (3) 9 months following initialcriticality, whichever is earliest. If the Startup Report does not cover allthree events (i.e., initial criticality, completion of Startup Test Program,and resumption or'ommencement of commercial operation), supplementaryreports shall be submitted at least every 3 months until all three eventshave been completed.

ANNUAL REPORTS"

6.9. 1.2 Annual Reports covering the activities of the unit as described belowfor the previous calendar year shall be submitted prior to March 1 of eachyear.

Reports required on an annual basis shall include:

'

~ A tabulation on an annual basis of the number of station, utility,and other personnel (including contractors) receiving exposuresgreater than 100 mrem/yr and their associated man-rem exposureaccording to work and job functions"" (e.g., reactor operations andsurveillance, inservice inspection, routine maintenance, specialmaintenance (describe maintenance), waste processing, and refuel-ing). The dose assignments to various duty functions may be esti-mated based on pocket dosimeter, thermoluminescent dosimeter (TLD),or film badge measurements. Small exposures totalling less than20X of the individual total dose need not be accounted for. Inthe aggregate, at least 80X of the total whole-body-dose receivedfrom external sources should be assigned to specific major workfunctions;

b. The results of specific activity analyses in which the primarycoolant exceeded the limits of Specification 3.4.8. The followinginformation shall be included: (1) Reactor power history starting48 hours prior to the first sample in which the limit was exceeded(in graphic and tabular format); (2) Fuel burnup by core region;(3) Clean-up flow history starting 48 hours prior to the firstsample in which the limit was exceeded; (4) History of degassingoperations, if any, starting 48 hours prior to the first sample inwhich the limit was exceeded; and (5) The time duration when thespecific activity of the primary coolant exceeded 1. 0 microcurieper gram DOSE EQUIVALENT I-131.

A single submittal may be made for a multiple unit station. The submittalshould combine those sections that are common to all units at the station.

""This tabulation supplements the requirements of 5 20.407 of 10 CFR

Part 20.



ANNUAL REPORTS (Continued)

The Annual Radiological Environmental Operating Reports shall includesummaries, interpretations, and an analysis of trends of the results of theradiological environmental surveillance activities for the report period,including a comparison with preoperational studies, with operational controls,as appropriate, and with previous environmental surveillance reports, and anassessment of the observed impacts of the plant operation on the environment.The reports shall also include the results of the. Land Use Census required bySpecification 3.12. 2.

The Annual Radiological Environmental Operating Reports shall includethe results of analysis of all radiological environmental samples and of allenvi,ronmental radiation measurements taken during the period pursuant to thelocations specified in the table and figures in the Offsite Dose CalculationManual, as well as summarized and tabulated results of these analyses andmeasurements in the format of the table in the Radiological Assessment Branch

'echnical Position, Revision 1, November 1979. In the event that some indivi-dual results are not available for inclusion with the report, the reportshall be submitted noting and explaining the reasons for the missing results.The missing data shall be submitted as soon as possible in a supplementaryreport.

The reports shall also include the following: a summary description ofthe Radiological Environmental Monitoring Program; at least two legible map "4covering all 'sampling locations keyed to a table giving distances and direc-tions from the centerline of one reactor; the results of licensee participa-tion in the Interlaboratory Comparison Program and the corrective actiontaken if the specified program is not being performed as required by Specifi-cation 3.12.3; reasons for not conducting the Radiological Environmental Moni-toring Program as required by specification 3.12.1, and discussion of alldeviations from the sampling schedule of Table 3.12-1; discussion of environ-mental sample measurements that exceed the reporting levels of Table 3.12-2but are not the result of plant effluents, pursuant to ACTION b. of Specifi-cation 3. 12. I; and discussion of all analyses in which the LLD required byTable 4.12-1 was not achievable.

Q,gytine, PAnva R

ggpp1 5 Qof8V L 4 ~ ~p+~~~~

pygmy <>g~ ~age.adolf". j~Y'e~'4~+<)

F~i.r g~ 5l4g l ~4 ~ g

AQ 6 OAI R,A,g >oL0aicAL ~iMPM~AL Cf~i~6 4(e i<A Entire>tn~pJ


A s ngle submittal may be made f r a multiple unit station.

+y Dne n~ gk~A mylar + ~g n~~~ SX'TB:50VNbPgy~<c cME sk II lac.tvJR. +~ ~~ al 5 e'~ 4't+I M &

Jl y) ~

6-16


SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT%

6.9. 1.4 Routine Semiannual Radioactive Effluent Release Reports coveringthe operation of th'e unit during the previous 6 months of operation shallbe submitted within 60 days after January 1 and July 1 of each year. M~

I

The Semiannual Radioactive Effluent Release Reports shall include a

summary of the quantities of radioactive liquid and gaseous effluents andsolid waste released from the'nit as outlined in Regulatory Guide 1.21,"Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes andReleases of Radioactive Materials in Liquid and Gaseous Effluents fromLight-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with datasummarized on a quarterly basis following the format of Appendix B thereof.For solid wastes, the format for Table 3 in Appendix B shall be supplementedwith three additional categories: class of solid wastes (as defined byI I II,« I T.O., TF A,TFF A.,~

AO I IFI ATION O T,A, . ~A, strong +iqh+ PacLaga~

The Semiannual Radioactive Effluen~eehse Report"to be submittedwithin 60 days after January 1 of each year shall include an annual summaryof hourly meteorological data collected over the previous year. This annualsummary may be either in the form of an hour-by-hour listing on magnetictape of wind speed, wind direction, atmospheric stability, and precipitation(if measured), or in the form of joint frequency distributions of wind speed,wind direction, and atmospheric stability."*~his same report shall include I

an assessment of the radiation doses due to the radioactive liquid and gaseouseffluents released from the unit or station during the previous calendar year.This same report shall also include an assessment of the radiation doses fromradioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to theiractivities inside the SITE BOUNDARY (Figure 5.1-1) during the report period.All assumptions used in making these assessments, i.e., specific activity,exposure time, and location, shall be included in these reports. The meteoro-logical conditions concurrent with the time of release of radioactive mate-rials in gaseous effluents, as determined by sampling frequency and measure-ment, shall be used for determining the gaseous pathway doses The assessmentof radiation doses shall be performed in accordance with the methodology and

paramet 'he OFFSITE DOSE CALCULATION MANUAL (ODC

FiPpFn T~ATFa ~ anrASANvaT.pnoxTn n c Artcc oALs'rT~QVScd 15 /iCci syF ~fig,l NefbrOIo ice-f rneaS~ re~

~el"list single submittal may be made for a multiple unit station. The submittal

should combine those sections that are common to all units at the station;however, for units with separate radwaste systems, the submittal shallspecify the releases of radioactive material from each unit.

*"An lieu of submission with the Semiannual Radioactive Effluent ReleaseReport, the licensee has the option of retaining this summary of requiredmeteorological data on. site in a file that shall be provided to the NRC

upon

request.'URKEY

POINT " UNITS 3 8 4 6-17


SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)

The Semiannual Radioactive Effluent Release Report to be submitted within60 days after January 1 of each year shall also include an assessment ofradiation doses to the likely most exposed HEHBER OF THE PUBLIC from reactorreleases from the previous calendar year<

Acceptable methods for calculating the dose contribution from liquid andgaseous effluents are given in Regulatory Guide 1.109, March 1976.

The Semiannual Radioactive Effluent Release Reports shall include a listand description of unplanned releases from the site to UNRESTRICTED AREAS ofradioactive materials in gaseous and liquid effluents made during the report-ing period.

The Semiannual Radioactive Effluent Release Reports shall include anychanges made during the reporting period to the PROCESS CONTROL PROGRAM (PCP)and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specifica-tions 6.13 and 6.14, respectively, as well as any major change to Liquid,Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.15.It shall also include a listing of new locations for dose calculations and/orenvironmental monitoring identified by the Land Use Census pursuant to Specifi-cation 3.12.2.

The Semiannual Radioactive Effluent Release Reports shall also includethe following: an explanation as to why the inoperabi ty of liquid or gaseouseffluent monitoring instfumentetion wes not corfecte within the time snecified

) ~in Specification 3.3.3.6 or 3.3.3.7, respectively; description of the eventsleading to liquid holdup tanks or gas storage tanks exceeding the limits ofSpecification 3.11.1.4 or 3. 11.2.6 ,

* rrI

MONTHLY OPERATING REPORTS

6.9. 1.5 Routine reports of operating statistics and shutdown experience,including documentation of all challenges to the PORVs or 'safety valves,shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commis-sion, Document Control Desk, Washington, D.C. 20555, with a copy to theRegional Administrator of the Regional Office of the NRC, no later than the15th of each month following the calendar month cov'ered by the repor't.

4AOIAN.PEAKING FACTOR LIMIT REPORT Re in« ~ ~~«+(Fr'.

9. l. 6 e „ or RATED THERMAL POWER (F established o

at least each reload core a vailable in the ControlRoom. The e established and implemente o le

TURKEY POINT - UNITS 3 Ec 4 6-18 '

4.%. l. (o The W(Z) function(s) for Base-Load Operation corresponding to a +2% band about the targetflux difference and/or a +3% band about the target fluxdifference, the Load-Follow function FZ(Z) and the augmentedsurveillance turnon power fraction, PT, shall be provided tothe U.S. Nuclear Regulatory Commission,

whenever PT is < 1.0. In theevent, the option of Baseload Operation (as defined in Section3.2.6.a [3]) will not be exercised, the submission of the W(Z)function is not required. Should these values (i.e., W(Z), FZ(Z)and PT) change requiring a new submittal or an amendedsubmittal to the Peaking Factor Limit Report, the

Pea.king Fa.car Liw'~4 Re+'+ saba.'ll 4. prov'cd'd file.gP Q Py~~ ~e~+ Q~fl-a 1 JPi k 4iLC ~p f cs t4 +cgeqyona.l Adwi< r ikaat~r cu 3 the Re sdcnk Znspecforg4,i~ gg Jag@ op <ha(r )wptervteafa<<w) voless04~v.h)L sc. cypovect bg f'ke. ~ 1@ lsd[~.

ping Q ~~~iatOALVY

AOMINISTRATIVE CONTROLS

/ ~ ~

WAHAb PEAKING FACTOR LIMIT REPORT (Continued) ~~/;„pz ~~The analytical methods used to generate the ~ Iimits shall be thosepreviously reviewed and approved by the NRC~ If changes to these methodsare deemed necessary they will be evaluated in accordance with 10 CFR 50.59and submitted to the NRC for review and approval prior to their use if thechange is determined to involve an unreviewed safety question or if such achange would require amendment of previously submitted documentation.

A repo ing the F„ limits for all core planes containincontrol rods and all unro lanes alo e plot of predicted(F~.PR ) vs axial core hei e l e for comparison) shallbe prof)ded to ocument Control desk with cop>es e ional Administ the Resident Lnspector within 30 da s of their im lemen

VC e'er'(C Or CO~~ieo<>~ Oeume~l GV ice)SPECIAL REPDRTS goop,rgio ') D CW>oSSg, cud/ o„co6.9.2 Special reports sha e su m e o e eg ona m n strator of theRegional Office of the NRC within the time perio specified for each report asstated in the Specifications within Sections 3.0~~ 4.0>cr WO.6. 10 RECORO RETENTION .

6.10.1 In addition to the applicable record retention requirements of Title 10,Code of Federal Regulations, the following records shall be retained for atleast the minimum period indicated.6. 10.2 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at eachpower level;

b. Records and logs of principal maintenance activities, inspections,repair, and replacement of principal items of equipment related tonuclear safety;

c. All REPORTABLE EVENTS;

d. Records of surveillance activities, inspections, and calibrationsrequired by these Technical Specifications;

e. Records of changes made to the procedures required bySpecification 6.8.1;

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests andresults; and

h. Records of annual physical inventory of all sealed source materialof record.

e»



RECORD

6. 10. 3Operati

a.

b.

C.

d.

e.

RETENTION Continued)

Records of facility radiation and contamination surveys;

Records of radiation exposure for all individuals entering radia-tion control areas;

Records of gaseous and liquid radioactive material released to theenvirons;

Records of transient or operational cycles for those unit componentsidentified in Table 5.7-1;

4',IY~The following records shall be retained for the duration of the ~~

ng License: JQcL~l l lQ

Records and drawing changes reflecting des gn modificationsmade to systems and equipment described n the Final SafetyAnalysis Report;

Records of new and irradiated fuel inventory, fuel transfers, andassembly burnup histories;

g. Records of reactor tests and experiments;

h.

k.

m.

n.

Records of training and qualification for current members of thefacility staff;

Records of inservice inspections performed pursuant to these TechnicalSpecifications; .

Recor of quality assurance activities required for the duration ofth 'erating License by the /vilityAssurance Manual;

go.c,l ~qRecor s o7 reviews performed for changes made to procedures orequipment or reviews of tests and experiments pursuant to10 CFR 50.59;

Records of meetings of the PNSC and the CNRB;

Records of the service lives o l hydraulic and mechanical snubbersrequired by Specification 3.7. including the date at which theservice life commences and associated installation and maintenancerecords;

Records of secondary water sampling and water quality; and


R cords of me tings of the PNS and the CtNRB

Run

or Enrovlsl

ords frth p

ironme tal Qual ication hich arns of p ragraph 13.

cove

RespeCOAl

rdsificati

ence

f the service lives of all snub e('s ren 3.13 includin the dat 'f whic the se

and associated insta ation mai

iredvice liten anc

ords.

6. 2

0~ Annual Radiological Environmental Monitoring Reports andrecords of analyses transmitted to the licensee which are usedto prepare the Annual Radiological Environmental MonitoringReport.

ADIA ON PR CTION PROGR M

P edures r personnel, radiation rotection shall be prepared consistentwit the req 'rements bf 10 CF Part 20 and shall be. approved,main ined and hered to for all ope tions involving personrietradiation

xposu e )

HI RA ATIO AREA1

6.12. In l u of the "control evice" o "alarm signal" requited byparag ph 20.203 2) of 1 CFR 20:

a Each igh Radia 'on Area 'hich e intensity g radiatio isreate than 100 Rem/hrtbut less t an 1000 mRem/hr shall

barr aded and enspicu sly pestItd as a High, RadiatidnA a and ntrance th eto shgl be cont oiled by issbance of a;Ra 'ation Work Per 't and~any indi idual or 'group ofindiv'duals rmitted t enter 'such area shall be provided

'ithradi tion mon oring 'gevice w ich continuously'icat the ra iation do rate in e area.4

b. Ea h Hig Radiati n Area in hich th intensit of radiation isgre ter th 1000 Rem/hr sh ll be siiPject to Pe provisronsof 12.1(a) above, d in a/dition hgked d+rs shall beprovi d to p vent u uthorized hptry ink such areas and thekeys shall be intaine under administrati e control.

POPAmendment Nos.123 and 116

-y <rC

Q

s /I

l~, 1C' 3 (QvtC(/

0


RECORD RETENTION Continued)

p. Records for Environmental gualification which are covered under theprovisions of 10 CFR 50.49.

6. 11 RADIATION PROTECTION- PROGRAM

6. 11. 1 Procedures for personnel radiation protection shall be preparedconsistent with the requirements of 10 CFR Part 20 and shall be approved,maintained, and adhered to for all operations involving personnel radiationexposure.

6. 12 HIGH RADIATION AREA

6. 12. 1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the"control device" or "alarm signal" required by paragraph 20.203(c), each highradiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiationsource or from any surface which the radiation penetrates shall be barricadedand conspicuously posted as a high radiation area and entrance thereto shallbe controlled by requiring issuance of a Radiation Work Permit (RWP). Indi-viduals qualified in radiation protection procedures (e. g., Health PhysicsTechnician) or personnel continuously escorted by such individuals may beexempt from the RWP issuance requirement during the performance of theirassigned duties in high radiation areas with exposure rates equal to or lesst than 1000 mR/h, provided they are otherwise following plant radiation protec-tion procedures for entry into such high radiation areas. Any individual orgroup of individuals permitted to enter such areas shall be provided with oraccompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates theradiation dose rate in the area; or

b. A radiation monitoring device which continuously integrates theradiation dose rate in the area and alarms when a preset integrateddose is received. Entry into such areas with this monitoring devicemay be made after the dose rate levels in the area have been estab-lished and personnel have been made knowledgeable of them; or

c. An individual qualified in radiation protection procedures with aradiation dose rate monitoring device, who is responsible for pro-viding positive control over the activities within the area andshall perform periodic radiation surveillance at the frequencyspecified by the Health Physics Shift Supervisor in the RWP.

'I

6.12.2 In addition to the requirements of Specification 6.12.1, areasaccessible to personnel with radiation levels greater than 1000 mR/h at 45 cm

(18 in.) from the radiation source or from any surface which the radiationpenetrates shall be provided with locked doors to prevent unauthorized entry,and the keys shall be maintained under the administrative control of the shift

. TURKEY POINT - UNITS 3 8c 4: 6"21


HIGH RADIATION AREA Continued)S vive rfthor pc.rS~oe,(

on duty nd/or health physics Doors shall remain locked pexcep uring periods of access by personnel under an approved RWP which shallspecify the dose rate levels in the immediate work areas and the maximumallowable stay time for individuals in that area. In lieu of the stay timespecification of the RWP, direct or remote (such as closed circuit TV cameras)continuous surveillance may be made by personnel qualified in radiation pro-tection procedures to provide positive exposure control over the activitiesbeing performed within the area.

For individual high radiation areas accessible to personnel with radia-tion levels of greater than 1000 mR/h that are located within large areas,such as PWR containment, where no enclosure exists for purposes of locking,and where no enclosure can be reasonably constructed around the individualarea, that individual area shall be barricaded, conspicuously posted, and a

. flashing light shall be activated as a warning device.

6.13 PROCESS CONTROL PROGRAM PC

re (owe. PNSM6.13.1 The PCP shall b y th prior to implementation.

6.13.2 Licensee-initiated changes to the PCP:

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

(1) Sufficiently detailed information to totally support therationale for the change without benefit of additional orsupplemental information;

(2) A determination that the change did not reduce the overallconformance of the solidified waste product to existing crite-ria for solid wastes; and

(3) Documentation of the fact that the change has been reviewedand found acceptable by the PNSC.

b. Shall become effective upon review and acceptance by the PNSC.

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:

a. 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:



OFFSITE DOSE CALCULATION MANUAL ODCM (Continued)

(1) Sufficiently detailed information to totally support therationale for the change without benefit of additional orsupplemental information. Information submitted should consistof a package of those pages of the ODCM to be changed with eachpage numbered, dated and containing the revision number,together with appropriate analyses or evaluations justifyingthe change(s);

(2) A determination that the change will not reduce the accuracy orreliability of dose calculations or Setpoint determinations; and

(3) Documentation of the fact that the change has been reviewed andfound acceptable by the PNSC.

b. Shall become effective upon review and acceptance by the PNSC.

6. 15 MA R CHANGES TO LI UID/ GASEOUS AND SOLID RADMASTE TREATMENT SYSTEMS"a

6.15. Licensee initiate major cha ges to the Radwaste Treatment Systems(liq d, gaseo , and so d): /a. Sh 1 be rep rted to t Commission,fn the Semia ual RadioactiVe

E fluent R ease Repo t for the period in which he evaluation Lsaseviewed the PNSC The discus on of each hange shall co ain:

(1) A ummary of he evaluatio that led t the determina on thate change uld be made n accordanc with 10 CFR 5 .59;

(2) Sufficien detailed inf rmation to otally support the reasonfor the hange withou benefit of dditional or s pplemental.inform ion;

(3) A de ailed descrip on of the quipment, comp nents, andpro esses involve and the i erfaces with o her plant ystems;

(4) A evaluation o the change'hich shows t e predicte releasesradioactive aterials i liquid and g eous efflu ts and/or

quantity of s lid waste at differ fro th'ose prev'ously pre-dicted in t License a lication and endments t ereto;

(5 An evaluat on of the c ange, which sh ws the exp tedmaximu~'xposuresto a MEMBER F THE PUBLIC n the UNRE RICTED ARE+

and to t e general p pulation that iffer from hose previ slyestimat d in the Li ense applicati n and amend ents there o;

(6) A corn rison of t predicted re eases of ra oactive m erials,in li uid and gas ous effluents and in solid aste, to the actualrele ses for the period prior o when the c nge is t be made;

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



MAJOR ANGES T I UID G OUS AND SOI ID RADWASTE TREATMENT SYSTEMS

Co inued

) An es mate of theiexposure t plant opera ng person as a

resu t of the cha6ge; and

(8) Documentatfop'of the fac that the c nge wae n $ ewed andfound acce table by th PNSC.

b. i hall becom effective u n review d accepta e by the SC.

//

4

'URKEY POINT - UNITS 3 4 4 6"24JtJV

Qw

T. S. Number 1. 0

Justification:The new RPI scales take into account the non-linearresponse of the RPI's that would otherwise be accountedfor by the Reference Position. The new scales allow theRCCO to directly compare the RPI position to the stepcounter position without having to refer to the ReferencePosition curve. Removal of the "Reference Position"alleviates some of the-unnecessary workload on the RCCO.

2.

3.

K,<<=0.95 is consistent with Amendment 132/126 (7/18/88) .

A CONTROLLED LEAKAGE definition has been added tocomplement the addition of CONTROLLED LEAKAGErestrictions added to Specification 3.4.6.2.

4 ~

5.

All of our shipments do not go directly to a burial site,some are shipped to a waste processor for volumereduction prior to burial.A limit of 24 months on "refueling" surveillances is amanagement action to limit time between surveillances.

Justifications:1. BASES — Per FSAR Table 4.1-9, the reactor vessel and

pressurizer are designed to the requirements of ASMESection III.

T.S. Number 2.2.1

Justifications:

2 ~

3.

The CTS does not address "allowable" values. The 1986version of RTS adopted the STS 2-column format containingnominal setpoints and allowables. The RTS allowableswere best estimate numbers determined by Westinghouseusing generic tolerances. However, FPL would requireadditional analysis to adequately support these values.

Pzr. pressure: No credit is taken for PORVs in thesafety analysis or the basis in the CTS. (PORVfunctional capability enhances the overall reliabilityof the RCS).

Turbine trip: P-7 blocks reactor trip on turbine trip.4 ~ Items 17.b(1) and 17.d are confusing. P-10 in 17.d,

which allows a block of PR trip during plant startup,should be greater than or equal to 104. P-10 reset pointin 17.b(1), which allows a block of at-power trips duringshutdown, should be less than or equal to 10%. Revisingthe wording to refer to 17.b(1) as the reset of the P-10bistable and 17.d as the tripping of the bistable willclarify the tech spec and better represent instrumentdesign.

Justifications:1 ~ Flexibility required for dual unit shutdown to prevent

a severe transient on the Florida electrical grid. Thiscomment is consistent with previous discussions andagreements during RTS working meetings regarding dualunit shutdown due to electrical components beinginoperable. (Also applies to BASES change.)

T. S. 3.0.6

Justification:1 ~ This comment provides a minor revision, for clarification

only, of CTS 3.0.5. This current specification wasapproved for inclusion in the CTS in Amendment 114/108dated June 27, 1985.

The rewritten portion of the comment closely matches thewording of the NRC letter dated April 10, 1980 to AllPower Reactor Licensees concerning the definition ofOPERABLE.

Although this tech spec is not currently in the deskreference STS, FPL feels that it is important to includethis clarification.(This justification also applies to BASES change.)

T. S. Number 4.0.3 BASES

Justifications:1. This change clarifies that the allowable outage time in

the ACTION Statement begins at the end of the 24 hourallowance.

T. S. Number 4.0.6 BASES

Justifications:1. This change was made to eliminate the need for

unnecessary duplicate testing.

Justifications:1 ~

2 ~

3.

BASES — Revised to reflect actual condition addressed bythe LCO. "Postulated steam line break" appearsinappropriate with RCS temperature below 200F.

Figure 3.1-1: Breakpoints should be shown for operatorclarity.During the first 60 EFPD of a cycle, it is possible thatan adjustment to the letdown curve may not have to bemade if there is good agreement between measured data anddesign figures.

Justifications:1. The term "equivalent" would mean 100+ pgm of 1950 ppm;this would require 2 charging pumps in Mode 5. No rate

requirements for boration in Mode 5.

Justifications:1. Consistency on reporting requirements. This is a generic

issue on 30 day reporting.

Justifications:1 ~ STS requires temperature verification once per 7 days.

There is no CTS requirement to verify temperature, hence,7 days is more restrictive than CTS.

2 ~ Certain areas of the plant may be inaccessible due toradiological considerations.

Justifications:1.

2 ~

STS requires temperature verification once per 7 days.There is no CTS requirement to verify temperature, hence,7 days is more restrictive than CTS.

Certain areas of the plant may be inaccessible due toradiological considerations.

T. S. Number 3 4.1.2.4

Justifications:1. STS requires temperature verification once per 7 days.

There is no CTS requirement to verify temperature, hence7 days is more restrictive than CTS.

Justifications:1. STS requires temperature verification once per 7 days.

There is no CTS requirement to verify temperature, hence,7 days is more restrictive than CTS.

2 ~ 3/4.1.2 BASES clarified to indicate background for RWSTtemperature; for ope'rator information.

Justifications:1. This change is consistent with related tech specs (i.e.,

3.1.2.1, 3.1.2.2 and 3.1.2.5). There is no CTSrecpxirement to verify temperature.

T. S. Number 3 4.1.3.1

Justifications:1. The new RPI scales take into account the non-linear

response of the RPI's that would otherwise by accountedfor by the Reference Position. The new scales allow thereactor operator to directly compare the RPI position tothe step counter postion without having to refer to theReference Position curve. Removal of the "ReferencePosition" alleviates unnecessary workload on the reactoroperator. (This justification also applies to BASESchange.)

Justifications:1.

2 ~

Revised for consistency with CTS 3.2.2 and related bases:"Reduction in power to 754 (3 loop) . . . will ensurethat design margins to core limits will be maintainedunder both steady-state and anticipated transientconditions." Power reduction for an indicatedmispositioned rod should be the same as a knownmispositioned rod. (Tech. Spec. 3.1.3.1 ACTION c.3.a.)Delete reference to Reference Position. Refer tojustification to 3/4.1.3.1 for further information.

Justifications:1. This surveillance will require a unit shutdown to be

performed. Surveillance 4.1.3.4a and b cover anymaintenance that could affect the rod drop times.

Justifications:1. Comment returns surveillance to STS wording which is

appropriate for Turkey Point.

Justifications:1. Performing surveillance for greater than a 6 hour period

provides negligible information. This comment was agreedto by NRC technical reviewer on April 7, 1988.

Adding a functional test will be consistent with othersurveillances in STS and meets the intent of thesurveillance.

Justifications:1 ~ The NRC proposed STS wording would require FPL to perform

additional analysis to generate the Peaking Factor LimitReport per T.S. 6.9.1.6. Xn addition, the NRC's proposedwording deleted the Augmented Surveillance Tech. Specs.which were approved by NRC on March 17, 1982, as LicenseAmendment 80/74. The Augmented Surveillance Tech. Specs.were developed as a contingency against reductions in theFq margin. FPL has experienced Fq margin reductions dueto steam generator tube plugging, changes in analyticalmethods, and unexpected errors in design codes.Equipment currently in place to implement the AugmentedSurveillance Tech. Spec. would have to be removed if thissurveillance is discontinued.

Justifications:

2.

There is no requirement in the current tech specs toverify the excore QPTR calculation with an incore fluxmap. Monitoring the incore tiltwith a flux map once per24 hours will be significantly more conservative than thecurrent tech spec.

With one power range detector out of service, theremaining power range detectors along with other controlroom instrumentation (i.e., incore T/C, Tavg, flows,etc.) will detect changes in core parameters.

Several plants currently use incore thermocouples todetermine QPTR. Zncore thermocouples provide a rapidmeans of determining the core relative radial powerdistribution for use on an on-line basis. Thermocouplesare currently used in QSPDS. When used in conjunctionwith moveable detector map, incore thermocouples can benormalized to provide accurate relative integrated fuelassembly power distribution measurements.

Justifications:1. The present equipment at Turkey Point does not allow

performing this surveillance without a plant shutdown andcontainment entry behind the biological shield wall.

Justifications:This change reflects the Plant design as shown inattached FSAR Figure 7.2-5. These trips do not applybelow P-7.

2. Provides clarification to operators on required safetyfunction based on increasing and decreasing power, i.e.,automatically enabling trips, etc.

3.

4 ~

This surveillance requires Plant to be in a shutdowncondition. An 18-month surveillance frequency may causeFPL to have to shutdown for refueling early or cause anunnecessary outage.

This surveillance has been revised to create consistencybetween the table and the table notations.

5. There is no LCO which corresponds to this surveillanceand the trip bypass breaker circuit is already tested byNote 13.

6. The Automatic Trip and Interlock Logic is not testableprior to mode 3; therefore, the surveillance requirementhas been removed.

7 ~

8.

Use of jumpers/fuses to place channel in trip conditionwas agreed to by NRC technical reviewer in FPL/NRCworking meetings.

STS wording is adequate to give guidance for 30-75< RTP.

9. Above P-7 and below P-8, 2 RCP "breakers open" arerequired to cause a reactor trip. Based upon Plantdesign, a single underfrequency channel/bus may not leadto a RCP breaker/reactor trip. A requirement for 2/buswill lead to a reactor trip due to an underfrequency oneither bus.

10. The P-10 logic circuitry provides a redundant signal.In the decreasing power mode, two (2) channels to tripare required (see Justification g2 above).

12. In the decreasing power mode, three (3) channels to tripare required (see Justification g2 above).

REACTOR TRLP SICNAL

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REF DWOT 5610 T LI SH. 2 IREV 81

Rev 4 7/86

FLORIDA POWER SI LIGHTCOMPANYTURKEYPOINT PLANT UNITS 3 Sc 4

REACTOR TRIP SIGNALS

FIGURE 7.2.5

Nr. Steve Cra'g

0

desires to reserve tha option to cock the, rods (shu'.di:wn banks) out ofthe core during Mode 3 for operational flexib) lity.To add~ass the issue, Hr. R)ck Mande has tnd3cated that FPL would 1)ke toknst1tute an administrative control during Node 3 that would require the rodcontrol system to be either 1) placed )n the "bank select" code with ashutdown bank as the bank selected, or 8) d1sabled by opening the reacto~tHp breakers. Nest)nahouse aoraes that th$ s $ s an acceptable means ofprevent)ng'ncontrolle8 rod withdrawal during Node 3, and $ s therefore an:effect$ ve means of rddressfny tha issue on a short-term bas)s.

Please call me 1f you have any questions.

Very truly yours

4.Q.VS.A. PearsonPro)ect Engineer

8AP:kph

cc: R.J. AcostaD,C, Bradford$ .6. Bra)nQ.E, CoeR.A. DeckerJ.A. OeNastryT.G. GrozanD.M. HaaseR.D. HankelJ.A. HandschuhK.N. HarrisJ.A. HughesY.A. Kam<nskasK.R. KnucklesC,U. La)sure

D. NantzR. Hende

J.E, NoabaK.H. NordttiyerO.C. Poteraylsk)G.A. RowanR.J. RodriguezS.H. ShepherdA.E, S5ebeF.H. Southwor th

C, Y)11ardR.bf. 'ifinnar dS.K. HathavanC,S. Smallwood

0@

'pQ jS «i C A « I «I II«

NCNlghaeo %ter ReactorBg&c Qemitlw IvIslons

p i' '«e ~ ~ O

Mcussecs'ayltcaOMse

le VN~t m~ ~mme

~ ~

)fat Kt N ~ Harr'i 5 ~ Vice PresidentTurLey Point NuC'learRorida Power and Ught CoeyanyP. 0. Boa 029100Niaai, Florida 33152

Dear Nr, Noaba:

florida Power I light CompanyTurkey Point Units 3 and I

UPDATE CONSISTENCY IEMEN SAFETY ANALYSISND TECHNICAl SPECIFICATIOiS COHCKQIINO

ER OF REACTOR COOLANT %PS O ERATION

?h June of thi s year you were notified of a potential unrev)ewed safetyquestion cOncerning an inconsistency between the safety analySis and the TechSpecs. This inconsistency applies to the weber of operating reactor coo)antpvips when $ n Me 3 {or the u)va)ent) of the Tech Specs.

Since that t<ie. Vestinghouse has Net w)th the NC staff at their reqveitto present the Vest<nphouse position on this issue and recoivnendatiohsfor resolution. The Iaterial presehted at this seeting is docvaented ih'Astihghouse letter NS-PR-2935. which $ s attached for your $ nfoveat)on(cover letter on1y).

The purpose of this letter is to update yov on the latest infomationconcerning this issue. Vestinghouse w) ll notify you of further developoentsas they occur,

If you have any questions. please contact ae,

Yery rvly yours,

. J, R(cherds, l4nagerPro)acts Depa~entSouth Area

HT/413LAttachient

cc: V. H. Rogers, Jr.K. N. HarrisH. D. itentzP. P. DeRoseH, E. YaegerC. J. Saker

H. N. PaduanoS. Q. BrainD, J. RichardsT. P. SullivanE.. C. Anderson gE. V. Rvtledge

ATIrlteaefnrIvons

Z/+pi.zNae <aeoogy0heen

Atare)war's N5

July 9, lN4f5 T~O{}3

r

lh. O. Kfsenhut> 1frectot"P~'>,lf,ihh~imMO Mori'olk AvenwQsh$ rgcon. O.C. f0555

3Q}er Two or al I Oi(At the s uest ~

Howe e, the sa ythe reactor coot'e ass nels;-.; .he n|Nber of operant'.ng pays have been note< or

r f rplaning w<ghfn 4+tinQ-y„.. rg gg p4 fn the acDcMAt)o Tba acChdkhts wh1cn

an 1'Ii ng at aero go~. or< >'4awiine brea< ~c 0;oc.ion an-" ~n< ~~ I.

ftgg s~ 'f,'fggl, Qstfnghous„ $ „ reviewed these ac:indents under the reducededitions of one va" ": :.x md Qe..;(~ an. stegsline break events >

Mest$ nghouse his djetarlin.; :.'.: :." inconsista~;y oc:»een the safety ana y1 sis

lear %, Efsenhut:

Nl%ER OF ) MATINS l CTN 00UR'NOBS lh NOE 3

Thfa letter foealfaes the latarfal .oreaente4'on zuni 15< 1984, «fth respectto the consfstency bebeen the Technfcal $yecfffcat<ons and the safety analysisfor the nosher of operatfny reactor coolant plus fn linda S. Thfs ~tiny washeld at the bluest of the NC staff fn order Co 4fsouss the Nestfnohouse detarafnat$ on of a potantfal unrevfee4 safety question for ttiree and I'our loop plantsfOr Wa faaue. EnC1OSed am tan (10) yrOprfetary Cayfea Of the Slf4eS an5 ten{10) ten-proprfeta~ cop<os. Also enclosed are one {1} copy of App1icat<on forllfthho14fny N 9 63 (Nn-proprfeCary} and one {1) copy of Afffdavft (ten-~mp~eea>f.

As part of an fat'oval rev$ m of a utf3fty's Tyrh Specs by the Nh4 Reac:orSys~ Sraneh. the staff asked ~t the safety analysts ass~tfons were concarn<ng the nosher of ooeratfny riactor coolant peys, partÃevtar ly at or neartarO pOWer. iltheuoh tht queathOn WS neVer fOealiy aaka4, Qeat<nghduSe reViewedthe ana1ysfa assertions with rlsyect to the Tech specs.

The muf~nt for oyeratfny reactor coolant plies under these conditions>s contaited fn Specfffoatkon $ .4,>.2 f the Sgehr4 Tech Specs. !n non-StandardTech Specs, the nqu$ ~ene fs contafned fn Speeff<cat<on 3.1. these Soecs statet'zt when x4e plant fs subcr<tial by the yhutdo~ «er;<n beereen 350'F (~HR cut-fn) and S47'F or 117'F {m~load cond$ tfons), there egest be two loops oyeraole,but only one loop has Co be actually operaifny.

v r Net an 1 s!s <n

and the Tech Spec wfll n5t feyaet the conclusfons presented fn the f0', Forthe bank withdrawal Am s*crftfcal event> Mestfnghouse has pirforiid calcv-latfons whfch show that the f85 dasfon basfs my not'e set when on'ty one pmyfs fn operatfon. Thus, the Nargfn of safety as daffned fn the basfs af Che7ech'Specs fs reducid.

Vestfnghouse has a1so peHomef ca'lcu1atfons f'r one puny operatfon assnfng&n.rialf's fc, but stfll consarvatfvi, reactfvfty fnsartfon ratis, The resultsof these calculatf0ns show that the M disfgn basfs fs set. Other 40'~'L)onsand aerials used fn these analyses are fdintfcal Co the FSN aethods of analyifsfor thfs event ~ Thus ~ Qostfnghousa fels that m sfgnfffcant safety hazardaxfsts.

lkstfnghovsa fs currently consfderfng long tate analytfca1 solutfons co ch<sfssue whfch wfll show that the 5IS disfgn basfs can be etc when on>y one reactorcaalanc pmp fs fn operation Io that the Tech Specs w)11 not need Co be changed.However, fn the short tarI, Qestfnghouse raaeeends that Che plants be operatedwfth the save maber of riactor coolant pumps fn operatfon as ms ass+ed fn theanalycfS,, lhtI Chat thfS fS nOt a realfStfe ~ufment when Che aslant fS COOl)ngdown prfor to yofng fnto %de 4 (RN operatfon). partfcularly for chose plantsfor whfch Chi analysfs assess all pcs fn ooaratfon, Thus, an alternatfve cohavfng mre than one puef'n operatfan fs to prevent rod wfthdrawal. Thfs wfllproc)ude the acefdenC fere Cakfne place. Although plysfcal privancfan of wfth-drawa1 wfll aceeplfsh thfs, awfnfstratfva procadwes say be proferable. Tleabflfty Co cack the mds parlay out of tha cora durfng Node 3 prov<des desiredoperatfng Roxfhf1fty. Furthereeri thara fs lN sochanfsm by whfch the soncrolds Can be autoeatfca 1y wfthdrawn fn Ma 3 due to a contro1 sysM errorjIncreased operator awareness during thfs tfm and adheranca Ca procedures weal>

also prevent the accfdent Am occurrfng.

I'fna1ly, whfle Mestfnghouse feels that ft fs approprface co cans<der bank

Hthdrawal when fn Noda 3, Nstfnghausa Res not fntand Co address Ch<s eventfn other mdks of operatfon (Standard Tach Spec Nodes 4 and 5), Sank w<chdriwa>

f~ suhcrftfcal fs a valfd scanarfo when Oofng free Hode 3 Co %de R ~vericonsfderatfon of bank Qthdr awal fn Nndes 4 and 5 fs unrealfstfc and ft fsCuestfonahle as to whether ft fs applfcahlo or ff ft fs a Condfcfon lI event,A9afn fnsmasid operator awareness rust he consf4erid whin avaluacfng Che

approprfaceniss of the event.

a w wE V I'l14V ~ S ~ AC II>lv s ~ ~ < ~ e e r~ ~ W W ~

C4rrespon4fenca «fCh tesPeot t0 She Qestfnghousa aRfdav/t or application~-~ for Athho'lding should v%femce N-$4-53, and should be addressed to

+. R. A. Neseaann, Nanaoar Regulatory and 4yfslat<ve Mfa)m, 1.0. lox 355,P)ttsburyh, Pennsylvania IQ ~ Other corresPondence or questions should bed<i~ecI to Nr. 4. L,. tittle, 4nager, Operatic Plant L,<eens<nI Suppo<i410/$1+5454.

'ery twice/Ours'sT3%NUSE

LKcfaff COQCVTION

N. P. Osbom]44

Enc) osures

, Rahei Jr.uclear Safety Oepartment

, ~RON:

PATE:SUBJECT:

Nuclear Safety DeprrrtmentRisk Assessment Technology284%303October 4> 1984FKlFLA Prevention of Rod Withdrawal in Mode 3

To: 8. A. Pearson

oc: H. Pe Osbornet Po Ae LoftusP. M. RobertsonN. N. Raymond

SOB 2-17

}5C 4-09AMNC 49AMNC 4&9ANNC 4-09A

~ 0

~it~ o A I

~ ~ ~ ~ 'I

Ref> 1. FPI 84-729, 8-'l5-842. HS-EPR 2935) 7-9 843. FP-FP-697)

82FP~~011

, Florida Pmer and Light has requested additional infixation regarding aninconsistency between the ssfety analysis and the Tech Specs with respect to

. the number of operating reactor ooolant peps when in Noae 3 (or theequivalent) ai defined in the Tech Specs. Reference 1 provided the latestupdate on this issue. The purpose of this letter is to provide confirmat$ .on ofthe safety analysis assumptions and FPhl.3s proposed administrative procedure toaddress this issue, as discussed with Rick Monde> Reactor Engineer at TurkeyPoint.

As discussed in Reference 2, a letter from Westinghouse to the NRC, the onlyevent impacted by this issue is the RGB bank withdrawal from subcritioal.

'hisaccident is defined as an mcontrol3,ed addition of reactivity to thereactor core caused by withdrawal of RCCAs resulting in a power excursion.Such a transient could be caused by a malfrirction of the automatic rod controlsystem or by operator error. The ana3ys5a assmes a reactivity insertion rateof 75 pm/sec> greater Chan the maxim'esulting from the simultaneoust withdrawal of the combination of two sequential control banks having themaximus. co-...coined worth at msximmr speed.

The most recent RCCA bank withdrawal from subcritioal analysis for Turkey Po5ntUnite 3 4 4 was pe!'fonaed as part of tn. positive If'tudy ir 1981 (Reference3) f ana assumes tnrgt all three reao.o. ac';iant ymps are oper r Ling. However,the Turkey Point Tech Specs state Chat »ln Hot Shutdown at least two ReactorCoo3ant Loops shall be operabLe end s..'~est one Reactor Cc03c.nt Loop shall be

. in operation"I (T.S. 3.4.1.4> pp:.,~-.'..'hus, the Tech SWcs are inconsistentwitn the smg'psi'.'-'.-pl'ion A; .;:....:. 3'."erence 2', $ t '.::'.,t, a realisticrequirerwnt to hsv; all three reLo~or <«]a... prxrrps operating. wnen the pLant, iscooling cwn prior to going to Cr ic E. ": .. '.. In acax".3cn, r.r~'nistrative

rocedures are preferatQe to phd,;eau „.:.I:,...~.. of rc" w'::,.r a, al for Turkeyint since the pLe>t desires to reser e one option to cocle the rods (shut, down

nenki) outof the,core during Itoce 3 for oocrational fleaioility.

E ~l'"'c issue, Rick Hend':ias in;;;e:cc .net FP44 v Dd liae

i',itut,e an attain$ strative oontrol during Hade 3 that would require the rodcntrol system to be either 1) placed i'n the «bank select«mode vfth s shutdgm

bank as the bank se3.ected> or 2) disabled by opening the reaotor tripbrewkers, Westinghouse agrees that this is an aooeptsble means of gev'entinguncontrolled rod withdrawal during Mode 3, and is therefore an effeotive meansof addressing the issue on a short-tens basis.

The foregoing has been telecopied to FP4L, and their ooaments have beinincorporated. Please fannaQy transmit this information to the oustaaer.Contaot me if there ire any questions.

G. H. HeberlePlant Transient hnalysis

Justifications:1. Allowing the standby RHR loop to be inoperable for

surveillance testing would cause only one loop to beOPERABLE for the two-hour period of time allowed forperforming the tests. The potential exists that theremaining RHR pump could be de-energized for one (1) hour(as allowed by another footnote) or that it wouldotherwise become inoperable, resulting in having nooperating loop for the specified time period. This doesnot represent a significant increase in risk because theRCS thermal capacity is sufficient to maintain the RCStemperature rise within acceptable limits during thistime period, while the RHR loop being surveilled isrestored to its OPERABLE state.

Justifications:1. BASES — The relief capacity of the pressurizer safety

valves is 293,300 lbs/hr per FSAR Table 4.1.3

T. S. 3 4.4.2.2

Justifications:1- This revised action statement allows cooldown of a unit

when safety valves are declared inoperable to performmaintenance. (Cooldown is considered a positivereactivity change.)

2 ~ The definition of RCS pressure boundary is consistentwith OMS protection that provides at least 2.20 squareinches to depressurize the RCS.

T. S. ~34. 4. 3

Justifications:1 ~ BASES — The maximum pressurizer water volume at 92%

indiated water level is 1133 cubic feet.2 ~ This surveillance should not be performed during power

operation.

Justifications:

2.

A revised specification has been proposed which coversonly the PORV block valves.

Reactor Coolant System overpressure protection isprovided by the Pressurizer Safety Valves as addressedin Specification 3/4.4.2.The Steam Generator tube rupture accident does requirea means to depressurize the Reactor Coolant System toreduce coolant leakage to the Secondary Side of the SteamGenerator. The primary means of depressurizing theprimary system is by use of the normal pressurizer spray.Auxiliary pressurizer sprays can be used as a backup.While the PORV can be used as a second backup, it is theleast desirable because it tends to reduce ReactorCoolant System inventory.ACTION has been revised to indicate HOT SHUTDOWN (Mode4) consistent with the APPLICABILITY (Modes 1-3).

T. S. 3 4.4.5

Justifications:1. Specification 3.0.4 provides restriction for MODE change.

Shutdown requirements are provided because Specification3.0.3 would apply if a steam generator becomes inoperableduring operation. Current wording is confusing tooperator because it does not provide all requiredactions.

2 ~

3.

The intent of this surveillance is to inspect everyrefueling interval.Consistency on reporting with 10 CFR 50.73 for LER's.

Justifications:1 ~ The CTS requirement is refueling (recently approved

amendment). Surveillance cannot be performed with unitat power or greater than 200 F due to ALARA and HPrestrictions on dose rates and stay times.

T. S. 3.4.4.6.2

Justifications:1 ~ CONTROLLED LEAKAGE requirements were added to explicitly

control flow from the RCP seals. The wording is similarto the STS.

2. This LCO requirement was revised to be consistent withCTS (NRC Order, dated 4/20/81) restrictions for valveleakage, which are explicitly called out in Table 3.4-1.

3.

4 ~

The added footnote is from CTS 4.17, Reactor CoolantSystem Pressure Isolation Valves. It allows flexibilityin measurement of valve leakage in order to reducepersonnel radiation exposure.

Consistent with CTS and 3.0.3.5. This surveillance was added due to the addition of an LCO

for CONTROLLED LEAKAGE. The wording is similar to theSTS.

6. The added wording excludes the requirement to verifyvalve leakage if work performed on the valve would notreasonably be expected to affect valve leakage; e.g.,painting.

7 ~ BASES —Wording was added to describe CONTROLLED LEAKAGE.The STS words describing safety injection flow were notincluded, as they were not applicable to Turkey Point.

8. BASES — Leakage from the RCS pressure isolation valvesis usually isolated during normal operation due tovarious system line ups. Wording was added to clarifythat only actual mass loss is considered in calculatingthe allowed limit of IDENTIFIED LEAKAGE.

9 ~ A surveillance is added to require testing of the checkvalves after actuation (flow through the valve), similarto the STS.

This testing is required prior to entering Mode 2 becauseoperating pressure is required to conduct the test.

Justifications:1 ~ A representative sample of the RCS can not be obtained

in this condition. This footnote is consistent withrecent FPL/NRC discussions at Turkey Point.

T. S. 3 4 '-8Justifications:

1. BASES — Due to the allowance given for transport ofreleases to site boundary (30 min.) and two (2) hoursfrom sample counting, isotopes with halflives, <10 minwould decay through 12 halflives which would make thedata very inaccurate even if these isotopes were stilldetectable. Industry assumptions are, that after 8 halflifes, the isotope is gone, i.e., <14 of isotopesoriginal activity remains, and after 120 min. (samplingallowance), 12 halflives for an isotope with a 10 min.half life, only .02~ is left.

*.Justifications:

1 ~ The ACTION statement has been reorganized for clarity tominimize confusion and avoid unnecessary thermal cyclingof the Plant.

2 ~ Added plant-specific information regarding specimenlocation and lead factor. Unit 3 capsule "V" has alreadybeen withdrawn.

T. S. 3 4.4.9.2

Justifications:1 ~ The change reflects the current Plant requirements.

2. The ACTION Statement has been reorganized for clarity tominimize confusion and avoid unnecessary thermal cyclingof the Plant.

T. S. 3 4.4.9.3

Justifications:1 As worded now, the Specifications require that HHSI

capability be isolated at exactly 380 F (when coolingdown) and implemented at exactly 380'F (when heating up).This is unnecessarily restrictive on the Plant. Therevised wording provides the Plant desired flexibility.

T. S. 3 4.4.10

Justifications:

'a ~

b.

Components of the Reactor Coolant System were notdesigned to provide access to permit inserviceinspections in accordance with Section XI of theASME Boiler and Pressure Vessel Code (1981 editionand addenda through Winter 1981).

1981 Code was not written when Turkey Point Units3 and 4 were designed.

c ~ The paragraph order has been revised for clarity.2.

a ~

b.

The changes proposed are consistent with applicablecode requirements.

In addition to the referenced changes, all Class l,Reactor Coolant System repairs on components shallmeet the original construction requirements for thatspecific repair.

c ~ Following repair on the RCS pressure boundary, therepaired area receives a hydrostatic presure testin accordance with ASME applicable code requirementswhich assure the structural integrity of therepaired area.

Justifications:1. The surveillance requirement frequency has been changed

to refueling invervals because the Plant must be shutdownto perform the surveillance. The 18 month requirementmay require an unnecessary shutdown and cycling of thePlant.

T. S. Number 3 4.5.1

Justifications:1. A limit of 24 months on "refueling" surveillance is a

management action to limit time between surveillances.

T. S. 3 4.5.2

Justification:1 ~ This special report is a duplicate requirement to

10 CFR 50.73 for ESF actuation.

2. As worded now, the tech specs recyire that HHSIcapability be isolated at exactly 380 F (when coolingdown) and implemented at exactly 380'F (when heating up).This is unnecessarily restrictive on the Plant. Therevised wording provides the Plant desired flexibility.

3.

4 ~

5.

Valve *-887 was removed from the list because it isoperated in the fully open position.Valve FCV-*-605 was added to the list because it is usedas an RHR throttle valve when it is necessary to bypassthe heat exchangers or valve HCV-*-758.

This comment has been added to provide flexibilityrequired for dual unit shutdown to prevent a severetransient on the Florida electrical grid. This commentis consistent, with previous discussions and agreementsduring RTS working meetings regarding dual unit shutdownsdue to electrical components being inoperable.

The correct position for valves 866A and B is "closed".FPL drawing 5610-T-E-4510 Sheet 2 shows these valves asnormally closed.

6. The surveillance requirement frequency has been changedto refueling intervals because the Plant must be shutdownto perform the surveillance. The 18 month requirementmay require an unnecessary shutdown and cycling of thePlant.

Justification:1. This special report is a duplicate requirement to

10 CFR 50.73 for ESF actuation.

2. This requirement states that all High Head SafetyInjection (HHSI) pump motor breakers shall be racked outwhen one or more RCS cold legs is less than or equal to275'F. This is impossible since the system is shared andmay be required operable for the other unit whose RCStemperature is greater than 275'F.

Justification:l. BASES

This change clarifies that the use of portableinstrumentation is acceptable. There is no permanenttemperature monitoring instrumentation installed on theRWST.

Justifications:1 ~ We need capability to test certain components, i.e.,

PASS, where manual containment boundary valves have tobe open.

2 ~

See recently approved amendments 114/108 dated 6/27/85.

Surveillance is not needed. Surveillances 4.6.1.3 ensurecompliance with 3.6.1.3 which maintains containmentintegrity. Also, surveillance is without time frame.

T. S. 3 4 '-1 '

Justifications:1 ~

2 ~

3 ~

L, is an option which Turkey Point does not intend toutilize. (Also applies to BASES.)

ACTION STATEMENT needs to address what must be done ifType B and C leakage exceeds .6 L, in modes 1, 2, 3 and4 (RCS temperature greater than 200'F).

This is consistent with containment integrity Tech Specs.

These surveillances are quoting a portion of 10 CFR 50App J. It has already been clearly established in4.6.1.2.1 that all types A, B, and C, tests will beperformed in accordance with 10 CFR 50 Appendix J. Itis confusing to be quoting "portion" of Appendix Jrequirements. (In fact, Item C.1 as written is inconflict with 10 CFR 50 Appendix J Section III.A.3.b.)law.

Justifications:1. .02L, is much less than is expected and has no basis.

This number, if implemented, would cause numerous unitshutdowns.

This paragraph represents a portion of combined leak rateallowable of 0.6 L,. 0.2 L, represents a value which, ifexceeded, willprovide timely indication of air lock sealfailure. However, the 0.6 L, limit of Tech Spec 3.6.1.1controls the overall containment leakage.

Justifications:1. Revised BASES for consistency with CTS.

Justifications:1. FPL analysis has determined that containment air

temperatures exceeding 120F up to 125F can beaccommodated for limited durations.

Justifications:1 ~ Change is made to maintain consistency on reporting

requirements.

2. Based on Turkey Point's CTS, all requirements foradjacent concrete surfaces have been met. The CTSrequire inspections on the end anchorage concretesurfaces, the mapping of the predominant visible concretecrack patterns, and the measurement of the crack widths.These inspections were done during the StructuralIntegrity Tests one-half year after the SIT and one yearafter the SIT. The inspection report determined that theconditions are satisfactory, therefore, the closeinspections were terminated.

Due to the design of the containment. at Turkey Point, theinspections of individual tendons and adjacent concreteareas are not easily performed. The equipment andpersonnel resources expended for this are not justifiedfor the probable limited benefits. Operations to removegrease from tendons and to inspect tendon wirebuttonheads for corrosion are already performed duringtendon surveillances; inspection of end anchorages andadjacent concrete should be done at the same time, ratherthan requiring an additional mobilization of personneland equiment for a second inspection at the time of ILRT.It is not desirable to perform this operation during anoutage; the need to position a crane near the containmentwill interfere with other outage-related activities.Also, note that the same tendons will not be inspectedat each surveillance; T.S. 4.6.1.6.1 requires randomselection.

Although the 5-year tendon inspection is greater than theILRT interval (40 months), the historical data supportsthis lower inspection frequency. In addition, proceduresare in use at Turkey Point which have acceptancecriteria based on normal containment pressure rather thanILRT pressure.

3. This visual inspection is already required by 10 CFR 50,App J, Section V.A (which is referenced in Tech Spec4.6.1.2.1). Both 10 CFR 50 App J and Reg. Guide 1.35,Sections B and C.3, refer to this as a "general"inspection intended to identify "widespread" problemareas. Therefore, no rigid acceptance criteria areprovided (as none were recommended in the revised TechSpecs). Additional discussion of acceptable inspectiontechniques has been added; requirements for recordkeeping have been added to facilitate the use of data forcomparisons to future inspections.

BASES — Selection of tendons may not be completelyrandom; certain tendons must be eliminated from possibleselection because they are inaccesible or would createhazards to personnel performing the surveillance, e.g.,tendons located near main steam vents.

Proposed Rev. 3 of Regulatory Guide 1.35 recommends theinspection of 24 of each group of tendons (dome, verticaland hoop). For Turkey Point, this equates to 3 dome, 4vertical and 5 hoop tendons.

This change will assure that the appropriate lift-offstress is used for each inspected tendon. Per the FSAR,the averacVe tendon prestress would be less than 0.7 timesthe ultimate strength; however, different tendons (evenwithin the same group) would have different effectiveprestress levels after elastic shortening. Zn addition,time-dependent prestress losses occur (creep, shrinkageand relaxation).The FSAR design stress levels are based on the end of the40-year life of the plant. Comparison of values obtainedat an intermediate point in plant life could result ina non-conservative assessment of prestress adequacy.

Justifications:The tech spec is reworded to clearly identify therequirements and to reflect previous agreements with NRCas noted below.

By letter dated November 28, 1978, the NRC requested alllicensees to respond to generic concerns aboutcontainment purging and venting during normal plantoperation. That letter stated that unlimited purgingduring normal operation would be permitted if licenseesdemonstrated that purge isolation valves were capable ofclosing against the dynamic forces of a design basisloss-of-coolant accident. The November 28, 1978 letteralso required that licensees evaluate the impact ofpurging on ECCS performance, the radiologicalconsequences of any design basis accident requiringcontainment isolation occurring during purge operationsand the containment purge and isolation instrumentationand control circuit designs. Pending completion of theabove, the NRC requested that licensees commit to ceaseor limit purging. FPL committed to limit purging forTurkey Point Units 3 and 4 during power operation ( 24power) to 200 hours per year for the site (200 hourstotal for both units). At that time, FPL stated that itintended to justify unlimited purging.

By letter dated December 13, 1979 (L-79-346), FPLprovided the results of the required evaluations. Theevaluation of the effect of containment purging on ECCSperformance indicated that the effect of purge operationupon the calculated pellet cladding temperatures issmall. An assessment of the incremental increase inradiological dose caused by containment purging duringa postulated loss-of-coolant accident (LOCA) indated thatthe anticipated total LOCA dose to be well within thelimits of 10 CFR Part 100. An evaluation of thecontainment purge instrumentation and control design didnot identify any single failure concerns.

The NRC in an evaluation dated August 31, 1981, indicatedthat the 200 hour purge limit, ECCS analysis, and theinstrumentation and control design, were acceptable. Ina letter dated February 10, 1983, the NRC provided theresults of a generic evaluation of the radiologicalconsequences of accidents while purging or venting atpower. To assure that the generic evaluation was validfor Turkey Point, the NRC verified the adequacy oftechnical specification limits on iodine equilibrium andvalve closure times. That evaluation also indicated thatthe dose contribution through open valves is small, andthat the total accident radiological consequences of such

such accident would be less than the dose guidelines of10 CFR Part 100.

FPL letters dated September 17, 1982 (L-82-407), March4, 1983 (L-83-120), and April 2, 1984 (L-84-86) providedinformation to demonstrate operability of the purge andvent valves. The NRC safety evaluation report datedAugust 17, 1984 concluded (subject to replacement of thebolts in the operators and installation of debrisscreens) that FPL had demonstrated the ability of thevalves to close against the buildup of containmentpressure in the event of a DBA/LOCA.

The August 17, 1984 NRC letter further stated that FPL'sproposed containment purge technical specificationsshould reflect the limitation of the opening angle forthe (48-inch and 54-inch) purge valves, and reflect (inthe basis) that the combined purges for both units willbe about 200 hours per year during power operation forthe site (200 hours total for both units). The proposedtechnical specifications provided by FPL as part of thetechnical specification upgrade project were consistentwith these requirements. Additional restrictions overand above what the NRC has reviewed and accepted are notjustified, and could impose a hardship on plantoperations. The guidelines for the upgrade project wouldpreclude the imposition of such requirements.

T. S. 3 4.6.2.2

Justifications:1. 2000 gpm flow rate cannot be verified during operation

unless the CCW system is put in accident configuration(e.g., two RHR Hx in service, all three ECC's operatingsimultaneously, NCC's isolated, CRDM's isolated, etc.).Plant does not contain individual flow indicators foreach cooling unit. Valve manipulation required tosupport this test would result in need to rebalance theCCW system which can only be done during shutdown.

Justifications:1 ~ This surveillance requires that the plant be in a

shutdown condition. An 18 month surveillance frequencymay cause FPL to have to shutdown for refueling early orcause an unnecessary outage.

T. S. 3 4.6.4

Justification:1 ~

2 ~

This table is derived from the CTS with the exceptionthat the steam generator blowdown valves were deletedfrom this table. The secondary system is considered tobe an extension of containment for containment isolationpurposes, and does not rely on any valves, includingblowdown valves, to provide containment integrity.Of the Table 3.6-1 isolation valves, only the purgevalves are explicitly called out in the LOCA analysis.Specific valve closure times have been eliminated fromthe tech spec.

3 ~ The removal of Notes 2, 5 and 6 is an editorial change.Note 1 was deleted because this information is notspecifically referenced in this tech spec.

Notes 4 and 7 were deleted due to removal of specificisolation times (see 2 above).

Justifications:1. Sample is obtained from the carbon tray after removal.

There is no provision to obtain a sample in accordancewith NRC Reg. Guide 1-52, Sections C.6.b.

2. This surveillance requires that the plant be in ashutdown condition. An 18 month surveillance frequencymay cause FPL to have to shutdown for refueling early orcause an unnecessary outage.

Justifications:1. There is no loop isolation capability from S/G's.

Justifications:1.

2.

This comment has been added to provide flexibilityrequired for dual unit shutdown to prevent a severetransient on the Florida electrical grid. This commentis consistent with previous discussions and agreementsduring RTS working meetings regarding dual unit shutdownsdue to electrical components being inoperable.

This surveillance requires that the plant be in ashutdown condition. An 18 month surveillance frequencymay cause FPL to have to shutdown for refueling early orcause an unnecessary outage.

Justifications:1 ~ Same ACTION Statement and logic train as Auxiliary

Feedwater tech specs.

2 ~ This surveillance requires that the plant be in ashutdown condition. An 18 month surveillance frequencymay cause FPL to have to shutdown for refueling early or,cause an unnecessary outage.

T. S. 3 4.7.1.4

Justifications:1 ~ This footnote is confusing. Zt is FPL s understanding

that an internal NRC position has been taken by sometechnical branches that this footnote should be removedfrom the STS.

T. S. 3 4.7.1.5

Justifications:1 ~ This change will allow reasonable time for corrective

maintenance and represents a 504 reduction from AOTallowed in CTS.

2 ~ Steam demand can exceed steam produced in MODE 3.

Justifications:1 ~ Use of the term "operable" has been modified to reflect

the fact that these pumps are not safety related. Thisconcept is consistent with CTS (Amendment 118/112, 8/86).Due to plant configuration, this path is normallyisolated.

2 ~

3.

This clarification ensures the unit(s) will not beshutdown because of inoperability of non-safety gradeequipment. This comment is consistent with CTS(118/112) .

A limit of 24 months on "refueling" surveillance is amanagement action to limit time between surveillances.

T. N. Number ~34.7.2Justifications:Note:

2 ~

The CCW 'Tech Spec was extensively discussed betweenNRC/FPL in Bethesda on March 15, 1988 (see attached NRCletter to FPL, dated 3/29/88). FPL's comments andjustifications reflect the discussion and agreements madeat that meeting.

The PTN design contains 3 heat exchangers, however, only2 heat exchangers, capable of removing design basis heatloads, are required. Although provisions are availablefor isolating passive failures, a passive failure of aheat exchanger is not a postulated design basis. NRCconcurred with this approach, in particular noting theseverity of the associated ACTION C (plant shutdown — asmarked up) and the extensive heat exchanger surveillanceprogram.

Note that the term "in service" was used and clarifiedvia footnote in lieu of the term "OPERABLE», whoseconventional definition may be misinterpretted in thisapplication."Valves, interlocks, and piping" are considered to beincorporated into the operability requirements for CCWpump/heat exchangers, or the operability of componentsserved by CCW, as appropriate. Accordingly, no specificreference to them is required in the LCO or ACTION.

3. The CCW design incorporated 3-100% capacity pumps. Thechange of AOT from 24 to 72 hours reflects the remotenessof a scenario which would disable the remaining pump(s)and provides additional time which may be required toeffect repairs. Additionally, the requirement tomaintain 3 pumps operable is more conservative thancurrent industry practices and the STS.

4 ~ See Justification 1 above. The wording, as marked up,reflects the LCO and system requirements.

5. With this pump inoperable, the plant still retains 2independent 1004 capacity CCW pumps. The CCW pumps havea good maintenance history and a 30 day AOT isreasonable.

6. These surveillances require the plant to be in shutdowncondition. An 18 month surveillance frequency may causeFPL to shutdown for refueling early or cause anunnecessary outage.

7 ~ BASES is revised to clarify design requirements for CCWand the reasoning behind the SURVEILLANCE wording.

8. Generic dual shutdown concern.

p,ll i(QII(4p

gy ~ gO

UNITED STATES

NUCLEAR REGULATORY COMMISSlONWASHINGTON. O. C. 20555

'tarch 29, 1988

'3/+.7. 0

Docket Nos. 50-?50and 50-251 RECElVED

AIRO5 assFlorida Power and Light Company

Turkey Point Units 3 and 4

LICENSEE:

FACILITY:

SUBJECT:

Nuctear Ucensfnt,

SUHHARY OF HEETING HELD WITH FLORIDA POWER AND

LIGHT COMPANY (FP&L) ON HARCH 15, 1988, REGARDINGTECHNICAL SPECIFICATIONS AND OPERABILITY OF INTAKECOOLING MATER AND COMPONENT COOLING WATER SYSTEHS

REFERENCE: TAC Numbers 63038 and 63039

A meeting was held in Rockville, Haryland on Harch 15, l988 with representativesof Florida Power and Light Company (FP&L) to discuss two matters related to

= Technical Specifications (TS) for,.the- Intake Cooling Mater ( ICM) and ComponentCooling Water (CCW) systems at Turkey Point Units 3 and 4. The first matterconcerned possible interpretations of "operability" of CCM heat exchangers.The second matter related to a revision of the TS for both systems as part ofthe orgoinq TS Revision Project.

2.

~0 erabilit of CCM Heat Exchan ers

In a letter to the licensee (FP&L) dated December 3, 1987, the staffindicated that each CCM heat exchanger should be declared inoperablewhen it becomes known that it cannot remove its design basis heat load(50~ of the required heat removal for the reactor unit). The staff hasnow improved this interpretation to address the matter from a systemsstandpoint. The staff agreed with FP&L that it is sufficient if two CCM

heat exchangers operating together can remove the total design basis heatload of two CCM heat exchangers, provided that a monitoring proqram is inplace to assure this capability continues to exist. The concern is thatfouling of the heat exchanger surfaces could degrade the heat removalcapability. To paraphrase, the staff agreed that it is permissible forthe plant to operate with one operating CCM heat exchanger which is notable to remove all of its design basis heat load, provided that a secondoperating CCM heat exchanger can fully compensate by removing more thanits design basis heat load, and provided that FP&L monitors the heatexchanger capability on a frequent basis. The licensee agreed to providea letter to the staff by Harch 18, 1988 agreeing to .these provisions. Thestaff will then issue a letter clarifying Enclosure 2 of the December 3,1987 letter.

Revision of TS for CCM and ICM

The staff indicated that it could nnt accept a relaxation of TSs whichwould entirely remove Allowable Outage Time (AOT) from the third ICM pump

~

~~

~~ ~

~

and the third CCW pump. This position is based on the current knowledge

of the importance of service water to core melt frequency as identifiedin the Byron PRA and is under generic review by the staff under GenericIssue 130. The staff suggested that simply relaxing the current 24-hourAOT for the ICW pumps to perhaps 72 hours was a better approach. Thestaff also suggested that the ICM strainers should be included in the TSwith an AOT. The licensee agreed to consider possible AOTs for the ICWand CCW pumps and the ICW strainers and propose these at a later date.In addition, the staff agreed that no AOT is necessary for the third CCM

heat exchanger because of the natural incentive for the licensee tomaintain the heat exchanger in an operable condition, and because of theTS interpretation described in (I) above.

The meeting agenda and reference material discussed at the meeting areprovided as Enclosure I to this letter. The attendance list is Enclosure 2.

4

Enclosures: As stated

Gordon E. Edison, Sr, roject Hanager-Project Oirectorate II-2Division of Reactor Projects-I/IIOffice of Nuclear Reactor Regulation

cc w/enclosures:See next page

T. S. Number 3 4.7.3

Justifications:Note:

1 ~

2 ~

The ICW Tech Spec was extensively discussed betweenNRC/FPL in Bethesda on March 15, 1988 (see attached NRCletter to FPL, dated 3/29/88). Unless otherwise notedbelow, FPL's comments and justifications reflect thediscussions and agreements made at that meeting.

"Two headers", as well as "valves, interlocks and piping"are considered to be incorporated into the operabilityrequirements for ICW pumps, or the operability ofcomponents served by ICW, as appropriate. Additionally,due to the open design and operation of the system, aspecific requirement for headers was deemed unnecessary.Accordingly, no specific reference to them is required.

A 3.0.4 exclusion is added, consistent with CCW tech specand NRC/FPL discussions.

3 ~

'4~

The ICW design incorporates 3-1004 capacity pumps. Thechange of AOT from 24 to 72 hours reflects the remotenessof a scenario which would disable the remaining pump(s)and provides additional time which may be required toeffect repairs. Additonally, the requirement to maintain3 pumps operable is more conservative than currentindustry practices and the STS.

These surveillances require the plant to be in a shutdowncondition. An 18 month surveillance frequency may causeFPL to shutdown for refueling early or cause anunnecessary outage.

5. BASES is revised to clarify that design "and operation"of this system ensures required cooling capacity. Thisclarification recognizes that operator actions may berequired in certain scenarios, as previously discussedwith NRC.

6. As discussed between FPL/NRC at the March 15th meeting,FPL is to propose wording regarding the basket strainers.This wording will be provided at a later date.

ill I'tOu%8

)0

I 0

O

UNITEDSTATESNUCLEAR REGULATORY COMMISSlON

WASHINGTON, o. C. 20555

~1arch 29, 1988

Docket Nos. 50-250and 50-251 RECElVKD

APROS N8Florida Power and Light Company

Turkey Point. Units 3 and 4

LICENSEE:

FACILITY:

SUB<>ECT:

Nucfear Uo.nsfnt.

SUMMARY OF MEETING HELD WITH FLORIDA POWER AND

LIGHT COllPANY (FP5L) ON MARCH 15, 1988, REGARDINGTECHNICAL SPECIFICATIONS AND OPERABILITY OF INTAKECOOLING WATER AND COMPONENT COOLING WATER SYSTEMS

TAC Numbers 63038 and 63039REFERENCE:

1. ~0 erabi lit of CCM Heat Exchan ers

In a letter to the licensee FPItL( ) dated December 3, 1987, the staffindicated that each CCW heat exchanger should be declared inoperablewhen it becomes known that it cannot remove its desiqn basis heat load(50" of the required heat removal for the reactor unit). The staff hasnow improved this interpretation to address the matter from a systemsstandpoint. The staff agreed with FphL that it is sufficient if two CCM

heat exchangers operating together can remove the total design basis heatload of two CCM heat exchangers, provided that a monitoring proqram is inplace to assure this capability continues to exist. The concern is thatfouling of the heat exchange surfaces could degrade the heat removalcapability. To paraphrase, the staff agreed that it is permissible forthe plant to operate with one operating CCM heat exchanger which is notable to remove all of its design basis heat load, provided that a secondoperating CCM heat exchanger can fully compensate by removing more thanits design basis heat load, and provided that FP5L monitors the heatexchanger capability on a frequent basis. The licensee agreed to providea letter to the staff by March 18, 1988 agreeing to these provisions. Thestaff will then issue a letter clarifying Enclosure 2 of the December 31987 letter.

S

2. Revision of TS for CCM and ICM

The staff indicated that it could nnt accept a relaxation of TSs whichwould ent.irely remove Allowable Outage Time (AOT) from the third ICMan d the third CCM pump. This position is based on the current knowledge

ir pump

A meeting was held in Rockvi lie, Maryland on March 15, 1988 with represent'ativesof Florida Power and Light Company (FP5L) to discuss two matters -related toTechnical Specifications (TS) for the Intake Cooling Water ( ICQ) and ComponentCooling Water (CCW) systems at Turkey Point Units 3 and 4. The first. matterconcerned possible interpretations of operability" of CCW heat exchangers.The second matter related to a revision of the TS for both systems as art ofthe ongoing TS Revision Project.

ems as par o

of the importance of service water to core melt frequency as identifiedin the Hyron PRA and is under generic review by the staff under GenericIssue 130. The staff suggested that simply relaxing the current 24-hourAOT for the ICM pumps to perhaps 72 hours was a better approach. Thestaff also suggested that the ICM strainers should be included in the TSwith an AOT, The licensee agreed to consider possible AOTs for the IChtand CCM pumps and the ICW strainers and propose these at a later date.In addition, the staff agreed that no AOT is necessary for -the third CCM

heat exchanger because of the natural incentive for the licensee tomaintain the heat exchanqer in an operable condition, and because of theTS interpretation described in (1) above.

The meeting agenda and reference material discussed at the meeting areprovided as Enclosure I to this letter. The attendance list is Enclosure 2.

Enclosures: As stated

cc w/enclosures:See next page

Gordon E. Edison, Sr. Project, ManagerProject Directorate II-2Division of Reactor Projects-I/IIOffice of Nuclear Reactor Regulation

T. S. 3 4.7.4

Justifications:l. The requirements of this tech spec are incorporated into

the CCW tech spec through the requirement to monitor andmaintain in service CCW heat exchangers capable ofremoving design basis heat loads. A tech spec for aspecific ultimate heat sink temperature is inappropriate;ultimate heat sink temperature is one of several factors(i.e., heat. exchanger performance characteristics, flow)involved in determining heat removal capability.

T. S. 3 4.7.5

Justifications:1 ~ As there is only one (1) CREVS, the word "the" was

substituted for "each" for clarity.2 ~

3 ~

4 ~

The added words had been deleted by the NRC. Theirdeletion would imply the requirement to visually inspectthe entire CREVS. This would require a major effort, aslarge parts of the system are not accessible withoutdisassembly.

CTS AOT is 3-1/2 days. STS is seven (7) days for aredundant system (e.g., two (2) independent systems) .

This comment has been added to provide flexibilityrequired for dual unit shutdown to prevent a severetransient on the Florida electrical grid. This commentis consistent with previous discussions and agreementsduring RTS working meetings regarding dual unit shutdownsdue to electrical components being inoperable.

T. S. 3 4.7.6

Justifications:1 ~

2.

This inspection cannot be done at power.

Footnote clarifies "inaccessible".3 ~ A limit of 24 months on "refueling" surveillances is a

management action to limit time between surveillances.

T. S. 3 4.7.8.2

Justifications:Per the Turkey Point Appendix R upgrade, the listedspray/sprinkler systems were superseded for the followingreasons:

a ~ 4160V Switchgear Room Louver Spray —This system wasrendered obsolete when the opening it was protectingwas filled with a 3-hour fire barrier.

b. EDG Building Hater Curtain — This manual system onthe outside of the EDG Building was superseded byan automatic suppression system in the EDG rooms.

c ~ Control Room Guardhouse Sprinkler System — Thissystem protected the old guardhouse, which wasconstructed of combustible materials. A non-combustible guardhouse has since been constructed.

Justifications:1. These hose stations are not all shared.

Justifications:1. BASES — This evaluation is inconsistent with the

requirements of the LCO.

T. S. 3 4.9.1

Justification:1 ~ RTS wording is revised to 0.95 per Amendment 132/126.

The restriction on K,<< is CTS and has been lessened bythis recent amendment. (This justification also appliesto BASES.)

2 ~

3.

The term "equivalent" would mean 100 gpm of 1950 ppm;this would require 2 charging pumps in Mode 6. No raterequirements for boration in Mode 6.

The tech spec is not requiring valves be "locked closed".

I'ustifications:

l.

2-

Due to expected maintenance activities during this mode,the suggested 4 hour time period may not be achievable.

A limit of 24 months on "refueling" surveillances is amanagement action to limit time between surveillances.

Justifications:1. Due to expected maintenance activities during this mode,

the suggested 4 hour time period may not be achievable.

Justifications:1 ~ Consistent with tech spec 3.3.2 (Table 3.3-2, item 3.c.4)

for ESFAS operability in Modes 1-4.

Justifications:1. There is no need for an upper limit on water level.

Justification:l. All design calculations are based on the time the fuel

was last critical, not the time the fuel was placed inthe Spent Fuel Pool.

2. Wording was clarified to state 120 days of decaynot 120 days in cask.

Justifications:1. Consistent with Tech Spec 3.3.2 (Table 3.3-2, item 3.c.4)

for ESFAS operability in Modes 1-4.

Justification:1 ~ BASES — Current Technical Specification 3.17 specifies

a maximum K,ff of 0. 95 applicable to the proposed two-region spent fuel pools at PTN. This is also discussedin RTS 5.6.1.Refer to Justification 3/4.9.1, g1.

T. S. Number 3 4.10.1

Justifications:1 ~ The term "equivalent" would mean 100+ gpm of 1950 ppm;this would require two (2) charging pumps. Tech Spec

3.1.2.3 ACTION allows two (2) charging pumps to beinoperable for a limited time.

Justification:1. Depending on specific unit configuration, the operating

staff needs flexibility to deal with the LCO withoutpotentially creating an unexpected operating condition.

Justifications:1. Recently approved evaluation by FPL's Fuels Analysis

Group allowed withdrawal of two banks during control rodtesting.

Justifications:1 ~

2 ~

3 ~

The Unit 3 and 4 containment bleed lines constitute asmall release rate to the plant vent. This bleed rateis continuous unless containment isolation occurs, atwhich time the bleed lines are automatically isolated.All realses via this pathway are continuously monitoredby the installed plant vent monitors for noble gas,particulate and iodine activities. This monitoring willaccount for all activity released from the containmentby the bleed line pathway.

Note 6: No need to sample if coolant activity has goneup unless the effluent monitor has also gone up.Therefore, the statement should require both conditions.(This wording is consistent with recently issuedspecifications at St. Lucie and Palo Verde.)

Note 7: FPL's comment returns the wording to a positiveformat where analysis is required when both requirementsare met; with the negative format, it appears analysisis required when either one or both of the conditions arenot met. Without this comment, this tech spec wouldrepresent a significant change to the CTS (Amendment103/97) .

Justifications:1 ~

2.

The plant has the ability to monitor only the in-servicetank, not all 6 tanks simultaneously. This comment isconsistent with previous FPL/NRC discussions.

Footnote has been added for consistency with Table 3.3-8, Item 2 which covers the use of grab samples forinoperability of continuous monitors.

T. S. Number 3 4.11.3

Justifications:1 ~ All of our shipments do not go directly to a burial site,

some are shipped to a waste processor for volumereduction prior to burial.Only "wet" radioactive wastes (not all radioactivewastes) are required to be dewatered or solidified.

Justifications:1 ~ BASES — Turkey Point's outside storage tanks, if

released, will be contained within the site boundary andmixed with our cooling canals which are completelyenclosed.

T. S. Number 3 4.12.1

Justifications:1 ~ The existing requirement (if sample is >1.0 units and >10

times most recent control sample) is consideredappropriate to trigger gamma isotopic analysis ofspecific samples. The monitoring program is conductedby the State of Florida under a joint Turkey Point/St.Lucie contract. This comment provides a consistentapproach between Turkey Point and St. Lucie and reflectsthe requirements of the CTS (RETS Amendment 103/97).

T. S. Number 3 4.12.3

Justifications:1. The State Laboratory participates in the EPA Inter-

comparison Program for media, isotopes and analysis notrequired by our program in addition to those that arerequired. The proposed wording would have FPL generatea report if the State chose not to participate in one ormore of the "other" media, isotopes or analysis.

T. S. Number 5.1 Fi ure 5.1-1

Justifications:1. This figure willbe redrawn to reflect the required level

of detail. An original of the redrawn figure will beprovided to the NRC.

T. S. Number 5.3

Justifications:1 ~

2.

Initial fuel load enrichment is historical informationand is not required. Reload fuel enrichment will be achanging value and will be adequately documented in thereload analyses provided in the FSAR.

This wording reflects recent CTS Amendment.

T. S. Number 5.7 Table 5.7-1

Justifications:1. These changes are the result of an engineering evaluation

of cyclic loading. An FSAR change is being processedwhich will change these values in FSAR Table 4.1-8.

T. S. Number 6.2

Justifications:The Operations Superintendent is a senior manager withresponsibility for operations, outage scheduling,chemistry, reactor engineering, and health physics. Inthe absence of the Plant Manager, the OperationsSuperintendent normally assumes the duties of the PlantManager. The Turkey Point Plant organization does nothave an Assistant Plant Manager position. The proposedrequirement that the Operations Superintendent eitherhold or have held an SRO license would continue to ensurethat he has the knowledge and experience commensuratewith his level of responsibility. The OperationsSupervisor, who has direct responsibility for operations,will continue to hold a current SRO license. Theserequirements provide rea'sonable assurance that decisionsand actions during normal and abnormal conditions willbe such that the plant will be operated in a safe andefficient manner.

T. S. Number 6.3

Justifications:1. This change is required for consistency with the proposed

change to 6.2.2.i.2. The March 28, 1980 NRC letter has been superseded. RO

and SRO qualifications are now covered by 10 CFR 55 andANSI 3. 1( 1981.

T. S. Number 6.4

Justifications:l. The 10 CFR 55 rec{uirements have been included in the body

of the document, and are no longer in Appendix A. TheMarch 28, 1980 NRC letter has been superseded by 10 CFR55 and ANSI 3.1, 1981.

T. S. Number 6.S

Justifications:

2 ~

These changes were made to reduce management time spenton PNSC activities. The problem of excessive PNSCmeetings and line management involvement was noted by theNRC in Safety Review Team Inspection Report 87-49(attached). In response to this concern, FPL proposesadapting the Palo Verde Unit 3 T.S. regarding technicalreview and control activities. This T.S. removes thedirect responsibility for review of various plantdocuments and programs from the PNSC. FPL believes thatthe delegation of review responsibility adequatelyresponds to the NRC's concern, and is consistent with theNRC's interpretation of PNSC responsibilities (variousmemos attached).

This type of generic composition of the CNRB has beenpreviously approved by the NRC for other plants (e.g.,Fermi II). It will eliminate the need for future licenseamendments due to changes in titles of CNRB members.

~4 itOy~C p

n4

Docket Nos. 50-250and 50-251

UNITED STATESNUCLEAR REGULATORY COMMISSION

WASHIHOTOlliD. C. 20555

February 10, 1988

7$. Ns

RECFnrED

FEB 12 SINudear uce&IC

Florida Power and Light CompanyATTN: Hr. C. 0. Woody

Execut1ve Vice PresidentP. 0. Box 14000Juno Beach, Florida 33408

SUBJECT: SAFETY REVIEW TEAH INSPECTION REPORT XO. 50-250/87-49)50-251/87-49

Gentlemen:

This letter forwards the report of the Safety Review Inspection conductedby Hr. C. Haughney and other NRC personnel during the period December 7-11,1987, of activities at the Turkey Point Power Plant, authorized by NRCOperating Licenses DPR-31 and DPR-41, and to thc discussion of our findingswith Hr. C. Baker and ot;hers at thc conclusion of the inspection.

The 1nspectfon consisted of an examination, on a sampling basis, of activi-ties fn the areas of safety rev1ew pursuant to 10 CFR 50.59, and the on-siteand off-site review coIIInfttees. The fnspcctfon included assessment of safetyrev1ew documentat1on, interviews with site and corporate personnel, and atten-dance at coaIIfttee meetings.

The inspection team found ev1dence of significant improvement fn the qualityof safety evaluations over the past year. Reaafnfng steps to assure contfnua-tfon of this trend include thc tfmly issuance of your comprehensfve draftprocedure covering 10 CFR 50.59 evaluations and assoc1ated training forreviewers.

Additional areas for improvement fdcntff1ed by the fnspect1on team include:improving the efficiency of Plant Nuclear Safety CoaIIfttee (PNSC) reviews,identffyfng currently valid safety evaluations, classification of guidancedocuments regarding act1ve check valve failures, the retent1on of designcalculations and timely completion of Request for Assistance evaluations(REAs).

Some of the items identified by thc team may be potential enforcement f1ndfngs.Any enforcement actions will be identified by Region II.In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosurew1ll be placed in the NRC Public Document Room.

florida Pawer and Light Ceapany -2- february )0, 1988

Should you have any questions concerning this inspection, please contact eeor Hr. L. Norrhol@ (301-492-0956) of this office.

Steven A. Varga, DirectorDivision of Reactor Pro5ects, I/IIOffice of Nuclear Reactor Regulation

Enclosure: Inspection Report 50-250/87-49; 50-251/87-49

cc w/enclosure: See next page

U.S. NUCLEAR REGULATORY CORIISSIOKOFFICE OF NUCLEAR REACTOR REGULATION

OIVISION OF REACTOR INSPECTION AND SAFEGUAROS

Report No.: 50-250/87-49 and 50-251/87-49

Oocket No.: 50-250 and 50-251

Licensee: Florida Power and Light CeapanyP.O. Box 14000Juno Beach, Florida 33408

Inspection At: Turkey Point Power PlantFlorida City, Florida

Inspection Conducted: December 7-11, 1987

Tea< Leader:ar es . auecial InSp s t n 8 anch, RR

at ne

Teae Members:e . or

Teaa I peO~ d

n isal and Pevelopeent Sec. 1

ivya e gne

/x~ A'. ~ ns, , o ree

Consultant: Gary J. verbeck

ate gne

Approved By:ar es . au , e

Special Inspec i Branch, IS, NRRe gne

This one-week team inspection evaluated safety review activities conductedby the licensee pursuant to Title 10 Code of Federal Regulations part 50.59.Related functions of the Plant Nuclear Safety Cemfttee (pNSC) and thcCorporate Nuclear Review Board (CNRB) were also assessed. The fnspectfonconsisted of record reviews, interviews w1th cognizant personnel, and atten-dance at meet1ngs. The purpose of the inspection was to determine whetherappropriate issues were subjected to 10 CFR 50.59 reviews, whether correctdeterminations were made with respect to unrevfewed safety qucst1ons, andwhether documentation of these reviews were complete 1n describing the basesand rationale for conclusions.

2.0 Sugar of Sf nificant Findin s

Steady improvement in the quality and completeness of 10 CFR 50.59 safetyevaluation documentation was observed over the past year. Further improve-ment and the continuatfon of current skills fn this area should be aided bylicensee initiatives to issue comprehensive procedural guidelines and toprovide training to cvaluators.

In general, recent safety evaluations reviewed were sufficiently detailed todemonstrate, as a stand-alone document, the logic and bases for determinatfonsregarding potential unrcv1ewed safety questions. One notable strength was thcuse of prior 10 CFR 50.59 review and Plant Nuclear Safety Coeafttcc (PNSC)authorization for temporary modifications to the plant. including Jumpersand lifted leads.

Some areas werc identiffed as weaknesses which should receive attention by thelicensee. Despite attempts at mitigation, the volume of material requiringPNSC review has resulted in long and frequent meetings, soee brfcf reviews,and diversion of management from normal duties. Some safety evaluationsreviewed were no longer valid, but no systca was fn place to identify these.Licensee attention should be directed to insuring that Request for EngineeringAss1stance (REA) evaluations were completed fn a tfmely manner. Finally, the .

licensee should reevaluate its position regarding the treatment of check valveactive failures and the retention of design calculations.

3.1 Design Change Process

Licensee procedures for all design and safety analyses performed by the PowerPlant Engineering Department were contained 1n gualfty Instruction JPE-gI-3.2,Revision 3, dated July 30. 1982. The gu1dancc provided in this procedure wasmin1mal with respect to 10 CFR 50.59 evaluations in that the definition ofunreviewed safety question was quoted from the rulc, and an additional statementwas made to the effect that thc discuss1on should clearly indicate the basesfor the conclusions reached.

Station Administrative Procedure 0190.15, Plant Changes and Nodfffcatfons(PC/H), dated October 26. 1987, prov1ded essentially the same informationas JPE-(1-3.2. Station Administrative Procedure 0190.22, Changes, Tests,and Erperiments, dated December 12, 1986, offered some additional guidance

Plant Nuclear Safety Comfttce (PHSC)

Inspection of PHSC consisted of discussions with licensee personnel, includingPNSC members, review of documents, and attendance at the PHSC meeting held onOecember 8, 1987.

One of thc events that prompted this inspection was PNSC Meeting No. 86-232held on August 31, 1986. That meeting was conducted by means of individualtelephone calls from the Sh1ft Technical Advisor to each of the PNSC members.PHSC Meeting 86-232 was conducted to gct concurrence from thc PNSC to start upUnit 4 without f1rst repairing a small, identified reactor coolant leak. Theleak was through an instrumentatfon port column assembly conoseal fitt1nglocated on the reactor vessel head. The plant was subsequently restartedand operated at power, Ourfng an outage fn March 1987, thc licensee deter-mined that a significant amount of boric acid from the leak had accumulatedon the reactor vessel head. At that time. the licensee also dctermfned thatcorrosion rates of materials. fn contact with the boric acid. may have beengreater than those on which the PNSC based the decision to startup the plantin August 1986. This event was documented fn licensee Letter No. L-87-186,dated April 27, 1987, to the U.S. Nuclear Regulatory Coawfssfon.- The conse-quences of this event indicated that thc practice of obtaining PNSC concurrenceby walking-around items or by serial telephone calls can result fn less thanadequate reviews.

The lfcensee revised Administrative Procedure (AP) 0110.4, 'Plant NuclearSafety Coanfttee General Procedure," dated October 6, 1987, to clarify and .

tighten the requirements for holding PHSC meetings. The October 6, 1987rev1sfon of AP 0110.4, dfd not allow walk-around PNSC concurrences. Theprocedure also required that telcon PHSC meetings were to bc conference calltype, with the members talking to each other. The following fteas could not beapproved utflfzfng the telcon PNSC meeting:

Any item which required a written safety evaluation for approval;Any item wh1ch involved a change fn the FSN or TechnicalSpccfffcat1ons;Any plant changes or modfffcatfons, controlled plant work orders,and process sheets.

One of the lfcensce personnel interviewed by the team was the Chafraan of thcPNSC, the Plant Manager-Nuclear. The Plant Manager-Nuclear dfd not cha1r manyof the PHSC meetings; instead, thc ma)ority of the PNSC meetings were chafredby the vice chairman, the Operations Superintendent-Nuclear. The plant managerstated that he intentionally dfd not chair many of the meetings because byhavfng the vice chairman conduct the meetings, a more independent review ofPHSC items was obtained. He felt this practice was necessary because, as plantmanager, he had to review each item and give final approval for ft to beissued.

Thc team verff1ed by review of licensee quality assurance documents and discus-sions with licensee quality assurance personnel, that thc licensee's qualityassurance department was monftorfng PHSC act1vf ties. Two of the gh documentsreviewed were gA audits gAO-PTX-86-723 and gAO-PTN-87-818. These annual auditswere conducted to verify that the PNSC was meeting thc requirements of TS 6.5.1and the related plant 1mplementing procedures. A third QA document reviewedwas Corrective Action Request (CAR), Unit 4 Conoscal Leak, CAR-87-019. This

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CAR reported noncompl fances 1dent1f1ed 1n a gA per foraance monftor1ng act1v1tyof events related to the Unft 4 conoseal leak. CAR-87-019 reported that PNSCNeet1ng 86-232 was conducted on August 31, 1986, by the sh1ft techn1cal advfsormak1ng 1nd1v1dual telephone calls to each PNSC meaber,.and that 1t was coeaonpract1ce to obta1n PNSC concurrences 1n th1s way rather than 'by conferencecalls, as now spec1f1ed by AP 0110.4.

The team made the follow1ng two observat1ons dur1ng the 1nspect1on:

I. The l1censee d1d not have a formal tra1n1ng program for PNSC membersor the1r alternates. In 1984, the 11censee held a one-day tra1n1ng.sess1on, wh1ch was taught by a contractor, for the PNSC oeabers. In June1987, the PNSC coord1nator prepared a PNSC tra1n1ng gu1del1ne manual wh1chconta1ned mater1al relat1ve to the operat1on of the PNSC. Th1s aanual wasa requ1red read1ng type tra1n1ng course. These tra1n1ng aanuals were sentto each of the PNSC members w1th a letter stat1ng that the manual was toprov1de tra1n1ng for the PNSC members and the1r alternates. At the t1me ~

of th1s NRC 1nspect1on. the PNSC coord1nator had rece1ved documentat1onback from four of the members show1ng that they and the1r alternates hadrev1ewed the mater1al referenced 1n the tra1n1ng aanual. At the ex1tmeet1ng, the 11censee stated that they were develop1ng tra1n1ng programsfor PNSC members and the1r alternates.

2. v ha d w1 h la volume of mat r1al that 1sv e d b NS . Th P n 0

1n ast n over two hour In a 1t1on to t two sche u emeet ngs every wee, ca meet ngs were frequently held. Through December13, 1987, 334 PNSC meetfngs had been held dur1ng 1987. 01scussfons w1th11censee personnel 1nd1cated that the w1n1ae of tw1ce-weekly aeet1ngshad been held for a number of years. Because of the volwe of mater1aland the frequency of the meet1ngs, the meet1ng agenda and package of 1temsto be rev1ewed were not prov1ded to the members unt1l they arr1ved at themeet1ng. Th1s pract1ce d1d not g1ve the members an opportunfty to faw)l-1ar1ze themselves w1th the 1teas on the agenda. To help allev1ate theproblems of the large number of 1tems to be rev1ewed. the PNSC has requ1redthat a sponsor be present at the aeet1ng to present each 1tem. Thesponsor prov1ded expert1se to the PNSC on each 1tea presented. Any 1temthat d1d not have a sponsor present at the meet1ng was tabled. In add1-t1on, any 1tem that any member quest1oned was sent back to the preparerto resolve these quest1ons pr1or to PNSC approval.

Th1s observat1on descr1 s a condft1on h 1 coeeon to man fac111t1 w1thtra t ona s

nd xs roce u 1 e an as a result the number o rocedures sub ect to

ur er c undedh fac tha man f rocedu c ld nd r c an es severa t1mes a

r. Tn rv

chan e rev1ew.

h n r nf n u and ro edure

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In addition to the requirement to revIcw procedures and procedure changes, TS6,5,1,6.f requires a revfew of facf1 1 ty operations to detect potcntfal safetyhazards. Mhcn PNSC members were asked how they conducted such a revfcw theywere fnft1ally unable to respond. After some thought, one member suggestedthat such a revIew was accomplished by the review of Licensee Event Reports(lERs). However, the team pointed out that review of LERs was sere explicitlycovered by TS 6.1.5.6.k, which required review of all reportablc events. Theteam also asked PNSC members whether they had,ever been requested to performspecial reviews and fnvestfgatIons by the CNRB. These revfews are a TSrequirement fn 6.5.1.6.g. None of the PNSC members fntcrvfewcd could recallconductIng such a review. Admittedly. both of these TS requirements are verybroad and thc absence of their rcvIews by the PNSC would not, fn and of itself,constitute a violation of TS. However, the team questioned PNSC members as towhether they had ever made those review requfrements meting agenda Items. Inthe team's view, the PNSC fs so burdened with the extensive revfcw of routineprocedure changes that ft fs effectively unable to devote time to reflect uponbroad safety Issues. This issue fs c ounded b the fact that PNSC membersare kc station mana crs o must v e t e r t se ann act v cs

an ours an su rv s on o e r e ar tmcn s. am cons rs ate c u s n ct vc c cou vc 0

a rnativ fr n ndn f n sCertain t s of ke roccdurcs such as cnc o ratin edu s and

cedurcs c ld vf d bic see n icatcd t at e would c n fd r

x mce fnn T a n n

nstftut n a ro riate adminfstratIve con ro cha es tha would a ow theareduce t c rocc ure review work oa on the NSC.

Based on this lfafted inspection, the team determined the followfng:

The PNSC was conducting adequate revfcws.

Adequate admfnfstratfvc procedures had been fmplemnted to control PNSCactivities so that TS requirements were being mat.

The gA department had performed monftorfng activities and audits ofPNSC activities.

6.B ~dlhThe Inspection team conducted an exit meeting on Decciabcr 11, 1987, to providea sNNaary of issues identified during thc Inspection. The licensee'sreprescntatIves at the ex1t meeting are Identified In Attachment A. The scopeof the inspection was dfscussed, the observations werc presented for each area1nspected, and team members responded to questions from the licenseerepresentatives.

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UNIT50 STAT55NUCI.EAR REQULATORY COMMISSION

IVAN'NCAN,P. C. 200%i

February 16, 1978

Docket No. 50-313

Arkansas Power I Light CompanyATTN: -Nr. Nlliam Cavanaugh. III

Executive Director, Generationand Construction

Post Office Box 551Little Rock, Arkansas 72203

Gentlemen:

By letter dated January 10, 1978, you requested an interpretation ofTechnical Specification 6.5.1.6 for Arkansas Nuclear One - lJnit No. l.This request was the result of a difference in interpretation betweenArkansas Power 5 Light Company staff and NRC Regional Inspection andEnforcement personnel, and concerns the technical specification wordingwhich states that: "The Plant Safety Comaittee (PSC) shall be

responsible for review of ...'. Your interpretation of this require-ment is that items of review for which the PSC is responsible could be

delegated to individuals or groups outside the PSC. You state thatthis interpretation is contrary to that of OINK personnel who requirethat the PSC perform all reviews and investigations.

The proper interpretation of this provision fs as follows:

The PSC may delegate revie~ responsibility to individualsor groups outside of the forwal composition of the PSC.

However, since the PSC is responsible for the performanceof these reviews, the review function is not completeuntil the PSC acts upon the results of the review in a

formal manner and documents such action. Documentation-@ay take the form of a siaple notation in 'the minutes ofa PSC aeeting or may be as elaborate as the preparationof a formal PSC report.

gaA alCig~

~ates

7-~ 6.sUNITED STATES

NUCLEAR REGUI.ATORY COMMISSlONWASHINGTON, D. C. 20555

'April 28, 1978

HEHORANDUH FOR: B. M. Grfer, Director, RIJ. P. O'Ref lly, Dfrector, RIIJ. G. Keppler, Director, RIIIG. L. Hadsen, Acting Director, RIVR. H. Engelken, Director, RV

FROH:

SUBJECT:

J. M. Snfezek, A/D for Field Coordination, DROI, IE

INTERPRETATION OF TECMNICAL SPECIFICATION REQUIREYiENTSFOR ON-SITE COMMITTEE REVIEW FUNCTION

On August 12, 1977, I forwarded a memorandum to all RO5NS Branch Chiefsregarding the on-site committee review function. Included fn thismemorandum was an interpretation by K. R. Goller that was intended toclarify how the on-site conmfttee was intended to function. By memo-randum to Karl Goller dated August 2, 1977, Jfm Hurray said that he .

had reviewed Goller's memorandum of June 27 and could not agree withthe legal conclusions and that he would be unable to support enforce-ment action fn those instances where the On-Site Review Committee dfdnot formally convene a quorum to conduct comfttee business. Sincethat tflie we have held several meetings with NRR and ELD. There appearsto be no conflict regarding the intent of collllfttee review requirements.NRR expects the on-site review committee to act on tasks in their areaof responsibility while the coomtttee fs 1n sess1on and to record theresults of their actions fn coamittee mfnutes. This fs not to saythat the fndfv1duals must meet to review issues, o&n y they must meetto act on the various'ssues and that these actions should be recorded.Certain portions of the on-site corefttee's overall responsfbf11ty maybe reviewed by subcomfttees or individuals who make recommendationsto the on-site colrmlfttee.'he on-site committee must act on theserecoreendatfons; however. ft fs not required that the on-site committeeactually conduct a detailed review of the issue fn question. DROIand ELD concur wfth NRR's views on this matter.

|lith regard to enforceabflfty of current technical specificationrequirements fn this area, we note that the above interpretation ofhow the colmlfttee should function fs completely consistent with NRC'slong standing interpretation of the on-site review function. While wefeel that this fntent fs implicit fn the language of the TS, we cannotdf'sagree with ELD's view that the language of the TS may not be

'I

Regional Directors7.s:

April 28, ] 9]3

enforceable should the matter be taken to a hearing. To resolve thisissue for future plants, NRR intends to revise the language of theSTS. Mith regard .to those facilities that are already licensed, itwas pointed out that the intent of this specification appears to beclearly understood by all but a few licensees. Therefore, to avoidunnecessary TS changes NRR will request licensees who are not meetingthe intent of. the specifications to submit a change to their TS. Inconsideration of the difficulties of a generic resolution to thisissue, we find this approach acceptable.

Please call if you have any questions on this matter.

cc: ROIHS BCs, Regions I-VEnforcement Coordinators,

Regions I-VE. L. JordanJ. Nurray, KLDJ. HcGoughT. J, Carter

H, Sniezek, A/D or FieldCoordination

Division of Reactor OperationsInspection, IE

CONTACT: G. L. Constable(492-8019)

T. S. Number 6.8

Justifications:1 ~

2 ~

This change reflects the current T.S. wording.

The footnote was deleted because FPL considers itinappropriate to include state regulations in the T.S.This footnote is not in the current T.S.

3 ~ Current T.S. 6.14 contains the same requirements as RTS6.8.4.a; however, no specific systems are identified.FPL's position is that inclusion of systems could beinterpreted as limiting the program to those listedsystems.

4 ~ This change reflects the changes to 6.5.1 and 6.5.2.

T. S. Number 6.9

Justifications:The initial startup reports were completed in 1972 and1973, and met the requirements of the original FSAR. TheFSAR has subsequently been revised several times. Itwould not be productive to compare the initial StartupReports to the current FSAR. All subsequent StartupReports will meet or exceed the current FSAR. Theproposed wording is taken from the current T.S.

2 ~

3.

4,

This wording is in the current T.S., and reflects FPL'sNRC-approved response to NUREG-0472.

There are no other uranium fuel cycle sources near theTurkey Point site.This change, which replaces the PS<R version with currentT.S. 6. 9. 3. d (except for reporting requirements), reflectsthe FPL-proposed change to T.S. 3.2.2. The NRC-proposedT.S. would require FPL to obtain additional analysis togenerate the proposed Peaking Factor Limit Report.Current T.S. 6.9.3.d was approved by the NRC onMarch 17, 1982.

5.

6.

These changes were made to conform with 49 CFR.

This is a duplication of the requirements of the AnnualEnvironmental Monitoring Report, and is not part of STS.

7 ~ This change is for consistency with other reportingrequirements of 10 CFR 50.4 and T.S.6.9.1.

T. S. Number 6.10

Justifications:1. These current T.S. words were recently approved

(Amendment 103/97), and are consistent with St. LucieUnits 1 & 2. Any change to these recpxirements wouldresult in program inconsistencies between FPL's nuclearsites.

T. S. Number 6.13

Justifications:1 ~ The current T.S. changed per NUREG-0472, contain the

requirement to have the PCP reviewed by the PNSC priorto implementation.

T. S. Number 6.15

Justifications:l. The NRC safety evaluation for License Amendment 103/97

stated that a specification addressing major changes tothe radwaste treatment systems was not required. Also,the intent of NUREG-0472 recpxirements are met since majorchanges would be reported in the annual FSAR update.

kg

,~5+~i'++'E >

Forwards comments on proof & review version of Tech Specs ...

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