Text

Proposed Tech Specs,Removing Certain cycle-specific Parameter for Relocation to Core Operating Limits Rept
	ML20006A421
Person / Time
Site:	Mcguire, Catawba, McGuire, 05000000
Issue date:	01/17/1990
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20006A418	List: ML20006A417; ML20006A421;
References
	NUDOCS 9001260241
	Download: ML20006A421 (165)
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ATTAClIMENT 1A 4

McGuire' Units 1 and 2 Technical Specifications Changes Requests i'

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INDEX ,

OEFINITIONS SECTION PAGE

1. 0 DEFINITIONS 1.1 ACTI0N........................................................ 1 '1. 2 ' ACTUATION LOGIC TEST.......................................... 1-1 i

1. 3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1

1. 4 AXIAL FLUX 0IFFERENCE......................................... 1-1

1. 5 CHANNEL CALIBRATION........................................... 1-1 i

1. 6 CHANNEL' CHECK................................................. 1-1

1. 7 CONTAINMENT INTEGRITY......................................... 1 ,

1.8 CONTROLLED LEAKAGE............................................ 1-2 1.9 CORE ALTERATION............................................... 1-2 I . l o , c ost optung umn s tronr . . . . . . . . . . . . . . . . . . . . . . . . :-s l 1-2 1 . )116' 00 S E EQ U IVA L E NT I- 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .L. . . .

i i 1 ),f.sf-AVERAGE DISINTEGRATION ENERGY............................... 1-2 l-i L 1.12' ENGINEERED SAFETY FEATURES RESPONSE TIME...................... 1-3 /

E 3 1.JafFREQUENCY IV N0TATION............................................ 1-3 l-

_1.gf'IDENTIFIEDLEAKAGE............................................ 1-3' l 1.g{MASTERRELAYTEST............................................. 1-3 l 1.j{ MEMBER (S)0FTHEPUBLIC....................................... 1-3 (

- 1 lf 0FFSITE DOSE CALCULATION NANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 fa l

it U OPERABILITY........................................ 1 I'

1. )fis OPERA 8LE 1.)z s , O P E RAT I ONA L MOD E - M0 0 E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 l 1.i6 PHYSICS TESTS................................. ............... 1-4 j .i p ><

1.g{ PRESSURE-BOUNDARYLEAKAGE..................................... 1-4 I

[ 1.,32' PROCESS CONTROL PROGRAM ...................................... 1-4 i m l 23 ll (Unit 2)

Amendment No.

McGUIRE - UNITS 1 and 2 I Amendment No. (Unit 1)

M

i 1 3 INDEX *

' DEFINITIONS .

SECTION- PAGE

1. M' PURGE - PURGING................................................ 1-5 I !

ve 1.)( QU AD RANT POWE R T I LT R AT I 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 .l.

t 25 1.,2!IRATED THERMAL P0WER............................................ 1-5 l u

6 1-6 '

1. )n' R E ACTO R B U I LD I NG I NT EG R I TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l . .

1.J7' REACTOR TRIP SYSTEM RESPONSE TIME.............................. 1-5 l a n

1. K REPORTABLE EVENT .............................................. 1-5 Tl r n .

1. 5 SHUTDOWN MARGIN................................................- 1-6 l n

1. 36' S I T E B 0 V N D A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 l 11 1.Jf SLAVE RELAY TEST............................................... 1-6' l <

n 1-6 1 4 50M DIFICATION.................................................

1. x S0uRCE CNECx................................................... 1-6 i w ;

.1.)4~ STAGGERED TEST BASIS.....................-...................... 1-6 l

,s 1.38' THERMAL P0WER..................................................

1-6 l n

1.38' TRIP ACTUATING DEVICE OPERATIONAL TEST......................... 1-7 l 7

17: l 1-7 igUNIDENTIFIEDLEAKAGE.......................................... l 1-7 1.gUNRESTRICTEDAREA.............................................. l SYSTEM........................... 1-7 l 1.J,9' VENTILATION EXHAUST TREATMENT .

1. pr V E NT I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 l bil

'1. M WASTE GAS HOLDUP SYSTEM........................................ 1-7 l Sf t TABLE 1.1, FREQUENCY N0TATION....................................... 1-8 TABLE 1,2, OPERATIONAL M0 DES........................................ 1-9 McGUIRE - UNITS 1 and 2 II Amendment No. (Unit 1)

Amendment No. (Unit 2)

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE l 2.1 SAFETY LIMITS ~

2.1.1 REACTOR C0RE................................................. 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............~.................

. 2 l FIGURE 2.1-1 UNITS 1 and 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS. I IN 0PERATION........................................... 2-2a FIGURE 2.1-2 (BLANK)............................................... 2-3 l l

2.2 LIMITING SAFETY SYSTEM SETTINGS l 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ............... 2-4 l TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS. . .. 2-5 i

BASES SECTION RG,E 2.1 SAFETY LIMITS l

2.1.1 REACTOR C0RE................................................. B 2-1 2.1,2 REACTOR COOLANT SYSTEM PRESSURE.............................. B 2-2 2.2' LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS................ B'2-3 McGUIRE - UNITS 1 and 2 III Amendment No. 88 (Unit 1)

Amendment No. 69 (Unit 2)

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CHANGES THis PAG

[DR INFORMATION ONL[E INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS E SECTION PAGE l 3/4.0 APPLICABILITY..............................................., 3/4 0 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1~ 80 RATION CONTROL Shutdown Margin - T > 200'F............................ 3/4 1-1 gg Shutdown' Margin - T ,g 5 200* F........................... 3/4 1-3 !

Moderator: Temperature Coefficient......................... 3/4 1-4 [

s FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL...... 3/4 1-5a Minimum Temperature for Criticality....................... 3/4 1-6 3/4.1.2 80 RATION SYSTEMS Flow Path - Shutdown...... ............................... 3/4 1-7 f 1

Flow Paths - Operating.................................... 3/4 1-8

' Charging Pump - Shutdown.................................. 3/4 1-9 l l

Charging Pumps - Operating................................ 3/4 1-10 -l Borated Water Source - Shutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-11 Borated Water Sources - Operati ng. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-12 3/4.1.3' MOVABLE CONTROL ASSEMBLIES I i

Group Height... .......................................... 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT l.

OF AN INOPERABLE FULL-LENGTH R00..................... 3/4 1 .1 Position Indication Systems - Operating................... 3/4 1-17.

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Position Indication System - Shutdown..................... 3/4 1-18 l' Rod Drop Time (Units 1 and 2)............................. 3/4 1-19 Shutdown Rod Insertion Limit.............................. 3/4 1-20 McQUIRE - UNITS 1 and 2 IV Amendment No. 60 (Unit 1)

Amendment No. 41(Unit 2) 1: - - . ._ .- - . . . - - . . -

INDEX 4

i LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS 1E.M M Control Rod Insertion Limits.............................. 3/4 1-21 6 ~--25 2.1 : - ; :._. ....i.ii ~.9- 11E "" 'J""" % C L^ T i-E K-E d ' ^ 1 a2

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3/4.2 POWER 0!STRIBUTION l.IMITS I

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3/4.2.1 AXIAL FLUX DIFFERENCE..................................... 3/4 2-1 I l 3/4.2.2 HEAT FLUX H0T CHANNEL FACTOR - F (Z)......................

q 3/4 2-6

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3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL .

FACT 0R.................................................... 3/4 2-14

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3/4.2.4 4 QUADRANT POWER TILT RATI0................................. 3/4 2-19 ' i 3/4.2.5 DN8 PARAMETERS............................................ 3/4 2-22 TABLE 3.2-1 DNS PARAMETERS....................................... 3/4 2-23 l 3/4.3 INSTRUMENTATION j 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................... 3/4 3-1 McGUIRE - UNITS 1 and 2 V Amendment No. (Unit 1)

Amendment No. (Unit 2) m

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INDEX LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.................. 3/4 3-2 ;

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES... 3/4 3-9 TA8LE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS....................................... 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM :

INSTRUMENTATION......................................... 3/4 3-15 i

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.................................... 3/4 3-16 i TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP01NTS...................... 3/4 3-25 ,

TABLE 3.3 5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............ 3/4 3-30 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS. . . . . . . . . . 3/4 3-34 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR P LANT O P E RAT I O N S . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 3/4 3-40 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS......................................... 3/4 3-41 TABLE 4.3.3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS............... 3/4 3-43 Movable Incore Detectors.................................. 3/4 3-45' -

Seismic Instrumentation................................... 3/4 3-46 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION................... 3/4 3-47 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS....................................... 3/4 3-48 Meteorological Instrumentation............................ 3/4 3-49 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............ 3/4 3-50 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................... 3/4 3-51 Remote Shutdown Instrumentation........................... 3/4 3-52 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION........... 3/4 3-53 McGUIRE - UNITS 1 and 2 VI Amendment No. 88 (Unit 1)

Amendment No. 69 (Unit 2) m

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INDEX i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................... 3/4 3-54 Accident Monitoring Instrumentation....................... 3/4 3-55 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-56 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................... 3/4 3-57 ,

Radioactive Liquid Effluent Monitoring Instrumentation.... I 3/4 3-66 ;

TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION.................................... 3/4 3-67 TABLE 4.3 8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-69 i Radioactive Gaseous Effluent Monitoring Instrumentation... 3/4 3-71 TABLE 3,3-13 RADI0 ACTIVE GASEOUS EFFLUENT M!sNITORING

, INSTRUMENTATION.................................... 3/4 3-72 TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-75 Loose-Part Detection System .............................. 3/4 3-78 l 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-79 3/4.4 REACTOR COOLANT SYSTEM i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup snd Power Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 L Hot Shutdown............................................. 3/4 4-3 l Cold Shutdown - Loops Filled............................. 3/4 4-5 l

1 l McGUIRE - UNITS 1 and 2 VII Amendment No. 98 (Unit 1)

Amendment No. 80 (Unit 2)

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SECTION PAGE i Cold Shutdown - Loops Not Ft11ed......................... 3/4 4-6 l 3/4.4.2 SAFETY VALVES I i

Shutdown................................................. 3/4 4-7 ;

Operating................................................ 3/4 4-8 i

3/4.4.3 PRES $URIZER.............................................. 3/4 4-9 o i 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 1

3/4.4.5 STEAM GENERATORS......................................... 3/4 4-11 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE l INSPECTED DURING INSERVICE INSPECTION............. 3/4.,4-16 j i

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION..................... 3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE -

Leakage Detection Systems................................ 3/4 4-18 :

Operational Leakage...................................... 3/4 4-19 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.... 3/4 4-21 P 3/4.4.7 CHEMISTRY................................................ 3/4 4-22 -

TABLE 3.4-2 REACTOP. COOLANT SYSTEM CHEMISTRY LIMITS. . . . . . . . . . . . . 3/4 4-23 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS

SURVEILLANCE REQUIREMENTS......................... 3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY.......................................... 3/4 4-25 l

FIGURE'3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL- '

POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

> 1 pCi/ gram DOSE EQUIVALENT I-131................ 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.....'....................................... 3/4 4-28 e

McGUIRE - UNITS 1 a.nd 2 VIII Amendment No. 32 (Unit 1)

Amendment No.13 (Unit 2)

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IMOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.9 PRESSURL, m mRE LIMITS Reactor Coolant 5ystem.................................... 3/4 4-30 i

FIGURE 3.4-2a UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE ,

UP TO 10 EFPY......................... 3/4 4-31 -

FIGURE 3.4-2b UNIT 2 REACTOR COOLANT SYSTEM

. HEATUP LIMITATIONS-APPLICABLE UP TO 10 EFPY......................... 3/4 4-32 FIGURE 3.4-3a UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 10 EPFY............................ 3/4 4-33 FIGURE 3.4-3b UNIT 2 REACTOR C0OLANT SYSTEM 000LOOWN LIMITATIONS-APPLICA8LE UP i TO 10 EPFY............................ 3/4 4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

j WITH0RAWAL SCHEDULE.................................. 3/4 4-35 Pressurizer............................................... 3/4 4-36 Overpressure Protection Systems........................... 3/4 4-37

j. 3/4.4.10 STRUCTURAL INTEGRITY...................................... .

3/4 4-39 3/4.4.11 REACTOR VESSEL HEAD VENT SYSTEM.,......................... 3/4 4-40 l

l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ,

Cold leg Injection........................................ 3/4 5-1 (Deleted]................................................. 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T,yg >_ 350'F.............................

3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - T,yg 5,350*F.............................

3/4 5-9 3/4.5.4 [0eleted).................................................. 3/4.5-11 3/4.5.5 REFUELING WATER STORAGE TANK............................... 3/4 5-12 McGUIRE - UNITS 1 and 2 IX Amendment No. 82 (Unit 1)

Amendment No. 63 (Unit 2) 1 1

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 1

3/4.6 CONTAINMENT SYSTEMS l 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity..................................... 3/4 6-1 l Containment Leakage....................................... 3/4 6-2 l l 1 Containment Air Locks..................................... 3/4 6-10 1 l

I Internal Pressure......................................... 3/4 6-12 J i

Air Temperature........................................... 3/4 6-13 Containment Vessel Structural Integrity. . . . . . . . . . . . . : . . . . . 3/4 6-14 4

Reactor Building Structural Integrity..................... 3/4 6-15 )

i Annulus Ventilation System................................ 3/4 6-16 Containmer.t Ventilation System........................... 3/4 6-18 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.................................. 3/4 6-20 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. 3/4 6-22 .

l 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors......................................... 3/4 6-31 Electric Hydrogen Recombiners............................. 3/4 6-32 Hydrogen Control Distributed Ignition System.............. 3/* -33 3/4.6.5 ICE CONDENSER Ice Bed................................................... 3/4 6-34 l

l McGUIRE - UNITS 1 and 2 X Amendment No. 94 (Unit 1)

Amendment No. 76 (Unit 2) l

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS .

SECTION PAGE Ice Bed Temperature Monitoring System..................... 3/4 6-36 Ice Condenser Doors....................................... 3/4 6-37 Inlet Door Position Monitoring System..................... 3/4 6-39 Divider Barrier Personnel Access Doors and i Equipment Hatches....................................... 3/4 6-40 l Containment Air Return and Hydrogen Skimmer System........ 3/4 6-41 i Floor Drains.............................................. 3/4 6-42 -

Refueling Canal Drains.................................... 3/4 6-43 Divider Barrier Sea 1...................................... 3/4 6-44 TABLE 3.6-3 DIVIDER BARRIER SEAL ACCEPTABLE PHYSICAL PROPERTIES.... 3/4 6-45 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................ 3/4 7-1

. TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION........................ 3/4 7-2 TABLE 3.7-2 (BLANK)............................................. 3/4 7-2 .

TABLE 3.7-3 STEAM LINE SAFETY VALVES PER L00P................... 3/4 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Specific Activity........................................ 3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.............................. 3/4 7-7 Main Steam Line Isolation Va1ves. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-8 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-9 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-10 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM............................. 3/4 7-11 FIGURE 3/4 7-1 NUCLEAR SERVICE WATER SYSTEM....................... 3/4 7-11a McGUIRE - UNITS 1 and 2 XI Amendment No. 88(Unit 1)

Amendment No. 69(Unit 2)*

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LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.5 STANDBY NUCLEAR SERVICE WATER PON0....................... 3/4 7-12 t 3/4.7.6 CONTROL AREA VENTILATION SYSTEM.......................... 3/4 7-13. !

3/4.7.7 AUXILIARY BUILDING FILTERED VENTILATION EXHAUST SYSTEM... 3/4 7-16 3/4.7.8 SNUBBERS................................................. 3/4 7-18 TABLE-3.7-4a SAFETY-RELATED HYDRAULIC SNUBBERS (UNITS 1 AND 2)....3/4 7-23 :

TABLE 3.7-4b SAFETY-RELATED MECHANICAL SNUBBERS (UNITS 1 AND 2)...3/4 7-26 FIGURE 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST ........... 3/4 7-29 ,

3/4.7.9 SEALED SOURCE CONTAMINATION............... .............. 3/4 7-30 3/4.7.10 Deleted 3/4.7.11 Deleted i

3/4/7.12 AREA TEMPERATURE MONITORING............................... 3/4 7-42 '

TABLE 3.7-6 AREA TEMPERATURE MONITORING.......................... . 3/4 7-43 3/4.7.13 GROUN0 WATER LEVEL......................................... 3/4 7-44 TABLE 3.7-7 AUXILIARY BUILDING GROUNDWATER LEVEL MONITORS.......... 3/4 7-45 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE...................... 3/4 8-8 McGuire - Units 1 and 2 XII Amendment No. 98 (Unit 1)

Amendment No. 80 (Unit 2)

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IMOEX LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1 1

SECTION PAGE TABLE 4.8-2 LOAD SEQUENCING TIMES............................... 3/4 8-9 Shutdown................................................. 3/4 8-10 3/4.8.2 D.C. SOURCES Operating................................................ 3/4 8-11 i

-TABLE 4.8-3 SATTERY SURVEILLANCE REQUIREMENTS................... 3/4 8-14 1 l

Shutdown (Units 1 and 2)................................. 3/4 8-15 )

3/4.8.3 ONSITE POWER O!STRIBUTION SYSTEMS 0perating................................................ 3/4 8-16 Shutdown................................................. 3/4 8-17 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective 0evices...................................... 3/4 8-18 TA8LE 3.8-la UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES............. 3/4 8-20 TA8LE 3.8.lb UNIT 2 CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROJECTIVE DEVICES............. 3/4 8-59 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... 3/49-1 3/4.9.2 INSTRUMENTATION.......................................... 3/4 9-2 3/4.9.3 DECAY TIME............................................... 3/4 9-3 3/4.9.4 CONTAINMENT SUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 C0mVNICATIONS........................................... 3/4 9-6 l 3/4.9.6 MANIPULATOR CRANE........................................ 3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L BUILDING.......... 3/4 9-8 FIGURE 3.9-1 REQUIRED PATH FOR MOVEMENT OF TRUCK CASKS............

3/4 9-9 XIII McGUIRE - UNITS 1 and 2

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

t SECTION PAGE ;

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION !

High Water Leve1......................................... 3/4 9-10 Low Water Leve1......................... ................ 3/4 9-11 1 3/4.9.9 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-12 3/4.9.10 WATER LEVEL - STORAGE P00L............................... 3/4 9-13 I i

3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM................. 3/4 9-14 '

3/4.9.12 FUEL STORAGE - SPENT FUEL P00L........................... 3/4 9-16 :

TA8LE 3.9-1 MINIMUM BURNUP vs. INITIAL ENRICHMENT FOR REGION 2 ST0 RAGE............................................. 3/4 9-17 q 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 REACTOR COOLANT LOOPS ................................... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 '

3/4.11 RADI0 ACTIVE EFFLUENTS 3/4 11.1 -LIQUID EFFLUENTS i Concentration............................................. 3/4 11-1 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-2 0ose...................................................... 3/4 11-5 Liquid Radweste Treatment System.......................... 3/4 11-6 Liquid Holdup Tanks....................................... 3/4 11-7 Chemical Treatment Ponds.................................. 3/4 11-8 3/4.11.2 GASE0US EFFLUENTS Oose Rate................................................. 3/4 11-9 McGUIRE - UNITS 1 and 2 XIV Amendment No. 88 (Unit 1)

. _ _ _ _ _ . _ _ _ __ Amendment No. 69 (Unit 2)_

i N3 CHANSE3 THIS PACE L

FOR INFC.*tMATION ONLYi ,

INDEX LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ,

SECTION PAGE TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-10 Dose - Noble Gases........................................ 3/4 11-13 Dose - Iodine-131 and 133, Tritium, and Radioective [

Materials in Particulate Form............................. 3/4 11-14 '

Gaseous Radwaste Treatment System......................... 3/4 11-15 Explosive Gas Mixture..................................... 3/4 11-16 Gas Storage Tanks......................................... 3/4 11-17 3/4.11.3 SOLID RADIDACTIVE WASTE................................... 3/4 11-18 ,

3/4.11.4 TOTAL 00SE................................................ 3/4 11-20 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 .[

TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES........................... 3/4 12-9 '

TABLE 4.12-1 MAXIMUM VALUES FOR TNE LOWER LIMITS OF DETECTION (LLD).................................... 3/4 12-10 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-13 3/4.12.3 INTERLA80RATORY COMPARISON PR0 GRAM....................... 3/4 12-15 l

XV McGUIRE - UNITS 1 and 2

. . . . . _ _ - .._. ___ __ __ . m

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t NO CHANGES THIS PA .

FOR INFORMATION ON .

INDEX \

BASES .

SECTION PAGE ;

3/4.0 APPLICABILITY................................................ B 3/4 0-1 L

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.......................................... B 3/4 1-2

3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-2 .

3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 t

3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP and ENGINEERED SAFETY FEATURES

ACTUATION SYSTEM INSTRUMENTATION........................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-2 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-5 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-2 L

3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 l' 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-3 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 l

L McGUIRE - UNITS 1 and 2 XVI Amendment No. 88 (Unit 1) l Amendment No. 69 (Unit 2)

f INFO_.1AT o oNLy INDEX

^

BASES SECTION PAGE i

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 ,

3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-7 ,

TA8LE B 3/4.4-1 REACTOR VESSEL TOUGHNESS (UNIT 1).................. B 3/4 4-9 REACTOR VESSEL TOUGHNESS (UNIT 2).................. B 3/4 4-11 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E > 1 MeV) AS A FUNCTION ,

OF EFFECTIVE FULL POWER YEARS................... B 3/4 4-12 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO 550*F TEMPENkIURE..................................... B 3/4.4-13 3/4.4.10 STRUCTURAL INTEGRITY...................................... B 3/4 4-17 3/4.4.11 REACTOR VESSEL HEAD VENT SYSTEM........................... B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 AC C UMU LATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-1 3/4 5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 -

L 3/4.5.4 [ Deleted]................................................. B 3/4 5-2 3/4.5.5- REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 .

3/4.6 CONTAINMENT SYSTEMS

-3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-4 3/4.6.3 CONT AI NMENT I SO LAT I ON V A LV E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 3/4.6.5 ICE CON 0ENSER............................................. B 3/4 6-5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 !

I McGUIRE - UNITS 1 and 2 XVII Amendment No. 76 (Unit 1) l Amendment No. 57 (Unit 2) l l

NO INFORMA FOR CHANGES THIS PAG INDEX BASES SECTION PAGE 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM.............................. B 3/4 7-3 3/4.7.5 STANDBY NUCLEAR SERVICE WATER PON0........................ B 3/4 7-3a !

TABLE B 3/4 7-1 NUCLEAR SERVICE WATER SYSTEM SHARED VALVES......... B 3/4 7-3b 3/4.7.6 CONTROL AREA VENTILATION SYSTEM........................... B 3/4 7-4 3/4.7.7 AUXILIARY BUILDING FILTERED VENTILATION EXHAUST SYSTEM.... B 3/4 7-4 3/4.7.8 SNUBBERS.................................................. B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.10 Deleted 3/4.7.11 Deleted 1 1

3/4.7.12 AREA TEMPERATURE MONITORING............................... B 3/4 7 6 .i 3/4.7.13 GROUN0 WATER LEVEL......................................... B 3/4 7-6 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND j ONSITE POWER DISTRIBUTION SYSTEMS........................ B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1 3/4.9.5 C OM4UN I C A T I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.6 MANIPULATOR CRANE......................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING........... B 3/4 9-2 McGUIRE - UNITS 1 and 2 XVIII Amendment No.98 (Unit 1)

Amendment No.80 (Unit 2)

, , , - , - . , . . - - - , --m,- , . . . . - . - .,.-.-.e .,

-+ , -+ .. -- as .-a .- ~ , -- . . - - _ . ~ , , . . . - -

l l

O PAQt, INFO gg ONLy i i

I INDEX BASES SECTION PAGE !

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 9-2 '

3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and STORAGE P00L............................................. B 3/4 9-3 3/4.9.11 FUEL HANDLING VENTI LATION EXHAUST SYSTEM. . . . . . . . . . . . . . . . . . B 3/4 9-3 ;

3/4.9.12 FUEL STORAGE - SPENT FUEL STORAGE P00L. . . . . . . . . . . . . . . . . . . . b 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS l

3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 ,

3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS <

3/4.11.1 LIQUID EFFLUENTS......................................... B 3/4 11-1 I

3/4.11.2 GASEOUS EFFLUENTS........................................ B 3/4 11-3 l

3/4.11.3 SOLIO RADI0 ACTIVE WASTE.................................. B 3/4 11-7 3/4.11.4 TOTAL 00SE............................................... B 3/4 11-7 i

t 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l

3/4.12.1 MONITORING PR0 GRAM....................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS.......................................... B 3/4 12-2 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... B 3/4 12-2 l-1 McGUIRE - UNITS 1 and 2 XIX Amendment No. 88(Unit 1).

Amendment No. 69(Unit 2)

~

R INroffANO' toy>A e ,ce, Y

t o

INDEX DESIGN FEATURES SECTION PAGE 5.1 SI T,E, 5.1.1 EXCLUSION AREA............................................... 5-1 5.1. 2 LOW POPULATION 20NE.......................................... 5-1 5.1. 3 MAP DEFINING UNRESTRICTED AREAS FOR RADI0 ACTIVE GASEQUS AND LIQUID EFFLUENTS............................... 5-1 FIGURE 5.1-1 EXCLUSION AREA....................................... 5-2 FIGURE 5.1-2 LOW POPULATION 20NE.................................. 5-3 FIGURE 5.1-3 SITE BOUNDARY FOR GASEOUS EFFLUENTS.................. 5-4 FIGURE 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS................... 5-5

5. 2 CONTAINMENT s 5.2.1 CONFIGURATION................................................ 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE.............................. 5-6

5. 3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.............................................. 5-6 5.3.2 CONTROL R00 ASSEMBLIES....................................... 5-6 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................. 5-7 5.4.2 V0LUME....................................................... 5-7 5.5 METEOROLOGICAL TOWER LOCATION.................................. 5-7 5.6 FUEL STORAGE 5.6.1 CRITICALITY.................................................. 5-7 5.6.2 DRAINAGE..................................................... 5-7 j i

5.6.3 CAPACITY..................................................... 5-7 l- 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT............................ 5- 7a TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS................... 5-8 McGUIRE - UNITS 1 and 2 XX Amendment No.88 (Unit 1)

Amendment v

No.69 (Unit 2)

F l

' NO CHANGES THIS PAoE .

FOR INFORMATION ON1 Y .

4 INDEX

ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY................................................. 6-1 6.2 ORGANIZATION 6.2.1 0FFSITE...................................................... 6-1 6.2.2 UNIT STAFF................................................... 6-1 FIGURE 6.2-1 (Deleted)............................................ 6-3 FIGURE 6.2-2 (De1etad)...........................;................ 6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION....................... 6-5 6.2.3 STATION SAFETY REVIEW GROUP (SSRG)

Function.................................................. 6-7 Composition............................................... 6-7 Responsibilities........................................... 6-7

. Authority................................................. 6-7 Records................................................... 6-7 6.2.4 SHIFT TECHNICAL ADVIS0R...................................... 6-7 6.3 UNIT STAFF QUALIFICATIONS...................................... 6-7 6.4 TRAINING....................................................... 6-7

6. 5 REVIEW AND AUDIT 6.5.1 TECHNICAL REVIEW AND CONTROL Activities................................................ 6-8 McGUIRE - UNITS 1 and 2 XXI Amendment No. 52 (Unit 1)

Amendment No. 33 (Unit 2)

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1 i

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INDEX ;

ADMINISTRATIVE' CONTROLS !

SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) :

Function.................................................. 6-9 [

t Organization.............................................. 6-10 l Review.................................................... 6-11 Audits.................................................... 6-11 Authority................................................. 6-12 Records................................................... 6-13 :

6.6 REPORTABLE EVENT ACTI0N........................................ 6-13 6.7 SAFETY LIMIT VIOLATION......................................... 6-13 I -

6.8 PROCEDURES AND PR0 GRAMS.............<.......... .............., 6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REP 0RTS.............................................. 6-16 Startup Report............................................ 6-16 Annual Reports............................................ 6-17 Annual Radiological Environmental Operating Report........ 6-18 Semiannual Radioactive Effluent Release Report............ 6-18 -

Monthly Operating Reports................................. 6-20 CORE OPtunr+ -

P;;ki ng F ac t^ ? Limi ts Report. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 .[l l

l -' i 1

J L ;

l l

l l' I l

l McGUIRE UNITS 1 and 2 XXII Amendment No. V (Unit 1) )

Amendment No. A (Unit 2) 1

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t ADMINISTRATIVE CONTROLS SECTION PAGE REPORTING REQUIREMENTS (Continued) 6.9.2 SPECIAL REP 0RTS.............................................. 6-21 ;

6.10 RECORD RETENTION.............................................. 6-22 6.11 RADIATION PROTECTION PR0 GRAM.................................. 6-23 6.12 HIGH RADIATION AREA.............................:............. 6-23 6.13 PROCESS CONTROL PROGRAM (PCP)................................. 6-24 ,,

6.14 0FFSITE DOSE CALCULATION MANUAL (00CM)........................ 6-25 ,

6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS, AND SOLIO W

,ASTE TREATMENT............................................. 6-26 P

4 McGUIRE - UNITS 1 and 2 XXIII Amendment No. 32 (Unit 1)

Amendment No.13 (Unit 2)

. - - - m.

1. 0 DEFINITIONS OL W A The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

Ella .

1.1 ACTION shall be that part of a Technical Specification which prescribes remedial asasures required under designated conditions.

gTUATIONLOGICTEST

1. 2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each gossible interlock logic state and verification of the required lo2ic output, she /.CTUATION LOGIC TEST shall include a continuity check, as a minimum, of output hvices.

ANAL.0GCHANNELOPERATIONALTES,1 1.3 An ANALOG CHANNEL OPERATIONAL MST shall be the in1ection of a simulated signal into the channel as close to tne sensor as practicable to verify OPERABILITY of alam, interlock andhr trip functions. The ANALOG CHANNEL of the alare inter-OPERATIONALTESTshallincludeadjustments,asnecessaryIthintherequIred lock and/or Trip Setpoints such that theTetpoints are w '

range and ac. curacy. -

AXIAL FLUX DIFFERENCE

1. 4 AXIAL FLUX DIFTERENCE shall be'the difference in normalized flux signals between 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 the channel such that it respcnds with the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alare, interlock and/or trip functions and may be !

performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible,' comparison of the channel indication and/or status with other i indications and/or status derived from independent instrument channels measuring the same parasater.

d McGUIRE.- UNITS 1 and 2 1-1 9

e s

]

DEFINITIONS-l CONTAIMIENT INTEGRITY 1

1.7 CONTA!WIENT INTEGRITY shall exist when: I l

a. All penetrations required to be closed during accident conditions )

are either:

1) Capable of being closed by an. OPERABLE containment automatic ,

i isolation valve system, or operator action during periods when L containment isolation valves may be opened under administrative a ,

controls pursuant to Specification 4.6.1.1.a; or i

2) Closed by manual valves, blind flan es, or deactivated i automatic valves secured in their c osed positions,

b. All equipment hatches are closed and sealed, ;

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

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

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

1 . .

CONTROLLED LEAKAGE l 1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied t 'o the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE' ALTERATION shall be the movement or manipulation of any component j

within the reactor pressure vessel with the. vessel head removed and fuel in

the vessel. Suspension of CORE ALTERATION shall not preclude completion of l- movement of a component to a safe conservative position.

Quirn@)

l 005E EQUIVALENT I-131

~

L 1.)#u DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) l which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid '

dose conversion factors used for this calculation shall be those listed in Jable III of TID-14844, " Calculation of Distance Factors for Pc,wer and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY

1. n l shall be the average (weighted in proportion to the concentration of l each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

1 McGUIRE - UNITS 1 and 2 1-2 Amendment No. (Unit 2)

Amendment No. (Unit 1) w

l N SERT @

l-

~ CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

i-1 These. cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is addressed in individual specifications.

(-

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i

i OEFINITIONS )

ENGINEERED SAFETY FEATURES RESPONSE TINE l i

1. h The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interv from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety ,

function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel ;

generator starting and sequence loading delays where appitcable. ,

c FREQUENCY NOTATION ,

-1.h The FREQUENCY NOTATION specified for the performance of Surveillance l Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.hIDENTIFIEDLEAKAGEshallbe: l

a. Leakage (except CONTROLLED LEAKAGE) into closed systees, such as pump seal or valve packing leaks that are captured and conducted to '

a sump or collecting tank, or _

b.- Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the '

operation of leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE, or 9

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System. .

MASTER RELAY TEST Ib

1. X A MASTER RELAY TEST shall be the energization of each master relay and l verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MENBER(S) 0F THE PUBLIC ,

1. M MEMBER ($) 0F THE PUBLIC shall include all persons who are not l occupationally associated with the plant. This category does not include

- employees of the licensee, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make L

l- deliveries. This category does include persons who use portions of the !

site for recreational, occupational, or other purposes not associated with the plant.

l

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McGUIRE - UNITS 1 and 2 1-3 Amerur No. @~ir I) h me u,, (6,r > > q

- . _ . _ _ _ _ _ _ _ _ _ _ _ _ ._ _ _ _ _ _ _ _ _ _ _ . _ ._. _ _ v

4 F

L 1

0 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (0DCM)

I l

is l' F The 0FFSITE DOSE CALCULATION MANUAL shall contain the methodology and

. l parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

OPERABLE - OPERASILITY

1. h A system, subsystem, train, component or device shall be OPERA 8LE or l l have OPERASILITY when it is capable of performing its specified function (s), l and when all necessary attendant instr a ntation, controls, electrical power, J cooling or seal water, lubrication or other auxiliary equipment that are ]

required for the system, subsystem, train, component, or device to perform its j function (s) are also capable of performing their related support function (s). ,

OPERATIONAL MODE - MODE 1I[TAnOPERATIONALMODE(i.e., MODE)shallcorrespondtoanyoneinclusive l combination of core reactivity condition, power level, and average reactor coolent temperature specified in Table 1 2.

PHYSICS TESTS _

1. h PHYSICS TESTS shall be those tests performed to measure the fundamental f nuclear characteristics of the core and related instrumentation: (1) described l in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR l 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1..N PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l 1eakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESSCONTROLPROGRAM(PCP1 u

1.J,0 The PROCESS CONTROL PROGRAM shall contain the provisions to assure that l the SOLIDIFICATION of wet radioactive wastes results in a waste form with properties that meet the requirements of 10 CFR Part 61 and of low level radioactive waste disposal sites. The PCP shall identify process parameters influencing SOLIDIFICATION such as pH, oil content, H content, ratio of solidification agent to waste and/or,0 content, additives necessary solids for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each waste type, based on laboratory scale and full scale testing or experience. The PCP shall also include an identification of conditions that must be satisfied, based on full scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of low level radioactive waste disposal' sites.

McGUIRE - UNITS 1 and 2 1-4 Arge m %, [4m ;)

A n t~rmm t% %,r q ,

- _ _ _ - . . _ _ -_ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . . _ _ . . _ _ ~ . _. v

DEFINITIONS !

PURGE - PURGING m .

i 1.J3' PURGE or PURGING shall be the controlled process of discharging air or { .

gas from a confinement to maintain temperature, pressure, humidity, concentra- !

tion or other operating condition, in such a manner that replacement air or ,

gas is required to purify the confinement. ;

QUADRANT POWER TILT RATIO !

W ,

1.7# QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore j detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated ,

output to the average of the lower excore detector calibrated outputs, '

whichever is greater. With one excore detector inoperable, the remaining .

three detectors shall be used for computing the average. ,

RATED THERMAL POWER u l 1.JY RATED THERMAL POWER shall be a total core heat transfer rate to the l ,

reactor coolant of 3411 MWt. j f

REACTOR BUILDING INTEGRITY 1.k REACTOR BUILDING INTEGRITY shall exist when:

. l

a. Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at ;

least one door shall be closed. l

b. The Annulus Ventilation System is in compliance with the requirements of Specification 3.6.1.8, and I

c. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE. l REACTOR TRIP SYSTEM RESPONSE TIME

1. [ The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time intervall from l when the monitored parameter exceeds its Trip Setpoint at the channel sensor i until loss of stationary gripper coil voltage.

1 REPORTABLE EVENT p )

ve : \

1.36' A REPORTABLE EVENT shall be any of those conditions specified in yl l

.Section 50.73 to 10 CFR Part 50. L l l

l L McGUIRE - UNITS 1 and 2 1-5 Amendment No.V (Unit 1)

Amendment No.y (Unit 2)

I

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DEFINITIONS

[

$ HUT 00WN S RGIN

1. k SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which 1 the reactor is subcritical or would be subcritical from its present condition j assuming all full-length rod cluster assemblies (shutdown and control) are T <

fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE SOUNDARY !

n 1.M The SITE SOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the Itcensee. l {

SLAVE RELAY TEST . '

lt

1. T A SLAVE RELAY TEST shall be the energization of each slave relay and j l verification of OPERASILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION n

1. R" SOLIDIFICATION shall be the immobilization of wet radioactive wastes l such as evaporator bottoms, spent resins, sludges, and reverse osmosis g concentrates as a result of a process of thoroughly mixing the waste type with 1 a SOLIDIFICATION agent (s) to form a free standing monolith with chemical and physical characteristics specified in the PROCESS CONTROL PROGRAM (PCP).

SOURCE CHECK 1.N'ASOURCECHECKshallbethequalitativeassessmentofchannelresponse l

when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS

1. h A STAGGERED TEST BASIS shall consist of: l

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

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

THERMAL POWER u

1. g THERMAL POWER shall be the total core heat transfer rate to the reactor j coolant.

Amendment No. (Unit 2)

McGUIRE - UNITS 1 and 2 1-6 Amendment No. (Unit 1)

DEFINITIONS

\ , TRIP ACTUATING DEVICE OPERATIONAL TEST

1. k A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operatin l

,' . Trip Actuating Device and verifying OPERABILITY of alare, or interlock an N trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include as necessary, of the Trip Actuating Device such that it actuates adjustment, at the requ ired Setpoint within the required accuracy, i UNIDENTIFIED LEAKAGE

1. N UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED l LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA v l l 1.)8' An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY J access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any l areawithintheSITEBOUNDARYusedforresidentialquartersorforIndustrial, l commercial, institutional .and/or recreational purposes. l 1

~

j VENTILATION EXHAUST TREATMENT SYSTEM ,

w. !

1.)9' A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or l

particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas i effluents. Engineered Safety Feature (ESF) Atmospheric Cleanup Systems are '

l not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING we 1..W VENTING shall be the controlled process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

WASTE GAS HOLDUP SYSTEM J

1. k A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to l reduce radioactive gaseous effluents by collecting Reactor Coolant System l

offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the ;

l l environment.

l l

McGUIRE - UNITS 1 and 2 1-7 A r eg,mm 9,, (m,,, O l Aetn m e No. (w 1) l-l

i REACTIVITY CONTROL SYSTEMS .

MODERATOR TEDERATURE COEFFICIENT .

i LIMITING CS STTION FOR OPERATION e !

3.1.1.3 The yl's g'u 'En'yrage [*g*gry co4{ficien"t a **

" '# *fMTC}' ' * *"shall

~ "

bet

*

ww'm " " "6'a,n" i' 4 t r ,xit. : tra.. th: lia.iti shown in Fi C

2. L;;e n; ;the tier,1.1 10 ' :!t: t/t/gure 3.1-0. .end- *

- i '

0L'

" T fe tt; ;il 7;;; witt4... ,I '

-d Ofns ;;fr;p;-um;ede;odh lif q,AT:0 T";""//. 70Z" ;;;;it'* - ;

APPLICAl!LITY: 3 pes * -- * " X I.S. M ? - MODES 1 and 2* only.# h rM ' er rffi::ti:

c v6:t 4.,n n (0.1.1.3,~.,-

en.) ,. ,r MODES 1, 2, and 3 only.# ,

ACTION:

[ l g gpree, <=r ceca

,,,,,y,,,

y; ' i

a. With the MTC incre positive than the limitu ';xift::ti::,3.1.1 ' . <

e n ,. operation in MODES 1 and 2 may proceed provided:

l 1. Control rod withdrawal limits are established and maintained '

l sufficient to restore the MTC to less positive than the MnMe '

am 6,-,rup,,fl$/Eca.a thr- in fi;u-- 3.1-0 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in H0T STAN08Y within I a l the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition d L to the insertion limits of Specification 3.1.3.6; *

2. The control rods are maintained within the withdrawal limits ik l established above until a subsequent calculation verifies'that i i the MTC has been restored to within its limit for the all rods 2 :

withdrawn condition; and I d

t 3. A Special Report is prepared and submitted to the Commission 1 pursuant to Specification 6.9.2 within 10 days, describing the J value cf the measured MTC, the interim control rod withdrawal t limits, and the predicted average core burnup necessary for ,

l l

restoring the positive MTC to within its limit for the all rods withdrawn condition.

h{s 1

ros With the MTC more negative than the" limit 01 !;nifinth;,3.1.1.M.

utworo w rut coLn ) '

i

b. 9' abous,.be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K,77 greater than or equal to 1.0.

See Special Test Exception 3.10.3.

l l

1 McGUIRE - UNITS 1 and 2 3/4 1-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

. _. ..._.. _ _ ._ _ . . . _ . _ _ _ _ . _ _ ___.u _...a _t._ ..._ _ __..___.

.- 1 l yu R_EAgIVITYCONTROLSYSTEMS' y;-

o-

Y SURVEILLANCE REQUIREMENTS y.

4.1.1. 3 The MTC shall be determined to be within its limits during each fuel

,. cycle as follows:: ,

' \ '-

The hfC .shall be measured and compared to the IfUlfif BOL 0limit IW Tkt CDsR

- a.- - %+

. "S;:fi'icetten 3.1.1.3:., 2:;; prior to initial-operation above 5% }

of RATED THERMAL POWER, after each fuel loading; and ;

we ve rra s nnousuce su~r urcmer ow we cosa -

m

b. 'The MTC shall be measured at any THERMAL POWER and compared tv l

.-0.2 w i0 ^ delte U;v'"F (all rods withdrawn, RATED THERMAL POWER '

condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppe. In the event this comparison indicates the MTC is more_ negative than -0.2 o 10 ' d;1t: L'L'*F, the MTC shall be remeasured, and comp red to the EOL MTC limit ;f 4;;ifice-

- >ncem, is M cotafti:r. 3.1.1.35.., at least once per 14 EFPD during the remainder of the {j fuel cycle, nr so* ren sunviewcq s.im,r sucmt* ow w cow, .

l 1

L .

l L

L l

l>

Il p

1 McGUIRE.- UNITS 1 and 2 3/4 1-5 Amendment No (Unit 1)

Amendment No. (Unit 2)

., i$.

_ - ... . - . - . . _ . - . . _ _ .. . 2 2 .. ~ -. - .. . . . . .

. . . - .. - - . - -.- -- - - - - - ' - - -- - - ^ ~ ~ ~ - ' ' -

l 1

1 O

Y n

1.0'-

7* - 0. 9 ====- ,

=

S .

! 0.8 - 1 j

z : - ,

E 0.7 )

O

\

Seceptable Unaccostante

$. 0.6 Operation Coeration 3 0.5 -

0.4 -

, p 3

.$ O.3-

-i

e. -

0.2 .

l

' O.1 -

g' -

6 0- 10 20 30 40 50 . 60 70 80 - 90 100 l

% of Rated Thermal Power FIGURE 3.1-0 MODERATOR TEMPERATURE COEITICIENT VS POWER LEVEL t

r l

McGUIRE - UNITS 1 and 2 3/4 1-Sa Amendment No. 60 (Unit 1)

. .. Amendment No. 41 (Unit 2)

.v. '

__._____/L_.__n______________

1 l

l REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVA8LE CONTROL ASSEMBLIES GROUP HEIGHT i

LIMITING CONDITION FOR OPERATION

]

3.1.3.1 All full-length shutdown and control rods'shall be OPERABLE and I positioned within i 12 steps (indicated position) of their group step counter 'l 1 demand position.

i APPLICABILITY: MODES 1* and 2*.

ACTION:

a. . With one or more full-length rods inoperable due to being immovable !

as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTOOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With more than one full-length rod misaligned from the group step.

counter demand position by more than i 12 steps (indicated position),

{

be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one full-length rod trippable but inoperable due to causes ,

other than addressed by ACTION a,, above, or misaligned from its-group step counter demand height by more than i 12 steps (indicated !

L position), POWER OPERATION may continue provided that within 1-hour:

l- 1. The rod is restored to OPERABLE status within the above alignment requirements, or g g ,p, w 3, g

2. The rod is declared inoperab)e and the remainder of the rods in the group with the inoperabTe rod are aligned to within i 12 steps of the inoperable red while maintaining the rod sequence o-and insertion limits of" 'g = 3.1-1. The THERMAL POWER level yl~

L i

shall be restricted pursuant to Specification 3.1.3.6 during.

subsequent operation, or

3. The rod'is declared inoperable.and the SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may.then continue provided that; a) A' reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

)

^5ee Special Test Exceptions 3.10.2 and 3.10.3.

McGUIRE - UNITS 1 and 2 3/4 1-14 Amendment No (Unit 1)

Amendment No (Unit 2)

L

- -- - -__________n . _____ _ _ ___ _ ____ _ _ _ _ _

4 t

l REACTIVITY CONTROL-SYSTEMS-

- ACTION .(Continued) _

c) A power distribution map is obtained from the movable incore detectors and F9(Z) and F H are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and 1

d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour' ]

and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron-Flux. l Trip Setpoint is reduced to less than or equal-to 85% l I

of RATED THERMAL POWER.

r 1

d. With more than one full-length rod trippable but inoperable due to i i J

causes other than addressed by ACTION a above, POWER OPERATION may l continut provided.that: .,

p sruumr n v.ts.

o ,'

1. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the rods in the bank (s) with +

O; the. inoperable rods a aligned to within i 12 steps of the ;

I inoperable rods whil maintaining-the rod sequence and ,

l '.

l.

insertion limits of m i;ute 3.1-1. The THERMAL POWER level {

- shall be restricted pursuant to. Specification 3.1.3.6 during: -

subsequent operation, and 5

2. The inoperable rods are restored to OPERABLE status within'72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

[9 a

- SURVEILLANCE REQUIREMENTS , 4.1.3.1.1 The position of each-full-length rod shall be determined to be 1

within the group demand limit by verifying the individual rod positions-at least once per 12' hours except during time intervals when the Rod Position.

Deviation Monitor is inoperable, then verify the group positions at least once l, per 4-hours.

L 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be-

. determined to be 0PERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

. t

\

McGUIRE .- UNITS 1 and 2 3/4 1-15 Amendment No. (Unit 1)

Amendment No. (Unit 2) -

--,w-+ --------------A-A------------

NO ggtE 3 I" FOR INFORMAT E,

ONLY ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D ,

Rod Cluster Control ~ Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment loss of Reactor Coolant.from Small Ruptured Pipes or from Cracks in '

Large-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 Coolant Accident)

Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

M i

a McGUIRE - UNITS 1 and 2 3/4 1-16

, - - e , , - - < ~~a . - ,- 0 se e --n-, - -- -_- - - - -

' REACTIVITY CONTROL SYSTEMS

SHUTDOWN R00 IN_SERTION LIMIT ,

LIMITING CONDITION FOR OPERATION i

3.1.3.5 All shutdown rods shall be f - E -R==:m m.rra #- r ms% ~sents.~ 4s senorser m we costorturn uimors gemar(coun).

APPLICA8ILITY: MODES-1* and 2*#.

~

ACTION:

. ruscarte nerm not ruster,aa s,nor $rnorier ou rut coLR, l

.With a maximum of one= shutdown rod d Q tj i, except for. surveillance V

l testing pursuant to Specification 4.1.3.1.2, within I hour either: "

scsroer '

a. duthpwithdraw'the rod,cor rv woww m insenrion Lowr spuinro sa wr coLR,og )

b. Declare the rod to be inoperable and apply Specification 3.1.3.1. .l 9

Y SURVEILLANCE REQUIREMENTS i vs W'M NE 'H$rnTIon sansr stictrato tH Nt COLRs 4.1.3.5' Each shutdown rod shall be determined to be%y'-i _ ==;: l l

[ -

1 1;'

a. Within 15 minutes prior to withdrawal <of any rods in Control H Banks A, B,:C or D during an-approach to reactor criticality, and- l

b. At least'once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. j 4

1: i o

"See Special Test Exceptions 3.10.2 and 3.10.3.

With'K,ff greater than or equal to 1.0. l i

McGUIRE - UNITS 1 and 2 3/4 1-20 A mt~emm vo. ( uu,r ,)

b e~* m ur No. tun.,1)

.u

REACTIVITY CONTROL SYSTEMS CONTROL R00 INSERTION LIMITS .

LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical _ insertion as ebaum:4m Q= 1 ; T $rrcosato w us coat ortsariae teners streariceun). (l l APPLICA8ILITY:~ MODES 1* and 2*#.

ACTION:

sercome w we eora iD With the control banks inserted beyond the 4beve-insertion limits'; except for g ,

surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the "-- ^^; -- , -:= i-senn.s cs~rs rmer,r, ,, wr ceca, oa j

c. Be in at least H0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

L "See Special Test Exceptions 3.10.2 and 3.10.3. '

j. #With K,ff greater than or equal to 1.0.

u E

.McGUIRE - UNITS 1 and 2 3/4 1-21 Amendment No. (Unit 1)

Amendment No. (Unit 2)

- - . . . . . . , . . - . - - . - - , . . . - . - . - . .1 '

a-

.t

(, ;

Que1EFw ,

2 (Fullywithdrawn) r 28 --

1 "

0 __ fi,(295,228) _- ,_(79%,22%j- 1,

/ -

j <

200 Y ^ n'~ . _!

i

' .' u_ .f--.f._

BANK B f' , ,

180

l' ' .-

I 160 (05, 162) . _

~

b 1 g

- (loot, 'I'6t..d_

i

= 140 l -

i f.

g t

f BANK.C _- l

~

5-120

~ '

~

U '

8: .

.~

.~ -

' h, . ,

2a 100

~

' ~

_- -(

8. <

L ,f '.

~

.- BANK D 9 ,

i f - - - - -

}

8.0 2 .-

_~ .

)

[. 60

-~ -

- d , i q

((05,47)

- ~-

~~

I ;

-40 -

l

~

? ,

1 t,

L 20 -

./ . 1 ,

'(30%,0) 0

~

-f L

0 20 40 '

60 80 100 (Fullyins ted) RelativePower(Percent) '

f, k

\ <,

5 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VS RELATIVE POWER l

McGUIRE - UNITS 1 and 2 3/4 1-22 Amendment No. (Unit 1) )

L. . . . . . - . . . - . . - . . . - - - - - - - - - - - -- - - - - - - - - -

Amendment K No. - - - -

(Unit 2) i

4 4

3/4.2 POWER DISTRIBUTION LIMITS -

3/4.2.1= AXIAL FLUX' DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION-3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

as spoororo w rue cose cresarm umors strosrteow s

a.- the allowed operational space #"- " '

- -;- L. -- 3.24 for RAOC operation, l or TM 5ftarsteov wtCoLR

b. within M p - 2* target band about the target flux difference during base

{l load operation.

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *.

ACTION:

a. For RAOC operation with the indicated AFD outside of the I's+ - 1 2-1 limits, srecorers v vr ca a,

1. Either restore the indicated AFD to within the %... G.2=1 cosa -)'

limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWERr ,

'within-30 minutes and reduce the Power Range Neutron Flux -

High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. For Base Load operation above APLND** with the indicated AXIAL FLUX-DIFFERENCE outside of the applicable target band about the target flux. difference:

cosa sprceroro

1. Either restore the indicated AFD to within the# target band l limits-within 15 minutes, or

2. Reduce-THERMAL POWER to less than APLND'of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes,

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL !

POWER unless the indicated AFD is within the t's -- 2.2-1 limits JP(corof, x sw we coon. . .l 1

l l

See Special Test Exception 3.10.2. l ND > " "'" ' " "'

- APL is the minimum allowable" power level for base load operation and wM4#s fatorer, i te F vvided in the p a :g F-- -

-!_^y -1 per Specification 6.9.1.9.

CORE oprpason conor) Agrou l

McGUIRE - UNITS 1 and 2 3/4 M Amendment No. Unit 1) .

l Amendment No. (Unit 2)

p j FOR supogyg% %, 7

. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS-

.4.2.1.1 The indicated AFD shall be determined to be within its limits-during

- POWER OPERATION above 50% of RATED THERMAL POWER by:

Monit6 ting the indicated AFD for each OPERABLE excore channel:

-a.

-1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring !

the AFD Monitoring Alarm to OPERABLE status.

- b. Monitoring and logging the indicated AFD.for each OPERABLE excore

. channel at least once per hour for the first 24. hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed

- to exist duri'ng the interval preceding each logging. ;

4.2.1.2 The indicated AFD shall-be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the G limits. -

4.2.1.3' When in Base Load operation, the target axial flux difforence of-

.g' -

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 i

. not applicable. ,,

4.2.1.4 When in Base Load operation, the target flux difference shall be

- updated at least' once per 31 Effective Full Power Days by either. determining the target flux difference in conjunction with the surveillance requirements '

of Specification 3/4.2.2 or by linear interpolatidn.between the most recently measured value and the calculated value at the end of cycle life. The provisions a of Specification 4.0.4 are not applicable. ,

1.

4 y

McGUIRE - UNITS 1 and 2 3/4 2-la Amendment NoA2 (Unit 1) ;

Amendment No.23 (Unit 2) ,

[ ;

.l OTHIS PAgg

)%POR INF MATioN ONLy i

t 1

P k

This page deleted.

l i

r 4

McGUIRE - UNITS 1 and 2 3/4 2-2 Amendment No A2 (Unit 1)

Amendment No23 (Unit 2)

1 Q..

q _ p/p

& L -

~. -

])

/

- : N ,,e L ,

f 2 '

l {

/

( + 30.1# 1. I ' '"

,= --

/ .I j r / !

\

(==an=a I \ .uandu . !; )

b f

\ a".

\ l t

3 x / =="="

7'%

l / l 10 i .

/ \

1 / \ '

T, I

es \ /

r

.S j w 'l A i o so --

/

K (St.MI i

( = 35. M )

/

u \ ,

t 3

j

, /\

ii se -

f \

/ '\ i i ac -

/ \

( / \

50 '

i / N

.N h

/ '

~

,, . .se -no -to * ,o .. a a. i l Axial Flux Difference (1 De a-1) ,,

i 1

l Figure 3.2-1 f '

?,

, Limits as a Function of Rated Thermal Power - - ~

f , )

___.._.- )

I 1

McGUIRE UNITS 1 and 2 3/4 2-3 Amendment No. (Unit 1)

Amendment No. (Unit 2)

,. : - - ..-. .-._._.___~_______----_'___- .

\

J ,

,, g-

'** INP PAag, ;

47

~

ONLY l

1:

l 7 .t t.

I'

.t

.c i

4'4..-

o ,

.r This page deleted. ..

4

') ? -

McGUIRE - UNITS 1 and 2 3/4 2-4 Amendment No.73 (Unit 1)

-p Amendment No.54 (Unit 2)-

csAN0cs 73'-

hR'N!'ORung*0.qy P A G e.

-j l ~

l J

l f --

. ?

5 s

i N

t 4 ,

-(

This page deleted.

i i

.-1

( f.

o ,

l-l - -

u - McGUIRE - UNIT 5 1 and 2 3/4 2-5 Amendment No.42(Unit 1) l-c i

Amendment No.23(Unit 2) ,

1

p. s

,--e, . . , - -

., -- .r

y .- -- -

L i

. POWER DISTRIBUTION. LIMITS l

3/4.2.2-HEATFLUXHOTCHANNELFACTOR-Fg L .

LIMITING CONDITION FOR OPERATION

[

4

~

3.2.2' F (Z) shal'l be limited by the following relationship:

9 1 ll" Fq (Z) 5 4 1 Z ayK(Z)} for P > 0.5 f, ' b P

l gj,,

Fq (Z) 1M p >{K(Z)} for P $ 0.5 g -l 1

wkrat s Fl9: NE Fe so~or ^1 MM1s91HERMAL fewfR(Rrr) sticofofo w Tut Cort ointATw&

p _ THERMAL POWER '* ors Rrrear (cotm4 i

RATED THERMAL POWER ' -

warm,n uo F - :^? ^^FoxAVovrW tet& K(Z)*ser the"f sctie:ettem =ieed ';;= rigs + t M 'or a giur, core height Wetiert- 3rrc ror, sv 7we co's.

APPLICABILITY: MODE 1.

ACTION:

With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% n F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ;. subsequent POWER "

OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K 4) have been reduced at least 1% (in aT span) for each 1% F (Z) exceeds the limit; and 9

b. Identify and correct the cause of the out-of-limit condition prior

, to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided Fq (Z)-

is demonstrated through incore mapping to be within its limit.

McGUIRE - UNITS 1 and 2 3/4 2-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

m a POWER DISTRIBUTION' LIMITS-SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 For RAOC operation,9F (z) shall be evaluated to determine if F9 (z) i

-is within its limit by:-

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.'

b. Increasing the measured F q (z) component of the power distribution map by 3% to account fer manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. !

Verify the requirements of Specification 3.,2.2 are satisfied.

c. Satisfyingthe,fg}1owingrelationship:

F M

x K(z) for P > 0.5 kb 0 (7) <Pik$$ x W(z) i F

M O(7)-[b:

St x K(z) for P < 0.5 kI; W(z) x 0.5 g mc ~maco f.,m m + rwrw or cose keer, ,

where F q (z) is the m suredF(z)increasedbytheallowangsfor 9

manufacturing toler nces and measurement uncertainty,4=44 is the i !

F limit, K(z) is 4' r '- G ,_.-- E -2. P is the relative THERMAL q

j POWER, and W(z) is the cycle dependent function that accounts for i power distribution transients encountered during normal operation. -

. . .; . _ _ ' r.; - _g . ..; : =-- d- ^ _; _ r_ s,per Speci-

'!Tkto, A v wmane strove ou me cet orenerw cmen ser.ar l!

fication 6.9.1.9.

d.

N Measuring Fq (z) according to the following schedule:

j

1. Upon achieving equilibrium conditions after exceeding by :

10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or 9

2. At least once per 31 Effective Full Power Days, whichever occurs first. ,

During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

McGUIRE - UNITS 1 and 2 3/4 2-7 Amendment No. (Unit 1)

Amendment No. (Unit 2) i

. x x -- -

.?

i POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

e. With measurements indicating maximus'

[F$(z)h over z ( K(z)-)

M has increased since the previous determination of Fq (z) either of the following actions shall be taken:

M

1) .F9 (z) shall be increased by 2% over that specified in Specifi-cation 4.2.2.2c. or M

2) Fq (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum [F (z)kisnotincreasing.

over z \ K(z))

f. With the relationships specified in Specification 4.2.~2.2c. above.

not being satisfied:

1) Calculate the percent F (z) exceeds its limit by the following j expression: 9 fmaximum

~

M 4)

Fg (z) x W(z) 1-1 5 x 100 for P > 0.5 1

' #~*

Ff'P - 4 K(z)

(( )}

\ ,

maximum M !

, F g (z) x W(z) h-1 x 100 for P < 0.5

-1 over I g,, .. m .

a ggj-xK(z{h Il i

-2) -One of the following actions shall be taken: 3 rrc e r,can.e.s,i -

a) Within15 minutes,'controltheAFDtowithinnew30 limits i which are determined by reducing the- AFD limits of t2s4 by l.

1% AFD for each percentqF (z) exceeds -its limits as deter-mined in-Specification 4.2.2.2f.1). Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset- H the AFD alarm setpoints to these modified limits, or b) Comply with the requirements of Spec'ification 3.2.2 for Fq (z) l exceeding its limit by the percent calculated above, or H T

c) Verify that the requirements of. Specification 4.2.2.3 for l Base Load operation are satisfied and enter Base Load- I operation.

1 l

\

l l

l McGUIRE - UNITS 1 and 2 3/4 2-8 Amendment No. (Unit 1)

Amendment No. (Unit 2) j i

P POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) ,

g. The limits specified in Specifications 4.2.2.2c, 4.2.2.2e., and 4.2.2.2f.

above are not applicable in the following core plane regions:

1. Lower core region from 0 to 15%, inclusive. 4

2. Upper core region from 85 to 100%, inclusive.

8 4.2.2.3 Base Load operation is permitted at powers above APL N0 1f the following conditions are satisfied: !

a. Prior to entering Base Load operation, maintain THERMAL POWER above ND s0 >a M e srcsw.wooma APL and les bd"or an ~

eq**ua^l to that allowed by Sp fication 4.2.2.2-for at least the pr s 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base lo operation surveillance (AFD withinN5hef target flux difference during this l time period.- Base Load operation is then permitted providing THERMAL ND OL ND POWER is maintained between APL and APL or between APL and 100% (whichever is most limiting) and FQ surveillance is maintained pursuant to Specification 4.2.2.4. APLBL is defined as:

APL BL = minimumover - x K(Z)Z r ]( x 100% f- '

F (Z) x W(Z)BL where: . F 0(z) is the measured F (z) increased by the all,o.wances for

, Q Q manufacturing tolerances and measurement uncertainty.g"the F r.c me,saro r Q limit) t 2. 22. K(z) i s* g : = ' atu ss " aa r~~no~

L 2-2, er cone nom. is the cycle dependent },

W(z)BL function that accounts for limited power distribution transients en-countered during' base load operation. Nuf=u= i; ;p = M l-

= Factor =t-iet h=2 m per S f$y,ktsb a n utu es An Rrputt

b. tesking't During Basespesoporainye Load operation, f:the conf 6kinaisus THERMAL POWERstror?ecification.

is decreased below 6.9.1.9.

ND APL then the conditions of 4.2.2.3.a shall be satisfied before re-entering Base Load operation.

l 4.2.2.4 During Base Load Operation F (Z) shall be evaluated to determine if

.F (Z) is within its limit by: 9 9

a. Using the movable incore detectors to obtain a power distribution ND l

map at any THERMAL POWER above APL ,

b. Increasing the measured F (Z) component of the power distribution 9

l map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

L 1.

APL" I$ NE hoMohw MwA6sg (puc.stu ptjocu)pom sqm gog gggg gogy opgg,op, In treeweenv~ 5.1.s .

McGUIRE - UNITS 1 and 2 3/4 2-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

L

m !

i POWER'0ISTRIBUTION LIMITS

]

SURVEILLANCE REQUIREMENTS'(Continued)

, c. Satisfyi he following relationship:

F (Z) 5,p x Z) for P > APLND )l!

g,3r! n where: F (Z) is the measured F (Z). *Theq F limit,i; 2.32.

9 j l1 m roonw nro r ttie M n Furocrio.s or <ont ysour. l K(Z) is y' r '- t : :.2-2. .P is the relative THERMAL POWER. l W(Z)gg is the cycle dependent function that accounts for limited ;

. power distribution transients encountered during M operation.

O -:: fc:th.. b girr in tn- Pe ding : = L-;=it ^-s " 5 per Specification 6.9.1.9. f/",utsu o ww Aac utm,r, ,-wcont orturwo cenerspra,r,)

d .' Measuring Ff(Z) in conjunction with target flux difference deter- l mination according to.the following schedule:

1. Prior to entering BASE LOAD operation after satisfying Section 4 4.2.2.3 unless a full core flux map has been taken in the previous.31~EFP0 with the relative thermal power having been ND maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and

2. At least once'per 31 effective full power days.

e. With measurements indicating b F (Z) maximum [ ] i overcZ-hasincreasedsincethepreviousdeterminationFf(Z)eitherofthe

'following actions shall be taken:

1. Ff(Z)shallbeincreasedby2percentoverthatspecifiedin q L

4.2.2.4.c, or

2. F (Z) shall be measured at least once per 7 EFPD until 2

l. successive maps indicate.that M

Fn(Z) maximum [. g ] is not increasing.

over Z i l; l

f. With the relationship specified in 4.2.2.4.c above not being l satisfied, either of-the following actions shall be taken:

W L 1. Place the core in an equilbrium condition where the limit in 4.2.2.2.c is satisfied, and remeasure F (Z), or l

L l

l McGUIRE - UNITS 1 and 2 3/4 2-9a Amendment No. Unit 1)

Amendment No. Unit 2) l l

j POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Conthued)

< 2. Comply with'the requirements of Specification 3.2.2=for Fq(Z) t exceeding its limit by the percent calculated with the following ,- ..

expression:

ND F (Z)

[(max. over z of ( ,x K(Z) x-W(Z)BL ) ) -1 ] x 100 _ for P > APL-gjrr' p-- _

g. The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above- -

. are not applicable-in the following core plan regions:

1. Lower core region 0 to 15 percent, inclusive. ;

2. - Upper core region 85 to 100 percent, inclusive, i 4.2.2.5..When Fq (Z) is measured for reasons otherfthan meeting the requirements of specification 4.2.2.2 an overall measured F q(z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing-tolerances- ,

and further increased by 5% to account for measurement uncertainty. .

i i

t j

l l

McGUIRE - UNITS 1 and 2 3/4 2-9b Amendment No. Unit 1)

Amendment No. (Unit 2) I

r-. -r 4

!No CHANGES ^

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i : McGUIRE - UNITS 1 and 2' 3/4 2-10 Amendment No.42(Unit 1)

Amendment No.23(Unit 2) )

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Amendment No.23(Unit 2) 4 N p -4 -ww h-b - -+ + e _ ree-p er ---e + we--e- -+---,-__ac v-ww

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3/4 2-12 Amendment No. (Unit 1)

McGUIRE - UNITS 1 and 2 (Unit 2)

Amendment No.

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- McGUIRE . UNITS 1 and 2 3/4 2-13 Amendment No.43(Unit 1)

Amendment No.24(Unit 2)

i POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING E N ITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant' System (RCS) total flow-rate and R shall be maintained within the region of allowable operation stosn j,

= 9: 3.2-3 for four loop operation:

citiihto ow we cm ortnavuta conon strosr(cosn>

Where: N t a.

R = **P [1. 0 + (1.0 - P)] ,

j e .

4 THERMAL POWER i,

b. P= ,

i '

c. Fh = Measured values of Fh obtained by using the movable incere ] d detectors _to obtain a power distribution meas tw mi.map. .The,awec.ured '

usor,es ca values of Fg shall be used to calculate R sinceUr: !. :

N

( .l includes penalties for undetected feedwater venturi fouling of j' i

0.1% and for measurement uncertainties of 1.7% for flow'and 4% ,

for incore measurement of F N , p , 'i

. 5 rwr cs5m

,,Ng7, ,/jl'/d,'""C*'">"

8 N

.i, APPLICABILITY: MODE 1.

ACTION:

1

.With the combination of RCS total flow rate and R outside the region of acceptable operation + 6 - : - N ,, , : :.; :: sprunr, su we cocs: fl

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either: ,

. 1. Restore the combination of RCS total flow rate and R 4' to within the above limits, or '

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ticGUIRE - UNITS 1 and 2 3/4 2-14 Amendment No (Unit 1)

Amendment No. (Unit 2)

-9e i , ' POWER DISTRIBUTION LIMITS '

LIMITING CONDITION FOR OPERATION ,

t ACTION: (Continued)

b. Within'24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside-the above limits, verify through incore flux mapping and RCS total flow rate comparison that j

~

the combination of R and RCS total flow rate are restored to within the-above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within'the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . ;

c.- Identify and correct the cause of'the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION h.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and

(-

RCS total flow rate' comparison, to be within the region of acceptable (

operation e -- en R . r 3.2-3. prior to exceeding the following

. THERMAL POWER levels: NEa w Wr coq ,

-l l

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

{ q

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS l

4.2.3.1 The rovisions of Specification 4.0.4 are not applicable.

.-4.2.3.2 The combination of indicated RCS total flow rate determined by process computer readings or digital voltmeter measurement and R shall be j within the region of acceptable operation ef 4 : 3.3.3 > minro w wrco u: l I

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel

. loading, and

b. At least once per 31 Effective Full Power Days, i sma,r,suncm .

4.2.3.3 TheindicatedRCS-tota [flowrateshallbeverifiedtobewithinthe <

L Ee 3.2 regionmostofrecently acceptable operatioV value of ' "E obtained per Specification 4.2.3.2, isat leas the assumed to exist.

obtained

{ ll 1

s 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

l 1.

McGUIRE - UNITS 1 and 2 3/4 2-15 Amendment No. Unit 1)

Amendment No. Unit 2)

L 5 .________________m_n_______.______-____

g

, F 3

. NO CHANGES THIC PAGE FOR INFORMATION ONLY ,

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Amendment No.43 (Unit 1) l McGUIRE - UNITS 1 and 2 3/4 2-16 Amendment No.24 (Unit 2) l

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1 McGUIRE - UNITS 1 and 2 3/4 2-18 Amendment No.42(Unit 1)

Amendment No.23(Unit 2) v

l t

l

m. 4 ,<t m lVITY CONTROL SYSTEMS l

l BASES :

3/4.1.! RORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A ufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently l subcritical to preclude inadvertent criticality in the shutdown condition. :

SHUTDOWN MARGIN requirements vary throughout core life as a function of The most restrictive fueldepletion,,RCSboronconcentration,andRCST,yfn.

condition occurs at E0L, with T,yg at no load operat g temperature, and is i assedettd with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% delta k/k is required to control the reactivity transient, Accordingly, the SHUT 00WN MARGIN requirement is based upon this limiting con-j !

i dition and 1s consistent with FSAR safety analysis assumptions. With T,yg i less than 200*F, the reactivity transients resulting from a postulated steam :

line break cooldown are minimal and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection. ,

3/4 1 1 1 MnDERATOR TEMPERATURE COEFFICIENT '

The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

i The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other l than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. p ,, ,, 3,, ,,,, ,,, ,

r-, s (m.c L ens (ted ,

l The most negative MTC value quivalent to the most positive oderator density coefficient (MDC), was o tained by incrementally correctin the MDC used in the FSAR analyses to nom nal operating conditions. These co rections involved subtracting the increm tal change in the MDC associated wi i a core condition of all rods inserted most positive MDC) to an all rods wit drawn condition and, a conversion for the rate of change of moderator densit with I temperature at RATED THERMAL P ER conditions. This value of the MDC as then transformed into the limiting TC value;-4.1 - 10 i _.i b 5?F. The HTC

, value -4 .2 - 10 ; m-h- L""** represents a conservative value (with L

corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting,MTC value.vf "_1 - - - , i En I1cGUIRF - UNITS 1 and 2 B 3/4 1-1 (Unit 1)

Amendment No.

, Amendment No. (Unit 2)

. _ _ _ . . _ . _ _ _ _ _ . _ . _ _ _ ._ _ _ s

REACTIVITY CONTROL SYSTEMS of OMy 4 BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUMTEMPERATUR&FORCRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F. This limitation is required to ensure: (1) the moderator temperatui<e coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimus RT temperature.

NDT 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facilit/ operation. The components required to I perform this function include: (1) borated water sources. (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat L Tracing Systems, and (6) an emergency power supply from OPERABLE diesel l

generators.

l With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of-either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay and cooldown to 200'F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires .

16,321 gallons of 7000 ppm borated water from the boric acid storage tanks or l 75,000 gallons of 2000-ppm borated water from the refueling water storage tank (RWST).

With the RCS temperature below 200'F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 300'F provides assurance

, that a mass addition pressure transient can be relieved by the operation of a I single PORV.

McGUIRE - UNITS 1 and 2 B 3/4 1-2 Amendment No. 42 (Unit 1)

Amendment No. 23 (Unit 2) m

l l REACTIVITY CONTROL SYSTEMS 1

BASES i,

l B0 RATION SYSTEMS (Continued) .

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200*F to 140'F. This condition requires either 2000 gallons of 7000-ppm borated water fros' the4oric acid storage tanks or 10,000 gallons of 2000 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentestion of the RWST also ensure a pH value of between 8.5 and 10.5 for the soldion recirculated within containment after a LOCA. This pH bend minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained. (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance'with the control rod alignment and insertion list (5,g,,

re t co inn a e.,sco.a m.,, w, snr,.~~ s.. ,ser,n.s u~. rs 4.e4 wome, , n,r Nr l The ACTION statements which permit limited variations from the basic .

requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement l of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or }

equal to 551*F and with all reactor coolant pumps operatinjVEnsures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are /

required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is /

inoperable. These verification frequencies are adequate for assuring that the/ # '

applicable LCO's are satisfied.

Mus uws nu.ar Ptn snucicoow s,q,s q, McGUIRE - UNITS 1 and 2 B 3/4 1-3 A, em,,w u,, (w.,,, )

th< sn m Ho, tury m

m NO CHANCEO THIS PAGE FOR INFORMATION ONLY REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued)

For Specification 3.1.3.1 ACTIONS c. and d., it is incumbent upon the plant personnel to verify the trippability of the inoperable control rod (s).

This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism.

During performance of the Control Rod Movement periodic test (Specifica-tion 4.1.3.1.2), there have been some " Control Malfunctions" that prohibited a control rod bank or group from moving when selected, as evidenced by the demand counters and DRPl*. In all cases, when the control malfunctions were corrected, the rods moved freely (no excessive friction or mechanical interference) and were trippable.

This surveillance test is an indirect method of verifying the control rods are not immovable or untrippable. It is highly unlikely that a complete con-trol rod bank or bank group is immovable or untrippable. Past surveillance and operating history provide evidence of "trippability."

Based on the above information, during performance of the rod movement test, if a complete control rod bank or group fails to move when selected and can be attributed to a " Control Malfunction," the control rods can be considered

" Operable" and plant operation may continue while ACTIONS c. and d. are taken.

If one or more control rods fail to move during testing (not a complete bank or group and cannot be contributed to a " Control Malfunction"), the affected control rod (s) shall be declared " Inoperable" and ACTION a. taken.

(

Reference:

W 1etter dated December 21, 1984, NS-NRC-84-2990, E. P. Rahe to Dr. C. (1 Thomas)

Digital Rod Position Indicators i

i McGUIRE - UNITS 1 and 2 B 3/4 1-3a Amendment No. 77 (Unit 1) ]

Amendment No. 58 (Unit 2)

i 3/4.2 POWER DISTRIBUTION LIMITS ;

i BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop- ,

erties to within assumed design criteria. In addition, limiting the peak linear ;

power density during Condition I events provides assurance that the initial '

conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria l limit of 2200'F is not exceeded. I The definitions of certain hot channel and peaking factors as used in these specifications are as follows: i 1

l FS (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 j l by the average fuel rod heat flux, allowing for manufacturing toler-l ances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of Fh the integral of linear power along the rod with the highest integrated power to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE ThelimitsonAgtFLUXOIFFERENCEJAFD)assurethattheF bound envelope ofb' tin $e# Io5aNze[aN1"keaNi[faYtNq(Z) is not exceeded uppe 7l during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER i

.for the associated core burnup conditions. Target flux differences for other l THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup i

l considerations.

l 1

l l

McGUIRE - UNITS 1 and 2 B 3/4 2-1 Amendment No. (Unit 1)

Amendment No. (Unit 2) w

POWER DISTRIBUTION LIMITS I BASES AXIAL FLUX DIFFERENCE (Continued) l ND

M "' *>

At power levels below APL , the limits on AFD are definedey 9-- i a-i- l 1.e. that defined by the RAOC operating procedure and limits. These limits were calculated in a manner such that expected operational transients, e.g. load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of .

time allowed outside of the limits at reduced power levels will not result in significantxenonredistributionsuchthattheenvelopeofpeakingfactorkD would change sufficiently to prevent operation in the vicinity of the APL power level, g m M C O L A, At power levels greater than APL , tw odes of operation are permissible;

1) RAOC, the AFD limitiof which are defined -- V t : 1. and 2) Base Load operation, which is defined as the maintenance of the AFD within att::K band t

8**

about a target value. The RAOC operating procedure above APL ND is the same as that defined for operation below APL ND However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions l inthemaximumallowedpowerorAFDinordertoguaranteeoperationwithFft) less than its limiting value. To allow operation at the maximum pemissible valueg the Base Load operating procedure restricts the indicated AFD to relatively small W

^

target band and power swings (AFD targ'e fb I Ed e k 'APLND i power < APLO ' or 100% Rated Thermal Power, whichever is lower). For Base Load operatTon, it is fl expected that the plant will operate within the target band. Operation outside of the target band for the short time period allowed will not result in signi-ficant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact from past operation on the Base Load operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months waiting period at a power level ND above APL and allowed by RAOC is necessary. During this time period load

changes and rod motion are restricted to that allowed by the Base Load pro-cedure. After the waiting period extended Base Load operation is permissible.

i: The computer determines the one minute average of each of the OPEP.ABLE l- excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are: 1) outside the allowed al power operating space (for RAOC operation), or 2) outside the l allowed AI target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation),

or 2) APLND (for Base Load operation). Penalty deviation minutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

McGUIRE - UNITS 1 and 2 B 3/4 2-2 Amendment No. (Unit 1)

Amendment No. Unit 2) m

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENMA .PY RI5E HOT CHANNEL FACTOR N The limits-on heat flux hot channel factor, RCS flow rate, and nuclear o' enthalpy rise hot channel factor ensure,that: (1) the design limits on peak I local power density and minimum DNBR are not exceeded, and (2) in the event of d a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS accep- l tance criteria limit rmt s,e.as ang ves.raro w n c can onurn s.~ ors nemr m u rsonce n e,s s.n.e.o.

4 Each of as specified these is measurable in Specifications 4.2.2 and 4.2.3. but will Thisnormally only be determine periodic surveillance is '

sufficient to insure that the limits are maintained provided: ;

a. Control rods in a single group move together with no individual rod l" -

insertion differing by more than t 13 steps from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described !:

in Specification 3.1.3.6; <

c. The control rod insertion limits of Specifications 3.1.3.5 and i 3.1.3.6 are maintained; and :"

d. The axial power distribution, expressed in terms of AXIAL FLUX L DIFFERENCE, is maintained within the limits, i L N

F ' ions a te.hrough t

AH will be maintainedmwithi.n m . n srza,suo its limits o n. ?rovided conc nearws Conditnneer.

e a) h

d. above are maintained. As noted onP - 3 ^ 3. RCS flow rate and pow,er < ,1 may be " traded off" against one another (i.e. , a low measured RCS flow d ,

rate is acceptable if the power level is decreased) to ensure that the calcu- d lated DNBR will not be below the design DNBR value. h TherelaxationofFhas ,

a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits, y e,s ne ursmto w we coa >

R as calculated in Specific ion 3.2.3 and used in"c 5, accounts N

for F 3g less than or equal to' M

'"*t*!"lue This va

- r " ' *

is used in the various accident

['q ,

l analyseswhereFfH influences parameters other than DNBR, e.g. , peak clad ten- (

perature, and thus is the maximum "as measured".value allowed. l Margin between the safety analysis limit DNBRs il ? .... ;.^^ .'_. O S h(,

E- ' p i;;i :: 8 , :_:;::t';:5) and the design limit DNBRs @ ^^ ~ i " e r M ' 1 .. ...: ^4/. ce' - ' : , =: t u riy is maintained. A fraction of this margin is utilized to accommodate the transition core DNBR penalty (2%) and ;

the appropriate fuel rod bow DNBR penalty (WCAP - 8691, Rev. 1). ,

When an qF measurement is taken, an allowance for both experimental. error and manufacturir'ig tolerance must be made. An allowance of 5% is appropriate j McGUIRE - UNITS 1 and 2 8 3/4 2-2a Amendment No. (Unit 1)

Amendment No (Unit 2)

I

NO FOR INFORMATION O Y

4 This page deleted.

's I

i 1icGUlpr . UNITS 1 and 2 B 3/4 2-3 Amendment No. 42 (Unit 1) l Amendment No. 23 (Unic 2)

i c - __ _ . _ . . - - _ _ - _ . - _ .

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE 9 HOT CHANNEL FACTOR (Continued) j j i

for a full-core' map taken with the Incore Detector Flux Mapping System, and a '

l 3% allowance is appropriate for manufacturing tolerance. >,

N When RCS flow rate and FAHare measure no additional allowances are l l mot erumt a me cu. < !

necessary prior to comparison with the li its of'rC,,:.^ ^ Measurement ?JJ errorsof1.7%forRCStotal'flowrateand4%forFhhavebeenallowedforin determination of the design DNBR value. hi J

o i The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate [J;, ;

indicators, Potential fouling of the feedwater venturi which might not be ? (

detected could bias the result from the precision heat balance in a non- /

conservative manner. Therefore, a penalty of 0.1% for undetected fouling of J the feedwater venturi is included in f r- ?l 3. Any fouling which might ?

bias the RCS flow rate measurement gtl'ater than 0.1% can be detected by ?i monitoring and trending various plant erformance parameters. If detected, ,

action shall be taken before performi subsequent precision heat balance i measurements, i.e., either the effect f the fouling shall be quantified and [,

compensated for in the RCS flow rate me$surement or the venturi shall be cleaned to eliminate the fouling. b .n,c n g,,, ,fn ,ng ,,, ,,, ,, ,, g ,

The 12-hour periodic surveillange of indicated RCS flow is sufficient to 9 l detect only flow degradation which 4ould lead to operation outside the accept- l able region of operation shown onh- - --_-

strume,

-l The hot channel factor F (z) is measured periodically and increased by a i.

cycle and height dependent power factor appropriate to either RA00 or Base I ,

Lead operation, W(z) or W(z)BL, to provide assurance that the limit on the p hot channel factor, F (z), is met. W(z) accounts for the effects of normal )

9 operation transients and was determined from expected power control maneuvers d over the full range of burnup conditions in the core. W(z)BLaccountsfor ],

P the more result in lessrestrictive severeoperating transient limitsvalues. allowed by) Base The W(z function Load for operation which.

normal operation p A n=f M 4 " 'a y br" t W per Specification 6.9.1.9. (;,

^~* not % v~ n~ ros we who onow~ Ast seu opse, o- m sen wguvuc w,,n nrns- d ll f

McGUIRE - UNITS 1 and 2 B 3/4 2-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

l ADMINISTRATIVE CONTROLS diws ww wsiQ k.1.9 PEAKING FACTOR LIMIT REPORT The W(z nctions for RAOC ar.d Base Load operation and the value fe d ll I (as require shall be established for each reload core and imp e 6nted prior ( )

to use. ?' !

< 1 The methodology used t enerate the W(z) functions r RAOC and Base Load c l ND Operation and the value for shall be th previously reviewed and ap- < l ods are deemed necessary they will proved by the NRC*. If changes these i i

be evaluated in accordance with 10 .59 and submitted to the NRC for re- ? l view and approval prior to their u he change is determined to involve an l unreviewed safety question or i uch a c e would require amendment of pre- i viously submitte.d documenta n.

A report containin t (z) functions for ROAC and Load operation and the value for AP (as required) shall be provided to th RC document con- ( ;

trol desk wit opies to the regional administrator and the re ent inspector (

within 30 s of their implementation. 4 ND Any formation needed to support W(z), W(z)BL and APL will be by req st rom the NRC and need not bi included in this report. I 1 i SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the l NRC Regional Office within the time period specified for each report.

1 l

l l

9- h, 1

WCAP-10216 " Relaxation or tonn: -t ?H +Mnet ControMn surveiiiam. TEM'(

[ '

_) (l l [ A irication". '

l

[ ;

l McGUIRE - UNITS 1 and 2 6-21 Amendment No. (Unit 1)

Amendment No. (Unit 2) m )

m y l

}lMSERT Q CORE OPERATING LIMITS REPORT t 6.9.1.9 Core operating limits shall be established and documented in !

the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

,' 1. Moderator Temperature Coefficient BOL and EOL limits and ;

300 ppm surveillance limit for specification 3/4.1.1.3, j

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, !

i

3. Control Bank Insertion Limits for Specification 3/4.1.3.6, !

4. Axial Flux Difference limits, target band, and APLND for '

[ Specification 3/4.2.1, >

TP

[ 5. Heat Flux Hot Channel Factor, F ,K(Z),W(Z),APL ND and W(Z)BL f r Specification 3/4.2.2, and RTP

6. Nuclear Enthalpy Rise Hot Channel Factor, F3g , and Power i Factor Multiplier, MFAH, limits f r Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC :

in: ,

1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALVATION METHODOLOGY", July 1985 (W Proprietary).

(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank ;

Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat ,

Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot thannel Factor.) ,

2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL 0FFSET ,

CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Off set Control) and 3.2.2 -

Heat Flux Hot Channel Factor (W(Z) surveillance requirements for Fg Methodology).)

t

3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

?

'The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core >

[

thermal-hydraulic limits, ECCS limits, nuclear limits such as L

1

F l L -

1

{ lN SE AT Q$ (caNrnswcoj l

r shutdown margin, and transient and accident analysis limits) of

! the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector, l'

t F

t r

5 P

E w ,, , ,---

ATTACID4ENT IB .

I Catawba Units 1 and 2 Technical Specifications Changes Requests

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k i' C 5

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/-

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f

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e !

DEFINIT!0NS {

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SECTION PAGE i 1.0 DEFINITIONS i i

1.1 ACT!0N........................................................ 1-1 !

1. 2 ACTUATION LOGIC TE5T.......................................... 1-l' ,

1 i

1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST..............................

1. 4 AXIAL FLUX DIFFERENCE......................................... 1-1 ;

1.5 ' CHANNEL CALIBRATION........................................... 1-1

1. 6 CHANNEL CHECK................................................. 1-1 i 1.7 CONTAINMENT INTEGRITY......................................... 1-2 .

1. 8 CONTROLLED LEAKAGE............................................ 1-2 !

CORE ALTERATION............................................... 1,-

.,2 - j

1. 9.. c o m m m ~ m or..... .. . . . _ ,. . .... . .. .... ;

1- 2 l 1.).6' o DO S E EQU IVALENT I- 131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

ENERGY.............................. 1-2 l

1. K I-AVERAGE DISINTEGRATION 1.)E ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 l <

n

1. W,,FREQUENCY NOTATI0N........................................... 1-3 l LEAKAGE........................................... 1-3 l

1. $ IDENTIFIED 1.HfgMASTERRELAYTEST............................................ 1-3 l

1. g , MEMBER (5) 0F THE PUBLIC...................................... 1-3 l :

1-3 l-1 4 0FFSITE DOSE CALCULATION MANUAL..!........................... ;

1. % OPERABLE - OPERABILITY....................................... 1-4 l

1. R , OPERATIONAL MODE .M0DE...................................... 1-4 l 1-4 1 +

1. K PHYSICS-TESTS................................................

l1 4 , PRESSURE BOUNDARY LEAKAGE...................... ............. 1-4 l 1 1.,g3 PROCESS CONTROL PR0 GRAM...................................... 1-4 l- l 1.JAq, PURGE - PURGING.............................................. 1-4' l

-1.2 % QUADRANT POWER TILT RATI0............................'........ 1-4 1-1.)8;g RATED THERMAL P0WER.......................................... 1-5 l, 1.)81, REACTOR BUILDING INTEGRITY................................... 1-5 l 1 -

1.J# REACTOR TRIP SYSTEM RESPONSE TIME............................

1-5 l ,

1.38i, REPORTABLE EVENT............................................. 1-5 l 1.)9;, SHUTDOWN MARGIN.............................................. 1-5 l 1.M,SITEB0VNDARY................................................ 1-5 l ;

1-5 j

1. K SLAVE RELAY TEST.............................................

CATAWBA - UNITS 1 & 2 I A w w . e r N ,, (y ,7 0

. Ahe+ensus Ho. (um t)

IL_R _ . _- . - _ _ . _ - , _ _ -

DEFINITIONS SECTION PAGE 1.)f'n SOLIDIFICATION............................................... 1-5 1 1 4 SOURCE CHECK................................................. 1-6 l 1.)s STAGGERED g TEST BASIS......................................... 1-6 1 1.3816 THERMAL P0WER................................................ 1-6 l 1 4 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 l 1.)Tg UNIDENTIFIED LEAKAGE......................................... 1-6 l 1 4 UNRESTRICTED AREA............................................ 1-6 l 1 4 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-6 l 1.68;, VENTING...................................................... 1-7 l l

1. % WASTE Gt.S HOLDUP SYSTEM...................................... 1-7 l TABLE 1.1. FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M00ES....................................... 1-9 e

l e

L t

t s

1 CATAW8A - UNITS 1 & 2 II A"f*"f" A* (""O l-1 Angrenter Nor (Muor 1) m

ND CHANCE 3 THl3 PAGE._,

i' FOR INFORMATION ONLY

~

l SAFETY LIMITS AM LIMITING SAFETY SYSTEM SETTINGS ;

t SECTION PAGE l r

2.1 SAFETY LIMITS ?

2.1.1 REACTOR C0RE................................................ 2-1 l 2.1.2 REACTOR COOLANT SYSTEM PRE 55URE............................. 2-1 I FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP01NTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 j

9 BASES SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE......................< ........................ B 2-1 ,-

l l 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 D

2.2 LIMITING SAFETY SYSTEM SETTINGS L 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS............... B 2-3 l l

[

l i i :

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n CATAWSA - UNITS 1 & 2 III L

I6 t , p - , , . - , - - - . . . + -. ,m, . ++--_.- - --___-_._.A.M.

p LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE ;

3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T avg > 200*F........................... 3/4 1-1 Shutdown Margin - T ,y, 1 200'F........................... 3/4 1-3 Moderator Temperature Coefficient........................ 3/4 1-4 Minimus Temperature f or Cri tical i ty. . . . . . . . . . . . . . . . . . . . . . 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 -

Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 'e Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - Operating........................ 3/4 1-12

-3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALY5ES REQUIRING REEVALUATION IN THE .,

EVENT OF AN IN0PERA8LE FULL-LENGTH R00................... 3/4 1-16 ,

l Position Indication Systems - Operating.................. 3/4 1 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop, Time............................................ .3/4 1-19

. Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Bank Insertion Limits. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-21

^iG";; i -i ^^: ;J.. . ;;;5hI;C; L;FiI5 'ish5d5 7-E;A-L ^OM"5

-..-...:. .. ; _ ; . ; , , a 2.f1 _ ;;-

l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1- AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 *

^^U^E : -

i ;;;L ;LE~ DI;;E:E ^E _IMIT; ai a 0-CTi^; ^;

fii-0 T-t==at 90=ti .. . 1/3 ^4 -

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F q (Z)..................... 3/4 2-5 l~

,- CATAWBA - UNITS-I & 2 IV Aa.ree,,r W. (% d Amenee A4. (w- r t)

~ _. -_ ___ _. . _ _ _ . . _ . -. _ _ . . . _ . . _ _ _ . _ . _ _ _ _ - . _ _ . . _ .

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RIOUIREMENTS i

SECTION PAGE

."ICJ: 11 : 4(Z) - -21. ZEEE ~Q(E; fl i ~..~:^TIOb gr ;;r.: '41GNT. ~9 2 2 - #- l 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R.................................. 3/4 2-9

.: : : _: rr c- : _=r-===

m : = = = -- = .4_,

- .:. :_:37 ;__ gr. .- ;-'-'

-- ' ".'~.'~"....

3/4.2.4 QU AD RANT POWE R T I LT RAT I 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-12 3/4.2.5 DNb PARAMETERS........................................... 3/4 2-15 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2' TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS........................... 3/4 3-27 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 1 INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-42 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-51 TABLE 3-3-6 RACIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS............. ....................... 3/4 3-52 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-54 Movable Incore Detectors................................. 3/4 3-55 Seismic Instrumentation.................................. 3/4 3-56 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-57 i TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-58 Meteorological Instrumentation........................... 3/4 3-59 i

CATAWBA - UNITS 1 & 2 V A n, . ~,, r 9. . (w~.ru 1 Arm.,n v.; Nr 4 1

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TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-60 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE I REQUIREMENTS.......................................... 3/4 3-61 l Remote Shutdown System............... ................... 3/4 3-62 )

TABLE 3.3-9 REMOTE SNUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-63 ]

TABLE 4.3-6 REMOTE $NUTDOWN MONITORING INSTRUMENTATION I SURVEILLANCE REQUIREMENTS................................ 3/4 3-64 Accident Monitoring Instrumentation...................... 3/4 3-65 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-66 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-68 Chlorine Detection Systems................................ 3/4 3-70 Fire Detection Instrumentation........................... 3/4 3-71 3/4 3-73 e TABLE 3.3-11 FIRE DETECTION INSTRUMENTATION ......................

Loose-Part Detection System.............................. 3/4 3-77 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-78 i

TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-79 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-81 !

Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-83 TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/4 3-84 TABLE 4.3-9 RADI0 ACTIVE GASEQUS EFFLUENY MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-88 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-91 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops Fi11ed............................. 3/4 4-5 Cold Shutdown - Loops Not Fi11ed......................... 3/4 4-6 CATAWBA - UNITS 1 & 2 VI

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FOR INFORMATION ONLy LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES Shutdown................................................. 3/4 4-7 0perating................................................ 3/4 4-8 ,

3/4.4.3 PRES $URIZER.............................................. 3/4 4-9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-12 -

TABLl! 4.4-1 MINIMUM NUMBER OF STEAM GENERATOR $ TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-17 ,

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-19 Operational Leakage...................................... 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4'4-22 3I4.4.7 CHEMISTRY...............c................................ 3/4 4-24 TABLE 3.4-2 REAC10R COOLANT SYSTEM CHEMISTRY LIMITS. . . . . . . . . . . . . . . 3/4 4-25.

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY ' LIMITS SURVEILLANCE l REQUIREMENTS............................................. 3/4 4-26 l

3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC l ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-29 '

t TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-30

3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO 16 EFPY................................. 3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -

APPLICABLE UP TO 16 EFPY................................. 3/4 4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITH0RAWAL SCHEDULE...................................... 3/4 4-35 Pressurizer.............................................. 3/4 4-36 i Overpressure Protection Systems.......................... 3/4 4-37 3/4 4-39 3/4.4.10 STRUCTURAL INTEGRITY.....................................

3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-40 l

l_ CATAWBA - UNITS 1 & 2 VII

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3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ;

Cold Leg Injection....................................... 3/4 5-1 t Upper Head Injection..................................... 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F........................... 3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - T avg

< 350'F........................... 3/4 5-9 ;

3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ;

Containment Integrity.................................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 TA8LE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS............ 3/4 6-5 Containment Air Locks.................................... 3/4 6-8 Internal Pressure........................................ 3/4 6-10 g ;

Air Temperature.......................................... 3/4 6-11 ;

Containment Vessel Structural Integrity.................. 3/4 6-12 Reactor Building Structural Integrity.................... 3/4 6-13 Annulus Ventilation System............................... 3/4 6-14 Containment Purge Systems..................'.............. 3/4 6-16 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/4 6-18 I

i 3/4.6.3 CONTAINMENT ISOLATION VALVES............................. 3/4 6-20 TA8LE 3.6-2. CONTAINMENT ISOLATION VALVES.......................... 3/4 6 22 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................ 3/4 6-38

Electric Hydrogen Recombiners............................ 3/4 6-39 i Hydrogen Mitigation System............................... 3/4 6-40 -

3/4.6.5 ICE CONDENSER Ice Bed.................................................. 3/4 6-41 Ice Bed Temperature Monitoring System.................... 3/4 6-43 Ice Condenser 0eors...................................... 3/4 6-44 Inlet Door Position Monitoring System.................... 3/4 6-46 CATAW8A - UNITS 1 & 2 VIII I m

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Divider Sarrier Personnel Access Doors and Equipment Hatches...................................... 3/4 6-47 Containment Air Return and Hydrogen Skimmer Systems...... 3/4 6-48 Floor 0 rains............................................. 3/4 6-50 Refueling Canal 0 rains................................... 3/4 6-51 Divider Barrier Sea 1..................................... 3/4 6-52 TABLE 3.6-3 DIVIDER SARRIER SEAL ACCEPTA8LE PHYSICAL PROPiRTIES... 3/4 6-53 3/4.6.6 CONTAINMENT VALVE INJECTION WATER SYSTEM ................ 3/4 6-54 3/4.7 PLANT SYSTEMS 3/4.7.1 TUR8!NE CYCLE Safety Va1ves............................................ 3/4 7-1 TA8LE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETP0!NT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION...................................... 3/4 7-2:

TA8LE 3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Specific Activity........................................ 3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-7 Main Steam Line Isolation Va1ves......................... 3/4 7-8 Condensate Storage System................................ 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-11 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM............................. 3/4 7-12 3/4.7.5 STAN08Y NUCLEAR SERVICE WATER PON0....................... 3/4 7-13 3/4.7.6 CONTROL ROOM AREA VENTI LATION SYSTEM. . . . . . . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.7 AUXILIARY SUILDING FILTERED EXHAUST SYSTEM............... 3/4 7-17 3/4.7.8- SNU88ERS................................................. 3/4 7-19 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUB 8ER FUNCTIONAL TEST........... 3/4 7-24 3/4.7.9 S EALED SOURCE CONTAMINATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-25 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System............................ 3/4 7-27 Spray and/or Sprinkler Systems........................... 3/4 7-29 CD: Systems.............................................. 3/4 7-31 Fire Hose Stations....................................... 3/4 7-33 !

CATAWBA - UNITS 1 & 2 IX m

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3/4.7.11 FIRE BARRIER PENETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-36 ;

3/4.7.12 GROUNDWATER LEVEL........................................ 3/4 7-38 3/4.7.13 STAND 8Y SHUTDOWN SYSTEM.................................. 3/4 7-40 3/4.8 ELECTRICAL POWER SYST[MS 3/4.8.1 A.C. SOURCES 0perating................................................ 3/4 8-1 i

TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-9 TABLE 4.8-2 LOAD SEQUENCING TIMES................................. 3/4 8-10 Shutdown..... ........................................... 3/4 8-11 3/4.8.2 D.C. SOURCES 0perating................,................................ 3/4 8-12 TABLE 4.8-3 BATTERY SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . . . . . . . 3/4 8-15 ,

Shutdown................................................. 3/4 8-16 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-17 Shutdown................................................. 3/4 8-18 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective 0evices..................................... 3/4 8-19 ,

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES....................................... 3/4 8-21 l

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... 3/4 9-1 L

3/4.9.2 INSTRUMENTATION.......................................... 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 9-7 l 3/4.9.6 MAN I P U LATO R C R AN E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-8 l 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.......... 3/4 9-9 j 1

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) FCQ INFO ~WATION ONLY LIMITING Can!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1......................................... 3/4 9-10 Low Water Leve1.......................................... 3/4 9-11 3/4.9.9 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-12 1

3/4.9.10 WATER LEVEL - STORAGE POOL .............................. 3/4 9-13 i 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM. . . . . . . . . . . . . . . . . 3/4 9-14 1

1 3/4.10 SPECIAL TEST EXCEPTIONS j 3/4.10.1 SHUTOOWN MARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 l 3/4.10.3 PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS.................................... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN.................... 3/4 10-5 ,

J 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR0 GRAM.................................................. 3/4 11-2 00se..................................................... 3/4 11-5 Liquid Radwaste Treatment System......................... 3/4 11-6 Liquid Holdup Tanks...................................... 3/4 11-7 l Chemical Treatment Ponds................................. 3/4 11-B 3/4.11.2 GASEQUS EFFLUENTS Dose Rate................................................ 3/4 11-9 TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM................................................., 3/4 11-10 Dose - Noble Gases....................................... 3/4 11-13 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form............................. 3/4 11-14 Gaseous Radwaste Treatment System........................ 3/4 11-15 Explosive Gas Mixture.................................... 3/4 11-16 Gas Storage Tanks........................................ 3/4 11-17 CATAWBA - UNITS 1 & 2 XI

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LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11.3 SOLID RADIDACTIVE WASTES................................. 3/4 11-18 3/4.11.4 TOTAL 00SE............................................... 3/4 11-19 ;

i 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES................................, 3/4 12-9 i

TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS................................................. 3/4 12-10 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-15 s

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3/4.0 APPLICABILITY............................................... B 3/4 0-1 )

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3/4.1 REACTIVITY CONTROL SYSTEMS 1

3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER O!STRIBUTION LIMITS................................... B 3/4 2-1 .

3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R............................................ B 3/4 2-2

' FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS .

THERMAL P0WER............................................. B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 ;

3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR C00LANT LOOPS AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 4-1 B 3/4 4-1 3/4.4.2 ' SAFETY VALVES.............................................

3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 1

3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY............................,............ B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-7 1 1

CATAWBA - UNITS 1 & 2 XIII

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BASES l SECTION PAGE i TABLE 8 3/4.4-1 REACTOR VESSEL 70VGNNESS.......................... 8 3/4 4-9 j FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF j FULL POWER SERVICE LIFE.................................. 8 3/4 4-11 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SNIFT OF RT NOT FOR REACTOR VESSELS EXPOSED TO 550'F............ B 3/4 4-12 ,

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3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/4 4-16 l 3/4.4.11 EEACTOR LOOLANT SYSTEM VENTS............................. B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. s 3/4 5-1 )

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... 8 3/4 5-1 l 1

3/4.5.4 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 l l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... 8 3/4 6-4 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 i 3/4.6.4 COM8USTIBLE GAS CONTR0L................................... B 3/4 6-4 3/4.6.5 ICE C0NDENSER............................................. 8 3/4 6-5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-2 ;

3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ 8 3/4 7-3 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM.............................. B 3/4 7-3 l 3/4.7.5 STAN08Y NUCLEAR SERVICE WATER PON0........................ B 3/4 7-3 I l 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM...................... B 3/4 7-3 3/4.7.7 AUXILIARY BUILDING FILTERED EXHAUST SYSTEM................ 8 3/4 7-4 3/4.7.8 SNU8BERS.................................................. B 3/4 7-4

[ 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS.................................. B 3/4 7-6

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3/4.7.11 FIRE BARRIER PENETRATIONS................................. B 3/4 7-7 l' 3/4.7.12 GROUN0 WATER LEVEL......................................... B 3/4 7-7 3/4.7.13 STAND 8Y SHUTOOWN SYSTEM................................... B 3/4 7-8 CATAW8A - UNITS 1 & 2 XIV

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BASES PAGE SECTION 3/4.8 ELECTRICAL POWER SYSTEMS j i

3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, ONSITE POWER DISTRIBUTION ................................ B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 l'

3/4.9 RFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4'9-1 3/4.9.2 ~INSTRUMO *JTION........................................... B 3/4 9-1 3/4.9.3 DECAv T m ................................................ B 3/4 9-1

, 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1 !

3/4.9.5 C0PNUNICATIONS............................................ B 3/4 9-2.

3/4.9.6 MANIPULATOR CRANE......................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT rUEL STORAGE POOL BUILDING. . . . . . . . . . . B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............. B 3/4 9-2 3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR' VESSEL and ,

STORAGE P00L.............................................. B 3/4 9-3 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM.................. B 3/4 9-3 3/4.10. SPECIAL TEST EXCEPTIONS ;

1 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS..... B 3/4 10-1

~3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1

'3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......................................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS........................................ B 3/4 11-4 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................. B 3/4 11-8 3/4.11.4 TOTAL 00SE............................................... B 3/4 11-8 1

CATAWBA - UNITS 1 & 2 XV m

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t 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 {

. 3/4,12.2 LAND USE CENSUS........................... . ............. B 3/4 12-1 {

,: 3/4.12.3 .INTERLABORATORY. COMPARISON PR0 GRAM...................... B 3/4 12-2 -l l

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5.1.1 EXCLUSION AREA.............................................. 5-1 5.1. 2 LOW POPULATION Z0NE.......................... .............. 1 5.1. 3 NAPS'OEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADICACTIVE GASEOUS AND LIQUID EFFLUENTS.................... 5- 1 FIGURE 5.1-1 EXCLUSION AREA....................................... 5-2 FIGURE 3.1-2 LOW POPULATION 10NE.................................. 5-3 i FIGURE 5.1-3 UNRESTRICTED AREA AND SITE B0UNDARY FOR RADI0 ACTIVE LIQUID EFFLUENTS................................ 5-4

)

FIGURE 5.1-4 UNRESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE GASE005 EFFLUENTS............................... 5-5 5.2 CONTAIMENT 5.2.1 CONFIGURATION............................................... >

5-1

.5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-6 1

5.3 REACTOR C0RE i i

5.3.1- FUEL ASSD8 LIES............................................. 5-6 .;

5.3.2 CONTROL R00 ASSEMBLIES...................................... 5-6 ;

5.4 REACTOR COOLANT SYSTEN 5.4.1 '0ESIGN PRESSURE AND TEMPERATURE............................. 5-6 j 5.4.2 V0LUME...................................................... 5-6 l

5.5 NETE0R0 LOGICAL TOWER LOCATION................................. 5-6 -;

5.6 FUEL STORAGE 5.6.1 CRITICALITY.................:............................... 5-7 5.6.2 .0RAINAGE.................................................... 5-7 5.6.3 CAPACITY.................................................... 5-7

' 5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-7 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................

5-8 F-

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CATAWBA - UNITS 1 & 2 XVII

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NO CHANGES THIS PAGE FOR INFORMATIONj ONLY !

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App 1N15TRATIVE W O \

l g[1{g PAGE 6.1 RESPONSIBILITY.............................................. 6-1 l

6.2 ORGANIZATION................................................ 6-1 6.2.1 0FFSITE and DNSITE ORGANIZATIONS.......................... 6-1 l l 6.2.2 UNIT STAFF................................................ 6-1 J FIGURE 6.2-1. DELETED............................................ 6-3 FIGURE 6.2-2 DELETED............................................ 6-4 i TA8LE 6.2-1 MINIMUM SHIFT CREW COMPOSITION...................... 6-5 6.2.3 CATAWBA SAFETY REVIEW GR0VP............................... 6-6 l Function.................................................. 6-6 1 Composition............................................... 6-6 l Responsibilities.......................................... 6-6 Records................................................... 6-6 6.2.4 SHIFT TECHNICAL ADVIS0R.................................... 6-6 6.3 UNIT STAFF QUALIFICATIONS................................... 6-6 6.4 TRAINING.................................................... 6-6 6.5 REVIEW AND AUDIT............................................ 6-7 6.5.1 TECHNICAL REVIEW AND CONTROL ACTIVITIES................... 6-7 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSR8)........................ '

6-8 Function.................................................. 6-8 Organization.............................................. 6-9 P Review..................................................... 6-10 Audits.................................................... 6-10 Records........................,.......................... 6-11 6.6 REPORTABLE EVENT ACTI0N..................................... 6-12 6.7 SAFETY LIMIT VIOLATION...................................... 6-12 6.8 PROCEDURES AND PR0 GRAMS..................................... 6-13 6.9 REPORTING REQUIRENENTS...................................... 6-15 6.9.1 ROUTINE REP 0RTS........................................... 6-15 Startup Report............................................,

6-15 f

CATAWBA - UNITS.1 & 2 XVIII Amendment No. 56 (Unit 1)

Amendment No. 49. (Unit 2)

W

l ADMINISTRATIVE CONTROLS j SECTION PAGE Annual Reports............................................ 6-16 l Annual Radiological Environmental Operating Report........ 6-16 Semiannual Radioactive Effluent Release Report............ 6-17 Monthly Operating Reports................................. 6-19 Co L;ftJ, JLt AAD",V.ereeki.v Fai.ec Limit 3 Report........................ 6 _l 6.' 9. 2 SPECIAL REP 0RTS........................................... 6-19 6.10 RECORD RETENTION........................................... 6-19 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-21 6.12 HIGH RADIATION AREA........................................ 6-21 6.13 PROCESS CONTROL PROGRAM (PCP).............................. 6-22 6.14' 0FFSITE 00SE CALCULATION MANUAL (00CM)..................... 6-23 6.15 MAJOR CHANGES TO LIQUID GASEOUS. AND SOLID RA0 WASTE TREATMENT 5YSTEMS.......................................... 6-23 l i

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7.0 SPECIAL TEST PR0 GRAM........................................ 7-1 i

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CATAWBA - UNITS 1 &-2 XIX Amt-va w 6- ("*r0 l Ar,eense no. tu-.r 1)

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1.0 DEFINITIONS _ONty The defined terms of this section appear in capitalized type and are applicable i throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions. ,j ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall 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 simulated-signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL ,

OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter- !

lock and/or: Trip Setpoints such that the Setpoints are within the required range and accuracy. !

~ AXIAL FLUX DIFFERENCE

1. 4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between 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 the i channel such that it responds within the required range and accuracy to known '

values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

' CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the chhnnel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CATAWBA - UNITS 1 & 2 1-1 m

4 1

DEFINITIONS e 1

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

'1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-2-of Specification 3.6.3.

b. 'All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

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

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERA 8LE.

s CONTROLLED LEAKAGE

-1.8o CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. 1

.. CORE ALTERATION ,

1.9 CORE. ALTERATION shall be the movement or manipulation of any component- '

within the reactor pressure vessel with'the vessel head removed and fuel in-the vessel.- Suspension of CORE ALTERATION shall not preclude completion of-movement of a. component to a safe conservative position. ;

f(3ksterGDE) 005 i-EQUIVALENT I-131

'1.I8' 00SE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic i mixture ~of I-131, I-132, 1-133, 1-134, and I-135 actually present. The thyroid 1 dose ~ conversion factors used for this calculation shall be those listed in Table"III of-TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." ,

E - AVERAGE DISINTEGRATION ENERGY 11 1,3t' I shall be the average (weighted in proportion to the concentration of j

-each radionuclide in the sample) of the sum of the average beta and gamma ,

energies per disintegration (MeV/d) for the radionuclides in the sample.

CATAWBA - UNITS ~1 & 2 1-2 Arr~ err ~r vo, (umr 0 Anownt~rNor (u-urQ n a

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! CORE OPERATING LIMITS REPORT l'.10The CORE OPERATING-LIMITS-' REPORT (COLR) is the unit-specific document ,

that provides core operating limits for the current operating reload cycle.-

These cycle-specific core operating limits shall: - be determined for each reload cycle in accordance.with Specificacion 6.9.1.9. Unit operation within these w -operating > limits is' addressed in individual spectfications. 3

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-a'> J - , , - , , - , _ , . ., , . . , , ,

,y DEFINITIONS ,

ENGINEERED SAFETY FEATURES RESPONSE TIME n

1..Z The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time j interval from when the monitored parameter exceeds its ESF Actuation Setpoint

i. at'the channel sensor until the ESF equipment is capable of performing its !

safety function (i.e., the valvas travel to their required positions, pump, discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION-

-1. h The FREQUENCY NOTATION specified for the performance of Surveillance l Requirements shall correspond to the intervals defined in Table 1.1.

_ IDENTIFIED LEAKAGE 1.hIDENTIFIEDLEAKAGEshallbe: l

a. Leakage (except CONTROLLED LEAKAGE)'into closed systems, such as pump t seal or valve packing leaks that are captured and conducted to a sump '

or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both i specifically located and known either not to interfere with the opera-tion of Leakage Detection Systems or not to be PRESSURE BOUNDARY-LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST

.n-1.yf A MASTER RELAY TEST shall be the energization of each master relay and j verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay. ,

1 MEMBER (S) 0F'THE PUBLIC 1.)hMEMBER(S)0FTHEPUBLICshallincludeallpersonswhoarenotoccupa- l tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

.This category does include persons who use portions of the site for recre-

.ational, occupational, or otner purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 18 1.)T .The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology l and parameters used in the calculation of offsite doses due to radioactive I gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ- l mental Radiological Monitoring Program. I

~ CATAWBA - UNITS 1 & 2 1-3 A er~,rs u r Me, (u-n l)

A ntwe rscur klo, kr1,)

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OEFINITIONS OPERABLE - OPERABILITY M-

1. W A system, subsystem, train, component or device shall be OPERABLE or )

have OPERA 8ILITY 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 are L '

required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE te 1..W An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive )

combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 11 1.je" PHYSICS TESTS shall be those tests performed to measure the fundamental l nuclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE B0UNDARY LEAKAGE 1.)d'PRESSUREBOUNDARYLEAKAGEshallbeleakage(exceptsteamgeneratortube l4 leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM 13 1.JE The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l ,

, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be. accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other require-ments governing the disposal of-radioactive waste.

PURGE - PURGING 1.jdPURGEorPURGINGshallbeanycontrolledprocessofdischargingairorgas j from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO w

1.J# QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore l detector calibrated output'to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

CATAWBA - UNITS 1 & 2 1-4 A n e,w c ur A, (hr t) h k't M M wr hjo, {%r t)

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DEFINITIONS RATED THERMAL POWER u

1.jHT RATED THERMAL POWER shall be a total reactor core heat trar!sfer rate to l the reactor coolant of 3411 MWt.

REACTOR BUILDING INTEGRITY ,

l 1.h-REACTORBUILDINGINTEGRITYshallexistwhen: l

a. Each door in each access opening is closed except when the access opening is-being used for normal transit entry and exit, then at

=least one door shall be closed,

b. The Annulus Ventilation System is OPERABLE, and c.- The sealing mechanism associated with each penetration (e.g., welds,=

bellows, or 0-rings) is OPERABLE. ;

REACTOR TRIP SYSTEM RESPONSE TIME j

st 1.) r The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from l when the monitored parameter exceeds its Trip Setpoint at_the' channel sensor until loss of stationary gripper coil voltage. .

., REPORTABLE EVENT

%d u i 1.5 r A REPORTABLE-EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50 l ]l

.SHUTOOWN MARGIN .

w 1.jW SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY ;

1.N The SITE BOUNDARY shall be that line beyond which the land is neither l owned, nor leased, nor otherwise controlled by licensee.

SLAVE RELAY TEST 3%

1.)r.A SLAVE RELAY TEST shall be the energization of each slave relay and l verification'of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION n

1. X SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

CATAWBA - UNITS 1 & 2 1-5 4 *8e ^/* (""",rO o A reeerm Ale. (4~

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- - DEFINITIONS-SOURCE CHECK'

1. k A SOURCE CHECK shall be the qualitative assessment of channel response l when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST' BASIS 3r 1.)4' A STAGGERED TEST BASIS shall consist of: l

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n 1 equal subintervals, and

b. The testing of one system, subsystem, train, or other designated ,

component at the beginning of each subinterval. 1 THERMAL POWER n

n 1. E THERMAL POWER shall be the total reactor core heat transfer rate to the-

- l-g reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST n \

1.)6' A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating'the !

Trip. Actuating Device and verifying OPERABILITY of alarm, interlock and/or gl '

trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include.

adjustment, as necessary, of the Trip Actuating Device such that it actuates at the-required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 1s-1.,7 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE l ,

or CONTROLLED LEAKAGE. i UNRESTRICTED AREA )

)?

1.38' An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY-- l access to which is not controlled by the licensee for purposes of protection of ,

^

individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, -

commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 90 1.X A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l L installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust. gases through activated 6 carbon adsorbers and/or HEPA filters for the purpose of. removing iodines or particulates from the gaseous exhaust stream prior to the release to the envi-ronment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

1 CATAWBA - UNITS 1 & 2 1-6 Amendment No. (Unit 1)

Amendment No. (Unit 2) n ,

DEFINITIONS VENTING j HI 1.A8' VENTING shall be the controlled process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration or other '

operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING. Vent, used in system, names, does not imply a VENTING process.

WASTE GAS HOLOUP-SYSTEM 4 a

1.K A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to l reduce radioactive gaseous effluents by collecting Reactor' Coolant System ,

offgases from the Reactor Coolant System and providing.for delay or holdup !

for the purpose of reducing the total radioactivity prior to release to the environment, 1

4 CATAWBA - UNITS 1 & 2 1-7 Anrnme~r W, Nr o Ame~ansn Ma , tumr u 4

REACTIVITY CONTROL SYSTEMS-M00ERATOR TEMERATURE COEFFICIENT LIMITING COISITION FOR OPERATION 3.1.1. 3 The moderator tagerature coefficieg*{MTC) N' shall beX wiku

'"**'**** *

TW tsaaert.

**' M "

'f':Og'}j

_ _ ,e.ith: Y,"'o%

07.;.7, tt,?r"*?'ieits" **' in Figure 3.1-Oy shown {_

c _. . _. . . . . - . ... .,, . . .. . -

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'II 7 _ 7 E EEI' TIE 5"redi nNen ur a b .a n '

~ '"'" " ' r .'Q 'i,.o' A '.J c 4 N F@ ve aiT ' . o n "~ ' ' ' ' "" "- ~ ~

APPLICABILITY: ';;d: 5 ::ti c s -3. 1. 1. 3:.# - MODES 1 and 2* only#.

f;::ificati a 3.1.1.^L. - MODES 1, 2, and 3 only#.

Ewa of esLL4 LIFE ttos) Lom or ACTION: -

i

a. With the MTC more positive than th imit/ vtcer,c,gwr

h r cout,2.1 0,-

ci;r: (

operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained '

s C go,w=srsnE/s}Ee[N(na:ufficient h:~T. in Tis ,; 2.1 to 0restore within 24the MTC hours to less or be-in positive than the p HOT-STANOBY /

'!:it:

-within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and

3. A Special: Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal-limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

Goh jrtortto IN THE COUt,

b. With the MTC more negative than the" limit M S;;;ificatien 0.1.1. 2.

2 ::, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , s

With K,ff greater than or equal to 1.

See Special Test Exceptions Specification 3.10.3.

CATAWBA - UNITS 1&2 3/4 1-4 Amendment No. (Unit 1)

Amendment No. (Unit 2) f '

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- REACTIVITY CONTROL SYSTEMS !

' SURVEILLANCE REQUIREMENTS j i

~

4.1.1.3 The MTC shall be determined to be within its limits during each. fuel j cycle as follows: 1 ser . a rc a nto w ru r c o ts l

a. The MTC shall'be measured and compared to the*' limit"Of 9 g re 2.1-0, p 5 '[ ~{

l prior to-initial operation'above 5% of RATED THERMAL POWER, after each fuel loading; and yt worr~ swwown sonsrsreconto oa me enn

b. The MTC shall be measured at any THERMAL POWER and compared to v

--3.2 1^ ' 3.k/k/ T (all rods withdrawn, RATED THERMAL POWER

' condition) within 7 EFPD after reaching an equilibrium boron' I

i concentration of 300 ppm. In the event this compar,ison indicates i the MTC is more negative thanf3 .2 x 10 t siv'k/^~, lthe MTC'shall be (- l l

1 remeasured, and compared to t ie EOL MTC limit of @;ifi;;tica !

- smont, w rxt cotA,3.1.1, ?b. , at least once per 4 EFP0 during the remainder of the l fuel cycle.

W 1*ren NHroMutc Li~rstrurire in twe cacaJ l

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CATAWBA - UNITS 1&2 3/4 1-5 Amendment No. (Unit 1)

Amendment No. (Unit 2)

'L (t n

a 1

'Nko y H'* 'PAag, N_oNty 1.0 i q '

1 7 0.9 _ l s

E 0,8 _

N.

0.7 Acceptable Unacceptable 1 T* 0.6 Operation peration

.v 4

{ 0.5 _ .i 8 1 0.4 _.

. :s"E j 0.3 _.

J

. 0.2 _ l 5 6

g. 0.1.

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i = 0 l l e

0 10 20' 30 40- 50 60 70 80 '90 100 II I

% of Rated Thermal Power ll l:

l l

l FIGURE 3.1-0 MODERATOR TEl1PERATURE COEFFICIENT VS. POWER LEVEL I

l I ll l

L l l

. CATAWBA - UNITS 1 & 2 3/4 1-Sa Amendment No. 14 (Unit 1)

Amendment No. 6 (Unit 2) s n a

r A -

, REACTIVITY CONTROL SYSTEMS ,

-3/4.1.3 MOVABLE CONTROL ASSEM8 LIES GROUP HEIGHT

$ LIMITING CONDITION FOR OPERATION l L

! 3.1.3.1 All full-length shutdown and control rods shall be OPERA 8LE'and positioned within *12 steps (indicated position) of their group step counter demand position.

3 APPLICABILITY: MODES 1* and 2*.

ACTION: . .

a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical-interference or L known to be untrippable, determine that the SHUTDOWN MARGIN require- .

L -ment of . Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be 'in J - HOT STAND 8Y-within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With more:than one full-length rod misaligned from the group step counter demand position by more than 212 steps (indicated position),

be in HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l;c c. With one full-length red trippable but inoperable due to causes l ' other than addressed by ACTION a., above, or misaligned from its- 4

group step counter demand height by more than *12 steps (indicated =

position), POWER OPERATION may continue provided that within 1 hour:

. 1. The rod -is restored to OPERABLE ~ status within the above alignment requirements, or a sneursarrea 3.tu, .

2. The rod is declared inoperable _and the remainder of the rods in

. the group with'the inoperable rod are aligned to within i 12 steps of the-inoperable rod'while maintaining the rod sequence _and insertion limits of4i;;= h-1=1&F4gure4rirlb, es spp14eable.

'The THERMAL POWER level shall be restricted pursuant to Specifi- 4. .

cation 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is -

performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents m remain valid for the duration of operation under these' conditions; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

1 CATAWBA - UNITS 1 & 2 3/4 1-14 Amendment No. (Unit 1)

Amendment No. (Unit 2)

t- . ]

J~ i s R '

REACTIVITY CONTROL SYSTEMS 1 LIMITING ColeITION FOR OPERATION

' ACTION'(Continued)-

c) A' power. distribution map is' obtained from the movable N

incore detectors and F0(Z) and F are verified to be AH within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

~

e d)- The THERMAL POWER level is reduced to less.than or-equal to 75% of RATED THERMAL POWER within the next hour-and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux 7 Trip Setpoint is~ reduced to less than or equal to 855 of RATED-THERMAL POWER . !

d. With more than one full-length rod'trippable but inoperable due-to i causes other than addressed by ACTION a above, POWER OPERATION may continue provided that:~ (P> t a metracerw 3.1 u. .

1. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the rods in the bank (s) with 2 the inoperable rods are aligned to within *12 steps of the. >

inoperable rods while maintaining the rod sequence and inser- i ,

tion limits of M ; = 3.1-1: 07 Mgure 3.1-lb, es ;.ppli::ble. S[

--The THERMAL. POWER level shall be restricted pursuant to' 1 Specification 3.1.3.6 during subsequent' operation,'and

2. - The inoperable rods'are restored to OPERABLE status within 4' f

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . .

SURVEILLANCE riOUIREMENTS "4.1'.3.1.1 The position of each full-length rod shall be determined to be-within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months eAcept during time intervals when the. Rod Position

. Deviation Monitor is inoperable, then verify the group positions at least once per 4: hours, i 4.1.3.1.2 -Each full-length rod not fully inserted in the core shall be ;

. determined to be OPERABLE by movement of at' least 10 steps in any one direction at' least once per 31 days.

CATAWBA - UNITS 1 &.2 3/4 1-15 Amendment No. (Unit 1)

Amendment No. (Unit 2)

_ . . . . .. . . . _ . _ _ _ . . _ _ _ . . ______ _____._,___.x m _____.___ _ _____

l t ' ..

a

" No rn 4NGes o ,

-TABLE 3.1-1 NMAr .* N G E.

ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00 t..

Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in large 'l '

-Pipes Which Actuates the Emergency Core Cooling System i

Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)

Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control .

Assembly Ejectiori) i i

I y

S

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i CATAWBA - UNITS 1 & 2 3/4 1-16 4

W

1 l

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REACTIv!TY CONTROL SYSTEMS SHUTDOWN R00 INSERTION LIMIT LIMITING CON 0! TION FOR OPERATION

'3.1.3.5 All shutdown rods shall'be 'rit: : Ud:: -

$ pts etnto W NL COPC CPERAT>w$ k,MITS PrfcKT(CoGM),

u-.tr,,- -s,% ..s,u ,,su APPLICABILITY: MODES 1* and 2*#.

ACTION:

restarro erv o we suorano~ w,,,3rtsasseo ou me c oun, With a maximum of one shutdown rod - :_.., _..._._-_a except for sutvail- )

i l lance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either: :

arrr.ar '

a. h11y uiu.. dthe rod.,At 1* "o n,~ mr owbraruou so~r 3 rec,Foto o*> me cota,on )

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

i SURVEILLANCE REQUIREMENTS w,n,,p we ,usenou umr spuuriro w me coun:.

4.1.3.5 Each shutdown rod shall be determined to be'" .; = ..... Jn: , l.

a. Within'15 minutes prior to withdrawal of any rods in Control Bank-A,-- 3 B, C, or D during an approach to reactor criticality, and l

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. !

l 1

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

A *r *ar-r ^' (""" ')

. CATAWBA - UNITS 1 & 2 3/4 1-20 u n,.o a-r no. tun,r o v

i - > -

i REACTIVITY CONTROL SYSTEMS .,

'I CONTROL SANK INSERTION LIMITS-LIMITING COMITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion a's e h en l W snsorier ru m cent o!EMTowV Litelts $tronT (coLn). ,

APPLICABILITY: . MODES la and 2*#.

ACTION:

'sn<onto su nt cou . l With the control banks inserted beyond the ebow insertion limits $ except for [ ,

surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than'or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the _^ : " _ - :, 2 sustorow smors spesovuto w sur cout, en l. I

c. Se in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be. determined to be.within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the-individual

rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 l

L.

1 1

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

l CATAWBA - UNITS 1&2 3/4 1-21 Amendment No. (Unit 1)

(Unit 2)

Amendment No.

L 1 ________ ___ - _ _ ___ a

s 3:

f917 F Gvf,Q ;; J= ^ l' 1 --

-(peny VIReherewn) /

gg p, gg (79 % . 328)

. .0 --

k 300 i

{ SANKB )

2. -

i

\

(100 % . 1611 gg (0% in -

qq . SANK C I 1,. -

100 = =

d BANKO a

to =

10 % . 47) g =-

20=

tm. 0) , ,. , , , ,

40 80 # 1# {

0 3 gpgggy i ,q) Relative Power (Percentl' FIGURE 3.11 $

RCD BANK INSERTION LIMITS VERSUS THERM AL POWER POUR LOOP OPERATION l

d

,r -

3/41 3 Amendment No. (Unit 1)

CATAWSA UNITS 1 and 2 Amendment No. -(Unit 2)

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CATAWBA.- UNITS 1&2 3/4 1-23 Amendment No. 39 (Unit 1)

Amendment No. 31(Unit 2)

.m- .

'/* * - -

i 3/4.2 POWER DISTRIBUTION LIMITS 3'/4.2.1 AXIAL FLUX O!FFERENCE (AFD) f LINITING CONDITION FOR OPERATION k

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

as streurste w we cost ortamws co~ ors percar(coen) j

a. the allowed operational spacew '----- h ~ :,.. 5. 2-1 for RAOC gl

< operation, or ,

m irront,w ms cota

b. within e::s3% target band"about the target flux difforence during -l l baseload operation -i APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER.* f ACTION:

1

a. For RAOC operation with the indicated AFD outside of the '"g = -

.i-1 $;

-limitsg struire w we coca,

1. Either restore the indicated AFD to within the C ._._ ^ ^ coce'

{' l limits within 15 minutes, or 1

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High-Trip setpoints to less than or. equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. For Base Load operation above APLND** with the indicated AXIAL FLUX' !

DIFFERENCE outside of the' applicable target band about the target flux difference:

cocs Spec ,,r .

1. EitherrestoretheindicatedAFDtowithinthe#targetSandlimits l4 within 15 minutes, or-ND

2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes. l

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER' unless the indicated AFD is within the 1 9 = i :-F limits frrcarst, -

x 7*

In NE COLR.

See Special Test Exceptions Specification 3.10.2.

(" *

h * "'

- APL ND is the minimum allowabl(e"spower level for base load operation and inM*

b = ='Md.,in the ^ , ^= d-l- ^eg:t per Specification 6.9.1.9.

n stesurare Qant opegnrmt,. ue,*T3 peront CATAWBA - UNITS 1&2 3/4 2-1 Amendment No. (Unit 1)

Amendment No. (Unit 2) v

N ui Fo ggg E8 YHIS PAGE. 1

. POWER DISTRIBUTION LIMITS TlON oNty LIMITING C0mITION FOR OPERATION

j. SURVEILLANCE REQUIREMENTS i l

4.2.1.1 The indicated AFD shall be determined to be within its limits during l:

POWER OPERATION'above 50% of RATED THERMAL POWER by: .

a. Monitoring the indicated AFD for each OPERABLE excore channel: j 1)- At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2)' At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring ;

the AFD Monitor Alarm to OPERABLE status,

b. Monitoring and logging the indicated AFD for each OPERA 8LE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least i

once per 30 minutes thereafter, when the AFD Monitor Alarm is- inoper-able. The logged values of the indicated AFD shall be assumed to. l exist during the interval preceding each logging. ;

c. The provisions of Specification 4.0.4 are not applicable.

t ,

4.2.1.2 The indicated AFD shall be considered outside of its limits when at '

least two OPERABLE excore channels are indicating the AFD to be outside the (

-limits.

~ 4.2.1.3 When in Base Load operation, the target axial 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 When in Base Load operation, the target flux difference shall be l updated at least once per 31 Effective Full Power Days by either determining I the target flux difference in conjunction with the surveillance requirements'of l Specification 3/4.2.2 or by linear interpolation between the most recently mea-- 1 sured values and the calculated value at the end of cycle life. The provisions I

- of. Specification 4.0.4 are not applicable, s

CATAWBA - UNITS 1&2 3/4 2-2 Amendment No.39 (Unit 1)

Amendment No.31 (Unit 2)

, l

. . -. -_._______---__-__s__u_ 1

I

+ ;

f 4 E H-a b _

l E

O gr k

B s

I A 120.1001

(10. 100) l ACCEPTABLE UNACCEPT Lt OPERATION f

OPERATIO E " ACCEPTABLE ,

PERATION 80 -

,. 6 80 - (21. 50)

(.38. 80) 40 - ,

l-i

,0 _

I I I I I I I I O 10 20 40 - 50 30 20 10 0 50 Flux Difference (411%

1

!s F

FIGURE 3.21 k

[ AXtAL FLUX OIFFERENCE LIMITS AS A' FUNCTION OF RATED THERM 1

l-.

(L' nit 1) t 3/423 Amendment No.

1- CATAWBA- UNITS 1 and 2 Amendment No. (L' nit 2) ,

" ~ - - . - - - - - .

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CATAWBA - UNITS 1&2 3/4 2-4 Amendment No. 39 (Unit 1)

Amendment No. 31 (Unit 2) 1 m

?

'a . . . -

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.-__'.__ EI. _ _ _ _ _ . _ _ _ _ . ___ , _ _ _ _ _ _ _ _ _ _ _ _ , . _ , _ _ _ ______ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ , _ , , _ _ _ _ _ _ , ,, _ .

p ,_ _ _

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[ POWER DISTRIBUTION LIMITS 3/4,2.2 NEATFLUXHOTCHANNELFACTOR-Fg f LIMITING CONDITION FOR OPERATION

3.2.2 F(Z)shallbelimigedbythefollowingrelationships

q {

Fq (Z) 1 ,{K(Z))forP>0.5 l Fg (Z) 1 4K(Z){forP10.5 l w,,vg, r,*rr , wt v as.w sr Rasro wrem pomo(nrr)trauroro ow we conc cyrwux p , THERMAL POWER , and mets arrear (wur,

, RATED THERMAL POWER u&amawtre r,vo K(Z) = the = " 1_ . _ - - - - - -- ap e 3.2-i for a given core height 4h Jrroter, w wr coca, APPLICABILITY: . MODE 1. k ACTION:

With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints witMn the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to e total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; sesequent POWER OPERATION- I may proceed provided the Overpower AT Trip Setpoints (value of K.) have been reducer'. at least 1% (in AT span) for each 1% Fq(Z) exceeds the limit, ar.d

b. Identify and correct the cause of the out-of-limit condition prior j~ to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be incren. sed provided Fq (Z) is demonstrated through incore mapping to be within its limit.

l l -

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l l' CATAWBA - UNITS 1&2 3/4 2-5 Amendment No. (Unit 1)

Amendment No. (Unit 2) i o

i i

POWER DISTRIBUTION LIMITS l SURVEILLANCE REQUIREMENTS , 4.2.2.1 The' provisions of Specification 4.0.4 are not applicable. N 4.2.2.2 For RA0C operation, F (z) shall be evaluated to determine if Fg(z) { ,

is within its limit by: 9 ;

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

i

b. Increasing the measured gF (z) component of the power distribution ,

map by 3% to account for manufacturing tolerances and further in-creasing the value by 5% to account for measurement uncertainties. ,

Verify the requirements of Specification 3.2.2 are satisfied, j '

c. Satisfying the ,)1owingrelationship: I M

U*) for P > 0.5 l .

FO (2) ~<P x W(z) !

M

U3)

Fg (z) $ for P 1 0.5 l W(z) x 0.5 M

m nn~nmro r,ca m a n,cw or c=nt ktow,

where F g (z) is the asured Fg (z) increased by the allowances for manufacturing tole ances and measurement uncertainty, 4 48 s the F

g limit,K(z)isy'1- .-

2;g, - -L P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

TY ^_ ::i:- i -p .... ::: 7,. air,- f m :r Li W ( _ : --4 per j Specification 6.9.1.9. T/3 krqw, wm mar vrorer, w me ccer om4we &<*as R trcar M

d. Measuring Fg (z) according to the following schedule: ;

1. Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) ;

9 J was last determined,* or

2. At least once per 31 Effective Full Power Days, whichever occurs i first. (

I l

l

During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a

. power distribution map obtained.

I CATAWBA - UNITS 1&2 3/4 2-6 Amendment No. (Unit 1)

Amendment No. ,( Unit 2)

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ . _ _ . ~ . __ v

i r

POWER DISTRIBUTION LINITS SURVEILLANCE REQUIREMENTS (Continued) .

[ e. With measurements indicating maximum F$(z) over z K(z)

N has increased since the previous determination of Fq (z) either of

'the following actions shall be taken:

N

1) F q (z) shall be increased by 2% over that specified in Specification 4.2.2.2c., or (

N

2) F q (2) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximus is not increasing.

F$(z) over z K(z)

f. With the relationships specified in Specification 4.2.2.2c. above 7 not being satisfied: s

1) Calculate the percent F q (z) exceeds its limit by the following expression:

Imaximun m

F (z) x W(z)

"li )

-I dx100 for P > 0.5 (over z r.** -

h fE x K(z) l :

P . l .

5

") 3 l F$(z)xW(z) -l a x 100 for P < 0.5 f(fmaximun 7

\

over z r,* -4M x g(g) 0.5 s ]

g* i i

2) One of the following actions shall be taken: .

5rruiremoea 5.1.s a) Within 15 minutes, control the AFD to within new D limits which are determined by reducing the AFD limits of k M by Tl, 1% AFD for each percentg F (z) exceeds its limits as deter- I mined in Specification 4.2.2.2f.1). Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these modified limits, or

{

b) Comply with the requirements of Specification 3.2.2 for i Fq (z) exceeding its limit by the percent calculated above, or c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation.

CATAWBA - UNITS 1&2 3/4 2-7 Amendment No. (Unit 1)

Amendment No. (Unit 2)

j l

POWER DISTRIBUTION LIMITS j l

SURVE!LLANCE REQUIREMENTS (Continued) )

~

l

g. The limits specified in Specifications 4.2.2.2c., 4.2.2.2e., and 3 i 4.2.2.2f., above are not applicable in the following core plane [ J regions: ;

1. Lower core region from 0 to 15%, inclusive .

2. Upper core region from 85 to 100%, inclusive. !

4.2.2.3 Base Load operation is permited at powers above APL if the following pi conditions are satisfied:

a. Prior to entering Base Load operation, maintain THERMAL P,OWER r 3 rt,n above I ime r a4-e As~r wr e w r t.: !

APL ND and less than or " qual to that allowed by Speci cation 4.2.2.2 j for at least the previou 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base Lo operation ,

surveillance (AFD within *stof target flux difference during this ll time period. Base Load operation is then permitted providing THERMAL :

ND O POWER is maintained between APL and APL ' or between APLND and $

l L 1005 (whichever is most limiting) and FQ surveillance is maintained 8

! pursuant to Specification 4.2.2.4. APL ' is defined as: l APLBL , minimum M x K(Z) -

) x 1005 }

over Z g M(Z) x W(Z)BL p

( where: FU(z) is the measured F (z) increased by the allowances for :

l-Q Q e manufacturingtolerancesandmeasurementuncertainty...msr"The n,e wnmatesro /.co as A ren~ # c.*r nen,a Q F limit,'

i

' - 2 . ^L. K(z) is*;ir - i= W e E s t. W(z)gg is the cycle dependent ;

j function that accounts for limited power distribution transients 1 encountered during base load operation. .L _J u ._ -J = h g i m. .

b 32 l

-tast:Rastor=% sit =W&as, pe r Spec i f i c ati on 6. 9.1. 9. J Fl",Wm,sn wme , Ast frecorate w we con ottunus unm prMr

b. During Base Load operation, if the THERMAL POWER is decreased below APL NO then the conditions of 4.2.2.3a shall be satisfied before j re-entering Base Load operation.

4.2.2.4 During Base Load Operation F (Z) shall be evaluated to determine if $

O F (Z) is within its limit by:

q L a. - Using the movable incere detectors to obtain a power distribution ND map at any THERMAL POWER above APL ,

b. Increasing the measured Fg (Z) component of the power distribution map by 3% to account for manuf acturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied. (,

Ah" n nur howomum Avowness (u.assEn attow) Po#ta Leveu F-a sest tono ortitanou w sn< orocsow 1.s.e. ;

CATAWBA - UNITS 1 & 2 3/4 2-7a Amendment No. (Unit 1)

Amtndment No. (Unit 2)

. _ ._.at JL_ _ ___..____ ____ .

l POWER DISTRIBUTION LIMITS i

SURVEILL'EE REQUIREMENTS (Continued) i

c. " Satisfying following re.lationship:

F (Z) 5, * "II) for P > APL ND il P x W(Z)BL rr'n where: F (Z) is the measuredg F (Z). *theq F limit,i: 2. ^ l K(Z) is Y - ' N ' -

2 P 1 N e $ Y $1ve THERMAL POWER. l '

W(Z)8L is the cycle dependent function that accounts for limited power distribution transients encountered during'M operation. % $4r i e M2: 2: 5'r : ' - u._ ; u -x ?.. .., ; ; ; t; : ^ ++, pe r :

Specification 6.9.1.9. rf"',mps wtgm smm, wwt (wr onerw wem Arner

d. MeasuringF$(Z)inconjunctionwithtargetfluxdifferencedeter- ;

mination according to the following schedule- .

1. Prior to entering BASE LOAD operation af ter satisfying surveil- ,

lance 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been

{ t ND for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and maintained above APL ,

2. At least once per 31 effective full power days.

e. With measurements indicating ,

maximum F over I I 0K(z) (*)3 M

has increased since the previous determination F (Z) either of the following actions shall be taken: ,

1. Fq (Z) shall be increased by 2 percent over that specified in 4.2.2.4c,'or

2. F (Z) shall be measured at least once per 7 EFPD until 2 successive maps indicate that maximum F (z) over z K(z) ) is not increasing,

f. With the relationship specified in 4.2.2.4c above not being $'

satisfied, either of the following actions shall be taken:

1. Place the core in an equilibrium condition where the limit in I

4.2.2.2c is satisfied, and remeasure F (Z), or CATAWBA - UNITS 1 & 2 3/4 2-7b Amendment No. (Unit 1)

Amendment No. (Unit 2)

.- ._. - _ - -.... - ..- - .- .- _ _ s a _ __ _ _ _ . - -

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i POWER DISTRIBUTION LIMITS l SURVEILL"E REQUIREMENTS (Continued) {

2. Comply with the requirements of Specification 3.2.2 for f Fq (Z) exceeding its limit by the percent calculated with the following expression: ,

N NO

[(max. over 1 ofF[ g (Z) x W(Z)BL ) ) -1 ] x 100 for P > APL gw' p X K(Z) .

g. The limits specified in 4.2.2.4c., 4.2.2.4e., and 4.2.2.4f.

above are not applicable in the following core plan regions:

f

1. Lower core region 0 to 15 percent, inclusive. ,

2. Upper core region 85 to 100 percent, inclusive, i 4.2.2.5 When F (Z) is measured for reasons other than meeting the requirements l [

9 of Specification 4.2.2.2 an overall measured Fg(z) shall be obtained from a power .

distribution map and increased by 3% to account for manufacturing tolerances ,

and further increased by 5% to account for measurement uncertainty.

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L CATAWBA - UNITS 1 & 2 3/4 2-7c Amendment No. (Unit 1)

Amendment No. (Unit 2)

L

_ . _ . _ _ _ . _ . _ . . . _ _ _ _ ..-__________2 .&________._________

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CATAWBA - UNITS 1 & 2 3/4 2-7d Amendment No. 39 (Unit 1)

Amendment No. 31(Unit 2)

. . - . . - . . . _ . . . . . - _ _ - . . . _ . . . , . --.--.,_AJL.,._,.__..-- - . _ _ . _ . -

l 1

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)

OT Ps 6 b

i 1.50 .- -

-- .- ~ a )

=: -

4 -

= _

~

1.25 _

=mr_ .

=

1.00 ==ei!!geme memenummmmmmegggg s== = ggg=,=,_=,

---+

_ .==_===.====_.__

EEimanggmau =-=-

._ _.' i

=

$ 0.75

.~-- 1 4

as '.;

^

0.50 CORE --

HEIGHT K(Z) -

- ~~

0.0 1.00 - -

6.0 1.00 0.25 0.94 10.3 -

12.0 0.65 ~

_-~ -

z_ '

_~

L 0 -

0 2.0 4.0 6.0 . 0 10.0 12.0 i CORE HEIGHT (FT) l

)

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l FIGURE 3.2-2 ,

K(Z) - NORMALIZED F (Z) q AS A FUNCTION OF CORE HEIGHT l

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CATAWBA - UNITS 1 AND 2 3/4 2-8 Amrun.r-rW.l%r0 l Arr~o e-rw S, (%,o j l , -. .- . . . _ - . _

_ _ _ _ . _ _ . , . , __ __ __ _ _ _ _______x_.a___ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..I

f l'

POWER 0!STR18UT10N LINITS 3/4.2.3 REACTOR C00.. ANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE NOT l' C _---- L FA6 'UR LIMITING M ITION FOR OPERATION ,

3.2.3 The combination of indicated Reactor Coolant System total flow rate and l ~

R shall be maintained within the region of pomissible operation ^" - "'; - {

k::$st,,for four loop operation, i snewao s er cost erranu wo..n sturtenn Where:

F"

a. Rs 9 I is # [1.0 + 4 2 (1.0 - P)]

- F/,", %et % ,

THERMA. PWER , esuk g b

-! P = RAT u THEIMA. POWER N N

Fg = Measured values of Fg obtained by using the novable incore i c.

detectors to obtain a power distribution -

esp. The esasured twc smat ur+,mre iv we cea N

l values of F g shall be used to calculate R since ~'._ _ :.: l includes penalties for undetected feedwater venturi fouling of F 0.1% and for esasurement uncertainties of 2.15 for flow and 45 .

N i

e* ry* mr rfor incore measurement of F M' APPLICNIN*.,, MDhk S,~,ueY.'" orra "" ngs,,u ' Up','r#"'Nr'c'

*"8 tn , .

M

l

a. With the combination of Reactor Coolant System total flow rate and R within '

the region of restricted operation ,

fl

- within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required by tigunE3:2E2: Nr 5' War sprceir, #w tws coa.

b. With the combination of Reactor Coolant System total flow rate and R within ( '

l the region of prohibited operation r _ ": _ -- i. 2-i; wrcarro ,v we c oa . J l

1. 'Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either: .

a) Restore the combination of Reactor Coolant System total flow ?

t rate and,R to within the region of pemissible operation, or j ;

b) Restore the combination of Reactor Coolant Systes total flow !

rate and R to within the region of restricted operation and l comply with action a. above, or ,

c) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER {

and reduce the Power Range Neutron Flux - High Trip Setpoint <

to less than or equal to 55% of RATED THERMAL POWER within j the next 3rtsorov, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .,e sur e n n,

2. Within24["ursofinitiallybeingwithintheregionofprohibited 1 ,

operation ; . N,_. _ :._ -

, verify through incere flux espping I ,

and Reactor Coolant System total flow rate comparisen that the com-bination of R and Reactor Coolant Systes total flow rete are restored -

to within the regions of restricted or permissible operation, or i reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

CATAWBA - UNITS 1 & 2 3/4 2-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

POWER O!$TRIBUTION LIMITS LIMITING nun! TION FOR OPERATION ACTION (Continued) i 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit )

required by ACTION b.1.c) and/or b.2., above; subsequent POWER OPERA- 1 i TION may proceed provided that the combination of R and indicated )

Reactor Coolant System total flow rate are demonstrated, through j incore flux mapping and Reactor Coolant System total flow rate 1 comparison, to be within the regions of restricted or permissible operation " :i,_. 3.5-3. prior to exceeding the following l l Jart,r,r , w w r ce u l THERMAL POWER levels: ;

a) A nominal 50% of RATED THERMAL POWER, b) A nominal 75% of, RATED THERMAL POWER, and c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95%

of RATED THERMAL POWER.

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

4.2.3.2 The combination of indicated Reactor Coolant System total flow rate i determined by process computer readings or digital voltmeter measurement and R shall be determined to be within the regions of restricted or permissible operation M Q ; ;.24 frecer,ra,s wr c.ca: )l

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days., ,

f $ Prs, Fore w put com 4.2.3.3 The indicated Reactor Coolant System total flow rate tobewithintheregionsofrestrictedorpermissibleoperationgallbeverified c' N -- 2-3 fl ,

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained value of R, obtained per Specification 4.2.3.2, is assumed to exist.

4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected

to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.

4.2.3.5 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

+

CATAWBA - UNIT 5 1 & 2 3/4 2-10 Amendment No. (Unit 1)

Amendment No. (Unit 2)

. . . - . __ . ~ . . _ _ _ . _ _ . _ -_.__.___..____..._m.2_.________ .

H f

ptNALiiss OP 4.1% POR UNDETICTED Ptt0WA1N VEN?ual *>>

r POUuNS ANS MSA3WREMSNT UNetRTA4NTilt OF 3.1% POR j pggy Aug 4 gg POR IN00Rt MEASURSMSNT OP 7" a ARS p',

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Ffeutt 3.2 3 EACTOR COOLANT SYSTEM TOTAL FLOW AATE VtA5US R 80U2 LOOPS IN CPERATION s l

Amendment No. '(Unit 1)

Amendment No. (. Unit 2)

CATa4A

UNff! 1 & 2 1/4 P.M

- - -- - -. v _ . _ _ . _

3/4.1 REACTIVITYCONTROLSYSTifjS 0 y E.

BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTOOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made suberitical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, boron concentration, and T,yg. The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled Reactor Coolant System cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% Ak/k is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T lessthan200'F,thereactivitytransientsresultingfromapostulatedstelil9 line break cooldown are minimal and a 1% Ak/k SHUTDOWN MARGIN provides adequate protection. .

'3/4 1...1 3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains Nithin the limiting condition assumed in the FSAR accident and transient analyses. .

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC value equivalent to the most positive moderator density coefficient (MOC), was obtained by incrementally correcting the MDC used in the FSAR. analyses to nominal operating conditions. These corrections D

e CATAWBA - UNITS 1 & 2 B 3/4 1-1 l

- m

.i' l

REACTIVITY CONTROL SYSTEMS J 1

BASES I i

j l

MODERATOR TEMPERATURE COEFFICIENT (Continued) penasod'*' W ru, or c,w t.in (rad 1 involved subtracting the increme al change in the MDC associat with a core l condition of all rods inserted ( st positive MOC) to an all rod withdrawn i condition and, a conversion for he rate of change of moderator insity with 1 temperature at RATED THERMAL P ER conditions. Thisvalueofthe MDC was then transformed into ttg_limitin TC value. 4 1 n =/ ti. The'MTC 1; l value rf 4 2 - 2 : - A represents a conservative value (with '

1 corrections for burnup and soluble boron) at a core condition of 300 ppe '

equilibrium boron concentration andJ obtained by making these corrections to l the limiting.MTC valuegri 2.1 ; = =t 3;.

to6

$l i The Surveillance Requirements for measuremerit of the MTC at the beginning :

and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the t reduction in boron concentration associated with fuel burnup.

t 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY l

This specification ensures that the reactor will not be made critical with the Reactor Coolant System a'erage v temperature less than 551'F. This i limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the P-12 interlock is above its setpoint, (4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (5) the reactor vessel is above its minimum RT temperature.

NDT 3/4.1.2 BORATION SYSTEMS l The Boron Injection System ensures that negative reactivity control is l available during each mode of facility operation. The components required to I

perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths. (4) boric acid transfer pumps, and (5) an emergency-power supply from OPERABLE diesel generators.

With the coolant average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The l

boration capability of either flow path is sufficient to provide a SHUTDOWN CATAWBA - UNITS 1&2 B 3/4 1-2 Amendment No. (Unit 1)

Amendment No. (Unit 2) 1

_,, - ,,.m. , . - ~ - _ .-._,~c-~ ._--_A_.__.k-___________-_____

R q 8 Page, REACTIVITY CONTROL SYSTDIS 04y BASES _

' BORATION SY$TEMS'(Continued)

MARGIN from expected operating conditions of 1.35 Ak/k after xenon decay and cocidown to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppe borated water from the boric acid storage tanks l or 75,000 gallons of 2000 pos borated water from the refueling water storage tank. .j With the coolant temperature below 200*F, one Boron Injection System is I acceptable without single failure consideration on the basis of the stable i reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron ,

Injection fystem becomes inoperable. .

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except '

the required OPERABLE pump to be inoperable below 285'F provides assurance that a mass addition pressure transient can be relieved by the operation of a -

single PORV.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k af ter xenon decay and cooldown from 200'F to 340*F. This condition requires either 906 gallons of 7000 ppe borated water from the boric acid storage tanks or 3170 gallons of 2000 pas borated water

~

from the refueling water storage tank.

e The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the refueling water storage tank also ensure a pH value of between 8.5 and 10.5 for the_ solution recirculated within containment after a LOCA. This pH band 4 minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minimus SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120 and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.

Since the Digital Rod Position System does not indicate the actual shutdown rod

' position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.

CATAWBA - UNITS 1 & 2 B 3/4 1-3

- , - . ..- - - , , ~ . . . - . . - . - . - - . . - . _ . - _ _ - - rL. J

< I L j

.' REACTIVITY CONTROL SYSTEMS L

i BASES c ,_

Offs AT3"W Guh*IM RE!*AT PER MOVA8LE CONTROL ASSE4 LIES (Continued) W "'* ** " 6 Mh __ j V w coan.s nn msar, n-,s we w roo ~ n=* instanow s,s..n ssq l

{

The ACTION statements which permit limited variations f[quener om"tTebasic,~ n,e <ane requirements are accompanied by additional restrictions which ensure that the ;

original design criteria are met. Misalignment of a rod requires measurement ;

of 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 confirm that the results remain valid during future operation.

- The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T,yg greater than or equal to 551'F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced -

during a Reactor trip at operating conditions.

Control rod positions and OPERABILITY af the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more fre-quent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that .he applicable-LCOs are satisfied. , ,

For Specification 3.1.3.1 ACTIONS c. and d., it is incumbent upon the plant personnel to verify the trippability of the inoperable control rod (s). tC This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism.

During performance of the Control Rod Movement periodic test (Specifica-tion 4.1.3.1.2), there have been some " Control Malfunctions" that prohibited I[ '

a control rod bank or group from moving when selected, as evidenced by the 4 demand counters and DRPI. In all cases, when the control malfunctions were

(

corrected, the rods moved freely (no excessive friction or mechanical inter- {

forence) and were trippable. , g This surveillance test is an indirect method of verifying the control 6 rods are not immovable or untrippable. It is highly unlikely that a complete j control rod bank or bank group is immovable or untrippable. Past surveillance 7 and operating history provide evidence of "trippability".

Based on the above information, during performance of the rod movement f 4 test, if a complete control rod bank or group fails to move when selected and i can be attributed to a " Control Malfunction", the control rods can be considered

" Operable" and plant operation may continue while ACTIONS c. and d. are taken. / ,

If one or more control rods fail to move during testing (not a complete (

bank or group and cannot be contributed to a " Control Malfunction"), the 4 affected control rod (s) shall be declared " Inoperable" and ACTION a. taken. [

h

(

Reference:

W 1etter dated December 21, 1984, NS-NRC-84-2990, E. P. Rahe to Ur. C. O. Thomas) {

Y CATAWBA - UNITS 1&2 B 3/4 1-4 -Amendment No. (Unit 1)

Amendment No.

(Unit 2) >

.,-+ - e--...-~--....-%w..,--+., -I

T e ,

[1 3/4,2 POWER DISTRIBUTION LINITS mW5 i

I The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operatinn) and II (Incidents of Moderate Frequency) l events by: (1) maintaining the calculated DNOR in the core greater than or equal i to design limit DNSR during nomal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding ,

mechanical properties to within assumed design criteria. In addition, ifniting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS ,

acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in l

.these specifications are as follows: i i '

F9 (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 average fuel rod heat flux, allowing for manufacturing tolerances en fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratier of Fh the integral of linear power along the rod with the highest integrated power to the average rod power; and 3/4.2.1 AXIAL FLUX DIFFERENCE >

m rtwr uuor,to m nq ces ortura somn strem(cous)

The limits o AXIAL FLUX DIFFERENCE (AfD) assure that the F (Z) upper q i bound envelope of{.22 times the normalized axial peaking ] e factor is not I during either normal operation or in the event of xenon redistribution following power changes.

I'

. Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal l position for staaty-state operation at high power levels. The value of the. ;

target flux difference obtained under these conditions divided by the fraction I of RATED THE N L POWER is the target flux difference at RATED THERMAL POWER for-the associated core burnup conditions. Target flux differences for other l THE N L poler levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup l considerations.

l CATA%4A - UNITS 1 & 2 8 3/4 2-1 Amendment No. (Unit 1)

Amendment No. (Unit 2) )

. }

\

__ ._ . - ~ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . - - _ . . ~ .__ . ______.________ JL_JL_

l POWERDISTRIBUTIONLIM[TS T l BASES ,

At power levels below APL NO , the limits on AFD are defined -tup la MC cou, h

----- T ? L 1.e. , that defined by the RAOC operating procedure and limits. i These limits were calculated in a manner such that expected operational tran-sients, e.g. , load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the ;

short period of time allowed outside of the limits at reduced power levels will not result in significant xenon redistribution such that the envelope of peak-ing factors would change sufficiently to prevent operation in the vicinity of '

the gpg NO power level.

'# '4 #e !

Atpowerlev$1sgreaterthanAPLNO , two modes o operation are permis- 1; -

sible; 1) RAOC, the AFD limitiof which are define -- --- -

and 2) Base 1 Load operation, which is defined as the maintenance

of the AFD within 4J*Mranc, NO band about a target value. The RAOC operating procedure above APL is the same as that defined for operation below APLND However, it is possible when .

following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee opera-tion with F (z) less than its limiting value. To allow operation at the maximum permissible0 value, thre Bas: Load operating procedure restricts the indicated i t

i I'

g I

l CATAWBA - UNITS 1 & 2 8 3/4 2-2 Amendment No. (Unit 1)

Amendment No. (Unit 2)

.. . _ _ _ _ . _ _ _ _ . . _ _ . _ . _ _ _ . _ _ _ _ . - . _ __ _ v _ _ _ . _ . _ 1

l l ,

POWER DISTRIBUTION LIMITS l

BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM i FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR !

AFD to relatively small target band and power swings (AFD target band Aisrtm,re #s 1 ND h pc ceut,M, APL i p,,,7 i gpt or 100% Rated Thermal Power, whichever is !

j i

lower). For Base Load operation, it is expected that the Units will operate j within the target band. Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no '

residual xenon redistribution impact from past operation on the Base load operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months waiting period at a power level above APL ND and allowed by RA0C is necessary. During this time period load changes and rod motion are restricted to that allowed by the Base Load procedure. After the waiting period extended Base Load operation is permissible.

The computer determines the one minute average of each of the OPERA 8LE excore detector outputs and provides an alarm message immediately if the AFD

(

for at least 2 of 4 or 2 of 3 OPERABLE excore channels are: 1) outside the allowed AI power operating space (for RAOC operation), or 2) outside the allowed AI target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation), or

2) APLND (for Base Load operation). Penalty deviation minutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

The limits on heat flux hot channel factor, coolant flow rate, and nuclear enthalpy rise hot channel f actor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit. mese umo Aar vswere su me come ome,* umrs pric.4r in I utur.mnw s.so.9.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than 2 12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; CATAWBA - UNITS 1 & 2 B 3/4 2-2a Amendment No. (Unit 1)

Amendment No. (Unit 2)

I n- >

m i

j NO CHANOES THQ PACE, j

(. FOR INFORMATION ONLY-j i

i i

I

)

i This page intentionally deleted. ;

I

i !

i t !

i i

i

-i k

CATAWBA - UNITS 1 & 2 B 3/4 2-3 Amendment No.39 (Unit 1).

Amendment No.31 (Unit 2)

-.-+~,-...,.---.w-. , . - - - , , , - - . . , .%, - - - , - - - - - - - - n-*, - . , -s-- - +e + - - - . .

l l

1 POWER DISTRIBUTION LIMITS 1

BASES j l

HEAT FLUX H0T CHANNEL FACTOR. and REACTOR COOLANT SYSTEM FLOW RATE ANO NUCLEAR j ENTHALPY RI5E HOT CHANNEL FACTOR (Continued) j

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and i

d. The axial power distribution, expressed in terms of AXIAL FLUX !

O!FFERENCE, is maintained within the limits.

F N

will be maintained within its limits provided Conditions a. throu?nb h d.

AH son m.st steunto w see tese wren 4 vsen sturte* :

above are maintained. As noted on " ; : 1 ^-1, Reactor Coolant System flow rate l t

andFhmaybe"tradedoff"againstoneanother(i.e.,alowmeasuredReactor Coolant System flow rate is acceptable if the measured Ffg is also low) to ensure that the calculated DNBR will not be below the design ON8R value. The relaxation ;

of F as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

y m.neserototow meco%

R as calculated in Specification 3.2.3 and used inr:. -- ^. ^ 3, accounts M t for F jN less than or equal to' "~'n!'n'*

This valu *-"e i*s *used in the various accident analyses where F H influences parameters other than DN8R, e.g. , peak clad temp -

i erature, and thus is the maximum "as measured" value allowed. The rod bow pen- l alty as a function of burnup applied for Fh is calculated with the methods de- )

scribed in WCAP-8691, Revision 1. " Fuel Rod Bow Evaluation," July 1979, and the maximum rod bow penalty is 2. M CNBR. Since the safety analysis is performed with plant-specific safety DN8R limits ' 1 " r n-' compared to the design ;

ONBR l imi ts, d 1, F _ ._ 1. ^ ^ . ~ : n : ^_ i n ', , '- M -nic-1 =d7' irr%. -

there is,s::30E thermal margin available to offset the rod bow penalty of 2.M DNBR. i s rnutar The hot channel f actor F (z) is measured perindically and increased by a :

cycle and height dependent power factor appropriate to either RAOC or Base Load L operation, W(z) or W(2)gg, to provide assurance that the limit on the hot channel factor, Fg (z), is met. W(z) accounts for the effects of normal oper- l, ation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. W(z)g( accounts for the more ,

L restrictive operating limits allowed by Base Load operation which result in less severe transient values. The W(z) function for normal operation % }

y. ' r e r -51 re--;-

_---- Lint-- ; -

per Speci fication 6. 9.1. 9.

A~o set whv~sn~ 1 n we soo orrunea er srtuestb w we cost opeurw+ von r> per ar l

CATAWBA - UNITS 1 & 2 8 3/4 2-4 Amendment No. (Unit 1) c-Amendment No. (Unit 2) o ,

.. - - ~ . . . . . ..

q NO CHANGES THIS PAGE.

FOR INFORMATION ONLY J

t' This page intentionally deleted t

CATAWBA - UNITS 1 & 2 8 3/4 '* Amendment No. 39 (U"It 1)

Amendment No. 31 (Unit 2) v

. l POWER DISTRIBUTION LINITS t BASES HEAT F.UX HOT CHANNE'. FACTOR. and REACTOR COOLANT SYSTEM FLOW RATE AND NU

t. RIMAL >Y RISE MUT CMM9fEL FACTDR (Continued)-

WhenReactorCoolantSystemflowrata,andFharega,sugd gno,gddigi g allowances are necessary prior to comparison with the limits of": L = ^ 2 4 l

. Measurement errors of 2.1% for Reactor Coolant System total flow rate and 4%

forFhhavebeenallowedforindeterminationofthedesignDN8Rvalue.

The measurement error for Reactor Coolant System total flow rate is based f

upon perfoming a precision heat balance and using the result to calibrate the i

Reactor Coolant System flow rate indicators. Potential fouling of the i'eedwater l venturi which might not be detected could bias the result from the precision ,

heat balance in a nonconservative manner. Therefore, a penalty of 0.1% for l I undetected fouling of the feedwater venturi is included in E =4 W. Any l l i fouling which might bias the Reactor Coolant System flow te measurement

! greater than 0.1% can be detected by monitoring and tr ng various plant I performance parameters. If detected, action shall bedaken before performing l subsequent precision heat balance measurements, i s ,., either the effect of the fouling shall be quantified and compensated for n the Reactor Coolant System flow rate measurement or the venturi shall be leaned to eliminate the fouling.

' mr r%er arm.r, .a me can.

The 12-hour periodic surveillance of indicatedJeactor Coolant System could lead to opera-flow is sufficient to detect only flow degradation whiegn ib 3.63.

tion outside the acceptable region of operation shown o (

l uror.co 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-l tion satisfies the design values used in the power capability analysis. r l

Radial power distribution measurements are made during STARTUP testing and l'

i periodically during power operation.

The limit of 1.02, at which corrective action is required, provides ONB ,

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 associated with the indicated power tilt.

The 2-hour. time allowance for operation with.a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for. uncertainty on Fq is reinstated by reducing the maximum allowed puwer by 3% for each percent of tilt in excess of 1. <

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to corfirm that

/

the normalized symmetric power distribution is consistent with the QUADRANT I POWER TILT RATIO. The incore detector monitoring is done with a full incore A"r mm m. (w-r o CATAWBA - UNITS 1 & 2 8 3/4 2-5 Am. wee s. Iun.m

_ _ _ ___ _ . . _ _ _ _ _ ~. . _. ._. _. _ , _ _ _ _ .

l l

1

~ ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) i 1

. The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio- l active materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall' include any changes made l during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the l OFFSITE DOSE CALCULATION MANUAL (00CM), as well as a listing of new locations l for dose calculations and/or environmental monitoring identified by' the land I use census pursuant to Specification 3.12.2. l MONTHLY OPERATING REPORTS ;

6.9.1.8 Routine reports of operating statistics and shutdown experience, in- l cluding documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission.

Attn: Document Control Desk, Washington, D.C. 20555, with a copy to the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report. ;

T'4 LNG FACTOR LIMIT REPORT _ i ns for RA0C and Base Load operation and t for k6.9.1.9 The W(z APLND (as required) shall be pr '

to the U.S. egulatory Commis- p Document Control Desk, Was C. 20555 with copies to the ? '

sion, Attn:

Regional Administrator and the R Inspec thin 30 days of their [ ,;

implementation.

Any info . 'n'eeded to support W(t), W(2)gg and APL will ineladad in + hie raaa*+ J J f p A hRC and need not ha-QErmt wmo wheprk CATAWBA - UNITS 1 & 2 6-19 Amendment No. (Unit 1)

Amendment No. (Unit 2) r--*- , , - - . _- m - , _ .. . . _ - , + . _ . _ . - _ _ _ _ _ _ _ _ _ _ ___.__-m_ __ A __ A - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - _

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CORE OPERATING LIMITS REPORT 6.9.1.9 ~ Core operating limits shall be established and documented in ;

i. the CORE OPERATING LIMITS REPORT before each reload cycle or l any remaining part of a reload cycle for the following: j

[ 1. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for specification 3/4.1.1.3, L

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, .;

3. Control Bank Insertion Limits for Specification 3/4.1.3.6, [

4. Axial Flux Difference limits, target band, and APLND for

, Specification 3/4.2.1, ;

ND

5. Heat Flux Hot Channel Factor, FfE,K(Z),W(Z),APL and W(Z)BL for Specification 3/4.2.2, and j

6. Nuclear Enthalpy Rise Hot Channel Factor, F RTP , and Power L

Factor Multiplier, MFAH, limits for Specification 3/4.2.3.

e The analytical methods used to determine the core operating ,

limits shall be those previously reviewed and approved by NRC in:

1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION ,

METHODOLOGY", July 1985 (W Proprietary).

i-(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Oifference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June ;

1983 (W Proprietary).

~

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - +

Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FgMethodology).) ,

3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (W Proprietary). '

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor). l l The core operating limits shall be determined so that all L applicable limits (e.g., fuel thermal-mechanical limits, core

thermal-hydraulic limits, ECCS limits, nuclear limits such as i

h

n-

'['

IN SE AT @ (ceNrsward) i shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector,

((" ,

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FOR INFORMATION ONLY

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CATAWBA - UNITS 1 & 2 6-19a. Amendment No. 39 (Unit 1)

Amendment No. 31 (Unit 2) 3

- - , . _ , t

ATTACHMENT 2 Justification and Safety Analysis The requested changes to McGuire and Catawba Nuclear Stations Technical Specifications 1.0, 3/4.1.1.3, 3/4.1.3.1, 3/4.1.3.5, 3/4.1.3.6, 3/4.2.1, 3/4.2.2, 3/4.2.3, and 6.9.1.9 remove certain cycle-specific parameter limits from the Technical Specifications and relocate them to a " Core Operating Limits Report". These changes result from NRC Generic Letter 88-16. which describes an alternative (the Core Operating Limits Report) that eliminates L the need for a license amendment (i.e., prior NRC review and approval) to update the cycle-specific parameter limits for a fuel cycle, facilitating 10 CFR 50.59 reviews for future core reloads.

Background / Justification:

Generic Letter 88-16, dated October 4, 1988, was issued to encourage licensees to prepare changes to Technical Specifications related to cycle-specific parameters. These Technical Specification changes will relocate cycle-specific parameter limits from Technical Specifications to the Core Operating Limits Report (COLR), Presently the parameter limits in the McGuire and Catawba Nuclear Stations Technical Specifications are calculated using NRC-approved methodologies. These limits are evaluated for every reload cycle and may be revised periodically as appropriate to reflect changes to cycle '

specific variables. This is an administrative burden on both the NRC and Duke Power Company.

The generic letter provided guidance for relocation of certain cycle-dependent l core operating limits from the McGuire and Catawba Technical Specifications.

This would allow changes to the values of core operating limits without prior approval (1.9., license amendment) by the NRC, so long as an NRC-approved methodology for the parameter limit calculation is followed. Instead, the !

future McGuire and Catawba core reloads and other revisions would involve a safety review in accordance with the requirements of 10 CFR 50.59.

Currently, for each parameter limit proposed for inclusion in the COLR, Duke Power Company utilizes the associated methodologies identified in the revised Administrative Controls section of this license amendment request for calcula- l tion of these parameter limits during McGuire and Catawba core. reload design 1 and-when any other revisions are made.

Description of Requested Changes:

The requested technical specification changes concern the relocation of several cycle-specific core operating limits for McGuire and Catawba from i Technical Specifications to the COLR. A new definition of the COLR will be added to the Technical Specifications. Additionally, certain individual Technical Specifications will be modified to note that cycle-specific para-meter limits are contained in the COLR. A COLR paragraph will be added to the Administrative Controls Section (which will replace the Peaking Factor Limit Report). A draf t of the COLR is also provided (see Attachment 2A) and the actual report will be required to be submitted to the NRC to allow continued trending of the cycle-specific parameters.

I

'The requested changes will reference the COLR for specific parameters and will ensure that cycle-specific parameters are maintained within the limits of the COLR. .The cycle-specific parameter limits proposed for relocation to the COLR

, as part of this license amendment request include:

1 (a) 3.1.1.3 Moderator Temperature Coefficient !

(b) 3.1.3.5 Shutdown Rod Insertion Limit .

(c) 3.1.3.6 Control Rod Insertion Limits ]

(d) 3.2.1 Axial Flux Difference !

(e 3.2.2 Heat Flux Hot Channel Factor !

(f 3.2.3 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor The requested changes are consistent with the requirements of 10 CFR 50.36 and i the staff's proposed policy for improving Technical Specifications, delineated )

in SECY-86-10. " Recommendations for improving TS." The policy allows that !

process variables such as core operational limits to be controlled by

-specifying them numerically in the Technical Specifications or by specifying !

the method of calculating their numerical values if the staff finds that the ,

correct limits will be followed in operating the plant. The requested revi-sion references the NRC-approved calculation methodologies. The development ,

of cycle-specific core operating limits will continue to be performed by the ;

referenced methodologies which has been accepted by the NRC. .

The requested changes to the Technical Specifications are also considered to

be improvements and are consistent with the NRC stated policy for improving Technical Specifications (52 FR 3788, February 6,1987). "

Bases / Safety Analysis:

The current Technical Specification method of controlling reactor physics

. parameters to assure conformance to 10 CFR 50.36 (which requires the lowest functional. performance levels acceptable for continued safe operation) is to specify the values determined to be within the acceptance criteria using an NRC-approved calculation methodology. As previously discussed, the methodo-logies for calculating these parameter limits have been reviewed and approved

by the NRC and are consistent with the applicable limits in the Final Safety Analysis Report (FSAR).

The removal of cycle dependent variables from the Technical Specifications has no impact upon plant operation or safety. No safety-related equipment, safety function, or plant operations will be altered as a result of this proposed

-change. Since the applicable FSAP. limits will be maintained and the Technical Specifications will continue to require operation within the core operational limits calculated by these NRC-approved methodologies, this preposed change is administrative in nature. Appropriate actions to be taken if limits are .

violated will also remain in the Technical Specifications.

This requested change will control'the cycle-specific parameters within the acceptance criteria and assure conformance to 10 CFR 50.36 by using the approved methodology instead of specifying Technical Specification values.

The COLR will document the specific parameter limits resulting from Duke Power Company calculations, including mid-cycle or other revisions to parameter values. Therefore, the proposed change is in conformance with the require-ments oi 10 CFR 50.36. ,

, . -- l

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1 1

Any changes to the COLR will be made in accordance with the provisions of 1

- 10 CFR 50.59. From cycle to cycle, the COLR will be revised such that the '

appropriate core operating limits for the applicable cycle will apply.

Technical Specifications _will not be changed.

Conclusions:

1 These requested technical specifications changes relocate certain cycle- l specific parameter limits from the Technical Specifications to a " Core Operating Limitt Report". Based upon the preceding justification, Duke Power l Company concludes that the requested amendments are desirable from an unneces- l sary resource burden on NRC and Duke Power Company standpoint, and from a Technical Specification Improvements Policy standpoint. Based upon the ,

preceding safety analysis, Duke Power Company concludes that the requested-amendments will not be inimical to the health and safety of the public. In addition, the NRC has previously reviewed and approved similar changes on

- other plants and issued a Generic Letter recommending such changes.

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-l'0 Core Operating Limits Report h

I This Core Operating Limits Reporr (COLR) for [NAME] Unit [X) Cycle (un]

c. has been prepared in accordance with the requirements of Technical Specification < 6.9.1.9.

n The Technical Specifications affected' by this report are listed below:

3 /4 .1'.1. 3 Moderator Temperature Coefficient-3/4.1.3.5 ' Shutdown Rod Insertion Limit.

3/4.1.3.6 Control Rod Insertion Limits 3/4.2.1' Axial Flux Difference 3/4.2.2 Heat Flux Hot Channel'Sactor 3/4.2.3 Reactor Coolant System Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor 1

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f is Typical McGuire/ Catawba Core Operating Limits Report '

C 2.0 Operatina Limits The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. Those limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.9.

2.1 Moderator Temps,rature Coef ficient (Specification 3/4.1.1.3) 1 2.1.1 The-Moderator Temperature Coef ficient (MTC) 10mits are: '

The MTC'shall be less positive than the liisits shown in Figure 1 (the BOL/ARO/HZP-MTC shall be less positive than O.7 x 10-4 AK/K/ *F ) . +

The EOL/Ag0/RT7 MTC shall be.less negative than -

- 4.1x10 AK/K, F. t t

2.1.2 The MTC Surveillance limit is:

The 300 ppefARO/RTP-MTC should be less negative than or equal to -3.2x10 AK/K/ F.

L wheret BOL stands for Beginning of Cycle Life L ARO stands for All Rods Out HZP stands for Hot Zero = THERMAL. POWER q EOL stands 'for End of Cycle Life l RTP stands for RATED THERMAL POWER I

l 1

Page 2 of 16

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% of Rated Thermal Power 4 FIGURE 1 r MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL i

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0 Typical McGuire/ Catawba Core Operating Limits Report >

2.2 Shutdown Rod Insertion Limit (Specification 3/4.1.3.5) x 2.2.1 The shutdown rods shall be withdrawn to at least 228 steps.

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E2. 3 Control Rod Insertion Limits (Specification 3/4.1.3.6) 2.3.1 The control rod banks shall be limited in physical insertion as shown in Figure 2. ,

=2.4 Axial Flux Difference (Specification 3/4.2.1) ,

_2.4.1 The AXIAL FLUX DIFFERENCE (AFD) Limits are provided in Figure S . ;

2.4.2 The AFD target band ,during base load operation is +3%, ~ -3%.

2.4.3 The load minimum allowabgg)(nuclear operation (APL design) is 85% of RATED power THERMAL level for base POWER.

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Control Rod Bank Insertion Limits I

vs. Percent RATED THERMAL POWER

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Page 5 of 16

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- 2.5' Heat Flux Hot Channel Factor - Fq(Z) (Specification 3/4.2.2) l F (Z) <-Fq 9

K(Z) , for P > 0.5 '

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,Q 2.5.2. K(Z) is provided in Figure 4. .f 2.5.3 W(Z) values are provided in' Figures 5 through 7.

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2.5.4 W(Z) B values for base load operation are provided in Figures 8 [

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Page 7 of 16

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,,

1.20 :1.000

= 1.40 1.000

= 1.00 1.000 -

1.s0 1.sse 1 3,no 2.00 1.839 ;

_ 2.20

1.333 J 2.40 1.504 1,ss 2.40 1.485.

"_ 2.80 1. des

" 3.00 1.442 i

3.20 1.42s, 1.50 3.40 1.420 -

= 3.s0 1.416 r

3.80 1.410 i 4.00 1.401 ;l 1.45 .. 4.20 1. 3M 1 s 4.40 1.378 l l

4.s0 1.3ss .l 4.80 1.339, 1 1.40 7L s.00 1.316 1 s.20 1.291, -;'

s.40' 1.270 1.3s

, s.40 1.253 s.80 -

1.242.

6.00 1.239 e4 m 4.20 1.235 1.30 4.40 1.224 3 r e.e0 6.s0 1.210' 1.196 <

7.00 't.101 .y 1.2g , 7.30 1.165- '

a 7.40 -1.144' :I 7.00 1.128 I 1.20

= 7.80' ,1.107 i

- 4.00 1.003 'i

' .O.20 1.000 '

, 0.40 '1.037 1.15 B.80 1.023 i a

s.80 1.021 !

_ t.00 1.019-

=

9.20 1.013 i 1.10 9.40 1.00s

,, 9.40 1.001 1 9.80 1.000 )

" 10.00 1.000 '

l' i.Os 10.20 1.001

+ =

10.40 1.000 l m

10.40 '1.000 I

,. 1.

10.80 1.000 .J 11.00 1.000 !

0 2 4 s s 10 12

= 11.20 1.00 !

11,40 3 , co,0 ;- j L BOTTOM = 11.30 1.0001 i CORE HEIGHT (FEET) TOP , it.s0 1.000-

+ 12.00 1.o00 Figure 5 l

l. Rt.0C W(Z) as a Function of Core Height

,,, 150 MWD /MTU

Top ans settes 155 smaluese as pova Teen Spes 4.2.2.38 Page 9 of 16 L -

NEIDff et0L (FtrT) w(2)

Typical McGuire/ Catawba Core Operating Limits Report

0.00 1.000

0.20 1.000

O.40 1.000

0.00 1.000

0.80 1.000 j

1.00 1.000 <

1.20 1.000 J

1.40 1.000 1

1.60 1.000 1 1.80 1.255 1.40 2.00 '. 1.243

- 2.20 1.122-2.40 1.220 2.40 1.208 .

2.80 1.196 .,J 3.00 1.191 1.25 3.20 1.193

' 3.40 1.202 3.40 1.211 ar 3.60 1.217-4.00 1.221' ,

1.20 4.20 . 1.224 4.40 1,. $27. .,

1.220 p- m 4.40

,, --w 4.40 1.222 5.00 1.221 5.30 1.228 1.23 " :5.40 1.221: .

, o

" 5.60 1.238

- - F~"' " 5.80 1.250 S.00 1.263

~

a oi ' M w S.20 1.273 N 6.40 1.277

- 1.20 K w - J L 3 uw' O.60 1.279 G.80 1.279 +

7.00 1.279 - .

f 7.20 1.278L 7.40 1.275-1,13 7.80 1.271 7.80 1.253 8.00 1.252- 4 8.20 1.228~ .

8.40 1.221 l

1.10 -3.60 1.205. "

8.40 1.191' 9.00 1.185-9.20 1.183 9.40 1.181 9.60 1.184 1.05 9.80 1.191 <

10.00 1.197 10.20 1.204

10.40 ;1.00C

--

10.40 1.00C 1

10.00 1.00C'

M . 00 1. E 4 6 8 10 12

11.20 1.00C O 2

11.40 i .00c CORE HEIGHT (FEET) TOP

11.s0 1.00c BOTTOM - .30 . ca

11.40 f. W p,. .

Figure 6 RAOC W(Z) as a Function of Core Height 6000 MWD /MTU

./

. Top arms sottes 15% Excluene as per Teen spee 4.2.2.28 t.

Page 10 of 16 L -

. . - . - - ~ - . . . - - . . . - . ~ . ~ - ~ . . - . - - . . - _ - . . - - . ~ _ . _ - . - - , - - - . . -

t

. . t m .

,1, tl Typical McGuire/ Catawba Core Operating Limits Report h

0.00 -

M, 1.000

.

13.30 1.000

7.40. 1.000 I

0.00 1.000-

0.80 1.000 l

f.00 1.000-

f. E 1.000

t.4EF 1.000 1.40

f.90 1.000 1.80

1.218 - F 3r4L 2.00 1.209 a n.

2.20 1.200 x -

2.40 1.190 <

1.35 g 2.40 1.184 1 n 2.80 1.187 g 3.00 1.188' 3.20 1.193 3.40- 1.205 3.40 1.221 l 1.30 y .O 1.236 4.00 1.251- -!

u.v w 4.20 - 1:983 J f-m 4.40 1.273

, l 4.00 1.278 -'

1.25 '

? 4.80 1.382

l. , 'S.00 1.281 l L

S.20 1.277 I I -

w 5.40 1.284

t. A ^ S.00 1.301 l N' '

t M ^ 5.80 1.323'

- 1.20 S.00 : 1.346 I n_ w S.20: 1.346 w 1w i- S.40 1.378 .l I

S.60 1.385 S.80 1.387 1.15 7.00 1.384 l'. '

r 7.20-7.40 1.380 1.370-7.80 1.359' l '

7.80 1.338-

! S.00 1.310 1.10 S.20 1.282' 8.40 1.249-

-S.60 1.222

^8.80 1.208 9.00 1.198 -

'1.05 9.20 1.190 9.40 1.181 9.60 1.140 S.80 1.183 10.00 1.187 10.20 1.191 1 ._____ __

e 10.40 1.000

'1

10.80 1.000 O 2 4 6 8 10 12

= 10.80 1.000'

11.00 1.000 BOTTOM

11.20 1.000' u CORE HEIGHT (FEET) TOP = $1.40 1.000

. 11.60 t.000 a 11.50 1. vw pgg , 7 a 12.00 t 000 RAOC W(Z) as a Function cf Core Height 11500 MWD /MTU a

e Top and Settes 10% Exclueen as per Teen Spec 4.2.2.26 Page 11 of 16

~L

, y)) .

. .h

);

HEIGHT SQL Typical McGuire/ Catawba Core Operating Limits Report U]

, n

C.20 3%-

1.000

0.40 1.000

- 0.60 1.000

= 0.80 1.000

i.00 1.000 - 1

+

1.20 1.000 !

1.40 1.000 !

1.60 1.000 i 1.00 1.099 :

1.14 2.00 1.057' t 2.20 *1.095 2.40 1.052 ,

2.60 1.050

. 2.80 1.048

. 3.00 1.046= t t,12 3.20 1.045 3.40 1.043

-3.60 1.041-3.00 1.039

,3-'

4.00 1.037' '

l 4.20 1.035 4.40 '1.033-1.10 4.60 1.032 4.80 1.030 6.00 1.028 .,

9.30 1.028 ' -'

S.40 1.023 :

S.80 1.021 1 1.08 5.80 1.021 .I 6.00 1.024 8.30 1.028 U 6.40 1.032 3 5.80 1.035:

" .S.80 1.038.

s 1.06 - - 7.00 1.041 I c

7.30 1.043 :!

". -7.40 1.048 7.60 1.047/

"a. [ ~ %__ ; 7.80 1.048 l

-, _ - 8.00 1.049 4 1.04 8.30 1.090 J

.- 8.40 1.050 8.80 1.080 T. ; 8.80 1.049

" . ;, ;^ 9.00 1.047 9.20 1.046

".' 9.40 1.045 ,

1.02 9.60 1.044 9.80 - 1.044 I 10.00 1.046 10.20 1.046 1.000

--~

10.40-

10.80, 1,000 t :::::::: 'l ------- p

10.80 1.000

11.00- 1.000 0 2 4 6 8 to 12

11.20 1.000

11.40 1.000 g

mor rom CORE HEIGHT (FEET) TOP * " 'o

** 90.

1 000 9.000 >

2~ , ::.:: .:::

Figure 8 W(Z)3g as a Function of Core Height 150 MWD /MTU

i. -

.! # . Top ans setten iss Exsiuese as pee Teen spec 4.2.2.4G Page 12 of 16

~ _. .. .. _ ._. _ _ _ . . _ . _ _ . . . . _ _ _ __ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . . _ _ . . . - _ _

.)

( .'

p, HillDff NOL'

( FitT ) w( 2 ) '.

Typical McGuire/ Catawba Core Operating Limits Report

0.00 1.000s

. *: 0.20 1.000 e

- * - 0.40 1.000

.

0.80 1.000 ~

0.s0 1.000

1.00' 1.000

1.20 1.000-

1.40 1.000

= 1.40 1.000 -

1.a0 1.070- ,

1.10 . 2.00 .1.006 2.20', l 1.043 2.40 1.058 2.00 t.053 2.80 't.049 J 3.00 1.044-

-2 3.20 1.039

-3.40' -1.035' l 1,03 7 3.40- 1.033 ;.

,l' Z 3.80 -1.031' <

, 4.00' 1.029'

. 4.20 1.026 m

4.40 1.024

" 4.00' 1.022- 4

" 4.80 -1.021-

" 5.00 1.020 a

, 9.H .1. M 9 -.' '

X *

-1.06

^

^ 5.80- .'1. 020 --

5.80 1.024 .{

. ~ X m

6.00 1.038 ' '

-N 6.20 1.033-

" 4.40 -1.039 b I 8.00 .1.038.

' , 5.50 .1~039 7.00- 1.039

't.04 X 7.20 1.041 m . - .r 7.40' 1.043 X

X j_ 7.00 1.050 7.g@ 1,Qgd ;

Y 1.088

" " 8.00 --

JE 8.30 1.061

B.40 p _

.-- 1.065

^^ , 8.40 '1.084- x

_Z. . B.80 1.0714 !

-m. 9.00 1.074 1.02 __

=9.20 1.077 9.40 '1.000 i 9.60 1.002 9.a0 1.Os2 (

10.00 1.084 10.20 1.088

10.40 1.000

10.40 1.000 1 ___--.___

10.00 1.000

11.00 1m 0 2 4 4 4 10 12 a 11.20' 1.000

11.40 1.000 BOTTOM

CORE HEIGHT (FEET) TOP U$

- ;; . =

Is

. . .:,x J

Figure 9 W(Z)E as a Function of Core Height 600Q MWD /MTU .

e Top are settom 15% excluese as por Teen Spes 4.2.2 +46 ]

L Page 13 of 16 !

a

'j,. 4 1 ,

a t

l 4

.1 J

'l l

Ht!SH7 (CL :)

-(FEET) W(2)

Typical McGuire/ Catawba' Core Operating Limits Report

0.00 1.000 l

0.20 1.000

- = 0.40 1.000

= 0.00 1.000 !

=- 0.80 1.000 i

,i 1

= 1.00 1.000- !

1.20 1.000 :

= 1.40 1.000

1.10 = 1.s0 1.000-1.s0 1.Os0 +

2. 00 -. 1.074 2.20 1.070 ,

, Jr 2.40- 1.064 2.60 1.057 -

- 2.80 '1.050

- 3. 00 .- 1.044 <

x * * ~

1.08 3.40 1.035- i

^^

3.60 1.032' 1

2: 3.80 1.030 l-t x

4.00 1.027' ')

4.20 1.025 '

1-m x 4.40- 1'.623 4

=

4.00' 4.40 1.023-1.027'-

n .,

5.00 1.035 -- :

1.08 x x 5.20- 1.042 !

x 5,40 1.044 L

' 5.60 1.064~

1 * " 5.80 1.099 U

3 4.00 S.20 1.083 1.067 6.40 1.089 .

8.00 1.070

- 4.80 1.070-1.04 7.00 1.069 7.20 1.068 -

x n 7.40 1.005 7.80

[l' n

7.80 8.00 1.082 -

1.057 1.054 1

1 5.20 1.001 mm S.40 1.005 8.00 1.069 1.02 .40 1.073 -

9.00 1.077 9.20 1.000 9.40 1.083 S 9.80 - 1.085 9.00 1.084 -

j' 10.00 1.091=

10.20 1.092

10.40 1.000

( i

10.00' 1.000 '

O 2 4

10.00 1.000 6 4 10 12

11.00 1.000 .

g o T T r) M

= 11.20 1.000 rn0c urtnwT IcccTi -no

11 dC 1 000

.~ ....

. 11.s0 1.ooo Figure 10 * '**** '***

W(Z)BL as a Function of Core Height 11500 MWD /MTU

. Top ano settes 15s eslumes as pse Teen Sees 4.2.2.4G Page 14 of 16

W'"" ,

W ; < I

?J p: i e ,

f Typical McGuire/ Catawba Core Operating Limits Report . .,

'1., ,

N

,0 2~ 6 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor - F 0"

_( Specification _3/4.2.3) ,

s

-N F

0" R=

F H- * (l * 'U' 4H *(~)}

where P = THERMAL POWER t

RATED THERMAL POWER i 2.6.1 F 6H "

2.6.2 NF H

= 0.3 -

2.6.3 The Acceptable Operation Region from the combination of Reactor Coolant System total flow and R is provided in Figure 11.

1 i

)).

.n .i r

l i'

l '

Page 15 of 16

'I

..~-

, . . . _ _ _ . . . . . ~ . _ . _ _ _ _ _ . _ _ _ . _ . _ . . _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ . . _ . . _ . _ . _ _

o-. .

~

+

y i

pgNAL1185 SP t.1% POR UNSETICTED FttOWAftR VEN?unt -

pegung ANs usasuntaasNT UNCERTaumSS Op 3.1% Pon pggy sug 4.8% POR IN00RS MEAsuRausNT OF Ph ARE i

INCLUSIS IN THIS PISURE -

1 l

l

+

'! l ' PER$NSSISId ' ', , ' '. '. .

.PR. NE.ISN.TE. D. . . , , !,

.......,,.b i

. .OP. E.R.AT. ION, , , . . . . i ..OPERAT.10N .,

!?.01 l 6

. .i.

' Anot.0N. , e

. ! t t

, t

f. ...

4 R.E.st. 0N . p # # . .

3

. t (1.000.38.760) . ..,..i-== w 1 1 . ii.

.C RESTEICTED CPE3ATION i '

-5 38.5 4 ' '** REGI.ON- - - - -

, (,0.k9,4,38.37,2) ', ,

i Ig

-- RE.ST. R.I.CT. E.D. O.P.ER. A.TIO. N, RE, G.IO. N. , . , ,' .

....... ... .c g "- POWER < 9,61 RTP

. . a rt r s rs . . ,. ,. . .., ,

' ' - r- ' '

3S.0 -

(0.988 37.985) i- - -

E "

' ' ' ' 7 ' '

'"" RES. i.11C. i.T.D. d.P.ER.I.TI. O.N. .b. .lo. '. . .t. ,'1 . . . . '. .

l. [C. . P.O.WE.1.1. 941, .R.TP. ... i-

- . . 1

, ..(0.982,37,.,597)

I- 37.5 * " . ...,.. . . . . . . , ,

- s - -

nee......

". '"aRESTRICTED OPERATION Ig

= === REGION '

'[

".""' FOUER 1 922 RTF ' ' . . , , . . ... ....

== . . . . ....

,' (0.9.77.i37.210)- .

k ",",,",,",, RESTRICTED OPERATIDsu a 4 I 37.0 * ",i--. REGION i r

,,,,,i, POWE,Ri . <,. 90%.-iRTF, ,.

. . . i i. ,6 . . .<.

m , (? ',9,71 s,3,6.822) ,l

+.

p 36.5 - i e _

c . . .

L

' t .'a L

, l,, c.,. e .',, .i. :... 1... 1.n N

R=F AH /FRTP 6H * (1 + MF AH (1 - P)]

p Figure 11 I

RCS Flow Rate vs R - Four Loops in Operation Page 16 of 16

[

r L. n

1 l

1 1

I ATTACHMENT 3 Analysis of Significant Hazards Consideration l l

l

Introduction:

i As required by 10CFR 50.91, this analysis is provided concerning whether the requested amendments involve significant hazards considerations, as defined by J 10CFR 50.92. Standards for determination that an amendment request involves '

no significant hazards considerations are if operation of the facility in

'accordance with the requested amendment would not: 1) involve a significant

. increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different- kind of. accident from any accident'previously evaluated; or 3) involve a significant redtetion :

in a margin of safety.

The requested amendments remove certain cycle-specific parameter limits from the McGuire and Catawba Nuclear Stations Technical Specifications and relocate them to a " Core Operating Limits Report".

Analysis: j The following discussion describes how the requested amendment satisfies each >

of the three standards of 10 CFR 50.92(c). *

1) The requested change does not involve a significant increase.in the probJtbility or consequences of an accident previously evaluated.

The removal.of cycle-specific core operating-limits from the McGuire and Catawba Technical Specifications has no influence or impact on the probability or consequences of any accident previously evaluated. The cycle-specific core operating limits, although not in Technical Specifications, will be followed in the operation of the plants. The requested amendment still requires exactly the same

! actions to be taken when or if limits are exceeded as is required by current Technical Specifications.

Each accident analysis addressed in the plant's Final Safety Analysis Report (FSAR) will-be examined with respect to changes-in cycle-dependent parameters, which are obtained from application of the NRC-approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses. This examination, which will be performed per requirements of 10 CFR 50.59, ensures that future reloads will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2) The requested change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

As stated earlier, the removal of the cycle specific variables has L no influence or impact, nor does it contribute in any way to the l

probability or consequences of an accident. No safety-related i-l:

1

h 2-equipment, safety function, or plant operations will be altered as a a result of this requested change. The cycle specific variables are :

calculated using the NRC-approved methods and submitted to the NRC to allow the Staff to continue to trend the values of these. limits.

The Technical Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.

Therefore, the requested amendment does not in any way create the possibility of a new or different kind of accident from any accident previously evaluated.

3)- The requested amendment does not result in a significant reduction.

in the margin of safety.

The margin of safety is not affected by the removal of cycle- 1 specific core operating limits from the Technical Specifications.

The margin of safety presently provided by current Technical Specifications remains unchanged. Appropriate measures exist to ,

control the values of these cycle-specific limits. The proposed amendment continues to require operation within the core limits as obtained from the NRC-approved reload design methodologies and ;

appropriate actions to be taken when or if limits are violated remain unchanged.

The development of'the limits for future reloads will continue-to :'

conform to those methods described in NRC-approved documentation.

In addition, each future reload will involve a 10 CFR 50.59 safety revf ew to assure that operation of the unit within the cycle specific limits will not involve a significant reduction in a ~ margin of' safety.

Therefore, the requested changes are administrative in nature and do not impact the operation of McGuire or Catawba in a manner that involves a' reduction in the margin of safety.

.The-Commission has provided guidance concerning the application of the standards for determining whether a significant hazards. consideration exists.

This guidance (51 FR 7750)-includes examples of the type'of amendments that are considered not likely to involve significant hazards considerations. The change-proposed is similar to the examples of administrative changes.1denti- -)

fied in 51 FR 7750. Additionally, the proposed change is consistent with the NRC policy for improving technical specifications (52 FR 3788) and the pro-

~ posed change is consistent with 10 CFR 50.36 and 10 CFR 50.59. Further, the

'NRC has previously determined that similar changes on other plants did not involve a significant hazards consideration.

Conclusions:

. Based on the preceding analyses, Duke Power Company concludes that the I requested amendments do not involve a significant hazards consideration.

l l l

l'.

_ ____ _ _ _ __ _ ______.

ML20006A421

Contents

Text

Reference:

3.2.2 F(Z)shallbelimigedbythefollowingrelationships

=

Reference:

Conclusions:

Introduction:

Conclusions:

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ML20006A421

Text

Reference:

3.2.2 F(Z)shallbelimigedbythefollowingrelationships

=

Reference:

Conclusions:

Introduction:

Conclusions:

Navigation menu

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