Text

Proposed Administrative Changes to Tech Specs
	ML17146A481
Person / Time
Site:	Susquehanna
Issue date:	08/05/1986
From:	; PENNSYLVANIA POWER & LIGHT CO.
To:	;
Shared Package
ML17146A479	List: ML17146A478; ML17146A481; PLA-2698, Forwards Application for Proposed Amends 85 & 40 to Licenses NPF-14 & NPF-22,respectively,revising Tech Specs for Administrative Purposes.Fee Paid;
References
	NUDOCS 8608110186
	Download: ML17146A481 (149)
	v • d • e

DEFINITIONS INOEX

~oCT ION DEFINITIONS PAGE 1.0 ACTIONo

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1 1

1.2 1.3 1.4 1.5 1.6 AVERAGE~thlhEXPOSURE....................................

1-1 AVERAGE,PLANAR LINEAR HEAT GENERATION RATE..................1-1 CHANNEL CALIBRATION..........................,.............

1-1 CHANNEL CHECK...,........;.....,...........................

l-l CHANNEL FUNCTIONAL TEST.'...................................

l-l 1.7 1.8 1.9

l. 10 CORE ALTERATION......................................

CRITICAL POWER RATIO.................................

DOSE E(UIVALENT I-131................................

E-AVERAGE DISINTEGRATION ENERGY..................;...

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1 2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE TIME...

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END-.QFWYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE 1.13 FRACTION OF LIMITING POWER DENSITY...................

1-2 TIME..

1-2 1-3

1. 14

l. 15 FRACTION OF RATED THERMAL POWER......;.....

FREQUENCY NOTATION.........................

1-3 1-3

1. 16 GASEOUS RADWASTE TREATMENT SYSTEM..........................

1-3

1. 17

1. 18

l. 19 IDENTIFIED LEAKAGE.........................

ISOLATION SYSTEM RESPONSE TIME.............

LIMITING CONTROL ROD PATTERN...............

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1-3 1-3 1.20 LINEAR HEAT GENERATION RATE................

1.21 LOGIC SYSTEM FUNCTIONAL TEST...............

1.22 MAXIMUM FRACTION Of LIMITING POWER DENSITY.

1.23 MEMBER(S) OF THE PUBLIC.........

.24 MINIMUM CRITICAL POWER RATIO....

'L.25 OFFSITE DOSE CALCULATION MANUAL.

SUSQUEHANNA - UNIT 1 8b0811018b Bb0805 PDR 'DOCY 0500Q387 P,DR P

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1-4 1-4 1-4 1-4 1-4

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INOEX SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow..........,........

2-1 THERMAL POWER, High Pressure and High Flow................

Reactor Coolant System Pressure...........................

2-1 Reactor Vessel Water Level..........................,.....

2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......

2-3 BASES 2.7 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...................

THERMAL POWER, High Pressure and High F1ow................

Reactor Coolant System Pressure...........................

Reactor Vessel Water Level................................

8 2-1 8 2-2 B 2-/S B 2P'S 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor ProtectIon Systee InstrLssentetfon Setpofnts........

B 2j(+

SUS(UEHANNA - UNIT 1

0

LIMITING CONOITIONS FOR OPERATION ANO SURVEILLANCE RE UIREMENTS SECTION 3/4@0 APPLICABILITYe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 SHUTOOWN MARGIN..................,..............,......

3/4. 1. 2 REACTIVITYANOMALIES........................,..........

3/4. 1. 3 CONTROL'OOS Control Rod Operability................................

PAGE 3/4 0"1 3/4 1-1 3/4 1-2 3/4 1-3 Control Rod Maximum Scram Insertion Times..............

3/4 1-6 Control Rod Average Scram Insertion Times................

Four Control Rod Group Scram Insertion Times.

Control Rod Scram Accumulators...............

Control Rod Orive Coupling.............................

Control Rod Position Indication........................

Control Rod Orive Housing Support............

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer Rod Sequence Control System............................

Rod Block Monitor..

3/4. 1.5 STANOBY LIQUIO CONTROL SYSTEM..........................

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE............

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3/4 2. 2 APRM SETPOINTS.....;..................................-

3/4. 2.3 MINIMUMCRITICAL POWER RATIO...........................

3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-11 3/4 1-13 3/4 1-15 3/4 1-16 3/4 1-17 3/4 1-18 3/4 1-19 3/4 2-1 3/4 3-j(+

3/4 3-P 5 3/4. 2. 4 UNEAR HEAT GENERATION RATE GE FUEL......

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3/4 2-A8 3/4 3-EON 4I SUSQUEHANNA - UNIT 1 Amendment No.

57

INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.4 REACTOR COOLANT SYSTEH PAGE 3/4.4. 1 RECIRCULATION SYSTEH Recirculation Loops.-. 3 ~.~g.. (BAk~

@C'srCak ti'~ L~y -

SS~yte-LOLLOP OPC,r~r'~

Jet Pumps...............................

Recirculatson Pumps.......................;..........

Idle Recirculation Loop Startup......................

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

3/4 4.3 REACTOR COOLANT SYSTEH LEAKAGE 3/4 4-1 3/++-~~

3/4 4"2 3/4 4-3 3/4 4-4 3/4 4-5 Leakage Detection Systems...........

Operational Leakage.................

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Reactor Coolant System...............................

Reactor Steam Dome...................................

3/4.4.4 CHEMISTRY...............................

'.3/4 4.5.'.

SPECIFIC ACTIVITY....................................

3/4.4.'6:

'RESSURE/TEMPERATURE LIMITS 3/4 4-6 3/4 4-7 3/4 4-10 3/4 4"13 3/4 4-16 3/4 4-21 3/4.4.7 MAIN STEAH LINE ISOLATION VALVES.....................

3/4 4-22 3/4.4. 8 STRUCTURAL INTEGRITY.................................

3/4.4. 9 RESIDUAL HEAT'EHOVAL ot Shutdown o

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H Cold Shutdown........................................

3/4.5 EHERGENCY CORE COOLING SYSTEHS 3/4.5. 1 ECCS - OPERATING.....................................

3/4.5.2 ECCS - SHUTDOWN...........................

3/4.5.3 SUPPRESSION CHAMBER.............................

3/4 4"23 3/4 4-24 3/4 4-25 3/4 5-1 3/5 5"6 3/4 5"8 SUS(UEHANNA - UNIT 1 Vi

INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIRENENTS PAGE REFUELING OPERATIONS (Continued) 3/4.9.8 WATER LEVEL -

REACTOR VESSEL..........................

3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL.................

3/4 9-12 3/4-9. 10 CONTROL ROD REMOVAL Single Control Rod Removal............................

3/4 9-13 MultiPle Control Rod Raeoval..........................

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION H

h igh Water Level......................................

ow Water Level.......................................

L 3/4. 10 SPECIAL TEST EXCEPTIONS 3/4 9-15 3/4 9-17 3/4'-18 3/4. 10. 1 PRIMARY CONTAINMENT INTEGRITY..

3/4. 10.2 ROD SEQUENCE CONTROL SYSTEM....

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3/4 10-3 3/4.10 '4.

RECIRCULATION LOOPS....................................

3/4 10-4 3/4. 10.

TRAINING STARTUPS.....................................

3/4 10-5'USQUEHANNA

- UNIT 1

BASES INDEX SECTION 3/4. 0 APPLICABILITY............................................

3/4.1 REACTIVITY CONTROL SYSTEMS PAGE B 3/4 0-1 3/4.1.1 SHUTDOWN MARGIN..................................

B 3/4 l-l 3/4.1. 2 3/4.1. 3 REACTIVITY ANOMALIES.............................

CONTROL RODS.....................................

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS.................'...

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.....,.............

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION B 3/4 1-1 8 3/4 1"2 B 3/4 1-3 8 3/4 1-4 3/4.2.2 3/4.2.3 3/4.2. 4 TEo

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RA APRM SETPOINTS...................................

MINIMUM CRITICAL POWER RATIO.....................

LINEAR HEAT GENERATION RATE......................

8 3/4 2-1 B 3/4 2g I B 3/4 2-P 2-B 3/4 2-j( 3 3/4.3 INSTRUMENTATION 3/4.3.1 3/4.3.2 3/4.3.3 3/4.3.4 3/4.3.5 REACTOR PROTECTION SYSTEM INSTRUMENTATION........

ISOLATION ACTUATION INSTRUMENTATION..............

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...............................

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RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION........................

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REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.......................

B 3/4 3"1 B 3/4 3-2 B 3/4 3-2 3/4 3-3 B 3/4 3-4 3/4.3. 6 CONTROL ROD BLOCK INSTRUMENTATION...............

B 3/4 3-4 SUSQUEHANNA - UNIT 1 X11

BASES SECTION INSTRUMENTATION (Continued)

INOfX PAGE 3/4. 3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation......

Seismic Monitoring Instrumentation........

Meteorological Monitoring Instrumentation.

Remote Shutdown Monitoring Instrumentation Accident Monitoring Instrumentation.......

Source Range Monitors.....................

Traversing In-Core Probe System..

Chlorine Detection System......

Fire Detection Instrumentation............

Radioactive Liquid Effluent Instrumentation..

Radioactive Gaseous Effluent Instrumentation.

Loose-Part Detection System..

8 3/4 3"4 8 3/4 3-4 8 3/4 3"5 8 3/4 3-5 8 3/4 3"5 8 3/4 3"5 8 3/4 3"5 8 3/4 3-5 8 3/4 3"6 8 3/4 3-6 8 3/4 3-6 8 3/4 3"6 3/4.3.8 TURBINE OVERSPEED PROTECT/ON SYSTEM.............

8 3/4 3-7 3/4. 3. 9 FEEDWTER/HAIN TURBINE TRIP.

SYSTEM INSTRUMENTATION....................

3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM........

3/4.4.2 SAFETY/RELIEF VALVES.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE ACTUATION 8 3/4 3-7 8 3/4 4"1 8 3/4 4-2-

3/4.4.4 Leakage Detection Systems.............,........

Operational Leakage..

CHEMISTRY.....

8 3/4 4"2 8 3/4 4-2 8 3/4 4"2 3/4.4. 5 SPECIFIC ACTIVITY.

3/4.4. 6 PRESSURE/TEMPERATURE LIMITS.....................

8 3/4 4-3 8 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES................

8 3/4 4-5 3/4.4.8 STRUCTURAL INTEGRITY 8 3/4 4-5 3/4. 4. 9 SUS(UEHANNA "

RESIDUAL HEAT REMOVAL UNIT 1 X111 8 3/4 4-5

INDEX BASES SECTION 3/4. 10 SPECIAL TEST EXCEPTIONS PAGE 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY...................

8 3/4 10-1 3/4. 10.2 ROD SEQUENCE CONTROL SYSTEM.

8 3/4 10-1 3/4. 10.3 SHUTDOWN MARGIN DEMONSTRATIONS..................

8 3/4 10"1 3/4.10.4 RECIRCULATION LOOPS 8 3/4 10" 1 3/4. 10. g TRAINING STARTUPS.......................

3/4. 11 RADIOACTIVE EFFLUENTS 3/4. 11. 1 LIQUID EFFLUENTS Concentration.......

Dose...,-

8 3/4 10-1 8 3/4 11-1 8 3/4 11-1 Liquid Waste Treatment System....,...............

8 3/4 11-2 3/4. 11.2 GASEOUS EFFLUENTS Dose Rate..

Dose-Noble Gases XOd.nC "I%I, 7i i~I~ ~gt gad TOnIAC(SC4a Dose-Particulate Form ln 8 3/4 11-2 8 3/4 11-3 8 3/4 11-3 Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System...........

8 3/4 11-4 Explosive Gas Mixture.

Main Condenser Venting or Purging.

8 3/4 11-4 8 3/4 11-5 8 3/4 11-5 3/4.11.3 SOLID RADIOACTIVE WASTE,......................

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8 3/4 11-5 3/4. 11. 4 TOTAL DOSE 8 3/4 11-5 SUSQUEHANNA - UNIT 1 xvi

LIST OF FIGURES INDEX FIGURE

3. l. 5-1

3. 1. 5-2 PAGE SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS........................

3/4 1-21 SODIUM PENTABORATE SOLUTION CONCENTRATION.........

3/4 1-22 3.2. 1-1 3.2. 1-2 3.2. 3-1 3.2.4.2-1 3..1. 1-t 3.4.6. 1-1 g.g.g. ~.-z.

Mls'~

B 3/4 3-1 B 3/4. 4. 6-1 5.1. 1-1 5.l. 2-1 5.1.3-1a 5.1. 3-lb 6.2. 1-1 6.2. 2-1 MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, GE FUEL TYPE BCR233 (2.33K ENRICHED)................

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE BUNDLE EXPOSURE, EXXON BxB FUEL.......................................

REDUCED FLOW MCPR OPERATING LIMIT...................

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE EXXON 8x8 FUEL..............

THERMAL POWER LIMITATIONS...........................

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE........................-..

gPgflCTSg VeE~Ct.

H6TA2 ~Q fCI2h'ElFIE 4 Y4 g&kTOll 46~<6< t~CSl3~

REACTOR VESSEL WATER LEVEL...

fAST NEUTRON FLUENCE (E>IMeV) AT 1/4 T AS A FUNCTION OF SERVICE LIfE....-.............

EXCLUSION AREA....................................

LOW POPULATION ZONE..;............................

MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS......................

OFFSITE ORGANIZATION..............................

UNIT ORGANIZATION.........;.......................

3/4 2" 3/4 2-/ 9 3/4 2/7 3/4 2-10 3/4 4-lb 3/4 4-18 s/+ 4-i9 8 3/4 3-8 8 3/4 4-7 5-2 5-3 5-4 5-5 6-3 6-4 SUSQUEHANNA -, UNIT. 1 xxi Amendment No.

57

LIST OF TABLES INDEX TABLE 1.2 2.2. 1-1 PAGE SURVEILLANCE FREQUENCY NOTATION...................

1-9 OPERATIONAL CONDITIONS................ -.....

1-10 REACTOR PROTECTION SYSTEM INSTRUMENTATION ETPOINTS e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~

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3. 2. 3-1 3.3. 1-1 MCPR OPERATING LIMITS FOR RATED CORE A 443........

REACTOR PROTECTION SYSTEM INSTRUMENTATION.........

3/4 3-2

3. 3. 1-2

4. 3.l. 1-1 I

'.3."2-1

'EACTOR PROTECTION SYSTEM RESPONSE TIMES..........

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.....................'....

ISOLATION ACTUATION INSTRUMENTATION...............

3/4 3-6 3/4 3-7 3/4 3-11 3.3. 2-2

3. 3. 2-3 4.3.2. 1-1 iSOLATION SYSTEM INSTRUMENTATION RESPONSE TIME....

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3"21 3/4 3-23 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS.....

3/4 3-17 31303 1

3.3.3-2 3.3. 3-3 4.3.3. 1-1

3. 3.4. 1-1 3.3.4. 1-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION............

EHERGENCY CORE COOLING SYSTEH ACTUATION INSTRUMENTATION SETPOINTS..............

EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES......

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........

AVOWS RECIRCULATION PUMP TRIP SYSTEM INSTRUHENTATION..............................'.....

ATWS RECIRCULATION PUMP TRIP SYSTEH INSTRUMENTATION SETPOINTS.........................

3/4 3-28 3/4 3-31 3/4 3-33 3/4 3"34 3/4 3-37 3/4 3-38 SUSQUEHANNA - UNIT 1 XX11 Amendment No.

29

LIST OF TABLES Continued INDEX TABLE 4.3.7.4-1 3.3.7. 5-1 4.3.7. 5-1 3.3.7. 9-1 3.3.7. 10-1 4.3.7. 10"1 3.3.7. 11-1 4.3.7.11-1 3.3.9-1

3. 3. 9"2

4. 3 ~ 9. 1-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREHENTS......................,,.

ACCIDENT MONITORING INSTRUMENTATION...............

ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.....................................,

FIRE DETECTION INSTRUMENTATION....................

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION..........................,,,......

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUHENTATION SURVEILLANCE REQUIREMENTS.........

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION..........................

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION..................................

FEEDWATER/MAIN TURBINE TRIP SYSTEH ACTUATION INSTRUMENTATION SETPOINTS.........................

FEEDWATER/HAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

PAGE 3/4 3"69 3/4 3-71 3/4 3-73 3/4 3"78 3/4 3"8$ ~

3/4 3-SP +

3/4 3-8P 7 3/4 3-9f 0 3/4 3"96 3/4 3"97 3/4 3"98 3.4.3. 2-1 3.4. 4-1 4.'4. 5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM...................................

3/4 4-15 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...

3/4 4-9 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............

3/4 4"12 4.4.6.l. 3-1 3.6.3"1 3.6.5.2-1 3.7.6. 5-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-"

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

PRIMARY CONTAINMENT ISOLATION VALVES...............

SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS o

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FIRE HOSE STATIONS 3/4 4-20 3/4 6-19 3/4 6"33 3/4 7-Q 2+

SUSQUEHANNA - UNIT 1 Amendment No.

29

0

LIST OF TABLES Continued INDEX TABLE 4.8. l.1. 2-1 4.8. 1.l. 2-2 4.8.2. 1-1.

3.8.4. 1-1

3. 8.4. 2"1

3. 11. 1. 1-1

4. 11. l.l. 1-1 DIESEL GENERATOR TEST SCHEDULE....................

UNIT 1 AND COMMON DIESEL GENERATOR LOADING TIMERS..

BATTERY SURVEILLANCE RE(UIREMENTS.................

PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.;..................

MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION.......'.................................

MAXIMUM PERMISSIBLE CONCENTRATION OF DISSOLVED OR ENTRAINED NOBLE GASES RELEASEO FROM THE SITE TO UNRESTRICTED AREAS IN LIQUID WASTE................

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS ROGRAM a

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P PAGE 3/4 8-7 3/4 8-8 3/4 8-14 3/4 8-24 3/4 8-29 3/4 11-2 3/4 11-3 4.11.2.1.2"1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS ROGRAM o o

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P 3/4 11-10

3. 12. 1"1 3+12.1 2

REPORTING LEVELS FOR RADIOACTIVITYCONCENTRATIONS IN ENVIRONMENTAL SAMPLES........:.................

3/4 12"9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.....

3/4 12-3

4. 12. 1-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE, A

'I NALYSIS

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3/4 12-10 B3/4.4.6"1

5. 7. 1-1

6. 2. 2-1 REACTOR VESSEL TOUGHNESS..........

COMPONENT CYCLIC OR TRANSIENT LIMITS..............

MINIMUM SHIFT CREW COMPOSITION....................

B 3/4 4-6 5-8 6"5 SUSQUEHANNA - UNIT 1 XXV Amendment No.

29

'h 0

)

1. 0.

OEF INITIONS The following terms are defined so that uniform interpretation of.hese specifications may be achieved.

The defined terms apoear in capi alized type and shall be applicable throughout these Technical Specifications.

ACTION

1. 1 ACTION shall be that part of a Specifics ioo <<hich pr.scribes reroecial measures required under designated chandi ions..

RAGE BUNOLE E

URE The AGE BUNOLE EXP RE shall be ual to the sum the axially eraged exposure all the fuel s in the spe 'ed bundle di 'd by the num of fuel rods in e fuel bundle.

AVER PLANAR EXP RE 1.2 The VERAGE PLANA XPOSURE sha be applicable a specific p -nar nesg and is equ to the sum o

the exposu of all the fue ods in the s

cified b die at the sp ified height ivided oy t umber of fuel ds in.he fu bundle AVERAGE PLANAR LINEAR HEAT'ENERATION RATE le3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of he LINEAR HEAT GENERATION RATES for all the fuel rods in the. soeci fied bundle at the specified height divided by the number of fuel rods in the fuel bundle:,

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, os the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK le5 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 indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST

.1e6 A CHANNEL FUNCTIONAL TEST shall be:

a. 'nalog channels

- the injection of a simulated sional into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b.

Bistable channels

- the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

SUSQUEHANNA - UNIT 1 1-1 Amendment No.. 45

1. 0 OEF INITIONS The following terms are defined so that uniform interpretation of these

'specifications'ay be achieved.

The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION l.I ACTION shai'I be that part of a Specification which prescribes remedia]

measures required under designated conditions.

AVERAGE 7gggt EXPOSURE I. 8 The AVERAGE BUNQLE EXPOSURE shall be qual to the suoi of the axia??y averaged eSIposure of all the fuel rods in the specified bundle divided by tne niaoer of fuel rods in the fuel bundle

The AVERAGE PLANAR EXPOSURE shall be aopl icable to a specific'lanar n ight and is equal to the suoi of the exposure of all he fuel rods in he soeci i?ed bund?e at the specified height divided by he nunber of fuel rods in tne fue:. 3uncle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT 'GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATTON RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1.4

'A'.CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary T'ange and accuracy to known values of the parameter which the channel monitol's.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observati'on.

This determination shall include, where

possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

'HANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a'simulated signal into the channel as close to the sensor is practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b.

Bistable channels - the injection of a simulated signal into the sensor, to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

SUSQUEHANNA - UNIT

3/4. 0 APPLICASILITY LIMITING CONDIT'ION FOR OPERATION 3.0.

1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals.

If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the

'equirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the Specification does not apply by placing it, as applicable, in:

1.

At least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limitinq Condition for Operation.

Exceptions to these requirements are stated in the >ndividual Specifications.

This specification is not applicable in OPERATIONAL CONDITION 4 or 5.

3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements.

This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements.

Exceptions to these requirements are stated in the individual Specifications.

SUSQUEHANNA " UNIT 1 3/4 0-1

APPLICABILITY SURVEILLANCE RE UIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CQNOITIONS or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

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

a.

A maximum allowable extension not to exceed 25K of the surveillance interval, but b.

The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a

Limiting Condition for Operation.

Exceptions to these requirements are stated in the individual Specifica ns.

Surveillance requirements do not have to be performed on inoperable equipmen 4.0.4 Entry into an OPERATIONAL CONDITION or othe~ specified applicable condition shall not be made unless the Surveillance Requirement(s}

associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, 4 3 components shall be applicable as follows:

Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g}, except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g}(6}(i}.

b.

Surveillance intervals specified in Section NI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing ins ection and testin activities activities Meekly At least once per 7 days Monthly At least once per 31 days quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days SUS(UEHANNA - UNIT 1 3/4 0-2

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE for GE fuel and AVERAGE BUNDLE EXPOSURE for Exxon fuel shall not exceed the limits shown in Figures 3.2. 1-1,EEnA 3.2.>-2. +

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or E

ETEE THERM E

0 EE..

ACTION:

With an APLHGR exceeding the limits of Figure 3 2. 1-1, or 3.2.1-2, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to 1ess than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2.3.-l and 3.2. 1-2:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

"See Specification 3.4. 1. 1.2.a for single loop operation requirements.

SUS(UEHANNA - UNIT I 3/4, 2-1 Amendment No.

57

THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3/4

-2 Amendment Np.

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FIGURE 3.2.i-1 O

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MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE

~

EXXON 8X8 FUEL FIGURE 3.2.1-2 SUSQUEHANNA - UNIT 1 3

vs z-/

Amendment No.

POWER DISTRIBUTION LIMITS 3/4.2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

Trio Setpoint Allowable Value S

SW + 59K)T S

.SSW +

2X)T SRB < (0.58W + 50K)T SRB

< (0.58W + 535)T where:

S and SRB are in percent of RATED THERMAL POWER, W

= Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr, T

= Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER OENSITY.

T is always less than or equal to 1.0.

APPLICABILITY:

OPERATIONAL CONDITION I, when THERMAL POWER is greater than or o

ATRO TIIERMAL Ell.

ACTION:

With the APRN flow biased simulated thermal power upscale scram trip setpolht and/or the flow biased neutron flux-upscale control rod block trip setpoint

'less conservative than the value shown in the Allowable Value column for S or SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or S

to be consistent with the Trip Setpoint value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL PNER to less than 25'f RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux"upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at

'east 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

ith M greater than the FRTP during power ascension up to 90K of RATED THERMAL.POWER, rathe~ than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to lOOX times

MFLPD, provided that the adjusted APRM reading does not exceed 100K of RATED,THERMAL POWER, the required gain adjustment increment does not exceed 10K of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

See Specification 3.4.1.1.2.a for single loop operation requirements.

/

SUS)UEHANNA - UNIT 1 3/a 2-pg Amendment No.

57

POWER DISTRIBUTION LIMITS 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO LIHITING CONOITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be:

a.

b.

greater than or equal to the applicable HCPR limit determined from Table 3.2.3-1 during steady state operation at rated core flow, or greater than or equal to the greater of the two values determined from Table 3.2.3-1 and Figure 3.2.3-1 during steady state operation at other than rated core flow.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or ir ril 'rr ir.

ACTION:

r With MCPR less than the applicable MCPR limit determined from Table 3.2.3-1 and Figure 3.2.3-1, initiate corrective action within 15 minutes and restore MCPR to within.the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25 of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

't SURVEILLANCE REOUIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or equal to the applicable MCPR limit determined from Table 3.2.3-1 and Figure 3.2.3-1:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at Ieast 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

~

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~

I fC froVASlo~h oY S~ecikic. ~ 't<i~ +to. 0 oN <C. re Qhhpggjgg SUSQUEHANNA - UNIT I 5'/4 2-P Amendment No.

4S

THIS. PAGE INTENTIONALLY LEFT BLANK.

'USQUEHANN

- UNIT 1 3/

Amendment No. 45

jllUIPMENT STA1III TABLE 3.2.3-1 MCPR OPERATING LIMITS FOR RATED CORE FLOW MCPR OPERATING LIMIT 1.

EOC-RPT and Main Turbine Bypass

OPERABLE, RBM setpoint

< 108K 2.

EOC"RPT Inoperab1e, Main Turbine Bypass

OPERABLE, RBM setpoint

< 108Ã 3.

Main Turbine Bypass Inoperable, EOC-RPT OPERABLE, RBM Setpoint

< 108K 4.

EOC-RPT and Main Turbine Bypass

OPERABLE, RBM Setpoint

< 106K 5.

EOC-RPT InoperabIe, Main Turbine Bypass

OPERABLE, RBM Setpoint

< 106K II 6.

Main Turbine Bypass Inoperab1e, EOC-RPT OPERABLE, RBM Setpoint

< 106K

1. 29 I. 33

1. 29

1. 25

1. 33

1. 26 SUSQUEHANNA - UNIT 1 3/4 2-f(

Amen4asnt No. 57

1.6 1.4 ACCEPTABLE REGION OF OPERATION E

1.3 C

~ wa CL

.0 1.2 Q.

40 60 00 70 00 eo Total Core Flow (% of Bated) 100 REDUCED FLOW HCPR OPERATIHG LIHIT FIGURE 3.2.3-1

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE OE FUEL LIMITING CONDITION FOR OPERATION 3.2.4.1 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed 13.4 kw/ft.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

~IIII IIIEE I

Ell ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.4.1 LHGRs for GE fuel shall be determined to be equal to or less than the 1 imit'.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUS(UEHANNA -UNIT 1 8

3/4 2-Amendment No.57

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE ENC FUEL LIMITING CONOITION FOR OPERATION 3.2.4.2 The LINEAR HEAT GENERATION RATE (LHGR) for ENC fuel shall not exceed the LHGR limit determined from Figure 3.2.4.2-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or f HATEO TIIERMAL 0

KR.

ACTION:

With the LHGR of any fuel rod 'exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.4.2 limit:

a.

b.

c LHGRs for ENC fuel shall be determined to be equal to or less than the At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL R00 PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-5a, Amendment No. 57

O

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10000 20000 30000 40000 60000 Average Planar Exposure (MWD/MT)

LINEAR HEAT GENERATION RATE (LHGB) LIMIT VERSUS AVERAGE PLANAR EXPOSURE EXXON 8X8 FUEL FIGURE 3.2.4.2-'I EXP LHGR 0

16.82 600 '16.82 2,600 16.10 6.070 14.71 7,730 14.19 10,290 14.13 13.090 14.06 16,910 14.06 18.760 14.00 21,690 13.93 24,420 13.93'7,28013.08 30,160 12.24 33.060 11.40 35.960 10A7 38,900.

9.66 41,830 8.66 44,760 7.77

TABLE 3.3.1-1 (Continued)'EACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveil1ance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

This function is automatically bypassed when the reactor mode s~itch is in the Run position.

r (c)

The "shorting links" shall be removed from the RPS circuitry prior to and du~ing the time any control rod is withdrawn" and shutdown margin demonstrations performed per Specification

3. 10.3.

(d)

The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the "shorting links" are

removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMS and 6 IRMS.

(e)

An APRM channel is inoperable if there are less than 2

LPRH inputs per level or less than 14 LPRH inputs to an APRH channel.

(f)

This function js not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(g)

This. function is automatically bypassed when the reactor mode switch is not in the Run position.

(h)

This function is not required to be.OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i)

With any control rod withdrawn.

(j)

This function shall be automatically bypassed when turbine first stage pressure is less than 108 psig or 17K of the value of first stage pressure in psia at valves wide open (Y.W.O) steam flow, equivalent to THERMAL POWER of about 24K of RATED THERMAL POWER.

(k)

Also actuates the EOC-RPT system.

ot required or control rods removed per Specification 3.9.10.1 or 3.9.10.2.

SUS)UEHANNA - UNIT 1 3/4 3-5 Amendment No. 9

TABLE 3.3.2-1 ISOLATION ACTUATION INSTRNENTATION TRIP FUNCTION ACT ON 1.

PRIHARY CONTAIISENT ISOLATION a.

Reactor Vessel Mater Level 1)

Low, Level 3 2)

Low Low, Level 2 3)

Low Low low, Level 1 1, 2, 3

1, 2, 3

20 20 20 b.

Drywell Pressure - High c.

Hanual Initiation 1>2>3 1, 2, 3 20 Y,Z,X 3 4AAA 5&A 20 1, 2, 3 20 d.

SGTS Exhaust Radiation-High R

e.

Hain Steaa Line Radiation-High C

2.

SECONDARY CONTAINHENT ISOLATION Reactor Vessel Mater level-Low Low, Level 2 Drltwell Pressure - High Refuel Floor High Exhaust Duct Radiation - High Railroad Access Shaft Exhaust Duct Radiation - High Refuel Floor Mall Exhaust Duct Radiation - High Hanual Initiation ao 1, 2,3and*

1, 2, 3

25 25 b.

c 25

0 d.

25 e.

~

25 24

0 1, 2, 3 and

~

HININN APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL SIGNAL s PER TRIP SYSTN b

CONDITION

TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUHENTATION SETPOINTS TRIP FUNCTION l.

PRIHARY CONTAINHENT ISOLATION a.

Reactor Ve'ssel Mater Level 1)

Low, Level 3 2)

Low Low, Level 2 3)

Low Low Low, Level 1

b.

Orywell Pressure - High c.

Hanual Initiation d.

SGTS Exhaust Radiation - High e.

Hain Steam Line Radiation - High 2.

SECONOARY CONTAINHENT ISOLATION TRIP SETPOINT

> 13.0 inches*

> -38.0 inches*

> -129 inches*

< 1.72 psig NA

<23.0 mR/hr

< P x full power background 7.b ALLOWABLE VALUE

> ll.5 inches

> -45.0 inches

> -136 inches

< 1.88 psig NA

<31.0 mR/Hr x full power background s.+

ao b.

C.

d.

e.

Reactor Vessel Mater Level-Low Low, Level 2 Orwell Pressure " High Refuel Floor High Exhaust Ouct Radiation - High Railroad Access Shaft Exhaust Ouct Radiation " High Refuel Floor Wall Exhaust Ouct Radiation - High Hanual Initiation

> -38.0 inches*

< 1.72 psig

< 2.5 mR/hr~

< 2.5 mR/hr.~

< 2.5 R/hr~

NA h

hali

> -45.0 inches

< 1.88 psig

< 4.0 mR/hr~

NA 3.

HAIN STEAH LINE ISOLATION

'a o b.

C.

d.

Reactor Vessel Water Level -Low Low, Level I

Hain Skean Line Radiation - High Hain Steam Line Pressure - Low Hain Steam Line Flow - High

> -129 inches~.

< ).O.X full power background

> 861 psig

< 107 psid

> -136 inches

< 8.4 X full power background

> &41 psig

< 110 psid

TRIP FUNCTION TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRNENTATIDN SETPOINTS TRIP SETPOINT ALLOWABLE VALUE HAIN STEAH LINE ISOLATI (Continued)

~ e.

Condenser Vacua - Low

> 9.0 jnches Hg vacuuw

> 8.8 inches.Hg vacuujm h.i.

Reactor Building Hain Steam Line

- Tunnel Tewperature - High Reactor Building Hain Steaw Line Tunnel h Tewperature - High Hanual Initiation Turbine Building Hain Steaa Line Tunnel Tewperature-High

< 177'F

< 99'F NA

<177 F

< 184 F

< 108 F NA

<184 F

h.

REACTOR WATER CLEANUP SYSTEH ISOLATION 5.

< 60 gpa

< 147 F or 118.3 FN

< 694F or 35.34'A

> "38 inches"

< 426 gpa NA a.

SKU d Flow" High b.

'RWCU Area Temperature - High c.

RMCU/Area Ventilation h Teaperature - High d.

SLCS Initiation e.

Reactor Vessel Mater level-Low Low, level 2 f.

RWCU Flow - High g.

Hanual Initiation REACTOR CORE ISOLATION.COOLING SYSTEH ISOLATION

< 80 gpa

< 154 F or 125.3 FN

< 78OF or 44.3 FN NA

> -45 inches

< 436 gpa NA a.

b.

C.

RCIC Steaa Line h Pressure - High RCIC Steaw Supply Pressure - Low RCIC Turbine Exhaust Diaphragw Pressure - High

< 177" H,~

> 60 psig

< 10.0 n~ig

< 189" H>O

> 53 psig

< 20.0 psig

6

AD m

TRIP FUNCTION TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUHENTATION SETPOINTS TRIP SETPOINT ALLOMABLE VANE REACTOR CORE ISOLATION COOLING SYSTEH ISOLATION Continued) d.

e.

g.

h.

1

~

3.

RCIC Equipment Room Temperature - High RCIC Equipment Room A Temperature - High RCIC Pipe Routing Area Temperature - High RCIC Pipe Routing Area h Temperature

.-. High RCIC Emergency Area Cooler Temperature -.High Hanual Initiation Drywell Pressure - High

( 1670F<g

< 89'F

< 167 F

<89F

<147F NA

<1.72 psig

< 174 F""

<98 F

< D4'F" FH

< 154 l.

NA

<1.88 psig O

CA 6.

a ~

b.

C.

d.

e.

HPCI Steam Line Flow - High HPCI Steam Supply Pressure

- Low llPCI Turbine Exhaust Diaphragm Pressure

- High HPCI Equipment Room Temperature High HPCI Equipment Room h, Temperature

- High HPCI Emergency Area Cooler Temperature High HPCI Pipe Routing Area Temperature - lligh HIGH PRESSURE COOLANT INJECTION 5YSTEH ISOLATION

< 350 inches HqO~

> 104 psig

< 10 psig

<167F

<&9F (147 F

< 167 F""

c 367 inches Ilzhe I ~

90 psig

< 20 psig

< 174 F

< 9& F

<154'F D4()FH

Y CA Cm TRIP FUNCTION TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS TRIP SETPOINT ALLOWABLE VALUE M

h.

HPCI Pipe Routing Area h Temperature - High Hanual Initiation Drywell Pressure - High

< 89 F

NA

<1.72 psig 98~F58 NA

<1.88 psig 7.

RHR SYSTEM SHUTDOWN COOLING/HEAD SPRAY NODE ISOLATION a.

Reactor Vessel Water Level-Low, Level 3

b.

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High c.

RHR Equipment Area h, Temperature - High d.

RHR Equipment Area Temperature-High e.

RHR Flow - High f.

Hanual Initiation g.

Drywell Pressure

- High

> 13.0 inches"

< 98 psig

< 89'~

< 167'P

< 25,000 gpm NA-

<1.72 psig

> 11.5 inches

< 108 psig

< 90.5'F

< 170.5 ~

< 26,000 gpm NA

<1.88 psig ee Bases lgure B 3/4 3-l.

O Ch NLower setpoints for TSH-G33-N600 E, F and TSH-G33-N602 E; F.

NII15 minute time delay.

B E4.

FEEOMATER/HAIN lURBINE TRIP SYSTEH ACTUATION INSTRUHENTATION SURVEILLANCE RE UIREHENTS FUNCTIONAL UNIT a.

Reactor Vessel Mater Level-High CHANNEL CHECK CHANNEL FUNCT10NAL TEST OPERATIONAL CHANNEL CONDITIONS FOR MHICN CALIBRATION SURVEILLANCE RE UIRED R

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed

< 80K of the rated pump speed, and a.

the following revised specification limits shall be followed:

1.

Specification 2. 1.2:

the MCPR Safety Limit shall be increased to 1.07.

2.

Table 2.2.1-1:

the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

Tri Set oint

< 0.58W +

5 Allowable Value 3.

Specification 3.2.1:

The MAPLHGR limits shall be as follows:

OE t:.

1i i p

I i i

Hg /3.2.11~

multiplied by 0.81.

I b.

Exxan fue1:

the limits specii'ied in Figure 3.2.1$ muitiplied by 0.81.

4.

Specification

3. 2.2:

the APRM Setpoints shall be as follows:

5.

Table 3.3.6-2:

follows:

Tri Set oint S

0.58W + 55K)T SRB

< (0.58W + 46K)T the RBM/APRM Control Rod Block Allowable Value S

0.58W + 58K)T SRB 0'58W + 49~ T Setpoints shall'e as a..

RBM - Upscale Tri Setgoint Al 1 owabl e Value 1.

< 0.66W + 35 2.

< 0.66W + 37~

0.66W + 40K 5.a. 1 and 5. a.2 shall be used in conjunction with the MCPR limits specified in Table 3. 2.3-1 for RBM Setpoints of'06K and 108K, respectively.

b.

APRM-Flow Biased Tri Set oint Allowable Value

< 0.

+46

'""'.ll b.

APRM and LPRM""" neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4.1.1.1-1.

c.

Total core flow shall be greater than or equal to 42 millio '/hr when THERMAL POWER is greater than the limit specified in Figur 3g.1.1.1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2", except during loop operation.¹ SUS(UEHANNA - UNIT 1 3/4 4-1c Amendment No.

56

3/4,5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.l ECCS - OPERATING LIMITING CONDITION FOR OPERATION

'.5.

1 The emergency core cooling systems shall be OPERABLE with:

a.

The core spray system (CSS) consisting of two subsystems with each subsystem comprised of:

1.

Two OPERABLE CSS pumps, and 2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

b.

The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of two subsystems with each subsystem comprised of:

1.

Two OPERABLE LPCI pumps, and 2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

c.

The high pressure cooling injection (HPCI) system consisting of:

1 ~

One OPERABLE HPCI pump, and 2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

d.

The automatic depressurization system (ADS) with six OPERABLE ADS valves.

APPLICABILITY:

OPERATIONAL CONDITION 1, 2.",*",¹, and 3","*,¹¹.

~

'he HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.

"*The ADS is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

¹See Special Test Exception

3. 10.P'.

¹¹One LPCI subsystem of the RHR system may be inoperable in that it is aligned in the shutdown cooling mode when reactor vessel pressure is less than the RHR shutdown cooling permissive setpoint.

SUSQUEHANNA " UNIT 1 3/4 5-1

CONTAINMENT SYSTEMS ORYWELL AVERAGE AIR TEMPERATURE LIMITING CONOITION FOR OPERATION 3.6. 1.7 Drywell average air temperature shall not exceed 135'F.

APPLICABILITY:

OPERATIONAL CONOITIONS 1, 2 and 3.

ACTION:

With the drywe11 average air temperature greater than 135'F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTOOWN within the'following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

'URVEILLANCE RE UIREMENTS areas 4.6.1A7 The drywell average air temperature shall be the arithmetical average of the higher temperature at a minimum of 3 of the following and shall be determined to be within the limit at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

A~~

Tu~

Mi44Le.

C Pe~a.s t

Elevation 797'8" 752'2" 737'11'r 720'zimuth 110 295 90 i

270'50 300 270 85 SUS(UEHANNA - UNIT 1 3/4 6"10 Amendment No. 36

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION Continued CP C

I 0

~r Vl Vt C-CP O dP s ~aEdl tO M Vl 0 S-Ct CP

~r C

tO 0 3 C

CP tO Ol Vl dl) tO Ol tO S

CP Cl

~ r

$-N OP E O

tO CP C 0

tt M C tO 0 Vl >>~p

~

dl Vl tt O. Vl CP CIP CD C

CP I

~rP Qrl E

C D O CD dP Vl M tO Olt I 3 C tO CPO CP )t 0'l te Cl CP IO 'r CP I I-Cl dP' 0 ) )

tO Vl I-DN tO dl CP DQ CP Vl tO CP Vl ADO CP

& Cll e>>

C.

d.

3.

Mith the suppression chamber average water temperature greater than 120'F, depressurize the reactor pressure vessel to less.

than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With only one suppression chamber water level indicator OPERABLE and/or with less than eight suppression pool water temper'ature indicators covering at least six locations OPERABLE, restore the inoperable indicatort,'s) to OPERABLE'status within 7 days or verify suppression chamber water level and/or temperature to be within the'imits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Mith no suppression chamber water level indicators OPERABLE and/or with less than one suppression pool water temperature indicator at at least six different locations OPERABLE, restore at least one water level indicator and at least one water temperature indicator at at least six different loca'tions to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e.

Mith the drywell"to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200'F.

SURVEILLANCE RE UIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:

a.

By verifying the suppression chamber water volume to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be less than or equal to 90OF; except:

l.

At least once per 5 minutes during testing which adds heat to the suppression

chamber, by verifying the suppressio'n chamber average water temperature less than or equal to 105'F.

2.

At least once per hour when suppression chamber average water temperature is greater than or equal to 90'F, by verifying:

a)

Suppression chamber average water temperature to be less than or equal to 110 F, and b)

THERMAL POWER to be less than or equal to 0'f RATED THERMAL POWER after suppression chamber average water temperature has exceeded 90'F for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION:

(Continued) b)

110'F, place the reactor mode switch in the Shutdown posi-tion and operate at least one residual heat removal loop in the suppression, pool cooling mode.

SUS(UEHANNA - UNIT 1 3/4 6"13 Amendment No.

36

0

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAIHMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6.5. 1 SECONDARY CONTAINMENT INTEGRITY"" shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and ".

ACTION:

Without SECONDARY CONTAINMENT-"'NTEGRITY:

a.

In OPERATIONAL CONDITION 1, 2,.or 3, restore SECONDARY CONTAINMEHT INTEGRITY nothin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In Operational Condition ", suspend handlinq of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 6.5. 1 ao b.

SECONDARY CQNTAINMEHT INTEGRITY shall be'demonstrated by:

Verifying at, least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the secondary contaimment is 4eee than or equal to 0.25 inch of vacuum water gauge.

QI CC68l Verifying at least once per 31 days that:

1a.

When the railroad bay door (No. 101) is closed; all Zone I and III hatches removable walls, dampeg, and doors connected to the railroad access bay are closed, or

')

~Onl Zone I removable ills and/or ddors are open to the raa 1 road access

shaft, or ii)

~0nl Zone III hatches gd/or dampers are open to the railroad access shaft.

1b.

When the railroad bay door (No. 101) is open; all Zone I and III

hatches, removable walls, dampers, and doors connected to the railroad access bay are closed.

"When irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

""Secondary Containment consists of Zone I, Zone II and Zone III or Zone I and Zone III when Zone IIis isolated from Zone I and Zone III.

~

Personnel ingress and egress through doors within the secondary containment is not prohibited by.his specification.

SUSqUEHANHA UNIT 1

3/4 6-31 AMEHDMEHT HO. 21

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued c

d.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber

housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:

l.

Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than

'.05K and uses the test procedures of Regulatory Positions

. C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 10,100 cfm i 10K.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 0.175K; and 3.

Verifying a subsystem flow rate of 10,100 cfm k 10K during system operation when tested 'in accordance with ANSI N510-1975.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 0.175K.

At least once per 18 months by:

l.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 13 inches Water Gauge while operating the filter train at a flow rate of 10,100 cfm t lOX.

2.

Verifying that the filter train starts and associated dampers'pen on each of the following test signals:

a.

Manual initiation from the control room, and b'.

Simulated automatic initiation signal.

3.

Verifying that the filter cooling bypass and outside air dampers open and the fan start on filter cooling initiation.

4.

Verifyin that the temperature differential across each heating coil i 7'F when tested in accordance with ANSI N510-1975.

SUSQUEHANNA - UNIT 1 Q 6-35 Amendment No.

35

0

CONTAINMENT SYSTEMS ORYWELL AIR FLOW SYSTEM LIMITING CONOITION FOR OPERATION 3.6.6.f Orywel1 unit ecol'er fans 1'/414 ABB, 1'/416 AM and recirculation rane lV418 A88 sha11 be OPERABLE at low speed.

APPLICABILITY:

OPERATIONAL CONOITIONS 1 and 2.

ACTION:

With one.or more of the above fans inoperab1e at 1ow speed, restore the inoperable fan(s) to OPERABLE status within 30 days or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.6;6.g Each of the fans required above sha11 be demonstrated OPERABLE at

. 1east once per 92. days by:

a.

Starting each fan at low speed from the contro1

room, and b.

Verifying that each fan operates for at 1east 15 minutes.

SUSQUEHANNA - UNIT1'/4 6-38 Amendment No. 46

CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 9

3.6.6./

The dryrrell and suppression chamber atmosphere oxygen concentcatron shall be less than 4X by volume.

APPLICABILITY:

OPERATIONAL CONDITION lK, during the time period:

s a.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15K of RATED THERMAL POWER, following startup, to b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing THERMAL POWER to less than 15X of RATED THERMAL POWER preliminary to a scheduled reactor shutdown.

ACTION:

With the oxygen concentration in the drywell and/or suppression chamber exceeding the limit, restore the oxygen concentration to within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE RE UIREMENTS 9

4.6.6.g The oxygen concentration in the drywell and suppression chamber shall be verified to be within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15K of RATED THERMAL POWER and at least once per 7 days thereafter.

SUSQUEHANNA - UNIT 1 3/4 6-39

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM LIHITING CONDITION FOR OPERATION 3.7. 1.2 Two independent emergency service water system loops shall be OPERABLE with each loop comprised of:

a.

Two OPERABLE emergency service water pumps, and b.

An OPERABLE flow path capable of taking suction from the spray pond and transferring the water to the associated safety related equipment.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and ".

ACTION:

'a 0 In OPERATIONAL CONDITION 1, 2, or 3:

oC 1.8 With one emergency service water pump inoperable, restore he inoperable pump to OPERABLE status within 7 days or be in least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With two emergency service water pumps inoperable, restore at least one inoperable pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With one emergency service water system loop otherwise inoperable, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be fn at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and fn COLO SHUTDOWN within the following

24. hours.

b.

In OPERATIONAL CONDITION 4, 5 or *:

1.

With one pump in an emergency service water system loop inoper-able, verify adequate cooling capability remains available for the diesel generators required to be operable by Specifica-tion 3.8. 1.2 or declare the affected diesel generator(s) inoper-able and take the ACTION required by Specification 3.8. 1.2.

2.

With two pumps in an emergency service water system loop inoper able or with the loop otherwise inoperable declare the associ-ated safety related equipment inoperable (except diesel gene-rators),

and follow the applicable ACTION statements.

Verify adequate cooling remains available for the diesel generators required to be operable by Specification 3.8.1.2 or declare the affected diesel generator(s) inoperable and take the ACTION required by Specification 3.8. 1.2.

"When handling irradfated fuel in the secondary containment.

Ahen any diesel generator is removed from service in order to do work asso-ciated with tying in the additional diesel generator and its associated emergency service water pump is inoperable, Action a.l shall read as, follows:

a.1 Wfth one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status when its associated diesel generator is restored to OPERABLE status per Specification 3.8. 1. 1.

SUS(UEHANNA - UNIT 1 3/4 7"2 Amendment No. 51

PLANT SYSTEMS ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7. 1.3 The spray pond shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and ".

ACTION:

a.

With the groundwater level at any spray pond area observation well greater than or equal to 663'ean Sea Level (MSL), prepare and submit ~a a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the high groundwater level and the plans for restoring the level to within the limit.

b.

With the spray pond otherwise inoperable:

l.

In OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and the emergency service water system inoperable and take the ACTION required by Specifications 3.7.1. 1 and 3.7. 1.2.

3.

In Operational Condition ", declare the emergency service water system inoperable and take the ACTION required by Specif'ication 3.7.1.2.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE by verifying:

The average water temperature, which shall be the arithmetical average of the spray pond water temperature at the surface, mid and bottom

levels, to be less than or equal to 88'F at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l~ "

b.

C.

The water level at the overflow weir is greater than or equal to 678'1" 4>

MSL USGS, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The groundwater level at observation wells 1', 3, 4

, 6, and 1113 to be less than 663'SL at least once per 31 day.+(

3 "When handling irradiated fuel in the secondary containment.

SUS(UEHANNA " UNIT 1 3/4 7"4 Amendment No. 36

HALON SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.4 The Halon systems in the following panel modules with the storage tanks having at least 95K of full charge full charge pressure:

U 1)700 Ii7701 Iii702 llf703 l)704 lf06 ljf730 IN731'i7732 lt APPLICABILITY:

Whenever equipment protected by the Halon E.

ACTION:

shall be OPERABLE weight and 9'f tl ag7O5 systems is required a.

With one or more of the above required Halon systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly Ip fire watch patrol.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.6.4 Each of the above required Halon systems shall be demonstrated OPERABLE.

a.

At least once per 31 days by verifying Chat each valve, manual power operated or automatic, in the flow path is in its correct position.

b.

At least once per 6 months by verifying Halon storage tank weight and pressure.

c.

At'least once per 18 months by:

2.

Performance of a flow test through accessible headers and nozzles to assure no blockage.

Performance of a functional test of the general alarm circuit and associated alarm and interlock devices.

SUS(UEHANNA - UNIT 1 3/4 7"22 Amendment No.

36

AWO TABLE 4.8.1.1.2"2 UNI 1 UNIT 2 DIESEL mDEEEEDD I Il EE OEVICE TAG NO.

62A-20102

.62A-20202 62A"20302 62A"20402 SYSTEM RHR Pump 1A RHR Pump 1B RHR Pump 1C RHR Pump 10 LOCATION 1A201 1A202 1A203 lA204 TIME SETTING 3 sec 3 sec 3 sec 3 sec 62A-.20102 62A"20202 62A"20302 62A"20402 RHR Pump 2A RHR Pump 2B RHR Pump 2C RHR Pump 20 2A201 2A202 2A203 2A204 3 sec 3 sec 3 sec 3 sec K116A K116B K125A K125B CS pp IA CS pp 1B CS pp 1C CS pp 10 1C626 1C627 1C626 1C627 10.5 sec 10.5 sec 10.5 sec'0.5 sec K116A K116B K125A K125B CS pp 2A CS pp 2B CS pp 2C CS pp 20 2C626 2C627 2C626 2C627 10.5 sec 10.5 sec 10.5 sec 10.5 sec 62AX2-20108 62AX2"20208 62AX2-20303 62AX2-20403 Emergency Service Water (ESW)

Emergency Service Water (ESW)

Emergency Service Water (ESW) 1A201 1A202 1A203 1A204 40 sec 40 sec 44 sec 48 sec 62X3-20304 62X3"20404 62X-20104 Control Structure Chilled Water System Control Structure Chilled Water System Emergency Switchgear Rm.

Cooler A 8I RHR SW pp H8V Fan A

OC877A OC8778 OC877A 60 sec 60 sec 60 sec SUSQUEHANNA " UNIT 1 3/4 8-8 Amendment No. 36

ELECTRICAL POWER SYSTEMS A.C.

SOURCES -

SHUTDOWN

~

LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a.

One circuit between the offsite transmission network and the onsite Class 1E distribution system, and b.

Two diesel generators with:

1.

An engine mounted day fuel tank containing a minimum of 325 gallons of fuel.

2.

A fuel storage system containing a minimum of 47,570 gallons of fuel.

3.

A fuel transfer pump.

APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5 and ".

ACTION:

a.

b.

With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment, operations with a potential for draini'ng the reactor vessel and crane operations over the spent fuel pool when fuel assemblies are st&ed therein.

In

addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel

flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.

The provisions of Specifichtion 3.0.3 are not applicable.

SURVEILLANCE ENTS l

4.8.1..2 At least'he above required A.C. electrical power, sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1. 1.1, 4.8. 1.1.2 and 4.8. 1.1.3, except for the requirement of 4.8. 1.1.2.a.5.

)s Kng h~4ia hen srradiateb-fuel/in the secondary containment> ~ c4ui~

Coae AL~rrods ed oprot<oM &ltd o. ~ ticJ fir'drow,~ Ae. reooIor vo>>(

SUS(UEHANNA - UNIT 1 3/4 8-9

D.C.

SOURCES -

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 sources a.

b.

As a minimum, Division I or Division.II of the D.C. el shall be OPERABLE with:

Division I consisting of:

1.

Load group Channel "A" power source, consisting a) 125 volt DC battery bank b)

Full capacity charger 2.

Load group Channel "C" power source, consisting a) 125 volt DC battery bank b)

Full capacity charger 3.

Load group "I" power source, consisting of:

a) 250 volt DC battery bank b)

Half-capacity chargers 4.

Load group "I" power source, consisting of:

a) a 24 volt DC battery bank b)

Two half-capacity chargers Division II consisting of:

l.

Load group Channel "B" power source,, consisting a) 125 volt DC battery bank b)

Full capacity charger 2.

Load group Channel "0" power source, consisting a) 125 volt DC battery bank b)

Full capacity charger 3.

Load group "II" power source, consisting of:

a).

250 volt DC battery bank b)

Full capacity charger 4.

Load group "II" power source, consisting of:

a) 2 24 volt DC battery bank b)

Two half-capacity chargers ectrical power of:

10610, 20610**

10613,20613*"

of:

10630, 20630""

10633, 20633*"

10650 1D653A, 106538 10670

10673, 10674 of:

10620, 2D620""

10623) 20623""

of:

10640, 20640""

10643, 20643""

10660 10663 10680

10683, 10684 APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5, and ".

ACTION:

a.

W fess than the above required Unit 1 125 volt and/or 250 volt DC l

jpoup battery banks OPERABLE, suspend CORE ALTERATIONS, handling of-adiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

b.

With less than the above required Unit 2 125-volt DC load group battery banks OPERABLE, either:

(s be ~~ 4~lcd When Aaa444ay irradiated fuelgin the secondary containmentg, and c4rtng

"Not required to be OPERABLE when the requirements of ACTION b have been satisfied.

CoRE AL1EAAfl0$5 ~

a'flaratrav>

With a

(lottatrl4I 'Fal'i'a<a ag 'Hla taaatai VIAAE(.

SUSQUEHANNA - UNIT 1 3/4 8-15 Amendment No.

3l

ELECTRICAL POWER SYSTEHS LIMITING CONDITION FOR OPERATION IA201 1B210

08516, OB517

1B216, 18217 1Y216 4) 208/120-volt A.C. instrument panels b)

.Load group Channel "C", consisting of:

1) 4160 volt A.C. switchgear bus 2) 480 volt A.C. load center 3) 480 volt A.C. motor control centers 1A203 18230

OB536, OB136

1B236, 1B237 1Y236 1B219" 4) 208/120 volt A.C. instrument panels c}

Isolated 480 volt A.C. swing bus, including:

1)

Preferred power source 2)

Preferred power source HG set 3)

Alternate power source 4)

Automatic transfer switch 3.Q.3.2 As a minimum, the following power distribution system divisiops shall be energized:

a.

For A.C. power distribution, Division I or Division II with:

l.

Division I consisting of:

a)

Load group Channel "A" consisting of:

1) 4160 volt A.C. switchgear bus 2}

480 volt A.C. load center 3) 480 volt A.C. motor control centers 2.

Division II consisting of:

a)

Load group Channel "B", consisting of:

1) 4610 volt A.C. switchgear bus 2) 480 volt A.C. load center 3) 480 volt A.C. motor control centers 4) 208/120-volt A.C. instrument panels b)

Load group Channel "0", consisting of:

1) 4160 volt A.C. switchgear bus 2) 480 volt A.C. load center 3) 480 volt A.C. motor control centers 4) 208/120 volt A.C. instrument panels c)

Isolated 480 volt'A.C. swing bus, including I)

Preferred power source 2)

Preferred power source MG set 3)

Alternate power source 4)

Automatic transfer switch 1A202 1B220

,OBS26, OB527 IB226, 18227 1Y226 1A204 1B240

OB546, OB146 IB246, IB247 lY246 1B229""

~The swing bus shall be OPERABLE if the Division I LPCI subsystem alone is fulfillingthe requirements of Specification 3.5.2.

"The swing bus shall be OPERABLE if the Division II LPCI subsystem alone is fulfillingthe requirements of Specification 3.5.2.

SUSQUEHANNA - UNIT I 3/4 8-19 Amendment No. 55

ELECTRICAL POWER SYSTEMS REACTOR PROTECTION SYSTEM ELECTRIC POWER MONITORING

'LIMITING CONDITION FOR-OPERATION 3.8.4.3 Two RPS electric power monitoring assemblies for each inservice RPS MG set or alternate 'power supply shall be OPERABLE.

APPLICABILITY: At 'a'll times.

ACTION:

a.

With one RPS electric power monitoring assembly for an inservice RPS MG set or alternate power supply inoperable; restore the inoperable power monitoring assembly to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS MG set or alternate power supply from service.

b.

With both RPS electric power monitoring assemblies for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring assembly to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

SURVEILLANCE RE UIREMENTS 4.8.4.3 The above specified RPS electric power monitoring assemblies shall be determined OPERABLE:

Qn(+

a.

By performance of a CHANNEL FUNCTIONAL TEST each time the ~W is in COLO SHUTOOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed within the previous 6 months.

b.

At least once per 18 months by demonstrating the OPERABILITY of overvoltage, undervoltage and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints:

1.

Overvoltage 2.

Undervoltage 3.

Underfrequency RPS Oivision A

< 128.3 VAC

> 110.7 VAC""

> 57 Hz RPS Oivision 8

< 129.5 VAC

> 111.9 VAC""

> 57 Hz

""Initial setpoint.

Final setpoint to be determined during startup testing following the first refueling outage.

Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion.

SUS(UEHANNA - UNIT 1 3/4 8-33 Amendment No. 41

3/4. 9 REFUELING OPERATIONS 3/4.9.

1 REACTOR NODE SWITCH C

S LIHITING CONDITION FOR OPERATION 3.9.

1 The reactor mode switch shall. be OPERABLE and locked ig the Shutdown or Reft 1 position.

When the reactor mode switch is locked in the Refuel position:

a.

A control rod shall not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE.

b.

CORE ALTERATIONS shall not be performed using equipment associated with a'efuel position interlock unless at least the following associ-ated Refuel position interlocks are OPERABLE for such equipment.

l.

2.

3.

4.

5.

APPLICABILITY'll rods in.

Refuel platform position.

Refuel platform hoists fuel-loaded.

Fuel grapple position.

Service platform hoist fuel-loaded.

OPERATIONAL CONDITION 5" ACTION:

~

a.

With the reactor mode switch not locked in the Shutdown or Refuel position as specified, suspend CORE ALTERATIONS and lock the reactor mode switch in the Shutdown or Refuel position.

b.

With the one-rod-out interlock inoperable, lock the reactor mode switch in the Shutdown position.

~ c.

With any 'of the above required Refuel position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment interlock.

See Special Test Exceptions

3. 10. 1 and 3. 10. 3.

"i"=':II;'..".',n - UNIT 1 3/4 9-1 Amendment No.

SPECIAL TEST EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.4.1.1.1 and 3.4.1.3 may be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the performance of:

a.

PHYSICS TESTS, provided that THERMAL POWER does not exceed 5X of RATED THERMAL POWER, or b.

The Startup Test Program.

APPLICABILITY:,OPERATIONALCONDITIONS 1 and 2, during PHYSICS TESTS and the Startup Test Program.

ACTION:

a0 b.

With the above specified time limit exceeded, insert all control rods.

With the above specified THERMAL POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE REOUIREMENTS

4. 10.4. 1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s at least once per hour during PHYSICS TESTS and the Startup Test Program.

4.10.4.2 THERMAL POWER shall be determined to be less than 5X of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

SUSQUEHANNA - UNIT 1 3/4 10-4 4

Amendment No.

56

0

SPE AL TEST EXCEPTIONS 3/4. 10.

OXYGEN CONCENTRATION LIHITING CO TION FOR OPERATION 3.10.5 The provis ons of Specification 3.6.6.4 may be suspended during the performance of the artup Test Program until ither the required 100K of RATED THERNL POWER trip te s have been completed or the reactor has operated for 120 Effective Full Powe Days.

APPLICABILITY:

OPERATIONA CONDITION 1.

ACTION With the requirements of the abov specification not satisfied, be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREH TS 4.10.5 The Effe ive Full Power Days of operation. shall be verified to be less than 120, by c culation, at least once per 7 days during the Startup Test Program.

SUSQUEHANNA " UNIT 1 3/4 10-5

SPECIAL TEST EXCEPTIONS 3/4. 10.

TRAINING STARTUPS 5

LIMITING CONDITION FOR OPERATION

3. 1025The provisions of Specification 3.5.

1 may be suspended to permit one RHR subsystem to be aligned in the shutdown cooling mode during training startups provided that the reactor vessel is not pressurized, THERMAL POWER is less than or equal to 1X of RATED THERMAL POWER and reactor coolant temperature is less than 200'F.

APPLICABILITY:

OPERATIONAL CONDITION 2, during training startups.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE RE UIREMENTS

4. 10.jfSThe reactor vessel sha1 1

be verified to be unpr assurized and the THERMAL POWER and reactor coolant temperature'shall be verified to be within the limits at least once per hour during training startups.

SUS(UEHANNA - UNIT 1 5

3/4 10-jf

RADIOACTIVE EFFLUENTS VENTING OR PURGING LIMITING CONDITION FOR OPERATION 3.11.2.8 VENTING or PURGING of the Mark II containment drywell shall be through the Standby Gas Treatment System.

APPLICABILITY: Whenever the drywell is vented or purged.

ACTION:

ae With the requirements of the above specification not satisfied, suspend all VENTING and PURGING of the drywell.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIRMENTS 4.11.2.8.1 The containment drywall shall be de'termined to be aligned for VENTING

[

or PURGING through the Standby Gas Treatment System within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or PURGING of the drywel l.

4.11.2.8.2 Prior to use of the purge system through the standby gas treatment system assure that:

a.

Both standby gas treatment system trains are OPERABLE whenever the purge system is in use, and b.

Whenever the purge system is in use during OPERATIONAL CONDITION 1 or 2 or 3, only of the standby gas treatment system trains may be used.

one.

SUSquEHANNA - UND 1 3/4 11-19 Amendment No. 36

RADIOACTIVE EFFLUENTS 3/4. 11.3 SOLID RADWASTE SYSTEM LIMITING CONDITION FOR OPERATION

3. 11.3 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM, for the processing'nd packaging of radioactive wastes to ensure compliance with 10 CFR Part 20, 10 CFR Part 71, and Federal regulations aoverning the disposal of the waste.

APPLICABILITY: At al 1 times.

ACTEON:

a ~

Mith the requirements of 10 CFR Part 20, and/or 10 CFR Part 71, not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

b.

With the solid radwaste system inoperable for more than 31 days, prepare 'and submit to the Commission within 30'days pursuant to Specification

6. 9. 2 a Special Report which includes the following information:

2.

Identification of the inoperable equipment or subsystems and the reason for inoperability, L

Action(s) taken to restore the inoperable equipment to OPERAStE

status, 3.

A description of the alternative used for SOLIDIFICATION and

~

packaging of radioactive wastes, and 4.

Summary description of action(s} taken to prevent

.a recurrence.

c..

The provisions of Specifications 3.0.3 and 3.0. 4 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 11.3. 1 The solid radwaste system shall be demonstrated OPERABLE at least once per 92 days by:

a.

Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROl PROGRAM, or b.

Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

SUS(UEHANNA UNIT 1 3/4 11"20 Amendment No.

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3/4. 1 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1. 1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reacto~

can be made sub-critical 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 subcritical to preclude inad-vertent criticality in the shutdown condition.

Since cove reactivity values will vary through cove life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be perfarmed in the co1d, xenon-free condition and shall. shaw 'the core to be sub-critical by at least R + 0.38K delta k/k or R + 0.28" delta k/k, as appro-priate.

The value of R in units of Z de;ta k/k is the difference between the~

The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be determined analytically ov by test.

The SHUTDOWN MARGIN is demonstrated by control rod withdrawal at the beginning of life fuel cycle conditions,"and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure.

Observation of subcriticality in this condition assures subcritica-Iity with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumptianiin the analysis of plant performance and can be best demonstrated at the time of fuel loading,

. but the margin must also be determined anytime a control rod is incapable of insertion.

3/4.1.2 Reactivit Anomalies

- Since the'HUTDOWN MARGIN requirement is small, a careful check on'ctual reactor conditions compared to the predicted conditions is necessary.

Any changes in reactivity from that of the predicted (pvedicted core k ff) can be eff determined from the core monitoring system (monitored cove k ff).

In the absence of any deviation in plant operating conditions or reactivity anomaly, these values should be essentially equal since the calculational methodologies are consistent.

The predicted core k ff is calculated by a 3D core simulation code as a function eff of cycle, exposure.

This is performed for projected or anticipated reactor operat-ing states/conditions throughout the cycle and is usually done prior to cycle operation. - The monitored core k

is the k f as calculated by the core monitor ing system for actual plant conditions.

Since the comparisons are easily done, frequent checks are not an imposition on normal operation..

A 1X deviation in reactivity fram that of the predicted is larger than expected for normal operation, and thH efare should be throughly evaluated.

A deviation as large as 1Z would. not exceed the design conditions of the reactor.

SUSQUEHANNA -:UNIT )

B 3/4 l-l '.

Amendment No.

45

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.4 CONTROL ROO PROGRAM CONTROLS (Continued)

O 280 cal/gm design limit to demonstrate compliance for each operating cycle.

If cycle-specific values of the above parameters are outside the range assumed in the parametric

analyses, an extension of the analysis or a cycle-specific analysis may be required.

Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for per-forming the Control Rod Orop Accident analysis are provided in XN-NF-80-19 Volume I.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation.

Two channels are provided.

Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

3/4. 1.5 STANOBY LI UIO CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold,'enon-free

shutdown, assuming that none of the withdrawn control rods can be inserted.

To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.

A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown require-ment.

There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing.

The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm.

The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the fillingof other piping systems connected to the reactor vessel.

The temperature requirement for the sodium penetrate solution is necessary to ensure that the sodium penetaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

Surveillance requirements are established on a frequency that assures a

high reliability of the system.

Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

(

Replacement of the explosive char ges in the valves at regular in arval s will assure that these valves will not fail because of deterioration of the charges.

SUS)UEHANNA - UNIT 1" B 3/4 1-4 Amendment No. 57

REACTIVITY CONTROL YSTEMS BASES 3/4.1.S CONTROL ROO PRO~GGi M CONT OLS (Continued)

Replacement wi 11 assure that charges.

of the exol si e charges in the valves at regular intervals t, ese valves~gill not rail because of'etarioration of the

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SUSQUEHANNA - UNIT 1 Amendment No.45

3/4. 10 SPECIAL TEST EXCEPTIONS BASES 3/4 10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel. tests are being performed during the low power PHYSICS TESTS.

3/4.10.2 ROD SE UENCE CONTROL SYSTEM In order to perform the tests required in the technical'pecifications it is necessary to bypass the sequence restraints on control rod movement.

The additional surveillance requirments ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.

3/4. 10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur.

These additional restrictions are specified in this LCO.

3/4. 10. 4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

4. 10. 5 OXYGEN CENTRATIO lief from the o

en concentra ion specifica ns is necessa in order to provi access to the imary containmeqt during the 'nitial start and testing pha of operation.

Without this access the start and test prh ram could be restr ted and delaye 3/4. 10.

5'TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling..RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.

SUSQUEHANNA " UNIT 1 B 3/4 10-1

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A~lc<<ED FIGURE 5.1.3-1a MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS ANO LIQUID EFFLUENTS SUS(UEHANNA - UNIT 1 5-4 Amendment No. p9

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ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Contfnued) 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a.

b.

The intent of the original procedure fs not altered.

The change fs approved by two members of the unit management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

The change is documented, reviewed fn accordance with Specification 6.5.1.6 or 6.5.3, as appropriate, and approved by the Superintendent of Plant-Susquehanna within 14 days of fmplementatfon.

6.8.4 The following programs shall be establfshect, implemented, and maintained:

a.

Prfma Coolant Sources Outsfde Contafaaent

~A program to reduce leakage frea those portions of systems outside containment that could contain highly radioactive fluids during a serious transfent or accident to as low as practical. levels.

The systems include the core spray, high pressure coolant infection, reactor core isolation cooling, reactor water cleanup, standby gas treatment, scram dfscharge, heat renoval, post accident sampling and containment afr monitoring systems.

pesi4c a(

The program shall include the followfng:

1.

Preventive maintenance and periodic visual inspection requirements, and 2.

Integrated leak test requirements for each system at refueling cycle intervals or less.

b.

In-Plant Radiation Monitorin c

A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

Post-accfdent Sam lfn A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates fn plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following:

1.

Training of personnel.

2.

Procedure for sampling and analysis, 3.

Provfsions for maintenance of sampling and analysis equipment.

SUS/UEHANNA " UNIT 1 6-16 Amendment No.49

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM Recirculation Loops...3'&. +~)..g9; bm gee'ss<gso~

Lsscsrrc - Sante Loop Opera.t sin Jet Pumps o

~

~ ~ I ~

~

~ ~ ~ ~

~

~ ~

~

~ ~ ~

~

~ ~ ~

~ ~ ~ ~

Recsrculation Pumps..................................

Idle Recirculation Loop Startup......................

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

3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...

Oper ati onal Leakage.........

3/4. 4. 4 CHEMISTRY.....,.....

3/4. 4. 5,<<SPECIFIC ACTIVITY........'..............~.............

3/4.4. 6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System...............................

Reactor Steam Dome.......

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES 3/4.4. 8 STRUCTURAL INTEGRITY............

3/4.4. 9 RESIDUAL HEAT REMOVAL ot Shutdown...

~ ~. ~ ~ ~ ~ ~ ~ ~.

~ ~ ~. ~......

~......

~....

~.

~

H Cold Shutdownl ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~

~ ~ ~

~

~ ~ ~

3/4. 5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING.....................................

3/4.5.2 ECCS - SHUTDOWN......................................

3/4.5. 3 SUPPRESSION CHAMBER..................................

PAGE 3/4 4-1 3/+ 0-re 3/4 4-2 3/4 4"3 3/4 4 4,

3/4 4-5 3/4 4-6 3/4 4-7 3/4 4"10 3/4 4-13 3/4 4-16 3/4 4-20 3/4 4-21 3/4 4"22 3/4 4-23 3/4 4-24 3/4 5-1 3/5 5-7 3/4 5-9 SUS(UEHANNA - UNIT 2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION REFUELING OPERATIONS (Continued)

PAGE 3/4.9.8 WATER LEVEL -

REACTOR VESSEL............:.............

3/4 9-11 3/4.9.9 WATER LEVEL - SPENT FUEL STORAGE POOL.................

3/4 9-12 3/4.9. 10 CONTROL ROD REMOVAL Single Control Rod Removal.............

3/4 9-U Multiple Control Rod Removal.....................

3/4 9-15 3/4. 9, 11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION Hign Water Level................

ow Water Level.......................................

L 3/4 9"17 3/4 9-18 3/4. 10 SPECIAL TEST EXCEPTIONS 3/4.10. 1 PRIMARY CONTAINMENT INTEGRITY..........,..............

3/4 10-1 3/4. 10.2 ROD SEQUENCE CONTROL SYSTEM.

3/4. 10. 3 SHUTDOWN MARGIN DEMONSTRATIONS.

3/4 10"2 3/4 10"3 3/4. 10. 4 RECIRCULATION LOOPS....',..............................

3/4 10-4 3/4. IO.P TRAINING STARTUPS 5

3/4 10-j/g SUS/UEHANNA - UNIT 2

BASES SECTION INSTRUMENTATION (Continued)

INDEX'AGE 3/4. 3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation..........

Seismic Monitoring Instrumentation............

Meteorological Monitoring Instrumentation.....

Remote Shutdown Monitoring Instrumentation....

Accident Monitoring Instrumentation...........

Source Range Monitors.........................

Traversing In-Core Probe System...............

Chlorine Detection System.......

Fire Detection Instrumentation................

Radioactive Liquid Effluent Instrumentation...

Radioactive Gaseous Effluent Instrumentation..

Loose-Part Detection System..

8 3/4 3-4 8 3/4 3-4 8 3/4 3"5 8 3/4 3-5 8 3/4 3"5 8 3/4 3-5 8 3/4 3"5 8 3/4 3-5 8 3/4 3-6 8 3/4 3-6 8 3/4 3-6 8 3/4 3"6 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEH.............

8 3/4 3-7 3/4. 4, 4 HEHISTRY s o

~ ~ ~ ~ ~

~ o ~

~

~ ~ ~

~

~ ~

~

~ ~ ~

C 3/4. 4. 5 SPECIFIC ACTIVITY.......................

3/4.4.6 PRESSURE/TEHPERATURE LIMITS 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.........................,.......

3/4.4 REACTOR COOLANT SYSTEM 3/4. 4.l RECIRCULATION SYSTEM..

3/4.4. 2 SAFETY/RELIEF VALVES...

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.......................

Operational Leakage.............................

8 3/4 3-7 8 3/4 4-1 3 3/4 4-g 3

8 3/4 4-2 8 3/4 4-2 8 3/4 4-2 8 3/4 4-3 8 3/4 4-4 3/4.4. 7 MAIN STEAM LINE ISOLATION VALVES................

8 3/4 4-5 3/4.4. 8 STRUCTURAL INTEGRITY....;.......................

8 3/4 4-5 3/4.4.9 RESIDUAL HEAT REHOVAL...........................

8 3/4 4-5 SUS(UEHANNA - UNIT 2

INDEX BASES SECTIGN 3/4. 10 SPECIAL TEST EXCEPTIONS PAGE 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY..................

8 3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM.....................

8 3/4 10-1 3/4. 10. 3 SHUTDOWN MARGIN DEMONSTRATIONS..................

8 3/4 10-1 3/4.10.4 RECIRCULATION LOOPS.............................

8 3/4 10-1 3/A.IO.P TRAINING STARTUPS 3/4. 11 RADIOACTIVE EFFLUENTS 3/4. 11. 1 LIQUID EFFLUENTS

~

~ ~ ~

~

8 3/4 10-1 Concentration.

~

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

~

0ose...............

~

Liquid Waste Treatment System....

3/4. 11.2 GASEOUS EFFLUENTS I

Dose Rate..................

Dose-Noble Gases...

8 3/4 11"1 8 3/4 11-1 8 3/4 11"2 8 3/4 11-2 8 3/4 11"3 Dose-Iodine-131, Tritium, and Radionucl Particulate Form.....................

ides in 8 3/4 11-3 Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System...........

8 3/4 11-4 Explosive Gas Mixturel ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

Main Condenser...................................

Venting or Purging...............................

8 3/4 11-4 8 3/4 11-5 8 3/4 11-5 3/4.11.3 SOLID RADIOACTIVE WASTE..........................

8 3/4 11-5 3/4. 11. 4 TOTAL DOSE..........

8 3/4 11-5 SUSQUEHANNA - UNIT 2 xvi

LIST OF FIGURES INGEX FIGURE

3. 1. 5-1

3. 1. 5-2 3.2. 1-1

3. 2. 1"2

3. 2. 1-3 3.2. 3-1

3. 2. 3-

~

~ I 3.

l. 1-1 h

3.4.6. -1 SODIUM PEHTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS...................,....

SODIUM PENTABORATE SOLUTION CONCENTRATION.........

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, INITIALCORE FUEL TYPE 8CR183 (LOW EHRICHMENT)....

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, INITIALCORE FUEL TYPE 8CR233 (MEDIUM ENRICHMENT).

MAXIMUMAVERAGE PCANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, INITIALCORE FUEL TYPE 8CR711 (NATURAL ENRICHMENT).................................:...-.

MINIMUM CRITICAL POWER RATIO. (MCPR)

VERSUS x AT. RATED FLOW....:...............................

) FACTOR.........................................

K THERMAL POWER LIMITATIONS..........................

MINIMUM REACTOR. VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE.............:.............

PAGE 3/4 1-21 3/4 1-22 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-8 3/4 2-9 3/4 4-lb 3/4 4-18 4.7. 4-1 B 3/4 3-1 B 3/4.4.6-1 REACTOR VESSEL WATER LEVEL.................-......

FAST NEUTRON FLUENCE (E)1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE.....-....................

B 3/4 3-8 B 3/4 4-7 SAMPLE PLAN 2)

FOR SNUBBER FUNCTIONAL TEST........

3/4 7-15

5. l. 1'-1

5. l.2-1 5.1. 3-la MAP DEFINING UNRESTRICTED AREAS FOR GASEOUS AND LIQUID EFFLUEHTS.......

RADIOACTIVE

~

~ ~ ~ 0

~

EXCLUSION AREA oeeo

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

~

~ ~ eo

~ ~

~

LOW POPULATION ZONE...............................

5-2 5-3 5-4

5. l.3-lb

6. 2. 1-1

~

MAP DEFINING UNRESTRICTED AREAS FOR GASEOUS ANO LIQUID EFFLUENTS.......

OFFSITE ORGANIZATION RADIOACTIVE 6-3

6. 2. 2-1 UNIT ORGANIZATION...............,......... -.......

6-4 SUSQUEHANNA - UNIT 2 XX11 Amendment No. 2

3/4. 0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0. 1

'Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the LiNting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a Specification'hall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals.

If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the AHkae requirements is not required.

ACTloH 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the Specification does not apply by placing it, as applicable, in:

1.

At least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation.

Exceptions to these requirements are stated in the individual Specifications.

This specification is not applicable in OPERATIONAL CONDITION 4 or 5.

I 3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements.

This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements.

Exceptions to these requirements are stated in the individual Specifications.

SUS(UEHANNA - UNIT 2 3/4 0-1

APPLICABILITY SURVEILLANCE RE UIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

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

a.

A maximum allowable extension not to exceed 25K of the surveillance interval, but b.

The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation, Exceptions to these requirements are stated in the individual Specifica ns.

Surveillance requirements do not have to be performed on inoperable equ1pmen 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, 8

3 components shall be applicable as follows:

a.

.Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

b.

Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boilm and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing ins ection and testin activities 'ctivities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannual.ly or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days SUSQUEHANNA - UNIT 2 3/4 0-2

REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a ~

b.

With the maximum scram insertion time of one or more control rods exceeding

7.0 seconds

1.

Declare the control rod(s) with the slow insertion time inoperable, and 2.

Perform the Surveillance Requirements of Specification 4. 1.3.2.c.

at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demonstrated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

a.

For all control rods prior to THERMAL POWER exceeding 4(C of RATED THERMAL POWER following CORE ALTERATIONS" or after a reactor shutdown that is greater than 120 days, b;

For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and c.

For at least 10K of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.

"Except rod movement.

normal control SUSQUEHANNA -UNIT 2 3/4 1-6

TABLE 3. 3. 1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

A channel may be placed in an inoperable status for up to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s-for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b)

This function is automatically bypassed when the reactor mode switch is in the Run position.

(c)

The "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn" and shutdown margin demonstrations performed per Specification

3. 10. 3.

(d)

The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRHS and 6

IRHS.

(e)

An APRH channel is inoperable if there are less than 2

LPRM inputs per level or less than 14 LPRM inputs to an APRH channel.

(f)

This function is not required "o be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification

3. 10. 1.

(g)

This function is automatically bypassed when the reactor mode switch is not in the Run position.

(h)

This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i)

With any control rod withdrawn.

(j)

This function shall be automatically bypassed when turbine firstBtage pressure is less than 108 psig or 17K of the value of first stage pressure in psia at valves wide open (V.W.O) steam flow, equivalent to THERMAL POWER of about 24K of RATED THERMAL POWER.

(k)

Also actuates the EOC-RPT system.

"Not required for control rods removed per Specification

3. 9. 10. 1 or 3. 9. 10.2.

SUS(UEHANNA - UNIT 2 3/4 3-5

TABLE 3.3.2-1 ISOLATION ACTUATION INSTRNENTATION TRIP FUNCTION 1.

PRIHNY CONTAIINENT ISOLATION

a. 'eactor Vessel i@ter Level 1)

Los, Level 3 ISOLATION SIGNAL S a

HININNt OPERABLE CHANNELS PER TRIP SYSTEH b

APPLICABLE OPERATIONAL CONDITION 1, 2, 3 ACTIN 20 b;

Co d.

eo 2)

Lee Ler, Level 2 3)

Low Let Low, Level 1 Drywall Pressure - High Hanual Initiation SGTS Exhaust Radiation-High Hain Steaa Line Radiation-High X

Y,Z NA

.2 2

2 l.

1, 2, 3 1, 2, 3 1, 2, 3 1, 2, 3 1,2;3 20 20 20 2I g*AA 5*** 20 20 2.

SECONDARY CONTAINNENT ISOLATION O

b.

C.

d.

e Reactor Vessel ltater l.evel-Ler Lee, Level 2 D~ll Pressure - High Refuel Floor High Exhaust Duct Radiation - High Railroad Access Shaft Exhaust Duct Radiation - High Refuel Floor Mall Exhaust Duct Radiation - High Hanual Initiation A*

A*

1, 2, 3 and e 1,213 1,2,3and~

25 25 25 25 25 24

TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUHENTATION SETPOINTS TRIP FUNCTION 1.

PRIHARY CONTAINNENT ISOLATION TRIP SETPOINT ALLOMABLE VALUE

'a 0 b.

C.

d.

e.

Reactor Vessel Mater Level 1)

Low, Level 3 2)

Low Low, Level 2 3)

Low Low Low, Level 1 Drywell Pressure - High Hanual Initiation SGTS Exhaust Radiation - High Hain Steam Line Radiation - High

>13.0 inches"

> -38.0 inches"

> -129 inches"

< 1.72 psig NA

< 23.0 mR/hr

< 3 X full power background

> 11.5 inches

> -45.0 inches

> -136 inches

< l.88 psig Q

< 31.0 mR/hr

< 3.6 X full power background 2.

SECONDARY CONTAINNENT ISOLATION a 0 b.

C.

d.

e.

Reactor Vessel Mater Level-Low Low, Level 2 Drywell Pressure - High Refuel Floor High Exhaust Duct Radiation - High Railroad Access Shaft-Exhaust Duct Radiation - High Refuel Floor Mall Exhaust Duct Radiation - High Hanual Initiation

> -38.0 inches*

< 1.72 psig

< 2.5 aR/hd

< 2.5 mR/hP

< 2.5 mR/hd NA

> -45.0 inches

< 1.88 psig

< 4.0 mR/hP

< 4.0 aR/h~

< 4.0 aR/hP'M NA 3.

HAIN STEAN LINE ISOLATION a0 b.

C.

d.

Reactor Vessel Mater Level - Low Low Low, Level 1

Hain Steam Line Radiation - High Hain Steam Line Pressure

- Low Hain Steam Line Flow - High

> -129 inches"'

3 X full power background

> 861 psig

< 107 psid

> -136 inches

< 3.6 X full power background

> 841 psig

< 110 psid

I TRIP FUNCTION TABLE 3.3. 2-2 (Continued)

ISOLATION ACTUATION INSTRNENTATION SETPOINTS TRIP SETPOINT ALLOWABLE VALUE C

HAIN STEAH LINE ISOLATION (Continued) e.

Condenser Vacua - Low f.

Reactor Building Hain Steam Line Tunnel Teaperature - High g.

Reactor Building Hain Steaa Line Tunnel h Teaperature - High h.

Hanual Initiation i.

Turbine Building Hain Steam Line Tunnel Teaperature " High 4.

REACTOR WATER CLEANUP SYSTEH ISOLATION a.

RWCU h Flow - High b.

RWCU Area Teaperature - High c.

RWCU/Area Ventilation h Teaperature - High d.

SLCS Initiation e.

Reactor Vessel Water Level-Low Low, Level 2

~

f.

RWCU Flow - High g.

Hanual Initiation

> 9.0 inches Hg vacuua

< 177 F

<99F NA

< 177F

< 60 gpa

< 147'f or 118.3 ff

< 694F or 35.3ofi NA

> -38 inches*

426 gpa NA

> 8.8 inches Hg vacua

< 184'F

< 108'f NA

< 184'F

< 80 gtxe

< 154oF or 125.34' 78'F or 44.3'FN NA

> -45 inches 436 ~

NA 5.

I.

REACTOR CORE ISOLATION COOLING SYSTEH ISOLATION

< 10.0 psig a.

RCIC Steaa Line h Pressure - High

< 153" Hs0 b.

RCIC Steaa Supply Pressure - Low

> 60 psig c.

RCIC Turbine Exhaust Diaphraga Pressure - High

< 165" He~

'> 53 psig

< 20.0 psig

TABLE 3. 3. 2-2 (Cont inued)

ALLOWABLE VALUE RCIC Equipment Room Temperature - High RCIC Equipment Room 6 Temperature - High

< 174 F~

<98F

< 174 FM

< 98 FNk

<167 P e.

< 89 F

< 167 FNN

< 89 FOP RCIC Pipe Routing Area Temperature - High RCIC Pipe Routing Area h lemperature - High RCIC Emergency Area Coolel Temperature - High Manual Initiation Drywell Pressure

- High h.

<147 F

NA

< 1.72 psig

< 154 F

NA

< 1.88 psig i ~

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION ISOLATION ACTUATION INSTRUMENTATION SETPOINTS m

TRIP FUNCTION TRIP SETPOINT REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION (Continued) d ~

a.

b.

C.

e.

g.

h.

l.

J

~

HPCI Steam Line Flow - High HPCI Steam Supply Pressure

- Low HPCI Turbine Exhaust Diaphragm Pressure - High HPCI Equipment Room Temperature - High HPCI Equipment Room h Temperature - High HPCI Emergency Area Cooler Temperature - High HPCI Pipe Routing Area Temperature - High HPCI Pipe Routing A/ea h Temperature - High Manual Initiation Drywell Pressure - High

< 275 inches H~O~

> 104 psig

< 10 psig

< 167 F

<89F

< 147F

< 167 FNN

< 89'FM NA

< 1.72 psig

< 292 inches H~O~

90 psig

< 20 psig

< 174 F

<98F

< 154'F

< 174 FM

< 98FN NA

< 1.88 psig

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 7.

RHR SYSTEM SHU DOWN COOLING/HEAD SPRAY MODE ISOLATION a.

b.

C.

d.

e.

Reactor Vessel Water Level-Low, Level 3

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High RHR Equipment Area h Temperature - High RHR Equipment Area Temperature - High RHR Flow - High Manual Initiation

> 13 '

inches"

< 98 psig

< 89'F~

< 167 P

< 25,000 gpm NA

> 11.5 inches

< 108 psig

< 90.5 ~

< 170.5 F~

< 26,000 gpm NA g.

Drywell Pressure

- High 1 ~ 72 psig

< 1.88 psig

See Bases Figure B 3/4 3-1.

HLower setpoints for TSH-G33-N600 E, F and TSH-G33-N602 E, F.

N15 minute time delay.

TABLE 3. 3. 6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RBM inoperable and take the ACTION required by Speci ficat ion 3. 1.4.3.

ACTION 61 With the number of OPERABLE Channels:

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

NOTES With THERMAL POWER >'30K of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9. 10. 1 or 3.9

~ 10.2.

Not required when eigh fewer fuel assemblies (adjacent to the SRMs

)

are in the core.

Ot (a)

The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30~ of RATED THERMAL POWER.

(b)

This function shall be automatically bypassed if detector count rate is 100 cps or the IRM channels are on range 3 or higher.

(c)

This function is automatically bypassed when the associated IRM channels are on range 8 or higher.

(d)

This 4'unction is automatically bypassed when the IRM channels are on range 3 or higher.

N (e)

This function is automatically bypassed when the IRM channels are on range l.

SUS(UEHANNA - UNIT 2 3/4 3-53 Amendment No 16

CA CII Cm TABLE 3.3.7.5-1 ACCIDENT HONITOR ING INSTRUHENTATION C:

I tO INSTRUHENT 2.

3.

4.

5.

6.

7.

8.

Reactor Vessel Steam Dome Pressure Reactor Vessel Mater Level Suppression Chaaber Mater Level Suppression Chaaber Mater Teaperature Suppression Chamber Air Temperature Primary Containment Pressure Drywell Temperature Drywell Gaseous Analyzer

.a.

Oxygen b.

Hydrogen 9.

Safety/Relief Valve Position Indicators 10.

Containment High Radiation ll.

Noble gas monitors"~

a.

Reactor Bldg. Vent b.

SGTS Vent c.

Turbine Bldg. Vent 12.

Primary Containment Isolation Valve Position 13.

Neutron Flux REQUIRED NUHBER OF CHANNELS 8,

6 locations 2

2/range 2

2 2

1/valve" 1

1/val ve HINIHUH CHANNELS OPERABLE 6, 1/location 1

1/range 1

1 1/valve" 1

1/val ve 1

ACTION 80 80 80 80 80 80 80 80 82 80 81 81 81 81 80 80 APPLICABLE OPERATIONAL CONDITIONS lo 2 1, 2 1,

2 1,

2 1, 2 1,

2 1, 2 1,N 24 l,f 2f 1, 2 1, 2 2 and'c<

1 2 and*""

1, 2

Acoustic monitor.

"*Hid-range and high-range channels.

"*"Mhen moving irradiated fuel in the secondary containment.

PSee Special Test Exception 3. 10. l.

ACTION 80-TABLE 3.3.7. 5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS a.

With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3. 3.7. 5-1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 81 "

With the number of OPERABLE channels less than requir ed by the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable channel(s) to OPERABLE status within 7 days of the event, or ACTION 82-2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the

'inoperability and the plans and schedule for restoring the system to OPERABLE status.

a.

With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel to OPERABLE status within 30 days or be jn at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b..

With the number of OPERABLE channels less then the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, restore at least one channel to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SUS(UEHANNA - UNIT 2 3/4 3-'72

TABLE 3.3.7.9"1 INSTRUMENT LOCATION FIRE ZOHE ROON OR AREA FIRE OETECTION INSTRUMENTATION r

INSTRUMENTS OPERABLE PHOTO-HEAT IONIELTION ELECTRIC ELEV.

TOTAL MIM.

0-22A 0-24D 0-24G 0-24G 0"25A 0-25B 0-25C 0-250 0-25E 0-268 0"26C 0-260 0"26F 0-26G 0-26H 0-26H 0-26H 0"26H 0-26l 0-26J 0"26H 0-26N 0-26P 0-26R 0-26S ntrol Buildin Filter Area Lower Relay Room Lower Relay Room PGCC Lowet Cable Spreading Rm.

South Cable Chase Center'able Chase North Cable Chase Lower Cable Spreading Rm.

South Cable Chase Center Cable Chase North Cable Chase Vestibule Shift Office Control Rm.

(Under Flr. Unit 1)"

Control Room (Under Flr. Unit 2)"

Control Room Control Rm.

(Above Clg)"

Operational Support Center Vestibule Soffit Control Room Soffit Control Room Soffit Soffit South Cable Chase 687'Bu NA 698'-14 4

698'-I" 4

698'-1" 54 HA 2

2 27 714'"0" 20 10 714'-0" 714'-0" 1

714'-0" 1

1 714'-0" 729 I 1ll 729E -14 729E -1N 729'1" 729'-1" 26 13 NA HA NA NA NA NA NA NA NA NA 729'-1" 729'-1" 729'-1" 729'-1" HA HA NA HA NA HA HA,

NA 729'-1" 729'1" 729'-1" 729'-1" 7291 1II 729'1" NA NA NA HA HA HA HA HA HA NA 1

1 729'-1" NA NA 11 6

4 2

,4 2

30 15 6

3 HA, NA NA'A NA HA 6

3 1

1 1

1' 1

1 1

18 9

15 8

10 5

P'9 P5 1

1 1

1 4

2 2

1 2

1 4

2 HA HA NA NA NA NA HA NA HA HA N

NA NA NA N

HA N

NA NA NA NA NA" HA HA NA HA HA HA HA HA HA NA HA NA HA HA HA HA NA HA

(

HA HA SUS(UEHANN - UNIT 2 3/4 3-78

INSTRUMENT LOCATION TABLE 3.3. 7. 9-1 (Continued)

FIRE OETECTION INSTRUMENTATION INSTRUMENTS OPERABLE FIRE ZONE a.

C 0-26T 0-26V 0-27A 0-27A 0-27B 0-27C 0-27E 0-27F 0"27G 0"27H 0"28A 0-28B 0-28C 0"280 0-28E 0"28F 0-28G 0-28H 0-28I 0"28J 0-28K 0-28L 0-28M 0"28N 0-28P 0-28Q 0"28R ROOM OR AREA ontrol Buildin (Continued)

Center Cable Chase North Cable Chase Upper Relay Room PGCC Upper Cable Spreading Rm.

Upper Cable Spreading Rm.

Upper Relay Room South Cable Chase

'enter Cable Chase North Cable Chase Equipment Room Equipment Room Battery Room Battery Room'attery Room Battery Room Battery Room Repair Shop Battery Room Battery Room Battery Room Battery Room Battery Room Battery Room South Cable Chase Center Cable Chase North Cable Chase HEAT ELEV.

TOTAL MIN.

753'-0" 754'-1" 754'-1" 754'-1" 754'-1" 771'-0" 771 I ~OII 771'-0" 771'-0" 771'-0" 771'OII 771'-0" 771'-Ou 771'"0" 771'Ow 771'-0" 771'-0" 771I -O'I 771'0" 771'-0" 771'-0" 771'-0" 25 4

2 1

1 1

1 "NA NA NA NA NA NA NA NA NA HA NA NA NA NA NA NA HA NA NA NA NA NA NA HA NA NA NA NA 1

.1 1

1 1

1 729'1" 1

1 754'1" 2

1 754'-1" 55 28 753'"0" 24 12 NA NA 2

30 NA NA 1

15 2

NA NA NA 4

1 1

1 2

1 1

NA NA NA 1

NA HA HA 2

2' 1

1 1

NA NA NA IONIZATION TOTAL MIN.

PHOTO-ELECTRIC TOTAL MIH.

NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA HA NA NA NA NA NA NA NA NA NA NA NA NA NA HA NA NA SUSQUEHANNA - UNIT 2, 3/4 3-79

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4.1. 1.2 One reactor coolant recirculation loop shall be in operation with the pump speed

< 90K of the rated pump speed, and a.

the following revised specification limits shall be followed:

I.

Specification

2. 1.2:

the MCPR Safety Limit shall be increased to 1.07.

2.

Table 2.2.1-1:

the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

Tri Set oint W+

Al 1 owabl e Value

< 0.58W + 5 3.

Specification 3.2.1:

The MAPLHGR limits shall be the limits specified in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3, multiplied by 0.81.

4.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

5.

Table 3.3.6-2:

follows:

Tri Set oint 8W + 55K)T SRB

< (0.58W + 46K)T the RBM/APRM Control Rod Block Allowable Value S

0.

8W +

8X)T SRB

< (0.58W + 49K)T Setpoints shall be as a.

RBM - Upscale Trio Set oint, Al 1 owab le Value 1.

< 0.66W + 35 2.

< 0.66W + 37K

< 0.66W + 40K 5.a. 1 and S.a.2 shall be used in conjunction with the MCPR limits specified in Figures 3.2.3-1a and 3.2.3-1b, respectively.

b.

APRM-Flow Biased Tri Set oint Allowable Value

< 0.5

+ 46

~5 b.

APRM and LPRM~" neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4.1.1.1-1.

c.

Total core flow shall be greater than or equal to 42 million s/hr when THERMAL POWER is greater than the limit specified in Figur 3x4.1.1.1-1.

/

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2", except during t o loop operation.¹ ACTION:

a.

With no reactor coolant system recirculation loops in operation, take the ACTION required by Specification 3.4. 1. l.l.

SUSQUEHANNA - UNIT 2 3/4 4-lc Amendment No. 26

'I

~

REACTOR COOLANT SYSTEM RECIRCULATION PUMPS

(

LIMITING CONOITION FOR OPERATION

~ V

~

V (I

I 3".4".1;3-'ecirculation pump speed mismatch shall be-,maintained'.within::

a.

5X'of each other, with core flow greater than:-or equal;;to 75'.

~ -",*-; ~ of rated core fl'ow.

~ ~

""" b. " ZOX of each other with(core flow less than 75K'f rated'ore flow.

~

. APPLICABILITY:

OPERATIONAL CONOITIONS 1" and. 2" when both, recirculation. loops are in operation.,'.

~

ACTION:.

~

(

r Nth the recirculation pump speeds di.fferent by more than the, specified-limits, either:

(

a.

Restore the recirculation pump speeds to within the specified limit",

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b'. 'eclar e 'the recirculation. loop of'he pump with. the slower speed in operation and. take. the ACTION.required by: Specification. 3 4..1 1 I.

~

4

~

(

, SURVEILLANCE RE UIREMENTS (V ~

4;4'.1.3'. Recirculation pumo speed mismatch shall be verified to be within the limits 'at.least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

'I "See Special Test Exception 3.10.4.

SUSQUEHANNA - UNIT 2 3/4 4"3 Amendme'nt No. 26

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4. 5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5. 1 The emergency core cooling systems shall be OPERABLE with:

a.

The core spray system (CSS) consisting of two subsystems with each subsystem comprised of:

1.

Two OPERABLE CSS pumps, and 2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

b.

The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of two subsystems with each subsystem comprised of:

l.

Two OPERABLE LPCI pumps, and 2.

An OPERABLE flow oath capable of taking suction from the suppression chamber and transferring'he water to the reactor vessel.

c.

The high pressure cooling injection (HPCI) syste~ consisting of:

1.

One OPERABLE HPCI pump, and 2.

An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

d.

The automatic depressur ization system (ADS) with six OPERABLE AOS valves.

APPLICABILITY:

OPERATIONAL CONOITION 1, 2" "" 0, and 3" "" @.

he HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.

""The AOS is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

See Special Test Exception 3. 10.$(

4%One LPCI subsystem of the RHR system may be inoperable in that it is aligned in the shutdown cooling mode when reactor vessel pressure is less than the RHR shutdown cooling permissive setpoint.

SUSQUEHANNA - UNIT 2 3/4 5-1

CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE LIMITING CONDITION FOR OPERATION

'.6. 1.7 Drywell average air temperature shall not exceed 135 F.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the drywell average air temperature greater than 1350F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS

~CPS 4.6. 1.7 The drywell average air temperature shall be the arithmetical average of the higher temperature at a minimum of three of the following~~

and shall be determined to be within the limit at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

I Izimeuu Tmp g,, HIdci(e

c. Roe~

f'mestee Elevation 797'8" 752 I 2ll 725'r 711'11'r 720'zimuth

1050, 2850 80 280 40

, 2600 80,.2700 e

SUSQUEHANNA - UNIT 2 3/4 6-10

CONTAINMENT SYSTEMS 3/4.6. 5 SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5. 1 SECONDARY CONTAINMENT INTEGRITY"" shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and ".

ACTION:

Without SECONDARY CONTAINMENT"" INTEGRITY:

a.

In OPERATIONAL CONDITION 1, 2, or 3, restor e SECONDARY CONTAINMENT""

INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In Operational Condition ", suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.6.5. 1 SECONDARY CONTAINMENT"" INTEGRITY shall be demonstrated by:

a.

Verifying at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the secondary containment is~ than or equal to 0.25 inch of vacuum water gauge.

grec Wm b.

Verifying at least once per 31 days that:

la.

When the railroad bay door (No. 101) is closed; all Zone I and III hatches, removable walls, dampers, and doors connected to the railroad access bay are closed,¹¹ or i)

~Onl Zone I removable walls and/or doors are open to the ra>lroad access shaft,8¹ or ii)

~Onl Zone III hatches and/or dampers are open to the ra lroad access shaft.8¹ lb.

When the railroad bay door (No. 101) is open; all Zone I and III

hatches, removable walls, dampers, and doors connected to the railroad access bay are closed.

"When irradiated fuel is being handled in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

""Secondary Containment consists of Zone I, Zone II and Zone III or Zone g and Zone III when Zone I is isolated from Zone Il and Zone III.

zc

¹¹Personnel ingress and egress through doors within the secondary containment is not prohibited by this specification.

SUS(UEHANNA -'UNIT 2 3/4 6-31

CONTAINMENT. SYSTEMS ORYWELL AND 'SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.6.6.3 The drywell and suppression chamber atmosphere oxygen concentration shall be less than 4X by volume.

APPLICABILITY:

OPERATIONAL CONDITION, during the time period:

a.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15K of RATED THERMAL POWER,'ollowing startup, to b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing THERMAL POWER to less than 15K of RATED THERMAL POWER preliminary to a scheduled reactor shutdown.

ACTION:

With the oxygen concentration in the drywell and/or suppression chamber exceeding the limit, restore the oxygen concentration to within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.6.3 The oxygen concentration in the drywell and suppression chamber shall be verified to be within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL'OWER is greater than 15K of RATED THERMAL POWER and at least once per 7 days thereafter.

SUSQUEHANNA - UNIT 2 3/4 6-41

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7. 1.2

. Two independent emergency service water system loops shall be OPERABLE with each loop comprised of:

a.

Two OPERABLE emergency service water pumps, and b.

An OPERABLE flow path capable of taking suction from the spray pond and transferring the water to the associated safety-related equipment.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and ".

ACTION:

a.

b.

In OPERATIONAL CONDITION 1, 2, or 3:

I.i With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 7 days or be in 4 least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN)within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I ~y 2.

With two emergency service water pumps inoperable, restore at least one inoperable pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With one emergency service water system loop otherwise inoperable, restore the inoperable loop to OPERABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN withir the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; In OPERATIONAL CONDITION 4, 5 or ":

1.

With one pump in an emergency service water system loop inoperable, verify adequate cooling capability remains available for the diesel generators required to be operable by Specification 3.8.1.2 or declare the affected diesel generator(s) inoperable and take the ACTION required by Specification 3.8.1.2.

2.

With two pumps in an emergency service water system loop inoperable or with the loop otherwise inoperable declare the associated safety related equipment inoperable (except diesel generators),

and follow the applicable ACTION statements.

Verify adequate cooling remains available for the diesel generators required to be operable by Specification 3.8. 1.2 or declare the affected diesel generator(s) inoperable and take the ACTION required by Specification 3.8. 1.2.

"When handling irradiated fuel in the secondary containment.

Alhen any diesel generator is removed from service in order to do work asso-ciated with tying.in the additional diesel generator and its associated emergency service water pump is inoperable, Action a.l shall read as follows:

a. 1 With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status when its associated diesel generator is restored to OPERABLE status per Specification 3.8. 1. 1.

SUSQUEHANNA - UNIT 2 3/4 7-2 Amendment No.

19

h

ELECTRICAL POWER SYSTEMS A.C.

SOURCES -

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shy]1 be OPERABLE:

a.

One circuit between the offsite transmission network and the onsite Class 1E distribution system,.and b.

Two diesel generators with:

1.

An engine mounted day fuel tank containing a minimum of 325 gallons of fuel.

2.

A fuel storage system containing a minimum of 47,570 gallons of fuel.

3.

A fuel transfer pump.

APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5 and ".

ACTION:

a 0 With less than the above required A.C. electrical power sources OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment, operations with a,potential for draining the reactor vessel and crane operations over the spent fuel pool when fuel assemblies are stored therein.

In

addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet above the reactor pressure vessel

flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as soon as practical.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1. 1. 1, 4.8.1. 1.2 and 4.8. 1. 1.3, except for the requirement of 4.8. 1. 1.2.a.5.

Is 4c<ng Banc{(ccf irradiated fuel in the secondary containment gy A

f<<<<io~> ~,K a, ~sWti4 $a SUS(UEHANNA - UNIT 2 3/4 8-10

1i

ELECTRICAL POWER SYSTEMS D.C.

SOURCES -

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, Division I or Division II of the D.C. electrical power sources shall be OPERABLE with:

b.

Division I consisting of:

1.

Load group Channel "A" power source, consisting a) 125-volt D.C. battery bank b)

Full capacity charger 2.

Load group Channel "C" power source, consisting a) 125-volt D.C. battery bank b)

.Full capacity charger 3.

Load group "I" power source, consisting of:

a) 250-volt D.C. battery bank b)

Half"capacity chargers 4.

Load group "I" power source, consisting of:

a) 1 24-volt D.C. battery bank b)

Two half-capacity chargers Division II consisting of:

1.

Load group Channel "B" power source, consisting a) 125-volt D.C. battery bank b)

Full capacity charger 2.

Load group Channel "D" power source, consisting a) 125-volt D.C. battery bank b),

Full capacity charger 3.

Load group "II" power source, consisting of:

a) 250-volt D.C. battery bank b)

Full capacity charger 4.

Load group "II" power source, consisting of:

a) i 24-volt D.C. battery bank b)

Two half-capacity chargers

'f:

10610"", 20610 10613"",

20613 of:

10630"",

2D630 10633"",

20633 20650

20653A, 20653B 20670

20673, 20674 of:

10620"",

20620 10623"",

20623 of:

1D640"", 20640 10643*", 2D643 20660 20663 20680

20683, 20684 APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5, and ".

ACTION:

a.

With less than the above required Unit 2 125-volt and/or 250-volt D.C.

load group battery banks OPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

')S L I~g haMtW "When bc~~ irradiated fuel)in the secondary containment~

o.n4 d~<ng

""Not required to be OPERABLE when the requirements of ACTION b have been satisfied.

~

~ Ag~Ay<ogS ~ O~craf<e~a

~<~4 a +mt<cd 4r Are T~ ~~ +C ~fofCCC(

~

SUS(UEHANNA - UNIT 2 3/4 8"16 Amendment No.7

ELECTRICAL POWER SYSTEMS REACTOR PROTECTION SYSTEM ELECTRIC POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.3 Two RPS electric power monitoring assemblies for each inservice RPS MG set or alternate power supply shall be OPERABLE.

APPLICABILITY: At al l times.

ACTION:

a.

b.

With one RPS electric power monitoring assembly for an inservice RPS MG set or alternate power supply inoperable, restore the inoper able power monitor ing assembly to OPERABLE status. within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS MG set or alternate powet supply from service.

With both RPS electric power monitoring assemblies for an inservice RPS MG set or alternate power supply inoperable, restore at least one electric power monitoring assembly to OPERABLE status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

SURVEILLANCE REOUI ReMENTS 4.8.4.3 The above specified RPS electric power monitoring assemblies shall be determined OPERABLE:

Qn<+

a.

By'performance of a CHANNEL FUNCTIONAL TEST each time thef4 is in COLD SHUTDOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed within the previous 6 months.

b.

At least once per 18 months by demonstrating the OPERABILITY of over-

voltage, undervoltage, and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actuation of the protective relays, tripping logic and output circuit breakers and verifying the following setpoints:

1.

Overvoltage 2.

Undervoltage 3.

Underfrequency RPS Division A

< 129.1 VAC

> 112.0 VAC

> 57 Hz RPS Division B

< 130.3 VAC

> 112.5 VAC

> 57 Hz SUS(UEHANNA - UNIT 2 3/4 8"35

3/4. 9 REFUELING OPERATIONS 3/4. 9.1 REACTOR MODE SWITCH LIMITING CONDITION FOR OPERATION 3.9. 1 The reactor mode switch shall be OPERABLE and locked in the Shutdown or Refuel position.

When the reactor mode switch is locked in the Refuel position:

a.

A control rod shall not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE.

b.

CORE ALTERATIONS shall not be performed using equipment associated with a Refuel position interlock unless at least the following associ-ated Refuel position inter locks are OPERABLE for such equipment.

1.

2.

3.

4.

5.

APPLICABILITY:

.-'."TION:

All rods in.

Refuel platform position..

Refuel platform hoists fuel-loaded.

Fuel grapple positi,on.

Service platform hoist fuel-loaded.

OPERATIONAL CONDITION 5" C.

With the reactor mode switch not locked in the Shutdown or Refuel position as specified, suspend CORE ALTERATIONS and lock the reactor mode switch in the Shutdown or Refuel position.

With the one"rod-out interlock inoperable, lock the reactor mode switch in the Shutdown position.

With any of the above required Refuel position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperable Refuel position equipment'nterlock.

" See Special Test Exceptions

3. 10. 1 and 3.10.3.

SUS(UEHANXA - UNIT2'/4 9-1

~

SPECIAL TEST EXCEPTIONS 3/4. 10. 4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION y.y t.l,l->

I

3. 10.4 The requirements of Speciffcatfons 3.4.1. 1. 1.and 3.4.1.3 may be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the performance of:

a.

PHYSICS TESTS, provided that THERMAL POWER does not exceed SX of RATED THERMAL POWER, or b.

The Startup Test Program.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2, during PHYSICS'TESTS and the Startup Test Program.

ACTION:

a.

b.

With the above specified time limit exceeded, insert all control rods.

With the above specified THERMAL POWER limit exceeded during PHYSICS

TESTS, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE RE UIREMENTS

4. 10..4. 1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at least once per hour during PHYSICS TESTS and the Startup Test Program.

4. 10.4.2 THERMAL POWER shall be determined to be less than 5X of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

SUS)UEHANNA - UNIT 2 3/4 10-4 Amendment No. 26

SPECIAL TEST EXCEPTIONS 3/4. 0.5 OXYGEN CONCENTRATION LIMITING NDITION FOR OPERATION 3.10.5 The pro isions of Specification 3.6.6.4 may e suspended during the performance of th Startup Test Program until eit r the required 100K of RATED THERMAL POWER trip ests have been completed or e reactor has operated for 120 Effective Full P

er Days.

APPLICABILITY:

OPEBATI AL CONOITION 1.

ACTION

~

With the requirements of the above s

cification not satisfied, be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENT 4.10.5 The Effect' Full Power Days of operation.shall be verified to be less than 120, by cal lation, at least once per 7 days d'u ing the Startup Test Program.

SUSQUEHANNA - UNIT 2 3/4 10-5

SPECIAL TEST EXCEPTIONS 3/4. 10.

TRAINING STARTUP5 LIMITING CONDITION FOR OPERATION

3. 10.$5 The provisions of Specification 3.5. I may be suspended to permit one RHR subsystem to be aligned in the shutdown cooling mode during training startups provided that the reactor vessel is not pressurized, THERMAL {%WER

>s less than or equal to lX of RATED THERMAL POWER and reactor coolant temperature is less than 2004F.

APPLICABILITY:

OPERATIONAL CONDITION 2, during training startups.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE RE UIREMENTS 5

4.10.$'he reactor vessel shall be verified to be unpressurized and the THERMAL POWER and reactor. coolant temperature shall be verified to be within the limits at least once per hour during training startups.

SUSQUEHANNA - UNIT 2 S

3/4 10-/(

RADIOACTIVE EFFLUENTS 3/4. 11. 3 SOLID RADWASTE SYSTEM LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM, for the processing and packaging of radioactive wastes to ensure compli,ance with 10 CFR Part 20, 10 CFR Part 71, and Federal regulations governing the disposal of the ~aste.

APPLICABILITY: At al 1 times.

ACTION:

a.

With the requirements of 10 CFR Part 20, and/or 10 CFR Part 71, not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

b. 'ith the solid radwaste system inoperable for more than 31 days, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:

C.

l.

Identificati n of the inoperable ecuipment or subsystems and the reason for inoperability, 2.

Action(a) taken to rector the inoperahie equipment to Op":RAB/E

status, 3.

A description of the alternative used for. SOLIDIFICATION and packaging of radioactive wastes, and 4.

Summary description of action(s) taken to prevent a recurrence.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 11.3.1 The solid radwaste system shall be demonstrated OPERABLE at least once per 92 days by:

a.

Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, ot b.

Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

SUS)UEHANNA - UNIT 2 3/4 11-20

3/4. 10 SPECIAL TEST EXCEPTIONS BASES 3/4. 10. 1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.

3/4. 10. 2 ROD SE UENCE CONTROL SYSTEM In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints on control rod movement.

The additional surveillance requirments ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod wor ths do not exceed the values assumed in the safety analysis.

3/4.10.3 SHUTOOWN MARGIN OEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur.

These additional restrictions are specified in this LCO.

3/4. 10. 4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

N geiger from the o

gen concentretio specificetio is necess in or r to pro de access to e primary contai ent during e initial s artup a

testi g phase of op ation.

Without is access th startup an test p

gram co d be restricte and delayed.

3/4.10. 5 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to. minimize contaminated water discharge to the radioactive waste disposal system.

SUS)UEHANNA - UNIT 2 8 3/4 10-1

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ML17146A481

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