ML050610214

Technical Specifications Bases Unit 2 Manual

ML050610214
Person / Time
Site:	Susquehanna
Issue date:	02/18/2005
From:	Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
028401
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Feb. 18, 2005 Page 1 of 2 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2005-7865 USER INFORA i CH*RO M EMPL#:028401 CA#: 0363 Add CSA2 Phone# 3194 TR MTA m RMATION:

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TSB2 - TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL REMOVE MANUAL TABLE OF CONTENTS DATE: 01/25/2005 ADD MANUAL TABLE OF CONTENTS DATE: 02/17/2005 CATEGORY: DOCUMENTS TYPE: TSB2 ID: TEXT 3.1.6 REMOVE: REV:0 ADD: REV: 1 CATEGORY: DOCUMENTS TYPE: TSB2 ID: TEXT 3.3.2.1 REMOVE: REV:0

Feb. 18, 2005 Page 2 of 2 ADD: REV: 1 CATEGORY: DOCUMENTS TYPE: TSB2 ID: TEXT LOES REMOVE: REV:54 ADD: REV: 55 CATEGORY: DOCUMENTS TYPE: TSB2 ID: TEXT TOC REMOVE: REV:5 ADD: REV: 6 UPDATES FOR HARD COPY MANUALS WILL BE DISTRIBUTED WITHIN 5 DAYS IN ACCORDANCE WITH DEPARTMENT PROCEDURES. PLEASE MAKE ALL CHANGES AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX UPON RECEIPT OF HARD COPY. FOR ELECTRONIC MANUAL USERS, ELECTRONICALLY REVIEW THE APPROPRIATE DOCUMENTS AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.

- .SSESMANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date: 02/17/2005 Procedure Name ,Rev Issue Date Change ID change Number TEXT LOES 55 02/17/2005 I . .

Title:

LIST OF EFFECTIVE SECTIONS TEXT TOC 6 02/17/2005

Title:

TABLE OF CONTENTS -

"rI  ! .:

I, ... J.

TEXT 2.1.1 1 10/27/2004 At . ,.t, I

Title:

SAFETY LIMITS (SLS) REACTOR CORE SLS-; ., . I =

'~~ ' ..'C~ .9l e'+. '

TEXT 2.1.2 0 11/18/2002

Title:

SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM (RCS).PRESSURE SL TEXT 3.0 0 11/18/2OO2,. .

Title:

LIMITING CONDITION FOR OPERATION-M(LCO)-APPLICABILITY

/-s <dq %,},;' ----

TEXT 3.1.1 0 k11/18/2002

Title:

REACTIVITY CONTROL SYSTEMS.SHUTDOWN MARGIN (SDM)

TEXT 3.1.2 0e 11/18/2002

Title:

REACTIVITY CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 0 11/18/2002

Title:

REACTIVITY CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 0 11/18/2002

Title:

REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 0 11/18/2002 -

Title:

REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 1 02/17/2005

Title:

REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Page I of 8 Re Report Date: 02/17/05

SSFS MANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL up, TEXT 3.1.7 0 11/18/2002

Title:

REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL (SLC) SYSTEM TEXT 3.1.8 0 11/18/2002

Title:

REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV) VEMT AND'DRAIN VALVES TEXT 3.2.1 0 11/18/2002

Title:

POWER.DISTRIBUTION LIMITS'AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

TEXT 3.2.2 0 11/18/2002

Title:

POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER'PATIO (MCPR)

TEXT 3.2.3 . 0 11/18/2002

Title:

-POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE (LHGR)

TEXT 3.2.4 0 11/18/2002

Title:

POWER DISTRIBUTION LIMITS AVEFRAGE'POWER RANTGE MONITOR (APRM) GAIN AND SETPOINTS X TEXT 3.3.1.1 0 '11/18/2002 .

Title:

INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3.3.1.2 0. 11/18/2002 ,

Title:

INSTRUMENTATION SOURCE RANGE MONITORA(SRM) INSTRUMENTATION TEXT 3.3.1.3 11/22/2004

Title:

OPRM INSTRUMENTATION TEXT 3.3.2.1 1 .02/17/2005

Title:

INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 0 11/18/2002

Title:

INSTRUMENTATION FEEDWATERX- MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 0 11/18/2002

Title:

INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION LDCN 3710 Re or Da e 02/17/05 .

of~ _

of a Report Date: 02/17/05

SSES AMAL 7 Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS.BASES UNIT 2 MANUAL TEXT 3.3.3.2 0 11/18/2002.

Title:

INSTRUMENTATION REMOTE SHUTDOWN -SYSTEM , .' '

TEXT 3.3.4.1 0 11/18/2002:

Title:

INSTRUMENTATION END OF CYCLE :RECIRCULAT-ION PUMP:TRIP (EOC-RPT)-INSTRUMENTATION TEXT 3.3.4.2 0. 11118/2Q02

Title:

-INSTRUMENTATION ANTICIPATED TRANSIEN'.'WITHOUT.'SCRAM RECIRCULATION PUMP TRIP-(ATWS-RPT) INSTRUMENTATION TEXT 3.3.5.1 2 01/2-41'2005.'

Title:

INSTRUMENTATION EMERGENCY CORE-COOLING.SYSTEM (ECCS) rINSTRUMENTATION -:'

TEXT 3.3.5.2 0 1/1l8.A2Q02.. . '.

Title:

INSTRUMENTATION REACTOR.CORE,:.ISOLATION.COOLING'.(RCIC) . SYSTEM INSTRUMENTATION TEXT 3.3.6.1 1 ll/.09/20Q4_

Title:..INSTRUMENTATION PRIMARY.-C.ONTAINMENT ISOLATION,.INSTRUMENTATION-I' TEXT 3.3.6.2 1 11t.091.2004

Title:

INSTRUMENTATION SECONDARY -CONTAINMENT :-ISOLATION:INSTRUMENTATION TEXT 3.3.7.1 0 11/18720.02. ,,

Title:

INSTRUMENTATION CONTROL ROOM EMERGENCY.-OUTSIDE AIR'SUPPLY (CREOAS) SYSTE14-INSTRUMENTATION TEXT 3.3.8.1 1 09/02/200.4

Title:

INSTRUMENTATION LOSS OF.POWER-.(LOP).:INSTRUMENTATION -  :: -'.'a TEXT 3.3.8.2 0 111/18/2002

.Title: INSTRUMENTATION REACTOR :PROTECTION4 ,SYSTEM (RPS).ELECTRIC POWER MONITORING

TEXT 3.4.1 2 11/22/2004

Title:

REACTOR COOLANT -SYSTEM .(RCS) ':RECIRCULATION LOOPS OPERATING .

TEXT 3.4.2 0 11/.18/.2002 '

Title:

REACTOR COOLANT SYSTEM (RCS) JET PUMPS ' .'-. a -  ;

Page 3 of- 8 - Report Date: 02/17/05

- SSES MANUAL Manual Name: TSB2 . I Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.4.3 0 11/18/2002

Title:

REACTOR COOLANT SYSTEM (RCS) SAFETY/RELIEF VALVES (S/RVS)

TEXT 3.4.4 0 11/18/2002

Title:

REACTOR COOLANT SYSTEM ARCS):.RCS'OPERATIONAL LEAKAGE TEXT 3.4.5 0 11118/2002

Title:

REACTOR.COOLANT SYSTEM :(RCS)' RCS:'PRESSURE ISOLATION VALVE (Ply) LEAKAGE.

TEXT 3.4.6 0 11/18/2002 -

Title:

REACTOR COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 0 11/18/2002:-

Title:

REACTOR COOLANT SYSTEM (RCS) -RCS'SPECIFIC ACTIVITY.

TEXT 3.4.8 0 11/18/2002-

Title:

REACTOR COOLNT:--SYSTEM.'(RCS) RESIDUAL HEAT. REMOVAL (RHR) SHUTDOWN COOLING SYSTE&,_/-

- HOT SHUTDOWN TEXT 3.4.9 0 11/18/2002

Title:

REACTOR COOLANT:SYSTEM-(RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM

- COLD SHUTDOWN TEXT 3.4.10 0 11/18/2002

Title:

REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 0 11/18/2002

Title:

REACTOR COOLANT SYSTEM (RCS) REACTOR STEAMS DOME PRESSURE TEXT 3.5.1 0 11/18/2002

Title:

EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM ECCS - OPERATING TEXT 3.5.2 0 11/18/2002

Title:

EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM ECCS - SHUTDOWN TEXT 3.5.3 0 11/18/2002

Title:

EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM RCIC SYSTEM Page 4 of 8 Report Date: 02/17/05

; SSES 7ANUAL Manual Name

TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS.BASES UNIT 2 MANUAL-TEXT 3.6.1.1 0 11/18/2002

Title:

CONTAINMENT.SYSTEMS.PRIMARY'CONTAINMENT-' -

TEXT 3.6.1.2 0 11/18/2002. .'

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT.-AIR.LOCK:'

TEXT 3.6.1.3 0 11/18/2002'. .. . . I I 7

Title:

CONTAINMENT SYSTZMS.PRIMARY CONTAINMENT *ISOLATION'VALVES (PCIVS) ., . , ':  ! j L '  :.

TEXT 3.6.1.4 0 11/18/2002:

Title:

CONTAINMENT SYSTEMS.CONTAINMENT:PRESSURE ,'Y ,

• I .  :  %.I I TEXT 3.6.1.5 0 11/.18/20.02f .

I - 7 , *bl;*

Title:

CONTAINMENT SYSTEMS DRYWELL AIR .TEMPERATURE .

TEXT 3.6.1.6 0 11/18/2002 '

Title:

CONTAINMENT SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAXERS TEXT 3.6.2.1 0 11/18/20C02 ' :1 .. . - ~. ,

Title:

CONTAINMENT.SYSTEMS- SUPPRESSION:POOI'-AVERAGE TEMPERATURE TEXT 3.6.2.2 0 11/18/2002 '.'  :,. . 1.

Title:

CONTAINMENT SYSTEMS SUPPRESSION POOL WATER LEVEL-.'.;.' - - . t ' '-

TEXT 3.6.2.3 0 11/18/2002

Title:

CONTAINMENT SYSTEMS RESIDUALHEAT REMOVAL (RHR) SUPPRESSION POOL COOLING -

TEXT 3.6.2.4 0 11/18/2002.

Title:

CONTAINMENT SYSTEMS RESIDUAL HEAT.REMOVAL' (RHR)-SUPPRESSION POOL SPRAY r TEXT 3.6.3.1 0 11/18/2002 ' I -- .

Title:

CONTAINMENT. SYSTEMS PRIMARY. CONTAINMENT HYDROGEN RECOMBINERS .9' ...

TEXT 3.6.3.2 0 11/18/2002.'

Title:

CONTAINMENT SYSTEMS DRYWELL :AIR, FLOW .SYSTEM' Page 5 -of, 8 r Date--'02/17/.05 Report De 0 ,

-SSES ANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.3.3 0 11/18/2002

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 1 01/03/2005

Title:

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 2 01/03/2005

Title:

CONTAINMENT SYSTEMS SECONDARY-ZCONTAINMENT ISOLATION VALVES (SCIVS)

TEXT 3.6.4.3 2 11/09/2004

Title:

CONTAINMENT SYSTEMS STANDBY GAS-TREA.rMENT. (SGT) .SYSTEM..

TEXT 3.7.1 0 11/18/2002 '

Title:

PLANT SYSTEMS RESIDUAL: HEAT.:REMOVAL'. SERVICE WATER (RHRSWY SYSTEM AND THE ULTIMATE HEAT SINK (UHS)

TEXT 3.7.2 1 11/09/2004 .

Title:

PLANT SYSTEMS, EMERGENCY. SERVICE WATER (ESW) SYSTEM . -

TEXT 3.7.3 0 11/18/2002 .'

Title:

PLANT SYSTEMS CONTROL.ROOM EMERGENCY OUTSIDE AIR 'SUPPLY. (CREOAS) SYSTEM TEXT 3.7.4 0 11/18/2002

Title:

PLANT SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM' TEXT 3.7.5 0 11/18/2002

Title:

PLANT SYSTEMS MAIN CONDENSER OFFGAS TEXT 3.7.6 1 01/17/2005

Title:

PLANT SYSTEMS MAIN TURBINE BYPASS SYSTEM ...

TEXT 3.7.7 0 11/18/2002

Title:

PLANT SYSTEMS SPENT FUEL STORAGE'POOL WATER'LEVEL TEXT 3.8.1 1 10/17/2003

Title:

ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING.

Report Date: 02/17/05 Page 6 Page 6 of of 8 8 Report Pate: 02/17/05

SSES MANUAL Manual Name

TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2'MANUAL' TEXT 3.8.2 0 11/18/2002

Title:

ELECTRICAL POWER SYSTEMS AC SOURCES -. SHUTDOWN ' ' -

TEXT 3.8.3 0 11/18/2002 A

Title:

ELECTRICAL POWER SYSTEMS-DIESEL'FUEL 'OIL,. LUBE'OIL, AND STARTING AIR TEXT 3.8.4 ' 0 11/18/2002'. A ' ('

Title:

ELECTRICAL POWER-SYSTEMS.DC SOURCES-'.-:OPERATING - -' "

0 11/18/2002- 1, .. I .

TEXT 3.8.5

Title:

ELECTRICAL POWER-I SYSTEMS-DC SOURCES---;SHUTDOWN'-"-' -' - -

TEXT 3.8.6 0 11/18/2002-'

Title:

ELECTRICAL POIWER.SYSTEMS BATTERY CELL'PARAMETERS'  :. -;ii TEXT 3.8.7 0 11/1872002. -

Title:

ELECTRICAL POWER-SYSTEMS DISTRIBUTION'SYSTEMS - OPERATING '_ - .-. A TEXT 3.8.8 0 11/18/2002 - i

Title:

ELECTRICAL POWER SYSTEMS DISTRIBUTION- SYSTEMS' - SHUTDOWN '.

TEXT 3.9.1 0 11/18/2002 '

Title:

REFUELING OPERATIONS REFUELING EQUIPMENT- INTERLOCKS TEXT 3.9.2 0 11/18/2002 -

Title:

REFUELING OPERATIONS REFUEL POSITION::ONE-ROD-OUT'INTERLOCK'---

TEXT 3.9.3 0 11/18/2002' '

Title:

REFUELING OPERATIONS CONTROL rROD:POSITION '- '.-2 '- ' -'I TEXT 3.9.4 0 11/18/2002

Title:

REFUELING OPERATIONS CONTROL ROD POSITION INDICATION TEXT 3.9.5 0 11/18/2002

Title:

REFUELING OPERATIONS CONTROL ROD OPERABILITY - REFUELING of 8 eotDt:0/7

.ae7 Page 7 . : of 8 : Report Date: 02/17/b5:

SSES MANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL.

TEXT 3.9.7 0 11/18/2002

Title:

REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL TEXT 3.9.8 0 11/18/2002

Title:

REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL TEXT 3.10.1 0 11/18/2002

Title:

SPECIAL OPERATIONS INSERVICE LEAN AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2 0 11/18/2002

Title:

SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 0 .1/18/2002

Title:

SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL--- HOT SHUTDOWN TEXT 3.10.4 0 11/18/2002

Title:

SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - COLD SHUTDOWN TEXT 3.10.5 0 11/18/2002

Title:

SPECIAL OPERATIONS SINGLE CONTROL ROD DRIVE (CRD) REMOVAL - REFUELING TEXT 3.10.6 0 11/18/2002

Title:

SPECIAL OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING TEXT 3.10.7 0 11/18/2002

Title:

SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING TEXT 3.10.8 0 11/18/2002

Title:

SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Report Date: 02/17/05 Page88 Page of of 88 Report Date:' 02/17/05

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)

B2.0 SAFETY LIMITS (SLs) .................. TSIB2.0-1 B2.1.1 Reactor Core SLs ........... TS/B2.0-1 B2.1 .2 Reactor Coolant System (RCS) Pressure SL ................................. B2.0-6 B3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY .............. B3.0-1 B3.0 'SURVEILLANCE REQUIREMENT (SR) APPLICABILITY; ........................... B3.O-10 B3.1 REACTIVITY CONTROL SYSTEMS . ........................................... B3.1-1 B3.1.1 Shutdown Margin (SDM) ............................................. B3.1-1 B3.1.2 Reactivity Anomaliis .... B3.1-8 B3.1.3 Control Rod OPERABILITY .. B3.1-13 B3.1.4 . Control Rod Scram Times : . .... B3.1-22 B3.1.5 . ,, Control Rod Scram Accurulators . ......................... ....... B3.1-29 B3.1.6 Rod Pattern Control :.... .. '. ' .. TS/B3.1-34 B3.1.7 Standby Liquid Control (SLC) System . /.........; .B3.1-39 B3.1.8 Scram Discharge Volumrr(SDV) Vent and Dra Vae... ... B3.1-47 B3.2 POWER DISTRIBUTION LIMITS ...................... TS/B3.2-1 B3.2.1 Average Planar Linear Heat Generation Rafie (APLHGR) . TS/B3.2-1 B3.2.2 Minimum Critical Powerio (MCPR)..'.TS/B3.2-5 ' -.

B3.2.3 r Linear Heat Generation 'Rate&(LAGR) R2......:............................ TS/B3.2-1 0; B3.2.4 Average Power Range M ritor,(APRM) Gain and Setpoints ....... .... TS/B3.2-14 B3.3 - INSTRUMENTATION N') ' ' .' TS/B3.3-1 B3.3.1.1 . Reactor Protect on S>s't (RPS) Instrumentation.. TS/B3.3-1 B3.3.1.2 Source Range Monitot4SRM) Instrumentation . .TS/B3.3-35 B3.3.1.3 Oscillation ,Po'werRange.Monitor , . ........................ TS/B3.3-43a B3.3.2.1 ' Control Rod'Blockinstrurnentation......................... TS/B3.3-44 B3.3.2.2' Feedater>- Main Turbine High Water Level Trip

-Instruiienttion.. ........... . . . .. B3.3-55 B3.3.3.1 , Post Accid~ent.Monitorirg (PAM) ,Instrumentation ..... s c..s*.TS/B3.3-64:

B3.3.3.2 ' Remote"Shutdown System .B3.3-76 B3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation f. . . . .. ................... B3.3-81 8::

B3.3.4.2 ' Antcipated Transient Without Scram Recirculation

-- PumpTrip (ATWS-RPT) Instrumentation .B3.3-92 B3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation..... ....  ;: .TS/B3.3-101 B3.3.5.2 Reactor Core Isolation 'ooling (RCIC) System Instrumentation ......... B3.3-135 B3.3.6.1 Primary Containment Isolation Instrumentation .B3.3-147 B3.3.6.2 Secondary Containment Isolation Instrumentation .TS/B3.3-180 B3.3.7.1 Control Room Emergency Outside Air Supply (CREOAS)

System Instrumentation .B3.3-192 (continued)

SUSQUEHANNA-UNIT 2 TS / B TOC - 1 Revision 6

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)

B3.3 INSTRUMENTATION (continued)

B3.3.8.1 Loss of Power (LOP) Instrumentation .TS/B3.3-206 B3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ..... B3.3-214 8.-

B3.4 REACTOR COOLANT SYSTEM (RCS) .................................. TSIB3.4-1 B3.4.1 Recirculation Loops Operating ................ .................. TS/B3.4-1 B3.4.2 Jet Pumps .............. B3.4-10 B3.4.3 Safety/Relief Valves (S/RVs) ............................... TS/B3.4-15 B3.4.4 RCS Operational LEAKAGE ............................... B3.4-19 B3.4.5 RCS Pressure Isolation Valve (PIV) Leakage ............................... B3.4-24 B3.4.6 RCS Leakage Detection Instrumentation ............................... B3.4-30 B3.4.7 RCS Specific Activity ............. B3.4-35 B3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ............................ B3.4-39 B3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown. B3.4-44 B3.4.10 RCS Pressure and Temperature (P/T) Limits ............................. TS/B3.4-49 B3.4.11 Reactor Steam Dome Pressure TS/B3.4-58 B3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM .TS/B3.5-1 B3.5.1 ECCS - Operating .. TS/B3.5-1 B3.5.2 ECCS - Shutdown . .8........... B3.5-19 B3.5.3 RCIC System: ............ TS/B3.5-25 B3.6 CONTAINMENT SYSTEMS ................... TS/B3.6-1 B3.6.1.1 Primary Containment .............. TSIB3.6-1 B3.6.1.2 Primary Containment Air Lock.............. B3.6-7 8;

B3.6.1.3 Primary Containment Isolation Valves (PCIVs) .............. TS/B3.6-15 B3.6.1.4 Containment Pressure .............. B3.6-40 B3.6.1.5 Drywell Air Temperature .............. TS/B3.6-43 B3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers .............. TS/B3.6-46 83.6.2.1 Suppression Pool Average Temperature .............. B3.6-52 B3.6.2.2 Suppression Pool Water Level .............. B3.6-58 B3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling. B3.6-61 B3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray ................. B3.6-65 B3.6.3.1 Primary Containment Hydrogen Recombiners ................................ B3.6-69 B3.6.3.2 Drywell Air Flow System ........................................ B3.6-75 B3.6.3.3 Primary Containment Oxygen Concentration ................................. B3.6-80 B3.6.4.1 Secondary Containment ......................................... TS13.6-83 B3.6.4.2 Secondary Containment Isolation Valves (SCIVs) .................... TS/B3.6-90 B3.6.4.3 Standby Gas Treatment (SGT) System ................................... TS/B3.6-100 (continued)

SUQUHNN-NI2 T/BO-2-Reiio_

SUSQUEHANNA - UNIT 2 TS/BTOC-2 -Revision 6

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)

B3.7 PLANT SYSTEMS ........................................... TS/B3.7-1 B3.7.1 Residual Heat Removal Service Water (RHRSW) System and the Ultimate Heat Sink (UHS) ....................................... TS/B3.7-1 B3.7.2 Emergency Service Water (ESW) System ................................ TS/B3.7-7 B3.7.3 Control Room Emergency Outside Air Supply (CREOAS) System .................................. TS/B3.7-12 B3.7.4 Control Room Floor Cooling System .................................. TS/B3.7-19 B3.7.5 Main Condenser Offgas ..................................B3.7-24 B3.7.6 Main Turbine Bypass System .................................. TS/B3.7-27 B3.7.7 Spent Fuel Storage Pool Water Level ................. ................. B3.7-31 B3.8 ELECTRICAL POWER SYSTEM .................................. B3.8-1 B3.8.1 AC Sources - Operating .................................. B3.8-1 B3.8.2 AC Sources - Shutdown .................................. B3.8-39 B3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air. ............................. B3.8-47 B3.8.4 *DCSources - Operating ........... ................. TS/B3.8-56 B3.8.5 DC Sources - Shutdown ............................ B3.8-70 B3.8.6 Battery Cell Parameters ............................ B3.8-77 B3.8.7 Distribution Systems - Operating ...... ...................... B3.8-84 B3.8.8 Distribution Systems - Shutdown ............................ B3.8-94 B3.9 REFUELING OPERATIONS ............................. TS/B3.9-1 B3.9.1 Refueling Equipment Interlocks ............................. TS/B3.9-1 B3.9.2 Refuel Position One-Rod-Out Interlock ........... .................. B3.9-5 B3.9.3 Control Rod Position ............. B3.9-9 B3.9.4 Control Rod Position Indication ............ .......................... B3.9-12 B3.9.5 Control Rod OPERABILITY - Refueling ...................................... B3.9-16 B3.9.6 Reactor Pressure Vessel (RPV) Water Level ................................. B3.9-19 B3.9.7 Residual Heat Removal (RHR) - High Water Level ......................... B3.9-22 B3.9.8 Residual Heat Removal (RHR) - Low Water Level ........................ B3.9-26 B3.10 SPECIAL OPERATIONS ..................................... TS/B3.10-1

• B3.10.1 Inservice Leak and Hydrostatic Testing Operation ...................... TS/B3.10-1 B3.10.2 Reactor Mode Switch Interlock Testing ....................................... :.B3.10-6 B3.10.3 Single Control Rod Withdrawal - Hot Shutdown ............................. B3.1 0-11 B3.10.4 Single Control Rod Withdrawal - Cold Shutdown ........................... B3.10-16 B3.10.5 Single Control Rod Drive (CRD) Removal - Refueling ................... B3.10-21 B3.10.6 Multiple Control Rod Withdrawal - Refueling ................................. B3.10-26 B3.10.7 Control Rod Testing - Operating ..................................... B3.10-30 B3.10.8 SHUTDOWN MARGIN (SDM) Test- Refueling ............................. B3.10-34 TSB2 TOC.doc 211105 SUSQUEHANNA - UNIT 2 TSI/B TOC- 3 Revision 6

SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVESECTIONS (TECHNICAL SPECIFICATIONS BASES)

Section Title Revision TOC Table of Contents 6 B 2.0 SAFETY LIMITS BASES Page TS / B2.0-1 1 Page TS / B2.0-2 2 PageTS/B2.0-3 3 PageTS/B2.0-4 4 Page TS / B2.O-5 Pages B2.0-6 through B 2.0-8 0 8 3.0 LCO AND SR APPLICABILITY BASES Pages B 3.0-1 through B 3.0-7 0 Pages TS/ B 3.0-8and TS / B 3.0-9 1 Pages B 3.0-10 through B3.0-12 0 Pages TS / B 3.0-13 through TS /B 3.0-15 1 B 3.1 REACTIVITY CONTROL BASES Pages B 3.1-1 through B 3.1-5 - 0 Pages TS /B3.1-6 and TS /B3.1-7 1 Pages B 3.1-8 through B 3.1-27 0 PageTS/B3.1-28 1 Pages B 3.1-29 through B 3.1-33 0 Pages TS / B 3.1.34 through TS /B 3.1-36 1 Page TS /B3.1-37. 2 Page TS / B3.1-38 1 Pages B 3.1-39 through B 3.1-51 0 B 3.2 POWER DISTRIBUTION LIMITS BASES Pages TS / B 3.2-1 through TS / B 3.2-4 1 Pages TS / B 3.2-5 and TS / B 3.2-6 2 Page TS / B3.2-7 I Pages TS /B3.2-8 and TS /B3.2-9 2 Pages TS / B 3.2-10 through TS / 83.2-19 1 B 3.3 INSTRUMENTATION Pages TS / B3.3-1 through TS i B 3.3-10 1 Page TS /B3.3-11 2 Pages TS / B3.3-12 through TS / B 3.3-27 1 Pages TS / B 3.3-28 through TS / B 3.3-30 2 Page TS / B 3.3-31 1 Pages TS / B 3.3-32 and TS B 3.3-33 2 Pages TS / B3.3-34 through TS /B 3.3-43 1 Pages TS / B3.3-43a though TS / B 3.3-43i 0 Pages TS / B3.3-44 through TS /B 3.3-54 2 Pages B 3.3-55 through B 3.3-63 0 Pages TS /B3.3-64 and TS /B3.3-65 2 Page TS /B3.3-66 4 SUSQUEHANNA - UNIT 2 TS / B LOES-1 Revision 55

SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVESECTIONS (TECHNICAL SPECIFICATIONS BASES)

Section Title Revision

,, I Page TS B.3.3-67 3 Page TS / B 3.3-68 4 Pages TS I B 3.3-69 and TS B 3.3-70 3 Pages TS / B 3.3-71 through TS I B 3.3-75 2 PageTS/B3.3-75a 4 Pages TS I B 3.3-75b through TS I B 3.3-75c 3 Pages B 3.3-76 through B 3.3-91 0' Pages TS I B 3.3-92 through TS I B 3.3-103 1 Page TS I B 3.3-104 2 Pages TS I B 3.3-105 and TS I B 3.3-106 1 Page TS / B 3.3-107 2 Page TS / B 3.3-108 1 Page TS/B3.3-109 2 Pages TS I B 3.3-110 through TS / B 3.3-115 1 Pages TS / B 3.3-116 through TS / B 3.3-118 2 Pages TS / B 3.3-119 through TS / B 3.3-120 1 Pages TS / B 3.3-121 and TS / B 3.3-122 2 Page TS / B 3.3-123 1 Page TS I B 3.3-124 2 Page TS / B 3.3-124a 0 Page TS / B 3.3-125 1 Page TS / B 3.3-126 2 Page TS / B 3.3-127: 3 PageTS/B3.3-128 2 Pages TS / B 3.3-129 through TS i B 3.3-131 1 Page TS / B 3.3-132 2 Pages TS / B 3.3-133 and TS. B 3.3-134 1 Pages B 3.3-135 through B 3.3-137 0 Page TS / B 3.3-138 Pages B 3.3-139 through B 3.3-149 0 1-Pages TS/ B 3.3-150 through TS / B 3.3-162 Page TS / B 3.3-163 2 1

Pages TS ) B 3.3-164 through TS I B 3.3-177 Page TS / B 3.3-178 2 Page TS / B 3.3-179 3 Page TS / B 3.3-179a 2 Page TS / B 3.3-180 3 Page TS/B3.3-181 - 2 Pages TS / B 3.3-182 through TS I B 3.3-186 1 Pages TS / B 3.3-187 and TS / B 3.3-188 2 Pages TS / B 3.3-189 through TS / B 3.3-191 1 Pages B 3.3-192 through B 3.3-205 0 Page TS / B 3.3-206,- 1 Pages B 3.3-207 through B 3.3-220 0 B 3.4 REACTOR COOLANT SYSTEM BASES Pages TS / B 3.4-1 and TS / B 3.4-2 .1 Revision 55 TS I B LOES-2 SUSQUEHANNA - UNIT SUSQUEHANNA -

UNIT 22 -TS I B LOES-2 Revision 55

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Section Title Revision Pages TS / B 3.4-3 through TS / B 3.4:9 3 Pages B 3.4-10 through B 3.4-14 0 Page TS / B 3.4-15 I Pages TS / B 3.4-16 and TS / B 3.4-17 2 Page TS / B3.4-18 1 Pages B 3.4-19 through B 3.4-28 0 PageTS/B3.4-29 -

Pages B3.3-30 through B 3.3-48 0 Page TS / B 3.4-49' 2 Page TS / B 3.4-50 1 Page TS / B3.4-51 2 Pages TS / B 3.4-52 and TS / B 3.4-53 1 Pages TS / B 3.4-54 and TS / B 3.4-55 2 Pages TS / B 3.4-56 through TS I B3.4-60 1 B 3.5 ECCS AND RCIC BASES Pages TS / B 3.5-1 and TS / B 3.5-2 1 PageTS/B3.5-3 - 2 Pages TS / B3.5-4 through TS B 3.5-10 1 Page TS / B 3.5-11 2 Pages TS / B 3.5-12 through TS / B3.5-14 1 Pages TS / B 3.5-15 through TS / B.3.5-17 2 PageTS/B3.18 .

Pages B 3.5-19 through B 3.5-24 0 Page TS/B3.5-25 1 Pages B 3.5-26 through B 3.5-31 0 B 3.6 CONTAINMENT SYSTEMS BASES Page TS / B 3.6-1 - 2 Page TS / B 3.6-1a 3 Pages TS / B 3.6-2 through TS / B 3.6-5 2 PageTS/B3.6-6 . 3 Pages TS / B 3.6-6a and TSi B 3.6-6b 2 Page TS/B3.6-6c 0 Pages B 3.6-7 through B 3.6-14 0 Page TS / B3.6-15 3 Pages TS / B 3.6-15a and TS I B 3.6-15b 0 Page TS / B 3.6-16 . 1 PageTS/B3.6-17 2 Page TS / B 3.6-17a , 0 Pages TS / B 3.6-18 and TS I B 3.'6-19 1 Page TS / B 3.6-20 2 Page TS / B 3.6-21 3 PagesTS/B3.6-21aandTS/B3.6-21b 0 Pages TS / B3.6-22 and TS / B 3.6-23 2 Pages TS / B3.6-24 through TS / B 3.6-26 1 PageTS/B3.6-27 3 Revision 55 TS/B LOES-3 SUSQUEHANNA - UNIT UNIT 22 TS / B LOES-3 Revision 55

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SUSQUEHANNA STEAM ELECTRIC STATION UST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)

Section Title Revision Page TS / B 3.6-28 6 Page TS / B 3.6-29 3 Page TS / B 3.6-29a 0 Page TS / B 3.6-30 2 Page TS /B 3.6-31 3 Pages TS / B 3.6-32 through TS I B 3.6-34 1 Pages TS / B 3.6-35 through TS I B 3.6-37 2 Page TS / B 3.6-38 1 Page TS/B3.6-39 4 Pages B 3.6-40 through B 3.6-42 0 Pages TS / B 3.6-43 through TS / B 3.6-50 1 Page TS / B 3.6-51 2 Pages B 3.6-52 through B 3.6-62 0 Page TS / B 3.6-63 1 Pages B 3.6-64 through B 3.6-82 0 Page TS / B 3.6-83. 3 Pages TS / B 3.6-84 and TS / B 3.6-85 2 Pages TS / B 3.6-86 through TS / B 3.6-87a 1 Page TS / B 3.6-88 2 Page TS / B 3.6-89 1 PageTS/B3.3-90 3 Pages TS / B 3.6-91 through TS I B 3.6-95 1 Page TS/B3.6-96: 2 Pages TS / B 3.6-97 and TS I B 3.6-98 1 Page TS / B 3.6-99 2 Page TS / B 3.6-99a 0 Pages TS / B 3.6-1 00 and TS / B 3.6-101 1 Pages TS / B 3.6-102 through TS / B 3.6-104 2' Pages TS / B 3.6-105 and TS I B 3.6-106 1 B 3.7 PLANT SYSTEMS BASES Pages TS / B 3.7-1 through TS / B 3.7-6 2 Page TS / B 3.7-6a 2 Pages TS / B 3.7-6b and TS I B 3.7-6c 0 Page TS/B3.7-7 2>

Page TS / B 3.7-8 1 Pages B 3.7-9 through B 3.7-11 0 Pages TS / B 3.7-12 and TS / B 3.7-13 1 Pages TS / B 3.7-14 through TS / B 3.7-18 2 Page TS / B 3.7-18a 0 Pages TS / B 3.7-19 through TS / B 3.7-26 1 Pages B 3.7-24 through B 3.7-26 0 Pages TS / B 3.7-27 through TS / B 3.7-29 2 Page TS / B 3.7-30 1 Pages B 3.7-31 through B 3.7-33 0 SI U 2S Q E H A N A S/ BU ON ES- Re i si n 5 SUSQUEHANNA - UNIT 2 TS / B LOES-4 Revision 55

SUSQUEHANNA STEAM ELECTRIC STATION

-IST OFEFFECTWIE SECTIONS (TECHNICAL SPECIFICATIONS BASES)

Section- Title Revision B 3.8 ELECTRICAL POWER SYSTEMS BASES Pages B 3.8-1 through B3.8-4 0 Page TS / B 3.8-5 I Pages B 3.8-6 through B3.8-8 0 Pages TS / B 3.8-9 through TS / B 3.8-11 I Pages B 3.8-12 through B 3.8-18 0 Page TS / B 3.8-19 I Pages B 3.8-20 through B 3.8-22 0 0-Page TS / B 3.8-23 Page B 3.8-24 0 Pages TS / B 3.8-25 and TS I B3.8-26 Pages B3.8-27 through B 3.8-37 0 Page TS/B3.8-38 I Pages TS / B 3.8-39 through TS / B 3.8-55 0 Pages TS / B 3.8-56 through TS / B 3.8-64 I PageTS/B3.8-65 2 Page TS / B3.8-66 2 Pages TS / B 3.8-67 through TS I B 3.8-68 I Page TS / B3.8-69 2 Pages' B 3.8-70 through B 3.8-99 0 B 3.9 REFUELING OPERATIONS BASES Pages TS / B 3.9-1 and TS / B 3.9-2 1 Page TS / B 3.9-2a I Pages TS / B 3.9-3 and TS / B 3.9-4 Pages B 3.9-5 through B 3.9-30' B 3.10 SPECIAL OPERATIONS BASES PageTS/B3.10-1 1 Pages B 3.10-2 through B 3.10-32 0 Page TS / B 3.10-33 Pages B 3.10-34 through B 3.10-38 I 0

PageTS/B3.10-39 TSB2 LOES 211105 SUSQUEHANNA - UNIT 2 .TS / B LOES-5 Revision 55

I -'. , " i; ! ' " .

PPL Rev. 1 Rod Pattern Control B 3.1.6 B3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP. The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1 and 2.

APPLICABLE The analytical methods and assumptions used in evaluating the CRDA SAFETY are summarized in References I and 2. CRDA analyses assume that the ANALYSES reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis.

The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for U02 have been shown to be insignificant below fuel energy depositions of 300 cal/gm (Ref. 3), the fuel damage limit of 280 cal/gm provides a margin of safety from significant core damage which would result in release of radioactivity (Refs. 4 and 5). Generic evaluations (Ref. 1 & 6) of a design basis CRDA have shown that the maximum reactor pressure will be less than the required ASME Code limits (Ref.7). The offsite doses are calculated each cycle using the methodology in reference I to demonstrate that the calculated offsite doses will be well within the required limits (Ref. 5). Control rod patterns analyzed in Reference 1 follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (continued)

SUSQUEHANNA - UNIT 2 TS I B 3.1-34 Revision I

PPL Rev. 1 Rod Pattern Control B 3.1.6 1- . 7-i, BASES APPLICABLE (e.g., between notches 08 and 12). The banked positions are established SAFETY to minimize the maximum incremental control rod worth without being ANALYSES overly restrictive during normal plant operation. For each reload cycle the (continued) CRDA is analyzed to demonstrate that the 280 c'aVgm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation for control rod patterns. These analyses consider the effects of fully inserted inoperable and OPERABLE control rods not withdrawn in the normal sequence of BPWS, but are still in compliance with the BPWS requirements regarding out of sequence control rods. These requirements allow a limited number (i.e., eight) and distribution of fully inserted inoperable control rods.

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 9) may be used provided that all withdrawn control rods have been confirmed to be coupled prior to reaching THERMAL POWER of <10% RTP. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled. When using the Reference 9 control rod sequence for shutdown, the RWM may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion, or may be bypassed and the improved BPWS shutdown sequence implemented under LCO 3.3.2.1, Condition D controls.

In order to use the Reference 9 BPWS shutdown process, an extra check is required in order to consider a control rod to be "confirmed" to be coupled. This extra check ensures that no Single Operator Error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Details on this coupling confirmation requirement are provided in Reference 9, which requires that any partially inserted control rods, which have not been confirmed to be coupled since their last withdrawal, be fully inserted prior to reaching THERMAL POWER of <10% RTP. If a control rod has been checked for coupling at notch 48 and the rod has since only been moved inward, this rod is in contact with it's drive and is not required to be fully inserted prior to reaching THERMAL POWER of <10% RTP. However, if it cannot be confirmed that the control rod has been moved inward, then that rod shall be fully inserted prior to reaching the THERMAL POWER of

<10% RTP. This extra check may be performed as an administrative check, by examining logs, previous (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.1-35 Revision i

PPL Rev. 1 Rod Pattern Control B 3.1.6 BASES APPLICABLE surveillance's or other information. If the requirements for use of the SAFETY BPWS control rod insertion process contained in Reference 9 are ANALYSES followed, the plant is considered to be in compliance with the BPWS (continued) requirements, as required by LOC 3.1.6.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 8).

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY,"

consistent with the allowances for inoperable control rods in the BPWS.

APPLICABILITY In MODES 1 and 2, when THERMAL POWER is < 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is

> 10% RTPj there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In.MODES 3, 4, and 5, since the reactor is shut dowh and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to < 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control (continued)

SUSQUEHANNA - UNIT 2 JS / B 3.1-36 Revision 1

PPL Rev. I Rod Pattern Control B 3.1.6 BASES ACTIONS A.1 and A.2 (continued) rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for moves needed to correct the rod pattem, or scram if warranted.

Required Action A.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO.3.3.2.1 requires verification of control rod movement by a qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. OPERABILITY of control rods is determined by compliance with LCO 3.1.3, "Control Rod OPERABILITY," LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators." The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.

B.1 and B:2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a qualified member of the technical staff.

When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.-

(continued)

SUSQUEHANNA - UNIT 2 TS /B 3.1-37 Revision 2

PPL Rev. 1 Rod Pattem Control B 3.1.6 BASES (continued)

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency was developed considering that the primary.

check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at

< 10% RTP.

REFERENCES 1. PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis," Section 2.8, July 1992, Supplement 1-A, August 1995, Supplement 2-A, July 1996, and Supplement 3-A, March 2001.

2. "Modifications to the Requirements for Control Rod Drop Accident Mitigating System," BWR Owners Group, July 1986.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800,;Section 15.4.9, Revision 2, July 1981.

5. 10 CFR 100.11.

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

7. ASME, Boiler and Pressure Vessel Code.

8. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

9. -NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," April 2003.

SUSQUEHANNA - UNIT 2 'TS/ B3.1-38 Revision I

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation B E . S BASES .

BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events.

During -low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA).

During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.

The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one average power range monitor (APRM) channel assigned to each Reactor Protection System (RPS) trip system supplies a reference signal for the RBM channel in the same trip system. This reference signal is used to enable the RBM. If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 2).

The purpose of the RWM is to control rod patterns during startup, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-44 Revision 2

r I -

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND insert blocks when the actual sequence deviates beyond allowances (continued) from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 1). The RWM is a single channel system that provides input into RMCS rod block channel 2.

The function of the individual rod sequence steps (banking steps) is to minimize the potential reactivity increase from postulated CRDA at low power levels. However, if the possibility for a control rod to drop can be eliminated, then banking steps at low power levels are not needed to ensure the applicable event limits can not be exceeded. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled.

To eliminate the possibility of a CRDA, administrative controls require that any partially inserted control rods, which have not been confirmed to be coupled since their last withdrawal, be fully inserted prior to reaching the THERMAL POWER of <10% RTP. If a control rod has been checked for coupling at notch 48 and the rod has not been moved inward, this rod is in contact with it's drive and is not required to be fully inserted prior to reaching the THERMAL POWER of <10%

RTP. However, if it cannot be confirmed that the control rod has been moved inward, then that rod shall be fully inserted prior to reaching the THERMAL POWER of <10% RTP. The remaining control rods may then be inserted without the need to stop at intermediate positions since the possibility of a CRDA has been eliminated.

With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.

(continued)

SUSQUEHANNA - UNIT 2 TS / B3.3-45 Revision 2

• PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 i i..

BASES (continued)

APPLICABLE 1. Rod Block Monitor SAFETY

• ANALYSES, LCO, and The RBM is designed to limit control rod withdrawal if localized neutron APPLICABILITY flux exceeds a predetermined setpoint. The RBM was originally designed to prevent fuel damage during a Rod Withdrawal Error (RWE) event while operating in the power range in a normal mode of operation. FSAR 15.4.2 (Ref. 10) (Rod Withdrawal Error- At Power) originally took credit for the RBM automatically actuating to stop control rod motion and preventing fuel damage during an RWE event at power. However, current reload analyses do not take credit for the RBM system. The Allowable Values are chosen as a function of power level to not exceed the APRM scram setpoints.

• The RBM function satisfies Criterion 4 of the NRC Policy Statement (Ref. 7).

Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block for this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.

Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS: Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,

trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

(continued)

SUSQUEHANNA - UNIT 2 TS / B3.3-46 Revision 2

. I . I PPL Rev. I Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE The RBM will function when operating greater than 30% RTP. Below SAFETY this power level, the RBM is not required to be OPERABLE.

ANALYSES, LCO, and APPLICABILITY 2. Rod Worth Minimizer (continued)

The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 2, 3, 4, and 5. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattem Control."

When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 7) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions. When using the Reference 11 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion, or may be bypassed and the improved BPWS shutdown sequence implemented under the controls in Condition D.

The RWM Function satisfies Criterion 3 of the NRC Policy Statement.

(Ref. 7)

Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 6). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY,"

and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-47 Revision 2

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 BASES '

APPLICABLE Compliance with the BPWS, and therefore OPERABILITY of the SAFETY RWM, is required in MODES I and 2 when THERMAL POWER is ANALYSES, LCO, < 10% RTP. When THERMAL POWER is > 10% RTP, there is no and APPLICABILITY possible control rod configuration that results in a control rod worth that (continued) could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 4 and 6). In MODES 3 and 4, all control rods are required to be inserted' into the core (except as provided in 3.10 "Special Operations');

therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.

3. Reactor Mode Switch-Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.

The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement. (Ref. 7)

Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.

During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, .the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2) provides the required control rod withdrawal blocks.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-48 Revision 2

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 BASES -(continued)

ACTIONS A.1 With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining.

OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 5 days is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.

B.1, If Required Action A. 1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.

The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.

C.1. C.2.1.1. C.2.1.2. and C.2.2 With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram.

Altefnatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM was not performed in the last calendar year, i.e. the last 12 months.

Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. A reactor startup with an inoperable RWM is defined as rod withdrawal during startup when the RWM is required to be OPERABLE. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 (continued)

SUSQUEHANNA - UNIT 2 .- TS / B3.3-49 Revision 2

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS C.1. C.2.1.1. C.2.1.2. and C.2.2 (continued) allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.

D.1 With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence.

Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.

E.1 and E.2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.

In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1.

Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted.

Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.

(continued)

SUSQUEHANNA - UNIT 2 TS / a 3.3-50 Revision 2

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Control Rod REQUIREMENTS Block instrumentation Function are found in-the SRs column of Table 3.3.2.1-1.

The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.

SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control Multiplexing System input. The Frequency of 92 days is based on reliability analyses (Ref. 8).

SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs and by verifying proper indication of the selection error of at least one out-of-sequence control rod. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE I when THERMAL POWER is 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Frequencies are based on reliability analysis (Ref. 8).

(continued)

SUSQUEHANNA - UNIT 2 TS / B3.3-51 Revision 2

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2A1 I . .

BASES SURVEILLANCE SR 3.3.2.1.4 REQUIREMENTS (continued) The RBM trips are automatically bypassed when power is below a specified value and a peripheral control rod is not selected. The power Allowable Value must be verified periodically to not be bypassed when 2 30% RTP. This is performed by a Functional check. If any RBM bypass setpoint is non-conservative, then the affected RBM channel is considered inoperable. Alternatively, the RBM channel can be placed in the conservative condition (i.e., enabling the RBM trip). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.

Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 24 month Frequency is based on the need to perform the Surveillance during a plant start-up.

SR 3.3.2.1.5 The RWM is automatically bypassed when power is above a specified value. .The power level is determined from steam flow signals. The automatic bypass' setpoint must be verified periodically to be not bypassed

• 10% RTP. This is performed by a Functional check. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in.

the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Frequency is based on the need to perform the surveillance during a plant start-up.

SR 3.3.2.1.6-A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs. -

(continued)

SUSQUEHANNA - UNIT 2 . .. .TS / 8 3.3-52 Revision 2

PPL Rev. 1 Control Rod Block Instrumentation B 3.3.2.1 BASES . -'

SURVEILLANCE SR 3.3.2.1.6 (continued)

REQUIREMENTS As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 24 month Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.

The 24 month Frequency is based on the need to perform portions of this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.2.1.7 CHANNEL CALIBRATION is a test that verifies the channel responds to the measured parameter with the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibration consistent with the plant specific setpoint methodology.

As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.

Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8.

The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

(continued)

SUSQUEHANNA - UNIT 2 TS / B3.3-53 Revision 2

4- PPL Rev.1 Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.8 REQUIREMENTS (continued) The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors -are possible.

REFERENCES 1. FSAR, Section 7.7.1.2.8.

2. FSAR, Section 7.6.1.a.5.7

3. NEDE-2401 1-P-A-9-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, Section S 2.2.3.1, September 1988.

4. "Modifications to the Requirements for Control Rod Drop Accident

.Mitigating Systems," BWR Owners' Group, July 1986.

5. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

6. NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17,"

December 27,1987.

7. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 32193)

8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

October 1988.

9. GENE-770-06-1, "Addendum to Bases for changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation, Technical Specifications," February 1991.

10. FSAR, Section 15.4.2.

11. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," April 2003.

SUSQUEHANNA - UNIT 2 . TS / B 3.3-54 Revision 2